View source for RBG-44035, Application for Amend to License NPF-47,changing TS Safety Limit 2.1.1.2, Reactor Core (Safety Limits), to Increase Two Recirculation Loop MCPR Limit to 1.13.Proprietary Info Encl.Proprietary Encl Withheld

{{Adams
| number = ML20217J036
| issue date = 08/05/1997
| title = Application for Amend to License NPF-47,changing TS Safety Limit 2.1.1.2, Reactor Core (Safety Limits), to Increase Two Recirculation Loop MCPR Limit to 1.13.Proprietary Info Encl.Proprietary Encl Withheld
| author name = Mcgaha J
| author affiliation = ENTERGY OPERATIONS, INC.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000458
| license number = 
| contact person = 
| document report number = RBEXEC-97-088, RBEXEC-97-88, RBF1-97-0240, RBF1-97-240, RBG-44035, NUDOCS 9708130406
| package number = ML19317C583
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 9
| project = 
| stage = Request
}}

=Text=
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==Subject:==
River Bend Station- Unit 1 Docket No. 50-458 License No. NPF                                      License Amendment Request (LAR) 97-18, Change to Technical Specification 2.1.1.2,
                                     " Reactor Core (Safety Limits)"
File Nos.:      G9.5, G9.42
                   - RBEXEC-97-088
                   ' RBF1-97-0240 RBG-44035 Gentlemen:
In accordance with 10 CFR 50.90, Entergy Operations, Inc. (EOI) hereby applies for amendment of Facility Operating License No. NPF-47, Appendix A - Technical Specifications, for River Bend Station (RBS). The proposed change to the Technical Specification Safety Limit 2.1.1.2, " Reactor Core [ Safety Limits)," will increase the two recirculation loop Minimum Critical Power Ratio (MCPR) limit to 1.13 and the single recirculation loop MCPR limit to 1.14. The revised limits are necessary to address Cycle 8 operation.
Entergy previously submitted a similar change to this Technical Specification. LAR 96-42, submitted November 15,1996 and supplemented thereafter, proposed that the Cycle 7 two recirculation loop MCPR limit be changed from 1.07 to 1.10 and the single recirculation loop limit be changed from 1.08 to 1.12. That amendment request is still necessary to address Cycle 7 operation.
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  . .            s          .w Lkee Amendment Request (LAR) 97-16, .
* Change to Technical Specification 2.1.1.2, Reactor Core [ Safety Limits)" -
August 5,1997-RBEXEC-97-088 RBF1-97-0240 RBG-44035 Page 2 of 3 Based on the guidelines in 10 CFR 50.92, Entergy Operations has concluded that this proposed amendment involves no significant hazards considerations. Attschment 2 provides the basis for this determination in a detailed description of the proposed changes, a justification for the proposed changes and the No Significant Hazards Considerations. Attachment 3 is a copy of the marked-up -.
TS pages. Attachment 4 is a copy of the General Electric proprietary analysis in support of this request.
This request has been discussed with the NRR project manager for RBS. It has also been reviewed and approved by the RBS Facility Review Committee and the Nuclear Review Board. If you have -
any questions regarding this request or require additional information, please contact Mr. T. W.
e Gates at (504) 381-4866.
Sincerely, S
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  ..                                                                                   l License Amendment Request (LAR) 97-16, Change to Technical Specification 2.1.1.2,
             . Reactor Core [ Safety Limits]"
             - August 5,1997 RBEXEC-97-088 RBF1-97-0240 '
RBG-44035 Page 3 of 3 cc:     Mr. David L. Wigginton U. S. Nuclear Regulatory Commission
                      ' M/S OWFN 1311-3 Washington, DC 20555 NRC Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington,TX 70611 Department of Environmental Quality Radiation Protection Division                                '
-                       P. O. Box 82135 Baton Rouge, LA 70884-2135

BEFORE TIIE UNITED STATES NUCLEAR REGULATORY COMMISSION LICENSE NO. NPF-47 IN TIIE MATTER OF ENTERGY GULF STATES,INC, CAJUN ELECTRIC POWER COOPERATIVE AND ENTERGY OPERATIONS,INC.
AFFIRMATION I, John R. McGaha, state that I am Vice President-Operations of Entergy Operations, Inc., at River Bend Station; that on behalf of Entergy Operations, Inc., I am authorized by Entergy Operations, Inc., to sign and file with the Nuclear Regulatory Commission, this River Bend Station License Amendment Request (LAR) 97-18, Change to Technical Specification 2.1,1.2,
           " Reactor Core (Safety Limits);" that I signed this letter as Vice President-Operations at River Bend Station of Entergy Operations, Inc.; and that the statements made and the matters set forth therein are true and correct to the best of my knowledge, information, and belief.
N fohn R. McGaha STATE OF LOUISIANA
          -PARISH OF WEST FELICIANA SUBSCRIBED AND SWORN TO before me, a Notary Public, commissioned in the Parish above named, this -5
* day of- lla atuit                              .1997.
0 (SEAL)
N OLLu$ L b N Claudia F. Hurst Notary Public My Commission expires with life

 -       Attachment 2 PageIof5 ENTERGY OPERATIONS INCORPORATED RIVER BEND STATION DOCKET 50-458/ LICENSE NO, NPF-47
                                             " Reactor Core [ Safety Limits]"
(LAR 97-18)
AFFECTED TECHNICAL SPECIFICATION (S):
Technical Specification 2.1.1.2, " Reactor Core [ Safety Limits]"
BACKGROUND:
GESTAR II and Amendment 25 to GESTAR describe how plant / cycle specific Safety Limit hiinimum Critical Power Ratio (SLhiCPR) analyses are used to confimi the calculated SLhiCPR value on a plant / cycle specific basis to assure that 99.9% of the rods in the core will avoid boiling transition. As a result of the upcoming RBS reload core, a new Technical Specification SLhiCPR is proposed for Cycle 8. The General Electric (GE) methodologv used for River Bend Station is the same methodology used by GE to assess SLhiCPR for other BWR's. Similar methodology was used in the SLhfCPR analysis for the previous Cycle (RBG-43326, RBG-43663, RBG-43920).
The SLhiCPR analysis is performed based upon the actual cycle 8 core design and the proposed Technical Specifications changes will be implemented prior to the startup from RBS refueling outage 7, currently scheduled for September 1997.
DISCUSSION OF REQUEST:
The SLhiCPR is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the SLhiCPR is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

aJ                        - Attachment 2                                                                                                :
Page 2 of 5 The SLMCPR is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the -
;.                           occurrence of boiling transition is determined using the General Electric critical power correlations. Details of the SLMCPR calculation are given in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel, GESTAR-II " The SLMCPR incorporates
                            . margin for uncertainty in the core operating state and for uncertainties which are dependent on fuel type, including fuel bundle nuclear characteristics, critical power correlation and manufacturing tolerances.
Plant / cycle specific evaluations are being performed for Cycle 8 using the method described in GESTAR II, and in GETAB. The cycle-specific Safety Limit MCPR evalur. tion has been performed by General Electric (GE) and is also included in this submittal as Attachment 4.
2                          - As stated in Attachment 4, the plant / cycle specific SLMCPR is analyzed with radial power
;.                           distribution maximizing the number of bundles at or near the limits. These methods are similar to those used in the previous cycle SLMCPR analysis.
                            ' Gli has established criteria to determine if the control rod patterns and resulting radial power distributions are acceptable. These criteria are documented in GE's Technical Design Procedures.             ,
4 They include no gross violations of Technical Specification operating limits (e.g. MCPR, MAPLHGR, LHGR), criticality (calculated, normalized Keff near one) ar.d total number of bundles within 0.20 of the MCPR of the core. Different rod patterns were analyzed until the criteria of the above parameters were met.
The SLMCPR analysis results are summarized in the following Table.
L                                                      Comparison of Gell Generic and River Bend Cycle 8 Core Description                                           Generic Gell                   River Bend Cycle 8 Limiting Cycle Exposure Point (MWD /STU)               PHE                            EOC Core MCPR (for limiting rod pattern)                   1.257                           1.316 MCPR Importance Parameter, MIP                         2.13                           4.79               _
R-Factor Importance Parameter, RIP                                                    31.4 %
    ~
g                                                                                   14.3 %
Calculated Safety Limit MCPR                           1.07-                           1.13                .
t                            Correlation of MIP and RIP directly to the calculated SLMCPR have been performed and it is demonstrated that the SLMCPR value of 1.13 is applicable to RBS Cycle 8. Additional 4                            evalustions have been performed by comparing various quantities (Table 1 of Attachment 4) for current RBS Cycle 8 core design and upon which the applicability of 1.13 SLMCPR is verified.
For single loop operations (SLO) the calculated SLMCPR for the limiting case is 1.14 as determined by plant / cycle specific calculations.
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  *<     Attachment 2 Page 3 of 5 This information is consistent with the methods used in the development of the RBS cycle 8 SLMCPR and is based on:
* General Electric Standard Application for Reactor Fuel (GESTAR II), NEDE-24011-P-A-11,
             ' hvember 1995 and Revision 13, August 1996. Note; revision 12 included proposed SLMCPR methods which were withdrawn in amendment 13 therefore, the Revision 11 and 13 material pertinent to the application is unchanged from Revision 11.
* R-Factor Calculation Method for gel 1, GE12 and GE13 Fuel, NEDC-32505P, November 1995.
* Methodology and Uncertainties for Safety Limit MCPR Evaluations, NEDC-3260lP, Class 111, December 1996 e    GENE proposed Amendment 25 to GESTAR 11, submitted December 13,1996. This submittal contains the interim implementing procedures used for RBS cycle 8 SLMCPR.
* General Electric Fuel Bundle Designs, NEDE-31152-P, Revision 5, June 1996.
PROPOSED CilANGE(S):
For Technical Specification 2.1.1.2 the changes to be made are:
Current TS Value           Cycle 7      Cycle 8 Value Value'        (Proposed)"
Two recirculation                1.07              1.10             1.13 loop operation One recirculation                1.08              1.12            1.14
_ loop operation
* Proposed via River Bend LAR 96-42 submitted November 15,1996 to cover Cycle 7 operation. NRC review ongoing.
         "        Required change to support Cycle 8 operation.
A mark-up of the applicable pages is included in Attachment 3.
NO SIGNIFICANT IIAZARDS CONSIDERATION:
The request does not involve a significant increase in the probability or consequences of an -
accident previously evaluated.
m   m

1 l     .
i*-            Attnnment 2 P;ge 4 of 5 The plant / cycle specific SLhtCPRs have been calculated using methods identical to those used by GE to assess the SLhiCPR for other BWRs. Similar methods were used to determine the value of the SLhiCPR for the previous cycle. These methods are within the existing design and licensing basis and cannot increase the probability or severity of an accident. The basis of the SLhiCPR calculation is to ensure that greater that 99.9% of all fuel rods in the core avoid transition boiling and fuel damage in the event of the occurrence of Anticipated Operational Occurrences (AOO) or a postulated accident.
Th SLhiCPR is used to establish the Operating Limit hiinimum Critical Power Ratio (OLhiCPR), Neither the SLhtCPR nor the OLhtCPR are initiators or affect initiators of an accident previously evaluated and therefore changes to the SLhiCPR do not increase the probability of any accident previously evaluated. The proposed changes involve the use of an accepted methodology in calculating the SLhiCPR and, since there it no change in the definition of the SLhiCPR, these changes will not affect the consequences of an,e accident previously evaluated. In addition, the proposed changes do not involve any changa in the way the plant is operated. Existing procedures will ensure that the SLhtCPR is not violated. Therefore, these changes have no effect on the consequences of an accident.
On these bases, there will be no increase in the probability or consequences of an accident previously analyzed as a result the proposed changes.
The request does not create the possibility of occurrence of a new or different kind of accident from any accident previously evaluated.
The proposed changes consist of SLhtCPR calculated from an accepted method of analysis which has been used by many BWRs. These changes do not involve any alteration of the plant and do not affect the plant operation. Neither the SLhiCPR nor the OLhtCPR can initiate an event, therefore a change to the SLhtCPR does not create the possibility of occurrence of a new or different kind of accident from any accident previously evaluated.
The request does not involve a significant reduction in the margin of safety.
The SLhiCPR is a Technical Specification numerical value to ensure that 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated. The proposed SLhiCPR change results from SLhtCPR analysis using the accepted methods as identified in the Attachment.
The margin of safety resides between the SLhiCPR and the point at which fuel fails. hiaintaining the hiCPR above the proposed SLhiCPR will maintain the margia of safety associated with GE's SLhiCPR methodology. Existing plant procedures will continue to ensure that the SLhiCPR is not violated.
3 Therefore, this request does not involve a reduction in the margin of safety.

Attachment 2 Page 5 of 5 Environmental Impact Consideration EOl has reviewed this request against the criteria of 10 CFR 51.22 for environmental considerations. Since this request involves (i) no significant hazard consideration, (ii) no significant change in the types or significant increase in the amounts of any efiluents that may be released offsite, and (iii) no significant increase in individual or cumulative occupational radiation exposure, EOl has concluded that the proposed change meets the criteria given in 10 CFR 51.22 (c)(9) for a categorical exclusion from the requirement for an environmental impact statement.
Notification of State Penonnel A copy of this amendment request has been provided to the State of Louisiana, Department of Environmental Quality - Radiation Protection Division.
REFERENCE
: 1. W. E. Russell (GE) to G. W. Scronce (Entergy), " Final Calculation of RBS Cycle 8 SLMCPR", GE Tech Letterbook 262 97-072, June 11,1997.
: 2. General Electric Standard Application for Reactor Fuel (GESTAR II), NEDE-240ll-P-A-13, August 1996.
: 3. General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO 10958-A, January 1977.
: 4. R Factor Calculation Method for Gell, GE12 and GE13 Fuel, NEDC-32505P, November 1995.
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