CNL-21-053, Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2, in Support of Atrium 11 Fuel Use (TS-535)
| ML21204A128 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/23/2021 |
| From: | Jim Barstow Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21204A127 | List: |
| References | |
| CNL-21-053 | |
| Download: ML21204A128 (982) | |
Text
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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 < 585 psig or core flow
< 10% rated core flow:
THERMAL POWER shall be~ 23% RTP. 2.1.1.2 With the reactor steam dome pressure~ 585 psig and core flow
~ 10% rated core flo;,----D 1.05 MCPR shall be ~ 4-eOO for two recirculation loop operation or ~ 4-: for single 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 within 2 hours: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods. BFN-UNIT 1 2.0-1 Amendment No.~~. 293,295,'.t99
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Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 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) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCi=>R Operating Limits for Specification 3.2.2; e Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal MINIMUM Power-High Scram setpoints used in the Automated BSP CRITICAL POWER Scram Region, and the BSP Boundary for Specification RATIO (MCPR) 3.3.1.1; and and MCPR99.9% (5) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.
- 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:
- 1. NEDE 24011 PA, Ret.1isioR 16, GeneFal Eleotrio StanElarc:I ApplioatioR for Reaotor Fl:lel, Ootober 2007. ~Deleted.
------~
- 2. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
(continued) BFN-UNIT 1 5.0-24 Amendmen~. 2J4., ~ .
~.23-1-,348--
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 10. XN Ni; 84 105(P)(/\) Volume 1 and Volume 1 Supplements 1 and 2, XCOBR/\ T: /\ Computer Code for B\l\lR Transient Thermal l=lydraulie Core Analysis , Exxon Nuelear Company, February 1987. ----
Deleted.
- 11. ANF-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
- 12. /\Ni; Q13(P)(/\) Volume 1 Revision 1 and Volume 1 Supplements 2, a and 4, COTR/\NSA2: /\ Computer Program for Boiling Water Reaetor Transient Analyses, /\d 11aneed Nuelear Fuels Corporation, August 1990. Deleted.
- 13. ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors , Framatome ANP, September 2005.
- 14. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
- 15. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation , August 2000.
- 16. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluations dated April 27, 2012, and July 31, 2014.
- 17. EMF-2292{P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
(continued) BFN-UNIT 1 5.0-24b Amendment No.~.~.~.
~.~, ~
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 18. EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
- 19. BA\J\I 10255(P)iA), Re\<ision 2, Cycle Specific 0I\/0M Methodology Using the RA.MONAS FA Code, AREVA NP, May 2006. Deleted.
- 20. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod
~-
~ ANP~
ary 2008. r Methodology for Boiling Water Reactors, AREVA NP, Revision 1 1 A Re*1ision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA ~ Marci 20-1-0.~ 2014 Inc.
- 22. ANP 3140(P) Revision 0, Brooms Ferry Units 1, 2, and 3 Improved K factor Model for ACEJATRIUM 10XM Critical Power Correlation ,
AREVA NP, August 2012. ( Deleted.
- 23. NEDC-33075P-A, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, Revision 8, November 2013.
Insert A (continued) BFN-UNIT 1 5.0-24c Amendment No. 234, 2$, ~ .
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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 < 585 psig or core flow
< 10% rated core flow:
THERMAL POWER shall be::;; 23% RTP. 2.1.1.2 With the reactor steam dome pressure ~ 585 psig and core flow
~ 10% rated core flow~
1.05 MCPR shall be~ 1.06 for t\¥0 recirculation loop operation or
~ 1.08 for single 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 within 2 hours: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods. BFN-UNIT 2 2.0-1 Amendment No. 253, 256, 270, 280, 311,~
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Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 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) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating bimits for Specification 3.2.2; e Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal MINIMUM Power-High Scram setpoints used in the Automated BSP CRITICAL POWER Scram Region, and the BSP Boundary for Specification RATIO (MCPR) 3.3.1.1; and and MCPR99.9% (5) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.
- 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:
- 1. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
- 2. XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
- 3. EMF-85-74(P) Revision O Supplement 1(P)(A) and Supplement 2 (P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
(continued} BFN-UNIT 2 5.0-24 Amendment No. ~ . 309, 385-
Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 5.6 Re~ ents (continued) 1?
- 4. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 5. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
- 6. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986. ~
- 7. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN 2, Siemens Power Corporation, October 1999.
- 8. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
- 9. XN NF 84 10e(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XGOBRA T:
A Computer Gode for BWR Transient Thermal 1-tydraulie Gore Analysis , EJOEon Nuolear Company, February 1987. Deleted. 10.ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
- 11. ANF 913(P)(A) 'Jolume 1 Re*1ision 1 and Volume 1 Supplements 2, 3 and 4, GOTRANS.'\2: A Computer Program for Boiling lJ.'-ater Reaetor Transient Analyses, Ad*1anoed Nuclear Fuels Corporation, August 1990. ~,:------J Deleted.
12.ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.
- 13. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
(continued) BFN-UNIT 2 5.0-24a Amendment No. 287, dGQ, 344-
Reporting Requirements 5.6
. ~ . -5.6.5 --- --
CORE - - -LIMITS OPERATING --- - -(COLR) REPORT - -(continued)
--.1 5.~ rting Re~uirements (continued) k
- 14. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 15. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluation dated February 15, 2013, and July 31, 2014.
- 16. EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 17. EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis:
Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
- 18. BAW 10256(P)(A), Re¥isi8R 2, Cycle Specific DIVOM Methsdslsgy UsiRg the R.A.MON.A.5 F.A. Csde, .A.REV.A. NP, May 2008. ~
Deleted.
- 19. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
- Revision 1 I Inc.
- 20. A~ A RQ.sR 0, ACE/ATRIUM 10XM Critical Power Correlation, AREV~ , March 20-1-0~ 2014
- 21. ANP 3140(P) Re¥i6i8R 0, BF8WR6 Ferry URits 1, 2, aRd 3 lmpro\led K faetor Mo~el for ACE//\TRIUM 1OXM Critical Pswer Csrrelati8R, AREVANP, August2012 . ~ Deleted.
- 22. NEDC-33075P-A, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, Revision 8, November 2013.
INSERT A (continued) BFN-UNIT 2 5.0-24b Amendment No. ~ . 309, 3-14-,a43.~
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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 < 585 psig or core flow
< 10% rated core flow:
THERMAL POWER shall be~ 23% RTP. 2.1.1.2 With the reactor steam dome pressure ;;:: 585 psig and core flow
~ 10% rated core flo~
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Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 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) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating Limit& for Specification 3.2.2; (4) T anual Backup Stability Protection (BSP) Scram Region Region I), the Manual BSP Controlled Entry Region (Region 11), the MINIMUM modified APRM Flow Biased Simulated Thermal Power-High Scram CRITICAL POWER setpoints used in the Automated BSP Scram Region, and the BSP RATIO (MCPR) Boundary for Specification 3.3.1.1; and and MCPR99.9% (5) The RBM setpoints and applicable reactor thermal power ranges for 1 each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.
- 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 ihe following documents: ,
- 1. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
- 2. XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
- 3. EMF-85-74(P) Revision O Supplement 1(P)(A) and Supplement 2 (P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
* (continued)
BFN-UNIT 3 5.0-24 Amendment No. ~ . ~ . 24&.~.268~
Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 4. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 5. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
- 6. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
- 7. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBUR~ 2, S~ ens Power Corporation, October 1999. * "---0
- 8. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, Deleted.
- 9. ~::Fry~g::e(P){A) Volsrne ~ 1 2, spplernen~ and XGOBRA T: A Computer Gode for BWR Transient Thermal l=lydraulic Core Analysis, Exxon Nuclear Company, February 1987.
10.ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
- 11. ANF 913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2,
~Deleted. 3 and 4, GOTRANSA2: A Computer Program f:or Boiling 'Nater Reactor Transient Analyses , Ad 1o1anced Nuclear Fuels Corporation , August 1990. (continued) BFN-UNIT 3 5.0-24a Amendment No.~. ~ . ~ .
~ . 2:e8, --2--7-G
Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 5.6
- 12. ANF-1358{P){A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.
- 13. EMF-2209{P){A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
- 14. EMF-2245{P){A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 15. EMF-2361{P){A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluations dated February 15, 2013, and July 31, 2014.
- 16. EMF-2292{P){A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 17. EMF-CC-074{P){A), Volume 4, Revision 0, BWR Stability Analysis:
Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
- 18. BA'IN 10255(P)(A), Re1Jision 2, CyGle SpeGifiG DIVOM Methodology Using the R/\MONA5 FA Code, ARE\/,'\ NP, May 2008. ~
Deleted. 19 BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Bailin Water Reactors, AREVA NP, February 2008.
- Revision 1
- 20. A~ A . . , ACE/ATRIUM 10XM Critical Power Correlation, AREV~ NP. March 2G-48.~ 2014 Inc.
- 21. a ANP 3140(P) Re¥ision 0, Browns Ferry Units 1, 2, and lmpro¥ed K faetor Model for ACE!ATRIUM 10XM Critioal Po1Ner Correlation, AREVA NP, August 2012. ~
Deleted.
- 22. NEDC-33075P-A, GE Hitachi Boiling Water Reactor Detect and Suppress Solution -Confirmation Density, Revision 8, November 2013.
INSERT A (continued) BFN-UNIT 3 5.0-24b Amendment No. 24-r, ~ . 2-45, ~ . :268.~.~. 293-
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Reactor Core SLs B 2.1.1 BASES BACKGROUND incremental, cladding deterioration. Therefore, the fuel (continued) 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 MCPR fuel cladding integrity SL ensures that during normal operation and during abnormal operational transients, at least 99.9% of the fuel rods in the core do not experience transition boiling.This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as the SLMCPR95/95, which corresponds to a 95% probability of a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur. Operation above the boundary of the nucleate boiling regime 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 sustain damage as a result of SAFETY ANALYSES normal operation and abnormal operational transients. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.The Technical Specification SL is set generically on a fuel product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95. 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 limit. (continued) BFN-UNIT 1 B 2.0-2 Revision 0
Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity SAFETY ANALYSES (continued) Critical power correlations are valid over a wide range of conditions per References 2, 5, and 59, extending to expected conditions below 23% THERMAL POWER. For core thermal power levels at, or above 23% rated, the hot channel flow rate is expected to be >28,000 lbm/hr, (core flow not less than natural circulation i.e., ~25%-30% core flow for 23% power); therefore, the fuel cladding integrity SL is conservative relative to the applicable range of the critical power correlations. For operation at low pressure/flow conditions, consistent with the low power region of the Power/Flow operating map, another basis is used as follows: The static head across the fuel bundles is due to elevation effects from water solid channel, core bypass, and annulus regions, is approximately 4.5 psid. The pressure differential is maintained by the water solid bypass region of the core, along with the annulus region of the vessel. Elevation head provided by the bypass and annulus regions produces natural circulation flow conditions balancing pressure head with loss terms inside the core shroud. Natural circulation principles maintain a core plenum to plenum pressure drop of approximately 4.5 to 5 psid along the natural circulation flow line of the Power/Flow operating map. When power levels approach 23% rated, pressure drop and density head terms are closely balanced as power changes, such that natural circulation flow is nearly independent of reactor power. The flow characteristic is represented by the nearly vertical portion of the natural circulation line on the Power/Flow operating map. For a core pressure drop of approximately 4.5 to 5 psid, the hot channel flow rate is expected to be >28,000 lbm/hr in the region of operation when core power is < 23% with a corresponding core pressure drop of about 4.5 to 5 psid. (continued) BFN-UNIT 1 B 2.0-3 Revision 0, 68, 104, 112 October 24, 2018
Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity (continued) SAFETY ANALYSES For example, References 5 and 9 test data, taken at low pressures and flow rates, indicate assembly critical power in excess of 4 MWt, for flow rates indicative of natural circulation conditions. At 23% rated power, assembly average power is < 1.2 MWt. When considering design peaking factors, hot channel power could be expected to be on the order of 2 MWt. Consequently, operation up to 23% rated core power, with normal natural circulation available, is conservative even if reactor pressure is less than the lower pressure limit of the critical power correlation. When reactor power is significantly less than 23% of rated (e.g., below 10% of rated), hot channel flow supported by the available driving head may fall below 28,000 lbm/hr (along the lower portion of the natural circulation flow characteristic on the Power/Flow map). However, the critical power supported by the flow, remains above actual hot channel power conditions. The inherent characteristics of BWR natural circulation make core power/flow follow the natural circulation line as long as normal annulus water level is maintained. Operation below 23% rated core thermal power is conservatively acceptable, even for reactor operations at natural circulation. Adequate fuel thermal margins are maintained for low power conditions present during core natural circulation, even though the flow may be less than the critical power correlation applicability range. The low pressure safety limit value of 585 psig has been determined to adequately bound the minimum pressure that might occur while reactor power is at or above 23% of rated. This condition would most likely be created by a pressure regulator failure open transient (PRFO) that results in a rapid depressurization of the vessel and a subsequent scram. Reference 8 provides a detailed evaluation of this transient event, and provides the basis for the low pressure safety limit of 585 psig. (continued) BFN-UNIT 1 B 2.0-4 Revision 0, 68, 97, 104, 112 October 24, 2018
Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The fuel cladding integrity SL is set such that no 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 MCPR95/95.However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties. The MCPR SL is determined using a statistical model combining 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 AREVA critical power correlations. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. References 2, 3, 4, 5, and 6 describe the uncertainties and methodologies used in determining the MCPR SL.The SL is based on ATRIUM 11 fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burned at the start of the cycle. (continued) BFN-UNIT 1 B 2.0-5 Revision 0, 68, 104 October 20, 2016
Reactor Core SLs B 2.1.1 BASES (continued) SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 7). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.
- 2. EMF-2209(P)(A), SPCB Critical Power Correlation, (as identified in the COLR).
- 3. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co -
Resident Fuel, (as identified in the COLR).
- 4. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
- 5. ANP-10298PA Revision 0ANP-10298P-A Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NPInc., March 20102014.
- 6. ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.Not used.
- 7. 10 CFR 50.67.
- 8. ANP-3245P Revision 1, Browns Ferry Evaluation of PRFO Low Pressure Technical Specification Value, AREVA Inc., February 2014.
- 9. ANP-10335P-A Revision 0, ACE/ATRIUM 11 Critical Power Correlation, Framatome Inc., May 2018.
BFN-UNIT 1 B 2.0-7 Revision 0, 29, 68, 97, 104 October 20, 2016
SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) BASES BACKGROUND SDM requirements are specified to ensure:
- a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;
- b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and
- c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions. APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2, 3, 9, SAFETY ANALYSES and 1011) assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LCO 3.1.6, "Rod Pattern Control"). Also, SDM is assumed as an initial condition for the control rod removal error during refueling (Ref. 4) and fuel assembly insertion error during refueling (Ref. 5) accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more (continued) BFN-UNIT 1 B 3.1-1 Revision 0, 68 October 18, 2012
SDM B 3.1.1 BASES ACTIONS E.1, E.2, E.3, E.4, and E.5 (continued) or other reasons. It is not necessary to perform the SRs needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Action must continue until all required components are OPERABLE. SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can be made subcritical from any initial operating condition. This can be accomplished by a test, an evaluation, or a combination of the two. Adequate SDM is demonstrated before or during the first startup after fuel movement, or shuffling within the reactor pressure vessel, or control rod replacement. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial measured value must be increased by an adder, "R", which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity. If the value of R is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required (Ref. 7) . (continued) BFN-UNIT 1 B 3.1-6 Revision 0
SDM B 3.1.1 BASES (continued) REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
- 2. FSAR, Section 14.6.2.
- 3. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," Section S.2.2.3.1, October 2007.Not used.
- 4. FSAR, Section 14.5.34.3.
- 5. FSAR, Section 14.5.34.4.
- 6. FSAR, Section 3.6.5.24.4.
- 7. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, October 2007.Not used.
- 8. NRC 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 9. XN-NF 19(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design Not_
and Analysis," (as identified in the COLR). _ used.
- 10. EMF-2158(P)(A), "Siemens Po1Ner Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- B2," (as identified in the COLR).Not used.
- 11. ANP-10333P-A Revision 0, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," Framatome Inc., March 2018.
BFN-UNIT 1 B 3.1-8 Revision 0, 68 October 18, 2012
Control Rod Scram Times B 3.1.4 BASES (continued) APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY ANALYSES control rod scram function are presented in References 2, 3, and 4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met. The scram function of the CRD System protects the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)", and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient, using the methodology in Reference 11. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Refs. 5, 8, and 9Ref. 10) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits. Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref. 7). (continued) BFN-UNIT 1 B 3.1-27 Revision 0, 68 October 18, 2012
Control Rod Scram Times B 3.1.4 BASES (continued) REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.
- 2. FSAR, Section 3.4.6.
- 3. FSAR, Section 14.5.
- 4. FSAR, Section 14.6.
- 5. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, October 2007. Not used.
- 6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-8754, September 17, 1987.
- 7. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 8. XN-NF 1Q(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).Not used.
- 9. EMF-2158(P)~I\), "Siemens Po1A1er Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- B2," (as identified in the COLR).Not used.
- 10. ANP-10333P-A Revision O "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," Framatome Inc., March 2018.
- 11. ANP-10300P-A Revision 1, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios," Framatome Inc., January 2018.
BFN-UNIT 1 B 3.1-34 Revision 0, 68 October 18, 2012
Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP. The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA). This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1, 2, 11, and 12Reference 13. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, 2, 11, and 12 Reference 13. CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated. Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for UO2 have been shown to be insignificant below fuel energyThe CRDA analysis ensures that the potential release of radioactivity does not exceed that assumed in the plants licensing basis dose evaluation. The criteria of Regulatory Guide (RG) 1.236 (Ref. 14) is utilized with the steady state release fractions of RG 1.183 (July 2000 version, Ref. 15) and using the DG-1327 (July 2019 version, Ref. 16) transient fission gas release correlations and fission gas multipliers in making this determination. Consequently, the calculated offsite dose will be well within the required limits (Ref. 5). (continued) BFN-UNIT 1 B 3.1-41 Revision 0, 68 October 18, 2012
Rod Pattern Control B 3.1.6 BASES APPLICABLE depositions of 300 cal/gm (Ref. 3), the fuel damage limit of SAFETY ANALYSES 280 cal/gm provides a margin of safety from significant core (continued) damage which would result in release of radioactivity (Refs. 4 and 5). Generic evaluations (Refs. 1, 6, and 11) of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal/gm) have shown that if the peak fuel enthalpy remains below 280 cal/gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref. 7) and the calculated offsite doses will be well within the required limits (Ref. 5).The potential for CRDA to exceed ASME code pressure limits was addressed in Reference 13. The most severe power pulse for the CRDA occurs at the cold conditions at atmospheric pressure. Therefore, a significant amount of energy is required to simply pressurize the core. Although local pressure increase occurs within a few assemblies, there is little change in the core system pressure in either the startup range or the power range. Therefore, the CRDA would not result in a reactor pressure which would cause increased stress to exceed the "Service Limit C" ASME Boiler and Pressure Vessel Code. Control rod patterns analyzed in References 1, 11, and 12Reference 13, follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation. Analyses are performed using the Reference 11 13 methodology demonstrating the 280 cal/gm fuel damage limitRG 1.236 (Ref. 14) criteria will not be violated during a CRDA while following the BPWS mode of operation. The evaluation provided by the generic BPWS analysis (Ref. 8) allows a limited number (i.e., eight) and corresponding distribution of fully inserted, inoperable control rods, that are not in compliance with the sequence. When performing a shutdown of the plant, an optional BPWS control room sequence (Ref. 10) may be used provided that all withdrawn control rods have been confirmed to be coupled. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled. When using the Reference 10 control rod sequence for shutdown, the Rod Worth Minimizer may be reprogrammed. (continued) BFN-UNIT 1 B 3.1-42 Revision 0, 68 October 18, 2012
Rod Pattern Control B 3.1.6 BASES (continued) REFERENCES 1. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," Section 2.2.3.1, October 2007.Not used.
- 2. Letter from T. Pickens (BWROG) to G. C. Lainas (NRC),
Amendment 17 to General Electric Licensing Topical Report, NEDE-24011-P-A, August 15, 1986.
- 3. NUREG-0979, Section 4.2.1.3.2, April 1983.
- 4. NUREG-0800, Section 15.0.1.
- 5. 10 CFR 50.67.
- 6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors," December 1978.
- 7. ASME, Boiler and Pressure Vessel Code.
- 8. NEDO-21231, "Banked Position Withdrawal Sequence,"
January 1977.
- 9. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 10. NEDO 33091-A , Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.
- 11. XN-NF 1Q(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, " (as identified in the COLR).Not used.
- 12. EMF-215S(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- B2," (as identified in the COLR).Not used.
- 13. ANP-10333P-A Revision O "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident {CRDA), " Framatome Inc., March 2018.
- 14. Regulatory Guide 1.236, "Pressurized -Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents,"
June 2020 (NRC ADAMS ML20055F490).
- 15. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 (NRC ADAMS ML003716792).
- 16. Draft Regulatory Guide 1327, "Pressurized -Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents," July 2019 {NRC ADAMS ML18302A106).
BFN-UNIT 1 B 3.1-46 Revision 0, 29, 68 October 18, 2012
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.11 REQUIREMENTS (continued) SR 3.1.7.11 verifies that each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manual, power operated, and automatic valves in the SLC System Flowpath provides assurance that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing or securing. This verification of valve alignment does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensures correct valve positions. REFERENCES 1. 10 CFR 50.62.
- 2. NEDC- 33860P, "Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate,"
GE Hitachi Nuclear Energy, October 2016. Section 2.8.
- 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 4. FSAR, Section 14.6.
BFN-UNIT 1 B 3.1-57 Revision 0, 29, 112 October 24, 2018
APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) BASES BACKGROUND The APLHGR is a measure of the average LHGR 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 abnormal operational transients and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in References 2 and 11. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), abnormal operational transients, and normal operation that determine the APLHGR limits are presented in References 11, 12, 13, 14, 15, and 16, 17, 18, 19, and 20. for AREVA fuel. (continued) BFN-UNIT 1 B 3.2-1 Revision 0, 40, 68, 104 October 20, 2016
APLHGR B 3.2.1 BASES APPLICABLE (Deleted by Tech Spec Bases Revision 104) SAFETY ANALYSES (continued) AREVA Fuel For AREVA fuel, the APLHGR limits are developed as a function of exposure and, along with the LHGR limits, ensure adherence to fuel design limits during abnormal operational transients. No power- or flow-dependent corrections are applied to the APLHGR (referred to as the maximum APLHGR or MAPLHGR). AREVA APLHGR limits are intended to be bound by the LHGR limits. The calculational procedure used to establish the AREVA fuel MAPLHGR limits is based on LOCA analyses as defined in 10 CFR 50.46, Appendix K. MAPLHGR limits are created to assure that the peak cladding temperature of AREVA fuel following a postulated design basis LOCA will not exceed the PCT and maximum oxidation limits specified in 10 CFR 50.46, Appendix K. The calculational models and methodology are described in References 11, and 12, and 19. The AREVA fuel MAPLHGR limits for two-loop operation are specified in the COLR. For single-loop operation, a MAPLHGR multiplier is applied to the MAPLHGR limit (Ref. 11). The multiplier is documented in the COLR. The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 6). (continued) BFN-UNIT 1 B 3.2-3 Revision 40, 68, 104 Amendment No. 236 October 20, 2016
APLHGR B 3.2.1 BASES LCO The APLHGR limits specified in the COLR are the result of the fuel design, DBA, and transient analyses. With only one recirculation loop in operation, in conformance with the re quirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent limit by an APLHGR correction factor (Ref. 5 and Ref. 10). Cycle specific APLHGR correction factors for single recirculation loop operation are documented in the COLR. APLHGR limits are selected such that no power or flow dependent corrections are required. Additional APLHGR operating limit adjustments may be provided in the COLR supporting other analyzed equipment out-of-service conditions. APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 4) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels 23% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required. (continued) BFN-UNIT 1 B 3.2-3a Revision 40, 68, 104, 112 Amendment No. 236 October 24, 2018
APLHGR B 3.2.1 BASES REFERENCES 1. (Deleted by Tech Spec Revision 104).
- 2. FSAR, Chapter 3.
- 3. FSAR, Chapter 14.
- 4. FSAR, Appendix N.
- 5. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 2, December 1997.Not used.
- 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 7. NEDC- 32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Units 1, 2, and 3," April 1995.Not used.
- 8. NEDO-30130-A, "Steady State Nuclear Methods,"
May 1985.Not used.
- 9. NEDO-24154, "Qualification of the One -Dimensional Core Transient Model for Boiling Water Reactors," October 1978.Not used.
- 10. NEDO-24236, "Brmvns Ferry Nuclear Plant Units 1, 2, and 3, Single- Loop Operation," May 1981.Not used.
(continued) BFN-UNIT 1 B 3.2-5 Revision 0, 40, 68, 104 October 20, 2016
APLHGR B 3.2.1 BASES REFERENCES 11. EMF-2361 (P)(A), Rev. 0, "EXEM BWR-2000 ECCS (continued) Evaluation Model," Framatome ANP Inc., as supplemented by the site-specific approval in NRC safety evaluation dated April 27, 2012, and July 31, 2014.
- 12. EMF-2292(P)(A), "ATRIUM TM-10: Appendix K Spray Heat Transfer Coefficients," (as identified in the COLR).
- 13. XN-NF 58(P)(A), "RODEX2 Fuel Rod Thermal -
Mechanical Response Evaluation Model," (as identified in the COLR).
- 14. XN-NF 19(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).
- 15. XN-NF 19(P)(A), Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," (as identified in the COLR).
- 16. BAW- 10247PA Revision 0, "Realistic Thermal -Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.
- 17. ANF 98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs,"
Advanced Nuclear Fuels Corporation, May 1995.
- 18. BAW-10247P-A Supplement 2P-A Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 2: Mechanical Methods, "
Framatome Inc., August 2018.
- 19. ANP-10332P-A Revision 0, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios, " Framatome Inc., March 2019.
- 20. ANP-10340P-A Revision 0, "Incorporation of Chromia -
Doped Fuel Properties in AREVA Approved Methods," Framatome Inc., May 2018. (continued) BFN-UNIT 1 B 3.2-5a Revision 0, 40, 68, 104 October 20, 2016
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 MCPR 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 abnormal operational transients and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated. Although fuel damage does not necessarily occur if a fuel rod actually experienced 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. (continued) BFN-UNIT 1 B 3.2-6 Revision 0
MCPR B 3.2.2 BASES (continued) APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the abnormal operational transients to establish the operating limit MCPR are presented in References 11, 12, 13, 14, and 15and 16. 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 (~ CPR). When the largest CPR is added tocombined with the MCPR SLSLMCPR99.9%, the required operating limit MCPR is obtained. The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Ref. 8). Flow dependent MCPR (MCPR f) limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref. 12) and the multichannel thermal hydraulics code (Ref. 13). The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System. 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 MCPR99.9% calculation are given in Reference 11. The cycle-specific Reload Analysis report (as referenced in the COLR) includes tabulation of fuel- and plant-related uncertainties for the parameters used in the MCPR99.9% statistical analysis. Power dependent MCPR limits (MCPR p) are derived from the MCPR99.9% value and the transient analyses based on determined by the three-dimensional BWR simulator code (Ref. 12) and the one- dimensional transient codes (Refs. 14 and 15)multi-physics evaluation model (Ref. 16). 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 scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 23 % RTP and the previously mentioned bypass power level. MCPRp limits above the bypass power level may be implemented with a flow dependency as a margin improvement option. At a given power level, MCPRp limits can be defined for specific ranges of core flow. The use of this approach expands the original basis for thermal limit administration under the ARTS methodology (Ref. 8), and can improve operating MCPR margin when operating at the lower core flows associated with the MELLLA+ operating domain. The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 7). (continued) BFN-UNIT 1 B 3.2-7 Revision 40, 68, 104, 112 Amendment No. 236 October 24, 2018
MCPR B 3.2.2 BASES (continued) LCO The MCPR operating limits specified in the COLR (MCPR99.9% value, MCPRf values, and MCPRp values) are the result of the Design Basis Accident (DBA) and transient analysis. Additionally MCPR operating limits supporting analyzed equipment out-of-service conditions are provided in the COLR. The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits, which are based on the MCPR99.9% limit specified in the COLR. APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 23% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 23% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 23% RTP is > 3.5. Studies of the variation of limiting transient behavior have 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 23% 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 provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels < 23% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required. ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient (continued) BFN-UNIT 1 B 3.2-8 Revision 0, 40, 68, 112 October 24, 2018
MCPR B 3.2.2 BASES REFERENCES 8. NEDC- 32433P, "Maximum Extended Load Line Limit and (continued) ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Units 1, 2, and 3," April 1995.
- 9. (Deleted by Tech Spec Revision 104).
- 10. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single- Loop Operation," May 1981.
- 11. ANP- 10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.
- 12. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- 82," (as identified in the COLR).
- 13. XN-NF 19(P)(A) Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," (as identified in the COLR).
- 14. ANF-Q13(P)(A) Volume 1, "COTRANSA2: A Computer Program for Boiling Water Reactor Tran sient Analyses," (as identified in the COLR).Not used.
- 15. XN-NF 105(P)(A) Volume 1, "XCOBRA-T: A Computer Code for BWR Transient Thermal- l=lydraulic Core Analysis,"
(as identified in the COLR).Not used.
- 16. ANP-10300P-A Revision 1 "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios, " Framatome Inc., January 2018.
BFN-UNIT 1 B 3.2-10a Revision 40, 68, 104 October 20, 2016
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and The OPRM Upscale Function provides compliance with APPLICABILITY 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 and (continued) 12, thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations. Reference 13 describes the Detect and Suppress - Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA).Reference 13 describes the Detect and Suppress Confirmation Density (DSS-CD) long-term stability solution. Reference 13 also defines the Confirmation Density Algorithm (CDA) which is used in conjunction with DSS-CD for oscillation detection. The analytical portion of DSS-CD has since been replaced by the Best-estimate Enhanced Option III methodology with CDA as described in Reference 15. The hardware side of DSS-CD has been retained and is unchanged from Reference 13. This hybrid stability solution is referred to as BEO-III with CDA. Reference 13 also describes the DSS-CDCDA Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY is based only on the CDA. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms. DSS-CDCDA OPERABILITY requires at least 8 responsive OPRM cells per channel. The DSS-CDCDA software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary. (continued) BFN-UNIT 1 B 3.3-16a Revision 45, 112, 119 January 8, 2020
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued) LCO, and APPLICABILITY The OPRM Upscale Function is required to be OPERABLE above a power level set at 5% of rated power below the lower boundary of the Armed Region, which is >18% RTP. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is ~23% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (f) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation. If any OPRM auto-enabled setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP
~ 23% and reactor recirculation drive flows 75% of rated , the associated channel is considered inoperable for the OPRM Upscale function. Alternatively, the auto-enabled setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE.
Note (f) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM upscale function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary. (continued) BFN-UNIT 1 B 3.3-16b Revision 45, 119 January 8, 2020
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, (continued) LCO, and APPLICABILITY An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel. Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded. There is no Allowable Value for this function. The OPRM Upscale Function is not LSSS SL-related (Reference 13) and Reference 14 confirms that the OPRM Upscale Function settings based on DSS-CDBEO-III with CDA also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology. (continued) BFN-UNIT 1 B 3.3-16c Revision 45, 119 January 8, 2020
RPS Instrumentation B 3.3.1.1 BASES (continued) REFERENCES 1. FSAR, Section 7.2.
- 2. FSAR, Chapter 14.
- 3. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 4. FSAR, Appendix N.
- 5. FSAR, Section 14.6.2.
- 6. FSAR, Section 6.5.
- 7. FSAR, Section 14.5.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
- 9. NEDC-30851-P-A , "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
- 10. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 11. MED-32-0286, "Technical Specification Improvement Analysis for Browns Ferry Nuclear Plant, Unit 2," October 1995.
- 12. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function," October 1995.
- 13. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor Detect and Suppress Solution Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.
- 14. GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution Confirmation Density (DCC-CD), Analytical Limit (TAC No. MD0277)," October 29, 2008, (ADAMS Accession No. ML083040052).
15.ANP-3907P R evision 0, "Application of BEO -III Methodology with the Confirmation Density Algorithm at Browns Ferry," Framatome Inc .. $SULO. BFN-UNIT 1 B 3.3-44 Revision 0, 40, 45, 119 January 8, 2020
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer SAFETY ANALYSES, LCO, and The RWM enforces the banked position withdrawal sequence APPLICABILITY (BPWS) to ensure that the initial conditions of the CRDA (continued) analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, 7, 12, 13, and 1415. The standard BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattern Control." When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 12) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions. When using the Reference 12 control rod insertion sequence for shutdown, the Rod Worth Minimizer may be programmed to enforce the requirements of the improved BPWS control rod insertion process or bypassed if it is not programmed to reflect the optional BPWS shutdown sequence, as permitted by the Applicability Note for the Rod Worth Minimizer in Table 3.3.2-1. The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10). Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed. (continued) BFN-UNIT 1 B 3.3-61 Revision 0, 68 Amendment 275 October 18, 2012
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer (continued) SAFETY ANALYSES, LCO, and Compliance with the BPWS, and therefore OPERABILITY of the APPLICABILITY RWM, is required in MODES 1 and 2 when THERMAL (continued) POWER is 10% RTP. When THERMAL POWER is
> 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 5 and 7).the power pulse in the power range is much broader than that in the startup range due to the presence of voids in the core. For this reason, CRDA in the power range remains non-limiting (Ref.
15). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
- 3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch -
Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis. The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10). Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position. (continued) BFN-UNIT 1 B 3.3-62 Revision 0
Control Rod Block Instrumentation B 3.3.2.1 BASES (continued) REFERENCES 1. FSAR, Section 7.5.8.2.3.
- 2. FSAR, Section 7.16.5.3.1.k.
- 3. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2 and 3," April 1995.
- 4. NEDE-24011-P-A-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).Not used.
- 5. "Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.
- 6. NEDO-21231, "Banked Position Withdrawal Sequence,"
January 1977.
- 7. NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"
December 27, 1987.Not used.
- 8. NEDC-30851-P-A, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.
- 9. GENE-770-06-1, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.
- 10. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 11. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option Ill Stability Trip Function," October 1995. (continued) BFN-UNIT 1 B 3.3-71 Revision 0, 40, 68 Amendment 276 October 18, 2012
Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 12. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod (continued) Insertion Process," July 2004.
- 13. XN-NF-80-19(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).
- 14. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- 82," (as identified in the COLR).
- 15. ANP-10333P-A Revision O "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," Framatome Inc., March 2018 (continued)
BFN-UNIT 1 B 3.3-71a Revision 0, 40, 68 Amendment 276 October 18, 2012
Recirculation Loops Operating B 3.4.1 BASES (continued) APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. 1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered (Ref. 1). The LOCA analysis accounts for the loop flow mismatch allowed by the LCO (Refs. 9 and 11). The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 14 of the FSAR. (continued) BFN-UNIT 1 B 3.4-3 Revision 0, 87 October 29, 2014
Recirculation Loops Operating B 3.4.1 BASES APPLICABLE Plant specific LOCA analyses have been performed assuming SAFETY ANALYSES only one operating recirculation loop. These analyses have (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Refs. 7 9 and 811). The Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain has not been analyzed for single recirculation loop operation. Therefore, single loop operation is prohibited in the MELLLA+ operating domain (Ref. 10). The transient analyses of Chapter 14 of the FSAR have also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoint is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 6). (continued) BFN-UNIT 1 B 3.4-4 Revision 45, 50, 119 Amendment No. 236 January 8, 2020
Recirculation Loops Operating B 3.4.1 BASES (continued) LCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLHGR Limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, 'MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 9 and 811. Single loop operation is prohibited in the MELLLA+ operating domain. This prohibition is consistent with Reference 10. APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. (continued) BFN-UNIT 1 B 3.4-5 Revision 45, 119 Amendment No. 236 January 8, 2020
Recirculation Loops Operating B 3.4.1 BASES (continued) REFERENCES 1. FSAR, Section 14.6.3.
- 2. FSAR, Section 4.3.5.
- 3. Deleted.
- 4. Deleted.
- 5. Deleted.
- 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 7. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single- Loop Operation," May 1981.
- 8. NEDC- 32484P, "Bro1A'ns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Not_
Analysis," Re11ision 6, February 2005. _ used.
- 9. ANP-3015(P), "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis," Revision 0, September 2011.ANP-3546P Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+), " AREVA Inc., March 2017.
- 10. NEDC-33006P-A, "General Electrical Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," Revision 3, June 2009.
- 11. ANP-3905P Revision 1 , "Browns Ferry Units 1, 2, and 3 LOCA Analysis for ATRIUM 11 Fuel," Framatome Inc.,
July 2021. BFN-UNIT 1 B 3.4-10 Revision 45, 87, 119 Amendment No. 236 January 8, 2020
ECCS - Operating B 3.5.1 BASES BACKGROUND RCS is still pressurized. If HPCI fails, it is backed up by ADS in (continued) combination with LPCI and CS. In this event, either the vessel would be manually depressurized or the ADS timed sequence would be allowed to time out and open the selected safety/relief valves (S/RVs) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core. Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool may be circulated through a heat exchanger cooled by the RHR Service Water System. Depending on the location and size of the break, portions of the ECCS may be ineffective; however, the overall design is effective in cooling the core regardless of the size or location of the piping break. All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipment. The CS System (Ref. 1) is composed of two independent subsystems. Each subsystem consists of two 50% capacity motor driven pumps, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The LOCA analysis (Refs. 13 and 16) requires both pumps in a subsystem (loop) to be OPERABLE for the subsystem to be OPERABLE. Failure of one CS pump results in the loss of the associated CS loop for LOCA mitigation. The CS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal, the CS pumps in both subsystems are automatically started (A pump (continued) BFN-UNIT 1 B 3.5-2 Revision 0
ECCS - Operating B 3.5.1 BASES APPLICABLE c. Maximum hydrogen generation from a zirconium water SAFETY ANALYSES reaction is 0.01 times the hypothetical amount that would (continued) be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;
- d. The core is maintained in a coolable geometry; and
- e. Adequate long term cooling capability is maintained.
The limiting single failures are discussed in References 13 and
- 16. For a large or small pipe break LOCA and events requiring ADS operation, selected battery failure is considered the most severe single failure. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage.
The ECCS satisfy Criterion 3 of the NRC Policy Statement (Ref. 14). LCO Each ECCS injection/spray subsystem and six ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System. The low pressure ECCS injection/spray subsystems are defined as the two CS subsystems and the two LPCI subsystems. With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 9 could be exceeded. All ECCS subsystems and ADS must therefore be OPERABLE to satisfy the single failure criterion required by Reference 9. (continued) BFN-UNIT 1 B 3.5-7 Revision 0
ECCS - Operating B 3.5.1 BASES ACTIONS E.1 (continued) The LCO requires six ADS valves to be OPERABLE in order to provide the ADS function. References 13 and 16 contains the results of an analysis that evaluated the effect of one ADS valve being out of service. Per this analysis, operation of only five ADS valves will provide the required depressurization. However, overall reliability of the ADS is reduced, because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day Completion Time is based on a reliability study cited in Reference 11 and has been found to be acceptable through operating experience. F.1 and F.2 If any one low pressure ECCS injection/spray subsystem, or one LPCI pump in both LPCI subsystems, is inoperable in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCI and the remaining low pressure ECCS injection/spray subsystem. However, overall ECCS reliability is reduced because a single active component failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Since both a high pressure system (ADS) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours is required to restore either the low pressure ECCS subsystem or the ADS valve to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 11 and has been found to be acceptable through operating experience. (continued) BFN-UNIT 1 B 3.5-11 Amendment No. 240 Revision 0 March 12, 2001
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 REQUIREMENTS (continued) The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (in accordance with the ASME OM Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of References 13, and 15, and 16. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established by testing or analysis or during preoperational testing. The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be 950 psig to perform SR 3.5.1.7 and 150 psig to perform SR 3.5.1.8. Adequate steam flow is represented by at least two turbine bypass valves full open for SR 3.5.1.7 and SR 3.5.1.8. Therefore, sufficient time is allowed after adequate (continued) BFN-UNIT 1 B 3.5-16 Revision 0, 50, 53, 81 October 16, 2013
ECCS - Operating B 3.5.1 BASES (continued) REFERENCES 1. FSAR, Section 6.4.3.
- 2. FSAR, Section 6.4.4.
- 3. FSAR, Section 6.4.1.
- 4. FSAR, Section 6.4.2.
- 5. FSAR, Section 14.6.3.
- 6. FSAR, Section 14.6.5.
- 7. 10 CFR 50, Appendix K.
- 8. FSAR, Section 6.5.3.
- 9. 10 CFR 50.46.
- 10. TVA BFN System Design Criteria BFN-50-7032, Control Air System - Units 1, 2, and 3.
- 11. Memorandum from R. L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.
- 12. 10 CFR 50, Appendix A, GDC 34, GDC 35, GDC 36, and GDC 37.
- 13. ANP-3377P Revision 3, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU), August 2015.ANP-3546P Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+), " AREVA Inc.,
March 2017.
- 14. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 15. GE-NE-B13-01755-2, "Relaxation of ECCS Parameters, Revision 2, December 1996.
- 16. ANP-3905P Revision 1, "Browns Ferry Units 1, 2, and 3 LOCA Analysis for ATRIUM 11 Fuel, " Framatome Inc.,
July 2021. BFN-UNIT 1 B 3.5-22 Revision 0, 112 October 24, 2018
ECCS - Shutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the Core Spray (CS) System and the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCS - Operating." For LCO 3.5.2, only one pump is required for an OPERABLE subsystem, as stated in the LCO Bases below. APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes in the long term cooling analysis (Refs. 1 and 3) for a postulated loss of coolant accident (LOCA). The ECCS requirements are significantly reduced during shutdown since a LOCA is not postulated to occur. However, some ECCS capability may be required to restore and maintain the reactor coolant level in the event of an inadvertent draindown. It is reasonable to assume, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. To provide redundancy, a minimum of two low pressure ECCS injection/spray subsystems are required to be OPERABLE in MODES 4 and 5. The low pressure ECCS subsystems satisfy Criterion 3 of the NRC Policy Statement (Ref. 2). (continued) BFN-UNIT 1 B 3.5-23 Revision 0
ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.3 (continued) REQUIREMENTS In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, RHR valves that are required for LPCI subsystem operation may be aligned for decay heat removal. Therefore, this SR is modified by a Note that allows one LPCI subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned (remote or local) to allow injection into the RPV, and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent RPV draindown should occur. REFERENCES 1. ANP-3377P Revision 3, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU), August 2015.ANP-3546P Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+), " AREVA Inc., March 2017.
- 2. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 3. ANP-3905P Revision 1, "Browns Ferry Units 1, 2, and 3 LOCA Analysis for ATRIUM 11 Fuel, " Framatome Inc., July 2021.
BFN-UNIT 1 B 3.5-29 Revision 0, 112 October 24, 2018
Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.2 (continued) REQUIREMENTS Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the pressure in either the suppression chamber or the drywell does not change by more than 0.25 inch of water per minute over a 10 minute period. The leakage test is performed every 24 months. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage and also in view of the fact that component failures that might have affected this test are identified by other primary containment SRs. REFERENCES 1. NEDC-33960P33860P
- - -_ _ _ , "Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate,"
GE Hitachi Nuclear Energy, October 2016. Section 2.6.
- 2. FSAR, Section 14.6.
- 3. 10 CFR 50, Appendix J, Option B.
- 4. NEI 94-01, Revision 3A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."
- 5. ANSI/ANS-56.8-2002, "American National Standard for Containment System Leakage Testing Requirements."
- 6. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
BFN-UNIT 1 B 3.6-6 Revision 0, 43, 112, 114 March 27, 2019
CAD System B 3.6.3.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.1 Containment Atmosphere Dilution (CAD) System BASES BACKGROUND The CAD System functions to maintain combustible gas concentrations within the primary containment at or below the flammability limits following a postulated loss of coolant accident (LOCA) by diluting hydrogen and oxygen with nitrogen. To ensure that a combustible gas mixture does not occur, oxygen concentration is kept < 5.0 volume percent (v/o), or hydrogen concentration is kept < 4.0 v/o. The CAD System is manually initiated and consists of two independent, 100% capacity subsystems, each of which is capable of supplying nitrogen through separate piping systems to the drywell and suppression chamber of each unit. Each subsystem includes a liquid nitrogen supply tank, ambient vaporizer, electric heater (unqualified), and a manifold with branches to each primary containment (for Units 1, 2, and 3). The nitrogen storage tanks each contain 2615 gal, which is adequate for 7 days of CAD subsystem operation (Refs. 4 and 5). The CAD System operates in conjunction with emergency operating procedures that are used to reduce primary containment pressure periodically during CAD System operation. This combination results in a feed and bleed approach to maintaining hydrogen and oxygen concentrations below combustible levels. (continued) BFN-UNIT 1 B 3.6-90 Revision 0, 112, 117 August 16, 2019
CAD System B 3.6.3.1 BASES (continued) REFERENCES 1. AEC Safety Guide 7, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, March 10, 1971.
- 2. FSAR, Section 5.2.6.
- 3. NRC No. 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 4. ANP-3403P_ Revision
___ 2 , "Fuel Uprate Safety Analysis Report for Browns Ferry - Nuclear
- -Plant AREVA
- -Units 1, 2, and 3," _ __
Inc., August 2015. Section 2.6.4.
- 5. FS1-0055816 Revision 1.0, "Evaluation of ATRIUM 11 Fuel and the Containment Atmospheric Dilution System,"
Framatome Inc., April 1, 2021. BFN-UNIT 1 B 3.6-96 Revision 0, 112 October 24, 2018
Control Rod Testing - Operating B 3.10.7 B 3.10 SPECIAL OPERATIONS B 3.10.7 Control Rod Testing - Operating BASES BACKGROUND The purpose of this Special Operations LCO is to permit control rod testing, while in MODES 1 and 2, by imposing certain administrative controls. Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), such that only the specified control rod sequences and relative positions required by LCO 3.1.6, "Rod Pattern Control," are allowed over the operating range from all control rods inserted to the low power setpoint (LPSP) of the RWM. The sequences effectively limit the potential amount and rate of reactivity increase that could occur during a control rod drop accident (CRDA). During these conditions, control rod testing is sometimes required that may result in control rod patterns not in compliance with the prescribed sequences of LCO 3.1.6. These tests include SDM demonstrations, control rod scram time testing, and control rod friction testing. This Special Operations LCO provides the necessary exemption to the requirements of LCO 3.1.6 and provides additional administrative controls to allow the deviations in such tests from the prescribed sequences in LCO 3.1.6. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, 2, 3, and 4 Reference 5. CRDA analyses assume the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analyses. The RWM provides backup to operator control of the withdrawal sequences to ensure the initial conditions of the CRDA analyses are not (continued) BFN-UNIT 1 B 3.10-36 Revision 0, 68 October 18, 2012
Control Rod Testing - Operating B 3.10.7 BASES APPLICABLE violated. For special sequences developed for control rod SAFETY ANALYSES testing, the initial control rod patterns assumed in the safety (continued) analysis of References 1, 2, 3, and 4Reference 5 may not be preserved. Therefore, special CRDA analyses may be required to demonstrate that these special sequences will not result in unacceptable consequences, should a CRDA occur during the testing. These analyses, performed in accordance with an NRC approved methodology, are dependent on the specific test being performed. As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LCO 3.1.6, and during these tests, no exceptions to the requirements of LCO 3.1.6 are necessary. For testing performed with a sequence not in compliance with LCO 3.1.6, the requirements of LCO 3.1.6 may be suspended, provided additional administrative controls are placed on the test to ensure that the assumptions of the special safety analysis for the test sequence are satisfied. Assurances that the test sequence is followed can be provided by either programming the test sequence into the RWM, with conformance verified as specified in SR 3.3.2.1.8 and allowing the RWM to monitor (continued) BFN-UNIT 1 B 3.10-37 Revision 0, 68 October 18, 2012
Control Rod Testing - Operating B 3.10.7 BASES LCO control rod withdrawal and provide appropriate control rod (continued) blocks if necessary, or by verifying conformance to the approved test sequence by a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test). These controls are consistent with those normally applied to operation in the startup range as defined in the SRs and ACTIONS of LCO 3.3.2.1, "Control Rod Block Instrumentation." APPLICABILITY Control rod testing, while in MODES 1 and 2, with THERMAL POWER greater than 10% RTP, is adequately controlled by the existing LCOs on power distribution limits and control rod block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.3.2.1. With THERMAL POWER less than or equal to 10% RTP, the provisions of this Special Operations LCO are necessary to perform special tests that are not in conformance with the prescribed sequences of LCO 3.1.6. While in MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LCO 3.10.3, "Single Control Rod Withdrawal - Hot Shutdown," or Special Operations LCO 3.10.4, "Single Control Rod Withdrawal - Cold Shutdown," which provide adequate controls to ensure that the
---_ I assumptions of the safety analyses of References 1, 2, 3, and 4 Reference 5 are (continued)
BFN-UNIT 1 B 3.10-38 Revision 0, 68 October 18, 2012
Control Rod Testing - Operating B 3.10.7 BASES SURVEILLANCE SR 3.10.7.2 REQUIREMENTS (continued) When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly loaded into the RWM prior to control rod movement. This Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has been added to indicate that this Surveillance does not need to be performed if SR 3.10.7.1 is satisfied. REFERENCES 1. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," October 2007.Not used.
- 2. Letter from T. Pickens (BWROG) to G. C. Lainas (NRC)
"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.Not used.
- 3. XN-NF 1Q(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified by the COLR).Not used.
- 4. EMF-2158(P)~I\), "Siemens Po1A1er Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- B2," (as identified in the COLR).Not used.
- 5. ANP-10333P-A Revision 0 "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA), " Framatome Inc., March 2018.
BFN-UNIT 1 B 3.10-40 Revision 0, 68 October 18, 2012
SDM Test -Refueling B 3.10.8 BASES (continued) APPLICABLE Prevention and mitigation of unacceptable reactivity excursions SAFETY ANALYSES during control rod withdrawal, with the reactor mode switch in the startup/hot standby position while in MODE 5, is provided by the intermediate range monitor (IRM) neutron flux scram (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), and control rod block instrumentation (LCO 3.3.2.1, "Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA). CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. For SDM tests performed within these defined sequences, the analyses of References 1, 2, 3, and 4Reference 5 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of References 1, 2, 3, and 4Reference 5 may not be met. Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, may be required to demonstrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Refs. 1, 2, 3, and 4Ref. 5). In addition to the added requirements for the RWM, APRM, and control rod coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test. (continued) BFN-UNIT 1 B 3.10-42 Revision 0, 68 October 18, 2012
SDM Test -Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.5 REQUIREMENTS (continued) Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events. SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water pressure ensures that if a scram is required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig, which is well below the expected pressure of approximately 1100 psig, ensures sufficient pressure for rapid control rod insertion. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room. REFERENCES 1. NEDE-24011-P-A, Rev. 16, "General Electric Standard Application for Reactor Fuel," October 2007.Not used.
- 2. Letter from T. Pickens (BWROG) to G. C. Lainas, NRC, "Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.Not used.
(continued) BFN-UNIT 1 B 3.10-48 Revision 0, 68 October 18, 2012
SDM Test -Refueling B 3.10.8 BASES REFERENCES 3. XN-NF 1Q(P)(A), Volume 1 "Exxon Nuclear Methodology (continued) for Boiling Water Reactors - Neutronic Methods for Design Not_ and Analysis," (as identified in the COLR). _ used.
- 4. EMF-2158(P)(A), "Siemens Power Corp oration Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN- B2," (as identified in the COLR).Not used.
- 5. ANP-10333P-A Revision O "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," Framatome Inc., March 2018.
BFN-UNIT 1 B 3.10-48a Revision 0, 68 October 18, 2012
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Framatome Inc. ANP-3857NP Revision 2 Design Lim its for Framatome Critical Power Correlations Pa e 4-1 4.0 DESIGN LIMITS The design limits for Framatome fuel are calculated in Table 1 using Equation (4). Table 1 MCPR9s19s Limits for Framatome Fuel SPCB ACE ACE AC E Correlation: ATRIUM 10 ATRIUM 10 ATRIUM 10XM ATRIUM 11 Reference 6 7 8 9 Number of Data Mean ECPR Std. Dev. 1-D_._ B ._O_w_e_n --' kga= s,= gs_
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t The design limit was rounded upward to two digits.
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