ML20113B197
| ML20113B197 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/06/1996 |
| From: | Bruce A, Keys T, Riley E TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20113B188 | List: |
| References | |
| NUDOCS 9606260242 | |
| Download: ML20113B197 (40) | |
Text
{{#Wiki_filter:. .- -. . . -- . . . I M wx m y L32 960531 8 01 wx.cou me, Cort Opencng Limas Report Revuion 1. Page ! l QA RECORC . Browns Ferry Nuclear Plant
- Unit 2, Cycle 9 4
4 l CORE OPERATING LIMITS REPORT (COLR) 4 m TENNESSEE VALLEY AUTHORITY Nuclear Fuel Division
- BWR Fuel Engineering Department
) 1 Prepared By: _ </___ Dau: f/31 d Alan L. Bruce, Nucleu Engineer BWR Fuel Engineering j k Verified By: . Earl E. Riley, Engine (ring Specialist Date: MJMM BWR Fuel Engineering Approved By: ___ Date:
- Manager
! T. BWR A.Fuel Key [En3 i neering Reviewed By; Date: [E # eacto'r Engineering Supervisor Reviewed By: fd __ Date: b
- POR irman i
9606260242 960621 PDR ADOCK 05000259 P PDR s
....7.-
TVA Nuclear Fuel Core Operating Linuts Report TVA-COLR-BF2C9 , Revision 1 Page 2 L i i I l l l Revision Log l Revision Rgg Descriotion Affected Panes i 0 3/14/% Initial Release All 1 6/6/% Revise OLMCPRs due to error i$1 6,15,16,17 GE generic Safety Limit MCPR analysis (BFPER960502) l l l l l l I l I 1
l TVA Nuclear Fuel Core Operatmg Limiu Report TVA COLR.BF2C9 Revisbn 1, Page 3
- 1. INTRODUCTION This Core Operating Limits Report for Browns Ferry Unit 2, Cycle 9 is prepared in accordance with the requirernents of Browns Ferry Technical Specification 6.9.1.7.
i The core operating limits presented here were developed using NRC-approved j methods (References 1 and 2). Results from the reload analyses for Browns Ferry Unit 2, Cycle 9 are documented in Reference 3. The following core operating limas are included in this report: 3
- a. Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specification 3.5.I) i
- b. Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.5.J)
! c. Minimum Critical Power Ratio Operating Limit (OLMCPR) (Technical Specification 3.5.K/4.5.K)
- d. Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Specification 2.1. A.1.c, Table 3.2.C, and Specification 3.5.L)
- e. Rod Block Monitor (RBM) Upscale (Flow Bias) Trip Setting and Clipped Value (Technical Specification Table 3.2.C)
- 2. APLHGR LIMIT (TECHNICAL SPECIFICATION 3.5.I)
The APLHGR limit for each type of fuel as a function of exposure is shown in Tables ! - 8. The APLHGR limits for the GE9B (GE8X8NB) and the Gell bundles are for the most limiting lattice at each exposure point. The specific values for each
. lattice are given in Reference 4.
._ .., . - . . _ _ - . . _ . _ _ _ . _ = _ _ -_ TVA Nuclear Fuel Core Operating Limits Repon TVA-COLR BF2C9 Revision 1. Page 4 , I
- 3. LHGR LIMIT (TECHNICAL SPECIFICATION 3.5.J)
The LHGR limit for unit 2 cycle 9 is fuel type dependent, as shown below: Fuel Tvoe LHGR Limit P8X8R/BP8X8R 13.4 kw/ft GE8X8NB/GE11 14.4 kw/ft .
- 4. OLMCPR (TECHNICAL SPECIFICATION 3.5.K/4.5.K)
- a. The OLMCPR is equal to the fuel type and exposure dependent MCPR limit at rated flow and rated power shown in Figures 1 - 3 multiplied by the Kr l shown in Figure 4, where; i T ~ Ts T = 0.0 or T* , whicheveris greater.
l D - To 1 L = 0.90 sec (Specification 3.3.C.1 scram time limit to 20% insertion from fully withdrawn ) 1 T, = 0.710 + 1.65 (0.053) n. n 1E T.e = ' - i n where; n = Number of surveillance rod tests perfon... i to date in cycle (including BOC test). I = Scram time to 20% insertion from fully withdrawn of the iO rod. N = Total number of active rods measured in Specification 4.3.C. I at BOC.
- b. For the performance of Surveillance Requirement 4.5.K.2.a (prior to initial scram time measurements for the cycle),
T = 1.0
_ ~ TVA Nuclear Fuel Cote Operating Limits Report TVA-COLR BF2C9 Revision 1. Page 5
- c. For the performance of Surveillance Requirement 4.5.K.2.b T shall be determined in accordance with 4.a above.
S. APRM FLOW BIASED ROD BLOCK TRIP SETTING (TECHNICAL SPECIFICATION 2.1.A.I.c, TA33LE 3.2.C, AND SPECIFICATION 3.5.L) The APRM Rod Block trip setting shall be: SRB s (0.58W + 57%) where: SRB = Rod Block setting in percent of rated thermal power (3293 MWt) W = Loop recirculation flow rate in percent of rated Note: Under certain conditions, the APRM Rod Block setting must be adjusted by FRP/CMFLPD. See Technical Specification 3.5.L.
- 6. RBM UPSCALE (FLOW BIAS) TRIP SETTING AND CLIPPED VALUE (TECIINICAL SPECIFICATION TABLE 3.2.C)
The RBM Upscale trip setting shall be: s (0.66W + 46%) where: Trip level setting is in percent of rated thermal power (3293 MWt)
% = Loop recirculation flow rate in
; 'rcent of rated RBM upscale flow-biased setpoint clipped at 112 percent rated reactor power.
i
_ n. I TVA Nuclear Fuel Core Operating Limits Report TVA-COLR BF2C9 Revision 1, Page 6
- 7. REFERENCES
< 1. NEDE-24011-P-A-11 " General Electric Standard Application for Reactor Fue November 1995.
- 2. NEDE-24011-P-A-11-US, " General Electric Standard Application for Reactor Fuel -- U.S. Supplement," November 1995.
- 3. J11-02761SRLR, Rev. 3 " Supplemental Reload Licensing Report for Browns Ferry Nuclear Plant Unit 2 Reload 8 Cycle 9," May 1996.
- 4. J11-02761 MAPL, Rev. O, " Lattice Dependent MAPLHGR Report for Browns Ferry Nuclear Plant Unit 2 Reload 8 Cycle 9," January 1996.
1 TVA Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C9 4 Revision 1. Page 7 I Table 1 i APLHOR Limits for Bundle Type GE9B-P8DWB319-9GZ 1 l (GE8X8NB) { l Most Limiting 1.attice 3 for Each Exposure Point i 1 Average Planar Exposure APLHGR Limit (GWD/ST) (kw/ft) i 0.00 11.64 I 0.20 11.68 1.00 11.77 2.00 ,11.88 3.00 12.01
- ~
4.00 12.12 l i 5.00 12.22 6.00 l 12.32 i 7.00 j 1 12.49 8.00 12.65 9.00 12.77 10.00 12.86 12.50 12.86 15.00 12.62 ! 20.00 12.06 ! 25.00 11.47-5 35.00 10.41 45.00 8.81 51.20 5.91 t i l
{ TVA Nuclear Fuel Core Operating Limits Report TVA-COLR.BF2C9 Revision 1, Page 8 Table 2 1 APLHGR Limits for Bundle Type GE9B-P8DWB325-10GZ ! (GE8X8NB) l 1 Most Limiting Lattice for Each Exposure Point Average Planar Exposure APLHGR Limit (GWD/ST) ! (kw/ft) l 0.00 10.97 ! 0.20 11.00 3 1.00 11.11 2.00 11.31
- 3.00 11.54 4.00 11.79 l 5.00 11.97 6.00 12.10 7.00 12.23 8.00 12.35 9.00 12.46 10.00 12.58 12.50 12.58 15.00 12.35 20.00 11.86 25.00 11.33 35.00 10.12 45.00 8.60 50.27 5.92 1
i l TVA Nuclear Fuel TVA-COLR-BF2C9 Core Operating Lirmts Report Revision 1. Page 9 Table 3 APLHGR Limits for Bundle Type GE9B-P8DWB326-7GZ l (GE8X8NB) l l Most Limiting lattice < for Each Exposure Point Average Planar Exposure APLHGR Limit (GWD/ST) (kw/ft) l 0.00 11.45 0.20 { 11.47 l 1.00 11.55 2.00 11.70 3.00 11.87 4.00 12.01 5.00 12.13 i 6.00 12.22 , 7.00 12.31 ! 8.00 12.41 l 9.00 12.51 10.00 12.63 12.50 12.64 15.00 12.41 20.00 11.91 25.00 11.35 35.00 10.15 45.00 8.72 50.49 5.93
/
TVA Nuclear Fuel Core Operating Limits Report RA COLR BF2C9 Revision 1, Page 10 Table 4 APLHGR Limits for Bundle Type BP8DRB301L (BP8X8R) - Average Planar Exposure APLHGR Limit (GWD/ST) (kw/ft) 0.00 11.32 0.20 11.33 1.00 11.40 2.00 11.50 3.00 11.61 4.00 11.73 5.00 11.86 6.00 11.93 7.00 11.98 8.00 12.03 9.00 12.11 10.00 12.19 12.50 12.34 15.00 12.47 20.00 12.58 25.00 12.03 35.00 10.82 43.52 9.17
TVA Nuclear Fuel TVA COLR-BF2C9 Core Operatmg Limits Repon Revision 1. Page 11 Table 5 APLHGR Limits for Bundle Type BP8DRB299 (BP8X8R)
. Average Planar Exposure APLHGR Limit l (GWD/ST) (kw/ft) 0.00 10.71 ;
0.20 10.75 1.00 10.86 2.00 11.01 3.00 11.19 4.00 11.35 5.00 11.47 6.00 11.62 7.00 11.78 8.00 11.95 9.00 12.09 l 10.00 12.20 12.50 12.37 15.00 12.47 20.00 12.37 25.00 11.77 35.00 10.65 42.13 9.36
TVA Nuclear Fuel Core Operating Linuts Report TVA-COLR BF2C9 Revision 1. Page 12 Table 6 APLHGR Limits for Bundle Type P8DRB284L (P8X8R) Average Planar Exposure APLHGR Limit (GWD/ST) (kw/ft) 0.20 11.20 1.00 11.30 5.00 11.80 ! 10.00 12.00 15.00
,12.00 20.00 11.80 25.00 11.20 30.00 10.80 35.00 10.20 40.00 9.50 45.00 8.80 I
~ --
TVA Nuclear Fuel l Core Operatmg Limits Repon TVA-COLR.BF2C9 Revision 1. Page 13 i Table 7 APLHGR Limits for Bundle Type GE11-P9 HUB 366-12G4.0 ' l (Gell) l l Most Limiting Lattice for Each Exposure Point l Average Planar Exposure APLHGR Limit (GWD/ST) (kw/ft) 0.00 9.55 0.20 9.62 : 1.00 9.80 2.00 ! 10.05 3.00 10.32 ' 4.00 10.58 5.00 10.84 6.00 11.11 7.00 11.39 8.00 11.67 9.00 11.81 10.00 11.95 12.50 11.92 15.00 11.72 17.50 11.48 20.00 11.20 25.00 10.51 30.00 9.84 35.00 9.19 40.00 8.55 45.00 7.91 50.00 7.26 55.00 6.59 56.17 6.42
TVA Nuclear Fuel TVA-COLR BF2C9 Core Operating Limits Report Revision 1, Page 14 j l Table 8 APLHGR Limits for Bundle Type GE11-P9 HUB 367-14GZ (Gell) l I Most Limiting Lattice for Each Exposure Point I Average Planar Exposure APLHGR Limit (GWD/ST) (kw/ft) 0.00 9.18 l 0.20 9.27 1.00 9.45 2.00 9.72 3.00 10.00 l 4.00 10.30 5.00 10.62 6.00 10.95 , 7.00 11.31 i 8.00 11.67 9.00 11.81 10.00 11.95 12.50 11.92 15.00 11.72 17.50 11.48 20.00 11.20 25.00 10.51 30.00 9.84 35.00 9.19 40.00 8.55 45.00 7.91 50.00 7.26 55.00 6.59 56.17 6.42
TVA Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C9 Revision 1. Page 15 Figure 1 MCPR Operating Limit for P8X8R/BP8X8R 1.34 .
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- 1.28 Use this value at BOC9 prior to performing scram time testing.
TVA Nuclear Fuel TVA-COLR BF2C9 Core Operatmg Limits Repon Revision 1. Page 16 4 Figure 2 . l MCPR Operating Limit for GE8X8NB : 1 1 1 1.38 . 4.3. . _ 1.33 -
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- 1.31 Use this value at BOC9 prior to performing scram time testing.
TVA Nuclear Fuel i Core Operating Limits Report TVA-COLR-BF2C9 l Revision 1. Page 17 ( Figure 3 1 MCPR Operating Limit for Gell 1.37 . 1 l
. 1 1.36 -***?-****'****************. . .
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- 1.33 Use this value at BOC9 prior to performing scram time testing.
t TVA Nuclear Fuel Core Operating Limits Report TVA-COLR BF2C9 ; Revision 1, Page 18 (Final) Figure 4 , GEXL-Plus Kf Curve OLMCPR Flow Corrections
- 1.4 '
I Ncne ! 1.3 - 112 0 % 1 IM 0 % 1025 % D 1.2 - l l l 1.1 - l 1.0 ' ' ' 20 30 40 50 SO 70 80 90 100 Total Core Flow (% rated) For 40% < WT1100%, Kr = MAX [ 1.0, A 0.00441*WT ] WT 1 40%, Kr = [A 0.00441*WT ] * [1.0 + 0.0032*( 40 WT )) WT > 100%, Kr = 1.0 where : WT = Percent of Rated Core Flow, and A = constant which depends on the Flow Control Mode and the Scoop Tube Setpoint as noted below. Flow Control Mode Scoon Tube Setnoint A MANUAL 102.5 % 1.3308 MANUAL 107.0 % 1.3528 MANUAL 112.0 % 1.3793 MANUAL 117.0 % 1.4035 AUTOMATIC N/A 1.4410 Note: Flow Corrections are given for P8X8R, BP8X8R, and GE8X8NB bundles. These corrections are conservative when compared to Gell flow corrections which do not have the additional correcnon factor of [1.0 + 0.0032 * (40 WT)] below 40% flow.
~ - r.
I Safety Assessment No: SA-COLR BF2C9 R1 4 Page 1 of 8 Document No. TVA COLR-BF2C9. Rev 1
)
a ' I !
- S A FETY ASSESSMENT i 1
A1. DESCRIPTION This safety assessment addresses: 1
- I Revision 1 of the Browns Feny Nuclear (BFN) plant unit 2 cycle 9 Core Operating Limits Report (COLR)(ref.1) e Revision of Appendix N of the BFN FSAR to incorporate Revision 3 of the BFN unit 2 cycle 9 Supplemental Reload Licensing Report (SRLR)(ref. 2)
Revision of Appendix H of the BFN FSAR to change the reference given for the Safe Limit MCPR from the GE Licensing Topical Report GESTAR II (ref. 4) to the applica BFN Technical Specifications. These revisions are being made to implement revised Operating Limit Minimum Critical Power Ratio (OLMCPR) values in response to a Problem Evaluation Report (PER) (ref. 3) 1 On 4/16/96 General Electric (GE) notified TVA (ref. 5) that an' error had been d the generic analysis performed to determine the gel 1 fuel type Safety Limit MCPR j (SLMCPR). GE found that the generic analysis does not produce conservative values of j SLMCPR for all plants and core designs. Their corrective action was to reevaluate all operating plants and change their design methodology to include unit and cycle specific ? analyses. BFN Unit 2 was one of the plants for which the current SLMCPR specified in the plant Technical Specifications was determined by GE (ref. 6) to be non-conservative. This in i turn affects the OLMCPRs since the OLMCPRs are determined by adding the delta CPRs i from the cycle specific transient analyses to the SLMCPR. Since current design practice allows for only one value of SLMCPR (the most limiting for all fuel types loaded) for the j core, the OLMCPRs for all fuel types (GE7, GE9, and gel 1) are affected. l 4 GE issued a revision to the current operating cycle SRLR documenting the revised OLMCPRs which are being incorporated into the Unit 2 COLR. Appendix N of the BFN ! FSAR contains the current SRLR for each unit and is being revised to include Revision 3 of the Unit 2 SRLR. ( A Technical Specification change to revise the Safety Limit MCPR is being processed in parallel with the FSAR and COLR revisions being assessed here. 4, i i
- Saftty Assessment No: SA COLR-BF2C9 R1 Page 2 of 8 Document No. TVA COLR BF2C9. Rev 1 A2. REFERENCES
- 1. TVA-COLR-BF2C9, Rev.1, " Browns Ferry Nuclear Plant Unit 2 Cycle 9 Core Operating Limits Report (COLR)."
- 2. GE Document J11-02761SRLR, Rev. 3, " Supplemental Reload Licensing Report for Browns Ferry Nuclear Plant Unit 2 Reload 8 Cycle 9," May 1996.
- 3. Problem Event Report, BFPER960502,4/18/96.
4.
" General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-11, November 1995; and the U.S. Supplement, NEDE-24011-P-A-11-US, November 1995.
- 5. TVA Memorandum BFE-842, T. A. Keys to J.L. Lewis, " Safety Limit MCPR (SLMCPR)," April 16,1996. [L32 960416 804) 6.
GE Letter CJP2:96-139, C.J. Papandrea to T.L. Hayslett, " Safety Limit MCPR Calculation; Verified Results and BF2C9 SRLR Revision 3," May 28,1996. [L38 960529 800]
- 7. BFNP Updated Final Safety Analysis Report, re dsed through Amendment 12.
. 8. BFNP Technical Specifications, Unit 2 (through Amendment No. 244).
! l 1
i Safety Assessment No: SA-COLR BF2C9 R1 Page 3 of 8 Document No. TVA-COLR-BF2C9. Rev. I ! i
- B. SAFETY ASSESSMENT CHECKLIST Completion of checklist required?
4 Yes X No f Checklist: ' 1 POTENTIALLY j i REDUCES i NUCLEAR I
- SAFETY N/A
\
- l. X Fire Protection (Appendix R)
- 2. _X_ Internal Floodmg Protection (MELB)
- 3. X Pipe Breaks
- 4. _X_, Pipe Whip
- 5. X Modification to Non Seismic areas in the CB/AB
- 6. X Jet Impingement Effects
- 7. X Seismic / Dead Weight
- 8. X Internal / External Missiles
- 9. _X_, Heavy Load Lifts or Safe Load Paths t
(NUREG.0612) j 10. _,X_ ' Toxic Gases
- 11. _X_ Hazardous Material l
i
- 12. _X_ Human Factors
- 13. X Electrical Separation / Isolation
- 14. X Primary Containment Integrity / Isolation 15 X Secondary Containment Integrity / Isolation i ' 16. _X_ Equipment Reliability i
- 17. _X_ Materials Compatibility
- 18. X Single Failure Criteria 19 X Control Room Habitability
- 20. X Environmental Qualification Category i 21. X Equipment Failure Modes
- 22. X Tornado or External Flood Protection
- 23. X Protective Coatings Inside Containment
- 24. X Water Spray / Condensation
! 25. X System Design Parameters ; 26 X Test and Retest Scoping Document j
. (Post Modification test) t
~
j Safety Assessment No: SA-COLR-BF2C9 R1 Page 4 of 8 i Document No. TVA COLR BF2C9. Rev 1 POTENTIALLY l REDUCES NUCLEAR SAFETY N/A
- 27. X Chemistry Changes or Chemical i Release Pathways
- 28. X Equipment Redundancy
' l
- 29. X Equipment Diversity l
- 30. X Physical Separation
! l
- 31. X Electrical Loads
- 32. X Response Time of Emergency Safeguards Equipment
- 33. X Safety Injection / Core Cooling Capability j
- 34. X Decay Heat Removal Capability
- 35. X Reactor Coolant Pressure Boundary
- 36. X Reactor Core Parameters
- 37. X Pipe Vibration
- 38. X Security System
- 39. X Scaffolding
- 40. X Electrical Breaker Alignment Changes
- 41. _X_, I TABS, Protection Relay Settings
- 42. X Compensatory Measures !
- 43. X Environmental Impact Statement (see SSP-13.3)
- 44. X Design Basis Document l
- 45. _X_ Radwaste System Changes l
- 46. X Valve Alignment Changes
- 47. X Shield Building Integrity (SQN/WBN)
- 48. X New Radioactive Effluent (Liquid or Gaseous) Release Pathways
- 49. X Temporary Shielding
- 50. X Instrument Setpoints
- 51. X AShE Section XI
- 52. X Shutdown Reactivity Control
- 53. X Ventilation cooling for electronic equipment
Safety Assessment No: SA-COLR.BF2C9 R1 Page 5 of 8 Document No. TVA-COLR-BF2C9. Rev.1 Checklist Justification: 36 Reactor Core Parameters Applicable operating MCPR limits are determined for each cycle through cycle-specifi reload licensing analyses as described in GE licensing topical report GESTAR II (ref Events analyzed for determining MCPR operating limits include core-wide transien turbine trip, generator load rejection, loss of feedwater heating, feedwater controller failure), control rod withdrawal error, and fuel loading error. The limiting delta-CPRs determined from these analyses are added to the SLMCPR to give the MCPR opera limits. On 4/16/96, GE noti 6ed TVA (ref. 5) that an error had been discovered in the generic SLMCPR analysis for the gel 1 fuel type. The effect of the error is a non conservative SLMCPR value. GE has fixed the problem by performing cycle specific SLMCPR analyses. For Unit 2 Cycle 9, the gel 1 SLMCPR increased from 1.07 to 1.09 (ref. 6). GE's current design practice allows for only one value of SLMCPR (the most limiting for all fuel types loaded) for the core. Since the previous SLMCPR value for the Unit 2 core was 1.07, it would now become 1.09 and OLMCPRs for all fuel types loaded (GE7, GE9, ' and gel 1) would increase by 0.02 delta CPR. GE has issued a revised SRLR (ref. 2) documenting the revised OLMCPRs resulting from reanalysis of the SLMCPR. Site Licensing is opting to implement a more conservative SLMCPR value of 1.10 instead of the actual cycle specific 1.09 value. A value of 1.10 was the largest SLMCPR calculated for all BWR plants reevaluated by GE and should likely be bounding for future Browns Ferry cycles. This will potentially avoid the need for later Technical Specification submittals. Thus, the revised OLMCPRs being incorporated in Revision 1 of the unit 2 COLR (ref.1) will actually be increased by 0.03 delta CPR above their original values. The increase from 1.09 to 1.10 is conservative (i.e., provides more margin to critical power) and does not adversely affect nuclear safety. The revised OLMCPRs are bounded by previous admmistrative restrictions placeo ar. operating limits made in response to the documenting PER (ref. 3). The Option A CPP operating limits were maintained in the process computer rather than switching to the Option B CPR operating limits following scram time testing. The current OLMCPRs loaded in the process computer are now equal to the revised Option B limits for the GE9/ Gell fuel types and 0.01 greater for the GE7 fuel type. The GE7 process computer OLMCPR could therefore be relaxed by 0.01, but since it is not predicted to be the limiting fuel type at any point in the cycle this change should not be necessary. Thus, no changes to the process computer are required. 4
, w-, - ,- ,- , , . , , , - - ,
. l i
Safety Assessment No: SA-COLR BF2C9 R1 Page 6 of 8 Document No. TVA-COLR-BF2C9. Rev 1
- Since the revised OLMCPRs provide the same operating margin to the Safety Limi 9
the original analysis provided, and the changes are in the conservative direction (i.e. increasing the OLMCPR values provides additional margin to critical power), the revi OLMCPRs do not reduce nuclear safety. a Is the change acceptable from a nuclear safety standpoint? i a Yes _X_ No Justification:. ' The change is acceptable from a nuclear safety standpoint based on review of the above checklist. C. POTENTIAL TECHNICAL SPECIFICATION (T/S) IMPACT Yes No _X_ Is a change to the T/S required for conducting or implementing the change (design or procedure), test, or experiment? Justification: The core operating limits for MCPR have been removed from the Technical Specifications and placed in the Core Operating Limits Report. Therefore, incorporation of new operating limits for the cycle 9 core design does not require a Technical Specification change. Note that a separate Technical Specification change package is being processed in order to change the Safety Limit MCPR. If the answer is "Yes", a T/S change is required prior to implementation or the activity needs to be revised or canceled. D. IS AN SE OBVIOUSLY REQUIRED Yes No _X_ If yes, provide basis for conclusion and proceed to Part F.
Safety Assessment No: SA COLR-BF2C9 R1 Page 7 of 8 Document No. TVA COLR-BF2C9. Rev 1 I l E. POTENTIAL SAFETY ANALYSIS LTIPACT 4 i Yes No _X_ is this a special test, or experiment not described in the SAR? 1 l l
- 1 Does the proposed activity affect significantly (directly or indirectly) any information presented in the SAR or deviate from the description given in the SAR
- l i
I Yes _X_ No By changing: the system design or functional requirements- I the technical content of text, tables, graphs, or figures or the radwaste system? (For radwaste changes see Note in i Appendix A for guidance.) l
' Justification:
i l The revised cycle specific Supplemental Reload Licensing Report (ref. 2) needo to be incorporated into Appendix N of the FSAR. Also, A ppendix H of the FSAR needs to be revised to change the refer.ence given for the Safety ',imit MCPR from the GE Licensing Topical Report GESTAR II (ref. 4) to the applicabh BFN Technical Specifications. l Does the proposed change involve new procedures or instructions or revisions thereof that significantly: Yes No N/A _ X _ Differ with system operation characteristics from that described in the SAR7 Yes No N/A _ X _ Conflict with or affect a process or procedure outlined, summanzed, or described in the S AR' Justification: This change does not involve new procedures, instructions or revisions as described in the SAR.
. -n l
l Safety Assessment No: SA-COLR-BF2C9 R1 Page 8 of 8 ) Document No. TVA-COLR-BF2C9. Rev i i 1 If the questions are answered "No" or "N/A," the activity may be implemented without further evaluation. If any question is answered "Yes," an SE (10 CFR 50.59) is required. { CONCLUSION Incorporation of revised OLMCPRs resulting from the reanalysis of the SLMCPR into the cycle specific COLR is acceptable from a nuclear safety standpoint. Due to necessary revisions to the FSAR and COLR, a 10 CFR 50.59 safety evaluation is required. No l Technical Specification revisions are required. I 1 l F. REVIEW AND APPROVALS Preparer: n I. MPR ct. / #4 Date: 3~/3 I/il, Name ' Signature Reviewer: $4' $ "EV / M d' Date: IN'/9 L Name Signature Other Reviewers (Reactor Engineering) hew 6. e / h.M Date: 6lel% Name Sigitature
(.---.. . _ . . - - . - . - . - . . - - . . _ . - . - . . - . - - . _ . . - - . - i .m { L32 960531 { 80s QA RECC Safety Evaluation No. SE-COLR BF2C9 R1 1 Page 1 of 6 Document No. TVA-COLR-BF2C9. Rev 1 i. 4 S A FETY EV A L U ATION I j I. INTRODUCTION i A. DESCRIPTION This safety evaluation addresses: i j e i Revision 1 of the Browns Ferry Nuclear (BFN) plant unit 2 cycle 9 Core Operating Limits Report (COLR)(ref 1) , e i Revision of Appendix N of the BFN FSAR to incorporate Revision 3 of the BFN unit 2 cycle 9 Supplemental Reload Licensing Report (SRLR)(ref 2)
- e j Revision of Appendix H of the BFN FSAR to change the reference given for the Safety Limit MCPR from the GE Licensing Topical Report GESTAR II (ref 4) to the i applicable BFN Technical Specifications. -
These revisions are being made to implement revised Operating Limit Minimum Critical Power Ratio (OLMCPR) values in response to a Problem Evaluation Report (PER) (ref i
- 3). On 4/16/% General Electric (GE) notified TVA (ref 5) that an error had been i discovered in the generic analysis performed to determine the gel 1 fuel type Safety Lim MCPR (SLMCPR). GE found that the generic analysis does not produce conservative values of SLMCPR for all plants and core designs. Their corrective action was to i
reevaluate all operating plants and change their desiga methodology to include unit and cycle specific analyses. BFN Unit 2 was one of the plents for which the current SLMCPR ). specified in the plant Technical Specifications was detennined by GE (ref 6) to be non- { j conservative. This in turn affects the OLMCPRs since the OLMCPRs are determined by adding the delta CPRs from the cycle specific transient analyses to the SLMCPR. Since i current design practice allows for only one value of SLMCPR (the most limiting for all i fuel types loaded) for the core, the OLMCPRs for all fuel types (GE7, GE9, and GE l 1) l are affected. '4 1 GE issued a revision to the current operating cycle SRLR documenting the revised OLMCPRs which are being incorporated into the Unit 2 COLR. Appendix N of the BFN i- FSAR contains the current SRLR for each unit and is being revised to include Revision 3 j of the Unit 2 SRLR. i
' A Technical Specification change to revise the Safety Limit MCPR is being processed in parallel with the FSAR and COLR revisions being evaluated here.
l l i
Safety Evaluation No. SE-COLR-BF2C9 R1 Page 2 of 6 ! Document No TVA COLR-BF2C9 Rev 1 i B. REFERENCES
- l. TVA COLR BF2C9, Rev.1," Browns Ferry Nuclear Plant Unit 2 Cycle 9 Core Operating Limits Repon (COLR)."
- 2. GE Document J11-02761SRLR, Rev. 3, " Supplemental Reload Licensing Report for Browns Ferry Nuclear Plant Unit 2 Reload 8 Cycle 9," May 1996.
- 3. Problem Event Report, BFPER960502,4/18/96.
- 4. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P A-11, November 1995; and the U.S. Supplement, NEDE 24011-P-A-11-US, November j 1995.
l S. TVA Memorandum BFE 342, T.A. Keys to J.L. Lewis," Safety Limit MCPR (SLMCPR)," April 16,1996. [L32 960416 804]
- 6. GE Letter CJP2:96-139, C.J. Papandrea to T.L. Hayslett, " Safety Limit MCPR Calculation; Verified Results and BF2C9 SRLR Revision 3," May 28,1996. [L38 960529 800)
- 7. BFNP Updated Final Safety Analysis Report, revised through Amendment 12.
- 8. BFNP Technical Specifications, Unit 2 (through Amendment No. 244).
l l l l l 4 e l I I 1
\
i Safety Evaluation No. SE-COLR-BF2C9 R1
' Page 3 of 6 Document No. TVA COLR BF2C9. Rev.1 i
$ IL SAFETY EVALUATION 1 i , A. ACCIDENTS EVALUATED AS THE DESIGN BASIS A.I. Deshga Basir Accidents and Anticipated Operational Transients: Revising the OLMCPRs potentially affects the initial conditions and resulting consequences of all Anticipated Opctational Transients (AOT). l A.2. Credible Failure Modes of Proposed Activity 4 i Revising the OLMCPRs could potentially affect fuel failures resulting during an AOT which is addressed in A. I. No other failure mechanisms of the fuel or any ) other plant equipment are identified as being affectedty the proposed change. I } B. EVALUATION OF EFFECTS i i B.l. May the proposed activity increase the probability of an accident previously j evaluated in the SAR? 4 h Ya_ No _X._. ) Justificatior.: I l The anticipated operational transients and design basis accidents described in the ' FSAR are initiated by operator errors or equipment malfunctions. Revising the OLMCPRs does not affect any of these initiating events and thus will not increase the probabihty of an accident previously evaluated in the SAR. j B.2. May the proposed activity increase the probability of occurrence of a f malfunction of equipment in portant to safety previously evaluated in j the SAR7 I Yes Na_X_ Justification: { i ( ' The revised OLMCPRs being implemented redect the results of a reanalysis of :Fe SLMCPR for Gell fuel. 'Ihe SLMCPR is set to ensure that 99.9% of the fuel r.c
[ _.
]
i
- Safety Evaluation No. SE-COLR-BF2C9 R1 Page 4 of 6 Document No. TVA-COLR-BF2C9. Rev.1 will not experience departure from nucleate boiling. The margin between the OLMCPRs and the onset of departure from nucleate boiling has been increased in order to keep the same probability that rods will not experience depanure from nucleate boiling. Therefore, the probability of fuer failure following any AOT has not changed. The revised OLMCPRs will not affect other failure mechanisms for the fuel such as corrosion, fretting, manufacturing defects, etc. which are unrelated to operating thermal margins. The revised OLMCPRs will also have no impact on failure mechanisms for other plant equipment besides the fuel. For these reasons, the revised OLMCPRs do not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.
B1 May the proposed activity increase the consequences of an accident previously evaluated in the SAR? Yes No X Justification: As discussed in B.2, the revised OLMCPRs being implemented are just as conservative as the original limits. Thus operating with the revised limits will provide the same probability of fuel fadure and resulting consequences following any AOTs. Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the SAR. B.4. May the proposed activity incream the consequences of a malfunction of equipment important to safety previeusly evaluated in the S AR? Yes No _X_ Justification: As discussed in B.2., the revised OLMCPRs being implemented are just as conservative as the original limits were intended. Thus, the resulting consequences following any AOTs will be the same es those in the original cycle-specific analysis The revised limits will not affect other failure mechanisms for the fuel which are unrelated to thermallimits. The revised limits will also not affect failure mechanisms or consequences of malfunctions of plant equipment other than fuel. Therefore, the proposed activity does not increase the consequences of a malfunction of equipment important to safety previously evaluated m the " AR.
j.. Safety Evaluation No. SE-COLR-BF2C9 R1 Page 5 of 6 Document No. TVA-COLR-BF2C9. Rev.1
< B.S. May the proposed activity create a possibility for an accident of a different type than any evaluated previously in the SAR?
Yes 4 No _X_ Justification: 5 The revised OLMCPRs being implemented restore the previous thermal margins for the fuel. The revised OLMCPRs will not affect other neutronics characteristics of the fuel such as its ability to provide negative reactivity feedback or prevent power oscillations, or cause the fuel to operate in any unanalyzed conditions. It is concluded that the proposed activity does not create the possibility for an accident of a different type than any evaluated l previously in the SAR. B.6. May the proposed activity create a possibility for a malfunction of a , different type than any evaluated previously in the SAR? 4
- Yes No _X_
Justification: The revised OLMCPRs being implemented will not cause the fuel to functio manner different from normal or to operate outside any design bases. The revised OLMCPRs also will not affect the functioning of any plant equipment other than the fuel. Therefore, the proposed change does not create a possibility for a malfunction of a different type than any evaluated previously in the SAR. i B.7. May the proposed activity reigce the margin of safety as defined in the bases 3 for any Technical Specification? 1 Yes No _X_ Justi6 cation: The revised OLMCPRs reflect the change in the SLMCPR and provide margin equa to the original limits Thus, the proposed change will not reduce the margin of safety as defined in the bases for any Technical specification. e 4
4 I Safety Evaluation No. SE-COLR-BF2C9 R1 i Page 6 of 6 Document No. TVA-COLR-BF2C9 Rev 1 C. UNREVIEWED SAFETY QUESTION DETERMINATION CONCLUSION The change, test, or experiment: i 4 _X_ Does not involve an unreviewed safety question. Involves an unreviewed safety question and must be revised, canceled, or reviewed by the NRC prior to implementation. D. REVIEW AND APPROVALS Preparer: Nn d. MNcf / Date: I/3I/fd Name Signature Reviewer: bh E- 'tEY / Date: N3M96 Name. Signature / Reviewer: / Date: (PORC/QR)* Name Signature Other Reviewers (Reactor Engineering) Ke w era c. u e.a / .M . A b b, Date: 6[6[96 Name ' Signature
- This review is not required for corporate level procedures.
(
- As required by Technical Specifications.
l ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, AND 3 TECHNICAL SPECIFICATION (TS) BASES CHANGE TS-377 MARKED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 l 3.10/4.10-14 3.10/4.10-14 3.10/4.10-13 II. MARKED PAGES l l See attached. l l
3.10. E S & (Cont'd) FEB 2 3195 l
- 2. Morssa, W. R., "In-Core Neutron Monitoring System for General i Electric Boiling Water Reactors," General Electric Company, Atomic
!- Fower Equipment Department, November 1963, revised April 1969 j (AFID-5706) 4 C. Scene ruel Pool Water
{ S e design of the spent fuel storage pool provides a storage location j for approximately 140 percent of the full core load of fuel assemblies j in the reactor building which ensures adequate shielding, cooling, and i reactivity control of irradiated fuel. An analysis has been performed ! which shows that a water level at or in excess of eight and one-half i feet over the top of the stored assemblies will provide shielding such ! that the ==w h calculated radiological doses do not exceed the lialts 1 of 10 CFR 20. The normal water level prwides 14-1/2 feet of
! additional water shielding. The capacity of the skimmer surge tanks is l
! available to maintain the water level at its normal height for three I days in the absence of additional water input from the condensate i storage tanks. All penetrations of the fuel pool have been installed l at such a height that their presence does not provide a possible
, drainage route that could lower the normal water 2evel more than I one-half foot. l l l l The fuel pool cooling system u designed to maintain the pool water l' j temperature less than 125'F during normal hast loads. If the reactor core is completely unloaded when the pool contains two previous
! discharge batches, the temperature may increase to ter than 125'F. The RER system supplemental fuel pool cooling mode sti be used under i ! these conditions to maintain the pool temperature t 4 than 125'F. I ) 3.10.D/4.10.D RASES E C46 Reactor Buildina Crana The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building. The controls for the 125-tor. hoist are located in the crema cab. The five-tem has both est and pendant controls. Am.vasent inspection of the load-bearing hoist wire rope essures ' maestion of signs of distress or wear so that corrections can be prompaly made if needed. The testing of the various limits and interlocks assures their proper 1 operation when the crane is used. 3.10.E/4.10.E i Snent Fuel Cank i The spent fuel cask design incorporates removable lifting trunniens. The visual inspection of the trunnions and fasteners prior to BFN 3.10/4.10-14 l TS 348 - TVA Letter 2 NRC em o=c_d 02/23/95
(.-.-.---..--.~..-_-.-..-.--...- - . -. -..- - .~ . -- _ . _ . _ _ - i - i I J.10 M (Cont'd) FEB 2 31!!!E ! m l 3. Morgan, W. R., "In-Core Neutron Monitoring System for General 1 Electric Boiling Water teactors," General Electric Company, Atoalc Power Equipment Department. November 1963, revised April 1969 (APED-5706) ! C. Seant Fuel Peel Water , l l The design of the spent fuel storage pool provides a storage location for appronimately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shieldias, cooling, and reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in ascess of eight and one-half , feet over the top of the stored assemblies will provide shielding such ! that the nazimum calculated radiological doses de met ascoed the limits
- of 10 Cyt 20. h seraal water level provides 14-1/2 feet of l additional water shieldias. N capacity of the akimmer surge tanks is i available to maintain the water level at its aeraal height for three days in the absence of additional water input from the condensate
, storage tanks. All penetrations of the fuel pool have been installed i at such a height that their presence does not provide a possible ! drainage route that could lower the normal water level more than one-half foot. N fuel pool cooling system is designed to asiatain the pool water temperature less than 125'F during normal heat leads. If the reactor core is completely mioaded wh a the peel contains two previoua 9
- discharge batches, the temperature may increase to ter than 125'F.
i The IIR system supplaneatal inel pool cooling med win used under ! these conditises to maintain the peel temperature oTless than 125'F. ! i l D. Reactar Buildina Crana b C06 i j The reactor building crane and 125-ten helst are required to be , ! operable for headling of the spent fuel la the roaster building. N l controis for the 125-ten heist are located in the crans cab. N five-tem has both cab and pendsat controls. l A visual inspecties of the load-bearing heist wire rope assures detectism of signs of distress er wear se that corrections can be promptly made if aseded. The testias of the varioes limits and laterleeks assares their proper operatica when the crana is used. E. annar Fuel cask The spent fuel eask design incorporates removable lifting trusalons. The visual inspection of the tra aiens and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior ba ilias. - h trenions anst be properly attached to the S cash ier liftin of the cash and the visual inspecttea assures correet installaties. ~ gyg 3.10/4.10-14 l TS 348 - TVA Lettar :. 4C gagg 3 Dated 02/23/95
, en I i l 3.10 A&831 (Cent'd) NOV 17195
. l
{ ' 4
- 2. Morgan, W. R., "In-Core Neutron Monitoring Syst e for General
! Electric Soiling Water Reactors," General Electric Ceapany, Atoale j Fower Equipasat Department, November 1964, revised April 1969 l ! (AFID-5704) i ! C. Scen$ Fuel Fool Water - l The design of the spent fuel storage pool provides a storage location ' for approximately 140 percent of the full core load of fuel asamblies in the reactor building which ensures adequate shielding, cooling, and ! reactivity control of irradisted fuel. An analysis has been performed ! which shows that a water level at or in excess of eight and one-half i feet over the top of the stored assemblies will provide shielding such that the anziana calculated radiological doses do not exceed the limits l of 10 CFR 20. The normal water level provides 14-1/2 feet of I additional water shielding. The capacity of the skimmer surge tanks is i i available to maintain the water level at its normal height for three days in the absece of additional water impet from the condesata storage tanks. All penetrations of the fuel pool hafe been installed at such a height that their presence does not provide a possible
- drainage route that could lower the normal water level more than l j one-half foot.
The fuel pool cooling system is desigand to asistain the pool water i temperature less than 125'F during normal heat loads. If the reactor i core is completely mioeded when the peel costalas two previous l discharge batches, the temperatures aar increase to greater i 125'F. N RER system supplemsstal fuel peel coelias ande be used eder these canditions to asintain the pool temperature t less 125'F. 1 Con l D. Ramater Buildina Crama 1 l l N reactor building crema and 125-tem heist are required to be
- 0FIEABLE for handling of the spent fuel la the remeter building. *Pe i
' centrols for the 125-ten hoist are located la the crene cab. The l five-ten has both cab and pendant centrols. i a visual inspection of the lead-bearing heist wire rope assures of signs of distress or wear se that corrections can be made if aseded. Thai testing of the various limits and interlocks assures their proper operaties when the crana is used. E. Snent Fuel Caak N spent fuel cask design incorporates removable lifting trunnions. N visual inspection of the trunnions and f' ; ears prior to i j RF5 3.10/4.10-13 T5 370 ) < Unit 3 Letter Dated 11.
.- . . - - - . _ . - . .- - . - _. . = . _ . ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, AND 3 TECHNICAL SPECIFICATION (TS) BASES CHANGE TS-377 REVISED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 3.10/4.10-14 3.10/4.10-14 3.10/4.10-13 I i II. REVISED PAGES l See attached. I 1
)
l l
3.10 BASES (Cont'd)
- 2. Morgan, W. R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactorn," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706)
C. Soent Puel Pool Water The design of the spent fuel storage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20. The normal water level provides 14-1/2 feet of additional water shielding. The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input from the condensate storage tanks. All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-half foot. The fuel pool cooling system is designed to maintain the pool water ta..perature less than 125*F during normal heat loads. If the reactor core is completely unloaded when the pool contains two previous discharge batches, the temperature may increase to greater than 125'F. The RHR system supplemental fuel pool cooling mode can be used under l these conditions to maintain the pool temperature to less than 125'F. 3.10.D/4.10.D BASES Reactor Buildina Crane The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building. The controls for the 125-ton hoist are located in the crane cab. The five-ton has both cab and pendant controls. A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed. l The testing of the various limits and interlocks assures their proper operation when the crane is used. 3.10.E/4.10.E Scent Fuel Cask The spent fuel cask design incorporates removable lifting trunnions. The visual inspection of the trunnions and fasteners prior to BFN 3.10/4.10-14 Unit 1 L. )
1-3.10 BASES (Cont'd)
- 2. Morgan,'W. R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactors," General Electric Company, Atomic Power Equipment Department, No/ ember 1968, revised April 1969 (APED-5706)
C. Scent Fuel Pool Water The design of the spent fuel storage pool'provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20. The normal water level provides 14-1/2 feet of additional water shielding. The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input from the condensate storage tanks. All penetrations of the fuel pool have been installed at such a height'that their presence does not provide a possible drainage route that could lower the normal water level more than one-half' foot. The fuel pool cooling system is designed to maintain the pool water temperature less than 125*F during normal heat-loads. If the reactor core is completely unloaded.when the pool contains two previous discharge batches, the temperature may . increase to greater than 125'F. The RHR system supplemental fuel pool cooling mode can be used.under l these conditions to maintain the pool temperature to less than 125'F. ; D. Reactor Buildina Crane The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building. The controls for the 125-ton hoist are located in the crane cab. The five-ton has both cab and pendant controla. A visual inspection of the load-bearing hoist wire rope assures I detection of signs of distress or wear so that corrections can be promptly made if needed. The testing of the various limits and interlocks assures their proper operation when the crane is used. E. Spent Fuel Cask The spent fuel cask design incorporates removable lifting trunnions. The visual inspection of the trunnions and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior handling. The trunnions must be properly attached-to the cask for lifting of the cask and the visual inspection assures correct installation. BFN 3.10/4.10-14 Unit 2 9
3.10 BASES (Cont'd)
- 2. Morgan, W. R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactors," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706)
C. Soent Fuel Pool Water The design of the spent fuel storage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half i feet over the top of the stored assemblies will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20. The normal water level provides 14-1/2 feet of additional water shielding. The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input from the condensate storage tanks. All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-half foot. The fuel pool cooling system is designed to maintain the pool water temperature less than 125*F during normal heat loads. If the reactor core is completely unloaded when the pool contains two previous discharge batches, the temperatures may increase to greater than 125*F. The RHR system supplemental fuel pool cooling mode can be used under l these conditions to maintain the pool temperature to less than 125'F. D. Reactor Buildina Crane The reactor building crane and 125-ton hoist are required to be OPERABLE for handling of the spent fuel in the reactor building. The controls for the 125-ton hoist are located in the crane cab. The five-ton has both cab and pendant controls. A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed. The testing of the various limits and interlocks assures their proper operation when the crane is used. E. Soent Puel cask The spent fuel cask design incorporates removable lif ting trunnions. The visual inspection of the trunnions and fasteners prior to BFN 3.10/4.10-13 Unit 3
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