ML20046C626
| ML20046C626 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/04/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9308110247 | |
| Download: ML20046C626 (10) | |
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GENucle:rEnergy GeneralEn ric Comwy l 115 Cur nor Avenue, kn Jose, CA 95125 August 4,1993 Docket No. STN 52-001 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 Schedule - Fuel Bundle ,
Mis-orientation (Item #14)
Dear Chet:
Enclosed is the SSAR markup addressing Fuel Bundle Mis-orientation Item #14. We plan' to include this in Amendment 32 which is scheduled to be transmitted to the NRC on August 31, 1993.
Please pcovide a copy of this transmittal to George Thomas.
Sincerely, YO Jack Fox Advanced Reactor Programs cc: Alan Beard (GE)
Norman Fletcher (DOE)
Caroline Smith (GE) mm, 100030 9308110247 930804 3p )i PDR ADOCK 05200001 , hV l.
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23AB100 Rw.1 l ABWR snzdsed sateryAztysis a: port \
15.4.5.4.2 Fast Runout of All Reactor internal Pumps !
This transient results in a slight increase in reactor vessel pressure-(Figure 15.4-3) and therefore represents no threat to the RCPB.
15.4.5.5 Radiological Consequences An evaluation of the radiological consequences is not required for this event because no radioactive material is released from the fuel.
15.4.6 Chemical and Volume Control System Malfunctions Not applicable to BWRs. This is a PWR event.
ps slo (AkcM 15.4.7p? ace %undle Accident 15.4.7.1 Identification of Causes and Frequency Classification 15.4.7.1.1 Identification of Causes The event discussed in this section is the loading of a fuel bundle in an improper location and subsequent operation of the core. Three errors must ocgr,f is event to take place in the equilibrium core loading. First, a bundle must byibdbinto a wrong location in the core. Second, the bundle which was supposed to be loaded where evvor the (vguyoccurred is also put in an incoa ct location or discharged. Third, the mhp=& bundles are overlooked during the core verification process performed following core loading.
Provisions to prevent potential fuel loading errors are included in the plant Operating Procedures / Technical Specification.
15.4.7.1.2 Frequency Classification This unlikely event occurs when a fuel bundle is loaded into the wrong location in the core. It is assumed that the bundle is misplaced in the worst possible location, and the plant is operated with the mislocated bundle. This event is categorized as a limiting fault based on the following data:
Expected Frequency: 0.002 events / operating cycle.
This number is based upon past experience.
15.4.7.2 Sequence of Events and Systems Operation 15.4.7.2.1 Sequence of Events gyQ The postulated sequence of events for th mbphced bundle accident (MBA) is presented in Table 15.4.6.
Reactivity and Power Distribution Anomalies - Amendment 31 15.4 11 b
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15.4.7.2.2 Systems Operation rm s\ a ea%d b mb \ u Ahd LLg error, undetected by incore instrumentation following fueling operations, may result in an undetected reduction in thermal margin during power operations. For the analysis reported herein, no credit for detection is taken and, therefore, no corrective operator action or automatic protection system functioning is assumed to occur.
15.4.7.3 Core and System Perfortnance Mislocated bundle analyses are not performed for reload cores because, based on analysis of data available from past reloads, the probability that a mislocated fuel bundle loading error will result in a CPR less than the safety limit is sufficiently small.
For ABWR initial core, the mismatch of exposures and integrated bundle power between misloaded bundles are less severe than the equilibrium cycle. Therefore, the consequence of a postulated MBA for the initial core is less severe than that for the equilibrium cycle. Consequently, the conclusion drawn from the reload core analysis as previously presented is applicable to the ABWR initial core. Hence, no specific analysis is required.
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 notated by 90 or 180 degrees. In reactors where the water gaps are non-uniform around the bundle or where the rod enrichment distribution is not quadrant-symmetrical, rotation can cause increases in local rod power through increased moderation. In the ABWR lattice, the rotation results in non-uniform water gaps and produces similar increases in local rod power.
The initiator for opening 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 i l
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 (ftrst attribute) and to check the orientation of the bosses (second attribute). The f other scan is with a TV camera positioned sufIiciently above the core to allow viewing one complete four-bundle cell for the following four attributes: boss on lifdng bail, channel fasteners, channel buttons, and " cell look alike". Two independent reviewers (checkers A and B) are recommended to verify tapes from the above procedure.
A generic model was 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 tha the probability of operating the reactor with a MOFB is 8.5x10-5 per cycle v
15.4 12 Reactivity and Power Distribution Anomalies - Amendment 31 I
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[4 (5.7x10 / year with a 18. month fuel cycle). This probability of operation with a MOFB 4
is lower than the probability of a large break LOCA (i.e.,10 per year).
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The COL applicanf wil ee l gojide an analysis to confirm that the consequences oi ms u.h i =Sngevent meet all requirements approved by the NRC. See Subsection 15.4)d.1 for COL license information. ;
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15.4.7.4 Barrier Performance An evaluadon of the barrier performance is not made for this event, because it is a mild and highly localized event. No perceptible change in the core pressure is observed.
15.4.7.5 Radiological Consequences i An evaluation of the radiological consequences is not required for this event because l
> no tea radioactive s git.-T 15 4,8 material is released for the fuel. {
15.4fRod Ejection Accident .
f.1 Identification of Causes and Frequency Classification ;
15.4)3 The rod ejection accident is caused by a major break on the FMCRD housmg, outer l tube or associated CRD pipe lines. Due to a break of this type, the reactor pressure l exerted on the CRD spud pushes down the hollow piston and the ballnut with a large [
force. The shaft screw and the motor are forced to unwind. A passive brake mechanism is installed in the FMCRD system to prevent the control rod from moving.The design ;
of the brake is presented in Section 4.6.1. The probability of the initial causes (i.e., a CRD pipe line break or housing break) is considered low enough to warrant its being categorized as a limiting fault. Even if this accident does happen, the brake prevents the ,
control rod from ejection. Should the brake fail, the check valve will serve as a backup ;
brake to prevent the rod ejection.
15.4[>.2 Sequence of Events and Systems Operation l If a major break occurs on the FMCRD housing, the reactor pressure will provide forces that could cause the shaft screw to unwind. The FMCRD brake mechanism prevents the !
rod from moving. Therefore, no rod ejection can occur, q 15.4./.3 Core and System Performance 0 The FMCRD brake mechanism prevents this event from occurring.There is no need to .
analyze this event. ,
-l 15.4/.4 Barrier Performance 9 An evaluation of the harrier performance is not made for this accident since there is no [
circumstance for which this event would occur.
15.4-13 Reaccrity and Power Distribution Anometies - Amendment 31
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9 15.4.y'.5 Radiological Consequences The radiological analysis is not required.
15.4,6 Control Rod Drop Accident ;
15.45.1 Features of the ABWR Fine Motion Control Rod Drives
\b As presented in Subsection 4.6.1, the Fine Motion Control Rod Drive (FMCRD) System has several new features that are unique compared with locking piston control rod 1 drives. .
t In each FMCRD, there are dual Class 1E separation-detection devices that will detect the separation of the control rod from the CRD if the control rod is stuck and separated from the ballnut of the CRD. The control rods are normally inserted into the core and l withdrawn with the hollow piston, which is connected with the control rod, resting on the ballnut. The separation-detection device is used at all times to ascertain that the hollow piston and control rod are resting on the balinut of the FMCRD.The separation- ;
detection devices sense motion of a spring-loaded support for the ball screw and,in turn, the hollow piston and the control rod. Separation of either the control rod from the hellow piston or the hollow piston from the ballnut will be detected immediat'ely.
When separation has been detected, the interlocks preventing rod withdrawal will operate to prevent further control rod withdrawal. Also,'an alarm signalwill be initiated in the control room to warn the operator.
There is also the unique highly reliable bayonet type coupling between the control rod ;
-i blade and the control rod drive. With this coupling, the connection between the blade ~
and the drive cannot be separated unless they are rotated 45 degrees.This rotation is ' ;
not possible during reactor operation. There are procedural coupling checks to assure ;
proper coupling. There is also the automated overtravel check in the RCIS logic during j automated operation. Finally, there is the latch mechanism on the hollow piston part of .
the drive. If the hollow piston is separated from the ballnut and rest of the drive due to a stuck rod, the latch willlimit any subsequent rod drop to a distance of 8 inches. More l
detailed descriptions of the FMCRD System are presented in Subsection 4.6.1.
i 15.4/.2 Identification of Causes and Frequency Classification iD For the rod drop accident to occur, it is necessary for such highly unlikely events as ;
failure of both Class IE separation-detection devices, or the failure of the rod block ;
interlock, and the failure of the latch mechanism to occur simultaneously with the '
occurrence of a stuck rod on the same FMCRD. This would permit hollow piston separation from the ballnut.
Alternatively, separation of the blade from the hollow piston would require either that i
the control ro 1 was installed without coupling and the coupling checks failed, or there Reactivity and Power Distribution Aromalien - Amendment 31 15.4 14 1
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15.4.8 Misoriented Fuel Bundle Accident 15.4.8.1 Identification of Causes and Frequency Classification 15.4.8.1.1 Identification of Causes The misoriented fuel bundle (MOFB) event discussed in this section is the situation in which a bundle has been loaded in the correct location but is rotated by 90 or 180 degrees. The rotation could result in non-uniform water gaps which could cause an increase in local rod power through increased 4
moderation. The initiator for a reactor with a MOFB is an operator placing the bundle into the core in a I
misoriented position. The next step in the accident progression is failure to detect the.MOFB. A I
verification procedure 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 on the bail (second attribute).The other scan is with a TV camera positioned sufficiently above the core to allow viewing one complete 4 bundle cell for the following four attributes: boss on lifting bail, channel fasteners, l channel buttons, and " cell look alike." Two independent reviewers (checkers A and B) are recommended to verify tapes from the above procedure.
15.4.8.1.2 Frequency Classification ;
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 consuucted to find this probability using human error rates from NUREG/CR-1278. The results show that the probability of operating the l reactor with a MOFB is 8.5x10-5 per cycle (5.7x10 5 / year with an 18 month fuel cycle). This probability of operation with a MOFB is lower than the probability of a large break LOCA (i.e.,10" per year). ;
However, since at the time of this submittal the NRC has not approved this classification, the MOFB has been treated as a moderately frequent event, and analyzed accordingly.
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15.4.8.2 Sequence of Events and Systems Operation 3
15.4.8.2.1 Sequence of Events The postulated secuence of events for the misoriented fuel bundle accident (MOFB) is presented ;
f in Table 15.4-7. :
i 15.4.8.2.2 Systems Opemtion A misoriented fuel bundle accident, undetected by in-core instrumentation following fueling operations, may result in an undetected reduction in thermal margin during power operations. For the
' analysis reported herein, no credit for detection is taken and, therefore, no corrective operator action or ;
automatic protection system functioning is assumed to occur. l 15.4.8.3 Core and System Perfonnance /gj_3 i
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The MOFB event was analyzed for a reference core loading utilizing a bundle which isjery similar to the reference fuel bundle design. This bundle design is defined in Tab AY of Reference 5. The l only difference in the MOFB bundle design slight modifications to the radial enrichment distribution i
which were made to reduce the AR-factor. The maximum AR-factor under rotated conditions was i determined to be 0.035. The bundle used in this analysis exhibited energy capabilities equivalent to the reference bundle design and the 15% thermal margin requirement was maintained. The infinite lattice void i
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- coefficients for both designs were compared and there was no change. The methods for analyzing the misoriented fuel bundle are described in detail in Reference and approved in Reference 5. The ACPR i l
calculated for this event is reported in Table 15.0 2. [15.4- 5 l isA_4
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15.4.8.4 Banier Performance An evaluation of the barrier performance is not made for this event because it is a mild and highly localized event. No perceptible change in the core pressure is observed.
15.4.8.5 Radiological Consequences f An evaluation of the radiological consequences is not required for this event because no radioactive material is released from the fuel.
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23A6100 R1v.1 ASWR Stand:rd Saf:ty An: lysis R:prit is structural failure of this coupling. Under such circumstances of this coupling failure, the rod drop accident can only occur with the simultaneous failure of both separation-detection devices (or the failure of the rod block interlock), together with the occurrence of a stuck rod on the same FMCRD.
In either case, because of the low probability of such simultaneous occurrence of these multiple independent events, there is no basis to postulate this event to occur.
tD 15.4.Jf.3 Sequence of Events and System Operation 15.4.8.3.1 Sequence of Events IO The bayonet coupling and procedural coupling checks will preclude the uncoupling of the control rod from the hollow piston of the FMCRD. If the control rod is stuck, the separation detection devices will detect the separation of the control rod and hollow piston from the ballnut of the FMCRD, and rod block interlock will prevent further rod withdrawal. The operator will be alar med for this separation.
There is no basis for the control rod drop event to occur.
\O 15.4g.3.2 Identification of Operator Actions No operator actions are required to preclude this event. However, the operator will be notified by the separation-detection alarm if separation is detected.
15.4.fl.4 Core and System Performance .l The performance of the separation-detection devices and the rod block interlocks sittually preclude the cause of a rod drop accident.
15.4.g.5 Barrier Performance iO An evaluation of the barrier performance is not made for this accident, since there is no !
circumstance for which this event could occur. ]
\U 15.4.S.6 Radiological Consequences I
The radiological analysis is not required.
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Reactivity and Power Distribution Anomalies - Amendment 31 15.4 15
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15.4.11 References 15.4-1 Deleted.
15.4-2 C.J. Paone andJ. A. Woolley, Rod Drop Accident Analysisfor Large Boiling Water Reactors, Licensing Topical Report, March 1972 (NEDO-10527, Supplements ,
I and 2).
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,. i TABLE 15.4-7 ,
SEQUENCEOFEVENTS OFTHE MISORIENTED FUEL BUNDLE ACCIDENT (1) During the core loading operation, a bundle is rotated 90 or 180 degrees.
(2) During the core verification procedure, this error goes undetected.
(3) The plant is brought to full power operation without detecting :
the misoriented bundle.
i (4) The plant continues to operate throughout the cycle.
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