ML20034G629
| ML20034G629 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 03/05/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9303100259 | |
| Download: ML20034G629 (6) | |
Text
I GENuclear Energy Gev'r (Wrc Compay l
U5 Cacw Awnae. San Jose. CA BM25 March 5,1993 Docket No. STN 52-001 i
l l
Chet Poslusny, Senior Project Manager.
Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Review Schedule - Chapter 6 DFSER Outstanding Items
Dear Chet:
l Enclosed is a markup of SSAR Chapter 6 addressing: Open items 6.2.5-3 and 6.2.6-6; and l
COL Action Item 6.2.5-1.
Sincerely, k Fox Advanced Reactor Programs cc: Bill Fitzsimmons (GE)
Norman Fletcher (DOE)
Bernie Genetti(GE) i JIW51 9303100259 930305 PDR ADDCK 05 1
^
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23A61EDAS Standard Plant bA
Response
vahes are opened during power operation. Hese are air-operated vahes with rapid closure times, present-This requirement is not applicable to the ABWR.
ing little opportunity for substantial releases from the It applies only to PWR-type reactors.
PCV in the event of a transient requiring containment isolation. Note that under the technical specifications, 19A.2.26 Isolation Dependability [ Item (2) containment inerting and purging with the larger (xiv)]
ventilation lines is permitted during power operation above 15% for limited periods at either end of the NRC Position operating cycle. The process of purging the contain-Provide containment isolation systems that:
ment with air also serves to remove any potential
[lI.E.4.2) activity for ALARA censiderations prior to actual personnel entry into the PCV.
(A) Ensure all non-essential systems are isolated automatically by the containment isolation The large ventilation valves will be tested regularly
- system, and after any vahr maintenance to assure that closing times are within the limits assured in the radiological (B) For each non-essential penetration (except in-design basis. See Subsection 19A33 for COL license strument lines) have two isolation barriers in information.
- sches, 19A.2.28 Design Evaluator [ Item (2) (xvi)]
(C) Do not result in reopening of the containment isolation valves on resetting of the isolation NRC Position
- signal, Establish a design criterion for the allowable (D) Utilize a containment set point pressure for ini-number of actuation cycles of the emergency core tiating containment isolation as low as is com-cooling system and reactor protection system consis-patible with normal operation, tent with the expected occurrence rates of severe over cooling events (considering both anticipated tran-(E) Include autornatic closing on a high radiation sients and accidents). (Applicable to B&W designs signal for all systems that provide a path to the only.) [II.E.5.1]
ensrons.
Response
Response This requirement is not applicable to the ABWR. It This item is addressed in Subsection 1A.2.14.
applies only to PWR-type (B&W designed) reactors.
19A.2.27 Purging [ Item (2) (xv)]
19A.2.29 Additional Accident Monitoring Instrumentation [ Item (2) (xvii)]
NRC Position NRC Position Provide a capability for containment purg-ing/ venting designed to minimize the purging time Provide instrumentation to measure, record and consistent with ALARA principles for occupational readout in the control room: (A) containment pres-exposure. Provide and demonstrate high assurance sure, (B) containment water level, (C) containment that the purge system will reliably isolate under acci-hydrogen concentration, (D) containment radiation dent conditions. [II.E.4.4) intensity (high level), and (E) noble gas effluents at all potential, accident release points. Provide for con-
Response
tinuous sampling of radioactive iodines and particu-lates in gaseous effluents from all potential accident The ABWR primary containment vessel (PCV) op-release points, and for onsite capability to analyze and crates with an inert atmosphere. During normal op-measure these samples. [ILF.1]
eration, all large valves in containment ventilation lines are closed Only small,2", nitrogen-makeup 3
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gg nev c Standard Plant influent and effluent lines of this group are 6.2.4.4 Test and Inspections isolated by automatic or remote.manualisolation valves located as close as possible to the The containment isolation system is scheduled to undergo periodic testing during containment boundary.
reactor operation. The functional capabilities 6.2.43.2.4 Evaluation Against Regulatory of power operated isolation valves are tested remote manually from the control room. By Guide l.11 observing position indicators and changes in the Instrument lines that connect to the RCPB and affected system operation, the closing ability penetrated the containment have 1/4-inch orifices of a particular isolation valve is demonstrated, and manual isolation valves, in compliance with Air-testable check valves are provided on Regulatory Guide 1.11 requirements.
influent emergency core cooling lines of the HPCF and RHR systems whose operability is relied 6.2.4.3.3 Evaluation of Single Failure upon to perform a safety function.
A single failure can be defined as a failure of a component (e.g., a pump, valve, or a utility A discussion of testing and inspection of such as offsite power) to perform its intended isolation valves is provided in Subsection safety functions as a part of a safety system.
6.2.1.6. Instruments are periodically tested The purpose of the evaluation is to demonstrate and inspected. Test and/or calibration points that the safety function of the system will be are supplied with each instrument. Leakage completed even with that single failure. integrity tests shall be performed on the Appendix A to 10CFR50 requires that electrical containment isolation valves with resilient systems be designed specifically against a single material semis at least once every 3 months.
passive or active failure. Section 3.1 describes 6.2.5 Cornbustible Gas Controlin the implementation of these standards as well as General Design Criteria 17,21,35,38,41,44, Containntent 54, 55 and 56.
The atmospheric control system (ACS-T31) is l Electrical as well as mechanical systems are provided to establish and maintan. en inert designed to meet the single failure criterion, atmosphere within the primary containment during regardless of whether the component is required all plant operating modes except during shutdown to perform a safety action. Even though a com-for refueling or equipment maintenance and ponent, such as an electrically-operated valve, during limited periods of time to permit access The is not designed to receive a signal to change for inspection at low reactor power.
state (open or closed) in a safety scheme,it is flammability control system (FCS-T49) is assumed as a single failure if the system compon-provided to control the potential buildup of ent changes state or fails. Electrically-oper-oxygen from design basis radiolysis of water.
ated valves include valves that are electric. The objective of these systems is to preclude ally piloted but air operated, as well as valves combustion of hydrogen and damage to essential that are directly operated by an electrical de-equipment and structures. /NSE# r- (.y. c
[h' g l vice. In addition, all electrically operated valves that are automatically actuated can also 6.2.5.1 Design Bases be manually actuated from the main control room.
Following are criteria that serve as the Therefore, a single failure in any electrical I
system is analyzed, regardless of whether the bases for design:
loss of a safety function is caused by a Since there is no design requirement for component failing to perform a requisite (1) mechanical motion or a component performing an the ACS or FCS in the absence of a LOCA and there is no design-basis accident in the unnecessary mechanical motion.
ABWR that results in core uncovery or fuel failures, the following requirements mechanistically assume that a LOCA f
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,3. i 23xsiooxa mfc Semn,Isrd PInnt Included in the leak rate test sun. mary report will be, a report detailing the containment in-spection, a report detailing any repairs nec: -
sary to pass the tests, and the leak rate test results.
6.2.6.5 SpecialTesting Requirements The maximum allowable leakage rate into the secondary containment and the means to verify l
that the inleakage rate has not been exceeded, as well as the containment leakage rate to the l
environment, are discussed in Subsections 6.23 (cj,,,pryc)
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6.2 References 1.
W.J. Bilanin, The G.E. Mark III Pressure Suppression Containment Analytical Model, June 1974, (NEDO-20533).
2.
F.J. Moody, Maximum Discharge Rate of i
Liquid-Vapor Mixtures from Vessels, Genernl Electric Company, Report No. NEDO-21052, September,1975.
3.
W.J. Bilanin, The G.E. Mark III Pressure Suppression Containment Analytical Model, Supplement 1, September 1975 (NEDO-20533-1).
t t; 4.
J.P. Dougherty, SCAM-Subcompartment 9
Analysis Method, January 1977, (NEDE-
}
21526).
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