View source for RBG-44457, Application for Amend to License NPF-47,revising License Condition 2.C(13) Concerning Final Feedwater Temp Reduction Analysis, LAR 97-16.Affidavit Supporting LAR & Proprietary GE Rept NEDC-32549P,dtd May 1997,encl

{{Adams
| number = ML20216F185
| issue date = 04/09/1998
| title = Application for Amend to License NPF-47,revising License Condition 2.C(13) Concerning Final Feedwater Temp Reduction Analysis, LAR 97-16.Affidavit Supporting LAR & Proprietary GE Rept NEDC-32549P,dtd May 1997,encl
| 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 = RBF1-98-0021, RBF1-98-21, RBG-44457, NUDOCS 9804160462
| package number = ML20036E328
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 19
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Entsrgy operations,Inc.
A R,ver Bend Station 5485 U S Highway 61 PO Box 220
{                                                                         St Francisville. LA 70775 Tol 504 3814374 Fax 504 3814872 John H. McGaha. Jr.
vice Pr% dent Operatens April 9,1998 l
U. S. Nuclear Regulatory Commission Document Control Desk, OPI-17 Washington,DC 20555

==Subject:==
River Bend Station - Unit 1 l                     Docket No. 50-458 l                     License No. NPF-47 i                     License Amendment Request (LAR) 97-16, Final Feedwater Temperature Reduction Analysis
(        File Nos.:      G9.5, G9.42 RBEXEC-98-064 RBF1-98-0021 RBG-44457
                                                                                                                   )
Gentlemen:                                                                                                i l
In accordance with 10 CFR 50.90, Entergy Operations, Inc. (EOI) hereby applies for                        i amendment of River Bend Station's (RBS) Facility Operating License No. NPF-47 License Condition 2.C(13) concerning the " Final Feedwater Temperature Reduction Analysis."
This change request will enhance the operation of the plant, thereby yielding economic benefit to RBS through extended power production. The request allows Final Feedwater Temperature
!        Reduction (FFWTR) to extend the fuel cycle by maintaining the core thermal power at or close to rated by delaying the start of normal coastdown and decelerating the electrical power fall-off rate during the normal coastdown period. FFWTR, if used at the end of cycle 7, would have extended the cycle by approximately 14 effective full power days (EFPD) and can have a direct impact on the capacity factor. Similar economic benefit could be expected for cycle 8. This improvement has been licensed for the other Boiling Water Reactor (BWR) plants in the United States including the Perry plant , a BWR6.
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9804160462 980409 PDR  ADOCK 05000458 P                     PDR

License Amendment Request (LAR) 97-16, Final Feedwater Temperature Reduction Analysis April 9,1998                                                                                   l RBEXEC-98-064 RBF1-98-0021 RBG-44457 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. Attachment 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 Technical Specification (TS) page.
Attachment 4 is General Electric (GE) report NEDC-32549P, dated May 1997. This report contains the plant specific safety analysis prepared by GE in support of this change to the RBS license. Attachment 4 contains information proprietary to GE. GE requests that the document be withheld from public disclosure in accordance with 10 CFR 2.790 (a) (4). The affidavits supporting this request in accordance with 10 CFR 2.790 (a) (4), are provided with Attachment 4.
The original analysis, provided in Attachment 4, demonstrates EOls use of NRC approved GE methodology for FFWTR for fuel cycle extension and provides those cycle specific limits of RBS fuel cycle 7 design. The analysis for fuel cycle 8, our current operating cycle, included analysis supporting FFWTR for fur.1 cycle extension. These analyses determined the FFWTR condition is acceptable and would not require changes to current operating limits.
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 Safety Review Committee. If you have any questions regarding this request or require additional information, please contact Mr. W. J. Beck at (504) 381-4206.
Sincerely, 1
_         t l      JRM/RJK/WJB/BMB/
attachments I

r     License Amendment Request (LAR) 97-16, Final Feedwater Temperature Reduction Analysis April 9,1998 RBEXEC-98-064 RBF1-98-0021 RBG-44457 Page 3 of 3 cc:     Mr. David I.. Wigginton (w/o Attachment 4)
NRR Project Manager U. S. Nuclear Regulatory Commission                       1 M/S OWFN 13-H-3 Washington, DC 20555 l
NRC Resident Inspector (w/o Attachment 4)                 l P. O. Box 1050                                            ,
St. Francisville, LA 70775                                i U. S. Nuclear Regulatory Commission (w/o Attachment 4)
Region IV 611 Ryan Plaza Drive, Suite 400 Arlington,TX 70611 t             LA Department of Environmental Quality (w/o Attachment 4)
(             Radiation Protection Division l             P. O. Box 82135 Baton Rouge, LA 70884-2135 Attn: Administrator

BEFORE THE UNITED STATES NUCLE / '' REGULATORY COMMISSION LICENSE NO. NPF-47 DOCKET NO. 50-458 IN THE MATTER OF ENTERGY GULF STATES,INC. 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 l
Station License Amendment Request (LAR) 97-16, Change to Operating License NPF-47 License Condition 2.C(13) concerning the " Final Feedwater Temperature Reduction Analysis;"
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.
                                                                                     ~ hn R'. McGaha i
STATE OF LOUISIANA PARISH OF WEST FELICIANA SUBSCRIBED AND SWORN TO before me, a Notary Public, commissioned in the Parish above named, this l? A day of 8461                  ,1998.
(SEAL) hh0Dtl L            >
Claudia F. Hurst Notary Public My Commission expires with life                                                      ,

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    . RBG-44457
  ~
Attachment 2 Page 1 of 10 ATTACIIMENT 2 l
1 ENTERGY OPERATIONS INCORPORA TED RIVER BEND STATION                                             i DOCKET 50-458/ LICENSE NO. NPF-47                                      l Final Feedwater Temperature Reduction (LAR 97-16) l DOCUMENT INVOLVED Technical Specifications: License Condition 2.C(13)                                                   l l
Reason for Request                                                                                    )
River Bend Operating License NPF-47 License Condition 2.C(13), Partial Feedwater Heating (Section 15.1. SER), states "The facility shall not be operated with partial feedwater heating beyond the end of the fuel cycle without prior written approval of the staff. During the normal       ,
fuel cycle, the facility shall not be operated with a feedwater heating capacity which would result   I in a rated thermal power feedwater temperature less than 320 F without prior written approval of the staff." Proposed changes provide the bases for removing the first sentence referencing partial feedwater heating beyond the end of fuel cycle; such changes will allow partial feedwater
                                                                                                            )
I heating beyond the end of the fuel cycle, including for the purpose of fuel cycle extension.
l Final Feedwater Temperature Reduction (FFWTR) can extend the fuel cycle by maintaining the            j core thermal power at or close to rated after normal coastdown delaying the onset of the normal power coastdown period. FFWTR, if used at the end of cycle 7, would have extended the cycle by approximately 14 effective full power days (EFPD) and can have a direct impact on the capacity factor. Similar economic benefit could be expected for cycle 8. This mode of operation is similar to Feedwater Heater Out of Service (FWHOS) with the evaluation including the coastdown period.
DESCRIPTION The current RBS operating license allows operation with partial feedwater temperature reduction up to 100 *F during a normal fuel cycle. The operation with partial feedwater heating is desirable in the event certain feedwater heater (s) or string (s) of heaters become inoperable or are removed from service to perform maintenance during a fuel cycle. Operational flexibility and l     plant capacity factor are improved if the plant continues to operate until full heating is restored.
This mode of operation (FWHOS) has been approved for use at RBS in Amendment No. 37.
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    . RBG-44457 I
Attachment 2 Page 2 cf 10 i
The amendment allows operation with a range of rated feedwater temperature from 420 'F to 320
      'F at rated core power, or with the range of feedwater temperature as defined in current RBS plant procedures at off rated condition with any combination of feedwater heating capacity.
Operation resulting in partial feedwater heating for cycle extension is currently prohibited as identified in RBS Safety Evaluation Report (SER) for the initial RBS license and in the SER for Amendment 37 because at that time the safety analyses for this end of cycle condition was not provided. RBS License Condition 2.C(13) implements this restriction until such time as the required analyses is provided for NRC review and approval. This amendment request provides FFWTR analyses and demonstrates the acceptability of operations with a range of rated thermal power feedwater temperature from 420 F to 320 'F for cycle extension or during coastdown conditions. This analysis is consistent with the assumptions of the FWHOS analysis.
At the nominal end-of-cycle conditions reactor power decreases to less than rated. This condition, commonly identified as coastdown, is when core reactivity is reduced below the level which can be compensated for by withdrawal of control rods. FFWTR offers cycle extension and improved capacity factor by maintaining core thermal power at rated conditions delaying the onset of the normal power coastdown period. The use of reduced feedwater heating during coastdown is similar to partial feedwater heating during the normal cycle except, in general, the core power distribution is most limiting at end of cycle.                                           l The elTect of a 100 'F reduction in feedwater temperature on the feedwater nozzle and sparger was evaluated by GE as described in the attached report. This evaluation was found to be applicable to the FFWTR and FWHOS operation since the same amount of feedwater temperature reduction (100 *F) is assumed in the analyses. RBS has previously implemented site procedures regarding feedwater nozzle and sparger usage factor per Amendment No. 37 authorizing FWHOS.
The proposed change deletes the first sentence of License Condition 2.C(13) and revises the second sentence to allow operating with partial feedwater heating for the normal fuel cycle and for cycle extension under existing plant administrative controls.
JUSTIFICATION General Electric Company (GE) has provided the necessary safety evaluation, included as Attachment 4 (NEDC-32549P), to support the requested changes. This safety evaluation shows that RBS can operate with FFWTR during the coastdown condition for cycle extension. Partial l     loss of feedwater heating during the normal fuel cycle has been previously implemented at RBS as described in RBS USAR Section 15.1.7. Therefore, upon the approval and implementation of this request, RBS may operate at full power with feedwater heating capacity resulting in 100 %
rated core thermal power feedwater temperature at or above 320 'F for the entire fuel cycle including the coastdown. This allowance includes the previous provision that the number of full power days at such conditions are monitored by administrative controls already implemented per Amendment No. 37.

1
  . RBG 44457 Attachment 2 Page 3 of 10 RBS has also implemented operation in the Increased Core Flow (ICF) domain as described in USAR Section 15.0.7. The initial condition of 100% power,107% core flow (100P/107F) at the most limiting end-of-cycle (EOC) exposure was also selected for FFWTR analyses. The EOC 100P/107F state point is bounding for the entire RBS licensed operating domain because all the       ,
control rods are out and the scram is less effective. This state point was also determined to be the most limiting condition with respect to the Critical Power Ratio (CPR) as documented in the cycle 7 Supplement Reload Licensing Report (SRLR), Reference 2.
The impacts of the FFWTR condition on plant operating limits and the fuel thermal-mechanical performance are cycle dependent. They have been analyzed for the RBS cycle 7 configuration and applicability has been verified for fuel cycle 8. The application of this condition will be verified on a cycle specific basis. Furthermore, accident events, such as Loss-of-Coolant Accident (LOCA) and Anticipated Transient without Scram (ATWS) were reviewed for any                 J potential impact due to FFWTR mode. Other design evaluations potentially affected by partial feedwater heating including containment LOCA loads, the mechanical integrity of the reactor          ;
internal components, and the feedwater nozzle /sparger fatigue usage were evaluated.
Evaluation results (Attachment 4) justify both FWHOS and FFWTR provided that the feedwater sparger and nozzle fatigue limits are not exceeded. These limits are protected by restricting the number of days allowable in this mode of operation. Plant procedures to track the number of days in FWHOS or FFWTR have been previously implemented per Amendment No. 37.
The analysis supporting this application, and provided as Attachment 4, was conducted using RBS fuel cycle 7 design. The limiting events for fuel cycle 8 have been reanalyzed confirming applicability to the current operating cycle. The results of these reanalyses are included in the    i cycle 8 SRLR, Reference 4.
l Limiting Operational Occurrences GE has analyzed the limiting Anticipated Operational Occurrences (AOOs) for FFWTR effects.
Esc limiting AOOs are selected from USAR Section 15 to be consistent with the procedures (GESTAR US Supplement S.2) used in RBS cycle 7 Supplemental Reload Licensing Report (SRLR) in which the following events were found to be limiting or near limiting for CPR events:
     . Feedwater Controller Failure (FWCF) Maximum Demand
     . Load Rejection With No Bypass (LRNBP)
* Turbine Trip with No Bypass (TTNBP)                                                              l
* Pressure Regulator Failure Downscale (PRFD) l 1
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  . RBG-44457 Attachment 2 Page 4 of 10 Approved methodology (GESTAR Section 3) was used by GE to generate applicable core nuclear characteristics for the limiting ICF/FFWTR at EOC exposure with an all-rod-out (ARO) and cycle conditions. Traditionally in work from GE, EOC refers to the exposure point at which all rods have been pulled out of the core at 100%P/100%F. This is also sometimes referred to as End Of Full Power Life and is the point at which a power coastdown would normally start without any cycle extension actions. Increased core flow is the first preferred cycle extension option. The exposure at which maximum flow from ICF is reached while still at 100% power, 100%P/107%F, is referred to as Extended EOC or EEOC. After cycle extension from ICF is exhausted, FFWTR would be used to maintain rated thermal power, delaying the normal thermal power coastdown. The exposure at which cycle extension from ICF+FFWTR is exhausted, where FW temperature has been reduced to 320 F and the reactor is at 100%P/107%F with ARO, is referred to as EEEOC. Transient calculations to support FFWTR operation are performed at the EEEOC exposure point. For the purposes of this LAR, which is explicitly          j addressing FFWTR operation, EOC refers to the EEEOC exposure point.                               1 The transients responses for PRFD, LRNBP, TTNBP are discussed in USAR Sections 15 2 and the transient description for FWCF is discussed in Section 15.1. FFWTR transient ardyses (Table 2-1 of Attachment 4) show, for FWCF and PRFD, a slight increase in ACPP.as a result of     ;
lower feedwater temperature. The ACPR responses for LRNBP and ITNBP show a less severe            l ACPR due to a slightly lower initial pressure. These results remain consistent with the FWHOS analysis.
The 100 'F Loss of Feedwater Heating Event (LFWH) described in USAR Section 15.1.1 assumes this accident is initiated at the rated feedwater temperature condition. The design evaluation of this event is included in RBS cycle 7 and 8 reload licensing analysis (Reference 2 and 4). As discussed in the GE report section 2 the LFWH was also analyzed with FFWTR conditions. LFWH initiated from reduced temperature operation is milder because of the reduced subcooling effect.
The FWCF event yields the most limiting CPR for gel 1 and GE8 fuel under FFWTR/FWHOS conditions for cycle 7 and 8. These CPR changes are still bounded by RBS overall EOC OLMCPR results as listed in RBS Core Operating Limits Report (COLR). Therefore no change to the OLMCPR limits were required for cycle 7 or 8. The limiting event (s) will continue to be evaluated on a cycie and fuel specific basis, for FFWTR condition and OLMCPR limit impact.
In addition this event will be evaluated as necessary with other operational changes which affect the CPR limits identified in the COLR and SRLR.
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  . RBG-44457 Attachment 2 Page 5 of10 i   Vessel Overpressure Analysis .
l l   The Main Steam Isolation Valve (MSIV) Closure with Flux Scram described in USAR Section i    15.2.4 assumed normal feedwater temperature. The MSlV closure event at reduced feedwater l   temperature condition would result in a milder vessel pressurization transient as compared to the
!   same event initiated at normal feedwater temperature due to a reduced steam rate associated with l   increased subcooling. Therefore, the predicted peak pressure at reduced feedwater temperature, l   as concluded by GE in Attachment 4, is bounded by the value reported in cycle 7 SRLR for the j   FWHOS event.
I   Accident Evaluaticas l
In addition to the evaluations conducted of operational occurrences addressed through CPR evaluations, accidents evaluated for the effects of operation in FFWTR mode include control rod drop, rod withdrawal error, fuel loading error and loss of coolant accidents. These events are evaluated per GESTAR (US Supplement S.5.2.5). GE has concluded the results are bounded by the current analysis as described in the attached GE report. They are summarized as follows.
Control Rod Drop Accident The Control Rod Drop Accident event for the banked position withdrawal sequence plants has been evaluated by GE as described in the attached report. The results will not be affected by the change in subcooling causing by the reduced feedwater temperature condition.
Rod Withdrawal Error The characteristics of the Rod Withdrawal Error (RWE) described in USAR Section 15.4 are such that the most important parameters affecting the response are the initial control rod l        pattern and the error rod position which are not affected by the reduced feedwater l
temperature condition. Therefore, the results of this event is bounded by the RWE at normal feedwater temperature as reported in the SRLR and re-analysis for FFWTR is not required.       I Fuel Loading Error                                                                             !
Operation with reduced feedwater temperature would not impact the design evaluations of the Fuel Loading Error (FLE) such as the mislocated and misoriented bundle since the           l primary input parameter (the bundle R-factor) is not affected by the change in feedwater temperature. Therefore, the FLE events are bounded by the FLE at normal feedwater condition reported in the SRLR.
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l   . RBG-44457 l ,
Attachmmt 2 Page 6 of 10 Loss of Coolant Accident Operation with reduced feedwater temperature at FFWTR or ICF/ has been evaluated by GE as described in the attached report. GE concluded that the peak clad temperature (PCT) for l          ICF and/or FFWTR operation is bounded by the current PCT (SRLR) at the normal feedwater temperature initial conditions.
l l     ATWS l     The impact of FFWTR on ATWS response has been generically evaluated for other BWRs and it l
has been shown that the fuel surface heat flux, vessel bottom pressure, and suppression pool l
l     temperature become less severe due to a reduction in initial steam flow at a lower feedwater temperature, and subsequently reduced mass / energy release into the wetwell. GE concluded in
;     section 2.5 of Attachment 4 that the results of the design evaluation ATWS (USAR Section 15.8)
;     enveloped the response of similar events initiated at FFWTR conditions for RBS.
I Containment LOCA Loads
,     The LOCA containment evaluation (USAR Section 6.2) assumed normal feedwater temperature.
The short term design basis LOCA containment response during the blowdown phase is governed by the vessel blowdown flowrate which is dependent upon the reactor initial thermal-hydraulics conditions such as the core inlet subcooling. GE has performed an evaluation on this l     effect since partial feedwater heating changes the core inlet subcooling. The resulting peak       i drywell-to wetwell differential pressure, the peak loads for pool swell, condensation oscillation  I and chugging are all within the current design basis of the containment LOCA load as evaluated     j in Attachment 4.
OTHER EFFECTS AND CHANGES In addition to the events and accident evaluations conducted to confirm the operation of the plant i I
in FFWTR the following additional issues were evaluated for changes. GE has concluded the results are bounded by the current analysis as described in the attached GE report.
Turbine First Stage Pressure i          Under operation with reduced feMwater temperature, the relationship between vessel steam flow and core thermal power chai., . Less steam flow is generated at a given thermal power and therefore the turbine first stage pressure (TFSP) is reduced. These effects have l
been evaluated in Amendment No. 37 changes previously implemented and the associated l
Technical Specifications changes are applicable to partial feedwater heating for cycle extension since it assumes the same amount of reduced feedwater temperature (100 *F).
I          Therefore, changes to the High Power or the Low Power setpoints are not necessary for FFWTR operation.
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  . RBG-44457
,   Attachmmt 2 Page 7 of 10 Feedwater Sparger and Nozzle Fatigue Monitoring of the feedwater sparger and nozzle fatigue usage factor is required to maintain their values below the limit of 1.0. Evaluation results (Attachment 4) justify both FWHOS and FFWTR provided that the feedwater sparger and nozzle fatigue limits are not exceeded.
These limits are protected by restricting the number of days allowable in this mode of operation. Plant procedures to track the number of days in FWHOS or FFWTR have been previously implemented per Amendment No. 37. These plant procedures are applicable for both FWHOS and FFWTR since the same amount of feedwater temperature reduction,100 F, is assumed in each safety evaluation. The forty-year average number of days allowable during a 365 day operating year at a 100 F temperature reduction is 61 days and at a 50 'F reduction is 256 days. Therefore, the addition of this operating condition will be within the current monitoring program and the feedwater sparger and nozzle fatigue usa.ge factor values will be maintained as required.
Thermal-Hydraulic Stability RBS has implemented the GE Service Information Letter (SIL) 380 recommendation. RBS is also in compliance with the interim measures of NRC Bulletin 88-07 and supplement. All fuel designs used in RBS meet stability performance criteria as illustrated by the decay ratio and stability exchtsion region analyses in GESTAR. The procedural requirements in the GE SIL and the NRC Bulletin will remain in effect until the long term stability solution is implemented at RBS. As described in the RBS response to Generic Letter 94-02 dated                J September 12,1994 (RBG-40869) these provisions have been found to be adequate to avoid oscillations and to detect and suppress oscillations should they occur.
In addition to the measures taken in response to the GE SII       ' % NRC Bulletin, RBS is expecting to implement additional protection to limit the it .. .        d effects of a stability condition. These efforts are in response to NRC Generic Letter 94 a2 and comprise the
                                                                                                           ]
implementation of Enhanced Option IA as recommended in the BWROG topical report.                  j Further information concerning this change is contained in our response to GL 94-02.              '
Reactor Internals Mechanical Integrity The reactor internals most impacted by acoustic and flow-induced load under partial feedwater heating condition are the shroud, shroud support, and thejet pumps. These design evaluations have been previously performed to support Amendment No. 37 and GE concluded that sufficient design margin exists to accommodate 100 *F feedwater temperature reduction. This conclusion is applicable to partial feedwater heating for cycle extension since it assumes the same amount of feedwater temperature reduction.
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    . RBG-44457 l ,
Attachment 2 Page 8 of10 The NRC issued Generic Letter 92-01, Revision 1, concerning the reactor vessel integrity l          fracture toughness and material surveillance requirements contained in 10 CFR 50. In the
;          RBS response to item 3.a of GL-92-01 Revision 1," Embrittlement Effects Of Operating At i          An irradiation Temperature Below 525 T"(RBO-37110 dated July 2,1992), administrative l          procedures were implemented and they will continue to ensure conformance with this enteria.
i
!     PROPOSED CHANGES The proposed changes (Attachment 3) delete the first sentence of License Condition 2.C(13) and revise the second sentence to allow operation with partial feedwater heating for the normal fuel l     cycle and for cycle extension. Plant procedures governing partial feedwater heating will be revised to include planned partial feedwater heating for the purpose of cycle extension.          .
l     No change in MCPR limits or MAPLIIGR limits as reported in the COLR is required to implement partial feedwater heating. FWCF is the most limiting event at reduced feedwater           {
temperature and its worst CPR is bounded by the EOC OLMCPR set forth in the COLR. The               1 MAPLIIGR reported in the COLR is still applicable to the partial feedwater heating operation since the LOCA is bounded by the normal feedwater temperature design evaluation results. The FFWTR mode of operation has been analyzed and accounted for in setting the operating limits in      I the COLR as part of the regular reload analyses. Future reloads will continue to be analyzed for FFWTR impact.
ENVIRONMENTAL IMPACT CONSIDERATION EOI has reviewed this request against the criteria of 10CFR51.22 for environmental considerations. The request does not affect any system discharging radwaste to the environment or monitoring any such discharge. Also, the request does not adversely affect any system            ,
designed to monitor or isolate gaseous radioactive effluents to the environment. Therefore, the request does not involve a significant hazards consideration, does not significantly increase the types or quantity of effluent that may be released offsite, and does not significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, EOI concludes that the proposed change meets the criteria given in 10CFR51.22 (c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.
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    . RBG-44 457 Attachment 2 Page 9 of 10 NO SIGNIFICANT HAZARDS CONSIDERATION
: 1. The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated.
The abnormal operational occurrences or accidents analyzed in the SAR have been l          examined for impact caused by partial feedwater heating during cycle extension or at coastdown condition. The limiting abnormal operation transients, including the Load
;          Rejection with no Bypass (LRNBP) event and the Feedwater Controller Failure (FWCF) l          maximum demand event, Turbine Trip with No Bypass (TfNBP) and Pressure Regulator l          Failure Downscale (PRFD) have been analyzed based upon the core nuclear characteristic at end-of-cycle (EOC) conditions including the effects ofincreased core flow and the proposed reduction in feedwater temperature with an all-rods-out condition.
l          The LOCA, fuel loading error, rod drop accident, rod withdrawal error, overpressure
!          protections and ATWS analyses have been evaluated for the effects of reduced feedwater
;          temperature operation and found acceptable. In addition, the case of the analyzed i
operational events the current fuel OLMCPR and MAPLHGR limits I cund those necessary l          for operation and therefore, are not affected by operation with FFWTR therefore, these
!          events are bounded by the current RBS analysis. Because the accident results are acceptable and the current operating fuel limits are unaffected, the consequence of an event previously l          evaluated remains unaffected.
l          The probability of an accident is not affected by the proposed changes since no systems or l          equipment which could initiate an accident are affected. Therefore, the proposed changes do I
not significantly increase tia probability or consequences of any previously evaluated l          accident.
l l     2. The request does not create the possibility of occurrence of a new or different kind of
;          accident from any accident previously evaluated, l
l          The FFWTR mode of operation is functionally similar to operation with Feedwater Heaters Out of Service (USAR Section 15.1.7). All abnormal operational transients or accidents have been evaluated and the most limiting cases have been analyzed for applicability for the l          FFWTR operation. Limits on MAPLHGR and OLMCPR (including the power and flow dependent MCPR) which are included in the Core Operating Limits Report (COLR) as part i          of the normal reload licensing process will continue to assure that operations are within the l          assumptions, initial conditions and assumed power distribution and therefore will not create i          a new type of accident. The proposed changes do not involve new setpoints, new system interactions, or physical modifications to the plant. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previous analyzed.
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  . RBG-44457                                                                                          i AttachmGnt 2 l
l   Page 10 of 10 l
l   3. The request does not involve a significant reduction in a margin of safety.
l        The proposed changes do not involve any setpoint changes and would allow steady state
!        power operation at off-rated feedwater temperature conditions as defined in current plant
!        procedures. The transient and accidents described in the SAR are evaluated for effects caused by the reduced feedwater temperature of 100 #F. As cer.ribed in Attachment 4 (NEDC-32549P), the FWCF is the most limiting transient under such condition and the required OLMCPR for this event is bounded by the EOC OLMCPR limits set forth in the RBS COLR. The thermal limits MCPR and LHGR curves, and the MAPLHGR limits                    '
establish limits on power operation and thereby ensure that the core is operated within the assumptions and initial conditions of the transient or accident analyses, i
Operation within these limits set forth by the MCPR limits, the LHOR limits and the           l MAPLHGR criteria will ensure that the margin of safety will be maintained to the same         I level described in the Technical Specifications Bases and the SAR. As a result the consequences of postulated transients or accidents are not increased. The MCPR safety limit, mechanical performance limits and overpressure limits are not exceeded during any transient or postulated accident at normal feedwater temperature or at reduced feedwater temperature condition. Therefore, the proposed changes to allow partial feedwater heating for cycle extension do not involve a significant reduction in margin of safety.
On this basis,it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92 and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change and (3) this action I
will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC final environmental statement.
REFERENCES
: 1. NUREG-989, Amendment No. 37,
: 2. 24A5188 Rev. 2, Supplemental Reload Licensing Report for River Bend Station Reload 6             ,
Cycle 7.                                                                                        l
: 3. General Electric Standard Application for Reactor Fuel (GESTAR-II), NEDE-20411-P-A.              ;
: 4. J1103150 Rev. O, Supplemental Reload Licensing Report for River Bend Station Reload 7            i Cycle 8.                                                                                      ]
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