ML17202V011
| ML17202V011 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/06/1991 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML17202V012 | List: |
| References | |
| NUDOCS 9103110002 | |
| Download: ML17202V011 (6) | |
Text
ATTACHMENT C PROPoSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACIL:IlY OPERATING LICENSE DPR-25 REVISED PAGES UNIT 3 <DPR-25) 1/2. l-l 6-19
0 c 1.1 SAFETY LIMIT FUEL CLADDING INTEGRITY Applicability:
The Safety Limits established to preserve the fuel cladding integrity apply to these variables which monitor the fuel thermal behavior.
Objective:
The objective of the Safety Limits is to establish limits below which the integrity of the fuel cladding is pre-.
served.
Spec if i cat ions:
A.
Reactor Pressure greater than SOO psii and Core Flow greater than 10 of Rated.
The existence of a minimum critical ~
ratio (MCPR) 1 ess than.
] sha 11 constitute a violation of the MCPR fuel cladding integrity safety limit.
When in Single Loop Operation, the MCPR safety limit shall be increased by 0.01.
DRESDEN III DPR-2S Amendment No. )13, IW, S4.
2.1 LIMITING SAFETY SYSTEM SETTING FUEL CLADDING INTEGRITY Applicability:
The Limiting Safety System Set-tings apply to trip settings of the instruments and devices which are provided to prevent. the fuel cladding integrity Safety Limits from being exceeded.
Objective:
The objective of the Limiting Safety System Settings is to define the level of the process variables at which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.
Specifications:
A.
1/2.1-1 Neutron Flux Trip Settings The limiting safety system trip settings shall be as specified below:
- 1.
APRM Flux Scram Trip Setting (Run Mode)
When the reactor mode switch is in the run position, the APRM flux scram setting shall be:
S less than or equal to [.SSW + 62] during
.Dual LooB Operation or S less than or equal to [.SSW +SS.SJ dur-ing Sing9e Loop Opera-tion with a maximum setpoint of 120% for core flgw equal to 98 x 10 lb/hr and greater, where:
S - setting in percent of rated thermal power.
DRE. III Amendment No. ~
DPR-25 6.0 ADMINISTRATIVE CONTROLS (Cont'd.)
- 3)
The Local Steady State Linear Heat Generation Rate (LHGR) for Specification 3.5.J.
- 5) The Minimum Critical Power Operating Limit for Specification 3.5.L.
This includes rated and off-rated flow* conditions.
- 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 Dresden Unit 3, the I
, 1 topical reports are:
- InseY+ A ~Yom r---=::.:~~~~~~~~~~~~~~~~~~~~"'--f*
follow1n9 po.5e j
- 1) U<H NF 512(P)(A). "XN 3 CPitieal Pewel" Cel"Pelati;R.I I
=tfi'.:'.::* r::."' J.-3-1:;::~:~::>t!
- 8:~~=: :::!:a;.::!:!:~~ Pewe* 1
- Tt1ser+ 'c' fv-owi
-f'ollow/n3 fO.je
- c.
- d.
- 3) XN-NF-79-7l(P)(A), "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors".
- 4) XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors".
- 5)
XN-NF-85-67(P)(A), "Generic Mecahnical Design for Exxon Nuclear Jet Pump Boiling Water Reactors Reload Fuel".
- 6)
XN-NF-81-22(P)(A), "Generic Statistical Uncertainty Analysis Methodology".
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.
The Core Operating Limits Report, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance to the NRC Document Control *Desk with
- copies to the Regional Administrator and Resident Inspector.
B.
Reportable Events Reportable events will be submitted as required by 10 CFR 50.73.
6-19
Inserts for Page 6-19 CDPR-25)
Insert 'A' ANF-1125
<A>, "Critical Power Correlation - ANFB". Insert 'B' ANF-524
<A>, "ANF Critical Power Methodology for Boiling Hater Reactors". Insert 'C'
- 7)
ANF-913
<A>, "COTRANSA2: A Computer Program for Boiling Hater Reactor Transient Analyses". ZNLD743/25 ATTACHMENT D EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison Company <CECo) proposes an amendment to Facility Operating License DPR-25 <Dresden Station Unit 3) to reflect the use of new Advanced Nuclear Fuels' <ANF) reload licensing methodologies beginning with Cycle 13 for Unit 3. As discussed in Attachment 'A' <Description of Amendment Request), CECo proposes to reference these NRC-approved methodologies and incorporate the resultant increase in the Minimum Critical Power Ratio <MCPR) Safety Limit (from 1.05 to l.08). CECo has evaluated the proposed amendment and concluded that it does not involve a significant hazards consideration. rn accordance with lO CfR
- 50. 92(C):
The proposed amendment does not involve a significant increase in the probabtlity or consequences of an accident previou~ly evaluated. The NRC-approved methodologies to* be referenced in the Technical Specifications are used to evaluate core operating limits and do not introduce physical changes to the plant. ANF will continue to analyze the same spectrum of limiting events for each reload under the new methodology. The increase in the MCPR Safety Limit adequately accounts for the effects of the new methods and potential effects of channel bow, and will continue to maintain fuel cladding integrity by ensuring that 99.9% of the fuel rods will avoid transition boiling during limiting anticipated operational occurrences. Therefore, the changes do not effect the probability or consequences of accidents previously evaluated~ The proposed amendment does not create the possibility of a new or different kind of acctdent from any' accident previously evaluated. The referenced NRC-approved methodologies. will cont1nue to be used to analyze limiting transfents, and do not introduce any physical changes to the plant; therefore, the possibility of a new or different kind of accident is not created. Stmilarly, the basis of the MCPR Safety Limit has not been changed and will continue to maintain fuel cladding integrity during limiting antici*pated operational occurrences. The proposed amendment does not involve a significant reduction in a margin of safety. The referenced NRC-approved methodologies will continue to ensure fuel -tlesign and- -1 kenslng crHeria -are met. The lncrease i-n the MCPR -Safety Limit reflects the new methods, bounds the effect of channel bow for Cycle 13, and provides additional conservatism to facilitate future reload ltcensing reviews under the provisions of 10 CFR 50.59. Therefore, the margtn between the safety limit and potential fuel failure after the onset of transition boiling is not decreased. ZNLD143/26 ATTACHMENT E ENVIRONMENTAL ASSESSMENT The proposed amendmerit to the Unit 3 Technical Specifications reflects the use of new, NRC-approved reload licensing methodologies beginning with Cycle 13, and the resultant increase in the MCPR Safety L1 mit. The new methodologies and MCPR Safety Limit increase will maintain the current margin of safety and fuel cladding integrity so that no environm~ntal impact will result. Additionally, the proposed amendment does not involve a significant hazards consideration as previously presented in Attachment 0. ZNLD743/27