ML20091C472

Proposed TS Sections 6.6-2 & 6.6-2a Re Analytical Methods Used to Determine Core Operating Limits

ML20091C472
Person / Time
Site:	Quad Cities, LaSalle
Issue date:	03/31/1992
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20091C469	List: ML20091C468 ML20091C472
References
NUDOCS 9204030230
Download: ML20091C472 (12)

Text

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9 AII6GlMENLB PROPOSED CilANCES TO APPENDIX A, lECHNICAL SPECIFICATIONS Of FACIL11Y OPERATING LICEP.SES DPR-29 AND DPR-30 Of3:29 D?l-30 6.6-2 6.6-2a s

9204030230 920331 PDR ADOCK 05000254 1 p PDR ZHlD/1516/8

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l l QUAD-CITIES DPR-29 whole body dose received from external sources : hall be assigned to specific major work functions.

3. Monthly Operating Report Routine reports of operating statistics and shutdown experientta shall be submitted on a monthly basis to the Director, Office cf Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the appropriate Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report. In addition, any changes to the DDCM shall be submitted with tne Monthly Oper.ating Rr" ort within 90 days of the ef fective dai4 of the change.

A report of major change to the radioactive waste treatmer,t systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the onsite review function. If such change is re-evaluated and noi, installed, notificatiren of cancellation of the change should

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be providad to the NRC.

4. Core Operating Limits Report

a. Core operating limits shall be established and documented in the CORE OFERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for tne following:

(1) The Rod Withdrawal Block Monitor Upscale Instrumentation Setpoint for Table 3.2-3 of Specification 3.2.C and for Specification 3.6.H.

(2) The overall average of the 20% insertion scram time data for Specification 3.3.C.

(3) The Average Planar Heat Generation Rate (APLHGR) for Specification 3.5.1.

(4) The Linear Heat Generation Rate (LHGR) for Specifica- j tion 3.5.J. l (5) The Minimum Critical Power Ratio (MCPR) for Specifica- I tion 3.5.K and 3.6.H.

(6) The K core flow MCPR adjustment factor for Specifica- /

7 s tion $.S.K. ,

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b. The analytical methods used to determine the core operating i h; limits shall be those previously reviewed and approved by NRC  !

in NEDE-24011-P-A, General Electric Standard Application for '

Reactor Fuel (latest approved revision). c  :

- - - A 6.6-2 Amendment No.

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1 INSERLA  !

b. The analytical methods used to determine the core operating limits shall 4 be those previously reviewed and approved by the NRC in the latest approved revision or supplement cf the topical reports describing the methodology. For Quad Cities Uritt 1, the toptral reports are:

(1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor fuel," (latest approved revision).

(2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of BHR Nuclear Design Methods," (latest approved revision).

(3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1,

" Benchmark of BHR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," (latest approved revision).

(4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2

" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

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, OUAD CITIES

, DPR-30 .

_ b. The anaIytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of the tcpleal reports describing the methodology.

For Quad Cities Unit 2, tt:a topical reports are:

(1) NEDE-24011-P A, " General Electric Standard Application for Reactor Fuel,' (latest approved revision).

(2) Commonwealth Edison Topical Report NFSR 0985, "Bonchmark of BWR Nuclear Design Mathods," (latest opproved revision).

1 (3) Commonwealth Edison Topical Report NFSR 0085, Supplement 1,

=

• Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisors, i " (latest approved revision).

(4) Commonwealth Edison Topical Roport NFSR 0085, Supplement 2,

" Benchmark of BWR Nuclear Design Methods Neutronic Licensing Analyses," (latest approved revision).

c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS

- limits, nuclear limits such as shutdown rnargin, and transient and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any mid cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector, r.

I-1976H ~ 6.6 2a Amendment No.

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4 AUACUMEMLC PROPOSED CilANGES TO APPEND 1X A, 1ECHNICAL SPECIFICATIONS of FACILITY OPr. RATING LICENSES NPf-11 and NPT-18 NPL .U NEE-18 6-25 6-25 ZNLD/1516/9-

9 ADMINISTRATIVE CONTROLS Semiannual Radioactive Effluent Release Report (Continued)

(1) The Average Planar Linear Heat Generation Rate (APLHGR) -

for Technical Specification 3.2.1.

(2) The minimum Critical. Power Ratio (MCPR) (including-20% ^

scram tirne, tau (t), dependent MCPR limits, and K, core flow MCPR adjustment factors) for Technical Specification 3.2.3. '

(3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4. -

.._.w' (4) The Rod Block Monitor Upscale Instrumentation Setpoints ,

1. ' for Technical Specification Table 3.3.6-2, ."

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$ ' ~[  % b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC -

z in NEDE-24011 P-A, General Electric Standard Application for '

Reactor Fuel (latest approved revision). _ ,

c. The core operating limits shall be determined so'that all applicable limits (e.g., fuel thermal-mechanical limits, core -

thermal-hydraulic limits, ECCS limits, nuclear limits such as '

shutdown margin, and transient and accident analysis limits) of -

the safety analysis are met. ,,

d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shail be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory .

Commission Decement Control Desk with copies to the Regional Administrator and Resident Inspector.

E. Deleted LA SALLE UNIT 1 6-25 AmendmentNo.)[

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. l INSERILA

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the  !

methodology. For LaSalle County Station Unit 1, the topical reports are:

(1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor fuel," (1atest approved revision),

(2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of BWR Nuclear Design Methods," (latest approved revision).

(3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1,

" Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," (latest approved revision).

(4) Commonwealth Edison lopical Report NFSR-0085, Supplement 2,

" Benchmark of BWR Nuclear Design Methods - Neutronic licensing Analyses," (latest approved revision).

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ADMINISTRATION CONTROLS _

Semiannual Radioactive Effluent Release Report (Continued)

(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

(2) The minimum Critical Power Ratio (MCPR) (including 20% I scram time, taL (t), dependent MCPR limits, and Kf core flow riCPR adjustment factors) for Technical Specification .

3.2.3. '

(3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4.

ry [6

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W (4) The Rod Block Monitor Upscale Instrumentation Setpoints g for Technical Specification Table 3.3.6-2. >

r b. The analytical methods used to determine the core operating

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limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for t __ Reactor fuel (latest approved revision).

c. The core: operating limits shall be determine so that all ,

applicable limits (e.g. , fuel thermal-mechanical 1imits, core /

thermal-hydraulic limits. ECCS Limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of ,

.the safety analysis are met. .

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d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions'or supplements thereto, shall be provided upon -

issuance, for each reload cycle, to the U.S. Nuclear Regulatory '

Commission Document Control Desk with copies to the Regional -

Administrator and Resident inspector. '

B. Deleted, s

LA SALLE - UNIT 2 6-25 Amendment NJ.754

JMSERT_A

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the methodology. For LaSalle County Station Unit 2, the topical reports are:

(1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor fuel, (latest approved revision).

(2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of BWR Nuclear Design Methods," (latest approved revision).

(3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1,

" Benchmark of BWR Nuclear Design Methods - Quad Citics Gamma Scan Comparisons," (latest approved revision).

(4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2,

" Benchmark of BHR Nuclear Design Mothods - Neutronic Licensing Analyses," (latest approved revision).

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1. AllACHMMLD SIGNIFICANT llAZARDS CONSIDERATION Commonwealth Edison Company proposes an amendment to facility Operating Licenses DPR-29, DPR-30, NPf-ll, and NPF-18 to include the NRC approved CECO Topical Report NFSR-0085 so that CECO can perform neutronic-licensing calculations. As discussed in Attachment A, CECO proposes to reference the topical Report in the Technical Specifications of Quad Clites and LaSalle County Stations.

CECO has evaluated the proposed amendment and concluded that it does not involve a significant hazards consir'eration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in ~

accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

The proposed amenament does not involve a significant increase in the probability or consequSnces of an accident previously evaluated becabse:

The NRC approved methodologies to be referenced in the Technical L Specifications are used to evaluate cere operating limits and do not introduce physical changes to the plant. The same spectrum of limiting events will continue to be analyzed using NRC approved methods for each reload. This amendment is administrative in nature and does not affect any accident initiators or initial assumptions used in plant accident

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

The proposed changes do not create the possibility of a new or dif ferent kind of accident from any accident previously evaluated because:

The referenced NRC approved methodologies will continue to be used to analyze limiting transients, and do not introduce any physical changes to the plant or the opention of the facility as described in the FSAR; therefore, the r >osed change does not create the possibility of a new or difft.ent kind of accident from any accident previously evaluated.

The proposed changes do not involve a significant eduction in margin of safety because:

The referenced NRC approved methodologies will continue to ensure fuel design and licensing criteria are met. The proposed amendment is purely administrative in nature and has no effect on g the margin of safety.

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- ATTACllHENT D (continued)

Guidance ha; been provided,in " final Procedures and Standards on No

.Sigt,1ficant Hazards Considerations," final Rule 51 FR 7744, for the application of standards to lic*nse change requests for determination of the: existence of significant hazards considerations. This documpnt ptovides examples of amendments which are and are not corisidered likely to involve significant hazards considerations. These proposed amendments most closely fit the example of a purely administrative change to the Technical Specification (e.(1) of 51 FR 7751).

The proposed amendments do not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations.

Theref)re, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.

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AllACllHLNI E ENVIRONMENTAL ASSESSMENT SIAllMNT APFLICABILITY REVILH Comwonwealth Edison has evaluated the proposed changes against the

- criteria fo" the identificatic'i of Ilcensing and regulatory actions  ;

requiring environnteatal assessmeat in accordance with 10 CFR 51.20. '

it has been deterniined that tief proposed changos meet the criteria for a categorical exclusion as provided under 10 CTR 51.22(()(9). This conclusion has been determined because the proposed cnanges do not pose a significant hazards consideration or to not involve a significant increase in the Amounts, and no significant changes in the a types, of effluents that may be released offsite. This request doe $

not involve a significant increase in individual or cumulative occt'pational radiation exposure, Therefore, the Environmental Assessmer.t Statement is not applicable for these changes. ,

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