ML20034H124
| ML20034H124 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 02/23/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9303150316 | |
| Download: ML20034H124 (5) | |
Text
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,, e ev:<wc v n ; :n unaa s.,me u ews February 23,1993 Docket No. STN 52-001 t
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 - Resubmittal of Open Item 15.3-1 and COL Action Items 17.1.1-1 and 17.2-1
Dear Chet:
Enclosed are SSAR markups revising our responses to Open item 153-1 and COL Action Items 17.1.1-1 and 17.2-1 originally transmitted in my letters dated February 16 and 17, 1993.
Please provide a copy of this tranmittal to George Thomas and Tim Polich.
Sincerely, k*f J ek Fox Advanced Reactor Programs cc: Norman Fletcher (DOE)
Bob Huang (GE)
Phil Novak (GE) 1 i
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Standard Plant I N S E F.T A worst possible location, and the plant is oper-drawn from the reload core analysis as previous.
ated with the mislocated bundle. This event is ly presented is applicable to the ABWR initial categorized as a limiting fault based on the core. Hence, no specific analysis is required.
following data:
4 15.4.7.4 Barder Performance j
'I Expected Frequency: 0.002 events / operating cyc!c.
1 An evaluation of the barrier performance is j
not made for this event, because it is a mild i
This number is based upon past experience.
and highly localized event. No perceptibte j
change in the core pressure is observed.
15A.7.2 Sequence of Events and Systems 1
Operation 15A.73 Radiological Consequences.
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i 15A.7.2.1 Sequence of Events An evaluation of the radiological conse-quences is not required for this event, because l
The postulated sequence of events for the no radioactive material is released for the misplaced bundle accident (MBA) is presented in fuel.
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Table 15.4-6.
15.4.8 Rod Ejection Accident f
15A.7.2.2 Systems Operation 15A.8.1 Identifiention of Causes and Frequency A fuel loading error, undetected by in-core Classification instrumentation following fueling operations, may result in an undetected reduction in thermal The rod ejection accio;nt is caused by a-margin during power operations. For the analysis major break on the FMCRD housing. outer tube or.
reported herein, no credit for detection is taken associated CRD pipe lines. Due to a break of and, therefore, no correctise operator action or this type, the reactor pressure exerted on the automatic protection system functioning is CRD spud pushes down the hollow piston and the i
assumed to occur.
ballnut with a large force... The shaft screw and '
the motor are forced to' unwind. A passive brake
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15A.73 Core and System Performance mechanism is installed in the FMCRD system to prevent the control rod from moving. The design This event is presented in Subsection S.2.5.4 of the brake is presented in Section 4.6.1. The -
j of Reference 1.
probability of the initial causes, i.e.', a CRD pipe line break or housing break, is considered l
l Mislocated bundle analyses are not performed low enough to warrant its being categorized as a j
for reload cores because, based on analysis of limiting fault. Even if this accident does date available from past reloads, the probability happen, the brake prevents the control rod from that a mislocated fuel bundle loading error will ejection. Should the brake fail, the check result in a CPR less than the safety limit is valve will serve as a backup brake to prevent sufficient small (see Reference S.2-58 of the rod ejection.
- l Reference 1).
15A.8.2 Sequence of Events and Systems j
For ABWR initial core, the mismatch of Operation exposures and integrated bundle power between mislocated bundle: are less severe than the If a major break occurs on the FMCRD housing.
equilibrium cycle. Therefore, the consequence of the reactor pressure will provide forces that a postulated misplaced bundle accident for the could cause the shaft screw to unwind. The
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initial core is less severe than that for the FMCRD brake mechanism prevents the rod from equilibrium cycle. Consequently, the conclusion moving. Therefore, no rod ejection can occur.
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T' I N SG R'T A The other potential type of bundle loading error that can occur is the mis-oriented fuel bundle (MOFB). In this case, the bundle is in the correct location but is rotated by 90 or 180 degrees. In reactors where the water gaps are non-uniform around the bundle or where the rod enrichmer.t distribution is not quadrant-symmetrical, rotation can cause increases in local rod power through increased moderation. In ABWR lattice, the rotation results in non-uniform water gaps and produces similar increases in local rod power.
The initiator for operating a reactor with a MOFB is an operator placing the bundle into the core in a mis-oriented position. The next step in the accident progression is failure to detect the MOFB. A verification process is recommended to detect a MOFB. This verification procedure requires two core scans. One scan is with an underwater TV camera positioned close enough to read the bundle serial numbers on top of the lifting bail (first attribute) and to check the orientation of the bosses (second attribute). The other scan is with a TV camera positioned l
sufficiently above the core to allow viewing one complete 4 bundle cell for the following four attributes: boss on lifting bail, channel fasteners, channel buttons, and " cells look alike". Two independent reviewers (checkers A and B) are recommended to verify video tapes from the above procedures.
A generic model was developed based on the recommended verification procedure to quantify the probability of operating a reactor with a MOFB. An event tree was constructed to find this probability using human error rates from NUREG/CR 1278. The results show that the probability of operating the reactor with a MOFB is 8.5x10-5 per cycle (5.7x10-5/ year with a 18 month fuel cycle). This probability of operation with a MOFB is lower than the probability of a large break LOCA (i.e.,10-4 per year).
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.. Starfdard Plant nev s control rod from the hollow piston of the FMCRD.
If the control rod is stuck, the separation de.
tection devices will detect the separation of the l control rod and hollow piston from the ba!! nut of the FMCRD and rod block interlock will ptevent further rod withdrawal. The operator will be i
alarmed for this separation.
There is no basis for the control rod drop l cvent to occur.
15.4.9.3.2 Identification of Operstor Actions No operator actions are required to preclude j
this event. However, the operator will be notified by the separation detection alarm if separation is detected.
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15.4.9.4 Cort and System Performance 15 4 10 COL t s ch **. h b~-
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The performance of the separation detection 1
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devices and the rod block interlocks virtually II'i IO*l Ne k W S ov mTa T\\#^
preclude the cause of a rod drop accident.
o k Ana k.riJ l 15.4.9.5 Barrier Performance Ev w
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g kee s O fMh An evalu" ion of the barrier performance is g ed%
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,p a not made sr this accident since there is no 4
j circumstance for which this event could occur.
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15.4.9.6 Radiological Consequences
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o The radiolop. cal analys..ts is not required.
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Ii 15.4J4 References 1.
General Electric Standard Application for Reactor Fuel--United States Supplement, r
l NEDE 24011 P-A US, (Latest approved revi.
sion).
2.
C. J. Paone and J. A. Woolley, Rod Drop Accident Analysis for Large Boiling Water Reactors, Licensing Topical Report, March
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1972 (NEDO-10527, Supplements 1 and 2).
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17.0 INTRODUCTION
Section 17.1 of this Standard Safety Analysis (7. 0.1 COL Licmae T,8e% don Report describes the Quality Assurance (OA)
Moe Program which is implemented by GE for the ABWR i7.O.t.I G.A Pro $v'avw2awdope d ua project. It is based upon the standard GE OA con 3Wh o, i
Program documented in the GE Nuclear Energy topical report NEDO 11209-04A (Reference 1) and Tk.c cot c(p phc M/hoIh the. additional information in this chapter s bo. ll pv-A p m e a. M i m pd i
describing and clarifying GE's interfaces and
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I the ABWR These technical associates are major f u" 4
- 4"b ^ Pk)M responsibilities with its technical associates on international corporations who are licensees of ei'SAdhe% t '7 I d -i-ba. ore d e GE's technology and have extensive independent p h s.c., o f' T s e b o M 1... T h experience in the design and construction of
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nuclear power stations.
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The standard program is used throughout GE cmd NO. A
\\ q - go.31r 'I (S4t-Nuclear Energy on all other nuclear power plant M 0) work and has been accepted by the Nuclear Regulatory Commission. It is in compliance with Title 10, Code of Federal Regulations, Part 50, Appendix B; ANSI /ASME N45.2; ANSI /ASME Snd Ae sAaki4 N45.2-series standards; and NRC Regulatory Guides
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with some NRC accepted GE Nuclear Energy psy,3,og "y
alternate positions.
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I The QA Program described in this chapter meets Regulatory Guide 1.28, Revision 3 and is organized to show its relationship to Reference 1, ANSI / ASME NOA-11983 and NOA la 1983, and GE's interfaces with its technical associates.
The terms and definitions of supplement S-1 of NOA la-1983 apply. Table 17.0-1 summarizes ABWR compliance with the quality related Regulatory Guides.
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