View source for CNRO-2019-00009, Stations - Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR, Revision 2

{{Adams
| number = ML19070A227
| issue date = 03/07/2019
| title = Stations - Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR, Revision 2
| author name = Halter M
| author affiliation = Entergy Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000416, 05000458
| license number = NPF-029, NPF-047
| contact person = 
| case reference number = CNRO-2019-00009
| document type = Letter type:CNRO, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 49
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Entergy Operations, Inc.
Enter~                                                                       1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5573 Mandy K. Halter Director, Nuclear Licensing 10 CFR 50.90 CNR0-2019-00009 March 7, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==Subject:==
Application to Revise Technical Specifications to Adopt TSTF-564, Safety  11 11 Limit MCPR, Revision 2 Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed Facility Operating License No. NPF-29 River Bend Station, Unit 1 NRC Docket No. 50-458 Renewed Facility Operating License No. NPF-47 In accordance with Title 1O of the Code of Federal Regulations (CFR) Part 50, Section 50.90, 11 Application for amendment of license, construction permit, or early site permit, Entergy 11 Operations, Inc. (Entergy) is submitting a request for an amendment to Renewed Facility Operating License, Appendix A, 'Technical Specifications (TS) for Grand Gulf Nuclear Station, 11 Unit 1 (GGNS) and River Bend Station, Unit 1 (RBS).
The proposed change revises the GGNS and RBS TS 2.1.1.2 Safety Limit (SL) value for the minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value, while still meeting the regulatory requirement for an SL on MCPR. In addition, the proposed 11 changes revise GGNS and RBS TS 5.6.5, Core Operating Limits Report (COLR), to require     11 inclusion of a 99.9% MCPR value into the cycle-specific COLA.
These proposed changes adopt Technical Specification Task Force (TSTF) Traveler 564, 11 "Safety Limit MCPR, Revision 2 into the GGNS TS and the RBS TS.
The Enclosure to this letter provides a description and assessment of the proposed changes.
Attachments 1.a and 1.b to the Enclosure provide the existing TS pages for GGNS and RBS, respectively, marked-up to show the proposed changes. Attachments 2.a and 2.b provide revised (clean) TS pages. Attachments 3.a and 3.b provide, for information only, marked up versions of existing TS Bases pages to show the proposed changes.

CNR0-2019-00009 Page 2 of 2 Entergy requests approval of the proposed license amendment by March 7, 2020. The proposed changes would be implemented within 90 days of issuance of the amendment.
This letter contains no new regulatory commitments.
Should you have any questions or require additional information, please contact Stephenie Pyle, Senior Manager, Fleet Regulatory Assurance at 601-368-5516.
In accordance with 10 CFR 50.91, Notice for public comment; State consultation," paragraph 11 (b), a copy of this application, with attachments, is being provided to the designated State Officials.
I declare under penalty of perjury, the foregoing is true and correct. Executed on March 7, 2019.
Respectfully,             . f
   ~hKd!vt Mandy K. Halter MKH/rws/jls

==Enclosure:==

Description and Assessment of the Proposed Changes Attachments to 

==Enclosure:==

1.a. Markup of Technical Specification (TS) Pages, Grand Gulf Nuclear Station, Unit 1 1.b. Markup of Technical Specification (TS) Pages, River Bend Station, Unit 1 2.a. Clean Technical Specification (TS) Pages, Grand Gulf Nuclear Station, Unit 1 2.b. Clean Technical Specification (TS) Pages, River Bend Station, Unit 1 3.a. Markup of Technical Specification (TS) Bases Pages, For Information Only, Grand Gulf Nuclear Station, Unit 1 3.b. Markup of Technical Specification (TS) Bases Pages, For Information Only, River Bend Station, Unit 1 cc:     NRC Region IV Regional Administrator NRC Senior Resident Inspector - Grand Gulf Nuclear Station, Unit 1 NRC Senior Resident Inspector - River Bend Station, Unit 1 State Health Officer, Mississippi Department of Health Louisiana Department of Environmental Quality NRC Project Manager - Entergy Fleet NRC Project Manager - Grand Gulf Nuclear Station, Unit 1 NRC Project Manager - River Bend Station, Unit 1

Enclosure CNR0-2019-00009 Description and Assessment of the Proposed Changes Application to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit Minimum Critical Power Ratio" Grand Gulf Nuclear Station, Unit 1 River Bend Station, Unit 1 (4 Pages) 1.0 Description 2.0 Assessment 2.1    Applicability of Published Safety Evaluation 2.2    Variations 3.0 Regulatory Analysis 3.1    No Significant Hazards Determination 4.0 Environmental Consideration 5.0 References

CNR0-2019-00009 Enclosure Page 1 of 4

==1.0     DESCRIPTION==

11 Entergy Operations, Inc. (Entergy) requests adoption of TSTF-564, "Safety Limit MCPR, Revision 2 (Reference 5.1) which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Grand Gulf Nuclear Station, Unit 1 (GGNS) and River Bend Station, Unit 1 (RBS) Technical Specifications (TS). The proposed amendment revises the TS safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.
2.0     ASSESSMENT 2.1     Applicability of Safety Evaluation Entergy has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated November 19, 2018 (Reference 5.2), as well as the information provided in TSTF-564. Entergy has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRG are applicable to GGNS and RBS, and justify this amendment request for the incorporation of the changes to the GGNS and RBS TS.
GGNS is currently fueled with Global Nuclear Fuel 2 (GNF2) fuel bundles. RBS is currently fueled with a mixture of GNF2 and GNF3 fuel bundles. The proposed SLMCPR Safety value in SL 2.1.1.2 for both GGNS and RBS is 1.07, consistent with the Table 1 values in TSTF-564 for GNF2 and GNF3 fuel. The proposed TS SLMCPR value and the proposed change to TS 5.6.5 is supported by References 5.3, 5.4, 5.5, and 5.6 (i.e., the General Electric (GE) documentation supporting the acceptance of TSTF-564 for GNF2 and GNF3 fuel.
The MCPR value that is calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences is referred to as MCPR 99 _9 %. Technical Specification 5.6.5, 11 "Core Operating Limits Report (COLR), is revised to require the MCPR 99 _9 % value to be included in the cycle-specific COLA 2.2     Variations Entergy is proposing the following variations from the TS changes described in TSTF-564.
These variations do not affect the applicability of TSTF-564 or the NRG SE to the proposed license amendment.
2.2.1   The GGNS and RBS TS utilize different numbering than the Improved Standard Technical Specifications (ISTS), on which TSTF-564 was based. Specifically, TSTF-564 revises TS 5.6.3, "Core Operating Limits Report (COLR)." The COLA TS number in both the GGNS and RBS TS is 5.6.5. These differences are administrative and do not affect the applicability of TSTF-564 to either GGNS or RBS.

CNR0-2019-0000 9 Enclosure Page 2 of 4 2.2.2   The GGNS and RBS TS contain requirements that differ from the ISTS on which TSTF-564 was based (i.e., steam dome pressure values in SL 2.1.1.1 and reactor thermal power in TS 3.2.2 Applicability). These differences are site-specific variations, and do not affect the applicability of TSTF-564 to either GGNS or RBS.
3.0     Regulatory Analysis 3.1     No Significant Hazards Consideration Analysis Entergy Operations Inc. (Entergy) requests adoption of TSTF-564, Safety Limit MCPR,"
11 Revision 2 (TSTF-564), which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Grand Gulf Nuclear Station, Unit 1 (GGNS) and the River Bend Station, Unit 1 (RBS) Technical Specifications (TS). The proposed change revises the GGNS and RBS TS safety limit on minimum critical power ratio (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition, to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of the two values applicable when one or two recirculation loops are in operation. TS 5.6.5 for GGNS and RBS, "Core Operating Limits Report (COLR)," are also revised to require the current SLMCPR value to be included in the COLR.
Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
: 1.      Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed amendments revise the TS SLMCPR and the list of core operating limits to be included in the COLR. The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure, for all accidents previously evaluated, that the fuel cladding will be protected from failure due to transition boiling. The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect functions of preventing or mitigating any accidents previously evaluated.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

CNR0-2019-0000 9 Enclosure Page 3 of 4
: 2.      Do the proposed amendments create the possibility of a new or different kind of accident from any previously evaluated?
Response: No The proposed amendments revise the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems, or components (SSCs). No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Do the proposed amendments involve a significant reduction in a margin of safety?
Response: No The proposed amendments revise the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in TSTF-564, changing the SLMCPR methodology to one based on a 95%
probability with 95% confidence level that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9% of the fuel rods are not susceptible to boiling transition, does not have a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce margin of safety.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on the above, Entergy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazard consideration" is justified.
3.2     Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

CNR0-2019-00009 Enclosure Page 4 of 4

==4.0       ENVIRONMENTAL CONSIDERATION==

The proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.

==5.0      REFERENCES==

5.1      Technical Specification Task Force (TSTF) Traveler 564, "Safety Limit MCPR, 11 Revision 2.
5.2      Final Safety Evaluations of Technical Specification Task Force Traveler TSTF-564, Revision 2, 11 Safety Limit MCPR, 11 Using the Consolidated Line Item Improvement Process, (CAC No. MG0161, EPID L-2017-PMP-0007), dated November 19, 2018, ADAMS Accession Nos. ML18299A054 (Cover Letter); ML18299A069 (Final Traveler SE); ML18299A048 (Package).
5.3      Letter from Brian R. Moore, Global Nuclear Fuel, to U.S. NRC, 11 Revised Information Supporting TSTF-564 Safety Limit Minimum Critical Power Ratio, 11 July 30, 2018, ADAMS Accession Nos. ML18212A018, ML18212A019, ML18212A020, ML18212A021.
5.4      GE Nuclear Energy, "General Electric BWR Thermal Analysis Basis (GETAB):
Data, Correlation and Design Application, 11 NED0-10958-A, January 1977, ADAMS Accession No. ML102290144.
5.5      GE Nuclear Energy, 11 Power Distribution Uncertainties for Safety Limit MCPR Evaluations," NEDC-32694P-A, August 1999, ADAMS Accession No. ML003740166.
5.6      GE Nuclear Energy, "Methodology and Uncertainties for Safety Limit MCPR Evaluations, 11 NEDC-32601-P-A, August 1999, ADAMS Accession No. ML003740166.

Enclosure, Attachment 1.a CNR0-2019-00009 Markup of Technical Specification (TS) Pages Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed Facility Operating License No. NPF-29 Unit 1 TS Pages 2.0-1 3.2-2*
5.0-18 Unit 1 TS Page 3.2-2 is included for reference only. There are no changes on the page.

SLs 2.0 2.0  SAFETY LIMITS (SLs) 2.1  SLs 2.1.1  Reactor Core SLs 2.1.1.1  With the reactor steam dome pressure< 685 psig or core flow< 10% rated core flow:
THERMAL POWER shall bes 21.8% RTP.
2.1.1.2  With the reactor         dome flow;;::: 10% rated MCPR shall operation or operation.
2.1.1.3  Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2   Reactor Coolant System Pressure SL Reactor steam dome pressure shall bes 1325 psig.
2.2  SL Violations With any SL violation, the following actions shall be completed within 2 hours:
2.2.1   Restore compliance with all SLs; and 2.2.2   Insert all insertable control rods.
GRAND GULF                               2.0-1      Amendment No. -+/--4 -+/--8-4-, M-9-,-- -+/--9-+/--, 203

INo Changes. Included for Reference I                      MCPR 3.2.2 3.2   POWER DISTRIBUTION LIMITS 3.2.2   Minimum Critical Power Ratio (MCPR)
LCO   3.2.2        All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.
APPLICABILITY:     THERMAL POWER~ 21.8     RTP.
ACTIONS CONDITION                    REQUIRED ACTION            COMPLETION TIME A. Any MC?R not within        A. l     Restore MCPR(s)  to      2 hours limits                              within limits.
B. Required Action and        B .1     Reduce THERMAL POWER     4 hours associated Completion               to < 21. 8% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                 FREQUENCY SR   3.2.2.1     Verify all MCPRs are greater than or equal        Once within to the limits specified in the COLR.              12 hours after
                                                                    ~ 21.8% RTP AND 24 hours thereafter SR   3.2.2.2     Determine the MCPR limits.                       Once within 72 hours after each completion of SR 3.1.4.1 Once within 72 hours after each completion of SR 3.1.4.2 Once within 72 hours after each completion of SR 3.1.4.4 GRAND GULF                              3.2-2                 Amendment N o . ~ 191

Reporting Requirements
: 5. 6 5.6    Reporting Requirements 5.6.2         Annual Radiological Environmental Operating Report.         (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.      The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3         Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year.      The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.        The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV. B. l.
5.6.4         Deleted
: 5. 6.5        Core Operating Limits Report       (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1)   LCO 3. 2 .1,  Average Planar Linear Heat Generation Rate (APLHGR),
: 2)   LCO 3. 2. 2,  Minimum Critical Power Ratio     (MCPR)
(including
: 3)   LCO 3. 2. 3,  Linear Heat Generation Rate    (LHGR),              power and flow
: 4)   Deleted dependent
: 5)   LCO 3. 3 .1.1, RPS   Instrumentation,  Table 3.3.1.1-1 APRM       limits, and the Function 2. f                                                     cycle-specific
: 6)   Deleted                                                           MCPR99.9%),
continued GRAND GULF                                 5.0-18          Amendment No .              188

Enclosure, Attachment 1.b CNR0-2019-00009 Markup of Technical Specification (TS) Pages River Bend Station, Unit 1 NRC Docket No. 50-458 Renewed Facility Operating License No. NPF-47 Unit 1 TS Pages 2.0-1 3.2-2*
5.0-18 Unit 1 TS Page 3.2-2 is included for reference only. There are no changes on the page.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1    SLs 2.1.1    Reactor Core SLs 2.1.1.1     With the reactor steam dome pressure < 685 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be~ 23.8% RTP.
2.1.1.2     With the reactor steam dome pressure :2:: 685 psig and core flow
:2:: 10% rated core flow:
                                                  ~
MCPR shall be :2:: 1.08 for t'wo recirculation loop operation or :2:. 1.10 for single recirculation loop operation.
2.1.1.3     Reactor vessel water level shall be greater than the top of active irradiated fuel.
: 2. 1.2  Reactor Coolant System Pressure SL Reactor steam dome pressure shall be        ~ 1325 psig.
2.2   SL Violations With any SL violation, the following actions shall be completed:
2.2.1   Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50. 72.
2.2.2   Within 2 hours:
2.2.2.1     Restore compliance with all SLs; and 2.2.2.2     Insert all insertable control rods.
2.2.3   Within 24 hours, notify the plant manager and the corporate executive responsible for overall plant nuclear safety.
(continued)
RIVER BEND                                     2.0-1              Amendment No. 84, Be, 99, 4-05, 444,~482-

INo Changes. Included for Reference I                         MCPR 3.2.2 3.2  POWER DISTRIBUTION LIMITS LCO 3.2.2           All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.
APPLICABILITY:      THERMAL POWER~ 23.8% RTP.
ACTIONS CONDITION                         REQUIRED ACTION           COMPLETION TIME A. Any MCPR not within             A.1 Restore MCPR(s) to within     2 hours limits.                              limits.
B. Required Action and             8.1 Reduce THERMAL POWER          4 hours associated Completion                to < 23.8% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                            FREQUENCY SR 3.2.2.1       Verify all MCPRs are greater than or equal to the      Once within limits specified in the COLR.                          12 hours after
                                                                         ?:: 23.8% RTP 24 hours thereafter RIVER BEND                                     3.2-2                 Amendment No. 84, 114

Reporting Requirements 5.6 Reporting Requirements 5.6.2        Annual Radiological Environmental Operating Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3        Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4        Deleted 5.6.5        CORE OPERATING LIMITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1)    LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),
: 2)    LCO 3.2.2, Minimum Critical Power Ratio (MCPR)(including power and flow dependent liroi~.
: 3)    LCO 3.2.3, Linear Heat Generation Rate (LHGR)(including power and flow dependent limits).
: 4)    LCO 3.2.4, Fraction of Core Boiling Boundary (FCBB)
: 5)    LCO 3.3.1.1, RPS Instrumentation (RPS), Function 2.b
: 6)     LCO 3.3.1.3, Periodic Based Detection System (PBDS)
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.
(continued)
RIVER BEND                                   5.0-18    Amendment No. 81 100 106 122 135,446

Enclosure, Attachment 2.a CNR0-2019-00009 Clean Technical Specification (TS) Pages Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed Facility Operating License No. NPF-29 Unit 1 TS Pages 2.0-1 3.2-2*
5.0-18 Unit 1 TS Page 3.2-2 is included for reference only. There are no changes on the page.

SLs 2.0 2.0  SAFETY LIMITS (SLs) 2.1  SLs 2.1.1  Reactor Core SLs 2.1.1.1  With the reactor steam dome pressure< 685 psig or core flow< 10% rated core flow:
THERMAL POWER shall be~ 21.8% RTP.
2.1.1.2  With the reactor steam dome pressure~ 685 psig and core flow~ 10% rated core flow:
MCPR shall be~ 1.07 2.1.1.3  Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2   Reactor Coolant System Pressure SL Reactor steam dome pressure shall be~ 1325 psig.
2.2  SL Violations With any SL violation, the following actions shall be completed within 2 hours:
2.2.1   Restore compliance with all SLs; and 2.2.2   Insert all insertable control rods.
GRAND GULF                           2.0-1        Amendment No. -3:-4 +/-8-4,- -3:--8-9, -3:-9-+/--;- 203

No change. Included for reference.
MCPR 3.2.2 3.2   POWER DISTRIBUTION LIMITS 3.2.2   Minimum Critical Power Ratio        (MCPR)
LCO    3.2.2        All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.
APPLICABILITY:      THERMAL POWER 2 21.8% RTP.
ACTIONS CONDITION                         REQUIRED ACTION           COMPLETION TIME A. Any MCPR not within         A. l        Restore MCPR(s) to      2 hours limits                                  within limits.
B. Required Action and         B.l         Reduce THERMAL POWER    4 hours associated Completion                    to< 21.8% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                   FREQUENCY SR   3.2.2.1     Verify all MCPRs are greater than or equal           Once within to the limits specified in the COLR.                  12 hours after 2 21.8  RTP 24 hours thereafter SR   3.2.2.2     Determine the MCPR limits.                           Once within 72 hours after each completion of SR 3. 1. 4. 1 Once within 72 hours after each completion of SR 3. 1. 4. 2 Once within 72 hours after each completion of SR 3.1.4.4 GRAND GULF                                   3.2-2                Amendment No. +/--re, 191

Reporting Requirements 5.6 5.6    Reporting Requirements 5.6.2         Annual Radiological Environmental Operating Report             (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.      The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3          Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year.      The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.        The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.l.
: 5. 6. 4        Deleted 5.6.5          Core Operating Limits Report       (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1)  LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),
: 2)  LCO 3.2.2, Minimum Critical Power Ratio (MCPR) (including power and flow dependent limits, and the cycle-specific MCPR 99 . 0 ) ,
: 3)  LCO 3.2.3, Linear Heat Generation Rate (LHGR),
: 4)  Deleted
: 5)  LCO 3.3.1.1, RPS Instrumentation, Table 3.3.1.1-1 APRM Function 2.f
: 6)  The Manual Backup Stability Protection (BSP) Scram Region (Region 1), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal Power - High trip function (Function 2.d) setpoints used in the OPRM Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1.
GRAND GULF                                 5.0-18      Amendment No. -3::-5-+, -3:-6-=7-, :l:-8-&, 205

Enclosure, Attachment 2.b CNR0-2019-00009 Clean Technical Specification (TS) Pages River Bend Station, Unit 1 NRC Docket No. 50-458 Renewed Facility Operating License No. NPF-47 Unit 1 TS Pages 2.0-1 3.2-2*
5.0-18 Unit 1 TS Page 3.2-2 is included for reference only. There are no changes on the page.

SLs 2.0 2.0  SAFETY LIMITS (SLs) 2.1  SLs 2.1.1    Reactor Core SLs 2.1.1.1    With the reactor steam dome pressure < 685 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be           ~ 23.8% RTP.
2.1.1.2    With the reactor steam dome pressure       ~ 685 psig and core flow
                         ~ 10% rated core flow:
MCPR shall be~ 1.07.
2.1.1.3    Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2    Reactor Coolant System Pressure SL Reactor steam dome pressure shall be       ~ 1325 psig.
2.2  SL Violations With any SL violation, the following actions shall be completed:
2.2.1   Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50. 72.
2.2.2   Within 2 hours:
2.2.2.1     Restore compliance with all SLs; and 2.2.2.2     Insert all insertable control rods.
2.2.3   Within 24 hours, notify the plant manager and the corporate executive responsible for overall plant nuclear safety.
(continued)
RIVER BEND                                     2.0-1             Amendment No. 84, Se, 99, ~ '
444,.:t-n,.~

INo Changes. Included for Reference I                         MCPR 3.2.2 3.2  POWER DISTRIBUTION LIMITS LCO 3.2.2           All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.
APPLICABILITY:      THERMAL POWER~ 23.8% RTP.
ACTIONS CONDITION                         REQUIRED ACTION           COMPLETION TIME A. Any MCPR not within             A.1 Restore MCPR(s) to within     2 hours limits.                              limits.
B. Required Action and             8.1 Reduce THERMAL POWER          4 hours associated Completion                to < 23.8% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                            FREQUENCY SR 3.2.2.1       Verify all MCPRs are greater than or equal to the      Once within limits specified in the COLR.                          12 hours after
                                                                         ~23.8% RTP 24 hours thereafter RIVER BEND                                     3.2-2                 Amendment No. 84, 114

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2         Annual Radiological Environmental Operating Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3         Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4         Deleted 5.6.5         CORE OPERATING LIMITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1)     LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),
: 2)     LCO 3.2.2, Minimum Critical Power Ratio (MCPR)(including power and flow dependent limits and the cycle-specific MCPR99.9%).
: 3)     LCO 3.2.3, Linear Heat Generation Rate (LHGR)(including power and flow dependent limits).
: 4)     LCO 3.2.4, Fraction of Core Boiling Boundary (FCBB)
: 5)     LCO 3.3.1.1, RPS Instrumentation (RPS), Function 2.b
: 6)    LCO 3.3.1.3, Periodic Based Detection System (PBDS)
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.
(continued)
RIVER BEND                                   5.0-18    Amendment No. 81 100 106 122 135 14 5          I

Enclosure, Attachment 3.a CNR0-2019-00009 FOR INFORMATION ONLY Markup of Technical Specification (TS) Bases Pages - For Information Only Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed Facility Operating License No. NPF-29 Unit 1 TS Bases Pages B 2.0-1 B 2.0-2 B 2.0-3 B 2.0-3a B 2.0-4 B 2.0-5 B 2.0-6 B 3.2-5 B 3.2-6 B 3.2-7 B 3.2-8 B 3.2-Ba

Reactor Core SLs B 2.1.1 B 2. 0 SAFETY LIMITS (SLs)
B 2.1.1    Reactor Core SLs BASES BACKGROUND          GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded *during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs}.
The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit
* is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an ~L, such that the MCPR is not less than the limit specified 1n Specification 2 .1.1. 2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result
~~~~~~~~~~~~--from thermal stresses, which occur from reactor operation 1his is accomplished by having a significantly above design conditions.
,afety Limit Minimum Critical
)ewer Ratio (SLMCPR) design          While fission product migration from cladding perforation is just as measurable as that from use related cracking, the
>asis, referred to as SLMCPR9s19s, thermally caused cladding perforations signal a threshold vhich corresponds to a 95%            beyond which still. greater thermal stresses may cause gross,
>robability at a 95% confidence       rather than incremental, cladding deterioration. Therefore,
~vel (the 95/95 MCPR criterion)       t     uel cladding SL is defined with a margin to the hat transition boiling will not       condi t
* that would produce onset of transition boiling
>ccur.                                 (i.e., MCP - 1.00). These conditions represent a
~~~~~~~~~~~~ __.significant dep                           re from the condition intended by design for planned operatio . The HCPR fuel cladding integrity 81 ensures that during normal operation and during AOOs, at least 99. 9% of the fuel rods in the core do not e1cperience transition boiling.
(continued}
GRAND GULF                              B 2.0-1                      Revision No. 0

Reactor Core SLs B 2.1.1 BASES BACKGROUND         Operation above the boundary of the nucleate boiling regime (continued)     could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
APPLICABLE         The fuel cladding must not                              of SAFETY ANALYSES    normal operation and AOOs
-he Tech Spec SL is set 1enerically on a fuel product JlCPR correlation basis as the   The Reactor Protection System setpoints (LCO 3.3.1.1, JlCPR which corresponds to a     "Reactor Protection System (RPS) Instrumentation"), in combination with other LCOs, are designed to prevent any
}5% probability at a 95%         anticipated combination of transient conditions for Reactor
:onfidence level that transition Coolant System water level, pressure, and THERMAL POWER
,oiling will not occur, referred level that would result in reaching the MCPR SL.
o as SLMCPR9s19s.
2.1.1.1    Fuel Cladding Integrity The use of the fuel vendor's critical power correlations are valid for critical power calculations at pressures 2 685 psig and core flows 2 10% of rated (Ref. 3, 5, and 6).
For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flow will always be
                                      > 4.5 psi. Analyses show that with a bundle flow of 28 x 10 3 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus the bundle flow with a 4.5 psi driving head will be > 28 x 10 3 lb/hr. Full scale (continued)
GRAND GULF                            B 2.0-2                        LBDCR 12035

Reactor Core SLs B 2.1.1 BASES APPLICABLE      2.1.1.1     Fuel Cladding Integrity    (continued)
SAFETY ANALYSES ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER> 44.2% RTP. Thus a THERMAL POWER limit of 21.8%
RTP [1.2 / (4408 MWt/800 bundles)] for reactor pressure
                          < 685 psig is conservative. Because of the design thermal hydraulic compatibility of the reload fuel designs with the cycle 1 fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.
l l n s e r t A ~2.1.1.2
'--~~-.,,..--                  MCPR
                    ;he MCPR SL eRslires slifficieRt co11servatisR1 i11 the operatiRg MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin bet*n*een calculated boiling transition Ci. e. ,
MCPR    1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. Reference 6 describes the methodology used in determining the MCPR SL.
The calculated MCPR safety limit is reported to the customary three significant digits (i.e., X.XX); the MCPR operating limit is developed based on the calculated MCPR safety limit to ensure that at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.
The fuel vendor's critical power correlations are based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is *.o'i thin a small percentage of the actual critical power being estimated. As long as the core pressure and flm, are within the range of validity of the correlations, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the (continued)
GRAND GULF                         B 2.0-3                            LBDCR 12035

Reactor Core SLs B 2.1.1 BASES inherent accuracy of tho fuel vondo::::-' s co::::-ro 1 at ion prov1 ao ~
reasonable degree o+/-: assurance that 99. 9?:: of tho rods in the core \.'ould not be susceptible to transition bo~ 2-ing during sustained operation at -:he ~~CPR ~L.       If boiling transit:on
           ,,.'ere to occur, thc::::-c is reason to bcl ~ eve that tho i~1tegr 1 t_/
GRAND GULF                     B 2.0-3a                                   LDC 98033

Reactor Core SLs B 2.1.1 BASES APPLICABLE      _2___.1__.__1....;...;.2"'--~MC__P__R (continued) of the fuel ',JOuld not be comproFHised. Significant test data accmHulated by the HRC and private organizations indicate that the use of a boiling transition limitat:on to protect against cladding failure is a very conservative approach.
Much of the data indicate that BWR fuel can survive for an rntended period of time in an environment of boiling tr::r.::i ~:c~~.
2.1.1.3                     Reactor Vessel Water Level During MODES 1 and 2, the reactor vessel ,,~ater level is required to be above the top of the active fuel to provide core cooling capability. With fuel                          the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should                                below       top of t0e active irradiated fuel during this                                   the
                           . V                                               .     ;:;
cooling capability                             lead to elevated cladding temperatures and clad perforation in the event that the water level becomes less than two-thirds                                 core heiaht.
reactor vessel water level SL has                     established at top of                           irradiated fuel to provide a point can be                           to also                      margin effect SAFETY                     reactor core                    are established integrity of the                           clad barrier to 1ve SL 2.1.1.2 ensure that the core operates                                the fuel design criteria.                           2.1.1.3 ensures that the reactor vessel water level is greater than the top of                           active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforation.
APPLICll.BILITY SLs 2 .1.1.1, 2 .1.1. 2, and 2 .1.1. 3 are applicable m all MODES.
(continued}
B 2.0-4                         Revision No. 0

PAGE INTENTIONALLY LEFT BLANK Reactor Core Sls B 2.1.1 BASES    (continued)
SAFETY LIMIT                    an SL may cause fuel        and create a           l l.
VIOLATIONS                            releases in excess of 10 CFR 50.67 limits Therefore,    is control rods (The                                  of the reactor water level to restore water level and vessel,     necessary, for ECCS              The 2 hour Time ensures that the remedial                  ensures that the              of an
                              +-                      oer    is GRAND GULF                              B 2.0-5                          LDC 01050

Reactor Core Sls B 2. .1 BASES   (continued)
I. 10 CFR 50, Appendix A,        CDC 10.
: 2. ANF-524 (P) (A),     Revision 2,  Supplements 1 and 2, November 1990.
: 3. EMF-220 9 ( P) (A),   Revision 2,            2003.
: 4. 10 CFR 50. 67,      "Accident Source Term."
: 5. NEDC-33383-P, Revision 1, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," June, 2008.
: 6. NEDE-24 011-P-A,      GESTAR- II.
GRAND GULF                             B 2.0-6

Insert A:
2.1.1.2 MCPR GE fuel The fuel cladding integrity SL 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. The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95%
confidence that rods are not susceptible to boiling transition, referred to as MCPR9s19s.
The SL is based on GNF2 fuel. For cores with a single fuel product line, the SLMCPR9s195 is the MCPR9s19s for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPRg 5,95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh- or once-burnt at the start of the cycle.

MCPR B 3.2 .2 B 3.2     POWER DISTRIBUTION LIMITS B 3. 2. 2   Minimum Critical Power Ratio       (MCPR)
BASES BACKGROUND           MCPR is a ratio of the fuel assembly pcwer that would result in the onset of boiling transition to the actual fuel assembly power.     'fire HCFR :3afet) bimH (Sl3) is ssi& Sl!leQ :t.l;iei:t 99.9-o or the fuel rods aooid boilin'j traFisiH@Ft if ;tl;;i,., Jiwit 1s not viola Led. (refer to the Bases fot1 Sn 2 .1.1. 2) . The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOO     Although fuel damage does not necessarily occur if and that 99. 9% of the fuel         a fue rod actually experiences boiling transition (Ref. 1),
ods are not susceptible to            the critical power at which boiling transition is calculated
>oiling transition if the limit is   to occur has been adopted as a fuel design criterion.
1ot violated. Although fuel The onset of transition boiling is a phenomenon that is tamage ...                            readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,
the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling).         Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.
APPLICABLE            The analytical methods and assumptions used in evaluating SAFETY ANALYSES      the AOOs to establish the operating limit MCPR are presented in the UFSAR, Chapters 4, 6, and 15, and References 2, 3, 4, and 5. To ensure that the MCPR             is not exceeded during any transient event              ccurs with moderate frequency, limitin         1ents have been analyzed to determine the rgest reduction in critical power ratio (CPR) . The types Safety Limit (SL)                   of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant                  decrease. The limiting trans the largest change in CPR ( CPR) . When the largest CPR is the required operating limit MCPR is ombined with the
;LMCPR99.9%,
               -j l~nsertB
                             ~
1....-                        ----- -~~                          (continued)
GRAND GULF                                B 3.2-5                              LBDCR 09037

MCPR B 3.2.2 BASES                                                ie               )~e      MCPR9 g 9% value and APPLICABLE            The MCPR operating limits ~rived from the transient SAFETY ANALYSES        analysi are dependent on the operating core flow and power (continued)                (MCPRf and MCPRp, respectively) to ensure adherence to design limits during the worst transient that occurs moderate frequency (Refs. 3, 4, and 5). Flow dependent limits are determined by steady state thermal hydraulic
                , and             methods using the three dimensional BWR simulator code (Ref.
: 6) and the steady state thermal hydraulic code (Ref. 2).
MCPRf curves are provided based on the maximum credible flow runout transient for Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control.
Power dependent MCPR limits (MCPRp) are determined by tA-e-approved transient           three dimensional BWR simulator code and the one dimensional analysis models              transient code (Ref. 7). The MCPRp limits are established for a set of exposure intervals. The limiting transients are analyzed at the limiting exposure for each interval.
Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 21.8% RTP and the previously mentioned bypass power level.
The MCPR satisfies Criterion 2 of the NRC Policy Statement*.--~~~~
            - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . . , , : : : : _ . ~ M C P R99 _9%
LCO                   The MCPR operating limits specified in the COL                           value, re the result of the Design Basis Accident (DBA) and transient                  MCPRt analysis. The MCPR operating limits are determined by the values, and larger of the MCPRf and MCPRp limit .                                    MCPRp
            --------------- ----~.....:=::. --------------- --1values APPLICABILITY                                limits are primarily derived from nsient analyses that are assumed to occur at high power levels. Below 21.8% RTP, the reactor is operating at a slow
, which are                       recirculation pump speed and the moderator void ratio is based on the                      small. Surveillance of thermal limits below 21.8% RTP is unnecessary due to the large inherent margin that ensures MCPR99.9% limit                   that the MCPR SL is not exceeded even if a limiting specified in the                  transient occurs.
COLR.
(continued)
GRAND GULF                                  B 3.2-6                                  LBDCR 12035

                !No Changes. Included for Reference I                 MCPR B 3.2.2 BASES APPLICABILITY  Studies of the variation of limiting transient behavior have (continued) been performed over the range of power and flow conditions.
These studies encompass the range of key actual plant parameter values important to typically limiting transients.
The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 21.8% RTP.
This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor (IRM) provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern.
Therefore, at THERMAL POWER levels< 21.8% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.
ACTIONS If any MCPR is outside the required limit, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limit(s) such that the plant remains operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limit and is acceptable based on the low probability of a transient or OBA occurring simultaneously with the MCPR out of speci fi cation.
If the MCPR cannot be restored to within the required limit within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to< 21.8% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to< 21.8% RTP in an orderly manner and without challenging plant systems.
(continued)
GRAND GULF                         B 3.2-7                       LBDCR 12035

MCPR B 3.2.2 BASES   (continued)
SURVEILLANCE        SR   3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is 2 21.8% RTP and then every 24 hours thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER reaches 2 21.8% RTP is acceptable given the large inherent margin to operating limits at low power levels.
SR 3.2.2.2 Because the transient analyses may take credit for conservatism in the control rod scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analyses. SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution.
The MCPR operating limit is then determined based either on the applicable limit associated with scram times of LCO 3.1.4, "Control Rod Scram Times," or the realistic scram times. The scram time dependent MCPR limits are contained in the COLR. This determination must be performed and any necessary changes must be implemented within 72 hours after each set of control rod scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.
REFERENCES          1. NUREG-0562, "Fuel Failures As A Consequence of Nucleate Boiling or Dry Out," June 1979.
: 2. NEDE-24011-P-A General Electric Standard Application for Reactor Fuel (GESTAR II).
: 3. UFSAR, Chapter 15, Appendix 15B.
: 4. UFSAR, Chapter 15, Appendix 15C.
(continued)
GRAND GULF                             B 3.2-8                        LBDCR 12035

MCPR B 3.2.2 BASES   (continued)
SURVEILLANCE        REFERENCES (continued)
REQUIREMENTS
: 5. UFSAR, Chapter 15, Appendix 150.
: 6. NEDE-30130-P-A, Steady-State Nuclear Methods.
: 7. NED0-24154, Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors.
GRAND GULF                           B 3.2-8a                        LBDCR 12035

Insert B:
MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition 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 approved Critical Power correlations. Details of the MCPR 99 _9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR 99 _9% statistical analysis.

Enclosure, Attachment 3.b CNR0-2019-0000 9 FOR INFORMATION ONLY Markup of Technical Specification (TS) Bases Pages - For Information Only River Bend Station, Unit 1 NRG Docket No. 50-458 Renewed Facility Operating License No. NPF-47 Unit 1 TS Bases Pages B 2.0-1 B 2.0-2 B 2.0-3 B 2.0-4 B 3.2-5 B 3.2-6 B 3.2-7 B 3.2-8

Reactor Core Sls B 2.1.1 B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND           GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,
MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MGPR fuel RIVER BEND                                   B 2.0-1                                Revision No. Q

Reactor Core Sls B 2.1.1 BASES BACKGROUND       Operation above the boundary of the nucleate boiling regime could result (continued)     in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.
Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
APPLICABLE      The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A~The reactor core Sls are established le &#xb5;reel11de
             ~~====::::ablished, The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR SL.
2.1.1.1        Fuel Cladding Integrity The use of the fuel vendor's critical power correlations are valid for critical power calculations at pressures ~ 685 psig and core flows ~ 10% of rated flow (Ref. 2. 7 and 8). For operation at low pressures or low flows, another basis is used, as follows:
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14. 7 psi a to 700 psi a indicate that the fuel (continued)
RIVER BEND                               B 2.0-2                               Revision No. 4&sect;9

Reactor Core SLs B 2.1.1 BASES APPLICABLE       2.1.1.1       Fuel Cladding Integrity SAFETY ANALYSES (continued)            assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors, this corresponds to a THERMAL POWER> 50% of original RTP. Thus, a THERMAL POWER limit of 23.8% RTP for reactor pressure < 685 psig is conservative.
Because of the design thermal hydraulic compatibility of the reload fuel designs with the cycle 10 fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.
2.1.1.2       MCPR The MGPR SL ensures sufficient conservatism in the operating limit MGPR limit that, in the event of an AGO from the limiting condition of operation, at least 99.9% of the fuel rods in the core *would be expected to avoid boiling transition. The margin betv..'een calculated boiling transition (i.e., MGPR 1.00) and the MGPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. Reference 6 describes the methodology used in determining the MGPR SL.
The calculated MGPR safety limit is reported to the customary three significant digits (i.e., X.XX); the MGPR operating limit is developed based on the calculated MGPR safety limit to ensure that at least 99.9%
of the fuel rods in the core are expected to avoid boiling transition.
The fuel vendor's critical power correlations are based on a significant body of practical test data, providing a high degree of assurance that the critical pov,er, as evaluated by the correlation, is *11ithin a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the correlations, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the inherent accuracy of the fuel vendor's correlation prm,'ide a reasonable degree of assurance that 99.9% of the rods in the core would not be susceptible to transition boiling during (continued)
RIVER BEND                                B 2.0-3                               Revision No. ~

Reactor Core SLs B 2.1.1 BASES sustained operation at the MGPR SL. If boiling transition were to occur, there is reason to believe that the integrity of the fuel ',Nould not be compromised. Significant test data accumulated by the NRG and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BVVR fuel can survive for an extended period of time in an environment of boiling transition.
2.1.1.3       Reactor Vessel Water Level During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes less than two-thirds of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.
SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.
SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.
APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
SAFETY LIMIT VIOLATIONS If any SL is violated, the NRC Operations Center must be notified within 1 hour, in accordance with 10 CFR 50.72 (Ref. 3).
(continued)
RIVER BEND                             B 2.0-4                             Revision No. 646

MCPR B 3.2.2 B 3.2   POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
BASES BACKGROUND         MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MGPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AO         . Although fuel damage does not n                          e ro actually experiences boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.
The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.
6 APPLICABLE        The analytical methods and assumptions use *n evaluating the AOOs to SAFETY ANALYSES   establish the operating limit MCPR are present in the USAR, Chapters 4, 6, and 15, and References 2, 3, anc!A( To ensure that the MCPR                   (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (LlCPR). When the largest LlCPR is                    added to the SLMCPR99 9 % Sb, the required operating limit MCPR is obtained.
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RIVER BEND                                  B 3.2-5                               Revision No. G

MCPR B 3.2.2 MCPR99.9%
BASES                                                              \
APPLICABLE      r i h e MCPR operating limits are derived fro~the transient analysis, and SAFETY ANALYSE            re dependent on the operating core flow and power state (MCPRt and (continued)            MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency. Flow dependent MCPR limits (MCPRt ) are determined by steady state thermal hydraulic urJTlr-u,  methods using the three dimensional BWR simulator code (Ref. 5) and the multi channel thermal hydraulic code (Ref. 2). MCPRt curves are provided based on the maximum credible flow runout transient for Non Loop Manual operation. Non Loop Manual operation bounds Loop Manual because Non Loop Manual operation can result in a more severe flow runout transient. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control. Non Loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow.                             approved J  analysis Power dependent MCPR limits (MCPRp) are determined b/fhe three dimensional BVVR simulator code and the one dimensional transient code (Ref. 2). The MCPR limits are established for a set of exposure intervals.
The limiting transients are analyzed at the limiting exposure for each interval. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 23.8% RTP and the previously mentioned bypass power level.
The MCPR satisfies Criterion 2 of the NRC Policy Statement.         MCPR              )
                                                                                     ~             P va 1ues LCO                    The MCPR operating limits specified in the COLR       ~
Design Basis Accident (OBA) and transient analysis. The MCPR result of the operating limits are determined by the larger of the MCPRt and MCPRp
               ~-
APPLICABILITY          The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 23.8% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 23.8% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.
Studies of the variation of limiting transient behavior have (continued)
RIVER BEND                                     B 3.2-6                             Revision No. 6--4&sect;

INo Changes.      Included for Reference     I                         MCPR B 3.2.2 BASES APPLICABILITY   been performed over the range of power and flow conditions. These (continued)  studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 23.8% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor (IRM) provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels
                < 23.8% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.
ACTIONS If any MCPR is outside the required limit, an assumption regarding an initial condition of the design basis transient analyses may not be met.
Therefore, prompt action should be taken to restore the MCPR(s) to within the required limit(s) such that the plant remains operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limit and is acceptable based on the low probability of a transient or OBA occurring simultaneously with the MCPR out of specification.
If the MCPR cannot be restored to within the required limit within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23.8% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23.8% RTP in an orderly manner and without challenging plant systems.
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RIVER BEND                              B 3.2-7                              Revision No. 6-4

MCPR B 3.2.2 BASES (continued)
SURVEILLANCE      SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is 2: 23.8% RTP and periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER reaches 2: 23.8% RTP is acceptable given the large inherent margin to operating limits at low power levels. The time interval based Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. NUREG-0562, "Fuel Rod Failures As A Consequence of Nucleate Boiling or Dry Out," June 1979.
: 2. XN-NF-80-19(P)(A) Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," January 1987.
: 3. USAR, Chapter 4, Appendix 48.
: 4. USAR, Chapter 15, Appendix 158.
: 5. XN-NF-80-19(P)(A) Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," March 1983 (As Supplemented).
6.
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