ML20210C596

Application for Amend to License DPR-6,changing Tech Spec Section 5.2.1.b by Adding Table of Uncertainty Factors Associated W/Analysis of Core thermal-hydraulic Parameters Heat Flux,Maplhgr & Max Bundle Power

ML20210C596
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	02/04/1987
From:	Buckman F CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20210C509	List: ML20210C506 ML20210C596 ML20210C634
References
NUDOCS 8702090423
Download: ML20210C596 (6)

Text

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CONSUMERS POWER COMPANY Docket 50-155 Request for Change to the Technical Specifications License DPR-6 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in the Facility Operating License DPR-6, Docket 50-155, issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Plant bv changed as described in Section I below:

I. CHANGES

A. Add to Section 5.2.1.(b), Table 1:

" Reload" " Reload" "I2" to the last column heading Il so that it will read II/I2 .

B. Add to Section 5.2.1(b), Table 1:

Footnote "(1)" to Minimum Critical Power Ratio at Normal Operation Conditions, Maximum Heat Flux at Overpower, Maximum Steady-State Heat Flux, Maximum Average Planar Linear Heat Generation Rate, Steady-State and Maximum Bundle Power.

Footnote "(2)" to Maximum Steady-State Power Level.

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C. Add to Section 5.2.1(b), Table 1, the following footnotes:

{

" Note 1: Actual Calculated Critical Power Ratio, Heat Flux, Average Planar Linear Heat Generation Rate, and Bundle l Power shall include uncertainties listed on Table 3 and be bounded by the Technical Specification limits found in Table 1.

Note 2: Maximum Steady-State Power level shall not exceed a power such that any thermal-hydraulic limit listed in Table 1, Table 2, Figure 1, or Figure 2 is exceeded."

D. Add to Section 5.2.1(b), Table 2, the following footnote and add an asterisk after LIMITS in the title:

"*MAPLHGR ratios shall include the uncertainty listed on Table 3."

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2 E. Add to Section 5.2.1(b), Table 2:

" Reload" " Reload" "I2" to the last column heading Il so that it will read 11/I2 .

F. Add to Section 5.2.1(b) the following table after Table 2:

" TABLE 3 UNCERTAINTY FACTORS APPLIED TO THERMAL-HYDRAULIC PARAMETERS Parameter Uncertainty Factor MCPR .1531 Heat Flux 9.4017%

MAPLHCR Ratio .0784 Actual Bundle Power / Max Allowed Bundle Power 3.7661%"

G. Add to Section 5.2.1(b), Figure 1, the following footnote and add an asterisk after RELATION in the title:

"*The maximum bundle peak linear heat generation rates shall include uncertainties listed for MAPLHGR in Table 3."

H. Add to Section 5.2.1(b), Figure 2, the following footnote and add an asterisk after MW in the title:

t

"* Maximum bundle power ratios shall include the uncertainty listed in Table 3."

II. DISCUSSION The proposed Technical Specification Changes requested are editorial in nature to allow inclusion of the Big Rock Point Reload I-2 fuel.

The Technical Specification limits specified in Section 5.2.1(b) define the restricting operating limits for the operation of the Big Rock Point core. Uncertainties which are associated with the parameters bounded by the limits have been applied, and with these uncertainties operating conditions are determined to conform with the Technical Specifications.

The Big Rock Physics Methodology Report - Revision 3 describes the approved methods to determine those uncertainties and the results of the analysis. (The acceptance of this document was approved by NRC letter LS05-83-02-019 from Dennis M Crutchfield to David VandeWalle dated February 9, 1983.) This Technical Specification change reflects the requirement for Big Rock Point conformance to the specified technical specification limits. Uncertainties shall be added to the associated calculated parameters.

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3 Minimum Critical Power Ratio The Critical Power Ratio is expressed as the ratio of Critical Power Level and Actual Power Level. This ratio for operating conditions must satisfy the following relationship:

Calculated Critical Power Ratio - Uncertainty > MCPR Limit OR Calculated Critical Power Ratio .1531 > 1.59 (1.61 for I fuel) OR Calculated Critical Power Ratio must be greater than 1.743 (1.763 for I fuel)

Heat Flux Heat Flux must satisfy the following relationship:

Calculated Heat Flux + uncertainty < Maximum Heat Flux Allowed per Technical Specification OR Calculated Heat Flux + Uncertainty < 322,100 BTU /HR - FT2 OR Calculated Heat Flux (1 + 9.4017%) < 322,100 BTU /HR - FT2 OR Calculated Heat Flux < 294,419 BTU /HR - FT 2 MAPLHGR i

The ratio of actual Maximum Average Planar Linear Heat Generation Rate to allowed Maximum Average Planar Linear Heat Generation Rate must satisfy the following relationship:

^

ff ff T + Uncertainty < 1; uncertainty = .0784 OR MAPLHGR Ratio < 92.16%

Bundle Power The ratio of actual calculated bundle power to the allowed bundle power of 5.2.1(b) Figure 2 must satisfy the following relationship:

Actual Bundle Power OR Allowed Bundle Power + Uncertainty < 1 Bundle Power Ratio < 96.23%

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4 Plant Core Follow Techniques To provide a constant assurance of compliance to the Technical Specification thermal-hydraulic limits, the plant reactor engineering group analyzes each notch to be withdrawn before it is released for actual withdrawal. This analysis is based upon updated and current fuel exposure, void fraction history, power history, control rod depletion, and control rod patterns.

Empirical verification for the accuracy of the code used for these calculations is assured through periodic comparison to flux shape measurements.

The thermal-hydraulic properties, with associated uncertainties, are compared to and must meet the requirements of Technical Specifications before the analyzed notch is released to operators.

This notch by notch analysis is initiated at the beginning of each cycle before thermal-hydraulic parameter to Technical Specification margins are less than 20% and continues throughout the cycle.

III. ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The mechanical, thermal-hydraulic, and neutronic analysis for Big Rock Point Keload I-2 is the same as that for Reload I-1. The design report previously issued for Big Rock Point Reload I-1 (Exxon Nuclear Company (ENC) report XN-NF-85-38(P), Rev 0) entitled, " Design Report for Big Rock Point I-1" is applicable for Reload I-2. This reload does not contain any fuel assemblies significantly different from those previously found acceptable by the NRC.

This change does not involve a significant increase in the probability or consequence of an accident previously evaluated because the limits are derived in a manner identical to that described'in Exxon Nuclear Corporation (ENC) report XN-NF-79-32, revision 1, Big Rock Point LOCA Analysis is using the ENC WREM NJP-BWR ECCS Evaluation Model - MAPLHGR Analysis.

The Technical Specification change includes uncertainty values assoc-iated with the thermal-hydraulic limits specified in Tables 1 and 2 and Figures 1 and 2. This change does not represent a departure from Big Rock Point's current methodology. Rather, it incorporates within the Technical Specifications uncertainties documented in the Big Rock Point

" Physics Methodology Report" Revision 3. The limits observed in operat-ing the core have been inclusive of these uncertainties. " Physics Methodology Report" Rev 3 has been approved by NRC letter from Dennis M Crutchfield to David J VandeWalle (LS05-83-02-019) dated February 9, 1983. The possibility of an accident of a different type previously evaluated is not created. This change does not involve any change in our methodology. The KN-NF-79-21 report covers the required spectrum of break locations, sizes and configurations for the Big Rock Point Plant. The addition of uncertainty factors and footnotes to the Technical Specifications only clarifies Big Rock Point Methodology for controlling operating parameters and does not represent a change in the margin of safety.

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5 Physics design calculations are performed prior to each cycle in accordance with the Physics Methodology Report Rev 3. Specific areas analyzed in this design package include core loading patterns, fuel exposures, control blade depletion, power distributions, thermal margins, reactivity limits, withdrawal sequence, rod and notch worths, scram

-behavior, power escalation rate, accident analysis, and reactivity coefficients. This package verifies that safety parameters are within values assumed in the Reference Transient Analysis (" Plant Transient Analysis of the Big Rock Point Reactor" dated July 1981 submitted by Consumers Power Company letter from RA Vincent to DM Crutchfield dated July 15, 1981 and approved by NRC SERs in letters dated 10/16/81, 2/8/82, 3/4/82, 3/11/82, 4/7/82, 6/18/82 and 6/25/82.)

As discussed in the Reference Transient Analysis document as modified by our Technical Specification change request dated December 20, 1982 and approved by NRC SER dated May 18, 1984 for conversion from a MCHFR limit to a MCPR limit, the transient causing the largest change fa critical power ratio is a turbine trip without bypass valve operation.

It was shown that maintaining thermal margins within the Technical Specification limits would ensure safe operation for this transient and preclude DNB.

LOCA analysis establishing MAPLHGR limits for the most limiting break have been performed by Exxon Corporation. (XN-NF-78-83 Rev 1 XN-NF-79-21, XN-NF-85-78). This most limiting break was identified to be a 0.375 FT2 split break between a recirculating pump and the downstream butterfly valve. Compliance with these limits ensure excessive cladding temperatures are not reached.

Maximum allowed fuel bundle power is based upon flow distribution requirements made upon the ECCS system. The bundle power is limited to assure that the core spray cooling supplied to each channel following a LOCA is sufficient to remove the decay heat generated by a fuel bundle l

at that maximum power level.

l Heat Flux limit is established to prevent fuel center line melting at l operating and overpower conditions.

Consequently this proposed change does not involve a significant hazards consideration.

IV. CONCLUSION The Big Rock Point Plant Review Committee has reviewed this Technical l Specification Change Request and has determined that this change does l

not involve an unreviewed safety question and therefore involves no l

significant hazards consideration. This change has also been reviewed 1

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6 under the cognizance of the Nuclear Safety Board. A copy of this Technical Specification Change Request has been sent to the State of Michigan official designated to receive such Amendments to the Operating License.

CONSUMERS POWER COMPANY By (M F W Buckman, Vice Presiden't Nuclear Operations Sworn and subscribed to before me this 4th day of February 1987.

b S Elaine E Buehrer, Notary Public Jackson County, Michigan My commission expires October 31, 1989 l

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