ML20059H698

Application for Amends to Licenses DPR-53 & DPR-69, Incorporating Changes to TS by Eliminating TS 3.3.3.2 & Relocating Limitations on Use of Incore Instrument Sys to Ccnpp UFSAR

ML20059H698
Person / Time
Site:	Calvert Cliffs
Issue date:	11/03/1993
From:	Denton R BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20059H702	List: ML20059H698 ML20059H705
References
NUDOCS 9311100177
Download: ML20059H698 (6)

Text

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L BALTIMORE  ;

GAS AND j ELECTRI '

s 1650 CALVERT CtJFFS PARKWAY- LUSBY, MARYLAND 20657-4702 i I

RoscRT E. DENTON j Vict PREstocNT ]

NUCLEAR ENERGf

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November 3,1993 U. S. Nuclear Regulatory Commission  ;

Washington,DC 20555 l ATTENTION: Document Control Desk

SUBJECT:

Calvert Giffs Nuclear Power Plant Unit Nos.1 & 2; Docke: Nos. 50-317 & 50-318 License Amendment Request; Relocation of Incore Instrument '

Reauirements

REFERENCES:

(a) 'Ihe Nuclear Regulatory Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 16,1993 (b) NUREG-1432, Standard Technical Specifications for Combustion Engineering Plants, dated September 1992 (c) CENPD-153-P, Rev 1-P-A, Eva!uation of Uncertainty in the Nuclear l Power Peaking Measured by the Self-Powered, Fixed In-Core l Detector System, dated May 1980 L

Pursuant to 10 CFR 50.90, the Baltimore Gas and Electric Company hereby requests an Amendment to Operating License Nos. DPR-53 and DPR-69 by the incorporation of the changes described below into the Technical Specifications for Calvert Giffs Unit Nos. I and 2.

DESCRIPTION The Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear.

Power Reactors (Reference a) contains four criteria which can be used to determine which-constraints on the design and operation of nuclear power plants a t appropriate for inclusion in the plant's Technical Specifications. The Final Policy Statement also encourages licensees to request changes in their Technical Specifications to implement the Policy.

Use of the Incore Instrument (ICI) System at the Calvert Giffs Nuclear Power Plant is governed by Technical Specification 333.2. This system does not meet the criteria in the Final Policy Statement

. and, consequently, was not included in NUREG-1432,

• Standard . Technical Specifications for Combustion Engineering Plants" (Reference b). This proposed change will eliminate Technical Specification 333.2 and relocate the limitations on the use of the ICI System to the Calvert Giffs i Updated Final Safety Analysis Report (UFSAR).

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.l HACKGROUND l

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The ICI System at Calvert Cliffs consists of 45 neutron detector strings positioned in the guide tube of selected fuel assemblies. Each detector string contains four rhodium neutron detector segments ,

located at 20,40,60, and 80% of core height. The neutron flux indicated by the detector segments is processed by a full-core power distribution system. ,

1 Specifications 3.2.1,3.2.2.1,3.23, and 3.2.4 provide limits on Linear Heat Rate (LHR), Total Planar Radial Peaking Factor (FTy), Total Integrated Radial Peaking Factor (FT,), and Azimuthal Power  ;

Tilt (T ).

q These core power distribution limits reflect the assumptions made in the UFSAR safety analyses and are measured using the ICI System. The ICI System is also used to recalibrate the Excore Neutron Flux Detector System as required by Specification 333.2. Specification 333.2,  ;

"Incore Detectors," provides specific requirements on the number and distribution of incore  :

detectors to be used when measuring the core power distribution limits and when recalibrating the -!

Excore Neutron Flux Detector System.

Baltimore Gas and Electric Company has submitted license amendment requests to relax the limits  :

on incore detector distribution twice in 1993 (March 9 and July 16). Both requests were approved. '  ;

Many licensees have requested relaxation of incore detector system requirements on a cycle-specific -j basis and such requests have been typically approved. The methodology for determining acceptable relaxations of the ICI requirements is well established. For Calvert Cliffs, relaxations are acceptable ,

if the measurement uncertainties are less than those assumed in Reference (c).

On July 16, 1993, the Commission issued a Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (Reference a). The Final Policy Statement contains four criteria which can be used to determine which constraints on the design and operation of nuclear -  ;

power plants are appropriate for inclusion in the plant's Technical Specifications. The ICI Sydem  !

does not m(et any of the four criteria as described below.

Criterion 1 states, " Installed instrumentation that is used to detect, and mdicate in the control room, a significant abnormal degradation of the reactor coolant system boundary" - The ICI System  !

provides no fonction which would indicate a degradation in the Reactor Coolant System boundary.

Criterion 2 states, "A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient Analysis that either assumes the failure of or  :

presents a challenge to the integrity of a fission product barrier." The discussion on Criterion 2  ;

provided in Reference (a) makes clear that core power distribution limits (e.g., hot channel factors)  :

are process variables included in Criterion 2. The proposed change would not eliminate those core l' power distribution limits from the Technical Specifications. De proposed change would relocate' from the Technical Specifications to the UFSAR the details of how those core power distribution limits are measured. The method of measuring a process variable is not included in Criterion 2. . i I

Criterion 3 states, "A structure, system, or component that is part of the primary success path'and ;

which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." He ICI System does not function to mitigate a Design Basis Accident or Transient.

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Criterion 4 states, "A structure, system, or component which operatmg experience or probabilistic  ;

safety assessment has shown to be significant to public health and safety." The ICI System is not- j included in the list of such structures, systems, or components ' included in Reference (a), and  ;

operating expe:ience and the plant-specific probabili.me safety assessment performed for Calvert ,

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Cliffs has not shown the ICI System to have a significant impset on public health and safety.

t Therefore, we have concluded that the requirements on the ICI System are not constraints on design and operation which belong in the Technical Specifications. This same conclusion was reached by the NRC Staff when preparing NUREG-1432, ' Standard Technical Specifications for Combustion Engineering Plants.", Reference (b). NUREG-1432 does not include Technical Specification  !

requirements on the ICI System. .

REOUESTED CIIANGE f Revise Technical Specifications as shown in Attachment (2). Specifically: 'l Remove cycle-specific footnotes from Specification 3.2.1,3 ? ? I and 3.23.  ;

Remove Specification 333.2.

Revise Surveillance 4.2.1.4.b to remove uncertainty factors applied to the ICI Sptem. i Remove the Table of Contents entry and BASES section for Specification 333.2. ,

SAFETY ANAINSIS I

The ICI System provides the spatial measurements of incore neutron flux that are used to verify conformance with the core power distribution limits of the technical specifications. These core power distribution limits are important assumptions in the analysis of accidents and transients, but j the ICI System itself has no safety function. Relocating the requirements for the ICI System from - ,

the Technical Specifications to the UFSAR will not affect the requirements on the monitored limits..

These limits must still be monitored and maintained. Plant procedures will control the number and >

distribution of incore detector measurements necessary to measure the core power distribution limits. Changes to those requirements must be evaluated under 10 Cl A E9.59 which will compare changes to the UFSAR. If proposed changes to the ICI System requirements do not meet the ,

requirements for licensee implementation under 10 CFR 50.59, they must be approsed by license l amendment.

• DETERMINATION OF SIGNIFICANT IIAZARDS

.The proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to not involve a significant hazards consideration, in that operation of the facility in accordance with the proposed amendments:

1. IVould not im>olve a significant increase in the probability or consequences of an accident previously evaluated.

The Incore Instrument (ICI) System is used to measure core power distribution for the j purpose of monitoring the technical specification limits on Linear Heat Rate, Total Planar Radial Peaking Factor, Total Integrated Radial Peakin ; Factor, and Azimuthal Power Tilt.

The ICI System has 'no safety purpose itself; it only measures values which have safety

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significance. No change to the monitored values is proposed. The proposed change will .f relocate requirements on the number and distribution of incore detectors used by the ICI ,

System when measuring these values from the Technical Specifications to the Updated Final  !

Safety Analysis Report (UFSAR). His will allow changes to the requirements to be made  :

without Commission approval as long as the changes meet the' criteria of 10 CFR 50.59. l' Changes to the ICI System requirements which do not meet the criteria of 10 CFR 50.59 must be approved by the Commission by license amendment.

Relocation of the requirements on the ICI System from the Technical Specifications to the UFSAR does not increase the probability or consequences of any accident previously r analyzed because the ICI System is neither a precursor or a mitigator for any analyzed '

accident. The ICI System is not credited in any safety analysis. The values measured by the ICI System are important parameters in many accident analyses; however, this proposed .l change does not remove or affect the limits on these values i e

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Therefore, the proposed change does not involve a significant increase in the probability or i consequences of an accident previously evaluated.  ;

2. Would not create the possibility of a new or different type of accident from' any accident I previously evaluated. i The proposed change does not represent a change in the configuration or operation of the  ;

plant The ICI System will continue to be used to monitor Technical Specification limits on core power distribution. The core power distribution Technical Specification limits are not i changed.

Herefore, the proposed change does not create the possibility of a new or different type of l accident from any accident previously evaluated.

3. Would not involve a significant reduction in a margin ofsafety.

i The ICI System makes no contribution to the margin of safety. The ICI System is used to -

measure core power distribution values which do contain a margin of safety. The limits on these values are not changed.

L Therefore, the proposed change does not involve a significant reduction in a margin of safety.  !

ENVIRONMENTAL ASSESSMENT I t

The proposed amendment would change requirements with respect to the installation or use of a l facility component located within the restricted area as defined in 10 CFR Part 20 or changes to an  :

inspection or surveillance requirement. We have determined that the proposed amendment involves t

no significant hazards consideration, and that operation with the proposed amendment would result in no significant change in the types or significant increases in the amounts of any efiluents that may l be released offsite, and in no significant increase in individual or cumulative occupational radiation ':'

exposure. Therefore, the proposed amendment is eligible for categorical exclusion as set forth in 10 CFR Part 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or ,

emironmental assessment is needed in connection with the approval of the proposed amendment.

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SCIIEDULE i This change is requested to be approved and issued by April 1,1994. However, issuance of this l amendment is not currently identified as having an impact on outage completion or continued plant i operation.

SAFE'IY COMMITTEE REVIEW  !

These proposed changes to the Technical Specifications and our determination of significant hazards  !

have been reviewed by our Plant Operations and Safety Review Committee and Offsite Safety Review Committee. They have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.

i Should you have any further questions regarding this matter, we will be pleased to discuss them with you.

truly yours, l1 STATE OF MARYLAND : l

TO WIT : l COUNTY OF CALVERT  :

I hereby certify that on the 3 day of Novernb ,1993, before a Notary Public of the State of Maryland in and for Calvert b mkthe V subscriber, personally appeared Robert E. Denton, being duly sworn, and states that he is Vice Plesident of the [

Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the '

foregoing response for the purposes therein set forth; that the statements made are true and correct  ;

to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporation. r WITNESS my Hand and Notarial Seal: - '

Notary Public My Commission Expires: aA.</ h , i lyate RED /BDM/dlm Attachments: (1) Unit 1 Technical Specification Revised Pages (2) Unit 2 Technical Specification Revised Pages y +,- -- , v- +

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.l cc: D. A. Brune, Esquire J. E. Silberg, Esquire  ;

i R. A. Capra, NRC - i D. G. Mcdonald, Jr., NRC T.T. Martin, NRC P, R. Wilson, NRC R. I. Mclean, DNR ,

J. H. Walter, PSC -

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