ML19257B066
| ML19257B066 | |
| Person / Time | |
|---|---|
| Issue date: | 09/14/1978 |
| From: | Jabbour K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19257B062 | List: |
| References | |
| REF-GTECI-B-48, REF-GTECI-CR, TASK-B-48, TASK-OR NUDOCS 8001150179 | |
| Download: ML19257B066 (40) | |
Text
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DRAFT OPERATING EXPERIENCE WITH BWR CONTROL ROD DRIVE COLLET RETAINER TUBES K. N. JABBOUR Division of Operating Reactors Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 1750 006 8001150 l
TABLE OF CONTENTS PAGE ABSTRACT.........................................................
1
1.0 INTRODUCTION
- DESCRIPTION OF THE BWR CONTROL R0D DRIVE SYSTEM................................................
2 1.1 Control Rods...................
2 1.2 Control Rod Drive...................
3 1.3 Control Rod Drive Hydraulic System....................
3 1.4 Description of the Collet Retainer Tube................
5 2.0 OPERATING EXPERIENCE........................................
7 2.1 C ra c k D i s c o v e ry........................................
7 2.2 Inspection Results.....................................
8 3.0 EVALUATION..................................................
10 3.1 Presumed Cause of the CRT Cracking.....................
10 3.2 Significance of the CRT Cracking.......................
11 4.0 CORRECTIVE MEASURE5........................
13 4.1 Interim Corrective Measures - Changes to Technical Specifications.......................................
13 4.2 Long Term Corrective Measures..........................
15 4.3 Interaction with the Office of Nuclear Reactor Regulation's Generic Technical Activities Program....
21
5.0 CONCLUSION
S.................................................
24 APPENDIX A - SAMPLE OF TYPICAL TECHNICAL SPECIFICATIONS..........
36 1750 007
ABSTRACT The NRC stiff has conducted a review of operating experience with the collet retainer tube (CRT) utilized in the control rod drive (CRO) systems of boiling water reactor (BWR) facilities because of the cracking which has been detected in these tubes at a number of operating BWR facilities.
This report describes the BWR CR0 system, the CRT cracking experience, the significance of the CRT cracking, and the corrective measures, both interim and long term, to reduce the likelihood of occurrence of such cracking in the future and to minimize the consequences of such cracking should it occur. This technical issue has been identified as Technical Activity B-48.
This report is meant to document the completion of the efforts for Task Action Plan B-48.
1750 008
1.0 INTRODUCTION
- DESCRIPTION OF THE BWR CONTROL R0D DRIVE SYSTEM 1.1 Control Rods The reactivity of a BWR's active core is regulated by the movement of cruciform shaped control rods which penetrate the bottom of the reactor vessel.
A typical BWR control rod is depicted i7 Figure 1.
The control rods which contain sections with neutron absorbing material provide basic control functions of the reactor during startup, normal operation, shutdown, and scram (rapid insertion of the control rods during certain transient and accident conditions).
The control rods are inserted into the core to decrease reactivity or are withdrawn from the core to increase reactivity. Each control rod can be individualIy moved by its associated control rod drive which is mounted in a housing at the bottom of the reactor vessel.
The systems associated with the operation of BWR control rods include the control rod drive (CRD) system and the reactor control and instrumentation systems.
1750 009
.. 1.2 Control Rod Drive The CRD is a double-acting mechanically-latched hydraulic cylinder using demineralized water as the operating fluid.
As shown in Figure 2, each CRD is an integral unit wholly container'.
in a housing extending from the bottom of the reactor vessel.
The control rod is connected to the CRD by a spud consisting of six spring fingers which enter a mating socket on the control rod.
The lower end of each CRD housing terminates in a flange which mates with the CRD flange.
1.3 Control Rod Drive Hydraulic System The basic components of the CRD hydraulic system for controlling the drive mechanism during positioning and scram operation are shown in Figure 3.
All components shown are typical of those that are associated with each rod drive.
The principal movable element of the system consists of the main drive piston, the index tube and the control rod coupled to the index tube.
The movable element is held in any chosen position by a collet which engages one of several notches in the index tube.
Gravity holds the tube notch against the latch since the entire mechanism is essentially at reactor pr' essure.
1750 010
. The control rod is moved by applying a pressure greater than reactor pressure to either the top or bottom of the main drive piston. When the reactor protection system calls for a reactor scram, all control rods are driven into the core at the maximum rate of speed.
When a scram signal is initiated, control air is vented from the scram valves allowing them to open by spring action.
Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric pressure and opening the inlet valve applies the accumulator pressure of 1500 psi to the bottom of the piston. Since the notches in the index tube are tapered on the lower edge, the latch is forced open by cam action allowing the index tube to move upward without restriction or3 r influence of the high pressure differential across the piston.
As the drive moves upward and accumulator pressure reduces to reactor pressure, the ball check valve opens letting reactor water complete the scram action.
If reactor pressure is low, such as during startup, the accumulator will fully insert the rod in the required time without assistance from reactor pressure.
The elapse of time from the opening of a reactor protection system sensor contact to the opening of the main scram contactor is less than 0.050 second.
1750 011
1.4 Description of the Collet Retainer Tube A collet retainer t. '
'CRT) is an internal tube located at the upper end of a CRD. The CRDs of a typical BWR are hydraulically operated stepping devices mounted in housings extending downward from the lower reactor vessel head that ir.5ert or withdraw corresponding control rods located in the core.
The drives operate within the reactor environment and utilize unheated reactor coolant as the operating fluid.
The CRT houses the locking collet and support piston that consists of a collet return spring and retainer, and an unlocking cam.
Figures 4 and 5 illustrate the present design of the CRT.
The CRT has three primary functions:
(1) to transfer the hydraulic unlocking pressure to the collet piston; (2) to provide an outer cylinder with a suitable inner wear surface for the metal collet piston rings; and (3) to provide mechanical support for the guide cap.
The guide cap also provides the upper rod guide or bushing for the drive mechanism.
During rod insertion, the ports in the upper end of the CRT allow displaced water to discharge from inside the drive irito the surrounding annulus.
1750 012
m The CRT is typically machined from Type 304 stainless steel (SS) tubing and then solution annealed.
The tube is next flash nickel plated, then copper plated, except for the portion of the inner surface that is to be nitrided. The tube is then nitrided and the copper plating is removed by a chromic-sulfric acid dip. The CRT was then welded to the outer tube, and the inside surfaces of the non-nitrided area were machined to finish dimensions.
1750 013
2.0 OPERATING EXPERIENCE 2.1 Crack Discovery During routine maintenance inspections of a General Electric de igned BWR in June 1975, dye penetrant inspections of CRD cc1ponents revealed fine cracks in some of the CRTs located at the upper end of the CRDs.
Subsequent inspections of other drives that had been in operation disclosed similar cracks.
Conventional metallography and scanning electron microscopy identified the cracking as intergranular in nature.
The cracks were generally circumferential, appearing mainly where the wall changes thickness, in the area between the ports.
The cracks had developed from the outside of the tube, but none of the cracks were through the wall.
They were generally shallow, less than half the wall thickness.
Many of the cracks were
" tight" and filled with oxide.
Figures 6 and 7 show typical cracking patterns in the CRT at early and advanced stages.
Because the CRT is fabricated from wrougt:t Type 304 stainless steel (55), which is in the sensitized (i.e., susceptible to stress corrosion cracking) condition as a result of the nitrid-ing cycle, the conclusion was that the interg-anular cracking was caused by a stress corresi chanism.
Figcre 8 depicts the dye penetrant indications ar. the c)ack and Figure 9'shows 1750 014
. an intergranular attack on c longitudinal strip between a pair of holes.
Operating experience obtained from 200 BWR years and results of the 779 CRTs* (out of about 4000 drives in service) inspected by the dye penetrant technique at 22 sites disclosed that partial cracking had occurred in 78 tubes at 11 of the sites.
Table 1 provides the details of this data for each plant.
2.2 Inspection Results Metallographic examinations were performed on CRTs from several plants.
In addition, studies were perfomed to better define carbon gradient effects as a possible role in susceptibility to cracking and to investigate a possible correlation between results of the dye penetrant examinations and the metallographic findings.
These inspect ons lead to the following conclusions:
a.
The cracking observed in the CRT is initiated by a stress-corrosion cracking mechanism.
The driving stresses are cyclically induced thermal residual stresses.
- As of June 1978 1750 015
9 b.
No nondestructi.e testino method of crack depth measurement was found that could be applied to parts containing dye penetrant indications.
c.
Dye penetrant testing may not tr'eal all cracks present on a field inspected CRT.
The undetected cracks are generally tight, shallow, and filled with oxide.
1750 016
N 3.0 EVALUATION 3.1 Cause of CRT Cracking The mechanical loads on the CRT are very small during normal operation (i.e., 175 psi).
The stress during drive withdrawal or scram at ambient pressure is approximately 500 psi.
The modes of CRD operation that produce the most severe stresses are hot scrams or full-stroke insertions without cooling water flow.
Stress analyses were conducted based on temperature distribution tests (which show that during a scram, cold water flows up the inside of the CRT and out the flow holes while hot water flows outside the CRT) and on the most severe operational modes.
These analyses indicated that through-wall temperature differentials on material located between the holes were suffi-cient to cause biaxial compressive yielding and when the temperature gradients are relaxed, residual tensile stresses are produced in the outside surface.
Evaluation of this stress pattern shows that for the portion between holes, crack extension occurs either shortly after the scram or in the times between scrams.
In contrast, cracking of the portion below the holes must occur only during each hot scram.
Two types of cracking 1750 017
occur at the change in section, circumferential between holes and vertical below holes.
The extent of either one is a func-tion of the number of scram cycles.
3.2 Significance of the CRT Cracking The structural integrity and the functional operability of the CRTs and CRDs are important because continued circumferential cracking could lead to a complete 360-degree severance of the tube, possibly preventing either insertion or withdrawal of the control rod.
This condition would be detected in any fully or partially withdrawn drive during the periodic surveillance required by each plant's Technical Specifications.
A complete severance failure of the CRT is highly unlikely although partial cracking has occurred.
No complete failure of a CRT has occurred in any of approximately 4000 drive ssemblies currently in operatica.
Because of the highly unlikely natura of the failure of one CRT, the chance that a large number of collet housing would fail completely at the same time is extremely remote.
This is primarily true because the distributions of failures by cracking mechanisms such as stress, corrosion and fatigue are not linear functions.
Since no collet housings have failed to dat'e, we 1750 018
have concluded that there would be very few, if any, failures during the time period corresponding to the total service life of the CRT.
However, due to the potential consequences of such failures, as described in Section 4.0, interim corrective measures have been implemented and long-term corrective measures have been planned.
1750 019
4.0 CORRECTIVE MEASURES 4.1 Interim Corrective Measures - Changes to Technical Specifications Until the latter part of 1975, the typical existing limiting conditions of operation (LCO) of an operating facility allowed plant operation to continue with up to eight inoperable CRDs.
The then existing surveillance requirements specified that daily surveillance of the conditions of all fully or partially withdrawn rods would not have to begin until three rods were found to be inoperable.
These existing conditions of operation and surveillance requirements did not sufficiently limit the possibility of operating for an extended period of time with a number of control rod drive mechanisms which cannot be moved.
Therefore, the NRC concluded that plant Technical Specifications should be changed as discussed below.
A sacole of typical Technical Specifications is enclosed in Appendix A.
a.
It was assumed that one stuck control rod does not create a significar.t safety concern.
However, if a rod cannot be moved and the cause of the failure cannot be determined, it was assumed that the rod could have a failed collet housing.
A potentially failed collet housing would be indicative of a problem which could eventually affect the scram capability of more than one control rod.
Since the 1750 020
m.
cracks appear to be of a type which propagate slowly, it is highly unlikely that a second control rod would experi-ence a failed collet housing within a short period of time after the first failure. Therefore, a period of time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> was allowed to determine the cause of failure.
This period is considered sufficient to determine if the cause of failure is not in the drive mechanism, yet short enouch to be reasonably assured that a second collet failure does not occur.
Therefore, the NRC required that the Limiting Condition for Operation, " Reactivity Margin -
Inoperable Control Rods", be expanded to require that if a control rod cannot be moved during normal operation, testing or scram, the reactor shall be shut down within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if the reason that it cannot be moved cannot be shown to be due to causes other than a failed collet housing.
b.
If a control rod drive cannot be moved, the cause of the stuck rod might be a problem aff ecting other rods.
To ensure prompt detection of any additional control rod drive failures which could prevent movement, the NRC required that the surveillance requirement for " Reactivity Margin - Inoperable Control Rods" be expanded to require surveillance every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of all partially and f'ully withdrawn rods if cne rod drive is found to be stuck.
1750 021
. Another typical Technical Specification that has long been in effect for operating BWRs requires exercising fully or partially withdrawn rods once every week as a part of corrosian prevention programs.
Although not intended by specification, this require-ment provides demonstration of CRT integrity.
However, the NRC has recently approved an amendment request (from a licensee of an operating reactor that exhibited no CRT cracks) to allow monthly exercising of one of the rods in each two-by-two array while the other three rods in the array will continue to be exercised weekly.
Such typical Technical Specification requirements coupled with the GE results on cracked CRTs (which show that the CRTs capa-bilities exceed the design number of hot scram cycles without impairing their function) provide a basis for concluding that operation of the CRD system will be satisfactory.
4.2 Long Term Corrective Measures The investigation of the cracking in the CRT indicated a need for modifying the present CRT design.
The following subsections discuss the design modifications, supporting materials, and the qualification testing program instituted to support this modification.
1750 022
As discussed in Section 2.2, the cracks observed in the collet retainer tubes of CRD mechanisms are caused by cyclically induced thermal residual stresses in a component sensitized by nitriding and exposed to an environment conducive to intergranular stress cerrosion cracking. To eliminate the stress co'rosion cracking problem, any or all of the three contributing conditions can be modified:
(1) susceptible material, (2) stress, and (3) environment.
The modified CRT development program has addresseo each of the three considerations by taking the following steps to improve the reliability of the CRD system for long-term operation:
a.
Elimiration of the material section change in order to reduce the discontinuity stresses.
b.
Change the material to cast 304L (ASTM CF-3) with controlled ferrite to provide resistance to stress corrosion.
c.
Recommendations to operating plant utilities to attempt to keep oxygen concentrations as low as practicable in the coolant for the drives to improve the environment to which the CRTs are exposed.
1750 023
4.2.1 Modified Design of CRT A two piece CRT hardfaced utilizing Colmonoy-6 in lieu of nitriding (Figure 10) has been developed to reduce thermally induced stresses and to eliminate the sensitization due to nitriding the tube.
The modified CRT design consists of a hardfaced cast 304L (ASTM CF-3) spacer and a cast 304L thin wall upper tube, both welded to the existing Type 30455 outer tube.
The improved design allows the upper tube to respond to thermally induced strains independently of the constraint of the spacer, thereby eliminating the section change stress effect.
To further reduce stresses, the new design has 12 cooling water exit ports (rather than three which are provided in the current CRT design).
This design modification reduces the circumferential variation in temperature below the flow holes during a scram cycle.
The modified design allows the use of Colmonoy-6, a nickel base hardfacing material applied by the flame spray process.
The colmonoy is applied to the inside diameter of the cast spacer to provide a wear surface where the collet seal rings contact the part.
1750 024
. The spacer also incorporates the collar, used in the original design to maintain concentricity between the collet retaining tube and the drive cylinder.
This arrangement has the added advantage of eliminating the axial stresses residual to the assembly weld.
4.2.2 Cool'nq Water Hole Design d
Both 6-and 12 port hole collet retainer tube configurations were evaluated with respect to temperature distribution during scram to identify a design with acceptably low circumferential temperature variations.
The tests were performed on a drive fitted with a prototype two piece CRT.
It was concluded from the test that the 12 port hole CRT :on-figuration (i.e., twelve 3/8-inch port holes circumferentially located 3.75 inches below the upper lip of the tube) provides a more uniform temperature distribution, both circumferentially and vertically, than a similar 6 port hole configuration.
Maximum vertical temperature gradient, below the flow orifice, is 180 F, while the maximum circumferential temperature gradient, 1/2 inch below the flow orifice, is 160 F.
The above maximums are associated with a 1510/575 psi accumulator scram at 1030 psi vessel pressure and an initial CRD probe temperatur'e of H50 025
250 F.
The maximum temperature changes occur approximately 3 seconds after scram initiation.
4.2.3 Colmonoy Hardfaci..g The current CRT design of wrought Type 304SS is hardfaced by nitriding.
Although the wear resistance of nitrided parts has been favorable, the nitriding thermal cycle leaves the material in the sensitized condition.
The modified design is hardfaced with Colmonoy-6, a nickel base alloy.
The Colmonoy is applied to the inner piece of the two piece design, which subjects only the spacer (inner piece) of the CRT to the hardfacing thermal cycle.
The colmonoy is machined to final dimensions following the welding of the spacer to the outer tube.
Since a nitrided surface cannot be machined, nitriding was not an option in the modified design.
4.2.4 Environment The cracking exhibited in some CRD collet retainer tubes is attributed to thermal cycles during hot scrams with exposure to aerated cooling water, which is aggressive to furnace-sensitized stainless steel materials.
The source of CRD cooling w'ater in 1750 026
. most BWRs is the condensate storage tank.
Typical CRD cooling water conductivity is 1 pmho/cm measured at 25 C, and the typical dissolved oxygen concentration is 5 ppm.
Somewhat higher or lower values can be observed depending upon operational practices and specific designs for each plant.
GE tests indicate that at any specific stress level there is a direct relationship between dissolved oxyger concentration and time-to-failure of furnace-sensitized stainless steel test specimens.
If high purity deaerated water (conductivity 1 0.1 pmho/cm measured at 25 C and dissolved oxygen 1 0.050 ppm) is used for cooling water, a significant increase in time-to-crack formation can be gained independent of any other metallurgical or mechan-ical improvements being considered.
This reduction in dissolved oxygen may thus increase the time-to-crack initiation of the present CRTs by a factor of 100.
Methods for providing high purity deaerated water to the CR0 system have been recommended by GE to BWR operating plant utilities to reduce the likelihood of CRT cracking.
The recom-mendation identifies approaches to achieving water quality limits and water sampling and monitoring additions.
The applicability of the recommendations has been successfully demonstrated in an operating BWR.
1750 027
4.3 Interaction with the Office of Nuclear Reactor Regulation's Generic Technical Activities Program NRC established a Category B Technical Activity (No. B-48) to address and resolve the issue of BWR Collet Tube Failures.
This Activity consists of reviewing several reports which address several aspects of the test program for the new CRT design.
These include:
l.
" Collet Retainer Tube Design Modification Development Report," General Electric Report NEDE-24004 (Proprietary),
June 1977.
2.
"XM-19 Materials Qualifications Report," General Electric Recort NEDE-21653 (Proprietary), July 1977.
3.
"CRD Components Program Report," General Electric Report NED0-21809 (Non-Proprietary), March 1978.
In the first report, GE presented its program to resolve the control rod collet retainer tube cracking.
It also presented its program to resolve the index and piston tubes materials degradation problem in the second report.
The third report is a non proprietary version ui the former reports.
1750 028
GE concluded that the cracking which occurred in components of the CRD fabricated from wrought Type 304 stainless steel with nitrided surfaces was due to the susceptibility of the material to intergranular stress corrosion cracking (IGSCC) in crevice areas.
GE further concluded that cast 304L (ASTM CF-3) is more suitable material for use in the CRT and XM-19 alloys are more suitable for the index and piston tubes of the CRD.
NRC agrees that the development and test data submitted by GE supports the conclusion that the use of these alloys will be an improvement over the use of unstabilized austenitic type 304SS.
The data indicate that the cast 304L has a greater resistance to crevice corrosion and can withstand service conditions which caused stress corrosion cracking in the 30455.
The XM-19 alloy has higher strength combined with greater resistance to sensitiza-tion and IGSCC than 30455.
Approximately 14 domestic operating BWRs have CR0 components fabricated from Type 30455 with nitrided surfaces.
As a result of the CRD component upgrade development program, La Salle Unit No. 1 is the first domestic plant which contains the redesigned CRT.
The CRT is fabricated from cast ASTM CF-3 alloy with Colmonoy hardfacing, and the index and piston tubes are fabricated from XM-19 alloy.
Hope Creek will be the next BWR plant to have the new components.
1750 029
The above-mentioned activity has recently been completed and a safety evaluation has been issued.
This NUREG report constitutes the final safety assessr. ant of the BWR collet tube cracking.
e 1750 030
5.0 CONCLUSION
S The CRT cracking that has occurred to date is generally shallow, inter-mittent, and very " tight", and has not impaired the CRD's ability to meet its functional requirements. The interim corrective measures (i.e.,
Technical Specifications) have provided additional assurance that failures will be identified before a significant number of CRDs would be affected.
These Technical Specification requirements assure early detection and replacement of failed CRDs in the unlikely event that a CR0 is stuck because of a separated CRT.
The long-term correct tve measures (i.e., design modifications) will substantially reduce the likelihood of CRT cracking in the futura.
The new drive design is being utilized for all BWR-6 plants and some BWR-4 and -5 plants which are under construction.
GE will provide further results from in-reactor testing as they become available.
1750 031
~ 25 -
TABLE 1 COLLET RETAINER TUBE INSPECTIONS NUMBER OF DRIVES NUMBER NUMBER REACTOR IN SERVICE EXAMINED CRACKED Brunswick 1 137 29 4
Cooper 137 38 1
Dresden 2 177 51 12 Dresden 3 177 87 23 Fitzpatrick 137 9
0 Millstone 145 43 0
Monticello 121 51 0
Nine Mile Point 129 34 3
Oyster Creek 137 100 0
Peach Bottom 2 185 21 7
Pilgrim 145 27 7
Quad Cities 1 177 15 0
Browns Ferry 1 185 17 7
Vermont Yankee 89 16 8
Overseas Reactors 763 241 6
Reactors for which no performed *gtion was CRT examin 1159 0
_0 4000 779 78 Note:
Ten percent of the collet retainer tubes examined were found to have cracks.
Percentage cracked remains unchanged from earlier inspections.
1750 032
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Com piece Design 1750 042
APPENDIO SAMPLE OF TYPICAL TECHNICAL SPECIFICATIONS LIMITING CONDITION FOR OPERATION 3.3 REACTIVITY CONTROL Applicability:
Applies to the operational status of the control rod system.
Objective:
To assure the ability of the control rod system to control reactivity.
Specification:
A.
Reactivity Limitations 1.
Reactivity Margin - Core Loading The core loading shall be limited to that which can be made subcritical in the most reactive condition during the operating cycle with the strongest operable control rod in its full-out position and all other operable rods fully inserted.
1750 043
2.
Reactivity Margin - Stuck Control Rods Control rod drives which cannot be moved with control rod drive pressure shall be censidered inoperable.
The control rod directional control valves for inoperable control rods shall be disarmed electrically and the rods shall be in such positions that Specification 3.3.A.1 is met.
In no case shall the number of non-fully inserted rods disarmed be greater than eight during power operation.
If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure the reactor shall be brought tc a shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless investigation demonstrates that the cause of the failure is not due to a failed control rod drive mechanism collet housing.
B.
Control Rod Withdrawal 1.
Each control rod shall be coupled to its drive or completely inserted and the control rod directional control valves disarmed electrically.
However, for purposes of ramoval of a control rod drive, as many as one drive in each quadrant may be uncoupled from its control rod so long as the reactoa is not in the shutdown or refuel condition and Specification 3.3.A.1 is met.
1750 044
2.
The control rod drive housing support system shall be in place during power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met.
SURVEILLANCE REQUIREMENT 4.3 REACTIVITY CONTROL Applicability:
Applies to the surveillance requirements of the control rod system.
Objective:
To verify the ability of the control rod system to control reactivity.
Specification:
A.
Reactivity Limitations 1.
Reactivity Margin - Core Loading:
Sufficient control rods shall be withdrawn following a refueling outage when core alterations were performed to 6emonstrate with a margin of 0.33% AK that the core can be 1750 045
made subcritical at any time in the subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.
2.
Reactivity Margi.1 - Stuck Control Rods Each partially or fully withdrawn operable control rod shall be exercised one notch at least once each week.
This test shall be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is continuing with three or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out.
The surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods has been reduced to less than three and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.
B.
Control Rod Withdrawal 1.
The coupling integrity shall be verified for each withdrawn control rod as follows:
1750 046
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when the rod is fully withdrawn the first time a.
subsequent to each refueling outage or after maintenance, observe that the drive does not go to the overtravel position; and b.
when the rod is withdrawn the first time subsequent to each refueling outage or after maintenance, observe discernible response of the nuclear instrumentation; however, for initial rods when response is not discernible, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumentation response.
2.
The control rod drive housing support system shall be inspected after reassembly and the results of the inspection shall be recorded.
1750 047
.