ML20059J117
| ML20059J117 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 10/28/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19311B188 | List: |
| References | |
| NUDOCS 9311120037 | |
| Download: ML20059J117 (12) | |
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GENucle:rEnergy l
Genera! Electric Company 175 Cume Avenue, SanJose. CA 95125 i
i October 28,1993 Docket No.52-001
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Chet Poslusny, Senior Proj,ect Manager i
Standardization Project Directorate Associate Directorate for Advanced Reactors i
and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Schedule - SSAR Supporting Information for Chapter 11 and Appendix 18F j
Reference:
(1) Letter, Jack Fox to Chet Poslusny, Submittal Supporting Accelerated ABWR Schedule - Revised Appendix 18F, l
DFSER Confirmatory Item 18.4.3-1, Dated June 9,1993 t
(2) Letter, R. C. Mitchell to R. C. Pierson, Reclassification of i
ABWR SSAR Proprietary Information, Resubmitted, Dated August 18,1992 Dear Chet-.
, provides SSAR markups of Section 11.0 and Appendix 18F referencing SSAR supporting information. The SSAR supporting information for Chapter 11 is l
provided m Enclosure 2. The SSAR supporting information for Appendix 18F was provided previously by Reference 1. is essentially the Amendment 31 proprietary version of SSAR Sections -
i issued are still applicable. proprietary affidavits under which they were originally!
11.2,11.3 and 11.4 and the Et should be noted that the pages of Enclosure 2 that i
contain the same mforn ation as SSAR Amendment 32 (non-proprietary) are not i
designated as proprietary in this transmittal.
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A final reclassification of the ABWR SSAR and SSAR supporting information is under preparation for transmittal by November 10,1993. This final classification will be similar to Reference 2.
Please provide a copy of this transmittal to Chandra.
Sincerely, Y4 Jack Fox Advanced Reactor Programs cc:
Alan Beard (GE) w/o Enclosure 2 Drawings Norman Fletcher (DOE) w/o Enclosure 2 Drawings JIW2M
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ENCLOSURE 1 SSAR MARKUPS OF SECTION 11.0 AND APPENDIX 18F REFERENCING SSAR SUPPORTING INFORMATION OCTOBER 1993 GE NUCLEAR ENERGY SAN JOSE, CA l
4
23A6100 Rn.1 i
ABWR studentsaferyAnalysis neport 11.0 Radioactive Waste Manda emet., b +b r s s Ae. chetplcr sg 6
11.1 Source Terms is peuvdd b w ec.u.i-s.
3 The information provided in this section defines the radioactive source terms in the reactor water and steam which serve as design bases for the gaseous, liquid and solid radioactive waste management systems.
Radioactive sotu ce term data for boiling water reactors has been incorporated in American National Standard ANSI /ANS-18.1 (Reference 11.1-1). The Standard proddes bases for estimating typical concentrations of the principal radionuclides which may be anticipated over the lifetime of a BWR plant. The source term data is based on the cumulative industry experience at operating BWR plants, including i
measurements at several stations through 1981. It therefore reflects the influence of a number of observations made during the transition period from operation with fuel of older designs to operation with fuel of current improved designs. The source terms speciSed in this section were obtained by applying the procedures of Reference 11.1-1 for estimation of typical source terms and adjusting the results upward as appropriate to assure conservative bases for design.
The sarious radionuclides included in the design basis term have been categorized as fission products or activation products and tabulated in the subsections which follow.
The lists do not necessarily include all radionuclides which may be detectable or theoretically predicated to be present. Those which have been included are considered to be potentially significant with respect to one or more of the following criteria:
(1) Plant equipment design (2) Shielding design (3) Understanding system operation and performance 1
(4) Measurement practicability (5) Evaluation of radioactidtyin efDuents to the environment 11.1.1 Fission Products 11.1.1.1 Noble Radiogas Fission Products Typical concentrations of the 13 principal noble gas fission products as observed in i
steam flowing from the reactor vessel are provided in the Source Term Standard ANSI /ANS-18.1 (Reference 11.1-1). Concentrations in the reactorwater are considered negligible because all of the gases released to the coolant are assumed to be rapidly transported out of the vesselwith the steam and removed from the system with the other non-condensables in the main condenser. As a consequence of the immediate removal Source Terms - Amendment 31 11.1 1 i
23A6100 Rsv.1 ABWR studadsdetyAulysis Report i
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11.1.5 Process Leakage Sources Process leakage results in potential release of noble gases and other volatile fission products via ventilation systems. Liquid from process leaks is collected and routed to the liquid-solid radwaste system. With the effective process offgas treatment systems now in use, the ventilation releases are relatively significant contributions to total plant releases.
Leakage of fluids from the process system results in the release of radionuclides into j
plant buildings. In general, the noble radiogases will remain airborne and will be released to the atmosphere with little delay via the building ventilation exhaust ducts.
Other radionuclides will partition between air and water and may plate-out on metal surfaces, concrete, and paint 'ladioiodines are found in ventilation air as methyl iodide and as inorganic iodine (particulate, elemental, and hypoiodous acid forms).
As a consequence of normal steam and water leakage in to the drywell, equilibrium dr)well concentrations will exist during normal operation. Purgmg of this activity from the dr>well to the emironmentwill occurvia the Standby Gas Treatment System and will make minor contributions to total plant releases.
1 Airborne release data from BWR building ventilation systems and the main condenser mechanical vacuum pump have been compiled and evaluated in Reference 11.1-4, which contains data obtained by utility personnel and from special in-plant studies of
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operating BWR plants by independent organizations and the General Electdc Company. Releases due to process leakage are reflected in the airborne release estimates discussed in Section 11.3.
11.1.6 References 11.1-1 American National Standard Radioactive Source Term for Normal Operation of Light Water Reactors, ANSI /ANS-18.1.
I1.1-2 Skarpelos,J.M. and R.S. Gilbert. TechnicalDerivation ofBMR 1971 Design Basis Radioactive MaterialSource Tenns, March 1973 (NEDO-10871).
11.1-3 Calculation of Releases of Radioactive Matedals in Gaseous and Liquid EfIluents from Boiling Water Reactors, NUPIG-0016, Revision 1 January 1979.
11.1-4 Martcro, T.R., Airborne ReleasesFmm BARsforEnvironmentalImpact Evaluations, March 1976 (NEDO-21159).
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i 23A6100 R1v. 2 ABWR standantsafetyAnalysis Rapon 18F Emergency Operation Information and Controls 18F.1 Introduction This appendix contains the results of an analysis ofinformation and control needs of the main control room operators. The analysis is based upon the operation strategies given in the ABWR Emergency Procedure Guidelines (EPGs) as presented in Appendix 18A and upon the significant operator actions determined by the Probabilistic Risk Assessment (PRA) and given in Appendix 19D.7. The minimum inventory of controls, displays and alarms from this analysis are presented in Tables 18F-1 through 18F-3 of this appendix. The informadon and controls identified from this anahsis do not necessarily include those from other design requirements 1
l (such as thos fr m Section 18.4.2.11, SPDS). Syph M M' Nh o^
PN h0 N M 198:-l.
i Information and control needs for each operation instruction or action are were I
developed through task analyses conducted in the following manner:
Each specific step in the EPGs (referred to as the EPG step) or specific operator a
action referenced in the PRA (herein referred to as the PRA step) was indhidually identified.
For each EPG step and PRA action, a summa 27 escription of the step or operator d
s action was developed.
Information needs of the operator to perform the specific EPG step or PRA m
operator acdon were then identified.
1 Next, the control functions that the operators perform to execute the actions s
specified in the EPG step or PRA operator action were identified.
The plant process parameters or other displays that are needed for execution of the s
individual EPG step or PRA operator action, were then identified.
Similar!y, the controls needed for the execution of the step, were identified m
Annunciators necessary for the execudon of the step were identified.
a Operator aids, such as supplementary procedures or other information needed for m
the execution of the step, were identified.
Displays used to provide a feedback to the operators to confirm that the specified s
control functions have been initiated or accomplished, were identified.
Position of control devices that provide feedback to the operators to confirm that s
proper controls are manipulated to the correct positions, were identified.
Emergency Operation Information and Controls - Amendment 32 18F-1
s-23A6100 Rsv.1 ABWR cunderdsareryAntysis aeron l
l Annunciators which provide feedback to the operators to confirm that proper I
e control actions are initiated or accomplished were identified.
Operator aids, which provide feedback to the operators to confirm that proper a
control actions are initiated or accomplished, were identified.
l The following operator actions are considered to be important operator actions in the l
ABWR PRA (refer to subsection 19D.7):
(1) Backup manual initiation of.HPCF (2) Recovery of feedwater following a scram l
(3) Use of condensate injection following scram with reactor depressurized I
(4) Control of reactor water level in an ATVS (5) Emergency depressurization of the reactor 1
(6) Alignment and initiation of firewater for RPV injection with ECCS failure (7) Alignment and initiation of firewater for drywell spray (8) Initiation of werwell spray using RHR (9) Isolation of water sources in an internal flooding These actions are already specified in the EPGs and are included in the analyses.
Based upon the results of those operator task analyses, the listings of controls, displavs and alarms that will be provided in the implemented ABWR design to support execution of the EOPs and PRA significant operator actions (as presented in Tables 18F-1.18F-2, and 18F-3), were generated.
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e ENCLOSURE 2 LIQUID, GASEOUS AND SOLID WASTE MANAGEMENT SSAR SUPPORTING INFORMATION OCTOBER 1993 GE NUCLEAR ENERGY SAN JOSE, CA
1 a
ABWR ssAn suportingintonnation Table of Contents List of Tables..
. I1.1-iii
.11.1-v List of Figures.
11.2 Uquid Waste Management..
.I1.2-1
..I1.2-1 11.2.1 Design Basis.
11.2.2 System Description.
.I1.2-4 11.2.3 Estimated Releases..
.I1.2-7 11.2.4 Tank Resistance to Vacuum Collapse..
.11.2-9 11.3 Gaseous Waste Management System..
. I1.Sl 11.3.1 General..
.I1.11 11.3.2 Design Criteria..
.11.Sl 11.3.3 Process Description.
.11.3-2 11.3.4 OfTgas System Description.
.11.S 4 11.3.5 Other Radioactive Gas Sources..
.11.S19
.11.S20 11.3.6 Instrumentation and Control.
11.3.7 Quality Control..
.I1.3-20 11.3.8 Seismic Design.
.11.3-21 11.3.9 Testing..
.I1.3-21 11.3.10 Radioactive Releases..
. II.S24 11.3.11 References..
.I1.S25 11.4 Solid Waste Management System.
.11.4-1 11.4.1 Design Bases..
.I1.4-1 11.4.2 System Description.
.I1.4-2 1
1 I
l Table of Contents - October 1993 11.1-i'il
I.
ABWR SSAR Supportinginfonnation List of Tables Table 11.2-1 Equipment Codes for Radwaste Equipment (from Table 1, RG 1.143).
.I1.2-10 Table 11.2-2 Capability of Liquid Radwaste Systems to Process Expected Wastes.
.I1.2-10 Table 11.2-3 Reactor Coolant Activity (RCA) Fraction..
.I1.2-11 Table 11.2-4 Capacities of Tanks, Pumps, and Other Components.
.I1.2-12 Table 11.2-5 Radwaste Eqvioment Performance Design Basis.
.I1.2-14 Table 11.2-6 Probable Inputs to I iquid Radwaste from Operational Occurrences.
.I1.2-15 Table 11.3-1 Estimated Air Ejector OITgas Release Rates Per Unit 3
(51 sm /br Inleakage).
.I1.S26 Table 11.S2 Offgas System 5fajor Equipment items..
.I1.127 Table 11.13 Equipment 5talfunction Analysis.
.I1.S29 Table 11.34 OfTgas System Instrument Setpoints.
.I1.3-32 Table 11.4-1 Expected Waste Volume Generated Annually by each " Wet" Solid Waste Source and Tank Capacities.
.I1.4-11 Table 11.4-2 Estimate of Expected Annual" Dry" Solid Wastes and Curre Count..
.I1.4-11 Table 11.4-3 Calculated Solid Waste Volumes and Curre Count.,
.I1.4-12 Table 11.4-4 Solid Waste hianagement System Components..
.I1.4-13 Table 11.4-5 Calculated Solid Waste Isotope Distribution.
.I1.4-15 I
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d List of Tables - October 1993
ABWR ssAn supportinginformation List of Figures Figure 107E6109 Radwaste System PFD (Sheets 1-2)
Figure 103E1634 Radwaste System P&lD (Sheets 1-36)
Figure 103E1603 OfTgas System PFD (Sheets 1-2)
Figure Jo3E1602 OfTgas System P&ID (Sheets 1-3) 11.1-iv List of Figures - October 1993