RBG-46583, License Amendment Request, Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, Core Operating Limits Report (Colr).

License Amendment Request, Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, Core Operating Limits Report (Colr).

ML062960299
Person / Time
Site:	River Bend
Issue date:	10/16/2006
From:	Hinnenkamp P Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RBG-46583
Download: ML062960299 (18)

Text

Entergy Operations, Inc.

River Bend Station 5485 U.S. Highway 61N St. Francisville, LA 70775 Tel 225 381 4157 UIV Fax 225 635 5068 dlorfin@entergy.com Paul D. Hinnenkamp Vice President, Operations RBG-46583 October 16, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, "Core Operating Limits Report (COLR)"

River Bend Station, Unit 1 Docket No. 50-458 License No. NPF-47

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for River Bend Station, Unit 1 (RBS). The proposed change will add a NRC previously approved topical report to the analytical methods referenced in Technical Specification (TS) section 5.6.5, "Core Operating Limits Report (COLR)." TS Section 5.6.5 requires core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, using the analytical methods previously approved by the NRC and referenced in section 5.6.5.b. Certain core operating limits are established based upon design basis Loss-of-Coolant Accident (LOCA) analyses. The current method of performing the LOCA analyses will be replaced by an updated method described in AREVA NP (formerly known as Framatome or Siemens) topical report, "EXEM BWR-2000 ECCS Evaluation Model".

RBS plans to use the updated methodology beginning with operating Cycle 15, currently scheduled to begin fall 2007. EXEM BWR-2000 has been approved by the NRC and is applicable to the RBS plant design and the fuel being used at RBS. RBS currently uses ATRIUM-10 fuel and will continue to do so during operating Cycle 15.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.

RBG-46583 Page 2 of 3 The proposed change does not include any new commitments.

The NRC has approved similar TS changes to allow the use of EXEM BWR-2000 for LOCA analysis at Susquehanna Units 1 and 2, Browns Ferry Units 2 and 3, LaSalle County Station Units 1 and 2, and Columbia Generating Station.

Entergy requested the same TS changes for Grand Gulf Nuclear Station (GGNS) by letter GNRO-2006/00015 dated May 8, 2006 (ADAMS Accession No.ML061310084). That request is currently under NRC staff review (TAC No. MD1496).

Entergy requests approval of the proposed amendment by August 31, 2007 to support Refueling Outage 14. Once approved, the amendment shall be implemented prior to Cycle 15 operation. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact Ron Byrd at 601-368-5792.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 16, 2006.

.Sincerely, Paul D. Hinnenkamp Vice President, Operations River Bend Station, Unit 1 PHD/RWB Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up) cc: Dr. Bruce S. Mallett U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 U.S. Nuclear Regulatory Commission Attn: Mr. Bhalchandra K. Vaidya MS O-7D1 Washington, DC 20555-0001

RBG-46583 page 3 of 3 cc: Louisiana Department of Environmental Quality Office of Environmental Compliance Attn: Mr. Jeff Meyers Surveillance Division P. O. Box 4312 Baton Rouge, LA 70821-4312

Attachment 1 RBG-46583 Analysis of Proposed Technical Specification Change to RBG-46583 Page 1 of 11

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-47 for River Bend Station, Unit 1 (RBS).

The proposed change will add a NRC previously approved topical report to the analytical methods referenced in Technical Specification (TS) section 5.6.5, "Core Operating Limits Report (COLR)." TS Section 5.6.5 requires core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, using the analytical methods previously approved by the NRC and referenced in section 5.6.5.b. Certain core operating limits are established based upon design basis Loss-of-Coolant Accident (LOCA) analyses. The current method of performing the LOCA analyses will be replaced by an updated method described in AREVA NP (formerly known as Framatome or Siemens) topical report, "EXEM BWR-2000 ECCS Evaluation Model"..

RBS plans to use the updated LOCA analytical method to determine specific core operating limits for operating Cycle 15 currently scheduled to begin fall 2007.

2.0 PROPOSED CHANGE

TS Section 5.6.5.b lists the analytical methods previously reviewed and approved by the NRC that are used to determine the core operating limits. The list includes the following topical reports:

18. XN-NF-80-19(P)(A) Volumes 2, 2A, 2B, And 2C, "Exxon Nuclear Methodology for Boiling Water Reactors EXEM BWR Evaluation Model", Exxon Nuclear Company, Richland, WA.

19. ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model", Advanced Nuclear Fuels, Richland, WA.

20. ANF-91-048(P)(A) Supplements 1 and 2, "BWR Jet Pump Model Revision for RELAX", Siemens Power Corporation, Richland, WA

21. XN-CC-33(A), "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual", Exxon Nuclear Company, Richland, WA.

These references are proposed to be revised as follows:

18. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP Richland, Inc.

19. Deleted

20. Deleted

21. Deleted to RBG-46583 Page 2 of 11 The TS change is intended to allow RBS to use the EXEM BWR-2000 methodology for performing LOCA analyses beginning with operating Cycle 15. The LOCA analyses will be used to establish the Average Planar Linear Heat Generation Rate (APLHGR) operating limits, imposed by TS 3.2.1. The NRC staff approved the EXEM BWR-2000 topical report by letter dated May 29, 2001 and found it to be acceptable for referencing in license applications.

RBS currently performs LOCA analyses using the EXEM BWR Evaluation Model in conjunction with the RELAX and HUXY codes (references 19, 20, and 21). The EXEM BWR ECCS Evaluation Model (current reference 18) will no longer be used beginning with operating Cycle 15 and is being deleted.

References to the topical reports for the RELAX and HUXY codes (references 20 and 21) no longer need to be listed because they are incorporated by reference into the EXEM BWR-2000 ECCS Evaluation Model topical report (new reference 18). This is consistent with other licensee amendments that approved the use of EXEM BWR-2000.

3.0 BACKGROUND

Core operating limits are established each operating cycle in accordance with TS 3.2, "Power Distribution" and TS 5.6.5, "Core Operating Limits Report (COLR)". These operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO). In addition, the APLHGR operating limits imposed by TS 3.2.1 also ensure that the peak cladding temperature (PCT) during the postulated design basis LOCA does not exceed the 2200°F limit specified in 10 CFR 50.46.

The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS section 5.6.5.b. The analytical methods currently listed support the determination of core operating limits by using those methods applicable to fuel supplied by General Electric (GE, currently known as Global Nuclear Fuels) or AREVA NP (formerly known as Framatome or Siemens). RBS is only using AREVA NP ATRIUM-10 fuel in the current operating cycle. RBS also plans to continue using ATRIUM-10 fuel in the next operating cycle, Cycle 15. Although River Bend is currently using AREVA NP fuel, reference 24, NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II)" is left in section 5.6.5.b to allow the option of using fuel supplied by Global Nuclear Fuels.

The requested TS change will add a previously approved AREVA NP topical report to the references listed in TS section 5.6.5.b. The topical report, EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model," describes an updated method of performing ECCS evaluations under design basis LOCA conditions. RBS currently performs the LOCA analysis using the EXEM BWR methodology described in topical report ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model." The LOCA analysis is used to establish the APLHGR limits required by TS 3.2.1 and TS 5.6.5.a.1).

One of the principal reasons for the updated methodology was to address some issues raised during a 1997 NRC Core Performance Inspection of Framatome (known now as AREVA NP).

The inspection included an assessment of the EXEM BWR evaluation model. The assessment determined that the benchmarking and validation of certain computer codes used to RBG-46583 Page 3 of 11 to support the evaluation model needed to be improved. These concerns were addressed by the development and approval of the EXEM BWR-2000 ECCS evaluation model. As part of the resolution of the NRC assessment issues, the older EXEM BWR evaluation model will no longer be used. This necessitates the need for the RBS LOCA analysis to be reanalyzed with the EXEM BWR-2000 model. The proposed TS change, if approved, will be implemented beginning with RBS operating Cycle 15.

4.0 TECHNICAL ANALYSIS

The APLHGR limits required by Technical Specification 3.2.1 are specified in the COLR and are the result of fuel design, design basis accident (DBA), and transient analyses. The APLHGR is a measure of the average linear heat generation rate of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design basis LOCA does not exceed the 2200°F limit specified in 10 CFR 50.46.

RBS currently uses the NRC approved EXEM BWR evaluation model for the LOCA analysis.

RBS proposes to use an updated AREVA NP LOCA analytical method beginning in operating Cycle 15. The updated method is described in topical report EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model." The NRC found the topical report to be acceptable for referencing in license applications and issued its safety evaluation providing the basis for acceptance by letter dated May 29, 2001.

The NRC review of the EXEM BWR-2000 model concluded the following:

The staff notes that from validation against test data, the large-break Design Basis Accident (DBA) Peak Clad Temperatures (PCTs) were conservatively calculated. The test results for small breaks show low temperatures, and the EXEM BWR-2000 model using evaluation model options bounds the temperature data. Furthermore, the EXEM BWR-2000 model adequately predicts the important LOCA phenomena.

The staff therefore concludes that the proposed EXEM BWR-2000 ECCS EM, as documented in Reference 1, 2, 4 and 5, is acceptable for referencing in BWR LOCA analyses, with the limitation that application of the revised evaluation model will be limited to jet pump applications.

Entergy has determined that the EXEM BWR-2000 evaluation model is acceptable for RBS LOCA analyses. The RBS plant design meets the limitation stipulated in the NRC safety evaluation since it is a BWR-6 plant which incorporates jet pumps in its design. RBS uses ATRIUM-10 fuel in its current operating cycle (Cycle 14) and plans to continue using ATRIUM- 10 fuel in Cycle 15.

A RBS plant specific LOCA analysis was performed by AREVA NP using the EXEM BWR-2000 evaluation model. The analysis assumed a full core of ATRIUM-10 fuel and used a generic ATRIUM-10 neutronic design that is expected to be conservative relative to actual cycle-specific designs. A cycle specific evaluation is performed each cycle to confirm that the generic fuel design remains bounding.

to RBG-46583 Page 4 of 11 The results of the new RBS LOCA analysis were compared with the current licensing basis analysis, which uses the older EXEM BWR evaluation model. Tables 1 through 5 included in this attachment provide the key input parameters used in both the current licensing basis analysis and the new analysis. The current calculated LOCA PCT for ATRIUM-10 fuel, using the older EXEM BWR model, is 1875°F. The LOCA PCT for ATRIUM-1 0 fuel using the new EXEM BWR-2000 model is 1969°F. As such, there is an increase in PCT of 94°F associated with the change in the LOCA analysis. However, it should be noted that the new analysis conservatively assumed that the low pressure injection permissive was 350 psia rather than the 450 psia assumed in the current analysis. This reduced pressure permissive assumption is conservative since the TS requires the permissive setpoint to be at or above 472 psig (approximately 487 psia). The new analysis also assumes that only 4 of 7 required Automatic Depressurization System (ADS) valves are available; whereas, the current analysis assumes 5 are available. The assumption is also conservative since the TS do not allow plant operation with less than 6 operable ADS valves. The resulting PCT of 1969 0 F, using the EXEM BWR-2000 model for the LOCA analyses, still affords adequate margin to the 22000F limit of 10 CFR 50.46.

In summary, the EXEM BWR-2000 evaluation model is an improved method of evaluating ECCS performance with LOCA analyses. The model has been reviewed and approved by the NRC and is applicable to the RBS plant design and the fuel being used at RBS. The application of the LOCA analysis model will continue to ensure that the APLHGR operating limits are established to protect the fuel cladding integrity during normal operation, AQOs, and the design basis LOCA.

5.0 REGULATORY ANALYSIS

5.1 Applicable Requlatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any General Design Criterion (GDC) differently than described in the Updated Final Safety Analysis Report (UFSAR).

10 CFR 50.36, paragraph (c)(5), states that the TS will include administrative controls that address the provisions relating to organization and management, procedures, record keeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

The COLR is required as a part of the reporting requirements specified in the RBS TS Administrative Controls section. The TS requires the core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and to be documented in the COLR. In addition, it requires the analytical methods used to determine the core operating limits to be approved by the NRC and described in the Administrative Controls section of the TS. The proposed TS changes ensure that these requirements are met.

Generic Letter (GL) 88-16, "Removal of Cycle-Specific parameters from Technical Specifications" and Technical Specification Task Force (TSTF) traveler TSTF-363, "Revise to RBG-46583 Page 5 of 11 Topical Report References in ITS 5.6.5, COLR" provide guidance on the method of referencing topical reports in the TS and the COLR. TSTF-363, which was approved by the NRC on July 6, 2000, requires the titles of the analytical methods to be included in TS 5.6.5.b and the complete identification (report number, title, revision, date, and any supplements) to be included in the COLR. The proposed TS changes are consistent with this guidance.

10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" requires each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding to be provided with an ECCS designed so that its calculated cooling performance following postulated LOCAs conforms to explicit criteria. Specifically related to this proposed TS change is a criterion that the calculated maximum fuel element cladding temperature shall not exceed 2200 F. In addition, the ECCS cooling performance must be calculated in accordance with an acceptable evaluation model. The proposed TS change allows use of an acceptable evaluation model that ensures conformance with these requirements.

5.2 No Sigqnificant Hazards Consideration Entergy proposes to use an updated Loss-of-Coolant Accident (LOCA) analysis method for River Bend Station (RBS) beginning with operating Cycle 15. The AREVA NP (formerly known as Framatome or Siemens) method of analysis, EXEM BWR-2000, will be used to determine the Average Planar Linear Heat Generation Rate (APLHGR) core operating limits imposed by Technical Specification (TS) 3.2.1. TS section 5.6.5.b must be revised to include a reference to the topical report that describes the updated method prior to the method being used to establish the APLHGR limits. Therefore Entergy proposes to revise TS section 5.6.5.b to replace references to the current method, EXEM BWR, with a reference to the updated method, EXEM BWR-2000.

Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Core operating limits are established each operating cycle in accordance with TS 3.2, "Power Distribution" and TS 5.6.5, "Core Operating Limits Report (COLR)". These core operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in theevent of any Anticipated Operational Occurrence (AOO). In addition, the Average Planar Linear Heat Generation Rate (APLHGR) operating limits imposed by Technical Specification 3.2.1 also ensure that the peak cladding temperature (PCT) during the postulated design basis LOCA does not exceed the 2200°F limit specified in 10 CFR 50.46. The APLHGR is a measure of the average linear heat generation rate of all the fuel rods in a fuel assembly at any axial location.

to RBG-46583 Page 6 of 11 The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS section 5.6.5.b. A change to TS section 5.6.5.b is requested to include an updated LOCA analysis method, EXEM BWR-2000.

The updated method will be used to determine the APLHGR operating limits imposed by Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and approved by the NRC and is applicable to the RBS plant design and the AREVA NP fuel being used at RBS. The application of the LOCA analytical model will continue to ensure that the APLHGR operating limits are established to protect the fuel cladding integrity during normal operation, AOOs, and the design basis LOCA.

The requested TS changes concern the use of analytical methods and do not involve any plant modifications or operational changes that could affect any postulated accident precursors or accident mitigation systems and do not introduce any new accident initiation mechanisms.

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

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed TS amendment will not change the design function, reliability, performance, or operation of any plant systems, components, or structures. It does not create the possibility of a new failure mechanism, malfunction, or accident initiators not considered in the design and licensing bases. Plant operation will continue to be within the core operating limits that are established using NRC approved methods that are applicable to the RBS design and the RBS fuel.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The ECCS performance analysis methods are used to establish the APLHGR limits required by Technical Specification 3.2.1. The APLHGR limits are specified in the COLR and are the result of fuel design, design basis accident (DBA), and transient analyses. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design basis LOCA does not exceed the 2200 0 F limit specified in 10 CFR 50.46.

The EXEM BWR-2000 evaluation model is an updated LOCA analytical method that.

has been approved by the NRC and is applicable to the RBS plant design and the fuel being used at RBS. A RBS plant specific ECCS performance analysis has been performed with the EXEM BWR-2000 evaluation model. This evaluation concluded to RBG-46583 Page 7 of 11 that the resulting PCT still afforded adequate margin to the 2200°F limit of 10 CFR 50.46 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE The NRC has approved the use of the EXEM BWR-2000 ECCS Evaluation Model, Topical Report EMF-2361 (P)(A), for use at the following plants.

PLANT NAME AMENDMENT No.

1. Columbia Generating Station Amendment No. 185 (ADAMS Accession No.:

ML031340744)

2. Susquehanna Steam Electric Amendment No. 231 (ADAMS Accession No.:

Station Unit 1 ML060730355)

3. Susquehanna Steam Electric Amendment No. 194 (ADAMS Accession No.:

Station Unit 2 ML050590044)

4. Browns Ferry Units 2 and 3 Amendment Nos. 287 and 245 (ADAMS Accession No.: ML033650142)

5. LaSalle County Station Units Amendment Nos. 174 and 160 (ADAMS Accession 1 and 2 No.: ML060120391)

K to RBG-46583 Page 8 of 11 Table 1 Initial Conditions EXEM EXEM Parameter BWR BWR-2000 Value Value Reactor Power (MWt) 3100 3100 Total Core Flow (MIb/hr) 90.4 90.4 Hot Assembly MAPLHGR (kW/ft) 12.5 12.5 Hot Assembly MCPR 1.16 1.16 Table 2 High Pressure Core Spray HPCS) Parameters EXEM EXEM Parameter BWR BWR-2000 Value Value Maximum Coolant Temperature (OF) 120 120 InitiatingSignals and Setpoints Water Level (in)1 454.82 454.82 High Drywell Pressure (psig) Not Used Not Used Time Delays Time for HPCS pump to reach rated speed 57 57 and injection valve wide open (sec)

Coolant Flow Rate Versus Pressure Vessel to Drywell AP (psid) Flow Rate (gpm) 0 4410 4410 200 4410 4410 1147 1260 1260 1177 0 0 1 Relative to vessel zero to RBG-46583 Page 9 of 11 Table 3 Low Pressure Coolant Injection LPCI) Parameters EXEM EXEM Parameter BWR BWR-2000 Value Value Reactor pressure permissive for opening 450 350 valves - analytical (psia)

Maximum Coolant Temperature (°F) 120 120 Initiating.Signals and Setpoints Water Level (in) 2 354.98 354.98 High Drywell Pressure (psig) Not Used Not Used Time Delays Maximum time for LPCI pumps to reach 67 67 rated speed (sec) 67_67 LPCI injection valve stroke time (sec) 47 47 Coolant Flow Rate Versus Pressure Vessel to Drywell AP (psid) Flow Rate (gpm) 0 4470 4470 20 4470 4470 222 0 0 2 Relative to vessel zero, to RBG-46583 Page 10 of 11 Table 4 Low Pressure Core Spray (LPCS) Parameters EXEM EXEM Parameter BWR BWR-2000 Value Value Reactor pressure permissive for opening 450 350 valves - analytical (psia) -

Maximum Coolant Temperature (OF) 120 120 InitiatingSignals and Setpoints Water Level (in) 3 354.98 354.98 High Drywell Pressure (psig) Not Used Not Used Time Delays Maximum rated speedtime for LPCS pump to reach (sec)577 [

LPCS injection valve stroke time (sec) 37 37 Coolant Flow Rate Versus Pressure Vessel to Drywell AP (psid) Flow Rate (gpm) 0 4950 4950 113 441.0 ,4410 263 0 0 Relative to vessel zero to RBG-46583 Page 11 of 11 Table 5 Automatic Depressurization System (ADS) Parameters EXEM EXEM Parameter BWR BWR-2000 Value [1] Value [2]

Number of Valves Installed 7 7 Number of Valves Available 5 44 Minimum flow capacity of available valves 4.0 at 1125 3.2 at 1125 (Mlbm/hr at psig)

InitiatingSignals and Setpoints WaterLevel (in)5 354.98 354.98 High Drywell Pressure (psig)6 2 2 Time Delays DelayTime (from initiating signal to time 6

120 120 valves are open) (sec) 4 Only 4 valves are assumed operable in the analyses to support two ADSVOOS operation and to eliminate the need to evaluate the single failure of an ADS valve as a separate single failure (total of three inoperable ADS valves) s Relative to vessel zero 6 The drywell high-pressure setpoint is exceeded before the water level setpoint is reached. Therefore, the ADS timer is assumed to start when the water level setpoint is reached.

Attachment 2 RBG-46583 Proposed Technical Specification Changes (mark-up includes affected Operating License page)

Attachment 2 to RBG-46583 Page 1 of 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

12) XN-NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Error Analysis, MCPRp for Plant Operations within the Extended Operation Domain", Exxon Nuclear Company, Richland, WA.

13) ANF-1358(P)(A), 'The Loss of Feedwater Heating Transient in Boiling Water Reactors", Advanced Nuclear Fuels Corporation, Richland, WA.

.14) EMF-1997(P)(A), 'ANFB-10 Critical Power Correlation", Siemens Power Corporation, Richland, WA.

15) EMF-1997(P) Supplement 1 (P)(A), "ANFB-10 Critical Power Correlation High Local Peaking Results", Siemens Power Corporation, Richland, WA.

16) EMF-2209(P)(A), 'SPCB Critical Power Correlation", Siemens Power Corporation., Richland, WA.

17) EMF-2245(P(A), "Application of Siemens Power Corporation's

-I Critical Power Correlations to Co-Resident Fuel", Siements Power EMF-2361(P)(A), Corporation.Richland WA "EXEM BWR-2000

--- - /

ECCS Evaluation 18) XN-NF-8-19(P)(A) Vol gmes 2, 2A, 2B. And 2C, 'Exn Nuclear I Model," Framatome Methodlogy for Boilin Water Reacto EXEM BWI Evaluation I ANP Richland, Inc.

• Model'. Exxon Nucle .a R In WA . L

19) ANF-91-048 )(A), "Advanc d Nuclear Fue Corporation I I--------------------I Methodolo for Boiling W er Reactors E EM BWR Evalu ion I Model", A anced Nuclea Fuel, Richland WA. . .

chii I

20. ANF 48(P)(A) Supp ments I and "BWR Jet Pum p/odel 1 Deleted I Revisio for RELAX, S- mens Power 0rporation, Richl nd, WA.

L-

21) 1 XN-C -33(A), "HUXY A-Generalized ultrod Heatup ode with 101 CFR Appendix K eatup Option sers Manual", E on Nuclear -

ILCm/any, Richland, WA

22) EMF-CC-074(P)(A), Volume 4, "BWRStability Analysis Assessment of STAIF with Input from MICROBURN-B2", Siemens Power Corporation, Richland, WA.

23) EMF-2292(P)(A), "ATRIUM-10 Appendix K Spray Heat Transfer Coefficients", Siemens Power Corporation, Richland, WA.

RIVER BEND 5.0-18b Amendment No. 1,22 to RBG-46583 Page 2 of 2 (3) EOI, pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3091 megawatts thermal (100% rated power) in accordance with the conditions specified herein. The items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications and Environmental Insert new Protection Plan Amendment The Technical Specifications contained /n Appendix I No.

A, as revised through Amendment No. and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 70 79 114 4124, 151