ML20002C967

Advises That Addition of Control Rod Drive Support Structure to Reactor Would Upgrade Engineered Safeguards Sys

ML20002C967
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/14/1969
From:	Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101150491
Download: ML20002C967 (7)

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General Omcas: 212 West Michigan Avenue. Jackson. M6chigan 49201. Area Code S17 788-0550 g

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January 11+, 1969

,'o g [ \\ N m-Dr. P. A. Morris Division of Reactor Licensing Uniteu States Atomic Energy Commission Washingtor., D. C.

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Dear Dr. M'rris:

Attention: Mr. D. J. Skovholt In line with our continuing policy of reviewing the op-erations and design of the Big Rock Point Nuclear Plant in the light of recent experience and current design criteria, we have come to the conclusion that the addition of a control rod drive support structure to the B.

Rock Point reactor would upgrade the engineered safeguards' rptems of the plant.

The control rod drive support structure is designed to prevent the ejection of a control rod drive in the unlikely event of a thimble failure. Consumers Power Company believes the addi-tion of this device vill make Big Rock Point equivalent to BWR plants currently authorized for construction with respect to the potential excursion caused by a control rod ejection. In our

- opinion, the addition of the control rod drive support structure does not involve an unreviewed safety quection or a change to the Technical Specifications of License DPR-6, Ibcket No 50-155, issued to Consumers Power Company on May 1,196!+, for the Big Rock Point 3

Plant.

DISCUSSION Design Analyses for the design were performed by General Electric Company and the actual design of the support structure was by Bechtel Corporation using General Electric criteria. Fabrication of the support structure is now in progress. Efforts are now being made to procure materials and fabricate components so that the structure might be installed during the next refueling outage in early 1969 The control rod drive support structure (see attached drawing) consists of a grid of steel plate beams which contain 32 I

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' Dr. P. A. Morris 2

January 14, 1969-supporting assemblies each of which is located directly below the lower.end of the 32 control rod drive mechanism housings. The grid

. structure is spring mounted on four (k) pipe columns, which.are an-chored to the concrete floor by drilled-in expansion type anchors.

The springs, which are mounted within the tops of the pipe columns,-will allow a three-inch maximum vertical displacement of-the grid and the supporting assemblies._ The clearance between the lower end of the control rod drive mechanism housings and each sup-port assembly will be one inch (plus or minus 1/16 inch) at the time of erection of the support assemblies.

Operation of the reactor will cause a downward thermal movement of the bottom end of the control rod drive housing. This movement has been measured during plant operation and found to be

- about 0 70 inch. The same measurement disclosed that the lateral movement at the lower ends of the control rod drive housings was very small and is not considered significant.

Upward thermal movement of the suppogt structure grid will be approximately 0.04 inch (based on a 70 F rise in ambient temperature in the control rod drive access room) between erection time and steady state plant operation. Normal operating clearance between the top of the support assemblies and the bottom of the con-trol rod drive housing is expected to be 0.26 inch during plant operation.

i Each individual supporting assembly is bolted in place in the grid structure and can be removed to provide access for removal, inspection, or maintenance of the individual control rod drive mechanism.

Lateral stability of the structure is provided by tension-compression struts to the perimeter concrete walls of the room.

The objective of the support structure is to prevent the rapid expulsion of a control rod from the reactor under the postu-lated condition of a thimble failure.

The design configuration of the aupport structure is that of a structure positioned below the control rod drives and designed for the maximum force which could be imposed by a ruptured control 1

rod thimble, so that axial motion would be limited or prohibited.

Therefore, even in the event of a circumferential thimble rupture, the control rod would not be ejected from the core.

Design Criteria The following design criteria were considered in the design of the Big Rock support structure:

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Dr. P. A. Morris 3

Janu:ry 14, 1969 1.

The support structure must be' adequate for a "one-time" loading condition.

4 2.

The minimum design loading to be considered-is a function of the sum of the jet force and dead load, times the impact factor, a.

The jet force can be expressed as equivalent to the vessel design pressure, times the gross cross-sectional area of the thimble.

b.

The dead load can be expressed as equivalent to the total weight of the control rod train, thimble, and the column of water over the failed thimble, The impact factor will vary with the amount of travel c.

permitted. The overall load consideration being minimized by the least amount of travel from expulsion to arrest.

g [(PA) + W]

Therefore L=F Where L = Minimum design load Fg = Impact factor P = Vessel design pressure (1,750 psi) 2 A = Area of thimble (n/h)(5 25 )

W = Total of control rod train, thimble, and vessel water weights (1,300 lb) a 3

2 3

g [(175 x 10 )(5 25 )(0 785) + 13 x 10 )

L=F 1 (3916 kips)

L=F 3

On the basis of core reactivity insertion, the suppo' structure must prohibit or limit the total possible control rod tral motion to less than (a normal rod withdrawal increment) three inches.

Thus, the reactivity addition by control rod motion, due to a thimble failure, is less than the reactivity addition of a single, normal withdrawal increment and will not initiate a nuclear excursion.

h.

The support structure design must provide clearance be-tween the thimble flange and the load-bearing support to. prevent con-tact of the two members, thereby allowing for the respective reactor vessel and thimble thermal expansion during plant operation.

1 5

Arrest of the failed thimble should be made at the thimble's lowest flange.

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Dr. P. A. Morris.

January 14, 1969 6.

Provision of easy access to the control rod drive mech-

.anisms to allow individual removal of in-core housings, control rod drive mechanism position indicator-probes (and -cables), as well as the entire control rod drive mechanism for inspection and maintenance must be provided.- The support structure should allow such removal, on a one-at-a-time basis, without disturbance to surrounding drive mechanisms and related equipment.

7 The support structure design provided must be installed within thef present available space in the control rod drive access No obstruction to other structures or equipment or hindrance room.

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of routine inspection and maintenance should be permitted.

8.

Rapidity of individual thimble drive support removal for access to.the drive mechanisms and ease of maintenance should be considered as a design goal.

Stress Analysis Evaluation of the stress levels and the potential for cyclic

- stress fatigue in the reactor vessel and control rod drive thimbles has defined the most critical area to be the welded joint between the thimble and thimble nozzle.

In the event of a failure of the thimble i

or connecting veld, the drive thimble would be ejected downward by a force equal to the vessel internal pressure multiplied by the area of the thimble exposed to that internal pressure. Assuming that the reactor is operating at its design pressure of 1,750 psig, the ejec-tion force on the thimble would be 37 9 kips. The weight of the con-

- trol rod assembly and housing is approximately 13 kips. The total downward force would then be approximately 39 2 kips. considering i

that the control rod drive housing will travel about 0.26-inch down-ward before encountering the resisting support assembly, and also considering that the support assembly will offer a resistance at each corner column of h2,500 pounds per inch of downward travel, calcula-tions reveal that a downward travel of 2.07 inches will result for the case where the failed assembly is nearest a torner column with a corresponding dynamic load factor of 2.29 This will result in a maximum dynamic load of 89 8 kips at that location. For the case where the failed assembly is near the center of the support grid, the deflection will be about 1 37 inches maximum. With a correspon-ding load factor of 2.46, this will result in a maximum dynamic load of 96 5 kips.

I It was assumed, in accordance with the above-mentioned de-

. sign criteria, that only one control rod housing thimble or connecting weld would fail at any one time. The design loading for each partic-ular component was based on assumed failure of that particular control rod drive which causes the highest loading of that component. The single failure assumption, in.our opinion, is proper because we are unable to postulate a reasonable set of mechanisms that would lead i

to multiple, coincident control rod thimble failures.

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Dr. P. A. Morris 5

January 14, 1969 The support structure has been designed at stress levels somewhat above the normal AISC working stress levels, but the maxi-mum stresses have been kept below 90% of yield stress levels in all components. Buckling due to lateral instability has been considered and in all components the buckling stresses are greater than the yield stresses.

HAZARDS CONSIDERATIONS In summary. ve believe the design of the control rod drive support structure is adequate to assure the support of a control rod drive mechanism in the unlikely event of thimble failure. The design of the support structure has adequate margin to satisfy present de-sign criteria.

In our. opinion, the addition of the above-described control rod drive support structure, to the Big Rock Point Nuclear Plant, does not present a significant change in the hazards consid-erations' described or implicit in the Final Hazards Summary Report.

Yours very truly, h

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2 ERC/dmb Robert L'. Haueter.

Attach Assistant Electric Production Superintendent - Nuclear CC: HDI'hornburg Div of Comp 1 USAEC 4

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