ML20044B196

Application for Amend to License DPR-51,changing Tech Spec 4.7.2, Control Rod Program Verification, by Removing Unnecessary Restriction in Method of Verification of Proper CRD Patching

ML20044B196
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/10/1990
From:	Carns N ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20044B197	List: ML20044B196 ML20044B198
References
1CAN079003, 1CAN79003, NUDOCS 9007180087
Download: ML20044B196 (7)

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• Buts" Carns m Nwe ti.t r y L July 10, 1990

. ICAN079003 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, DC 20555

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Technical Specification Change Request -

Control Rod Verification Program (TS 4.7.2)

Gentlemen:.

We hereby request an amendment to Operating License No. DPR-51 for ANO-1 with the attached submittal .of a proposed change to Technical Specification Section 4,7.2, " Control Rod Program Verification". The proposed change would remove an unnecessary restriction in the method of verification of proper control rod drive patching, and would also more accurately reflect the conditions under which patch verification is required.

.A copy of the proposed change is attached for your review and approval.

The circumstances of this proposed amendment are not of an exigent or emergency nature; however, it is our desire to implement this change as soon as possible.

In accordance with 10CFR50.91(a)(1), we have evaluated the proposed change using the criter'a in 10CFR50.92(c) and has determined that said change involves no.sigr.ificant hazards consideration. The bases for this determination Lre also attached for your review.

We request that the effective date of this change be 30 days after NRC issuance of the amendment to allow for distribution and procedural revisions i necessary to implement the change.

Very truly yours, NSC/1g Attachments

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9007180087 900710 PDR ADOCK 05000313 OO[

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U. S. NRC P. age 2 July 10, 1990 cc: Ms. Greta Dieus, Director Division of Environmental Health Protection State Department of Health 4815 West Markham Street Little Rock, AR 72201 Mr. Robert Martin U. S. Nuclear Regul story Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlin0 ton, TX 76011 NRC Senior Resident Inspector Arkansas Nuclear One - ANO-1 & 2 Number 1, Nuclear Plant Road Russellville, AR 72801 Mr. Thomas W. Alexion NRR Project Manager, Region IV/ANO-1 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-D-18 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Mr. Chester Poslusny HRR Project Manager, Region IV/ANO-2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-D-18 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852

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STATE OF ARKANSAS )

) SS COUNTY OF LOGAN )

I, N. S. Carns, being duly swo.n, subscribe to and say that I am j Vice President, Operations ANO for Entergy Operations, Inc., that I have full authority to execute this oath; that I have read the document numbered ICAN079003 and know the contents thereof; and that to the best of my knowledge, information and belief the statements in it are true.

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T l N. S. Carns SUBSCRIBED AND SWORN T0 before me, a Notary Public in and for the County and State above named, this /M day of dd/ ,

1990.

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PROPOSED TECHNICAL SPECIFICATICN CHANGE LICENSE AMENDMENT REQUEST IN THE MATTER OF AMENDING LICENSE NO. OPR-51 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 9 u ~

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PROPOSED CHANGE The proposed change would revise ANO-1 Technical Specification (TS) 4.7.2.1 (TS page 104) and the corresponding Bases to provide an improved method of assuring proper control rod drive mechanism patching during shutdown conditions by removing the two-inch upper travel limit. The proposed change would also ,

more accurately reficct the conditions under which patch verification is required.

BACKGROUND The Control Rod Drive (CRD) System translates the reactor control signal, either automatic or manual to a linear motion of the control rods.

Primarily, the CRD function is to maintain and control the nuclear chain reaction; it also serves an equally important function with respect to the Reactor Protective System. Upon a reactor trip signal, power to the control rod drive mechanisms (CRDMs) is de-energized, thus allowing the control rod assemblies (CRAs) to fall by gravity into the core, reducing neutron population, causing the core to go suberitical.

There are 68 CRDMs, one for each control rod assembly. The mechanisms are divided into eight separate groups. Groups 1 through 4 are referred to as safety groups. During reactor startup and all periods of operations, these groups are maintained at their full out position. Their function is to provide adequate shutdown capability upon reactor trip. Groups 5, 6, and 7 are the regulating groups; they are used to gain criticality and to control 5 the power output of the core. Group 8, partial length or axial power shaping control rods (APSRs), are used to control a,.lal flux distribution.

Each CRDM has a position indicator assembly. This assembly indicates the l vertical position of the control rod within the core by determining the l absolute position of the leadscrew, Reed switches, enclosed in a square fiberglass housing, mounted on the outside of the upper motor tube mark the position of the leadscrew by the magnetic field induced as the permanent magnet passes the immediate vicinity of the switch. As the magnet continues ,

l on past, the reed switch returns to its normally open position.

In addition'to the absolute position indication, each CRDM has a relative l position indication. The relative position indication receives its signal 1 by monitoring the input pulses to each PRDM motor. Every other phase of l the six phase input leads is measured, and this signal is converted into a relative position indication read-out on the position indication panel (PIP). The analog PIP can display either the absolute or relative position indication.

A patch panel is used to reassign control of specific drives to a different power supply. CRDM Group 8 is permanently patched through the panel and cannot he reassigned different power supplies. Repatching is completed while the reactor is shutdown, normally during a refueling outage. The instrumentation cables and the power leads must be reconnected during a patch; both are done separately.

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DISCUSSION As descriveu above, the control rod drive patch is changed only during a refueling shutdown. There are some maintenance activities which do not disturb the patch, and it is not practical to perform the patch verification each time the patch panels are opened and then relocked, as ;s strictly required by the present languaga of TS 4.7.2.1. ANO therefore proposes to modify this requirement to more accurately reflect the conditions under which a patch verification is required (i.e. , when the patch is reconnected).

Af ter each repatching, ANO-1 TS 4.7.2.1 requires a comparative check to verify that a designated control rod (by core position) is operating in its programmed functional position and group. To demonstrate that the proper rods are responding as shown on the unit computer printout, each CRDM is exercised by a movement of two inches or less. AP&L desires to remove this upper travel limit and re-word the specification to indicate "an amount of travel suf ficient to verify proper patching".

Recent modifications to some of the position indicator assemblies have been made. These modifications include replacing the existing type "B" rod position indicators with the new type "R4C". For both the present and previous position indicators, the indication on the unit computer changes from "0" incrementally at the first state change of the reed switch matrix which represents the "0" position. Due to physical construction (i.e., location and spacing of the reed switches vertically along the CRDM) this occurs ideally at 1.5 inches for the old type "B" position indicator. For the new type "R4C", this occurs ideally at 1.0 inches.

When exercising the CRDM per TS 4.7.2.1, halting rod movement at the first indicated rod position change on the computer will provide an " indicated" movement of less than 2.0 inches. The analog errors (12.0" for type "B";

12.5" for type "R4C") do not apply to this function since the required function is simply indication of movement. However, removal of the 2" upper limit will assure the intent of TS 4.7.2.1 is not misinterpreted with respect to inclusion of the system accuracy limits. Furthermore, it is possible that a faulty first reed switch could exist, which could result in exceeding the 2" limit before the operator is aware a problem exists. Removal of this upper limit will improve the repatch verification process while eliminating any potential confusion in regard to system accuracy.

The proposed Technical Specification resembles the B&W Standard Technical Specifications (STS) in that no upper travel limit is established._ The major concern during this verification is the prevention of inadvertent criticality. With TS 4.7.2.1 as proposed, CRDM movement could actually be exercised over the full range of travel. This could be done while maintaining a subcritical mode. Reactivity conditions are controlled by several specifications. During power operations, TS 3.5.2 requires a shutdown margin of at least 1% Ak/k to be maintained, assuming the highest worth control rod is fully withdrawn. In other modes of operation, TS 1.2 requires a minimum 1% Ak/k shutdown margin. Minimum conditions for criticality are provided in TS 3.1.3. In addition to t.he requirements of the above specifications, reactivity conditions are controlled administratively by ANO procedure.

DETERMINATION OF SIGNIFICANT HAZARDS ANO has performed an analysis of the proposed change in accordance with 10CFR90.91(a)(1) regarding no significant hazards consideration, using the standards in 10CFR90.92(c), as follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed change would not increase the probability or consequences of any accident previously evaluated since the proposed change would not cause the plant to exceed the shutdown margin.

The proposed change is bounded by the Rod Ejection Accident discussed in FSAR section 14.2.2.4. When the reactor is subtritical, the boron concentration is maintained at a level that ensures that the reactor is at least one percent subcritical with the control rod of greatest worth fully withdrawn from the core. Thus, a rod ejection will not cause a nuclear excursion when the reactor is suberitical and all the other rods are in the Core.

Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated The proposed change would not create the possibility of a new or dif ferent kind of accident from any previously analyzed since it would not introduce new systems, failure modes or other plant perturbations. Only the mode of operation is affected by this change. Eacn CRDM could be exercised over its full range of travel instead of within its current 2 inch range.

Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety The proposed change would not involve a significant reduction in the margin of safety since the ability to control criticality would continue to be maintained with the same shutdown margin. If patching problems occur, the reading from the additional reed switches (over the full range of the position indicator assembly) would assist troubleshooting and repair operations.

Therefore, based on the reasoning presented above and the previous discussion of the amendment request, ANO has determined that the requested change does not involve a significant hazards consideration.

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