ML20077A495
| ML20077A495 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 05/03/1991 |
| From: | Marriott P GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| EEN-9134, MFN-043-91, MFN-43-91, NUDOCS 9105100147 | |
| Download: ML20077A495 (57) | |
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GE Nuclear Errergy hiay 3,1991 hiFN No.043 91 Docket No. STN 50 605 EEN 9134 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Charles L. hiiller, Director Standardi7ation and Non Power Reactor Project Directorate
Subject:
GE Responses to the Discussion items of November 15, 1990, GE/NRC Radiation Protection liranch Conference Call Enclosed are thirty four (34) copies of a draft amendment to Chapter 12 of the ABWR SSAR, which addresses all of the subject discussion items.
It is intended that GE will amend the SSAR with these changes in a future amendment.
Sincerely,
/
P. W. hiarriott, hianager Regulatory and Analysis Services hi/C 382, (408) 925 6948 cc: F. A. Ross (DOE)
D. C. Scaletti (NRC)
R. L Pedersen (NRC)
D. R. Wilkins (GE)
J. F. Quirk (GE) 1 9105100147 910503 fDR f ADOCK0500g5 Doy i Y
ABWR :w - !
SIRIldard l'lnnt pvn have been redesigned as a result of (1) locating equipment, instruments, and inservice testing, sampling stations, which require routine maintenance, calibration, operation, or 12.1.2.2 3 Equipment Design Considerations to inspection, for case of access and minimum Limit Component Itadiation locls required occupancy time in radiation areas; (1) Equipment and piping were designed to reduce (2) laying out plant areas to allow remote or the accumulation of radioactive materials in mechanical operation, service, monitoring, the equipment. The piping, where possible, or inspection of highly radioactive was constructed of seamless pipe as a means equipment; and to reduce radiation accumulation on the scam. The filter demineralizers in the RWCS (3) providing, where practleable, for and FPCS are backwashed and flushed prior to transportation of equipment or components maintenance. requiring service to a lower radiation area.
(2) Equipment designs include provisions for limiting leaks or controlling the fluid that 12.1.2 3.2 blinimizing Radiation 1cels in does leak. This includes piping the Plant Access Artm und Vicinity of tqulpment released fluid to tne sumps and the use of ,
drip pans with drains piped to the floor Facility general design considerations draine directed toward minimizing radiation levels in plant access areas and in the vicinity of (3) The materials selected for use in the equipment requiring personnel attention include primary coolant system consist mainly of the following:
austenitic stainless steel, carbon steel and low alloy steel components. (1) separating radiation sources and occupied areas where practicable (e.g;, pipes or (4) The system design includes a RWCS and a ducts containing potentially high condeasate demineralizer system on the radioactive fluids not passing through n reactor feedwater. These systems are occupied areas; G designed to limit the radioactive isotopes in the coolant. (2) providing adequate shiciding between radiation sources and access and service (5) External recirculation pumps and areas. Of special note, the reactor recirculation piping were replaced by pressure vessel shield wall in the upper p internally mounted recirculation pumps. drywell extends to within four inches of
- Such pumps can be removed easily as an the upper drywell ceiling thus permitting integral or package unit for maintenance continued operation in the upper drywell outside the lower drywell radiation zone, during refueling and providing shellding in the case of a refueling accident; 12.1.23 Facility Layout General Design Considerations for hlaintaining Radiation (3) locating equipment, instruments, and Exposurn ^ f ' " A sampling sites in the lowest practicable radiation zone; 12.1.2 3.1 Slinimiting Personnel Time Spent in Radiation Areas (4) providing central control panels to permit remote operation of all essential Facility general design considerations to instrumentation and controls from the minimize the amount of personnel time spent in lowest radiation zone practicable; radiation areas include the following:
l l
Amendment 1213
AllWR m am^t iuv n Standanil'lant (5) where practicable for package units, separating highly radioactive equipment from less radionetite equipment, instruments, and controls; (6) providing means and adequate space for utilizing moveable shielding for sources within the service area when required; (7) providing rueans to control contamination and to facilitate decontamination of potentially contaminated areas where practicable; (8) providing means for decontamination of service areas; Amendment t214t l
ABWR nmont.
Mallilald]!llin! IWV lI 1
(9) providing space for pumps and valves outside of highly radioactive areas; (10) providing remotely operated centrifugal discharge and/or backflushable filter systems for highly radioactive rudwaste and cleanup systems; (11) providing labyrinth entrances to radioactive pump. equipment, and valve rooms; (12) providing adequate space in labyrinth l
entrances for easy access; (13) maintaining ventilation air flow patterns from areas of lower radioactivity to areas l of higher radioactivity; and (14) providing both automatic logic control and mechanical stop devices for control of the transverse in core (TIP) probe to prevent withdrawl of the radioactive portions of the T!P onto the cab!c spoolers.
12.1.3 OperallonalConsiderations Out of AllWR Standard Plant scope. See Subsection 12.1.4.3 for interface requirement.
12.lA Interfaces 12.1.4.1 Regulatory Guide 8.10 Compliance with Regulatory Guide 8.10 shall be demonstrated (See Subsection 12.1.1.3.2),
12.1.4.2 Regulatory Guide 1.8 Compliance with Regulatory tiuide 1.8 shall be demonstrated (See Subsection 12.1.1.3.3).
12.1.4.3 Occupational Radiation Exposures Applicants referencing the AllWR design will provide the criteria and/or conditions under which varicus operating procedures and techniques shall be provided to ensure that occupational radiation exposures one AIARA are implemented (See Subsection 12.1.3).
Amendment 12.1-4
AllWIl 234<.uort.
Slanilanil'[ ant twv, ri SECTION 12.2 CONTENTS Sec1hm lille Page 12.2.1 ContitlanLSources 12.2 1 12.2.1,1 Source Terms 12.2-1 12.2.1.2 Reactor, Radwaste, and Turbine Building Sources 12.2 1 12.2.1.2.1 Reactor Vessel Soureca 12.2 1 12.2.1.2.1.1 Radiation from the Reactor Core 12.2-1 12.2.1.2.1.1.1 Oeneral 12.2-1 12.2.1.2.1.1.2 Physical Data 12.2 1 12.2.1.2,1,1 3 Core Boundary Neutron Fluxes 12.2 2 12.2.1.2.1.1.4 Gamma Ray Source Energy Spectra 12.2 2 12.2.1.2.1.1.5 Gamma Ray and Neutron Fluxes Outside the Vessel 12.2-2 12.2.1.2.1.1.6 Deleted 12.2 2 12.2.1,2.2 Radioactive Sources in the Reactor Water, Steam and Offgas 12.2 3 12.2.1.2.3 Radioactive Sources in the 1IPCF System and the LPFL Mode of the RilR System 12.2 3 12.2.1.2.4 Radioactive Sources in the Reactor Shutdown Mode of the ResidualIIcat Remova! System 12.2 3 12.2.1.2.5 Radioactive Sources in Reactor Core Isolation Cooling System 12.2 3 12.2.1.2.6 Radioactive Sources in Radwaste Systems 12.2 3 12.2.1.2.6.1 Radioactive Sources in the Reactor Water Cleanup System 12.2 3 12.2.1.2.6.2 Radioactive Sources in Liquid Radwaste System 12.2 4 12.2 il Amendment
ABWR umma SinndardflanL RIXJJ SECTION 12.2 CONTENTS (Continued)
Secllon Tilk Page 12.2.1.2.63 Radioactive Sources in the Gaseous Radwaste System 12.2 4 12.2.1.2.6.4 Radioactive Sources in the Solid Radwaste System 12.2 4 12.2.1.2.6.5 Radioactive Sources in the Fuel Pool Cleanup System 12.2-4 12.2.1.2.6.6 Radioactive Sources in the Suppression Pool Cleanup System 12.2 4 12.2.1.2.7 Radioactive Sources in the Piping and hiain Steam Systems 12.2 4 12.2.1.2.7.1 Radioactive Sources in the hiain Steam System 12.2 4 12.2.1.2.7.2 Radioactive Crud in Piping and Steam Systems 12.2 4 12 2.1.2.8 Radioactive Sources in the Spent Fuel 12.2 4 12.2.1.2.9 Other Radioactive Sources 12.2-4.1 12.2.1.2.9.1 Reactor Stattup Sourcc 12.2-4.1 12.2.1.2.9.2 Radioactive Sources in the Control Rod Drive System 12.2-4.1 12.2.1.2.9 3 Radioactivity in the Transverse in Core Probe 12.2-4.1 12.2.1.2.9.4 Radioactivity in tne Reactor Internal Pumps 12.2 4.1 12.2.1.2.10 Post Accident Radioactive Sources 12.2 4.1 12.2.1 3 Turbine Budding Sources 12.2 4.2 12.2.2 Alt. borne and Llauld Sources for Environmental fansideration 12.2 4.2 12.2.2.1 Production of Airborne Sources 12.2 5 12.2.2.2 Deleted 12.2 5 12.2.23 Airborne Sources During Refueling 12.2-6 12.2.2.4 Annual Average Doses 12.2 6 12.2 iii Amendment
ABWR zw=41.
sum. n Standard Plant SECTION 12,2 CONTENTS (Continued)
Ruhtu Title l' age 12.2.2.5 Liquid Releases 12.2 6 12.2 3 Interfung 12.2 6 12.23.1 10CFR20 and GDC61 Compliance 12.2 6 12.2 3.2 Turbine lluilding Compliance 12.2 6 12.2.4 References 12.2 6.1 12.2 iiia Amendment
ABWR mamt.
niv n Siendard Plant I
SECTION 12.2 TAllLES Table Title l' age 12.2-1 Ba:Je Reactor Data 12.2-7 12.2-2 Core Boundary Neutron Fluxes 12.2 11 12.2 3 Gamma Ray Source Energy Spectra 12.2 12 12.2-4 Gamma Ray and Neutron Fluxes outside the Vessel Wall 12.2 15 12.2 5 Radiation Source 12.2 17 12.2-6 Fission Product Gamma Source Strength in RI!R licat Exchanger 12.2 18 12.2 7 Fission Product inventory in the RilR lleat Exchanger 2 Ilours After Shutdown 12.2 19 12.2 8 Reactor Coolant Concentration Values Entering the RCIC Turbine 12.2 20 12.2 9 CUW Filter Demineralizer 12.2-21 12.2-10 Reactor Water Cleanup, Regenerative llent Exchanger Tube Side (curies) 12.2 22 12.2 11 Reactor Water Cleanup, Non Regenerative lleat Exchanger Tube Side (curies) 12.2 23 12.2-12 Reactor Water Cleanup, Regenerative 1-leat Exchanger Shell Side (curies) 12.2-24 12.2 13 Liquid Radwaste Component Inventories 12.2 25 12.2-14 Offgas SystemInventories 12.2.26 12.2-15 Solid Radwaste Component inventories 12.2 27 12.2-16 FPC Filter Demineralizer 12.2-28 12.2-iv i
Amendment
ABWR mamt.
myn Standard Plant TAllLES (Continued)
Ildde 31 tis Eitge 12,2 17 Radioactive Sources in the Suppression Pool Cleanup System 12.2-29 12.2 18 Rcdioactive Sources in the Control Rod Drive System 12.2-30 12.2-19 Annual Airborne Releases for Offsite Dose Evaluations (Curies) 12.2 31 12.2 20 Comparison of Airborne Concentrations with 10CFR20 Concentration Limits 12.2 32 12.2 21 Average Annual Doses from Airborne Releases 12.2 33 12.2-22 Annual Average Liquid Releases 12.2 33.1 12.2 23 Liquid Pathway Dose Analysis 12.2 33.3 12.2 24 Activity 1.evels of the Transversing in core Probe System 12.2 33.4 12.2-25 Activity Levels in the Reactor Internal Pump 12.2 33.5 12.2 26 Activity in the Turbine Moisture Separator / Reheater 12.2 33.6 12.2-27 Acthity in the Turbine Condenser 12.2 33.7 12.2 28 Acthity in the Condenser Demineralizer 12.2 33.8 ILLUSTRATIONS Elgurs Ittic Enige 12.2-1 Radiation Source Model 12.2-34 12.2 v i
Amendment
ABWR 2mma Standlird Plant iam n 12.21(ADIATION SOURCES were used to develop sources in equipment con-taining reactor water or t.tcatn.
12.2.1 Contained Sources 12.2.1.2 Reactor, Raduaste, and Turbine 12.2.1.1 Source Terms ilullding Sources With the exception of the vessel and drywell The information in this section is divided shields, shielding designs are based on fission into two catagories, the reactor vessel sources product and activation product sources consis. (Subsection 12.2.1.2.1) and the sources from the tent with Section 11.1. For shiciding, it is remaining areas (Subsections 12.2.1.2.2 through conservative to design for fission product sour- 12.2.1.2.9) . Included in these areas arc the ces at peak values rather than an annual average, sources from the radwaste building (Subsection even though experience supports a lower annual 12.2.1.2.6) and the turbine building (Subsection average than the design average (Reference 1). 12.2.2.3). Table 12.2 5 presents a listing of it should be noted that activation products, the sources, excluding the reactor vessel, principally Nitrogen 16, control shielding calcu- including a reference to the source location on lations in most of the primary system. In areas the arrangement drawings in Section 12.3.
where fission products are significant, conserva-tive allowance is made for transient decay while 11 the same time providing for transient increase 12.2.1.2.1 Reactor Vessel Sources of the noble gas source, daughter product forma-tion and energy level of emission. Areas where 12.2.1.2.1.1 Radiation from the Reactor Core fission products are significant relative to Nitrogen 16 include: (1) the condenser off gas 12.2.1.2.1.1.1 General system downstream of the steam jet air ejector; (2) liquid and solid radwaste equipment; (3) The information in this section defines a re-portions of the RWCS; and (4) portions of the actor vessel model and the associated gamma and feedwater system downstream of the hotwell neutron radiation sources. This section is de-including condensate treatment equipment, signed to provide the data required or calcula-tions beyond the vessel. The data selected were For application, the design sources are not chosen for any given program, but were cho-grouped first by location and then by equipment sen to provide information for any of several type (e.g., reactor building, core sources). The shield program types. In addition to the source following paragraphs represent the source data in data, calculated radiation dose levels are various pieces of equipment throughout the plant. provided at locations surrounding the vessel.
General locations of equipment are shown in the These data are given as a potential check point general plant arrangement drawings of the Section for calculations by shield designers. ,
1.2. Specific Acceptance Criterion 11.6 of See-tion 12.2 provides that in addition to the loca- 12.2.1.2.1.1.2 Physical Data tion of contained sources, their approximate size and shape be shown. Though this has not always Table 12.21 presents the physical data re-been included, the source strength or concentra- quired to form the model in Figure 12.21. This tion has been provided in Chapter 12 tables and model was selected to contain as few separate detailed geometry has been provided in Table regions as possible to adequately portray the 12.21 for the reactor and in Chapter 5 for the reactor. Table 12.21 provides nominal dimen-main steam. sions and material volume fractions for each boundary and region in the reactor model. To describe the reactor core, Table 12.21 provides thermal power, power density, core dimensions, core average material volume factions and reactor power distributions. The reactor power distributions are given for both radial and In Chapter 12 the reactor water concentrations axial distributions. These data contain Amendment 12.2 1
ABWR mm StantlanU'htnl _
iam n j carried out for a mean element and appropriate is maintained from the pump into the pressuie decay time. sessel to minimite contamination of the lower pump housing and components. A complete 12.2.1.2.9 Other Radioacthe Sources description of the internal pump is given in Subsection 5.4.1. Contamination of the pump 12.2.1.2.9.1 Reactor Startup Source nevertheless occurs primarily on the upper impeller and components into the lower pump The reactor startup source is shipped to the housing. Table 12.2 25 presents the expected site in a special cask designed with shielding. levels of contamination based upon operating The source is transferred under water while in experience.
the cask and loaded into beryllium containers.
This is then loaded into the reactor while re- 12.2.1.2.10 Post Accident Padioactive Sources maining under water. The source remains within the reactor for it lifetime. Thus, no unique The AllWR general design criteria limits shielding requirements are required af ter potential radiation exposure from accidents both reactor operation. to plant personnel and to the public by the use of containment and treatment of accident 12.2.1.2.9.2 Radioacthe Sources in the sources. The following describes those features Control Rod Drhe System of the AllWR germane to post accident radiation sources in the primary containment, reactor The control rod drive (CRD) source term data building, radwaste building, and the turbine are provided in Table 12.218. The system is building.
described in Subsection 3.9.4.
The primary containment is an inerted steel 12.2.1.2.9.3 Radioacthity in the Transverse lined pressure boundary capable of containing in. core probe all accidents sources with minimal leakage to the environment or other plant urcas.
The transverse in-core probe (Tip) system Sufficient redundancy in the ECCS and spray consists of a probe and a stainless steel cable systems exist to insure within a reasonable which is run into and out of the core such that probability that this primary boundary will not the probe and up to 12 feet of cable are execed deign criteria. In the case of a activated. The probe is desciibed in Subsection degraded core event additional passive features 7.7.1.6.1 and is automatically controlled and such as the suppression pool and passive flooder indexed to its in-core position. For system have been incorporated to flood the maintenance, the probe is manually withdrawn into containment and scrub airborne fission a shielded assembly area in which a shielded products. Therefore, for all but the most container is used to hold the probe. Iloth improbable accident scenarios, radioactive automatic logic control and mechanical stops sources from the pressure vessel will be prevent the probe and activated sections of the contained in the primary containment.
cable from withdrawl beyond the shielded room and containct. Table 12.2 24 describes the levels of With respect to the reactor building, the radioactivity expected from the probe and cable. overall plant design has divided the reactor Since there are two specific types of probes, a building into three separate and independent neutron and a gamma, both types are described in divisions. ECCS components are contained in Table 12.2-24. cach division in separate isolated rooms such that the failure of one system in one division 12.2.1.2.9.4 Radioactivity in the Reactor will not affect in any way components in another Internal Pumps division. Releases of radioactive material either in the form of water or steam (airborne)
The reactor internal pumps, RIP, are located are contained in and isolated to a large extent on the lower exterior portion of the pressure in the compartment in which it might occur by vessel and connect to an impeller located in the the use of water tight doors and area radiation pressure vessel. A constant flow of clean water monitors which isolate the HVAC system from the Amendment 12.241
ABWR 2miut.
SlitudlinlI'lant 1uv u compartment. Divisional separation under such applicant as called out in Suosection 12.2.4.
conditions is complete. Sumps are designed to Tables 12.2 26 through 12.2 28 provide estimates detect and alarm in the event of leaks in excess of inventories for the moisture separator, of one gallon per minute establishing a threshold condenser, and condenser demineralizer. The for Icht before break on the larger water offgas system which is vaulted in a two meter carrying piping systems. All connections to the concrete vault in the turbine building is primary containment not terminating in the described in Sub.Section 12.2.1.2.6.3.
reactor building meet GDC54,55,56, and 57.
Therefore,in the event of an accident involving 12.2.2 Airborne and 1.ltluid Sources for radioactive sources in the primary containment or F.nsironmentill Consideration reactor building such sources would be contained and isolated for further treatment and This Subsection deals with the sources and decontamination, parameters required to evaluate airborne and liquid releases during normal plant operations k 1.ikewise potential releases in the radwaste for compliance with 10CFR20 criteria building will be contained by isolating the radwaste building atmosphere and scaling a water releases in the building which is scismically qualified and steel lined to prevent any potential water releases. Such potential I
releases are discussed in Section 15.7.
The turbine building contains no major sources of releasable radioactivity (discounting N 16 because of the 7.7 second half life) and potential releases are limited to liquid releases of low activity water from the feedwater and condenser system. Two other sources exist which contain radioactive species but in a form not amenabic for release. The potential for accident sources from these two sources, the offgas system and condenser demineralizers, is reduced due to heavy shielding and compartmentalirir.g these components.
Estimates on sources and location for limiting design basis events are found in Chapter 15 and sources for degraded core events as a function of probability are found in Chapter 10.
12.2.1.3 Turbine llullding Sources Turbine building sources are primarily dominated by N 16 in the steam flow from the pressure vessel. This N 16 source results in significant gamma shine from the main steam lines (on the order of 1.5 2.0 Rad /hr contact), the turbines, moisture separators, and reheaters.
Eatimates of typical BWR sources and gamma shine are given in Reference 11. Since the geometry of the radiation source is dependent on the exact turbine conCguration used, the specific details for the turbines and turbine reheaters are interface requirements for the referencing Amendment 12.2-t2
1 ABWR mmu Slalldard i'lant RIV U The altborne radiological releases from 12.2.2.1 Production of Airborne Sources building heating, ventilating, and air conditioning and the main condenser mechanical Design efforts are directed towards keeping vacuum pump have been compiled and evaluated in contained all the radioactive material, whether References 3 and 5.
it is in a solid, liquid or gaseous form; however, the unavoidable leaks from proces.s Based upon the above conditions and values in systems and some processes in refueling and References 2 and 4, airborne releases to the decontamination lead to airborne radioactivity. environment are summarized in Table 12.2 21.
Leakage of fluids from the .ocess system will Approximately 5,100 Ci/ plant /yr of noble result in the release of r .* >nuclides into plant radiogases are released; one half of this total buildings. In general, the noble radiogases will is released from the turbine building. The remain airborne and will be released to the at. total particulate release rate per plant is mosphere with little delay via the building ven- approximately 86 ci/yr; the annual release of tilation exhaust duct. The radionuclides will Co 60 is less than 0.03 Ci.
partition between air and water to approach equilibrium conditions. Airborne iodines will 12.2.2.2 Deleted
'platcout" on most surfaces, including pipe, concrete, and paint. A significant amount of radiciodine remains in air or is desorbed from surfaces. Radiolodines are found in ventilation air as methyl lodide and as inorganic iodine which is here defined as particulate, elemental and hypoiodous acid forms of iodine. Particu-lates will also be present in the ventilation exhaust air.
The average annual release of I 131 is given in Table 12.2 20. The basis for these releases is as follows:
i I
(1) a cslendar year consisting of 300 days of power operations and one refueling /
maintenance shutdown period; (2) a concentration 1131 in reactor water of 2.3 pCi/kg; (3) a carryover of I-131 from reactor water to
! steam of 1.5%;
l l (4) forward. pumped heater drains; (5) a noble gas release rate of 15,000 pCi at t - 30 min and an 1131 release rate of 100 l
pCi/sec at t = 0; and 1
(6) 24 drywell purges per year,365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br /> be-tween each purge.
(7) Meteorology as provided in Subsection l 11.3.10.
l l
l 12.2d Amendment i
AIMR :wwoxi.
F1m H Slanila131.fIRni annual doses to unrestricted areas subject to l
I airborne and liquid releases, l'or airborne dose calculations, isotopic release were taken from Table 12.2 20 assuming a 0.5 mile exclusion boundary. Releases were assumed from plant stack since all major (reactor building, turbine building and radwaste building) ventilation systems pipe to the stack for normal releases.
Since a site meteorology is not definitively defined, astatistical approach was used to evaluate the releases over a series of metrologics discussed in lieferences 6 and 7.
12.2.2.3 Airborne Sources During liefueling Doses ' vere calculated using methodologies and conversion factors consistent with Regulatory The airborne radioactivity during refueling in Guides 1.109 and 1.111 as implemented in the containment is expected to be similar to that References 8 and 9. The results of the airborne observed in operating stations. Experience at evaluations is given in Table 12.2-21. For the operating HWR has shown that airborne ingestion doses given in Table 12.2 21, radioactivity can result from fhe water in the ingestion values given in Tabic E 5 of reactor cavity execeding 100 F and flaking of Ilegulatory Guide 1.109 were used.
cobalt dioxide (coo,) from the dryer and separator if their surfaces are allowed to dry.
Other potential airborne sources could occur during vessel head venting and fuel movement. 12.2.2.51.lquid Releases The airborne radioactive material sources resulting from reactor vessel head and internals The AllWR is designed not to release removal have been determined from operating plant radioactive liquid eff!uents, llowever, under experience. The major radioisotopes found were certain conditions of high water inventory, up 1131, Co 60, and hin-54, with Nb 95, Zr 95, to 0.1 Curie per year excluding tritium may be Ru 103, and Cc 144 at moderate concentrations, released as described in Subsection 11.2.3.
and with Cc 141, Cs 137, Co 58, and Cr 51 at low These releases are given in Tabic 12.2 22 and concentrations. The radlogetive particulates form the basis for estimating doses using ranged as high as 2 x ig Ci/cc and the methodologies consistent with Regulatory Guide 1131 as high as 4 x 10' Ci/cc. 1.113 as implemented in Reference 10. The results of the liquid release, assuming To minimire the containment airborne dillution factors described in Subsection radioactivity contribution due to removal of the 11.2.3.2, are shown in the dose evaluation in reactor pressure vessel head: Tabic 12.2 23.
(1) the steam dryer and separator surfaces will 12.2.3 Interfaces bc kep! wel or covered; 12.23.1 10CFD20 and GDC61 Compliance (2) the fuct pools are cooled through heat exchangers of large capacity; and the Applicants rcIctencing the ABWR design will provide source tabics and operational criteria (3) ventilation system on the refueling poolis to insure compliance with respect to worker designed to sweep air from the pool surface restrictions of 10CFR20 and GDC61, and remove a large portion of potential airborne contamination. 12.23.2 Turbine Hullding Compliance 12.2.2.4 A)erage Annual Doses Applicants referencing the ADWR design will provide source tabics for the turbine and For compliance with 10CFR50, Appendix 1, turbine reheater, and will provide gamma sh;ne evaluations have been made to determine average calculations for the turbine complex to insure Amendment 12.2 4
ABWR urumri.
Standard Plant nw n turbine building gamma shine both offsite and to surrounding buildings is within applicable limits.
12.2.4 References
- 1. J.E. Smith, Noble Gas E.rperience in Boil-ing li'ater Reactors, Paper No. A 54, presented at Noble Gases Symposium, Las Vegas, Nevada, September 24, 1974.
- 2. Airborne Releases from B1l'Rs for Environ-mentalImpact Evaluations, NEDO-21159 2 l (1977).
- 3. American Nuclear Society, ANS-18.1, Table 5.
- 4. Airborne Releases from Bil'Rs for Enytron-mentalImpact Evaluations, NEDO 21159 March 1976.
- 5. Calculation of Releases of Radioactive Materials in Gascous and Liquid Effluents from Boiling if'aterReactors (llWR-GALE Codc)
U.S. NRC NUR EG-0016 Rev.1, January 1979.
- 6. 1. IIall, et al, Generation of Typical Meteorological Years for 26 SOLMET Stations, l Sandia National Laboratory Report S A N D78 1601 (1978).
- 7. D.C. Aldrich, et al, Technical Guidance for Siting Critena Development, NUREGlCR 2239 (1981).
- 8. E.W. Dradley, Gamma and Beta Dose to Man from Noble Gas Release to the Atmosphere GEMAN Code. NEDO-25132A, April, .1980.
- 9. E.W. Brr,dley and V.D. Nguyen, Radiation Erposurefrom Airborne Effluents the REFAE code, NEDO-25257, July,1980.
- 10. P.P. Standcavage and D.G. Abbott, Liquid Discharge Doses LIDSR Code, NEDM 20609 01, Aug,1976.
- 11. Rogers, D.R., Bil'R Turbine Equipment N-16 Radiation Shiciding Studies, GE NEDO 20206, December 1973.
Amendment 12.2-6.1
l AInVR - m6iwAt StandanLElnot Rev. ii Table 12.2 S Radiation Sources Source Table E2I Drawing lwallan Approximate Geometry 1216 RilR lleat Ihchanger 1211 (R1,RI') Rt Cylndr (r = 0.9m,1 = 7m)
(R6,RA)
(R6,RF) 1218 RCIC Turbine 1231 (R6,RC) Rt Cylndr (r = .5m,1= 0.7m) 12.2-9 CUW Miter Demmeralizer 1213 (R2,RD) 2 Tanks, Rt Cylndr(r 6m,1=3.3m) 12.2-10 RWCU Regen llcat thchanger 1232 (R1,RC) Rt Cylndt (r = .4m,1 = 6.8m) 12111 RWCU Non-llegen llcat Ihchanger ~123-1 (RI,RC) Rt Cylndt (r = .4m,1 = 5.5m) 12113.1 LCW Collector Tank 12 S37 rr1317 2 Tanks, Rt Cylndr (r = 4.m 1-9.4m) 12.2 .3.2 LCW Filter 12S 39 frEM 12 Rt Cylndr (r= 5m,1= 2.5m) 12.2 133 LCW Dcmineralizer 12S39 TDSt 11 Rt Cylndt (r".6m,1= 2.8m) 12113.4 LCW Sample Tank 12S38 frEM B 2 Tanks, Rt Cylndt (r = 4.m,1 = 9.4m) 12.2 13.5 IICW Collector Tank 123-37 Inst 13 Rt Cylndr (r = 2.2m,1 = 43m) 12 S 13.6 1ICW Dcmintralizer 12139 TFISI20 Rt Cylndr (r = .6m,1 = 2.Sm) 12114 Offgas 12S50 (FF,T2) Tank 1, Rt Cylnd r (r = .6m,1 = 7.6m)
Tanks 2 9, Rt Cylndt (r = 1.im,1 = 7.6m) 12.2 15.1 CUW Dackwash Receiving Tank 12 3-1 (R2,RD) Rt Cylndt (r = 2.2m,1 = 5.7m) 12.2-15.2 CF Dackwash Reeciving Tank 12 S 49 (ID,T4) Rt Cylndr (r = 2.2m,1= 5.7m) 124 153 Phase Separator 12S38 TilSt 30 2 Tanks, Rt Cylndr (r = 2.4m,1 = 6.0m) 12115.4 Spent Resin Storage Tank 12S38 THSt 31 Rt Cylndt (t = 2.0m,1 = 5.7m) 12.2 15.5 Concentrated Waste'."ank 12S37 DISt 35 Rt Cylnd: (r = 1.5m,1= 4.4m) 12115.6 Sol Dryer Feed Tank 12all ((13139 Rt Cylndr (r = 1.6m,1 = 3.2m) 12 S 15.7 Sol Dryer (outlet) 12339 ITEM 55 Rt Cylndr (r = 0.2m 1 = 3.2m) 12.2-15.8 Sol Peletizer 123-3' TrEM 58 Rt Cylnd r (t = 0.4m,1= 2.5m) 12.2 15.9 Sol Mist Separator (steam) 12S39 inst 56 Rt Cylndr (r = 0.lm,1= 2.8m) 12 1 15.10 Sol Condenser ,2.3 40 r111157 Rt Cylndt (r = 0.2m,1 = 1.4 m) 12.2-15.11 SolDrdm 123-39 (2,D) Rt Cylndr (t = 03m,1- 0,8m)
Dox (L5nul.5mxim) 12S16 ITC Filter Demineralizer 12S3 (R2,RD) Rt Cytndr (r = 0.7m,1 = 3,4m) 12 S 17 Suppression Pool Cleanup System 12 3-3 (R2,RA) Rt Cylndt (r = 0.7m,1 = 3.4m) 12.2 18 Control Rod Drive System 12S2 (R4,RF) Distributed Sourec 12S24 Transverse Incore Probe 1232 (R4,RD) Distributed Source 12.2 25 Reactor Internal Pumps 1.2 3b TMSL 3000 Distributed Source 12.2-26 Turbine Mositure Separators 123-52 (T6llTI) Distributed Source 12127 Turbine Condenser 12S$3 TD TG Distributed Source 12.2 28 Condenser Filter Dcmineralizer Fi't:r 123-31 (TC-T2) 3 Tanks, Rt Cylndt (r = 1.4m,1 = 6.lm)
Dcmineralizer 12151 (TClD) 6 Tanks, Rt Cytndr(r = 1.7m,1 = 5.lm)
Applicant Spent Fuel Storage 1215 (R4,RF) See Drawings Arrendment 12.2 17
u
- ABWR -- meim Sla.tulanLPl.aiit nmn Table 12.2 20 COMPARISON OF AIRBORNE CONCENTRATIONS WITil 10CFR20 CONENTRATIONS LIMITS Maximum Technical Annual Averane Airborne Specification Release 10CFR20 Nuclide - (Clhr) Concent7)
(pCi/cm tion Concent7)
(pCl/cm MPC tion Kr-83m 8.4E 04 53E 17 1,4E-15 3E 08 Kr-85m 2.1E + 01 13E 12 1.9E-11 1E 07-Kr-85 5.7E+02 3.6E 11 3.6B-11 3E-07 Kr-87 2.5E + 01 1.6E 12 4.2E 11 2E-08 Kr-88 3.8E4 01 2,4E 12 6.4E 11- 2E 08 Kr-89 - 2.4E 4 02 1.5E 11 4.1E 10 3E-08 Kr-90 33E-04 2.1E 17 5.6E-16 3E 08 Xc-131m 5.0E+01 J 3.2E 12 - 3.2E 12 4E 07 Xc-133m - 8.7E 02 5.5E-15 8.8E-15 3E 07 Xc-133 - 2.4 E + 03 1.5E 10 1.1E-09 3E 07 Xc 135m 4.0E+02 2.5E-11 6.6E-10 3E 08
~ Xe-135 4.5E+ 02 ' 2.3E 11 7.5E;10 1E 07 Xe-137 - 5.0E+02 3.2E 11 8.5E-10 3E 08 Xe-138 - 43E+02 2.7E 11 6.8E-10 3E-08 .
Xc139 4.1E 04 2.6E-17 7.0E-16 3E 08 I131- 2.6E 01 ' 1.6E 14 4.9E 13 1E 10 1132 2.2E + 00 1.4E 13 4.2E-12 3E-10
-1133 :1.7E+00 - 1.1E-13 33E 12 4E 10 1134 3.BE + 00 2.4E 13 7.2E 17 6E-10 1135 -2.4E+00 1.5E-13 4.6E 12 1E 09 .
II 3 6.7E + 01 ' 4.2E 12 4.2E-12 2E 07 C-14 -- 93E+00 5.9E 13
- 5.9E 13 - 1E 07 Na 24 -4.0E 03 2.5E 16 2.5E-16 SE-09 P 32 _ 9.2E 04 . 5.9E-17 5.9E-17 2E 09
- Ar 41 6.7E+ 00 - 43E-13 43E 13 4E 08 Cr-51 ' 3.5E-02 2.2E 15 5.2E-15 4E-07 Mn 4.9E-03 3.1E-16 8.5E-15 1E-08
< Mn 56 3.6E-03 ' 23E 16 .23E-16 2E-08 Fe-55 6.5E 03 4.1E 16 4.1E 16 3 8-08 Fe-59 6.5E-04 4.1E 17 9.2E-16 SE-09 Co 58 - - 2.4E 03 1.55 23E 15 3E-08 Co-60 1.1E 02 ' 6.8E 16 1.6E 14 1E h Ni 63 6.5E-06 4.1E-19 2E-09 4.1E-19 Cu-64 1.0E-02 63E-16 63E-16 4E-08 Zn 8.1E-03 5.1E 16 13E 14 2E-09 Rb 4.2E 2.6E 18 2.6E 18 - 3E 08
- Sr-89 . 5.6E-03 3.5E 16 9.5E 15 3E 10
. St 90 6.7E-05 4.2E-18 4.4E-17 3E-11 Amendment 12.2 32
ABWR mimat.
tw n S.timdard Plan! _
Table 12.2 20 COMPARISON OF AIRilORNE CONCENTRATIONS WITli 10CliR20 CONENTRATIONS LIMITS Maximum Technical Apnunt Avernce Airborne Speclucation Relense 10CI'R20 Nuclide (Cl/yr) Concentrp)
(pCl/cm tion Concentnition (pCl/cm ) MPC Y-90 4.6E 05 2.9E 18 2.9E 18 3E-09 Sr-91 1.0E 03 6.6E 17 6.6E 17 9E-09 St 92 7.8E-04 5.0E 17 5.0E 17 1E 08 Y-91 2.4E M 1.5 E-17 1.5E 17 1E-0)
Y 92 6.1E 04 3.9E-17 3.9E-17 1E 08 Y 93 1.1E 03 7.1E-17 7.1E 17 SE 09 Zr 95 1.2E-03 7.8E 17 23E-15 1E-09 Nb-95 1.7E-03 1.1E 16 3.1E-15 3E 09 Mo-99 1.5E-02 9.4E-16 2.2E 14 7E 09 Tc-99m 3.1E 04 2.0E 17 2.0E-17 SE-07 Ru 103 4.9 E-04 3.1E-17 7.1E-16 3E-09 Rh 103m 1.1E-04 73E-18 73E 18 2E4M Ru-106 1.9E 05 1.2E 18 1.2E 18 2E-10 Rh-106 1.9E 05 1.2E 18 1.2E-18 3E4E Ag 110m 6.6E 07 4.2E 20 13E-18 3E-10 St>-124 1.7E-N 1.1E 17 33E-16 7E 10 Te-129m 2.2E-N 1.4E-17 1.4E-17 1E-09 Te-131m 7.7E 05 4,9E 18 4.9E-18 6E-09 Te-132 1.9E-05 1.2E-18 1.2E 18 4E 09 Cs-134 1.7E 04 1.1E 17 1.1E 17 4E-13 Cs 136 8.1E-05 5.2E-18 5.2E 18 6E 09 Cs 137 4.7E N 3.0E 17 3.0E-17 SE-10 Cs 138 1.7E 04 1.1E-17 1.1E 17 'lE 08 Ba 140 13E-02 8.4E-16 2.2E 14 1E-09 l a 140 1.8E-03 1.1E 16 1.1E-16 4E 09 Cc-141 8.7E 03 5.5E-16 1.6E-14 SE 09 Cc 144 1.9E-05 1.2E 18 1.2E-18 2E 10 Pr 144 1.9E 05 1.2E 18 1.2E-18 3E-08 W-187 1.9E 04 1.2E-17 1.2E-17 1E-08 Np-239 1.2E 02 7.4E 16 7.4E 16 2E4E Amendment 12.2-El
ABWR uuima nev n Standard Plant ___
Table 12.2 24 ACTIVITY LEVELS OF THE TRANSVERSING IN CORE PROllE SYSTEM Decay Time (day) Rad /hr @ 1 meter Major Isotopes Gamma Probe Sensor 0.00139 5.61 Mo-56, Al-28, Ti 51 0.0417 3.20 Mn 56, Na 24, Ni 65 1.0 0.0133 Mn-56, Ma 24, Cu-64 2.0 0.00384 Na 24, Co-60, Cr 51 Cable 0.00139 53.5 Mn-56, Mg-27, Ni 65 0.0417 41.2 Mn-56. Ni-65, Fe 59 1.0 0,1M Mn !6, Fe-59, Mn 54 2.0 0.018 Fe-59, Mn-54, Cr-51 Neutron Probe Sensor 0.00139 3.382 Mn 56, Al 28,Ti 51 0.0417 2,142 Mn-56, Na 24, Ni 65 2.0 0.00378 Co-60, Na-24, C0-58 Cable 0.00139 45.1 Mn 56, Mg 27, Ni 65 0,0417 34.8 Mn-56, Ni-65, Fe 59 -
1.0 0.091 Mn-56, Fe-59, Mn 54 2.0 0.0189 Fe-59, Mn-54, Co 60 Amendment 12.2 MA
ABWR :wim.e Standard Plant nev u Table 12.2 25 ACTIVITY LEVELS IN Tile ItEACTOR INTEltNAl, l'UMI' compont rd int!
Impeller 4 24 R/hr Upper Motor 4(412(nmR/hr Motor 80 300 mR/hr Lower motor casing 70-500 mR/hr Amendment 12233.5
< ABWR unamic Standard Plant .
nev.n Table 12.2 26 ACTIVITY IN THE TURillNE MolSTURE SEPARATOR /REllEATER hetmn 11111tg Isotopr.s Curles KR 83M 1.7E 03 NA.24 2.7E 03 KR 85M 3.0E 03 : - P 32 - 5.2E-05 KR 85 1.2E 05 CR-51 1.6E 03 KR 87. 9.8E 03 M N 54 1.8E-05
' KR 88 9.6E 03 - MN 56 1.4 E-02 KR 89 6.2E 02 FE 55 2.6E-04 KR 90 1.4E-01 FE-59 7.8E-06 KR 91 - 1.6E 01 CO-58 5.2E 05 XE 131M 9.9E 06 CO 60 1.0E 04 -
XE-133M 1,4E-(M N163 2.6E 07 XE 133 : 4.2E 03 CU 64 7.9E-03 XE-135M 1.3E 02 ZN 65 5.2E 05 XE 135 / 1.1E-02 SR 89 - 2.6E 05 -
XE 137 7.6E-02 SR 90 1.8E-06 XE 138 - 4.5E-02 Y 90 1.8E-06 XE-139 ' 1.4E-01 SR 91 1.1E-03
- XE-140 '1,5E 01 SR 2.9E 03 XE-144 2.8E 04 Y 91 1.0E 05 Total - 8.3E Y 92 1.7E-03 1
% 93 1.1E 03 1131 1.9E-02 ZR 95 2.1E-06 I132 1.7E 01' NB 95 '2.1E 06 I133 :13E-01 MO-99 5.2E 04 1134 2.8E-01 TC 99M 5.2E-04
- Id35 : 1.8E-01 RU 103 5.2E 06 Total 7.8E 01. Ril 103M 5.2E 06-RU 106 < 7.8E-07.
RB 89 1.7E-03 RH 106- :7.8E 07 ,
CS-134. 7.0E-06~ AG 110M 2.6E .. CS 136 ~ 4.7E TE 129M 1.0E-05
,, CS 137 1.9E-05 TE-131M . 2.6E-05 CS-138 3.2E4T3 TE 132 2.6E 06, Total - 4.9E-03 BA 140 '1.0E-04 LA 140 1.0E 04
.N 16. - 3.9E + 03 CE 141 : 7.8E 06 CE 144l 7.8E-07
, 11 3 ' 7.9E 01 PR 144- -7.8E 07 a W-187 - 7.9E-05 NP-239 2.1E-03 Total 3.7E-02 l ' li; i
(
L Amendment 12.2 33.6
,.. ni
ABWR 2 min Stamlard Plant w Table 12.2 27 ACTIVITY IN THE TURBINE CONDENSER isotopes Curles hoicpn Curles KR-83M 2.6E-01 NA 24 3.9E-03 ER 85M 4.5E 01 P 32 7.7E 05 KR-85 1.8E-03 CR 51 23E-03 KR 87 1.5E + 00 MN 54 2.7E-05 KR 88 1.5E + 00 MN 56 2.1E-0'2 KR 89 7.8E + 00 FE-55 3.8E-N KR 90 7.9E + 00 FE-59 1.1E-05 KR 91 2.2E + 00 CO-58 7.7E-05 XE-131M 1.5E 03 CO-60 1.5E-04 XE-133M 2.2E 02 NI 63 3.8E-07 XE 133 6.4E 01 CU 61 1.2E 02 XE 135M 1.9E + 00 ZN 65 7.7E 05 XE-135 1.7E + 00 SR 89 3.8E 05 XE 137 9.8E + fX) SR-90 2.7E 06 XE-138 6.6E+ 00 Y 90 2.7E 06 XE-139 93E+ 00 SR 91 1.6E 03 XE 140 3.6E + 00 SR 92 43E 03 XE-144 3SE-03 Y 91 1.5E-05 Total 5.5E+ 01 Y 92 2 5E 03 Y-93 1.6E-03 1-131 2.8E-02 ZR 95 3.0E-06 1132 2.4E 01 NB 95 3.0E 06 I133 1.9E-01 M O-99 7.7E M I.134 4.1E 01 TC 99M 7.7E-04 E13$ 2.7E-01 RU 103 7.7E-06 Total 1.1E + 00 Ril 103M 7.7E 06 RU-106 1.1E 06 RB 89 23E 03 Rll-106 1.1E 06 CS 134 1.0E 05 AG 110M - 3.8E-07 CS 136 7.0E 06 TE-129M 1.5E-05 CS-137 2.8E-05 TE-131M 3.8E-05 CSd33 4.7E 03 TE-132 3.8E 06 Total 7.0E 03 BA 140 1.5E-04 LA 140 1.5E 04 N 16 3.8E + 02 CE-141 1.1E 05 CE-144 1.1E-06 11 3 1.2 + 00 PR 144 1.1E 06 W 187 1.2E N NP-239 3.1E-03 Total 5.5E 02 Amendment 12.2 33.7
ABWR uma Sitindard Plant net n Table 12.2 28 ACTIVITY IN TIIE CONDENSElt DEMINEltALIZER Demineralizer l'ilter Deminerallier Filter Isotopes Cudu CUUiu Isotopes C11du furb l-129 1.9E-08 SR 92 1.4 E-01 1-131 6.6E+ 01 Y-91 2.0E-01 3.1E 01 1132 6.8E + 00 Y-91h1 1.1E-01 1 133 4.8E + 01 Y 92 1.4E 01 2.0E 01 1134 4.4 E + 00 Y-93 1.5E-03 3.7E 01 -
I 133 U.il+,21 ZR-93 33E-09 Total 1.5 E + 02 ZR-95 8.6E-N 1.2E-01 NB-95hi 3.4 E-06 4.9 E-04 RB-89 7.2E 03 NB-95 6.6E-N 8.6E 02 CS-134 1.5E + 00 h10 99 1.2E + 00 CS 135 2.0E 05 TC 99hl 6 lE-01 CS136 2.7E 02 TC&) 6.6E4)6 CS 137 6.0E + 00 RU 103 6.7E-04 13E-01 C5-D8 3.1 E-02 Ril 103M 6.7E-04 13E 01 Total 7.6E + 00 RU 106 8.lE-04 3,6E 02 Ril-106 8.1E 04 3.6E-02 NA 24 73E 01 AG-110Ni 2.0E-04 1.1E-02 P 3.2E 01 AG 110 2.7E 06 1.5E 04 CR-51 1.9E + 01 TE 129h1 3.0E 61 hlN 54 1.lE + 00 4.1E-01 TE-129 9.4E 02 h1N 56 33E-01 6.0E-01 TE 13th! 2.8E 02 FE-55 4.4 E-01 1.2E4 01 TE-131 3.1E 03 FE-59 1.2E 03 2.1E 01 TE-132 3.7E 03 CO 58 8.1E-01 8,6E 01 BA 137h1 5.7E + 00 CO.60 1.4 E + 01 2.5 E + 00 BA-140 5.7E 01 Ni 63 43E-02 6.5E-03 LA-140 5.7E 01 1.0E + 00 CU-64 1.8E + 00 CE-141 8.2E-N 1.7E 01 ZN 65 53E+ 00 CE 144 13E-03 7.0E-02 SR-89 5.9E.01 PR 144N1 9.6E-06 5.0E N SR-90 5.8E-01 PR-144 13E 03 7.0E 02 Y 90 5.8E 01 W-187 2.6E-04 63E 02 SR-91 3.8E-01 NP 239 2.2E + 00 PU-239 1.8E M Total 5.8E + 01 1.9E + 01 Amendment 12.2-33 3 l
ABWR uraman wv, n Standard Plant SECTION 12.3 CONTENTS (Continued)
Section Elle B1gg 1233.2.2 Drywell 123-13a 1233.23 Reactor Building 123-13b 1233.2.4 Radwaste Building 123-13b 12 3.4 Area Radiation and Airborne Radioacthity Monitors 123-14 12 3.4.1 System Objectives 123-14 12 3.4.2 System Description 12 3-14 123.43 System Design 123-14 12 3.5 Post Accident Access Reautrements 123 15 12 3.6 Post Accident Radiation Zone Mamt 123 15 12 3.7 jnttrffiltt 12 3-15 12.3.8 Esferences 123 15 TAllLES Table 11.(19 Bigg 123 1 Computer Codes Used in Shielding Calculations 123 16 123 2 Typical Nickel and Colbalt Content of Materials 123-17 12 3-3 Area Radiation Monitor, Reactor Building 123 17.1 12 3-4 Area Radiation Monitor, Control Building 123 17.2 123 5 Area Radiation Monitor, Senice Building 12 3-17.2 123 6 Area Radiation Monitor, Radwaste Building 123-173 12 3-7 Area Radiation Monitor, Radwaste Building 12 3-17.4 i
12 3-i11 Amendment
l ABWR 2 mum ImV, H SlHDdard 1%nt ALARA. An example of this situation is the that could lead to radioactive crud deposi.
RWCS circulation pumps. Pumps adjacent to tion. Connections are available for conden-other highly radioactive equipment are also sate or demineralized water flushing of the shielded to reduce the maintenance exposure, heat exchangers. For highly radioactive for example, in the radwaste system, systems, such as the RilR and the RWCU, separate chemical decontamination con-Whenever possible, operation of the pumps nections are required in Subsection 12.3.7 and associated valving for radioactive for use in decontamination procedures, systems is accomplished remotely. Pump Instrumentation and valves are remotely conttol instrumentation is located outside operable to the maximum extent possible in high radiation areas, and motor- or the shielded heat exchanger cubicles, to pneumatic operated valves and valve reduce the need for entering these high extension stems are employed to allow radiation arcac.
operation from outside these areas.
(4) Valves
'2) Instrumentation Valve packing and gasket material are Instruments are located in low radiation selected on a conservative basis, accounting areas such as shielded valve galleries, for environmental conditions such as corridors, or control rooms, whenever temperature, pressure, and radiation possible. Shielded valve galleries provided tolerance requirements to provide a long for this purpose include those for the RWCS, operating life. Valves have back seats to FPCC, and radwste (cleanup phase separator, minimize the leakage through the packing, spent resin tank, sud waste evaporator) Straight-through valve configurations were systems. Instruments requitud a be located selected where practical, over those which in high radiation areas due to operations exhibit flow discontinuities or internal requirements are designed such that removal crevices to minimize crud trapping. Teflon of these instruments to low radiation areas gaskets are not used.
for maintenance is possible. Sensing lines are routed from taps on the primary system Wherever possible, valves in systems in order to avoid placing the transmitters containing radioactive fluids are separated or readout devices in high radiation areas, from those for " clean" services to reduce For example, reactor water level as well as the radiation exposure from adjacent valves recirculation system pressure sensing and piping during maintenance.
instruments are located outside the drywell.
Pneumatic or mechanically operated valves Liquid service equipment for systems are employed in high radiation areas, containing radioactive fluids are provided whenever practical, to minimize the need for with vent and backflush provisions, entering these areas. For certain Instrument lines, except those for the situations, manually operated valves are reactor vessel, are designed with provisions required, and in such cases extension valve for backflushing and maintaining a clean stems are provided which are operated from a fill in the sensing lines. The reactor shielded area. Flushing and drain vessel sensing lines may be flushed with provisions are employed in radioactive condensate following reactor blowdown. systems to reduce exposure to personnel during maintenance.
(3) Heat Exchangers For areas in which especially high radiation lleat exchangers are ;onstructed of stainless levels are encountered, valving is reduced steel or Cu/Ni tubes to minimize the pos- to the maximum extent possible with the bulk sibility of failure and reduce maintenance of the valve and piping located in an requirements. The heat exchanger design adjacent valve gallery where the radiation allows for the complete drainage of fluids levels are lower.
from the exchanger, avoiding pooling effects 12.3-2 Amendment
ABWR 2 m m^t Standard Plant wvn 12.3A Area Radiation and Airborne 123.43 System Design Radioactivity Monitors The area radiation monitoring detectors This section defines and describes the area provided in each plant building are listed in radiation system that monitors the gamma Tables 12.3-3 through 12.3-7 along with area radiation levels throughout the plant except location maps shown in Figures 12.3-56 through within the containment. The gamma radiation 12.3 73, Also, these tables specify the levels within the containment (drywell and sensitivity range of each channel as designated suppression chamber) are monitored continuously below along with requirements for local area by the containment atmospheric monitoring system alarms.
(CAMS) as described in Subsection 7.6.2. Four gamma sensitive ion chambers (two per divisions 1 The channel sensitivity covers the following
& 2) are provided by CAhts to monitor for airborne ranges:
radioactivity up to 107 rads per/hr. Those four sensors are located at the penetrations a) Range 10-2 to 102 mR/hr 11 ;
listed in Table 6.2-8. The area radiation (Iligh Sensithity) monitoring system is classified as non safety, 3
b) Range 10-1 to 10 mR/hr hl 123.4.1 S,ystem Objecthes (hf edium Sensithity)
The purpose of the area radiation monitoring c) Range 1 to 108 mR/hr - L (Low system is to warn plant personnel of excessive Sensithity) gamma ray levels in service areas including the areas where nuclear fuel is stored or handled, to d) Range 102 to 106 mR/hr - LL (Low record and indicate the monitored gamma radiation Low Sensithity) levels in the control room at selected locations within the various plant buildings, and to e) Range 10'I to 10' mR/hr - VL provide audible local -iarms at key locations (Very low Sensithity) where abnormal radiation levels could endanger plant personnel. There are two radiation detectors that are located in the fuel storage and handling area, 123.4.2 System Description one is positioned to monitor the radiation near the fuel pool and the other is placed in the The area radiation monitoring system fuel handling area to monitor the radiation that consists of gamma sensitive detectors, associated may result from accidental fuel handling.
digital radiation monitors, auxiliary units, Criticality detection monitors for this area are local audible warning devices and multipoint not needed to satisfy the criticality accident recorders. The detector signals are digitized requirements of 10CFR70.24, because the ABWR and optically multiplexed for transmission to the design utilizes specialized high density fuel radiation monitors. Each monitor has two storage racks that preclude the possibility of adjustable trip circuits for alarm initiation, criticality accident under normal and abnormal one high radiation level trip and one downscale conditions. The new fuel bundles are stored in trip. The downscale trip circuit operates on racks that are placed at the bottom of the fuel loss of power or when gross equipment failure storage pool. A full array of loaded fuel occurs. Auxiliary units are provided in local storage racks are designed to be suberitical by areas for radiation indication and for initiating at least 5% delta k. Refer to Sections 9.1 and the sonic alarms on abnormal levels. The 9.2 for details.
electronics are powered from the non 1E vital 120 Vac source while the recorders are powered from The detectors and radiation monitors are the 120 Vae instrument bus. responsive to gamma radiation over an energy range of 80 kev 7 hicV. The energy dependence Amendment 12.3-14
ABWR 2mmat.
Standard Plant ittiv. n will not exceed 20% of point from 100eV to 3 MeV. The overall system design accuracy is within 9.5% of equivalent linear full scale recorder output for any decade.
The trip alarra setpoints will be established in the field following equipment installation at the site. The exact settings will be based on sensor location, back ground radiation levels, expected radiation levels, and low occupational radiation exposures.
Each channel is calibrated based on a pseudo input signal to confirm accurate monitor response. The detectors are calibrated using standardized traceable radioactive source in order to establish the linearity and sensitivity of the channel for subsequent calibration. The area radiation monitoring system is designed to accommodate periodic surveillance testing.
The area radiation monitoring instru-mentation is designed and properly located to provide early detection and warning for personnel protection to insure that occupational radiation exposures will be as low as is reasonably achieved (ALARA) in accordance with guidelines stipulated in Reg Guide 8.2 and 8.8.
The area radiation monitoring system in-cludes instrumentation provided to assess the radiation conditions in crucial areas in the reactor building (the RHR equipment areas) where access may be required to service the safety related equipment during post LOCA per Reg Guide 1.97.
Amendment 12.3-18.1
ABWR ==
ru m n Standard Plant 12.3.5 Post Accident Access the environment and penetration leakage from the Itequircruents PASS system. Sources of radiation in each area are limited to gamma shine from the reactor The locations requiring access to mitigate the building and potential leakage from monitor consequences of an accident during the 100 day system such as the PASS. These sources are post-accident period are the control room, the considered minimal including the stack monitor technical support center, the remote shutdown room which contains only instrumentation with panel, the primary containment sample station their associated penetrations for monitoring (post accident sample system), the health physics stack effluent.
E f acility (counting room), and the nitrogen gas supply bottles. Each area has low post LOCA 12.3,6 Post Accident Itadiation 5
radiation levels. The dose evaluations in Zone Maps Subsection 15.6.5 are within regulatory guidelines. The post accident radiation zone maps for the areas in the reactor building are presented in Access to vital areas through out the reactor Figures 12.3-25 through 12.3 36. The zone maps building / control building / turbine building represent the maximum gamma dose rates that complex is controlled via the service building. exist in these areas during the post accident Entrance to the service building and access to period. These dose rates do not include the the other areas are controlled via double locked airborne contribution in the reactor building.
secured entry ways. Access to the reactor building is via two specific routes, one for Post accident zone maps of the control clean access and the second for controlled building and turbine building are presented in access. During an event such as a design basis Figures 12.3 54 and $5 respectively. The zone accident, the service building / control building maps are designed to reflect the criteria are maintained under filtered ilVAC at a positive established in Subsection 3.1.2.2.10.
pressure with respect to the environment. Air infiltration is minimized by positive flow via 12.3.7 Interface double entry ways. Therefore, radiation exposure is limited to gamma shine from the reactor The applicant will insure that the RilR and 8 building, turbine building, main steam line the RWCU systems are installed with separate 5 access corridor, and skyshine. This shine is connections to facilitate chemical decon-minimized by locating highly populated areas tamination of the systems, below ground.
12.3.8 lleferences During a design basis accident event, access l to remote shutdown panel, nitrogen bott!cs, and 1. N. M. Schaeffer, Reactor Shielding for the PASS and monitor systems is controlled from Nuc/ car Engineers, TID 25951, U.S. Atomic the service building via the controlled acccas Energy Cornmission (1973).
l way. These corridors are not maintained under filtered positive pressure so that personal 2. J. H. Ilubbell, Photon Cross Sections, protection equipment (radiation protection suits, Attenuation Coefficients, and Energy breathing gear, etc.) will be required in the Absorption Coefficients from to KcV to 100 access corridor. Primary contamination would (icv, NSRDS NBS20, U.S. Department of occur from leakage through the PASS system and Commerce, August 1969, air infiltration from the environment. Both pathways are considered minimal and minor 3. Radiological Ucalth Handbook, U.S.
contamination under even the most adverse Department of Elealth, Education, and conditions is expected. Welfare, Revised Edition, January 1970.
The reactor building vital areas are all 4. Reactor Handbook, Volume Ill, Part B, E.P.
located of f the controlled access way and Blizzard, U.S. Atomic Energy Commission contamination is limited to air infiltration from (1962).
12 3-15 Amendment
ABWR m6mw.
nov n Standard Plant
- 5. Lederer, liollander, and Perlman, Table of Isotopes, Sixth Edition, (1968).
- 6. M.A. Capo, Polynornial Approxirnation of Garntna Ray Buildup Factors for a Point Isotropic Sourec, APEX 510, November 1958.
- 7. Reactor Physics Constants, Second Edition, ANL-5800, U.S. Atomic Energy Commission, July 1963.
- 8. ENDF/B-Ill and ENDF/B.IV Cross Section Libraries, Brookhaven National Laboratory.
- 9. PDS 31 Cross Section Library, Oak Ridge National Laboratory.
- 10. DLC 7, ENDF/B Photo Interaction Library, i
Amendment t2.1111
p } '1 i
23A6100Al, A Standard Plant - imv. n L. 4 Y p -
Table 12.3 3 AREA RADIATION MONITORS- ,
P' -
REACTOR BUILDING i t:
Sensitivity Local
.& Location &Descrintion Finure #- Bangt Alarms -1
-1. Reactor area (A)-4F 123-62 - H' X
^
y' -
- Fuel storage pool area (B)-4F '12 3-62 LL
- 5' - R/B 4F south area 12 3-62 H
- 6 _.
ER/B 4F SE area 123 62 H- X 7' .R/B 3F NW area 123 60- H 8: 1R/B 3F SE arcaf =123 60 H -X 9' - CUW control panel area-B3F ~ 123-56 H T10 R/B equipment hatch-B2F 123 57 H X i
- 11 ~ lHCU area (A) B3F . 123 56 hi X
- 12 HCU area (B) B3F : 12 3-56 hi- X 13; SRV/htSIV valve maintenance room 3F 123 63-- ht X 14: R/B 1F SE area 12 3-49 H X j i 15 i RPV instrument rack room (A) B1F. 123 58 H -X' d 16 RFVinstrument rack room (B)-B1F 12 3-58_ H X j
-17 L 123-58 H "
R/B B1F SE area .
' TIP drive machine room El1500 '123-57 hi X 18 !
- 19 f. TIP machine equipment room El1500 '123-57 -L =X l 20 Core cooling water sampling room hf4F 12.3-61 hi X '
' 21 - CRD maintenance room B2F ' 12 3-57 hl X 022 _ -- R/B B2P SE area 12 3-57 H X 23 R/B B2F NW area 123-57 H .- .X- s
- = 24 -1 LR/B B3F NW area RHR 'A" equip area .123 56 YL- X .,
'~
. 25 '- ~ R/B B3F SE area RHR "B" equip area 12 3-56 VL X' 1 4
.~
y i f J. -
Amendment -- 12.3-17.1 i y+ ., , , . , - . - . .w.v.+ -.w--s-. -- .e,,-v>~,.-r.- m uevs
ABWR ummat.
imv n Standard Plant Table 12.3 4 AREA RADIATION MONITOltS CONTitOL llUILDING Sensitivity b's laation &Descripli.gn l'icure # linngt
'l Main Control Room 123 M 11 2 Passage Way Underneath Steam Tunnel 122 M 11 3 RilCW"A" Area El 1315 123 M i!
4 RBCW"II" Arca El 1315 123 64 11 5 RBCW "C" Area El .1315 123-M 11 Table 12.3 5 AREA RADIATION MONITORS SERVICE IlUILDING Sensitivity
&h Imation & Description Fleure # llangg 1 Service Building Tech. Support Center 123-64 11 l
12.3-17.2 Amendment
ABWR uumi.
Standard Plant tutv. ri l
Table 123 6 AREA RADIATION MONITORS RADWASTE IlUILDING Sensitivity Local hih location & Description Fleure # lyingt Alarms 1 R/W Building Control Room-El16000 123 68 11 2 Maintenance area 41 El16000 12 3-68 11 X 3 Maintenance Area #2-El16000 123-68 11 X 4 R/W Building 11VAC Exhaust El 1600 12 3-68 11 5 R/W BuildingTruck Area El7300 123 67 11 6 MSW Compactor Area El7300 12 3-67 11 7 Corridor to Aux. Building El7300 12 3-67 11 X 8 Equip Rack Area #1 El-0200 12 3-66 11 9 Equip Rack Area #2 El-0200 123 66 11 10 R/W Building MSW Control Room El-0200 12 3-66 11 11 Rad Waste Sampling Room El 6500 123-65 II 12 MSW Equipment Arca El 6500 12 3-65 II X 13 R/W Equipment Rack Area #1 El 6500 123-65 11 14 R/W Equipment Rack Area #2 El 6500 123 65 11 i
Amendment 12.3 17,3
ABWR zwimi.
nirv. n Standard Plant Table 12.3 7 AREA RADIATION MONITORS TURilINE BUILDING Sensitivity Local
& location & Description Fleure # Rangt Ala[H13 1 T/B Operating Area - Level 4 123 73 11 X 2 MSR Area #1 Level 4 123-72 11 3 MSR Area #2 Level 4 12 3-72 11 4 IIP Heaters Area #1 Level 3 12 3-71 H 5 IIP IIcaters Area #2 - Level 3 12 3-71 11 6 S.IAE Area Level 3 1^23-71 H X 7 Filter Maintenance Area Level 3 123-71 11 8 RFP Area Level 2 123-70 11 9 Off GasSampling Area Level 2 123-70 M X 10 Condensate Water Sampling Area - Level 2 12 3-70 M 11 Condensate Pumps Area Levell 123-69 M Amendment 123 17.4
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ABWR . u-1.
Standard Plant un 12A DOSE ASSESSMENT Most of the measurements indicate that the dose rates received during MSIV maintenance Dose assessment is an important part of increased in proportion la the activity in the determining and projecting that the plant design and recirculation lines. Since these have been proposed methods of operation assures that eliminated this contribution is reduced to vero.
occupational radiation exposure will be as low as Significant dose rate reductions are expected from reasonably achievable. Dose assessment depends the RIIR return flow through the feedwater line, upon estimates of occupancy, dose rates in various Based on past experience the contact dose rate on occupied areas, number of personnelinvolved in this type of line can be expected to be from 50 to reactor operations and surveillance, routine 200 mrem /hr. At an average of 4 meters the dose maintenance, waste processing, refueling, in-service rate is expected to be 4 to 15 mrem /hr. Cold legs inspection, and special maintenance. historically show smaller activity build ups than warm legs. It is estimated that the average field The goal is to reduce the exposure associated will be 4 mrem /hr for this activity.
with each phase of plant operation and maintenance to the minimum level consistent with practical (2) SRV considerations for accomplishing each task. To achieve this goal, the ABWR design includes SRV maintenance will primarily be affected by numerous significant design improvements to reduce better access to SRVs and a reduction in the occupational exposures from past experience. The radiation field by elimination of the recirculation design improvements include the climination of piping. The radiation source comes from the recirculation piping and valves, improved water RllR return flow in the feedwater with an chemistry and low cobalt alloys at the cooling water estimated 50 mrem /hr contact dose rate at a boundary, reduced equipment maintenance and distance of 3 meters. Improved access has not improved access, RHR discharge to the feedwater been accounted for in this estimate with the piping, overhaul handling and refueling devices, radiation field estimated at 5,5 mrem /hr.
multiple main steam line plugs, automatic MSIV seat tapping system and reactor vessel stud tensioner. In (3) CRD assessing the collective occupational dose, each potentially s:gnificant dose causing activity was CRD maintenance will significantly reduced with evaluated. an assumed maintenance schedule of drive and spool pieces overhauled per year and 20 drive motors inspected per year. A automated handling machine will be used with the maintenance crew reduced from five to four. Drawing from a 12.4.1 Drpell Dose combination of US and European experience, maintenance times for the preparation of the area Estimates of the task radiation for drywell under the vessel is 2 days and for removal time for components are as follows: one drive is 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Motor and spools are like drives and require 10 man hours each. An (1) MSIV cffective (averaged over all conditions) dose rate of 15 mrem /hr is estimated.
MSIV maintenance is expected to be reduced by use of MSIV overhauling devices, use of main steamline plugs, automatic MSIV lapping sptem. The radiation field will be reduced by elimination of the external recirculation piping, and RHR return through the feedwater system.
In the data base neither the overhauling aids or the lapping devices were used. Use of (4) Inservice inspecti m automatic systems will result in overall l reduction in maintenance times of 50% inservice inspection is reduced by removal of Amendment IMt
ABWR nu m Standard Plant nev n recirculation lines, elimination of 2 nonle (2) Refueling it.spcctions per year, elimination of shield i penetrations and plugs removal associated with Refueling will be reduced by use of an automated lines, improved automation equipment, and refueling platform where no personnel are located on reduced vessel welds. It is estimated that over the platform itself. The improved fuel, inspection 50% of the experienced man hours are saved. equipment and increased remote operations The improvements in vessel accessibility significantly reduce the refueling floor exposure. Fuel l through better insulation reduces the drywell sipping will not be required based upon the anticipated )
dose by over 50% For this case it is fuel. l considered that the contact dose rate on the feedwater line is less than half the contact dose (3) CRD rate on recirculation lines. It is expected that automation will permit 25% of the work to be CRD rebuilding time is reduced to 2 rebuilds per done remotely for work represented by lower year. The effective dose rate is 4 mrem /hr. l exposures.
(4) RWCU !
(5) RIP Heat Exchangers 1 RWCU maintenance includes 2 pump inspections l RIP heat exchangers assume 3 pumps and 1 per year with canned motors based upon the use I heat exchanger will be serviced per year. of an improved motor design. The estimated dose Estimates are based upon European experience rate 5 mrem /hr. l assuming an effective dose rate of 15 mrem /hr.
(5) RHR RHR maintenance consists of two series of (6) General Drywell inspections per year on two pumps. The maintenance time per pump per inspection is General drywell work is reduced by removal of estimated to be 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />, recirculation lines, reduced inspection of snubbers, HVAC units, and associated (6) General Reactor Building components. The drywell average dose rate is redue d because of the elimination of the General Reactor Building maintenance work ree'culation lines. It is e:.timated that contact includes the hydraulic control units, pumps, steam dwe rate against recirculation lines are more tunnel compartments, and instrumentation than twice the expected contact dose rate from equipment in the secondary containment.
the feedwater line containing the RHR return flow. This is based on the temperature of the feedwater line flow. The effective dose rate is l estimated to be 5 mrem /hr.
12.4.2 Reactor Building Dose Estimates of the task radiation for the reactor building components are as follows:
(1) Vessel Vessel access and assembly will be reduced by use of stud tensioner over % bolts. Estimated dose rate is 3 mrem /hr.
Amendment 12.4 2
ABWR naamat.
Standard Plant .ladi !
Table 12.41 !
PROJECTED ANNUAL RADIATION EXPOSURE FOR AllWR I'.xperienced Time and Dose Rate Projected Dose ,
Work Task flaurs mrem /hr Person Rem !
4 Drywc!! !
, MSIV 2000 4 8 SRVs 200 5.5 1 RIPS 200 15 3 Misc Valves 1500 4 6 CRD 200 15 3 LPRM/11P 200 40 8 ISI i200 5 6 Instrumentation 1000 $ 5 Other 1300 5 6.5 Reactor Iluilding Vessel Acccas/ Reassembly 2000 3 6 Refueling 2000 0.5 1 CRD Rebuilding 200 4 0.8 RWCU 200 5 1 RilR' . 200 10 2 Valves / Pumps - 600 4 2.4 l Instrumentation 600 3 1.8 l Other ~ 800 3 2.4 i
.)
Turbine Ilullding ,
Turbine Overhaul 15000 0.3 4.5 Valves / Pumps 1000 3 3 ,
Condernate System - 1000 3 3 0;hrt -10000 0.1 1 l
Radwaste Building 3500 1.5 5.25 Work at Power 3500- 5 17.5 TOTAL 98 i
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