ML20073G095
| ML20073G095 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 04/26/1991 |
| From: | Marriott P GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| EEN-9131, NUDOCS 9105030120 | |
| Download: ML20073G095 (88) | |
Text
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.. GE Nuclear Energy r,mvcewsconux v5 Cwtw Arynw Sw ha CA 35U5 April 26,1991
~
l MFN- No. 040 91 i Docket No. STN 50 605 EEN 9131 -
Document Control Desk-- 1 U.S. Nuclear Regulatory Commission _!
Washina, ton, D.C. 20555
}
- Attention: Charles L. Miller, Director Li Standardization and Non Power Reactor Project Directorate ; -
Subject:
Summary Status of GE/NRC' March 4 6,L1991[ Meeting c.i- -I
' Plant Systems Open Items:
s
Reference:
Summary Status'of GE/NRC March 4 6, 1991, Meeting on -
O ei #t srsiems one i' ems. ues se.o28 92. a tea u <ca 28; 1991-
~
Enclosed are thirty fo_ur-(34) copies of the second portion ~of the OE res onses to'the subject'open -.
' items. The first portion of these responses was provided to the NRC via he above reference.'
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>It is intended that:GE will amend the SSARJas appropriate, with these responses' inia future amendment.' y j
iSincerely - ,
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P. Wi Marriott, Manager s s Regulatory andl Analysis Services : m
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SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
,g PLANT SYSTEMS OPEN ISSUES t a Number Subiect Action Comments 3.11 LOUIPMENT OUALIFICATION 3.11(1) Time Margin GE Response provided 3/28/91.
3.11(2) IEEE Edition GE Response provided 3/28/91.
3.11(3)a Gammma Accident Dose GE Response provided 3/28/91.
3.11(3)b Worst Case Expected GE Response provided 3/28/91.
Environment CLARIFICATIQN OF PAST ISSUES AND RESPONSES FOR APPENDIX 3I Chemical Environmental Conditions Response provided 3/28/91.
Spray and/or Submergence Response provided 3/28/91.
Enviroment Data Beta Radiation Enviroment Data Response provided 3/28/91.
Limiting Accident Scope Response provided 3/28/91.
) Significant Enveloping Abnormal Response-provided 3/28/91.
Environmentally Mild or Harsh Response provided 3/28/91.
Zones Typical Equipment Located in Response provided 3/28/91.
Zones Limited Locations of Safety-Related Response provided 3/28/91.
Equipment Inconsistancy in Presssure Units Response provided-3/28/91.
Deletion of Tables Response provided-3/28/91.
+
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SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
() PLANT SYSTEMS OPEN ISSUES inn er Subiect Action Comments 3.5.1.1 PROTECTION OF SAFETY-RELATED EOUIPMENT 3.5.1.1 Separation of Satety-Related GE Response provided 3/28/91 (1) and non Safety-Related Equipment 3.5.1.4 DESIGN BASIS TORNADO (1) ANS 2.3 to SRP GE GE still evaluating impact.
Also, pursuing E-7 recurrence interval.
- 3. 5.2 PROTECTION OF CHARCOAL DELAY TANKS 3.5.2 Relative position of tanks in None Closed.
(1) turbine building.
3.6.1 WORST CASE FLOODING 3.6.1 Total failure of non-Seismic NRC NRC committed to providing a (1) Piping Systems reference regulatory basis.
](s 3.6.1 STEAM TUNNEL 3.6.1 Analysis of Steam Tunnel for GE Response provided on pages (1) Pipe Breaks 3.6-5 and 3.6-27.
3.6.1 HIGH ENERGY PIPING LINES 3.6.1 Exemption of Selected High GE Response provided 3/28/91 (1) Energy Pipes 3.6.1 DBA RUPTURE OF HIGH OR MODERATE ENERGY LINE 3.6.1 Habitability of Control Room Due (1) to Pipe Break and DBA Analysis
- Habitability of Control Room GE See response to 3.6.1(1) (Steam '
Portion Tunnel) above.
- DBA Analysis Portion GE Response provided on.pages 6.2-22, 22a, 23,.and.44 attached.
O
SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON PLANT SYSTEMS OPEN ISSUES Number Subiect Action Comments 6.2.6 CONTAINMENT LEAKAGE TESTING 6.2.6 Systems not Vented or Drained GE Response provided 3/28/91.
(1) (Type A) 6.2.6 Systems not be be Vented or GE Response provided 3/28/91. 4 (2) Drained
- 6.2.6 Type B Tests at Power GE Response provided 3/28/91.
j(3) 6.2.6 Air Lock Seal Testing GE Response provided 3/28/91.
(4) 6.2.6 Penetrations GE Response provided 3/28/91.
(5) 5.2.6 ECCS Isolation Valve Test GE Response provided 3/28/91.
(7) Type C 6.2.6 List of CIVs for C Testing GE Response provided 3/28/91.
a
- 3. 6 List of Valves Reverse Tested GE Response provided 3/28/91.
'(8)b l5. 2. 6 Testing of valves with no 30-Day GE Response provided 3/28/91.
'(8)c Seal
'6.2.6 Containment Purge Isolation Time NRC NRC will consider further (8)d and will discuss with GE at a later time.
6.2.6 Secondary Containment Inleakage/ GE Response provided 3/28/91.
(10) on-ILRT 6.2.6 Control of Test, Vents, and Drains GE_ Response provided 3/28/91.
(11) 6.2.6 ESF System Leak Testing GE Response provided 3/28/91.
(12) 6.2.6 Type C Tests for Containment None Resolved.
(13 Boundary Lines 9)
SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
-~ PLANT SYSTEMS OPEN ISSUES Nhmber Subiect Action Comments 6.5.3 FISSICN PRODUCT CONTROL SYSTEMS & STRUCTURES 6.5.3 Supression Pool Scrubbing Factor None Resolved.
A(1) 6.5.3 Standby Gas Treatment Single GE Response will be provided in A(2) Filter Train May.
6.5.3 Single filter reliability, GE See item 6.5.3 A(2) above.
B Availability (2) (1) 6.5.3 SGTS Instrumentation NRC This item is to be revisited.
B(2) (2) 6.5.3 Effects of Routine Operational GE Response will be provided in B(2)(3) use of the SGTS on its May in conjunction with Reliability and Availability the item 6.5.3 A(2).
for use During Post-Accident Conditions
/~N, 6.5.1 ESF ATMOSPHERE CLEANUP SYSTEM
%e 6.5.1 Normal Air Handling System NRC Provided on Amendment 16. NRC (1) will review.
6.5.1 Confirmatory Item (1) - Intake GE Response will be provided CI(1) Design Capacity following Chap.15 reanalysis.
6.5.1 Fire Protection for CR ESP Filter None Resolved.
(2) System 6.5.1 ESF Components List NRC Provided on Amendment 16. NRC (3) will review, 6.5.1 Redundancy of ESF. Filter Trains None Resolved.
CI(2) for CR Intake 3
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SUMMARY
STATUS OF GE/NRC FaRCH 4-6, 1991 MEETING ON
[~'g PLANT SYSTEMS OPEN ISSUES Ud er Subiect Action Comments 6.4 CONTROL ROOM HABITABILITY SYSTEMS i 6.4 Location of Makeup Air Inlets GE Resolved pending recalculating (1) all of the Chapter 15 radiological accidents.
6.4 Protection from Confined Area None Resolved.
(2) Releases 6.4 Instrumentation NRC ProvidedLon Amendment 16. NRC (3) will review.
6.4 Positive Pressure in Control and oGE - -Response provided-3/28/91.- '
(4) Mechanical equipment Rooms.
6.4 Thickness of Charcoal Adsorber GE Response providedI3/28/91.
CI(1) 15.7.3 LIOUID RADWASTE TANK FAILURE 15.7.3 Design of Radwaste1 Substructure GE Response will be provided
-following Chap. 15-reanalysis.
9.2.9 MAKEUP WATER SYSTEM (CONDENSATE) 9.2.9 Automatic-~Switchover of Suction NRC NRC will-review and evaluate.
-9.2.9- Auto Switchover for'SP-Cleanup- GE. .ResponseTprovided-3/28/91.- ;
.(2) Pump-Suction 9.2.9 Analysis for Potential Flooding _ GE Response'provided on page
.(3) From Failure of MUWC System-
- '9.2-16' attached.
9.2.9-Required CST Inventory GE Response provided.on;pagel (3) 9.2-16: attached.
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SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
- ; PLANT SYSTEMS OPEN ISSUES h-J Number Subiect Action Comments 9.2.10 MAKEUP WATER WATER SYSTEM (PURIFIED)
@.2.10 MUWP interface with SR Systems GE Response provided 3/28/91.
(1)
@.2.10 Domineralized Water Makeup and GE Response provided on page (4) Storage Tank Capacities 9.2-13 attached.
@.2.10 Water Supply Specifications GE Response provided 3/28/91.
(Sb)
@.2.10 Applicant Scope GE Response provided on page (Sc) 9.2-13 attached.
9.2.11 BEACTOR BUILDING GOOLING WATER _tSTEM (RCW) 0.2.11 Missile Protection None Resolved.
(1) 0.2.11 Heat Exchangers GE Response provided on pages
-' 9, 9.2-6 and 9.2-19.1 attached.
l NRC Will reexamine the basis for need to address 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> shutdown heat load.
p.2.11 Sizing of Heat Exchangers GE Response provided on page (3) 9.2-13 attached, Four Hour Shutdown with Loss None Resolved in Amendment 14.
f.2.11 (4) of AC Power
.2.11 Protection of RCW from HELB/MELB GE Response provided 3/28/91.
(5)
. 2.11 Service Water System Description GE Response provided on pages (7) and Interface with Sea Water 9.2.11 and 9.2-25d attached. .
P&ID will be provided in May. ]
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SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
'- PLANT SYSTEMS OPEN ISSUES Item Number Subiect Action Comments 9.2.12 HVAC NORNAL COOLING WATER SYSTEM 9.2.12 HVAC Isolation Valves Seismic (1) Category Part a: Secondary Containment None Resolved.
Isolation Valves Part b: Seismic Category I Class. GE Response provided 9/28/91.
of Primary Containment Penetrations Part c: Leakage Concerns NRC NRC will review.
3.2.12 Number of Chillers and Pumps GE Response will be provided in (2) in the System May.
9.2.13 HVAC EMERGENCY COOLING WATER SYSTEMS 13 Missile Protection GE Response provided 3/28/91.
3.2.13 Protection from Water Hammer GE Response provided 3/28/91.
(2) 3.2.13 Chemical Feed Tank GE Response will be provided in (5) May.
9.2.13 Number of Divisions and GE Response ~will be provided in (6) Associated Cooling for EDG May.
9.2.13 Referenced Number of P& ids GE Response will be provided in
- (7) May.
3.2.13 Pressure and Functional Testing GE Response provided 3/28/91.
(8) t l
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SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
- PLANT SYSTEMS OPEN ISSUES Item Number Subiect Action Comments 10.2 TURBINE GENERATOR
- 10.2 Periodic Tests of Turbine Valves GE Response provided on page (1) 10.2-8 attached.
10.3 MAIN STEAM SUPPLY SYSTEM 10.3 Main Steam Line Classification GE GE is still dicsussing with (1) Mechanical Engineering Branch.
10.4.2 MAIN CONDENSER EVAUATION SYSTEM 10.4.2 Radiation Monitoring of Exhaust GE Response provided on page (1) 10.4-5 attached.
104. 3 TURBINE GLAND SEAL SYSTEM 10.4.3 Local Exhaust Radiation GE Response provided on pages (1) Monitoring 10.4-6 and 10.4-7 attached.
A
( )3 Interface Regarding the Switch- GE Response provided on page
'rt) over to Auxiliary Steam Supply 10.4-17 attached.
10.4.4 TURBINE BYPASS SYSTEM 10.4.4 Turbine Bypass Valves GE Provided in Amendment 16.
(1)
- 10. 4. 5 CIRCULATING WATER SYSTEM' 10.4.5 CWS SSAR Table Reference GE Provided in Amendment 15.
(1) 10.4.5 Flooding Protection GE Response provided on page (2) 10.4-10 attached.
- Several clarifying changes were made to Chapter-10 in addition to responding to the open items.
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SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON-(~N x,,,/ PLANT SYSTEMS OPEN ISSUES I
Item Number Subiect Action Comments >
10.4.7 CONDENSATE AND FEEDWATER SYSTEM 10.4.7 Size of Feedwater Line GE Resolved.
(1) 10.4.7 Power Source for Motor Operated NRC NRC will provide further (2) Gate Valve. guidance.
10.4.7 CFS Seismic Category and Group _NRC Amendment 14 addressed this (3) Classifications item. NRC will review.
9.1.3 SPENT FUEL POOL COOLING AND CLEANUP-SYSTEM
- 9.1.3 Isolation of'FPC from Suppression 'GE Response provided on-page (1) Pool Cleanup System 9.1-5 attached. .;
9.1.3 Emergency Source of SPF Water. NRC NRC=will-evaluate-GE's response ;
(2) to previous: questions. -(See GE .
response ;on.page-9.1-5 attached)
\3 FPC Design Seismic Classification NRC NRC will reevaluate previous Cd
\_ similar-information.
9.1.3 FPC Design-single Active Failure,~ GE Response provided'on page (4) LOOP Sizing of-Heat.Exchangers 9.1-51 attached.-
-9.1.3 Provision of a FPC System lGE Response provided.on page:
(5)' Components Description Table . 9.1-3 attached. <
9.1.5 OVERHEAD HEAVY LOAD HANDLING SYSTEM 9.1. 5 - - OHLHS ! Design /RG 1.29f and RG 1.13 GE See Attachment.1.-
(A.1) 9.1.5 Non-Seismic CategeryfI Load. .GE- See Attachment 1..
.(A.2) Handling-Equipment.
9.1.5 Refueling' Bridge Crane References GE- See Attachemnt 1.
(A.3)- and Seismic': Classifications. .
9.1. 5 : Housing.of? Load Handling-Equipment GE- See. Attachment 1..
(A.4)_ for Steam Tunnel Servicing 2
-9'=1.5
. Spent Fuel Crane. Lifting Height: .GE' :See Attachment =1..
(B.1)
-veral clarifying. changes were made tcr Section 9.1.3 in addition to- 4
'Tesponding to the open items.-
_ . _ ________mm_________-_ .__ _ _ _ _ ___s.
SUMMARY
STATUS OF GE/NRC MARCH 4-6, 1991 MEETING ON
,3 j PLANT SYSTEMS OPEN ISSUES w .a humber Subiect Action Comments l
p.1.5 Control of Heavy Load Movement GE See Attachment 1.
(B.2) Over Spent Fuel Pool 9.1.5 Protection of Safety-Related GE See Attachment 1.
(B.3) Equipment During Heavy Load Oper.
9.1.5 Additional Details Concerning GE See Attachment 1.
(C.1) Hoists
.1.5 Single-Failure Criteria for GE See Attachment 1.
C.2) Hoists and lifting Devices
.1.5 Limit and Safety Devices and c.3) OHLHS FMEA
- Limit and Safety Devices GE See Attachment 1.
Portion
- OHLHS FMEA Portion GE See Attachment 2.
.1.5 Heavy Load Operations in the GE See Attachment 1.
{C.4) Control Building l 11.3.1 GASEOUS WASTE MANAGEMENT
- 9. 3.1Monitoring l
of the Exhaust the Turbine Building from GE See, response to item 10.3(1).
(1) 1.3.1 Sensitivity of Secondary Cont. GE -Response provided on page-(2) Exhaust Monitoring 11.5-3 attached.-
$1.3.1 Relative-Location of the. Plant- NRC Provided in' Amendment-16. NRC
~(3) Release Point will' evaluate.
11.4.1 SOLID WASTE _ MANAGEMENT' SYSTEM 1.4.1 Compliance with<10CFR61 .GE Response provided 3/28/91.
(1)-
11.4.2 EVALUATION FINDINGS i1.4.2 Radwaste Storage Capacity ' GE
_ Response provided 3/28/91..
-(1) -
$ 1. 4 '. 2 Cement Glass as a Waste GE Response;provided 3/28/91.
l .( 5) ' Solidification Agent
'1.4.2 Incinerator Description- GE Response provided 3/28/91.
) (1):
4.2 'Inconsistenciesiin-Addressing < GE Response will be provided in (5) (2) Estimated Waste Shipments May.-
bl.4.2E Illegible P& ids None Resolved.
- 5) (3)
--- - - - . - - . - . .~ _ - - .- - - . . - . .. - ~ . - ..
11.5.1 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND
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SAMPLING SYSTEM *
- 11.5.1 Classification of Exhausts as GE Response provided on page
' (1.a) Non-Radioactive 11.5-1 attached.
i J11.5.1 Direct Effluent Release Paths GE Response provided on pagrs (1.b) to the Environment 11.5-1 and 11.5-8 attached.
! 11.5.1 Reactor Service Water Effluent GE Response provided 3/28/91.
(1.c) Monitoring l11.5.1 Continuous Monitors Channel GE Response provided on pages (2) Ranges and Sensitivities 11.5-13,14,15,16,17,18 and 21 attached.
(3) Monitoring ,
!11.5.1 Plant Vent Exhaust Sampling GE Response provided on pages (4) 11.5-20 and 11.5-22-attached.
l11.5.1 Design and Qualification of GE/NRC .Both NRC and GE will review (5) Accident Monitoring the adequacy of the info.
Instrumentation -provided in Amendment 16.
11.2 LIOUID WASTE MANAGEMENT SYSTEM l(
(This additional open-item was not discussed during the meeting)1
~
. Local Alarm Capability for the GE- Response:provided on next page.
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Condensate Storage' Tank i
- Several clarifying: changes were made:to Section 11.5 iniaddition to.
responding to the.open' items.- i 6
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1 11.2 Liquid Waste Management System - Additional Open Item i
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- i. Discussion of condensate storage tank overflow is provided in L Response to Question 430.156. Local display of the-condensate ,
- storage tank water level is not provided since there are no i operator functions (such as starting or stopping of pumps) at-the condensate storage tank.
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ATTACHMENT l
/h U 9.1.5 Overhead Heavy lead Handling Systems 9.1.5.1 Design Bases The equipment covered by this subsection handle items considered as heavy loads that are handled under conditions that mandate critical load handling compliance.
Critical load handling conditions include loads, equipment, and operations, which if inadvertent operations or ec uipment malfunctions either separately or in combinatic 1, could cause; () a release of radioactivity, (2) a criticality accident, (3) the inability to cool fuel withm reactor vessel or spent fuel pool or (4) prevent safe shutdown of the reactor. This includes risk assessments to spent fuel and storage pool water levels, cooling of fuel pool water, new fuel criticality. This includes all components and equipment used in moving any load weighing more than one fuel assembly including the weight of its associated handling devices (i.e., one ton).
The reactor building crane as desi ned shall provide a safe and effective means for transporting heavy loads includin he handhng of new and spent fuel, plant equ,ipment and service tools. Saf handling includes design considerations for maintaining occupational radiation exposures as low as practicable during transportation and handling.
Where ap 31icable, the a apropriate seismic category, safety class quality group, ASME, A NSI, industrial and electrical codes have been identified (see Table 3.2-1 and Table 9.1-6). The designs will conform to the relevant requirements of General
()
v Design Criterion 2,4 and 61 of 10CFR Part 50 Appendix A.
The lifting ca?acity of each crane or hoist is designed to at least the maximum actual or anticipatet weight of equipment and handling devices in a given area serviced.
The hoists, cranes, or other hfting devices shall comply with requirements of ANSI N14.6, ANSI B30.9, ANSI B30.10 and NUREG 0612 Subsection 5.1.1(4) or 5.1.1(5).
Craaes and hoists are also designed to the criteria and guidelines of NUREG 0612 Subsection 5.1.1(7), ANSI B30.2 and CMAA-70 specifications for electrical overhead traveling cranes, including ANSI B30.11, ANSI B30.16. and NUREG-0554 as applicable.
9.1.5.2 System Description 9.1.5.2.1 Reactor Building Crane The Reactor Building is a reinforced concrete structure which encloses the reinforced concrete containment vessel (RCCV), the refueling floor, new fuel storage vault, the storage pools for spent fuel and the dryer and separator and other equipment. The reactor building crane provides. heavy load lifting capability for the refuelin g floor. The main hook (150 ton capacity) will be used to lift the concrete shield b ocks, drywell head, reactor pressure vessel (RPV head insulation, RPV head, dryer / separator strong back, RPV head strongback carousel, new fuel shipping containers, and spent fuel shipping cask. The or erly placement and movement aaths of these componentsby the reactor building main crane precludes transport 01 these heavy loads over the spent fuel storage pool or over the new fuel storage vault.
The RB crane will be used during refueling / servicing as well as when the plant is online. During refueling / servicing, the crane handles the shield plugs, drywell and
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s reactor vessel heads, steam d er and separators, etc. (see Table 9.17). Minimum crane coverage must include kB refueling floor laydown areas, and RB equipment storage, pit. During, normal plant operation the crane will be used to handle new fuel shipping contamers and the spent fuel shi 3 ping casks. Minimum crane coverage must include the new fuel vault, the RB equipment hatches, and the spent fuel cask loading and washdown pits. A description of the refueling procedure can be found in section 9.1.4.
The RB crane will be interlocked to prevent movement of heavy loads over the spent fuel storage portion of the spent fuel storage pool. Since the crane is used for handling large heavy objects over the open reactor the crane is of type I design. The reactor building crane shall be designed to meet the single-failure-proof requirements of NUREG 0554.
9.1.5.2.2 Other Overhead load Handling Systems 9.1.5.2.2.1 Upper Drfwell Servicing Equipment i The upper drywell arran ement provides servicing access for the main steam ,,
isolation valves (MSIVs feedwater isolation valves, safety relief valves (SRVs),
Emergency Core Cooli g Systems (ECCSs) isolation valves, an.1 drywell cooling coils, fans and motors. Access to the space is via the RB through either the upper drywell personnellock or equipment hatch. All equipment is removed through the upper drywell equi 3 ment hatch. Platforms are provided for servicing the Feedwater and mainsteam iso' ation valves, safety relicf valves, and drywell coofmg equipment -
D
() with the object of reducing maintenance time and operator exposure. The MSIVs.
SRVs, and feedwater isolation valves all weigh in excess of 2000 kg. Thus are considered heavy loads.
With maintenance activity only being performed during a refueling outage, only safe shutdown ECCS piping and valves need be protected from any inadvertent load drops. Since only one division of ECCS is required to maintam the safe shutdown condition and the ECCS divisions are spatially separated, an inadvertent load drop that breaks more than one division of ECCS is not credible. Inaddition, two levels of piping support structures and equipment platforms separate and shield the ECCS-pipmg from t be heavy loads transport path.
This protection is adequate such that no credible load drop can cause either (1) a release of radioactivity, (2) a criticality accident, or within reactor vessel or spent fuel pool;etherefore, upper drywellth(3) the inability servicing to cool fuel equipment is not subject to the requirements of subsection 9.1.5. .
9.1.5.2.2.2 Lower Drywell Servicing Equipment handling and 1 The lower drywell transportation (L/D) operations for_arrangement RIP, and FMCRD. provides The lower for servicing,drywell OHLHS -
consists of a rotating equipment service platform. chain hoists, FMCRD removal l machine, a RIP removal machine, and other special purpose tools.
The rotating equipment platform arovides a work surface under the reactor vessel:
to support the weight of personne , tools, and equipment and to facilitate g transportation moves and heavy load handling operations. The platform rotates .
3600 in either direction from its stored or " idle" position. The platform is designed to accomnwdate the maximum weight of the accumulation of tools and equipment l
77 plus a maximum sized crew. Weights of tools and equipment are specified in the V mterface control drawings for the equipment used in the lower drywell. Special hoists are provided in the lower drywell and reactor building to facilitate handling of ,
these loads.
(1) Reactor Internal Pump Servicing There are 10 RIPS and their supp_orting instrumentation and heat exchangers in the L/D that require servicing. The facilities provided for servicing the RIPS include:
(a) L/D equi ament alatform with facilities to rotate the motor from vertical to horizonta. and p ace it on a cart for direct pull out to the RB The equipment platform rotates to facilitate abgnment with the installed pump locations.
(b) Attachments points for rigging the RIP heat exchanger into place. The RIP heat exchanger can be lowered straight down to the equipment platform.
(c) Access to the RIP equipment platform is via stairs. There is a ladder access to the RIP heat exchanger mamtenance platform.
(d)The L/D equipment tunnel and hatch are utilized to remove the RIP motors from the lower drywell.
(e) The RIP motor servicing area is directly outside the L/D equipment hatch.
V The 10 RIPS have wet induction motors in housings which protrude into the lower drywell from the RPV bottom head. These are in a circle at a radius of 3162.5 mm from the RPV centerline. For service, the motor is removed from-below and outside, whereas the diffuser, impeller and shaft are removed from above and inside the RPV.
The motor, with its lower flange attached, weighs approximately 3300 kg, is 830 mm in diameter and 1925 mm high. The flange has cars" that extend from two sides,1800 apart. These ears, which are used to handle the motor, increase the flange diameter to 1200 mm for a width of 270 mm.
The motor, suspended from jack screws, is lowered straight down out of its housing onto the equipment platform. The motor is then moved, circumferentiall through thetpment equ,yremoval and lifted onto a tunnel L/D equipment rail mounted transport and hatch. The motor is cart for direc transported horizontally out of the contamment and into the motor service shop immediately adjacent to the L/D equipment hatch.
The RIP servicing equipment includes the cart to transport the motor from the service area through the equipment hatch to the L/D equipment platform. The'-
interface for this ec ul ament is the rails on the equiJ) ment platform that permit locating the motor 3e ow.its nozzle on the RPV The servicing equipment includes a chain hoist for rotating the RIP motor from horizontal to vertical and -
a hydraulic lift to raise it from the equipment platform to its installed position m below the RPV. Facilities are provided for handling s,tud tensioners, blind
) flanges, other tools, drains and vents used in RIP servicing.
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Servicin of the RIP heat exchan er, such as removal of the tube bundle, will be
'v) accompl shed by rigging to attachment points on the RPV pedestal and structural steel m the area. A direct vertical removal path is provided from the heat exchanger installed position to the equipment platform. The operation is performed by a chain hoist. This is considered to be a nonroutine servicing operation.
These RIPS are serviced only when the reactor is in a safe shutdown mode. In addition, there is no safety related equipment below either the RIPS or the RIP heat exchangers. Inadvertent load drops of either component can not cause 1 a release of radioactivity, (2) a criticality accident, or (3) the inability either to cool (fu)el within reactor vessel or spent fuel pool; therefore, the RIP servicing equipment is not subject to the requirements of subsection 9.1.5.
(2) Fine Motion Control Rod Drive Servicing There are 205 FMCRDS in the L/D that require servicing. There are two types of servicing operations: replacement of the FMCRD drive mechanism and -
motor and seal, replacement. Separate servicing equipment is provided for each of these operations.
(a) The FMCRD drive servicing machine has its own mechanisms for rotating and raising FMCRD drive assemblies from a carrier on the equipmem platform to their installed position. This servicing machine interfaces with the L/D equipment platform, which permits positioning the servicing machme under any of the 205 FMCRDs.
jq
'd (b) A separate machine and cart are provided for servicing FMCRD motors and seal assemblies immediately and outside the L transp/D equipment hatch.orting them to the serv related equipment below either component. Inadvertent load There is no safety RD serviemg equipment can not cause drops by the FMC 1) a release of
- 2) a criticality accident, or 3 the inability to cool uel within ,
radioactivity,l(or reactor vesse spent fuel pool; therefor (e,)the FMCRD servicing equipm '
not subject to the requirements of subsection 9.1.5.
9.1.5.2.2.3 Mainsteam Tunnel Servicing Equipment
- The mainsteam tunnelis a reinforced concrete structure that surrounds the mainsteam lines and feedwater lines. The safety related valve area of the mainsteam tunnelis located inside the reactor building. Access to the mainsteam tunnelis during a refueling / servicing outage. At this time MSIVs or Feedwater Isolation valves and/or feedwater check valves may be removed using permanent overhead monorail t and placed on floor.y7e hoists. Transported by monorail out of the steam t ceiling hatch by valve service shop monorail. During shutdown, all of the piping and -
valves are not rec uired to operate. . Any load drop can only dama e the other valves i or piping within t?ie main steam tunnel. Inadvertent load drops b the mainsteam tunnel sevicing equipment can not cause either (1) a release of ra ioactivity, (2) a .
criticality accident, or (3) the inability to cool fuel within reactor vessel or spent fuel O pool; therefore, the mainsteam tunnel servicing equipment is not subject to the J requirements of subsection 9.1.5.
I
O Q 9.1.5.2.2.4 Other Servicing Equipment Outside the RCCV, the mainsteam tunnel, and the refueling floor no safety related component of one division shall be routed over any, portion of a safety related portion of another division. The ABWR has three independent and separate ECCS divisions. A load drop accident in one division causing the complete loss of a second division is not credible. Hence inadvertent load drops can not cause either (1,) a release of radioactivity,(2) a criticalit accident, (3 within reactor vessel or spent fuel pool, orsafe 4) shutdown preventofthe) the the inability to cool fu reactor; therefore, all servicing equipment ocated outside the RCCV, the mainsteam tunnel, or the refueling floor are not subject to the requirements of subseetion 9.1.5.
9.1.5.3 Applicable Design Criteria For All OHLH Equipment All handling equipment sub ect to heavy loads handlin,; criteria will have ratings consistent with lifts required and the design loading wi .1 be visibly marked.
Cranes / hoists or monorail hoists will pass over the centers of gravity of heavy equipment that is to be lifted. In locations where a single monorail or crane handles several pieces of equipment, the routing shall be such that each transported piece will pass clear of other parts. If, however, due to restricted overhead space the transported load cannot clear the installed equipment, then the monorail may be offset to provide transport clearance. A lifting eye offset in the ceiling over each piece of equipment can be used to provide a Y hft so that the load can be lifted upward until free and then swung to position under the monorail for transport. !
(s1 V Pendant control is required for the bridge, trolley and the auxiliary hoist to provide efficient handling of fuel shipping contamers durmg receipt and also to handle fuel during new fuel inspection. The crane control system will be selected considering the long lift required through the equipment hatch as well as the precise positioning requirements when handling the RPV and drywell heads, RPV internals, and the RPV head stud tensioner assembly. The control system will provide stepless regulated variable speed capability with high empty hook speeds. Efficient handling of the drywell and RPV heads and stud tensioner assembly require that the control system provide spotting control. Since fuel shipping cask handling involves a long duration lift, low speed and spotting control, thermal protection features will be incorporated.
Heasy load equipment is also used to handle light loads and related fuel handling tasks. Therefore, much of the handling systems and related design, descriptions, operations, and service task information of Subsection 9.1.4 is applicable here. The cross reference between the handling operations / equipment and Subsection 9.1.4 is provided in Table 9.17. See Table 9.1-8 for a summary of heavy load operation.
Transportation routing drawings will be made covering the transportation route of every piece of heavy load removable equipment from its installed location to the appropriate service shop or building exit. Routes will be arranged to prevent congestion and to assure safety while permitting a free flow of equipment being _d serviced. The frequency of transportation and usage of route will be documente based on the predicted number of times usage either per year and/or,per refueling n
- i or service outage.
Safe load paths / routing will comply with the requirements of NUREG 0612 -
Subsection 5.1.1(1).
1
A k.) 9.1.5.4 Equipment Operating Procedures Maintenance and Service Each item of e ii ii ill be described on an interface control This will(includ)quipment requ r ng serv c ng wdiaj; ram IC e pull space for internal parts, access for tools, handling equipment, and alignment requirements. The ICD will specify the weights of large removable and describe installed lifting parts, show the location of their centers accommodations such as eyes and trunnions. An of gravity, instruction manual will desenbe maintenance procedures for each piece of equipment to be handled for servicing.
Each manual will contain suggestions for riggmg and lifting of heavy parts and identify any special lifting or handling tools required.
All major handling equipment components: cranes, hoist, etc., will be provided with an operating instruction and maintenance manual for reference and utilization by operations personnel. Handling equipment operatin procedure will comply with the requirements of NUREG 0612 Subsection 5.1.1( ).
The operational programs for maintenance and servicing are described in Subseetion 9.1.5.6.
9.1.5.5 Safety Evaluations The cranes, hoists, and related lifting devices used for handling heavy loads either satisfy the single failure proof guidelines of NUREG 0612 Subsection 5.1.6, including NUREG-0554 or evaluations are made to demonstrate compliance with p) the recommended guidelines of Section 5.1, including Subsection 5.1,4 and 5.1.5.
(J The ec ui) ment handling components over the fuel pool are designed to meet the single Jai ure proof criteria to satisfy NUREG 0554. Redundant safety interlocks and limit switches are provided to prevent transporting heavy loads other than spent fuel by the refueling bride crane over any spent fuel that is stored in the spent fuel storage pool.
A transportation routing study will be made of all planned heavy load handling moves to evaluate and minimize safety risks.
Safety evaluation of related light loads and refueling handling tasks in which heavy load equipment is also used are covered in Subsection 9.1.4.3.
9.1.5.6 Inspection and Testing Heavy load handlin equipment is subject to the strict controls of Quality Assurance (QA),incorporatin the requirements of Federal Regulation 10CFR50, Appendix B.
Components define as essential to safety have an additional set of engineering specified " Quality Requirements" that identify safety related features which require specific QA verification of compliance to drawmg/ specification requirements.-
Prior to shipment, every lifting equi) ment compcnent requiring inspection will be reviewed by QA for compliance anc thra the required records are available.
Qualification load and performance testing, including nondestructive examination n (NDE) and dimensionalinspection on heavy load handling equipment will be j performed prior to QA acceptance. -Tests may include load capacity, safety overloads, life cycle, sequence of operations and functional areas.
l l
pment is received at the site it will be inspected to ensure no dama e has When occurre e[during transit or storage. Prior to use and at periodic intervals each piece of equipment will be tested agam to ensure the electrical and/or mechanical functions are operational including visual and, if required, NDE inspection.
Cranc inspections and testing will comply with requirements of ANSI B30.2 and NUREG 0612, Subsection 5.1.1(6).
9.1.5.7 Instrumentation Requirements The m6arit of the heavy load handling equiprent is manually operated and controlfed the operator's visual observations. This type of operation does not necessitate t e need for a dynamic instrumentation system.
Load cells may be installed to provide automatic shutdown whenever threshold limits are exceeded for critical load handling operations to prevent overloading.
9.1.5.8 Operational Responsibilities Critical heavy load handling in operation of the plant shallinclude the following documented programs for safe administration and safe implementation of operations and control of heavy load handling systems:
(1) Heavy Load Handling System and Equipment Operating Procedures.
(2) Heavy Load Handling Equipment Maintenance Procedures and/or Manuals.
(U)
(3) Heavy Load Handling Equipment Inspection and Test Plans; NDE, Visual, etc.
(4) Heavy Load Handling Safe Load Paths and Routing Plans.
(5) OA Program to Monitor and Assure Implementation and Compliance of Heavy . '
Load Handling Operations and Controls.
(6) Operator Qualifications, Training and Control Program.
O
ATTACHMENT 2
,c\
\_j RESPONSE TO OHLHS PORTION OF OPEN ITEM 9.1.5(c.3)
As discussed in Section 15B.1, FMEAs are provided for two ABWR systems and one major component which presents a significant change from past ABWR designs. Specifically, FMEAs are included in Appendix 15B fors (1) control drive systems (with emphasis on the fine motion control rod drive),
(2) essential multiplexing system, and (3) reactor internal pump.
Regulatory Guide 1.70 requires FMEAs to be performed on selected subsystems of Chapters 6,7 and 9. However, GE considers that tFs plant nuclear safety operational analysis (NSOA) of Appendix iSA and the probabilistic evaluations of Appendix 19d adequately address single failures for those systems and components which are similar to past BWR designs. Since the design of the ABWR
, OHLHS instrument and control system is similar to past designs,
) GE believes that it is unnecessary to perform a FMEA on the OHLHS
' / instrument and control system.
/~
f
./
ABM 23^sioors Rrv s Standard Plant O 3.4 WATER LEVEL (FLOOD) DESIGN 3.4.1.1.1 Flood Protection from External V Sources The types and methods used for protecting the ABWR safety related structures, systems and Seismic Category I structures that may be components from external flooding shall conform affected by design basis floods are designed to
$ withstand the floods postulated in Table 2.01 to the guidelines defined in RG 1.102, %
using the hardened protection approach with $
Criteria for the design basis for protection structural provisions with incorporated in the against external flooding shall conform to the plant design to protect safety related requirements of RG 1.59. The design criteria for structures, systemt., and components from protection against the effects of compartment postulated flooding. Seismic Category I flooding shall conform to the requirements of structures required for safe shutdown remain ANSI /ANS.56.11. The design basis flood levels accessibic during all flood conditions.
are specified in Tab!c 3.41.
Safety related systems and components are 3.4.1 Flood Protection flood protected either because of their location above the design flood level or because they cre This section discusses the flood protection enclosed in reinforced concrete Seismic Category measures that are applicable to the standard ABWR I structures which have the following plant Seismic Category I structures, systems, and requirements:
$ components for both external flooding and R postulated flooding from plant component failures.- These protection measures also apply (1)lesswall thicknesses than.two feet; below flood lev to other structures that house systems and components important to safety which fall within - (2) water stops provided in all construction the scope of plant specific. joints below flood level; 3.4.1.1 Flood Protection Measu.ts for Seismic (3) watertight doors'and equipment batches
- Category I Structurer installed below design flood level; and The safety related systems and _ components of (4) waterproof coating of externalsurfaces.- g' the' ABWR Standard Plant are located in the g reactor, control, and radwaste buildings which . Waterproofing of foundations and walls of g are seismic category 1 structures. These Seismic Category 1 structures below grade is -
' structures together with those identified in accomplished principally by the use of water- CD damage flood protection of safety.related addition to water stops, waterproofing of the [,
Table 3.41 are protected against external flood stops at expansion and construction joints. In systems and_' components is provided for all plant structures that house safety relate A systems and components is provided up to m (3_ $
- gl postulated design flood levels and conditionsg described k Table 2.01. P
- from co:.oponent failures in the building compart. external surfaces from exposure to water.-
ments does not adversely affect plant safety nor .-
i - t w s ERJ does it. represent any hazard to the public. Additional specific provisions for flood - 3,41,1,l.-
protection include administrative procedures to Structures which house the safety related assure that all watertight doors and hatch ' 9 2 D.
~"
covers are locked in the event f of a flood
-gl equipment g idertified in Tableand 3.41,'offer flood of Descriptions protection are-If local seepage occurs through the -
these .warning.
I
' structures are provided in Subsection 3.8.4 and walls, it is controlled by sumps and sump pumps.
3.8.5. Exterior or access openings and-In the event of a flood, flood levels ~take a j l penetrations level are identified that in Table are 6.2below
- 9. the design relativelyflood long time to develop. This allows D
G Amendment i6 341 i l
m sse r 2 4. i . i. i 9 7.9 ($)
f '
l The flood protection measures that are described above also guard t
against flooding from on-site storage tanks that may rupture.
- The largest is the condensate storage tank that has a capacity of 2,110 cubic meters. This tank is constructed form stainless steel and is located between the turbine building and the
! radweste building where there are no direct entries to-these buildings. All plant entries start one_ foot:above grade. .Any
$ flash flooding that may result from tank rupture vill. drain;away j from the site and cause no damage to site-equipment.
i 3
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ABWR 234siooxt uv n
.. Standard Plant i
O)( requirement for redundant separation is The design criteria for restraints is given in m e t. Other redundant divisions are Subsection 3.6.2.3.3.
available for safe shutdown of the plant and no further evaluation is performed. ~ 3.6.1.33 Specific Protection Measures (4) If damage could occur to morr than one -(1) Nonessential systems and system components -
division of a redundant essential system are not required for the safe shutdown of
, within-30 ft of any high energy piping, the reactor, nor are they required-for the-other protection in the form of barriers, limitation of the offsite release in the shields, or enclosures is used. These event of a pipe rupture. However, wblic methods of protection are discussed in Sub- none of this equipment is needed during or i section 3.6.1.3.2.3.' Pipe whip restraints following a pipe break event, pipe whip ~ l
-i as discussed in Subsection 3.6.1.3.2.4 are protection is considered where a resulting -
used if protection from whipping pipe is not f ailure of a nonessential system or '
possible by barriers and shields. component could initiate or escalate the pipe break event in an essential system or 3.6.13.2.3 Barriers, Shields, and Enclosures component, or in another nonessential system whose failure could affect an essential Protection requirements are met through the system.
protection afforded by the walls, floors, columns, abutments, and foundations in many (2) For high energy piping systems penetrating cases. Where adequate protection is not already through the containment.. isolation valves present due to spatial separation or_ existing. are located as close to the containment as plant features, additional barriers, deflectors, possible.
or shields are identified as necessary to meet m the functional protection requirements, ) (3) The pres _sure, water level, and flow sensor
- instrumentation for those essential systems,
~
Darriers' or sbleids that are identified a3 which are required to function following_a
' pipe rupture, are protected.
necessary by the use of specific break locations L in the drywell and steam tunnel are designed for .
i the specific loads associated with the particula (4) High energy fluid system pipe whip -
break location. restraints and protective. measures are designed so that a postulated break in one
. Barriers or shields that are identified as- pipe could not,in turn, lead to a rupture IM necessary by the HELSA evaluation (i.e., based on of other nearby pipes or' components if the l
3.G.I.3.2.3 no specific break locations), are designed foc secon'dary rupture could result in worst case loads. The closest high. energy pipe consequences that woul_d be considered
- location and resultant loads are used to size the unacceptable for the initial postulatedL
- barriers, break.
- g.g LO #' 3.6.1.3.2.4 Pipc Whip Restraints - _.
(5) For any_ postulated pipe rupture, the - "
Peou . . .
. structural integrity of the _ containment..
- - Pipe whip restraints are used where pipe break structure is' maintained.
- In: addition, for pro:cetion requirements could not be satisfied those postulated ruptures classified as a using spatial separation, barriers, shleids, or! ' loss of reactor coolant, the design leak ~ q enclosures alone. : Restraints are located based ; : tightness of the cont _ainment fission product on the specific break locations determined in ac- barrier is maintained?
cordance with Subsections 3.6.2.1.4.3 and 3.6.2.-
1.4.4 After the restraints' are located, the (6) Safety / relief valves (SRV) and the_ reactor I
piping and essential systems are evaluated for- : core-isolation cooling (RCIC) system steam.
jet impingement and. pipe whip.' For those cases .line are located ~and restrained so.that a m . where Jet impingement damage could still occur, pipe failure would not prevent depressuri. I i ; tarriers, shleidst or enclosures are utilized. zation.- '
J
. Amendment 7 34 5-t
, + - -* * - y , ..m
i m s m 3. c. i .u. 3 .
L Barriers or shields that are identified as necessary by the use of specific break locations in the drywell are designed for the specific loads associated with the particular break location.
The steam tunnelis made of reinforced concrete 2m thick. A steam tunnel
. subcompartment analysis was performed for the xntulated rupture of a meinsteam line and for a feedwater line mainsteam line break was fou(see subsection 6.2.L.3.1).
pressure The peak from a feedwater pressure ,
line break was found to be 3.9psig. The steam tunnelis c esigned for the effects of !
an SSE coincident with a high energy line break inside the steam tunnel. Under this conservative load combination, no failure in any portion of the steam tunnel was found to occur; therefore, a high energy line break inside the steam tunnel will not effect control room habitability.
The MSIVs and the feedwater isolation and check valves being inside the tunnel-shall be designed for the effects of a line break. The details of how the MSIV and feedwater isolation and check valves functional capabilities are protected against stulated pipe failures will be provided by the ap the effects referencing theof ABthese gR design see subsection 3.6A1, items 4 and 6 .plicant i a
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- I i
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.. v
\ 23A6100AE Standard Plant Rrv n including deadweight and SSE (inertial) (1) A summary of the dynamic analyses components, applicable to high energy piping systems O in accordance with Subsection 3.6.2.5 of Shielded Metal Are (SMA%9 and Submewed Are Regulatory Guide 1.70. This shall (SA%S Weldt includc:
The flow stress used to construct the master (a) Sketches of applicable piping systems curve is 51 ksi showing the location, size and .
orientation of postulated pipe breaks The value of SI used to enter the master and the location of pipe whip curve for SMAW and SAW is restraints and jet impingement barriers.
SI = M (Pm+Pb+Pe)Z (8)
(b) A summary of the data dcTeloped to where select postulated break locations including calculated stress
= the combined primary bending stress, intensities, cumulative usage factors Pb i acluding deadweight and seismic and stress ranges as delineated in components. BTP MEB 31.
Pe = combined expansion stress at normal (2) For failure in the moderate energy piping operation. systems liste d in Table 3.6 6, .
descriptions showing how safety related ;;;
Z = 1.15 (1.0 + 0.013 (OD 4)] for SMAW, systems are protected from the resulting $
(9) jets, flooding -and other adverse-environmental effects.
Z = 1.30 (1.0 + 0.010 (OD-4)] for SAW, O
V (10) (3) Identification of protective measures provided against the effects of ,
and postulated pipe failures in each of the y systems listed in Tables 3.61, 3.6 2 and OD = pipe outer diameter in inches. 3.6-4.
When the allowable flaw length is determined (4) The details of how the MSIV functional from the master curve at the appropriate St. capability is protected against the 3 value, it can be used to determine if the- effects of postulated pipe failures. ;
required margins on load and flaw size are met using the following procedure. (5) Typical examples,~ if any, where
( protection for safety related systems and For the method of load combination described - components against the dynamic effects of in item (5), let M = 1.4, and if the pipe failures include their enclosure in sq L
allowable flaw length from the master curve suitably. designed structures or s' is at least equal to the leakage size flaw, compartments (including ~any. additional then the margin on load is met, drainage system or. equipment:
3.6.4 Interfaces 3.6.4.2 Lenk Before Break Analysis Report b IN 38RT ' >
3.6.4.1 Details of Pipe Break Analysis Results 3 G.9 I and Protection Methods As required by Reference 1, an LBB analysis 3,4, g' report shall be prepared for'the piping systems The following shall be provided by the proposed for inclusion from the analyses for the *#
M^"
a};nlicant referencing the ABWR design (See = - dynamic effects due to their failure. The report Su'bsection 3.6.2.5): . shall include only the piping stress analysis Poenua i
Amendment 10 3.6 27-q l
O l
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INSGRT 3,G,4,(
(6) The details of how the feedwater line check and feedwater isolation valves functional capabilities are protected against the effects of postulated pipe failures, O
l ABWR Standard Plant 08 ^ N'" #
h 2am trv c I forms a part of the secondary containment room design, the peak differential pressures are boundary is designed to at least Seismic Category not to exceed the design differential pressure.
O I and ASME Section 111, Class 3 requirements.
Some lines have no specialisolation provisiens 6.2.33.1.2 Design Features and are not ASME Section !!! or Selsmic Category I if an analysis shows that exfiltration would Tbc following paragraphs are brief not occur in the event of failure of that pipe descriptions of the compartments analyzed for
[i.e., the .1/4 in, water gage pressure pressurization. A more detailed description differential would be maintained]. will be found in Subsection 3.8.l[ Figure 6.2 37 s h o w s t h e s c h e m a t i c l a y o u t of,,,thL . w el For architectural openings the inicakage is secondary containment compritments with the ble d based on 1/4 in, water gage pressure katerconnected vent pathsafFigures 6.2 28 "n"
differential. All doors have a vestibule with a through 6.2 36 are the plan and elevation second (outer) door. llVAC and electrical drawings showing cotoponefut and equipment penetrations are designed to minimize leaks, and locations and configurations 9 Tables 6.2 3 and HVAC system is designed and tested for isolation r6.2 4 tabulate tbc compartment free volumes and '
under accident conditions, initial room conditions, flow path parametersj land blowout panel characteristlesO WSp.*T Table 6.2 9 provides a listing of secondary g, , q, , 3 , g , t h containmeat openings. All piping and cabletray 6.2.33.1.2.1 Reactor Cort Isolation Cooling penetrations will be scaled with a scaling (RCIC) Compartment compond for leakage and fire protection. All g doors are vestibule type with card reader access The RCIC compartment is located in the
- security systems that are monitored (See secondary containment at El(-)13200 mm. Tbc Subsection 13.6.3.4). The HVAC penetrations are design basis break for the RCIC compartment is designed to close on a design basis accident (See the double end-d bres.k of the 6.in RCIC steam Subsection 9.4.3 on reactor building HVAC). supply line. This line is a high energy line Testing procedure and frequency can be found in out to the normally closed isolation valve the plant technical specifications. Inside the RCIC compartment,+edhpplies high energy steam to the RCIC turbine m the event of 6.2.33 Design Evaluation rea. tor vessel isolation. .*-- iW tsa'r 6 4. 3. 3.1 @
The design of the secondary containment 6.233.1.2.2 Reactor Water Cleanup (RWCU) boundaries is described in tbc preceding Equipment and Yalves Rooms subsection. Evaluation of this design, such that all regulatory requiremt nts are met, are given in The RWCU equipment (pump, beat exchanger, the following subsections: and filter /demineralizer) and valves rooms are located in the 00 2700 quadrant of the (1) 6.5.1 Standby Gas Treatment System reactor building. The floor elevations are from
(-)L3200 mm to (.)200 mn with separate rooms for J (2) 9.4.5 Reactor Building liVAC System the equipment ar ! valves /High energy piping iconnects the equipment and valve rooms and is 6.233.1 Compartment Pressurization routed to the steam tunnel and the primary Leontaintnent vessel through special pipe chases.
6.2.33.1.1 Design Bases 6.23.3,1.23 Main Steam Tunnel 1 i me af 6. 7. 3 3.\ h The design of the secondary containment compartments with rcspect to pressurization is The reactor building main steam tunnel is based upon the worst case DBA rupture of a high located betreen the primary containment vessel or moderate energy line postulated to occur in and the turbine building /The steam tunnD cach compartment (see Subsection 3.6.2 for (houses the higb~ energy and radi6 active main rupture details). The rupture producing the ! steam and feedwater lines along with some greatest blowdown mass arid enthalpy is selected Lportions of the RCIC, RHR and RWCU pipint>
~
for the analysis of each compartm t. For the INSFT G.2.'3.3.) 6.2 22 Amendment 9 INSERT G . 2. 13 .1
i 2 DOA poemma op 3 G.I(t)
ABWR msm.
Standard Plant nry e We9 The DBA for the steam tunnelis the double.
O ended break of one of the 28 in. main steam lines.
which turbine16 routedThe building. from the reactor steam vessel to the}
tunnel blowout J
l e IN 45 ST G.2.3*3.1 h -
I.
(panels event ofvent into the turbine the postulated DBA. building inJthe 6.23J.13 Design Evaluatloa 4
0l ewe. p; ;h ;;; ;;d !: ;!::: rf ;::
4 IwSEAT G. 7. 3. %. l h O
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Amendment 9 6.2 22a
. - . . ~ . . , , - ~
D B A Po fLT \0 $J OF f .
MM ~
23A6100AD Standard Plant p1T n i I blowwt Pods are und in pbC8 of *F
- vent pathways when the environmental conditions accomplished by isolation of lines or ducts that of one compartment must be isolated from the penetrate the containment vessel. Actuation of
{
covironment in another compartment. The panels the contalnment isolation system is are designed to open upon a differential pressure automatically initiated at specific limits
] gM psid.::d ::: ::.:::d :: h f:!!y :;m d defined for reactor plant operation. After the vfta 0.1 m. 1 -..,;L... i -. isolation function is initiated,it goes through
) .- to completion.
! The RELAP4 computer program is used to j [ calculate the mass and energy release 6.2.4.1 rates and Design Bases i the resultant compartment pressures and , )
temperatures. A detailed discussion of the 6.2.4.1.1 Safety Design Bases l
I(methodology can be foundand assumptions in Subsection used in the(1) 6.2.7 of Referene progt Containment isolation valves provide the 1 /adTbc annual conditions for tne analysis include necessary isolation of the containment in l
l the assumption of 102% rated reactor power and the event of accidents or other conditions l the compartment pressures, temperatures and and prevent the unfiltered release of i relative humidity to maximize the mass and energy containment contents thnt cannot be
] 5 release rates, perrnitted by 10CFR$0 or 10CFR100 limits.
.) Leaktightness of the valves shall be 6.2.3.4 Tests and inspections verified by Type C test.
I Testing and inspection of the integrity of (2) Capability for rapid closure or isolation
, secondary containment will be made as part of the of all pipes or ducts that penetrate the i
testing of the STGS (Subsection 6.5.1). containment is provided by means that
- provide a containment barrier in such pipes Status lights and alarms for door opening of or ducts sufficient to maintain leakage secondary containment will be tested periodically within permissible limits.
7 by their operation, with observation of lights
) and alarms. Leakege testing and inspection of (3) The design of isolation valving for lines all other architectural openings will be made as penetrating the containment follows the they are utilized periodically, requirements of General Design Criteria 54
' through 57 to the greatest extent
- 6.2.33 Instrumentation Requirements practicable consistent with safety and l reliability.
l By their nature, electrical penetrations of
' secondary containment do not have any (4) Isolation valves for instrument lines that I
instrumentation requirements. Piping and HVAC penetrate the drywell/ containment conforms l penetrations instrumentation requirements are to the requirements of Regulatory Guide discussed as part of each system's description in 1.11.
this SAR. Details of the initiating signals for isolation are given in Subsection 7.3.1.1.10. (5) Isolation valves, actuators and controls l are protected against loss of their safety Certain doors are fitted with status function from missiles and postulated indication lights. effects of high. and moderate energy line 6.2.4 Containment Isolation Systern (6) Design of the containment isolation valves
- The primary objective of the containment and associated piping and penetrations isolation system is to provide protection against meets the requirements for Seismic Category releases of radioactive materials to the envir. I components.
onment as a result of accidents occurring in the .
l systems inside the containment. The objective is (7) Containment isolation valves and associated r
Amendment 2 6.2 23 i
DBA POWr\ou OF '3 6.N h I MM Standard Plant 2W1rCAB arv. c l
6.2.7 References
- 1. WJ. Bilanta, The G.E. Mark 111 Pressure l Suppression Containment AnalyticalModel, l June 1974, (NEDO.20533).
{ 2. FJ. Moody, Maximum Discharge Rate of
) Liquid Vapor Mixtures from Vessels, General Electric Company, Report No. NEDO.21052, September,1975, i 3. W.J. Bilanin, The G.E. Mark !!! Pressure Suppression Containment Analytical Model, l
Supplemeat 1, September 1975 (NEDO 20$331).
- 4. fidaho National Engineering Laboratory]-
- RELAP4/ MOD 5..A Computer Program for 4 ,9 g g p, 3 g,q,7 Transient Thermal.Hydraulle Analysis of f Nuclear Reactors and Related Systems, User's j (Manual, September,1976 (NUREG.13351 d
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Ammendment 11 p E-I l
INSCFTS FOR DBA PORT)ou 0 F 3. G. l [l)
(),ailure,in f conjunction with a worst case single active component For this analysis, a worst case single active component failure OB is defined as the failure to close of an isolation valve which separates the reactor pressure vessel from the high or moderate energy pipe break in the secondary containment.
() Table 6.2-3tabulatesthefreevolumes, initial environment conditions and DBA break characteristics for the compartments which are analyzed. Table 6.2-4 enumerates the flow path and blowout panel characterjatics.
OD Inthe thesteam event of amixture
/ air postulated design into is directed basisadjoining high energy line break, compartments and is eventually purged into the ateam tur..el.
()ThedesignbasisbreakfortheRWCUsystemcompartmentnetworkis either an 8-in or 6-in double-ended break of the water supply line. This depends upon which break diameter produces the I N 5s9TS maximum pressurization in the break compartment and any adjoining FOR compartments. This high energy piping, which connects the RWCU 6 2.3'3*l equipment , originates at the reactor pressure vessel. After
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being routed through the RWCU system the high energy line is directed back to the reactor pressure vessel through special pipe chases and the steam tunnel. In the event of a postulated design basis high energy line break, the steam / air mixture is directed into adjoining compartments and eventually purged into the steam tunnel.
( ) These lines originate at the reactor pressure vessel and are routed through the main steam tunnel to the turbine building. In the event of a postulated design basis high energy line break, the preseurized steam / air mixture de held up dn the main steam tunnel and purged into the turbine building through blowout panels, as required.
()Theblowdownmassandenthalpyreleaseratesforthehighenergy line breaks are determined using Moody's homogeneous equilibrium i model. A discussion of the methodology and assumptions used in this model can be found in Reference 2. The resulting compartment pressures and temperatures are calculated by the engineering computer program SCAM. A detailed discussion of the y methodology and assumptions used in this program can be found in Reference 4.
E"Y J. P. Dougherty, SCAM - Subcompartment Analysis Method,' January G .2 +7 1977, (NEDE-21526).
ABM 2asioorn
, Standard Plant arv s 4
9.1.3 Fuel Pool Cooling and Cleanup The FPC system cools the fuel storage pool by Sptem transferring the spent fuel decay heat through two 6.55 x 106 Blu/br heat exchangers to the-9.1J.1 Design Bases reactor building closed cooling water system l (RCW). Each of the two heat exchangers is de.
4 signed to transfer one half the system design heat load. The system utilizes two parallel 250
- The fuel pool cooling and cleanup (FPC) system m3/hr pumps to provide a system design flow of l
shall be designed to remove the de.:sy heat from 500 m3/hr. Each pump is suitable for l
the fuel pool, maintain pool water level and continuous duty operation. The equipment is
- quality and remove radioactive materials from the located in the reactor building.
l pool to minimize the release of radioactivity to the environs. The system pool water temperature is main.
l tained at or below 125'F. The decay heat The FPC rystem shall: released from the stored fuel is transferred to the RCW.fThe residual beat removal system (RHR)
(1) minimize corrosion product buildup and shall fcan supplement the FPC system to remove the l
control water clarity, so that the fuel additional heat generated should the reactor be
- assemblies can be efficiently handled under. defueled beyond the design. basis 35% batch.
) water; l
Fuel storage pool water is circulated by (
(2) minimize fission product concentration in means of overflow through skimmers around the
~
the water which could be released from the periphery of the pool and a scupper at the end 1
pool to the reactor building environment; of the transfer pool. The overflow is collected in the fuel pool drain tanks and the flow passes (3) monitor fuel pool water level and maintain a through the heat exchangers and filter.deminera.
water level above the furt sufficient to liters and back to the pool through the A provide shielding for normal building occu. Ldiffusers. gg Clarity and purity of the pool water are 9'I' d (4) maintain the pool water temperature below maintained by a combination of filtering and 'ion M P.
125'F under normal operating condi. exchange. The filter.demineralizers maintain 4
tions.
is set to The temperature establish an acceptablelimit of 125'F environ. with pHtotal rangecorrosion of 5.6 to 8.6 product at 250F formetals j.
compat. at 30 ppb 4 ment for personnel working in the vicinity ibility with fuel storage racks and other equip.
- of the fuel pool. The design basl4 normal ments. Conductivity is maintained at less than.
heat load from spent fuel stored in i se pool 1.2 pS/cm'at 25'C and chlorides less than' :
ii'
- is the sum of decay beat of the mos' recent .20 ppb. Each filter unit in the filter. demi.
s 35% baah plu the heat fron; the pr; v;out 4 neraliset subsystem hart.dcquste capac!ty to -
fuel batches The RHR system will be used . maintain the desired purity level of the pools to supplem t the FPC system u der the under normal operating conditions. The flow -
maximum ond condition as d ined in rate is designed.to be approximately that on 9.13.2. required for two complete water changes per day-(cun menoS gsegey p t
9 sug,f, g for the fuel transfer and storage pools. The-maximum system flow rate istwice that needed to
' acriptionO-9.1.3.2 SystempMt
- 9.1-il malatain the specified water quality,
. 9. l.3 c f (57 The PC system (Figures 9.1 a and b, and 9.10 maintains the spent fue storage pool' L The FPC system is designed to remove-
, below the desired temperature an acceptable suspended or dissolved impurities from the-radletion. level-and at a d ree of clarity _ l following sources:
necessary to transfer and i tvice the fuel' .
t bundles - -l (1) dust or other airborne particles;-
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t i During refueling prior to 21 days following shutdown, the i reactor (shutdown cooling).and fuel pool cooling are j provided jointly by the Residual Heat Removal (RHR) and FPC i systems in parallel. The reactor cavity communicates with
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the fuel pool since the reactor well. is flooded and the fuel
- j. gates are open. RHR suction is taken from the vessel i shutdown suction lines, pumped throu6h RHR heat exchangers i and discharged into the upper pools to improve water clarity.
for refueling. For the FPC system, fuel pool-water is circulated by means of overflow through skimmers around the periphery of the-pool and a scupper at the end.of the
- transfer pool . drain tanks, pumped through the FPC heat
- exchangers and filter demineralizers and back to the. pool j through the pool diffusers.
- After 21 days, the fuel gates are closed. At' this point FPC
- system provides solely the fuel pool cooling function. ,
However, when the reactor is-defueled more than the- 4
- design br. sis 35% batch -(maximum heat load condition), RHR-
- can provide supplemental cooling-to remove additional decay-
- heat. RHR' supplemental. cooling ~ suction is.taken from the-skimmer surge tank . passed through RHR
- heat exchanger and back to the-fuel pool.
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ABWR mum nyn Standard Plent (2) surface dirt dislodged from equipment control room and a local panel. Pump low suc.
( immersed in the pool; tion pressure automatically turns off the pumps. A pump low discharge pressure alarm is (3) crud and fission products emanating from the indicated in the control roem and on the local reactor or fuel bundles during refueling; panel. The circulating purnp motors can be powered from the diesel. generators if normal (4) debris from inspection or disposal opera. power is not available. Circulating pump motor tions; and loads are considered nonessentialloads and will be operated as required under accident (5) residual cleaning chemicals or flush water. condittons.
A post. strainer in the effluent stream of the The water level in the spent fuel storage filter.demineralizer limits the migration of poolis maintained at a belght which is suffi.
filter material. The filter holding element can cient to provide shielding for normal building
- withstand a differential pressure greater than occupancy. Radioactive particulates removed i the developed purnp head for the systern. from the fuel pool are collected in filter.de.
mineralizer units which are located in sbleided The filter.demineralizer units are located cells. For these reasons, the exposure of plant separately in shielded cells with enough clear. personoci to radiation from the FPC system is ance to permit removing filter elements from the minimal. Further details of radiological vessels, considerations for this system are described in Chapter 12.
Each cell contains only the filter.deminera.
lizer and piping. All valves (inlet, outlet. The circulation patterns within the reactor recycle, vent, drain, etc.) are located on the well and spent fuel storage pool are established outside of one shielding wall of the room, by placing the diffusers and skimrners so that together with necessary piping and headers, particles dislodged during refueling operations lOV instrument elements and controls. Penetrations are swept away from the work area and out of the through shielding walls are located so as not to pools, compromise radiation shielding requirements.
Check valves prevent the pool from siphoning The filter.demineralizers are controlled from in the event of a pipe rupture, a local panel. A differential pressure and conductivity instruments provided for each Heat from pool evaporation is handled by the filter.demineralizer unit indicate when backwash building ventilation system. Makeup water is is required. Suitable alarms, differential pressure indicators and flow indicators monitor provided through for +a6 remote.ooerated e 9.1.3.3 Safety Evaluation u UM of i the condition of the filter.demineralizers.
QJ(' p+eo^ c fclog 3 (81 The maximum possible heat loaT6the decay PC s,y)s System lostrumentation is provided for both automatic and remote manual operations. A low. heat of the full core load of fuel at the end of low level switch stops the circulating pumps when the fuel cycle plus the remaining decay heat of the fuel pool drain tank reserve capacity is the spent fuel discharged at previous refuel.
reduced to the volume that can be pumped in ing : the maximum capacity of the spent fuel l approximately one minute with one pump at rated sto se poolis 270% of a core. The temperature capacity (250 m3/hr). A level switch is the fuel pool water may be permitted to rise provided in the fuel pool to alarm on high stid to approximately 140'F under these condi. I low level. A temperature element is provided to tions. During cold shutdown conditions, if Xst oisplay pool temperature in the main control room, appears exceed thatthe 125'F, theoperator fuel pool can temperature connect the will l FPC system to the RHR system. Combining the ca.
The circulating putups are controlled from the pacities enables the two systems to keep the
- 'd L yow c k o W v-4. 0 9.14 Amendment 16
. (,C L AR l A C A*TWd
, ABM i m io w i mn Standard Plant i
water temperature below 125 F. The RHR system f Snseme C. og i A makeup watcr system and poo water level O will be used only to supplement the fuel pool cooling when the reactor is shut down. The instrumentatio n are provided to replace reactor will not be started up whenever portions evaporative and leakage losses. Makeup water 3 9
of the RHR systems are needed to cool the fuel during normal op: ration will be supplied from pool. The connecting piping from the fuel condensate. Thefuppression pool eleanup system storage pool to the RHR system is designed can be used as a[ source of makeup water in case a h'I' Seismic Cregory I and can be isolated, assuming of failure of the normal makeup water system 3m a single active failure, from the remainder of 9,g,3,3 the fuel pool system. Connections from the RHR system to the FPC system provide a Seismic Category I, These connections may also be utilized during safety related makeup capability to the spent emergency conditions to assure cooling of the fuel pool. The FPC system from the RHR 9,),3 spent fuel regardless of the availability of the connections to the spent fuel pool are seismic g) fuel pool cooling system. Tbc volume of water in Category 1, safety related.
the storage pool is such that there is enough 4 - IN SU IC 3. l 'I*
9.t . S g beat absorption capability to allow sufficient From the foregoing analysis, it is concluded time for switebing over to the RHR system for that the FPC system meets its design bases, h
g3gp emergency cooling.
---> 9.1.3A Inspection and Testing Requirements I'l M The 1400 F temperature limit is set to assure l hat the fuel building environment does not No special tests are required because,
@ t exceed equipment environmental limits, normally, one pump, one beat exchanger and one filter.demineralizer are operating while fuelis The spent fuel storage pool is designed so stored in the pool. The spare unit is operated that no single failure of structures or equipment periodically to bandle abnormal beat loads or to will cause inability to: (1) malatain irradiated replace a unit for servicing. Routine visual fuel submerged in water; (2) re establish normal inspection of the system components,instrumen-
,. fuel pool water level; or (3) remove decay beat tation and trouble alarms is adequate to verify from the pool, in order to limit the possibility system operability, of pool leakage around pool penetrations, the pool is lined with stainless steel. In addition 9.1.3.5 Radiological Considerations to providing a high degree of integrity, the lining is designed to withstand abuse that might Tbc water level in the spent fuel storage occur when equipment is moved about. No inlets, pool is maintained at a belght wb!ch is suffi-outlets or drains are provided that might permit cient to provide shielding for normal building the pool to be drained below a safe shielding occupancy. Radioactive particulates removed level. 1.ines extending i clow this level are from the fuel pool are collected in filter.
equipped with siphon breakers, check valves, or demineralizer units which are located in other suitable devices to prevent inadvertent sblelded cells. For these reasons, the exposure pool drainage. Interconnected drainage paths are of plant personnel to radiation from the FPC provided behind the line.t welds. These paths are system is minimal. Furtber detalls oI designed to: (1) prevent pressure buildup behind radiological considerations for this and other l the liner plate; (2) prevent the uncontrolled systems are described in Chapters 11,12, and loss of contaminated pool water to other rela- 15.
tively cleaner locations within the containment or fuel bandling area; and (3) provide liner leak .
detection and measurement. These drainage paths are designed to permit free gravity drainage or pumping to the equipment drain tank.
Amendment 16 9.15
INSERT 9.1.3.3 @ kg v s
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During the initial stages of refueling, the reactor cavity communicates with the fuel pool since the reactor well is flooded and the fuci pool gates are open. Decay heat removal is provided jointly by PJ1R and FPC systems and che pool temperature kept below 1400F. Evaluation studies concluded that after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay following shutdown (fuel pool gates open), the combined decay heat removal capacity of the 1-RHR and 1.FPC heat exchangers (single active failure 0 postulated) can keep the pool temperature well below 140 F.
The RHR FPC joint decay heat removal performance evaluation is shown in Table 9.1 12.
INSERT-9.1.3.3 h 9 l* }
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... utilizing either one of the two Seismic Category I makeup lines: (1) the makeup line to the spent fuel pool storage (2) or the makeup line to the dryer / separator storage pool. Makeup water from the dryer / separator fool is delivered to the spent fuel pool by opening pool gates to the spent fuel pool.
Both FPC and SPCU systems are Seismic Cr.tegory I Quality Croup C design with the exception of the filter demineralizer portion which is shared by both systems.
Following an accident or seismic event, the filter demineralizers are isolated from FPCS cooling portion and the SPCU by two block valves in series at both the inlet and outlet of the common filter demineraliser portion. Seismic Category I Quality Croup C bypass lines are provided'on both FPC and SPCU systems to allow continued flow of cooling and makeup water to the spent fuel pool.
Cf . l , '.
INSERT 9.1.3.3 @ -
P ,
-Furthermore, firehoses can be used as an alternate makeup source. The fire protection standpipes in the reactor building and their water supply (yard main,-one motor driven pump and water source) are seismically' designed. The motor' ;
driven pump is powered.from a bus which has.a safety-related. ;
diesel generator as one of its power sources. A second i seismically designed pump, directly driven by a diesel- ,
engine is also provided. I x
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4 9. l . 5 Table 9. I- ll (5)
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FUEL POOL COOLING HEAT EXCHANGER AND PERFORMANCE DATA 4
) Number of units 2 Seismic Category I design and l analysis l l
Types of exchangers Horizontal U Tube /Shell Maximum primary / secondary 16.0 kg/cm2 g/14.0 kg/cm2g side pressure Design Condition Normal heat load operating mode Primary side (tube side) performance data:
(1) Flow 250 m3 /h (2) Inlet temperature 52 0 maximum (3) Allowable pressure drop 0.7 kg/cm2 ,,x, (4) Exchanged heat 1.65 x 106 kcal/h Secondary side (shell side) performance data:
1 (1) Flow 280 m3 A (2) Inlet temperature 35 C maximum (3) Allowable pressure drop 0.7 kg/cm2 max.
(4) Type of cooling water RCW water m,
i 9.1-22.4
-f}
N/ l 9 5 l TABLE S.I-Ig RHR FPC JOINT DECAY HEAT REMOVAL PERFORMANCE TABLE (150 HOURS FOLLOWING SHUTDOWN)
RilR FPC Cooling Maximum llent Pool Temp. Maximum Cooling Time Pool Temp. To Max. Temp.
Loops Load @ time - 0 @ time - 0 Combination to - 150 hrs to - 150 hrs From t - 0 2 RilR HX's
+ 11 x 10 6 125 F MF t-0 2 FPC IIX's kcal/hr 2 RHP.IlX's
+ 11 x 10 6 125 F 125"F t-O 1 FPC HX kcal/hr q 1 RHR llX
+ 11 x 10 6 125 F 1290 p - 8 hrs Q' 2 PFC l!X's kcal/hr 1 RHR llX
+ 11 x 10 6 125 F 1360F ~ 12 hT8 1 FPC HX kcal/hr I i
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9.1 - 7.2 A
4 l-
! ABM isisio w nrv n Standard Plant normally closed to flow can be tested to ensure .
l j d operability and integrity of the system.
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1 Flow te the various systems is balanced by
- means of manual valves at the individual takeoff points.
$ 9.2.11 Reactor Building CoolingWater l System l 9.2.11.1 Design Bases
! 9.2.11.1.1 Safety Design Bases i !
(1) The reactor buildlag cooling water (RCW)-
system shall be designed to remove heat from plant auxiliaries which are required for .
4 safe reactor shutdown, as well as those l auxillaries whose operation is desired
[. following a LOCA, but not essential to safe shutdown.
The heat removal capacity is based on the
- heat removal requirement during LOCA with the maximum ultimate heat sink temperature,
- 950F, As shown in Table 9.2 4, the heat removal requirement is higher during other plant operation modes, such as shutdown at 4
. bours. However, the RCW system ise designed to remove this larger amount of l beat ::E:: :i: - U:;;.:: i..; .' ' :. . ;'.. ko mee4 Se rege r4 m 4 M J'\^
- r e r : ::x; n .::. 5 4 u M * ^ 5.A."T.t.t.7 .
l - (2) The RCW system shall be designed to perform 9.k ll ( S)
- its required cooling functions following a
- LOCA, assuming a single active or passive failure.
(3) The safety related portions and valves isolating the.nonsafety related portions of i
O ;
Amendment 26 :9.23,1
!.- ; y L-
ABM 2s m o w i Standard Plant REV B but not to a temperature that would damage Q
C/
System cotoponents and piping materials are selected where required to be compatible with the equipment or require an immediate shutdown.
available site cooling water in order to minimize 9.2.11.4 Testing and Inspection Requirements corrosion. Cathodic protection of the tubir.g side of the heat exchanger shall be provided. The RCW system is designed to permit periodic Adequate corrosion safety factors are used to in. service intpection of all system components assure the integrity of the system during the to assure the integrity and capability of the life of the plant. system.
During all plant operating enodes, all The RCW system is designed for periodic pres-divisions have at least one RCW cooling water sure and functional testing to assure: (1) the pump operating. Therefore,if a LOCA occurs, the structural and leaktight integrity by visible RCW cooling water system required to shut down inspection of the components; (2) the the ;' ant safely is already in operation. If a operability and the performance of the active loss of offsite power occurs during a LOCA, the components of the system; and (3) the pumps momentarily stop until transfer to standby operability of the system as a whole, diesel generator power is completed. The pumps are restarted automatically according to the Tbc tests shall assure, under conditions as diesel loading sequence. If a LOCA occurs, most close to design as practical, the performance of nonsafety.related components are automatically the full operational sequence that brings the isolated from the RCW system. Consequently, no system into operation for reactor shutdown and operator action is required, following a LOCA, to for LOCA, including operating of applicable start the P.CW system in its LOCA operating mode, portions of the Reactor Protection System and the transfer between normal and standby owe All beat exchangers and pumps will be required sources. 4 - t N suT 9 2, l[,4 g,2,1)(2) during the following plant operating conditions, in addition to LOCA: shutdown at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Tbc RCW system is supplied with a chemical C shutdown at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> and hot standby with loss of addition tank to add chemicals to each AC power, division. The RCW system is initially filled with deminera lized water. A corrosion Loss of one RCW division will result in loss inhibitor can be added if desired. These of RCW cooling to every other RIP (five total) as measures are adequate to protect the RCW system shown on RRS P&lD (Figure 5.4 4) and will cause from the ill effects of corrosion or organic those five RIPS to runback to minimum speed. The fouling.
RIP M G set in the same electrical division, I which is cooled by the same RCW division which The RCW system is designed to conform with i failed and powers two more RIPS, would stop by the foregoing requirements. Initial tests shall l M.G set cooling water prptection. This would be made as described in Subsection 14.2.12. l completely shutdown three RIPS and would have the i resulting total of seven RIPS either at minimum 9.2.11.5 Instrumentation and Control speed or stopped. Assuming the event began at Requirer.1 cots full power on the 100% Control Rod Line, the resulting temporary reactor power would be All equipment is provided with either globe i approximately 60% power. The operator would then or butterfly valves to give the capability for correct the RCW problem or initiate a normal manual control. These valves are accessible plant shutdown. downstream of the equipment for regulation of flow through the equipment or for balancing the The drywell cooling system can perform its circuits. The isolation valves to the nonessen-function after the loss of any RCW division, tial RCW system are automatically and -
With only one RCW division and one drywell cooler remote manually operated, operating, the drywell temperature will increase (Q)
Amendment 14 9.2-6
O t w % CL T 9 . '2 . t t . 4 9 . 2. I t ( 2 ))!
These tests shall include periodic testing of the heat removal capability of each RCW heat exchanger. Each of-these heat exchangers has been designed to provide 20%
margin above the heat removal capability required for LOCA in Tables 9.2-4 a, b and c. The revised heat removal capac-ity of the heat exchangers is shown in Table 9.2-4d. This 20% margin is provided to compensate for the combined of-fects of fouling and tube plugging. When this margin is no longer present, the heat exchanger heat removal capacity will be increased by either cleaning or retubing.
O 4
3
ABWR msmi Standard Plant _ uv s inhibited demineralized water through the shell systems are preoperationally tested is side of two of the three 50% capacity TCW heat accordance with the requirements of Chapter 14.
dp
i exchangers in service. The heat from the TCW system is rejee.ted to the turbine service water The components of the TCW system and system which circulates water on the tube side of associated instrumentation are accessib!c during
/ the TCW system heat exchangers, plant operation for visual esamination.
Periodic inspections during normal operation are The standby TCW sptem purup b autornatically made to ensure operability and lategrity of the started on detection of low TCW system pump system. Inspections include measurements of discharge pressure. The standby TCW syste u heat cooling water flows, temperatures, pressures, exchanger is placed in service muually, water quality, corrosion. erosion rate, controf positions, and set points to verify the system The cooling water flow rate to the condition.
electro. hydraulic control (EHC) coolers, the turbine lube oil coolers and aftercoolers, and 9.2.14J Instrumentation Appliestion generator exciter air cooler is regulated by centrol valves. Control valves in the coaling Pressure and temperature indicators are water outlet from these units are throttled in provided where required for testing and response to temperature signals from the fluid balancing the system. Flow indicator taps are e being cooled, provided at strategic points in the system for initial balancing of the flows and verifying -
The flow rate of cooling water to all of the flows during plant operation.
other coolers is manually regulated by individual throttling valves located on the cooling water Surge tank high and low level and TCW pump outlet from each unit. discharge pressure alarms are retransmitted to the main control room from the TCW local control The minimum system coolingwater temperature panels.
.. is maintained by adjusting the TCW system heat exchanger bypass valve. Makeup ilow to the TCW system surge tank is
(* Initiated automatically by low surge tank water The surge tank provides a reservoir for level and is continued until the normal levei is small amounts of leakage from the system and for reestablished, the expansion and contraction of the cooling fluid with changes in the system temperature and Provisions for taking TCW system water is cannected to the pump suction. samples are included.
Demineralized makeup water to the TCW system *.
is controlled automatically by a level control 9.2.15 Reactor Service Water System valve which is actuated by sensing surge tank-level. A corrosion inhibiter 6 manually added 9.2.15.1 Design Bases to the system.
9.2.15.1.1 Safety Design Bases 9.2.1.f.3 Safety Evaluation (1) The reactor service water (RSW) system' The TCW system has no safety design bases shall be designeAto remove heat fror!!'~~~d and serves no safety function, thegeactor coolln'g water system which*~~P is (e' quired for safe reactor :hutdown, 9.2.14.4 Tests and inspections and which also cools those aiMilaries '
whose operation is desired following a All major components are tested and LOCA, but not essential to safe inspected as separate components prior to shutdown.
fnstallation, and as an integrated systr.m after .
Installation to ensure des lgn performan:e. The (2) The R$W system shall be' designed to fh.
- ( ,I )
- ' in three cQtvtst'onS h 4. 4hr4te SiVISichS 04 b .
(cL-AtsM LcA% od)
ABWR 2miooai Standard Plant avn Seismic Category I and ASME Code, (1) flooding, spraying or steam release due Section lit, Class 3 Quality Assurance to pipe rupture or equipment failure; O B, Quality Group C, IEEE.279 and IEEE.308 requirements. (2) pipe whip and jet forces resulting from postulated pipe rupture of nearby high (3) The R$W system shall be protected from coergy pipes; flooding, spraying, steam impingement, pipe whip, jet forces, missiles, fire (3 missiles which result from equipment and the effect of f ailure of any failure; and non Seismic Category I equipment, a required. (4) fire.
(4) The RSW system shall be designed to meet Liquid radiation monitors are provided in the the foregoing design bases during a loss RCW system. Upon detection of radiation leakage of preferred power, in a division of the RCW system, that system is isolated by operator action from the control 9.2.15.1.2 Power Generation Design !!ases room, and the cooling load is met by another division of the RCW system. Consequently, The RSW system shall be designed to cool the radioactive contamination released by the R$W reactor building cooling water (RCW) as required system to the environment does not exceed during: (a) normal operation; (b) emergency allowable limits defined by 10CFR100, shutdown; (c) normal shutdown; and (d) testing.
System low point drains and high point vents 9.2.15.2 System Description are provided as required.
The RSW system provides cooling water durit.g System components and piping materials are various operating modes, during shutdown and selected to be compatible with the available post.LOCA operations. The system removes heat site cooling water in order to minimize O from the RCW system and transfers it to the corrosion. Adequate corrosion safety factors ultimate heat sink. Figure 9.2 7 shows the RSW are used to assure the integrity of the system system diagram. CMoM depb+d duting the life of the plant.
9.2.ll u p 4A m wMe 9. 2-W .
W The RSW system is able to function during During #11 plant operating modes each abnormally high or low water levels and steps are division shall have at least one service water taken to prevent organic fouling that inay degrade pump operating. Therefore,if a LOCA occurs, system performance. These steps include trash the system is already in operation. If a loss racks and provisions for biocide treatment (where of offsite power occurs during a LOCA, the pumps discharge is allowed). Where discharge of momentarily stop until transfer to standby blocide is not allowed, non blocide treatment diesel generator power is completed. The pumps will be provided. Thermal backwashing capability are restarted automatically according to the will be provided at sea water sites where dieselloading sequence. No operator action is infestations of macrobial growth can occur. required, following a LOCA, to start the RSW system in its LOCA operating mode.
9.2.15.3 Safety Evaluation 9.2.15.4 Testing and Inspection Requirements The components of the RSW system are separated and protected to the extent necessary The RSW system is designed for periodic to assure that sufficient equipment remains pressure sud functional testing to assure:
operating to permit shutdown of the unit in the event of any of the following (Separation is (1) the structural and lenktight integrity applied to electrical equipment and by visible inspection of the components; instrumentation and controls as weit as to mechanical equipment and piping.):
9 Amendment tt 9.2 12
ABWR mamui Prv n Standard Plant 9.2.17 Interfaces l'~)
Q 9.2.17.1 Ultimate fleat $1nk Capability Interface requirements pertaining to ultimate heat sink capability are delineated in Subsection 9.2.5 as follows:
Subnetion 11 tit 92.5.1 Safety Desip Bases 9.2.5.2 Power Generation Desip Bases 5
92.5.6 Evaluation of UHS Performance 9.2.5.7 Safety Evaluation 9.2.5.8 Conformance to Regulatory Guide 1.27 9.2.5.9 Instrumentation and Alarms 0.2.5.10 Tests and Inspections 9.2.17.2 Makeup Water System Capability Tbc raw water treatment and preparation of
(~ the demineralized water is sent to the makeup
(
water system (purified) described in Subsection 9.2.10. , g g g p,7 g , ,2 .\ 7 2. 9- 0N The makeup water preparation system shall be located in a building which does not contain any saf ety re'ated structures, systems or components. If the system is not available, E demineralized water can be obtained from mobile 5 equipment. The system shall be desiped so that any failure in the system, including any that cause floodin6, shall not result in the failure of any snicty related structure, system or component.
9.2.17.3 Potable and Sanitary Water System The potable and sanitary water system shall be desiped with no interconnections with systems E having the potential for containing radioactive
$ materials. Protection shall be provided through the use of air gaps, where necessary. (See Subsection 9.2.4),
i gm 5
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j 9.2 13 Amcodment 16 l
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i t
I 9.2,M.1 (9. 2,t o ( 4)
I N sstt:T I.
- i
- 1. The. demineralized water. preparation system shall consist of l
{ at least two divisions-capable of producing at least 200 gpm (
t of domineralized water each. Storage of domineralized water i shall be at least 200,000 gallons. If additional demineral-
) ized water is needed during peak usage periods, rented por- -
table demineralizers-shell be used as required. .,
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ABM 9 2 9N us iooxii Standard Plant _ mv n s TABLE 9.2 3 CAPACITY REQUIREMENTS FOR CONDENSATE STORAGE TANK .
Function Canacity RecuIred dead space top of pool 7,900g
{ (Note 1) nortnal operation variation 264,000g and receiving volume for plant startup return water s
minimum storage volume 66,000g dead space middle of pool 34,320g (Note 1)
{ water source for 150,480g station blackout (Note 2) dead space bottom of pool 34,320g (Note 1)
Total 557,020g l
NOTE (1) These values are based on a bottom area of 1,400 ft,2 m rs,1, e n t,, ,., ,
r ,, ,. r ,, ,r , , n r n n n g ,, r r n - , fg . p cy c i e D,000 ga!!c?! l- sasser
-.cp-',ctonfe',g,igh*'c">!
ever .m 1, ,a ,
7n, ,;;u , u ue ,, g ., g ... ,, .g yew .,
to r, no e geglceggg vev.. r g g ,y p cyc ::,;;;
-wate-fr^:e 'h: : pp: :::= p::.'.
(2) \v o +er for op eecchon o F g.c. t c is ko ker h rom O 4. C Ce CNW 3 C&C Storm S Yo eb Gwd O C SuPP"'55\0*
pcok AS de $ C n bc.c3 s w -\-6 e P G s oE Appew dix 19 A.
S e..e n&A f *f
, Amen &nent 14 h 8 V" 6%f8 9.2 16
. - ~ - . - - . - . _ . - . . . - - . . . -
i 9. 7. 9 ( d .
l O
No to S S AM , Ta @ t 9. 2 - 3 har i 4 e x 4 c b a n 3 4.
louw m e d 8s.a.s1 m c tordswjl g.
Normal alignment for removal of decay heat is with the condensate storage. tank. Water for RCIC operation is taken from either the condensate storage' tank or the suppression '
l pool as described in~the EPGs of Appendix18A. The volume of water in these two sources, as required by paragraph ;
3.3.2 of Regulatory Guide 1.155, is sufficient to permit core cooling during' station blackout for a duration of eight hours. The-switchover from the, condensate storage tank- to the suppression pool (or the reverse) is performed using station de power and-is not dependent upon either offsite ac power systems or onsite emergency ac power systems.
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A.BWR 2suimui Standard Plant vn' s TABLE 9.24d i DESIGN CHARACTERISTICS FOR REACTOR v' BUILDING COOLING WATER SYSTEM COMPONENTS RCW Pumps (Tw o per Die mn)
RCW (AT /(B) RCW (C)
Discharge Flow Rate 5,720 gpm/ pump 4,840 gpm/ pump Pump Total Head 82 psig 75 psig Desip Pressure 200 psig 200 psig Desip Temperature 1580F 1580F RCW Heat Exchangert (Tb<* c P*< N W'"
RCW (AT /(B) RCW(C)
A5 6 A 2-Capacity (fov eo ch ex10 BTU /h Ebt106 BTU /h WJ e 4 denyv )
RCW Surge Tanks y, E
Capacity Equal to 30 days of normalleakage 3
l
'd Desip Pressure Static Head Desip Temperature 1580F RCW Chemical Addition Tanks Desip Pressure 200 psig Desip Temperature 1580F RCW Piping Desip Pressure 200 psig Desip Temperature 1580F
_l
~^
Amendment 16 9.2 19.1
3.2.ll(7) v Table 9.2- 13 Reactor Service Water System RSW Pumps (two per division)
Discharge Flow Rate 7,920 gpm Pump Total Head 50 psi Design Pressure 115 psi Design Temperature 122 F RSW Piping and Valves Design Pressure 115 psi Design Temperature 122 F i
l i
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l 3,2- 2 5 d
- ABWR isAsioou Standard Plant Rn A (1) Visual examination of all accessible surfaces (c) Deposits on stems and other valve parts IV i
of rotors which could interfere with valve operation l l
l (2) Yhual and surface examination of alllow. (d) Distortions,mkalignment pressure buckets .
Inspection of all valves of one type will bc l
- (3) 100 percent visual examination of couptiags conducted if any unusual condition is discovered l
- and coupling bolts l
10.2A Evaluation l
- The inservice inspection of valves important to overspeed protection indudes the following: The turbine. generator is not nucleat safety
. telated and is not needed to effect or support a safe j
(1) All main stop valves, cetrol valves, extrac. shutdown of the reactor.
tion nonreturn vc!vs, and CBIVs will be
! tested under load. Test controls installed on The turbine is designed, constructed, and in-j the main control room turbine panel and spected to minimize the possibility of any major
, permit full stroking of the stop valve, control component failure.
valves, and CBIVs. Valve position indication la provided on the panel. Noload reduction The turbine has a redundant, testable overspeed is necessary before testing main stop and trip system to minimize the possibility of a turbine control valves, and CBlVs. Extraction conte. overspeed event.
turn valves are tested by equalizing air i
pressure across the air cylinder. Movement Unrestrained stored energy in the extraction of the valve arm is observed upon action of steam system has been reduced to an acceptable the spring closure mechanism, minimum by the addition of nonreturn valves in
( tg (2) Main stop valves, control valves, extraction selected extraction lines.
, $ nonteturn valves, and CBlVs will be tested at The turbine generator equipment shielding re-
, 0 . 2. IeastA = :t E!y, ::d ::: ;L '
quirements and the methods of access control for all N -
- r - c W h & L:t: . .. .'..... a areas of the turbine building ensure that the dose d criteria specified in 10CFR20 for operating personnel
^' '**
- p testocewillperbemonth, verifiedclosm b of ry e tionch[A during of the--- are not exceeded.
valve motion.mm ,+op v dy,e ,co4n Lv o lv e. % .I c. e iv - - All areas in proximity to turbine generator Tightness tests of the main stop and control equipment are zoneJ according to expected valves are performed at least once per main- occupancy tiraes and radiation levels anticipated ienance cycle by checking the coastdown under normal operating conditions, characteristics of the turbine from no load with each set of four valves closed alternately. Specificatic,n of the various radiation zones in accordance with expected occupancy is listed in (3) All main stop valves, main control valves, and Chapter 11 l
- CBlVs will be inspected once during the first
' three refueling or extended maintenance if deemed necessary during unusual shutdowns. Subsequent inspections will be occurrences, the occupancy times for certain areas l scheduled so that each valve is inspected at 3 will be reduced by administrative controls enacted by to 5 year laterval and at least, one valve of health physics personnel, each type is inspected after each fuel cycle or 31/3 year interval, whichever is less. The The design basis operating concentrations of I inspections will be conducted for: N 16 in the turbine cycle are indicated in Section 12.2.
(a) Wear oflinkages and stem packings The connection between the low pressure turbine exhaust bood and the condenser is made by (b) Erosion of valve seats and stems - means of a stainless steel expansion joint, r e os e. b v a l v ' U w ol
- 3I M' ob s e rvq3bnce 4b e. v at d vweeosskon k h c.lo.sm M cMoe gd ,4 moves 3%e oW' + o a M b CA
- M *I '
aa kom. )
L .
ABWR 234sioors Standard Plant hA co ra (2)liigh condenser pressure turbine trip at 22 plant power operation, and to ingA the turbine b inches Hg vaccum m::..h :;:er exhaust st the3 beginning of each R*
start up. 53 s1avv- $
. (3) Bypass valve closure at 12 inches Hg vacuum 10.4.2.1 Design Bases (4) Main steam isolation valve closes at,7 to 10 inches Hg vacuum 10.4.2.1.1 Safety Design Basea Condenser pressure is an input to the reactor The MCES does not serve or support any safety recirculation system. Recirculation pump runback is function and has no safety design bases.
Initiated upon the trip of a circulating water pump i when condenser pressure is higher than some site 10.4.2.1.2 Power Generatlon Design Bases specific preset valve. Runback is automatically initiated when required to avoid a turbins trip on Power Generation Dette Bath One The MCES high condenser pressure, is designed to remove air and other power cycle non.
condensable gases from the condenser during plant 10.4.1.53 Temperature startup, cooldown, and power operation and ex)4sg M them to the offga:Tsystem or turbine bulT&ni gg Temperature is measured in each LP turbine r:1. :; tr cahausp53s4*m. $$
exhaust hood by pneumatic tempersture controllers.
The controllers modulate a control valve in the water Power Generation Deeln Bath Two . The MCES spray line protecting the exhaust hoods from over. establishes and maintains a vacuum in the condenser hewing. during power operation by the use of steam jet att ejectors, and b Circulating water temperatures are monitored early startTup, y the mechanical vacuum pump d upstream and downstream of each condenser tube bundle and are fed to the plant computer and a main 10.4.2.2 Description control room regorder for use during periodic condenser perforrNance evaluations. The condenser evacuation system is !!!ustrated in Figure 10.41. The system consists of two 1004ca. R p.
10.4.1.$ 4 leakage pacity, double stage, steam jet air ejector (SJAE) $6 units (complete with intercondenser) for power plant 4 Leakage of circulating water into the condenser operation, and a mechanical vacuum pump for use shellis monitored by the on.line instrumentation and during startup. The last stage of the SJAE is a the process sampling system described in Subsection noncondensing stage. One SJAE unit is;normally in MR 93.2. operation and the other is on standby. 66 i conductivity of the condensate is continuously During the initial phase of startup, when the i' monitored at selected locations in the condenser, desired rate of air and gas removal exceeds the l Conductivity and sodium are continuously mcnitored capacity of the steam jet air ejectors, and nuclear at the discht. , of the condensate pumps. High steam pressure is not adequate to operate the air I
condensate co.. uctivity and sodium content, which ejector units, the mechanical vacuum pump estab-indicate a condenser tube leak, are individually lishes a vacuum in the main condenser and other
~
alarmed in the main control room. part: of the power cycle. The discharge from y.A vacuum pump is then r pp to the turbine buil 10.4.2 Main Condenser Evacuation rer exhausSInte%ere is then RRlittle System =;:..:_
no emuent r'adioactivity r present. Radiation detecto 66 in theb 1.... . r. , f.' ..p and plant vent Noncondensable gases are removed from the ralarm in the main control room if abnormal power cycle by the main condenser evacuation radioacthiry is detected (see Section 7.6). Radiation system (MCES). The MCES removes the hydrogen monitors are provided on the main steam lines which and oxygen produced by radiolysis of water in the trip the vacuum pump if abnormal radioactivity is reactor, and other power cycle noncondensable detected in the steam.being supplied to the gases, and exhausts them to the offgas system during condenser.
A,.%e %,sM3 e o w?CW4**4 Amendment 11 ew b 4sd Sg M M - 10A4 l
L
ABWR 2m m Standard Plant a., A The steam jet air ejecto aced in sersice to formed prior to plant operation in accordan:e with remove the gases from the main condenser after a applicable codes and standards, pressure of ebout 10 to 15 in Hg absolute is estab.
lished in the main condenser by the mechanical Components of the system are continuously men.
vacuum pump an'd when sufficient nuclear steam Itored during operation to casure satisfactory perfor.
pressure is available, py mance. Periodic inservice tests and inspections of
( the evacuation system are performed in conjunction j During normal power operation thy steam Jet with the scheduled maintenance outages.
air injectors are normally driven byp;.d;;..;d-steam, with the vnaln steam supply on automatic 10.4.23 Instrumentation Applications yg standby. The main steam supply, boguever,is
$$ normally used during startup and low load operation, l.,ocal and remote indicating devices for puch and auxiliary steam is available for normal use of the parameters as pressure, temperature, and flow steam jet air ejectors during early startup, should the indicators are provided as required for monitoring mechanical vacuum pump prove to be unavailable. the system operation.
10.4.2J Evaluation 10.4.2.5.1 Steam Jet Air Ejectors The offges from the main condenser is one Steam pressure and flow is continuously moni.
l source of radioactive gas in the station. Normally it tored and controlled in the ejector s{ cam supply includes the activation gases nitrogen 16, oxygen 19, lines. Redundant pressure controllergsense steam and nitrogen.13, plus the radioactive noble gas pressure at the second stage inlet and modulate the parents of strontium.89, strontium.90, and stearn supply control valves upstream of the air cesium 137. An inventory of radioactive contami. ejectors. The steam flow transmitters provide inputs pg.,
nants in the effluent from the steam jet air ejectors is evaluated in Section 11.3.
to logic devices. These logic devigs provide for $$
Isolating the offgas flow from the air mjector unit on O a two.out of three logic, should the steam flow drop G Steam supply Io the second stage jeetor is maintained at a minimum specified flow to ensure adequate dilution of hydrogen and prevent the offgas below acceptable limits for offgas stream dilution.
10.4.2.5.2 Mechanical Vacuum Pump l
from reaching the flammable limit of hydrogen.
Pressure is measured on the suction line of the j lh mechanical vacuum pump by a pressure switch.
Upon reaching a preset vacuum, the pressure switch The MCES has no safety.related function as energizes a solenoid valve which allows additional discussed in Section 3.2. Failure of the system will seal water to be pumped to the vacuum pumpf Seal not compromise any safety related system or compo. pump discharge pressure is locally monitored. Seal nent and will not prevent safe reactor shutdown. water cooler discharge temperature is measured by a temperature indicating switch. On high temperature, Should the system fall completely, a gradual the switch activates an annuciator in the main control reduction in condenser vacuum would result from room. The vacuum pumpJ::h:g; 2:n- E the buildup of noncondensable gases. This reduction 0.; ...L'.., RR in vacuum would first cause a lowering of turbine %:Sud P " o r.;si:t :nf: + =S:
The#@2 vacuum. .:pump is N6 cycle efficiency due to the increase in turbine tripped and its discharge valve is closed upon exhausts' pressure. If the MCES remained recching a main steam high. high radiation signal, pR inoperable, condenser pressure would then reach the 66 turbine trip set point and a turbine trip would result. 10A.3 Turbine Gland Seal System The loss of condenser vacuum incident is discussed in Subsection 15.2.5. The turbine gland seal system (TGSS) prevents }
the escape of radioactive steam from the turbine 10.4.2.4 Tests and inspections shaft / casing penetrations and valve stems and prevents air inleakage through subatmospheric l
Testing and inspection of the system is per- turbine glands, chs bo 9 fEMust co m p<=r+ve shew J s ts#*m AScI dob bM M ah moJom o F-Ae ck'rovd*3 gut p Pgm Am.um noi t i s he io e % .)-beie r-d so,s e. -b +b. m o M;p * ' * V" l fn A ewd % cdmo s p% n 1
ABWR nusoon ;
Standard Plant Rev A l 10.43.1 Design Bases . supplied from the main steam line or auxiliary steam Q header. Above approximately 50% load, however, V 10.43.1.1 Safety Design Bases sealing steam is normally provided from the heater drain tankvent header. At allloads, gland sealing i.!
I I The TGSS does not serve or support any safety can be achieved using auxiliary steam 60 that plant function and bas bo safery design bases. power operation can be maintained without appreclable rr tioactivity releases even if highly 10.43.1.2 Power Generation Design Bases abnormallevels if radioactive contaminants are present in the process steam, due to unanticipated Power Generation Desien Bash One The TGSS is fuel failure in the reactor, designed to prevent atmospheric air leakage into the !
turbine casings and to prevent radioactive steam The outer portion of all glands of the turbine and leakage out of the casings of the turbine generator, main steam valves are connected to the 6 1and steam condenser which is mJntained at a slight vacuum by Power Generat!on Detten Bath Two . The TGSS the exhauster blower. During plant operation, the returns the condensed steam to the condenser and gland steam condenser and one of the two installed g exhausts the noncondensable gases, via the turbine 100% capacity motor. driven blowers are in Rg g
' buildinggonsilah system, to the plant vent. operation. The exhauster blower g $ $'
ce 9 e b e/te.* % s+ A 'S&;;: to th turbine buildingsn..;..;a. m !
Power Generation Denien Bath Three . The TGSS sye++m exhaus elland steam condenser is ( ;
has enough capacity to handle stearn and air flows cooled by maln condege fylg g4 j
resuhing from twice the normal packing clearances, p,.g ,.Mo -
10.433 Evaluation dst chegeel 10.43.2 Description The turbine gland st.al system is designed to lyggg-10.43.2.1 General Description
prevent leakage of radioactive steam from the main turbine shaft glands and the valve stems. The The turbine gland sealing system is illustrated in high. pressure turbine shaft seals must accommodate
[ l Figure 10.4 2. The turbine gland seal system consists a range of turbine shell pressure from full vacuum to of a sealing steam pressure regulator, sealing steam approximately 220 psia. The low. pressure turbine header, a gland steam condenser, with two full. shaft seals operate against a vacuum at all times.
capacity exhauster blowers, and the associated The gland seal outer portion steam air mixture is piping. valves and instrumentation. exhausted to the gland steam condenser via the seal vent annulus (i.e., end glands) which is maintained at 10.43.2.2 System Operation a slight vacuum. The radioactive content of the sealing steam which eventually exhausts to the plant The annular space through which the turbine vent and the atmosphere is evaluated in Section 113 shaft penetrates the casing is scaled by stearr and makes a negligible contribution to overall plant lyggg supplied to the shaft seals. Where the gland seats
- radiation relesse. In addition, the auxiliary steam .
operate against positive pressure, the sealing steam system is designed to provide a 100% backup to the acts as a buffer and flows either inwards for collec- normal gland seal process steam supply. A full ca-tion at an intermediate lenkoff point, or, outwards pacity gland steam condenser is provided, and and into the vent annulus. Where the gland seals equipped with two 100% espacity blours. 1 l
operate against vacuum, the sealing steam either is drawn into the casing or leaks outward to a vent Relief valves on the seal steam header prevent i annulus. At all gland seals, the vent annulus is excessive seal steam pressure. The valves discharge- l maintained at a slight vacuum and also receives air in to the condenser shell. !
Icakage from the outside From each vent annulus, .
j the air steam mixture is drawn to the gland steam 10.43.4 Tests and laspections !
condenser. i Testing and inspection of the system will be per.
The seal steam header pressure is regulated formed prior to plant operation. Components of the - i Rg, automatically by a pressure controller. During . system are continuously monitored during operation
$$ startup and low load operation, the seal steam is k
U Amendment 11 10 M f
i
ABWR 23A6t0W Standard Plant an2 to ensure that they are functioning satisfactorUy. Power Generation Dnta Bath Two . The TBS is O Periodic tests and inspections may be performed in conjunction with maintenance outages, designed to bypass steam to the main condenser -
during plant startup and to permit a normal manual cooldown of the reactor coolant system from a hot 10AJJ lastrumentation Application shutdown ceindition to a point consistent with initia, tion of residual heat removal system operation.
10AJJ.1 Gland Steam Coedenser Exhausters Power Generation Deslan Basis Three The TBS is 10A.33.1.1 Pnssure designed, in conjunction with the reactor systems, to provide for a 40 percent electrical step load reduc.
Gland steam condenser ev.hauster suction pres, tion without reactor trip. The systems will also allow sure is continuously monitored and reported to the a turbine trip but without lifting the main steam main control room and plant computer. A low rebet and safety valves.
vacuum signal actuates a main contro! room annun.
clator. 10AA.2 Description 10AJJ.1.2 level $4 Neb 10AA.2.1 General Description Water levels in the gland steam con enser drain The TBS is shown in Fi6ure 10.31, Main Steam leg are monitored and makeup is added required System. The TBS consists of a three valve chest that to maintain loop sealintegrity. Abnorma are is connected to the main steam lines upstream of the.
annunciated in the main control room, turbine stop valves, and of three dump lines that connect separately each regulating valve outlet to one -
10A33.2 Sealing Steam &ader condenser shell. The system is designed to bypass 33 percent of the rated main steam flow directly to the
' ealing steam header pressure is monitored and condenser. The system and its components are shown reported to the main control room and plant com. In Figur:s 10A-10 and 10.4-11.
O puter. Header steam temperature is also measured V and recorded. The turbine bypass system,in combination with the
~
IN%9.T 4 rf. actor systems, provides the capability to shed 40 percent of the turbine. generator rated load witif.,out ?
10 A.345.l.310.4.4 Turbine Bypass System
- cactor trip and without the operation of relief and uA.3- The turbine bypass system (TBS) provides the safety valves. A load rejection in excess of 40 percent (0 capbility to discharge main steam from the reactor is expected to result in reactor trip but without
" -- directly to the condenser to minimize step load. operation of any steam relief and safety valve, reduction transient effects on the reactor coolant _
system. The system is also used to aischarge main 10AA.2.2 Component Description steam during reactor hot standby and cooldown op-erations. One valve chest.is provided and houses three ,
individual bypass ulves. Each bypass valve is an 10AA.1 Design Bases angle body type valve operated by hydraulic fluid pressure with spring action to close. The valve chest 10.4A.1.1 Safety Design Bases ' assembly includes hydraulic supply and drain piping;
. three hydraulic accumulators, one for each bypass :
The TBS does not serve or support any safety valve; servo valves; fast acting servo valves; and, valve
. function and has no safety design basis, positien transmitters.
10AA.1.2 Power Generation Design Bases The turbine bypass valves are provide'd with a separate hydraulle fluid power unit.- The unit Power Generation Desien Basis One The TBS has . inctndes high. pressure fluid pumps, alters, and heat the capacity to bypast 33 percent of the rated main exchangers; High pressure hydraulic fluid is -
steam flow to the main condenser. provided at%c bottom valve nLtuator and drained O
V 10 4-7 Amendment to
l l N SBS-T 1 0 4 .'i . 5 . ). 3 %%. S 10.4.3.5.1.3 Effluent Monitoring The TGSS effluents are first monitored by a system dedicated continuous radiation monitor installed on;the gland:-steam.
condenser,exhauster blower discharge. High monitor readings-are t
alarmed in-the main control room. -'The system effluents are then - 3 discharged to the turbine building =. compartment exhaust' system and the-plant vent stack where further effluent radiation monitoring.
is performed. See=SubsectionLie.4.10.1 %or-. interface. requirements-pertaining to the radiologicalianalysis of-the TGSS effluents.:
O 6
a
=
O f 4
t i ) ,1~.f h-
____._.____.__._____._1__._____ _ _ _ _ _ = _m_-_ -- _ - - _ _ _ - _ - _ . - _ - - - - - - - . - - --- -
ABWR 23 mow Standard Plant _-- nm a suction side of the drain pump. This switch will tion valves are interlocked with the circulating water automatically stop the pump in the event oflow - pumps so that when a pump is started,its discharge
[3 L/
water levelin the standpipe to protect the pump from excessive cavitation, valve will be opening while the pump is coming up to speed, thus assuring there is water flow through the pump. When the pump is stopped, the discharge 10.43.3 Evaluation valve closes autornatically to prevent or minimize
- backward rotation of the pump and motor.
The CWS is not a safey.related systeml however, a flooding analysis of the turbine building is- Level switches monitor water levelin the con-performed on the CWS postulating a complete denser discharge water boxes and provide a permis.
rupture of a single expansion joint The analysis sive for starting the circulating water pumps.These assumes that the flow into the condenser pit comes level switches ensure that the supply piping and the from both the upstream and downstream side of the condenser are full of water prior to circulating water break and, for conservatism, it assumes that one piunp startup thus preventing water pressure surges system isolation valve does not fully close, from damaging the supply piping or the condenser.
Based on the above conservative assumptions, To satisfy the bearing lubricating water and shaft the CWS and related facilities are designed such that sealing water interlocks during startup, the circulat-the selected combination of plant physical arrange- ing water pump bearing lubricating and shaft seat ment and system protective features ensures that all flow switches, located in the lubricating seal water eredible potential circulating water spills inside the supply lines, taust sense a minimum flow to provide turbine building remain confined inside the con- pump start permissive.
denser pit. Further, plant safety is ensured in case of multiple CWS failures or other negligible probability Monitoring the performance of the circulating CWS related events by the plant safety related gen- water system is acccmplished by t'ifferential pressure eral flooding protection provisions that are discussed transducers across each half of the condenser with in Section 3.4. remote differential pressure indicators located in the ;
A main control room. Thermal element signals from
- the supply and discharge sides of the condenser are
() 10.4.5.4 Tests and Inspections transmitted to the plant computer for recording, The CWS and related systems and facilities are display and conden<cr performance calculations, tested ar.d checked for leakage integrity prior to initial plant startup and, as may be appropriate, To prevent icing and freeze up when the ambient following major maintenance and inspection. temperature of the ultimate heat ' sink falls below.
32 F, warm water from the discharge side of the..
All active and selected passive components of condenser is recirculated back to the screen house the circulating water system are accessible for intake. Thermal elements, located in each condenser inspection and maintenance / testing during normal supply line and monitored in the main control room,-
power station operation. are utilized in throttling the warm water recirculation valve, which maintains the minimum-inlet tempera.
10.4.L5 Instrumentation Applications ture of approximately40 F.
Temperature monitors are provided upstream W '
10.4.6 . Condensate Cleanup Systern >
and downstream of each condenser shell section.
The condensate cleanup systemMCS) purifies L Indication is provided in'the control room to. and treats the conuensate as required to maintain - 'i H
identify open and closid positions of motor operated reractor feedwater purity,'using filtration to remove butterDy valves in the CWS piping. . corrosion products, ion exchange to remove con- _
denser leakage and other impurities, and water .
All major circulating water system valves which treatment additions to minimize corrosion /cro'sion .
control the flow path can be operated by local product releases in the power cycle. -
controls or by remote manual switches located on q the main control board. The pump discharge isola; ) N
- 10. 4.5 6 F Io o cl- P r o4 e cM o n [
- l. .Arnendment 3 _
g g; gb% s4 3 0 , T '3 ( W ' . _h"* 1oA10 -
Jk Co'C W 8 f L f >
- o n n% o heq J a ~ cN
ABMTt nui0w Standard Plant nm x F through the auxiliary condens,ers (off. gas recombiner
{s $
s condenser / coolers, gland steam condenser, and steam jet air ejector condensers) and maintains condensate pum; minimum flow. Measurements on pump suction and discharge pressures are 7tovided for all mmps i in the system.
The high pressure feedwater heater isolation valves are interlocked such that if a string of heaters wire to be removed fro:n senice the extraction non, return valves and/or isolation valves for those g beaters would automatically close and the beater
$ string bypass valve open. The low pressure feedwa.
ter Mater isolation valves are interlocked such that, if a string of featers were removed from senice, the extractions to the Wected heaters which are equiped with nonaturn valves would automatically close.
Sampling means are provided for rnonitoring the ,
quality of the condensate and final feHwater, as described in Subsation 9.3.2. Temperature mea, surements are provided for each :tage of feedwater heating. Steam pressure measurements are prosided at each feedwater heater. Levelinstrumentation and controls are provided for automatically reputating the heater drain flow rate to maintain the ;Dper levelin each feedwater heater shell or heater drain
,G tank. Hig5. level control valves provide automatic ij dump-to. condenser of beater drains on detection of high !cvelin the heate shell. ,
, The totat water volurne in the condensate and ,
feedwater system is maintained through automatic makcu.: and rejection of condensate to the conden, sate storage tank. The system makeup and . ,jection are controlled by the condenser betwell level controllers.
10A.8 Steam Generator Blowdctm System (PMR)
Not applicable to ABu%.
10A.9. Auxiliary Feedwater System (PMR)
Not applicable to ABWR.
l N Sliit%T - s 10 9.10 to.g 3 tow f4)
Amendment 11 10A.17 O
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l 10.6.10 Interfaces 10.6.10.1 Radiological of the TGSS effluents
! The Applicant referencing the ABWR design shell-perform a i radiological analysis of the TGSS effluents based on conservative site specific parameters. From this' analysis,Lthe Applicant
- shall determine the various actions to-be taken.if and when the TGSS effluent.radation monitor' detects preset levels of effluent contaiminations, including the level.at which the TGSS steam supply will be switched over to auxiliary steam.'(See Subsection .j 10.3.5,1. '
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[ Standard Plant REV D SECTION 11.5 l
[
CONTENTS i
Section Illlt Eagt i
11.5.1 Deslan Bases 11.5 1 11.5.1.1 Design Objectives 11.5 1 i
l 11.5.1.1.1 Radiation Monitors Required for Safety & Protection 11.5 1 .
l 11.5.1.1.2 Radiation Monitors Required for Plant Operation 11.5 1 11.5.1.2 Design Criteria 11.5 2 4
11.5.1.2.1 Radiation Monitors Required for Safety 11.5 2 11.5.1.2.2 Radiation Monitors Required for Plant Operation 11.5 2 ~
11.5.2 Svstem Descrintion . 11.5 3 11.5.2.1 Radiation Monitors Required for Safety 11.5 3-11.5.2.1.1 Main Steamline (MSL) Radiation Monitoring 11.5 3' 11.5.2.1.2 Reactor Building HVAC Radiation Monitoring 11.5 3 u s........
(N"M) 4# . 11.5 4 1
11.5.2.1.3 Fuel Handling Area Ventilation Exhaust i
Radiation Monitoring 11.5 4~
-11 5.2.1.4 Standby Gas Treatment Radiation Monitoring . 11.5 4 L 11 5.2.1.5 Control Building HVAC Radiation Monitorin6 :11.5 5 11.5.2.2 Radiation Monitors Required for Plant Operation - 11.5 6 l'.5.2.2.1 -
Off. gas Pretreatment Radiation Monitoring 11.5 6 11.5.2.2.2 Off gas Post Treatment Radiation Monitor I
l 11.5 6 l 11.5.2.2.3 Carbon Bed Vault Radiatioe Monitoring' System 11.5 7 s
11.5.2.2.4- Plant Vent Discharge Radiation Monitoring 11.5 8
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ABWR 23xetxxx Standard Plant nrv e SECTION 11.5 O coxrtxTs <ce ti##ea)
SectIon Hllt Eagt 11.5 S $ 5 Radwaste Efuuent Radiation Monitoring 11.5-8 11.5 " -. Reactor Building Cooling Water Radiation Monitoring 11.5 9 11.5.2.2./ Radwaste Building HVAC Radiation Monitoring 11.5 9
%g% Bai tcA ns Com gv 6:4 11.5.2.2/ p % Exhaust Radiation Monitoring 11.5 9 G k4u:4-11.522[ f Turbine Gland Condenser ND" Radiation Monitoring 11.5 9.1 IO Dr>well Sumps Drain Line Radiation Monitoring 11.5-9.1 11.5.2.2 [
11.5 3 Emuent Monitorine and Samtline 11.5 9.1 11.5 3.1 Basis for Monitor Location Selection 11.5 10 11.5 3.2 Expected Radiation Levels 11.5 10 11.533 Instrumentation 11.5 10 11.5 3.4 Setpoints 11.5 10 11.5.4 Process Monitorine and Sameline 11.5 10 11.5.4.1 Implementation of General Design Criterion 60 11.5 10 11.5.4.2 Implementation of General Design Criterion 64 11.5 10 .
11.5.4 3 Basis for Monitor Locatiot. Selection 11.510 ~
11.5.4.4 Expected Radiation Levels 11.5 10 11.5.4.5 Instrumeatation 11.5 10 11.5.4.6 Setpoints 11.5 11-11J.5 Calibration and Maintenance 11.5 11 11.5.5.1 Inspection and Tests 11.5 11 11.5.5.2- Calibration 11.5 11 11.5.5 3 Mainteaanec 11.5 12 11.5 lii Amendrncnt 16 1
ABWR useiooxx Standard Plant REV D SECTION 11.5 CONTENTS (Continued)
Sectlon Title East 11.5.5.4 Audits and Verifications 11.5 12 l
SECTION 11.5 TABLES Table Title East 11.5 1 Process and Effluent Radiation Monitoring Systems 11.5-13
! 11.5 2 Process Radiation Monitoring System (Gaseous
! and Airborne Monitors) 11.5 16 11.5 3 Process Radiation Monitoring System (1iquid
}
j Monitors) 11.5 18 11.5-4 Radiological Analysis Summary of Liquid
=
Process Samples 11.5 19 11.5 5 Radiological Analysis Summary of Gaseous 9 Process Samples 11.5 20 11.5 6 Radiological Analysis Summary of Liquid I Effluent Samples 11.5 21-l 11.5 7 Radiological Analysis Summary of Gaseous .
Effluent Samples 11.5-22 l
11.5 iv 9
Amendment 16 l
ABM - 224sioo4x Standard Plant arv. s 11.5 PROCESS AND EFFLUENT l RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 113.1.1.2 Radiatloa Moaltors Required for The process and effluent radiological Plant Operation monitoring and sampling systems are provided to allow determination of the content of radioactive The main objective of this radiation monit.
materialin various gaseous and liquid process oring is to provide operating personnel with and effluent streams. The design objective and measurements of the content of radioactive criteria are based on the following requirements: materialin all effluent and important process streams. This +4e*+ demonstratNa of (1) Radiation lastrumentation required for compliance with plant normal operational safety and protection. technical specifications by providing gross radiation level monitoring and by collection of (2) Radiation instrumentation required for _ halogens and particulates on filters (gaseous monitor and plant operation. effluents) as required by Regulatory Guide m ss p. r - 1.21. Additional objectives are to initiate n.c 11.5.1 Design Bases discharge valve isolation on the offgas or n,p liquid radwaste systems if predetermined release u . ai 11.5.1.1 Design ONectives rates are exceeded, and to provide for sampling _
- at certain radiation monitor locations to allow
"' 5' ' 11 3.1.1.1 Radiation Monitors Required for determination of specific radionuclide content.
Safety and Protection The process radiation monitoring systems also The main objective of this radiation monitor. provides the following design objectives:
ing'is to initiate appropriate protective action to. limit the potential release-of radioactive - (1) Monitors gaseous effluent streams materials from the reactor vessel and primary.and gd d secondary containment if predetermined radiation (a) Plant @ discharge A*od Ska ok levels are exceeded in major process / effluent .A.4 l streams. Another objective is to provide control (b)h",.y ,,_:7'wwbaust r=c room personnel with an indication of the radiation levels in the major process / effluent _ (c) Radwaste building ventilation exhaust streams plus alarm annunciation if high radiation ,
levels are detected. (d) Turbine gland steamp:_ g:. :r __p .
meer ;;;p ;hs.
The process radiation monitoring system provides the following design objectives: (2) ' Monitors liquid effluent streams (1) Main steamline tunnel area radiation ' (a) 'Radwaste effluent radioactivity monitoring; (b) Drywell sumps drain radioactivity 1 (2) Reactor building heating, ventilating, and air conditioning (HVAC) exhaust air rad- ,
intion monitoring; (3) J Monitors gaseous process streams (3) Fuel handling area HVAC exhaust air rad- (a) Off. gas pre treatment sampling lation monitoring;
- (b) Off gas post treatment sampling (4)- Control building HVAC air suply radiation monitoring; ~ (c) Carbon' bed vault gross gamma radiation levels (5) standby' gas treatment system.offgas Yadiation monitoring Amendment 16 : 11 5-1
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(4) Monitorsliquid process streams (10) Register full seale ouiput if radiation
- i. detection exceeds full scale i l (a) Reactor building closed cooling water j intersystem radiation leakage (11) Use instrumentation with sensitivities and l ranges compatible with anticipated 11J.1.2 Design Criteria radiatioa ieveIs.__
l
! WSENT- >
.d t t . S.1. '4 11J.1J.1 Radiation Moaltors Required for (The applicable General Design Criteria of l Safety f-10CFR50, A}Oendix A, are 2, 4, 13, 20, 21, 22, 23, 24, and 28 specified in Section 7.6.2.2. l The design criteria for the main steamline and :The process radiati on safety related sub. 1 containment ventilation exhaust plenum radiation systems shall meet the design requirements for l monitoring includes the following functional Safety Class 2,~ Seismic Category I, systems ,
requirements: along with the quality assurance requirements of 10CFR50, Appendix B. )
l_(1) Withstand the effect of natural phenomens' L I
(e.g... earthquakes) without loss of 11J.1.2J Radiation Monitors Required for i capability to perform their functions; Plant Operation l (2) Perform the intended safety functions in the The design criteria for opejational radiation:
environment resulting from normal and - monitoring wirwH includeg the follo_ wing abnormal conditions (e.g., loss of HVAC and ~ functional requirements:
isolation events) _ -
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(1) Provide continuous indication of radiationi l (3) ' Meet the reliability, testability, indepen. levels in the main control room' i
dence, and. failure mode requirements of
[ engineered safety features _~(2) Provide _ warning of increasing radiation q
- levels indicative of abnormal conditions by 1 (4) Provide continuous output in the main alarm annunciation - . j control room i .
_(3) Insof ar. sis- practical,f provide selft -_ l )
! l (5) ' Permit.- checking of the operationat - monitoring of components to the extent that availability _of cach channel during reactor- - power' failure' or component malfunction - ,
operation with provisions for calibration 'causes' annunciation and: discharge valve 1.
j' function and instrument checks isolation channel trip Jli I
! l(6) Assure an extremely high probability of' -(4) i Monitor a sample representative of the bulk' 1 O
l accomplishing safety functions.in the event: stream or volume ,
j -l = Lof anticipated operational ~ occurrences _.
p '(5) Incorporate provisions for calibration, and. '
y l ~(7)LI nitiate promp.t protective action prior to - : functional checks exceeding plant technical specification .
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-(6):_L Use _ instruments with: sensitivities 'and!
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i- l(8) ' Provide warning'of increasingiradiation - - radiation levels ;
i levels Indicative of abnormal conditions by = . .. . _ _.
i p .l. alarm annunciation 1(7) L Register fulliscale' output tif radiation: 1 1 detection exceeds full' scale.1 4 i l(9)f insof ar as practikal,~ provide Lselft m Li I
monitoring of compotents to the extent that: The radiation system'that monitors discharges 3 power failure or comeonent malfunction L from the gaseousland liquid radwaste. treatment
] causes annunciation an l channel trip } system shall_have provisions to. alarm and to} t initiate Lautomatic' closure off the : waste; N
V j Amendment 16 - 11.5 2 i w
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Design criteria of this system are based on meeting the relevant requirements of General Design Criteria -(GDC) 60, 63, and 64 of 10CFR50, Appendix A, in accordance with SRP 11.5 of NUREG-0800.
l These GDCs are in addition to those GDCs that are specified in t
Section 7.6.2.2 for system instrumentation.
if Also, the system is designed to meet the applicable provisions of 10CFR20.106, RG 1.21 and RG 1.97.
The safety-related-process radiation monitoring subsystems;are i
classified. Safety class 2,-Seismic' Category 1. These subsystems conforaL to the quality assurance requirements of 10CFR50,' Appendix B.
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ABM ursiooxx Standard Plant REV B discharge valve on the affected treatment system is visually displayed on the affected radiation prior to exceeding the normal operation limits monitor. A high.high or inoperative trip in the specified in technical specifications as required radiation monitor results in a channel trip by Regulatory Guide 1.21. gwhich is provided to the reactor protection F system (RPS) and to the leak detection and The applicable General Design Criteria ok isolation system (LDS). Any two out of.four l 10CFR50, Appendix A, are 60,63, and 64 in . channel trip results in initiation of main accordance with the Standard Review Plan for steamline isolation valve closure, reactor (Section 11.5 (NUREG 0800), scram, main condenser mechanical vacuum pump (MVP) shutdown, and MVP line discharge valve
.U.S.2 System bescription closure. A high trip actuates a MSL high control room annungator common to .all 11.5.2.1 Radiation Monitors Required for Safety channels. High and low trips do not result in a channel trip. Each radiation monitor 9:@ ^--
Information on these monitors is presented in displays the measured radiation leveh% > m e jg Table 11.51 and the arrangements are shown in Subsection 7.6.1.2. 11J.2.1.2 Reactor Building HVAC Radiation Monitoring gi % 6,,,, ,4 d W.
11.5.2.1.1 Mala Steamilne (MSL) Radiation Monitoricg This system monitors the radiation level in th reactor building ventilation system -
This subsection monitors the gamma radiation exhaust duct. A high activity level in the level exterior to the main steamlines in the MSL ductwork could be due to fission gases from a tunnel. The normal radiation levelis produced leak or an accident, primarily by coolant activation gases plus smaller quantities of fission gases being The system consists of four redundant instru-transported with the steam. In the event of a ment channels. Each channel consists of a gross release of fission products from the core, digital gamma sensitive GM detector and a the monitoring channels provide trip signals to ecstrol room radiation monitor. Power is the leak detection and isolation system, supplied to each channel, A. B, C, and D monitors from vital 120 Vac Divisions 1,2, 3 The MSL radiation monitors consists of four and 4 respectisely. A two pen recorder powered redundant instrument channels. Each channel from the 120 Vac instrument bus allows the consists of a local detector Guhdion output of any two channels to be recorded by the chamber) and a control room radiation monitoriuse of selection switch _egTET detectors are with a trip auxiliary unit. Power for channelst located adjacent to the exhaust ducting upstream-A, B, C, and D monitors is supplied from vital of the ventilating system isolation valves,b 120 Vac divisions 1,2,3 and 4 respectively, All four channels are physically and electrically Each radiation monitor has four trip circuits:
independent of each other. two upscale, one downscale and one inoperative similar to MSL radiation monitors.
The detectors are physically located near the -
main steamlines (MSL)just downstream of the out-board main steamline isolation valves in the Qu/ h2 M / #
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stream tu'2nel. The detectors.are geometrically 4 . u s arranged and are capable of detecting significant increases in radiation level with any number of b< ["*'7 p 4
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main steamlines in operation. Table 11.51 lists 4*/-#**
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the location and range of the detectors. 4,4 j. %<tr .M' " b \. -
Each radiation monitor has four trip jag eru Ma<M #
circuits: two upscale (high high and high), one g_/g'fMM.-.y /uc g,
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downscale (low), and one' inoperative. Each trip ye j O _ _ ts TD 4 hs@
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i MM 23A6100AK Standard Plant REV D p ' A high.high or inoperative /downscale trip in the radiation monitor results in a channel trip which is provided to LDS. Any two.out of.four channel trips will result in the initiation by LDS of the standby gas treatment system (SGTS) and in the isolation of the secondary containment (including closure of the containment purge and 113.2.13 Fuel Hand 11og Area Ventilation j vent valves and closure of the reactor building Exhaust Radiation j ventilating exhaust isolation valves).
i This subsystem monitors the off gas radiation Tbc high.high trip will initiate an alarm in level in the fuel handling area ventilation the control room common to all channels, exhaust duct. The system consists of four channels which are physically and electrically A downscale inoperative trip is displayed on independent of each otheg. Each channel the radiation monitor and actuates a control room consists of a digital gamma.senytive GM detector annunciator common to all four channels. and a control room radiation monitor. Power for channels (A, B, C, and D) is supplied from the The high radiation trip is provided and vital 120 Vac divisions 1, 2, 3 and 4 actuates a control room annunciator common to all respectively, channels.
Each radiation monitor has four ciicuits: two upscale, one dowescale and one inoperative similar to the MSL radiation tuonitors. This Each radiation monitor will display the subsystem performs the same trip functions as measured radiation level, those described in Subsection 11.5.2.1.2 for the reactor building HVAC exhausi radiation 4L5&lM T h e monitoring.
11.5.2.1.4 Standby Gas Treatment Radiation Monitoring This subsystem monitors the off. gas radiation levelin-the SGTS exhaust duct to the stack using four channels.
Two ionization chamber detectors are physically located downstream of the exhaust and heat removal fans and dampers on the exhaust duct to the stack. Two other scintillation detectors are used during off. gas sarnpling of 9 the gas exhaust to the stack.
Amendment 16 11.s.4
MM 23A6100AK Rrv n Standard Plant ad p, The subsystem consists of four instrumented inlet air, from any source, L i ? *;n will l Q channels. Each channel consists of a detector provide isolation of intake of leakage from and a main control room radiation monitor, accident sources escaping from other plant buildings.
Power for the channels is supplied from the non lE vital 120Vac scurce.
Each radiation monitor has four trip circuits:
two upscale, one/ inoperative and one donscale.
All trips are displayed on the appropriate radiation monitor and each aeruates a common main control room annunciator for high high, high and low / inoperative indications. Each radiation channel consists of a digital gamma sensitive GM detector and a radiation 11J.2.13 Control Building HVAC Radiation monitor which is located in the control room.
Monitoring Each radiation monitor has four trip The control building HVAC radiation circuits: two upscale, one/ inoperative and one monitoring subsystem is provided to detect high downscale. All trips are displayed for the radiation level in the normal outdoor air supply, appropriate radiation monitor and each actuates automatically close the outdoor air intake a controt room annunciator, dneponyand the exhaust dampers, and initiate automatically the outdoor air cleanup system in -
p the emergency racirculation air supply loop. The
( emergency recircu ation fans shall be started and area exhaust fans stopped on high radiation.
The radiation monitors for each of the '
control building HVAC systems consist nf f^ur c
redundant channels to monitor the air *to the-building. Each radiation monitor is physically
- [/lf -
, separated and powered from separate vital 120 Vac divisional power busses. Failure of one channel will not cause isolation of the HVAC system.
~
- o. another channelis provided to monitor for radioactive contamination in the air that is being supplied to' the control room complex - )
downstream from the supply fans. The detector is located at the commori HVAC duet that supplies ~ the ,
air to the various areas in the controibuildingc Power to this monitor is provided from the non.1E vital 120 Vac bus.
The monitors sW meet the requirements for Class 1E components to. provide-appropriate
. reliability. The system will warn of the-O presence of significant air contamination-in-V Amendment 16 '11J 5
- ABM :sssiooxx Standard Plant REV B
- cL a 11.512 Radiation hfonitars Requirsd for Plant The radlation level output y the monitor can
- / N Operation be directly correlated to t e concentration of
&1 the noble gases by usirg eminutomatic vial
. Information on these monitors is presented in sampler panel to obtain a grab sample. To draw Table 11.51, a sample, sampler a serum holder,#the bottle sample !inesisare inserted evacuatedinto a[
11.5.2.2.1 OG gas Pretreatment Radiation and a solenold. operated sample valve is opened Afonitoring to allow offgas to enter the bottle. The bottle is then removed and the sample is analyzed in This subsystem monitors radioactivity in the the counting room with a multichannel gamma condenser offgas at the discharge of the delay pulse height analyzer to determine the concen.
pipe after it has passed through the offgas tration of the various noble gas radionuclides.
condenser and moisture separator. The monitor A correlation between the observed activity and detects the radiation level which is attributable the monitor reading permits calibration of the to the fission gases produced in the reactor and monitor, transported with steam through the turbine to the condenser. 11.5.2.2.2 Off. gas Post.htatment Radiation blonttoring i A continuous sample is extracted from the I offgas pipe via a stainless steel sample line. This system monitors radioactivity in the it is then passed through a sample chamber and a offgas piping downstream of the offgas systern sample panel before being returned to tbc suction charcoal adsorbers and upstream of the offgas side of the steam jet air ejector (SJAE). The system discharge valve. A continuous sample is sample chamber is a stainless steel pipe which is extracted from the offgas system piping, passed internally polished to minimize plateout. It can through the offgas post treatment sample panel be purged with room air to check detector for monitoring and sampling, and returned to the -
response to background radiation by using a offgas system piping. The sample panel has a-W three way solenoid. operated valve. The valve is pair of filters (one for particulate collection Q controlled by a switch located in the maic control room. The sample panel measures and and one for halogen collection) in parallel (with respect to flow) with two identical indicates sample line flow. Two seeeeeeeerr e1% detectors. Two radiation monitors i A S: detectors are positioned adjacent to the in the main control room analyze and visually
[ vertical sample chamber and are connected to display the beasured gross radiation level,
' ( radi tion monitors in the main control room.
b 3dal g ow a-s.wh. o k (GM l Power is supplied from 120 Vac instrument bus for radiation monitor and detector and for the sample and vital sampler panels. The sample panel shielded chambers can be purged with room air to chech detector response i The radiation monitor has four trip circuits: to background radiation by using solenoid valves
- two upsple (high high and high), one downscale operated from the centrol room. The sample l and one inoperative. panel measures and indicates sample line flow.
l A solenoid operated check. source for each-I The trip outputs are used for alarm function detector assembly operated from the control room only. Each trip is visually displayed on the can be used to check operability of the gross radiation monitor and actuates a control room radiation channel.-
annunciator: offgas high high, offgas high, and offgss downscale/ inoperative. High or low sample Power.is'supp!!ed from a 120 Vac instrument l line flow measured at the sample panel actuates a bus to the radiation monitors and to the two pen E main control room offgas sample high low flow -
l annunciator.
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Amendment 16 11.5-6 s._- ___-__ _
MM 2sA61ooAK arv.n.
Standard Plant recorder. A 120.Vac local bus supplies the sample panel.
Each radiation monitor has four trip circuits:
two upscale (high high and high), one downscale (low) and one inoperative. Each trip is visually displayed on the radiation monitor. The trips {
actuate corresponding main control room j annunciators: offgas' post treatment high high radiation, offgas post treatment high radiation, and offgas post treatment downscale/ inoperative.
High or low 4 ample flow measured at the sample panel actuates zi .. ..;: ..;. a"... x: 4he.
& - ;' abnormal flow annunciatop m da cedr el v o om .
A trip auxiliary unit in the control room takes the high high and downscale trip /inoper, ative outputs to initiate closure of the offgas system discharge and bypass valves.- The high high trip setpoints are determined so that - 'i valve closure is initiated prior to exceeding technical specification limits. Any-one high upscale trip initiates closure of offgas system bypass line valve and permits opening of the treatment line valve.
A vial sampler. panel sletilar to the pre.
- tr@w sampler panelis provided for grab .
sample collection to allow isotopic analysis and
. gross monitor calibration.
11J.2.23 Carbon Bed Vault Rad!ation Monitoring - ,
The carbon vault is monitored for gross.
gamma radiation level with a single instrument . ..
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channel. The channelincludes a digital sensor ,
and converter, and a radiation monitor. The radiation monitor is located in the main control
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x room. The channel provides for sensing and
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. readout of gross gamma radiation over a range of'. i six logarithmic decades (1 to 10 6 mR/hr).
1 l The monitor has-one adjustable upscale trip-L circuit forl alarm and one downscale trip for finstrument trouble. Power is supplied from 120. i
--Vac instrument bus.- 7
. Amendment 16 - 11.51 ~
._____x-_. -- _ _ _:_--__w__ - _=
ABM Standard Plant Areach 4+** :sisiooxx uv. n 11J.2.2.4 Plant Vent Discharge Radiation Nith the exception of the radwaste system g Monitoring effluent, the streams monitored normally contain '
U; t
only background levels of radioactive i This system monitors tbc plant vent discharges materials. Increases in radiation level may be for gross radiation level during normal plantindicative
[ of heat exchanger leakage or, operation and collects halogen and particulate gulpment malfunctionf- 11.5.1
- samples for laboratory analysis. Also, this system utilizes a high. range radiation mon tor -
(l.O that measures fission products in plant gas ous effluents during and following an accident. A *The., d\ s ch co-te J Ovmwed representative sample is continuously extracted from the ventilation ducting through two isokinetic probes in accordance with ANSI N13.1 MS Cowwwen f kcu k and passed through the containment ventilation WMk incb ck e., H\/ A C sample panels for monitoring and samphng, and ;
4 returned to the ventilation ducting. Each sample panel has a pair of filters (one for particulate
_g g gg $4 '
collection and one for halogen collection ) in parallel (with respect to flow) for continuous gg /g 4 4 >
Faseous radiation sampling. The gross radiation detection assembly conslats of a shielded gggggQ Q chamber, beta. gamma. sensitive GM tubes, and a check source. The extended range detector S erM CA b]5 ;
i assembly consists of an ionization chamber which measure radiation 1eyeIs up to 11.5.2.2 3 Radwuste Emuent Radiation p10/ce.
mci A radiation monitor in the main ' Monitoring control room analyzes and visually displays the Q measured radiation level. This subsystem continuously monitors the C/ radioactivity in the radwaste effluent prior to The sample panel shielded chambers can be its discharge j anci clro m g c..
purged with room air by using two scienoid valves operated from the control room to check detector Liquid waste can be discharged from the response to background radiation, thus-checking sample tanks containing liquids that have been operability of the gross radiation channel, processel through one or more treatment systems such as evaporation, filtration, and ion Power is supplied from 120 Vac local bus for. exchang. Prior to discharge, the liquid is the radiation monitor and for the sample panel, extracted from the liquid-drain treatment
, process pipe, passed through a liquid sample The radiation monitor initiates trips for panel which contains a detection assembly for alarm indications on high high, high, and low gross radiation monitoring, and returned to the ;
radiation from each detector assembly. Also, the process pipe. The detection assembly consists sampled line is monitored for high or low flow of a' scintillation detector mounted in a-indications and alarming. shielded sample chamber equipped with a check ,
source. A radiation monitor in the control room Table 11.5 2 presents the gaseous and airborne -analyzes and visually displays the measured monitors for the effluent radiation monitoring gross radiation level, system.
- The sample panel chamber and lines can be 113.2.23 Uquid Process and Emuent drained to allow assessment of background
. l Monitoring buildup - The panel measures and indicates sample line flow : A solenoid operated check These subsystems monitor the gamma radiation source operated frorn the control room can be l levels of liquid process and effluent streamsj - used to check operabili_ty of the channel..
Amendment 16 1158
23A6100AK Standard Plant an n l Based on acceptable radiation levels. The trip signals are annunciated in the O discharge is permitted at a specified release radwaste building control room and in the main rate and dilution rate.
nre t.
control room.
The radiation monitor has feirr trip circuly. Each radiation monitor visually displays the Two upscale trips (high high and high)*fone radiation level and supplies an output signal to the computer.
downscale/:-!p, ::d ::: ino erative trip, E
% Li3 bb h 3wPs co
- M P*"
- A s$am moschec\r. s ov y-ce i.s provs cle cl 4o v" ch om W 4I 4h w pea does c-4. u.sc.8d
- cl +4.o /*o 5 0f f+** 4 ' " e,&bghon, e W ed P v= P.AI48 de two upscale trips and the low downscale/
inoperative trip actuate annunciators in the main control room and in the radwaste building control room. Table 11.5 3 describes the liquid monitors used for p ocess radiation monitoring.
11.5.2.2 9 Reactor Bulldlog Cooling Water Radiation Monitoring This subsystem consists of three channels: one for each RCW A, B and C loop for monitor'ng intersystem radiation leakage into the reactor building cooling water system.
Each channel consists of a scintillation p detector which is located in a well near the RCW V heat exchanger exit pipe. Radiation detected from the three channels are multiplexed and fed into a common peeeses radiation monitor. This monitor provides individual channel trips on high radiation level and downscale/ inoperative indication for annunciation in the control room.
Power to the monitors is provided from the non.1E vital 120 Vac source.
7 Ehoush 11J.2.24 Radwaste Building HVAC Redistion & N e b .a. b \ h wp Co M av' heed Ebs'.
Monitoring 11J.2.2/ C"g . /.m " r " J.6 Monitoring System This subsystem monitors the radwaste building _x ventilation discharge to the stack, including This subsystem monitors the(gvent-radwaste storage tank vents, for gross radiation discharge in the turbine building :; :;r::: cogad' level. The system consists of two redundant ++++s for gross radiation levels. The Instrument channels, each channe a local monitoring is provided by four channels (two detector, a converter, and a ma' control room . redundant sets). Two redundant channels monitor radiation monitor. Power is pplied to each. radiation in t'ne equipment area air exhaust duct channel by the 120.Vac: instrument bus. .and the other two redundant channels monitor the Nvmej radiation in the SJAE area air exhaust duct.
Each channel uses'a digital detector located j
Each radiation monitor provides two trip adjacent to the monitored exhaust duct. The -
circuits: one for upscale (high) radiation and outputs from' each set = of- detectors are one for downscale/ inoperative trip. multiplexed and then fed into two separate Amendment 16 1t.5 9
ABWR zu6imax Standard Plant REV D q process radiation monitors for display, recording are monitored for radioactivity releases in Q and annunciation. Each monitor provides alarm accordance with Criterion 64 of General Desica trips on radiation high and on radiation low Criteria,10CFR50, Appendix A, as follows:
(downscale/ inoperative).
g g h h d 'E!r Ptick41c HedM .g 11.5.2.2.8 Turbine Gland Condenser ::? 1* ....
P r "; - ": -!Y - %....;ag This subsystem monitors the off gas releases to the stack from the turbine gland seal system, H H 4: c::: n -i .._ _ -""~ r" r The off gas releases are continpously sampled and monitored for noble gas *3 by a scintillation detector. The output signal is multiplexed and then fed to a shared radiation monitor in the main control for display, recorJjng and annunciation. This monitor provides aben tripf a ku t j oa' on radiation high and radiation low (downscale/ inoperative). one o A grab sample of the off gas is provided for laboratory analysis. Also, samples of halogens and particulates are collected on filters for periodic analysis.
A gamma source check is provided for channel O calibration purposes.
( 10 11.5.2.2.jlDrywell Sumps Drain Line Radiation Monitoring This subsystem monitors the radiation levelin the liquid waste that is transferred in the drain line from the drywell LCW and HCW sumps to the s radwaste system. One' monitging chann I is - S provided fou:she demediusdnntrgeach sump .Each dVCi N channel uses an ionization chamber which is p located on the drain line from the sump , (p -tvo* g output from each sensor is multiplexed and then gui
- fed into a shared psamass radiation monitor for t S o k y
- q c.i n ,
display, recording and annunciation.-
y bee (0 4k v et.
The radiation monitor provides im trip circuits: two upscale (radiation high high and high), one downscale[nedumme inoperative. The high high signal is used to close the outboard isolation valve in its respective drain line.
All trips are annunciated in the main control, rootn.
11.5.3 Eftluent Monitoring and Sampling All potentially radioactive effluent materials (On)
Amendment 16 11.5-9.1 l
______---__a-
23A6100AK -
Standard Plant nry n L
(1) 'tiquid releases are monitored for gross -(1) Off gas post. treatment (9 gamma radioactivity;
\._/ (2) Reactor building coeriadeanef HVAC air (2)'fvolid wastes are monitored for gross gamma -
exhaust, radioactivity; and (3) Fuel handling area air exhaust.
(3) ggaseous releases are monitored for gross 3 2 g gamma radioactivity. (4) Drywell Sump Liquid Waste drain 11.53.1 Basis for Monitor Location Selection N EO*'I' ' N I.M Monitor locations are selected to assure that 113.4.2 Implementation of General Design all effluent materials comply with regulatory Criteria 64 i requirements as covered in Regulatory Guide 1.21.
Radiation levels in radioactive and poten.
11.53.2 Expected Radiation levels tially radioactive process streams are monitored for radioactivity releases. These include: l Expected radiation levels are within the ranges specified in Tables 11.5 2 and 11.5 3. (1) Main steamlin[- ,
11.5.3.3 Instrumentation (2) Off. gas pre-treatment and post.treatmen$
The process radiation monitors used for (3) Carbon bed vaul measuring radioactivity are listed in Table 11.5 1. (4) Intersystem leakage into reactor building.
cooling water ( .
Grab samples are analyzed to identify and-(m'j quantify the specific radionuclides in effluents 11.5.43 Basis for Monitor Location Selection and wastes. The results from the sample analysis, are used to establish relationships betw:en the Monitor locations are selected to assure gross gamma monitor reading; and concentrations compliance with Regulatory Guide 1.21 in that or release rates of radionuclides in continuous . semple' points ar,e located where there is a.
effluent releases. minimum of disturbance due to fittings and other physical' characteristics of _the equipment and 11.5.3.4 Setpoints - = components. Sample nozzles are inserted into the flow or liquid volume to ensure sampling the The radiation level _ trip setpoints for bulk volume of pipes and tanks, in the case of actuation of automatic control features thatt both liquid and. gas flow,1 care is taken to initiate actuation of isolation valves, dampers assure that individual sa~mples are actually or diversion valves are specified in the plant ^ represcritative of the effluent mixture. A more technical specifications as indicated in Table detailed discussion is given in~ ANSI N13.1; 11.5 1.
21.5.4.4 Expected Radiation hvels
. Expected radiation levels are listed in Tables 11.5.4. Process Monitoring and Sampling 11.5-2 and 11.5 3.
i 11.5.4.1 Implementation of General Design 11.5.4J Instrumentation Criterion 60 -
The process radiatl'on monitors used for All_ potentially significant radioactive dis- . measuring radioactivity are listed in Table charge paths are equipped with a control system - 11.5 1.
O
-U to automatically isolate-the discharge on indi-cation of a'high radiation level.- These include:
Amendment 16 ' 11.5.t0 i
23A6100AK Standard Plant Rev n C
O Grab samples are analyzed to identify and (4) tontrol building HVAC
! quantify ihe specific radionuclides in process 4 streams. The results from the sample analysis (5) reactor building cooling water system are used to establish relationships between the SpMqqM-enLE gross gamma monitor readings and concentration (6) SGTS and radionuclides in the process streams. y (7) turbine buildingy _ _ : .f ~
11J.4.6 Setpoints O (8) Offgas pretreatmen@
The radiation trip set points for the various c monitors are listed in Table 11.51. (9) carbon bed vault (
11.5.5 Calibration and Maintenance The following monitors include built.in check sources and purge systems which can be operated 11.5.5.1 Inspection and Tests from the main control room:
O During reactor operation, daily checks of < (1) offgas post treatmenf system operability are made by observing channel behavior. At periodic intervals during reactor v(2) pplant vent dischargef ~
operation, the detector response of each monitor L provided with a remotely positioned check source , (3) Aquid waste discharge will be recorded together.with the instrument sGTS background count rate to ensure proper function -(4) c";r ='f ing of the monitors. Any detector whose re- g sponse cannot be verified by observation during -(5) rtdwaste building exhaust gg, normal operation or by using the remotely post- a g4 cu m ,,, o ,g tioned check source will have its response M 9 checked with portable check source. A record will be maintained showing the background b"'"' '~
F7
^
radiation level and the detector response. 115.5.2 Calibration The system has electronic testing and cali. The continuous radiation monitor calibration brating equipment which permits channel testing is according to certified National Bureau of without relocating or dismounting channel compo. Standards of commercial radionuelide standards, nents. An internal trip test circuit adjustable and is accurate to at least ' + or 15E The over the full range of the readout meter is used source detector geometry during primary cali-for testing. Each channel is tested at least bration is identical to the sample detector geo-semiannually prior to performing a calibration metry in actual use. Secondary standards which check. Verification of channel operation and were counted in reproducible geometry during the trip function will be done at this time if it can primary calibration are supplied with each con-be done without jeopardizing plant safety. The tinuous monitor for calibration after installa.
test will be documented, tion. Each continuous monitor is calibrated during plant operation or during the refueling The following monitors have alarm trip outage if the detector-is not readily ac-circuits which can be tested by using test -cessible. A calibration can also be performed signals or portable gamma sources: by using liquid or gaseous radionuclide stan-g y dards or by analyzing particulate iodine or gas-(1) dain steamlinep cous grab samples with laboratory instruments. g (2) feactor building HVA The offgas pretreatment monitor shall respond
- p. to a gross gamma signal obtained from the ,
periodic analyses of grab samples. The readout l.
(3) ' fuel handling area HVA units shall be mR/hr per mci /sec.
Amendment 16 115t1
l l 23A6100AK Stondard Plant RI?V B s The following monitorsdiset= A gross which would affect the calibration, a recali.
r ; y ' 4e-a.
1 gamma signal c2:_ ::d 'r :h r.. '
- ""7:2s bration is performed at the completion of the c ' M - p!:_ :: n_2, . ._ in counts / min work.
t (1) 6ff gas post treatment 7 - 1133.4 Audits and Verifications mt (2) Pplant M dischargf- Audits and verification during normal plant operation are out of scope for the Standard ABWR Plant.
(3) ( dwaste ra v'
effluen*45U$'
e.(kush (4) ggland steam condenser and =:= - v .ii p (g) k d a b ild E C.oel/^5 " "' 'Y "
pvonda w aau v< s.d s
- 2 The following monitors are calibrated toyd the gross gamma dose rate in mR/hr:
( I Fe bube P s
n, (1) iham steamline; F. gg,p, hg g (2) eactor building HVAC[ [ C;,.D ywcl).5u I'78 L '
- f-(3) fuel handling area HVAC[ ;
(4) carbon bed vaultI C
(5) control building HV,A Ts.l.Qtdig g[ 4~ Ceydm b MW 'L -
(6) eE t name _
a (7) tadwaste building HVAC exhaust [
1133.3 Maintenance l All channel detectors, electronics, and recorder are serviced and maintained on an annual-basis or in accordance with manufacturers
- recommendations to ensure reliable operations. ;
Such maintenance-includes cleaning, lubrication, and assurance of free movement of the recorder in addition to the replacement or adjustment of any components required after performing a test or calibration check. If any work-is performed Amendment 16 11.5-12 l
MN 21A6100AK Standard Plant prv s TABLE 1151 .
PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS th==l= Setootnt _s Monitored No. Detector Sample IJne Channel Warning ACF Process of Ins or Detector Bangt Alann IIin Ssah Chan.
ash L< estion (Q 1) gi,3,i 7
A. Saferv Related Monitors Main steam- 4 Immediately 1106 mR/hr above full technical 6 dec. log line bne.1 downstream is specification of plant main gg-c } power backpound, awm steamline ^ ~f ' below trip isolation valve Reactor 4 S/C Exhaust duct 0.01 to above back- technical 4 dec, log building upstream of 100 mR/hr pound, specification HVAC exhaust ven. below trip e x host tilation isolation valve
- Control / S/C Intake duct 0.01 to above back- technical 4 dec, log building , upstream of 100mR/hr pound, specification HVAC ow intake venti- below trip hpf ly lation isola-tion valve f d./Jo /0 crai . 4 Standby gas 2 S/D SGTS exhaust Hi-te above back- technical / dec, log treatment air duct 100 2/b pound, specification system oM-jes downstream fo-15 h /o-6 below trip 2 IC .of exhaust up +e te' above back. None 6 dec log and beat re- pCum pound moval fans #"'t 8 and dampers Fuel 4 S/C Locally above 0.1 to 103 above back- .
4 dec, log bandling , operating mR/hr pound, technical area aw floor below trip specification CEhA ush
.p 4 CLc444 s 80v ud O' '^
Amendment 16 11.5 13
l ABWR uxswoxx
- Standard Plant prv n TABLE 11.51 -
PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS (Continued) tim Setnoints Monitored No. Detector Sample Line Channel Warning ACF Process of Im or Detector Banas altist Iria Ssals Chan. LesalJan (yde Q B. Moniters Recuired for Plant Ooeration f
/o A' #0<t -
4echice.l Radwaste 1 S/D Sample line N above back- -Meee. - 5 dec. log liquid dis- counts / min pound, Ef t edicda-ch below trip Reactor 3 S/D RCW Hx 10 to 10 above back. None 5 dec. log building line exit counts / min pound cooling water system 6
Offgas 2 GM B Sample line 10 to 10 above back- technical 5 dec. log che m el- pos4- counts / min pound, specincation vanh -9_:t 4vedm.,k below trip O 6/C V Offgas 1 Brt Sample line 1 to 10 6
at tech spec None 6 dec. log mR/hr report level eherce:ri uul: Met 4ve pW dmd-
${C c 4 Charcoal 1 ih4 On Charcoal 1 to 10 above. __
None 6 dec, los vault vault h#C mR/hr backpound ex ha w s4-L6 e. -,
6 Plant 1- GM B Sample line 10 to 10 at quarterly None 5 dec. log discharge counts / min tech spec
/o*'34 /ogg level-1 IC Sample line above back. None 6 dec. log pound, below trip Radwaste 2 GM B Exhaust ducts 0.01 to 100 above back- None 4 dec. log building WM. mR/hr ~ pound, vent - below trip Amendment 16 11.5 t4 l
l l
ABWR m am^x Standard Plant REV B TABLE 11.51 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS (Continued) hSetnoints Monitored No. Detector Sample une Channel Warning ACF EE2c:n of Tug or Detector Range Alarm Itia Scalt Chan. Location nm (t{oh 1) 11.5 1 B. Monitors Reouired for Plant Outstion a 4 S/C Exhaust duct 0.01 to 100 above back. None 4 dec. log c
j exhaust mR/hr ground
/4 to' &
Drywell 2 IC Drain line 4:01: nod 80 above back- Technical /dec. log sump assa from LCW & mR/hr ground specification SpJ HCW sumps d din Sl.D o.1 do lofcpm Tedese 1 1;C Sample line nyttrTUS above back- None 6 dec. log Glands 4ca% pGefer- ground Condeser 6 h
\ W e.us4 Msc6 fc. ,
ad e ' ^ W*
ACF Automat' ControlFunction '
, ,e n._ . , _ _ - - _ ,n &
GhE B BeIa NGhbetector "'
C"C C ;. L.a 1. CLd-IC loo Chamber S/C. Di; ital Gamma Sensitive GM Detector S/D Scintillation Detecter '
u . 5.1 Note 1. MU The channel range specified in this table is the equipment measuring or'dispicy range of the indicated parameter. Refer-to Tables-11.5-2 &.
11.5-3 for the dynamic detection range of the monitoring channel' expressed as-' concentration in units of microcuries per cubic centimeter, referenced to a specific nuclide.
O Amendment 16 11.5 15.
l 1. 5. I sg NTi nts ABWR ( 2.) TAous mm Standard Plant Rrv s TABLE 11.5 2 t
PROCESS RADIATION MONITORING SYSTEM (GASEOUS AND AIRBORNE MONITORS)
Radjation Configu-h%h Ddec W Principal Radionuclides Erpected Alarms Monitor git].qIL, 2 33 Senstthity Egggg Measured Aethits* * & Trios
/0'fVo/0 M/ j tt ic -
Offgas Offline B-Gert Pernew. 0.25 epm /pCi/ 1M MssiW,63* .5.4p6. fyjo%fp'c c Wb HlL chae4e.L le+e em 3 gin INOP/ Low vM +_m Filter, P C5 - i37 High pat + keh'"4 ledine High High Elter, T T-13l 7 00
' 'O " ^'00A'Icc.
Offgas Adjacent tid mR/hr 1@~IO6- Noble gas -F "A B;Gertew >
CA44400 to y e/ '
t/ C,'[(C @ fi.Lsion pf o. S~
vh sample I ducts High High cha:nber Ngh P
" "' ~"' 3 t~IS#
39 Cus Hl4 Main Steam. Adjacent G C fx1010 100 106 Coolant -100 mR/hr INOP line Radia- to steam At::p mR/hr activation law High tion lines
@e /R/hrcoJ4- gases High High 1 Ou loM/4bC</ec C' Charcoal GEssse s/c
-Dpli o.rme/Ar[cc jb /0-84 0 if - if Noble gases Negligible Lew /A9 4 vault mAfbr- Ep 44O/me# ,
Tis S{c o, Cme {hv l0 l iv 4 fl0
^::[, G6iuse @;:S S: f ^:l 10 W NdEI #
M es&p,umme;t exhaust h6 pcvto'8aCi[cc cym A b.//U Xe- 133 Xe ISS~
AfDM. 4[:'NOPr mR/br[- - -Hi-p Dr /hg4 Es High
~b
% 1 A<. 0 5' '0 / rfo fo 10 4 {le' 40fcc.
Reactor Offlume S/C $$t mR/hr OE - 17 noble gases 4 r10*2 tsopg; building HVAC , -k-155+ g w ait exhaust P" to-J.aCl/e lc @.u CilCC ge 155' g;s .
-l' Hia Hi Pt a.d. ved - .s. Gerg
.j.
lo /o lo Ye-\33 + 4,.rfff f 1:du c.gh .
F44s* 250 epm /pCi/ M MMI
^
U"-" Offline - Hish/ Low -
0444 h Pffsikq , o g/cc g3 b3hp 6terrP. AddCC , cs,gg g - LNop/ Low Hi;h (no*=dnage) i+4ia*
filter.- T - 2 - 13 / - -High High p
- Sensitivity based upon this radicnuclide.
- Erpected activi:ies are estimated based on cistingplants, Q. l Amehdment 16 11.5-16
g , ,g , g E W ~T
- t5 .
L ( 2) TACbD ABM Standard Plant _
23mmo:
prv n TABLE 11.5 2 O PROCESS RADIATION MONITORING SYSTEM (GASEOUS AND AIRBORhT MONITORS)(Continued)
DF' " '. Principal Radiation Conngv- D$$nIwa Radlonuclides Erpected Alarms Monitor tlLilpIt I.gg SensitMtv Eanat Measured Aet M h** $1rtes I go 2 Ofiline ; M.&ilb~!Ogp to 10 8 Xe 133' N/0 'p[gl(( Hisb/1.ow Main Stack (High Range) .aC gi/cc mC1/4c flow ac A 4/ W./CC .
INOP/L.ow High 14 o, S- /0# h /li XC-j3 '[(C.
Ra;' waste Omec B.GM ett mR/hr 0+tet90 - Mp46 M Lew 49 4. Sp building 0T Lle n W hr h f" 80~3.nC,/,[ mRfhy[(f INOP Wh 1,w kg/MP ventilatiou p,.g, p .#f< . Cg - t % 1 discharse * : m.r tio r r- t5 l nw hit.
-~ 1
- g-
/.3.8 Cp*u /p-fo/0 7 g,g$g m jg4,(,j Glandsteam C.Tline S/D IM - 4spr condenser seul fr pli4v.p%Ci/cc cr 'n'.n Cs-111 #
led+ftpF' lNOP[D:ct 4f y,,,
a . . dC@C A
El4"' S T- t5l Mi '
O . S' /8 dD/0-/ mgh.49 (wer Control Bldg. Oferec- 3/C asi mR hr &etet90 Neeff Negligible lh%g l
" "- 4 i52
- w CC 1,3,c ^tr r"';,.c%-pyCc I,lscreggn to-7fo jc*l m gyjgA.a0th ,
Standby Gas Offline m S/D .1erstte" 10 nf'.*C//cc Cedar 4ep, MHigh
~'Ticatment 'M 2/ pfti/cc, fr /IS He W, kr.s~ 1, gqh Exhaust Talk 1C 6eff" /o 'iipio 10 Greer 107 Lw[m3P IMi#ec.,,o j{eCi/cc '4e bit k8'IgaCi/cc-l , YSC / (G jg3f tot y f t/e Ykl
- f Fuel Handling 3
S/C MI Guser MM INef7H1gh
! Area Exhaust .H mR/hr . . _ /L "
Mobk mRytt ,t,ow
,,Lt CllCC GastS mqk Hqk P(WJICi(CCC5-\l'l[ &h9 Loa
@g,4 c ldt 'b NOF
? -
f e !bt-godsh et Ckt.veul h' Htv T. -
4 g, y , go ...- -' Cs-\51
_- g,. g r - *~6I <
- Sensitivity based upon this radionuclide.
" Etpecte:! activities c?e estimated and are based on existingplants.
'~~')
O.
6 11.5 11 Amcadment 16 s
- ll.bl O WP WC (1) 'T A3W 3
! MN nA61xAX Standard Plant nv n
- TABLE 11.5 3 PROCESS RADIATION MONITORING SYSTEM (LIQUID MONITORS)
- Opa .l Principal l
Radladoo Conngv. bjMd4 0de.LL Radlonudides Erpected Alarms hionitor -Bha 7]! ARAC Measured Aethit 7 & Trips
! gal 04 [gn,, hn$ n[ -7 1 a /d'h 4 i Radwaste effluent kline Omeme- 10 d1@ Cs D7' s Hightbewl % .
. radiaden monitor are Co 60 C (( Hi ls' p C fec Lea got ;
! (p* pe cc .r o + 4 r /0' I'* I L Reactor building Inline Gemme. 10W10s Cs D7' 1bles!DPw cooling water system f adladon 1100 Saint open 4O/[CC Co 60 .# / (c [4High[U@@
L*J i
monitor CrmfAl0**
pe e CC ,g.2 4 jy g Drywell 5 ump Inline fe W Gross u@ x *0 2 High Highblibe Drain 36%R/y h .p Ganuna , 8 W
I o C. /c c c3-a7 @C//cf
.a High
- Low /INOP 15 elt 4<
Sensitivity be.ted upon this radionuclide.
Expected activities are estimated and are based on aistingplants.
l l
I l
I l
1
-1 i
I l
I i I%
Amendment 16 - 113 15 I
L l
'i
ABWR zwimo:
Standard Plant try g TABLE 11.5 4 r l RADIOLOGICAL ANALYSIS
SUMMARY
OF LIQUID PROCESS SAMPLES Grab Sample SensitMty ,
Samole DeseHorlon Frenuenev Analysis gGbal Purnose i
- 1. Reactor Coolant Filtrate Dauy(a) Gross gamma 104 Enluate reactor water ae-ddy .
l Crud Dauy (8) Grou gamma 104 Enluate crud acdvity Filtrate Weekly (b) 1 131,1 133 10*7 Evaluate fuel cladding -j integrity Crud and Strate Weekly Gamma spectrum $ a 10*7 Determine radionuclides <
. present in system
- 2. Reactor water deanup Biweekly Grou gamma 104 Enluate cleanup i system effielency
- 3. Condenser demineralizer Influent Monthly Gross gamma 104 Evaluate Effluent Monthly Gross gamma '104 Evaluate demineralizer ,
performance - l Eul.u,4c Ak vaaliencSwik
- 4. Condensate storage tank Weekly Grcu S y 104 Td buri Y; O' 5. Fuel pool Gter -
l demineralizer !
Inlet and outlet Periodicauy Gross B y 104 Enluate system perform.
ance LC W st
- 6. %Insee couector Periodically Gross $.y 104 Evaluate system performe -l 3 tanks (4)
! ance 1 l MCv)
- 7. Mao,% collector PeriodicaUy Gross p y 104 Evaluate system perform.
ance j tanks (2)
- HSD sampic :
- 8. Ghemsembraste tank s (2)- Periodically Gross A 7 104' Evaluate sptem perform- I ance, !
' 1
]
f 4'
, i l
l .;
l l
Amendment 16 11.5 19 l
a l.
ABM 2xtioarx nrv s Standard Plant TABLE 11.5 4 RADIOLOGICAL ANALYSIS SUhth1ARY OF LIQUID PROCESS SAh!PLES (Continued)
Grab Sample Sent!tivity Samele DeseHetfon frecu rney Analysts E.CMml Purnose Setsd wash Het %< e.
- 9. (Evaporator bottoms)$ PeriodicaDy Grou / 7 104 Comparimausf aethity with that determined by drum readings H C VJ
- 10. F ;- "-- distillate bK Periodically Gross 4 v 104 Evaluate evaporator pet.
tanks 46) (bayeva Q formance
~b id s ng t, L II D[gh *d I'*S W eci4 l. G rosJg - Y 10 En lue.l c.
) L c4 Re4<.
(0) Daily meansfive timesper week.
Ib) Performed more frequently ifincrease noted on daily gamma count.
TABLE 11.5 5 RADIOLOGICAL ANALYSIS SUhth!ARY OF GASEOUS PROCESS SAh!PLES Sample Sensitbity Samole Descrintion Frecuenev Analvsts afJ/.ml Purnese
- 1. Containment atmosphere Periodically Gross a & 4 10 11 Determine need for respi-(dr>weu) and prior to Tritium 10' 0 ratory equipment entry
- 2. Offgas monitor sample WeeUy Gamma spectrum 10 10 Determine offgas aethiry
- 3. Off 6as vent sample WeeUy Gross 6(a) 10 11 Determine offgas system I.131(b) 30 10 cleanup Gamma speetrum 10 10
(*) On paniculatefilter.
(b) On charcoalcanridge.
I Amendmem 16 11520'
m6 tow;
/ MN Standard Plant m'A
)
TABLE 11.5 6 RADIOLOGICAL ANALYSIS
SUMMARY
OF LIQUID EFFLUENT SAMPLES s
Sample Sensitivity Analysis gCMal Purnost Er.mnle Descrintion Freauenev Gross p' 6e.m=4 10-7 Effluent discharge record
- 1. Detergent drain tanks Batch (a)
=
C- : .i 4
5 x 10 7 Efnuent discharge record Liquid radwaste Weekly (b) Ba/La 140 and
! 2.
I.131 j effluent Monthly Gamma spectrum 5 x 10 7 - '
! Composite of all Tritium 10 5 tanks discharged e
Gross alpha 10 7 >
i Dissolved gas (C) 10 5 ,
{
Quarterly Sr/89/90 FY 10-8 Grossg Gamma 10 7 Effluent discharge record
- 3. Circulating water Weekly grab 3
Tritium 10 5 (backup sample)
- decant line of continuously collected proportional sample G
U Gves GA--M. 10 EGw.J Di3ch.7 pcog
( 1.5.1 A Cac44 S(Mc' W*Kly TW N m lo
~d
[q) W4%
(0) if tank is to be discharged, analyses will be performed on each batch. If tank is not to be discharged, analyses will be performedpen*odically to evaluate equipment perfonnance.
l (b) Typical batch of average release. All other samples are proponionalcomposites.
(c) If no discharge event occurs during the week, frequency shall be so adjusted. .
i e
f L/
, 11.5-21 Amendment 2 l ..
r = .-,.,e-. y- ,-
--ww -
-%+
MN 23A6tocAK Standard Plant krv n Q
o TABLE 11.5 7 RADIOLOGICAL ANALYSIS
SUMMARY
OF GASEOUS EFF JENT SAMPLES Sample Sensitivity Samnie Descrintion Freauency Analysis uCl/mi Purnose
?ls.4 AT- Eids,g h p,5,i 1. R;e;;ca E '.!?:..;;d.e ; Weekly Gross B (a) 10 11 Emuent record g4) % Tmt0 s rseg 4 I 131(b) w#, 10 10 M3 10~i Ba/La 140 Monthly Gamma spectrum (a) 10 10 Quarterly St 89 and 90(a) 10 11 '
Gross alpha (8) 10 11 433 and 13W 10*1T['
lum 106 j ,
4 Dis:c bm;JL , dem; /.;et~..-
^ < deva E";;;:teceed- . L 2.
P Gland Steam Condenser As above As above Emuent record
- d.......b.,-df;s c.< h 4 s tkwp
.O (0) On particulatefliter. .
(b) On charcoalcartridge.
- This m c, k.j e s d beu 5 h5 re k
- E'" d ' M ,
'e C4 - { G f m geu,a so,u ia n L.w.y , e. aeon soug y ,
Q.) '
i Amendment 16 _ g1,$,22
.