ML18033A360
| ML18033A360 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/31/1988 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18033A358 | List: |
| References | |
| TVA-RLR-002, TVA-RLR-002-R02, TVA-RLR-2, TVA-RLR-2-R2, NUDOCS 8809080020 | |
| Download: ML18033A360 (43) | |
Text
TVA-RLR-002 Revision 2
July 1988 TENNESSEE VALLEY AUTHORITY BROGANS FERRY NUCLEAR PLANT UNIT 2, CYCLE 6 RELOAD LICENSING REPORT PDR0
'020 ge0826 AEOLK Ou0002A0 PDC 5503B
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Tennessee Valley Authority
1 Revision 2
July 1988 I.
Introduction This reload licensing report presents the results of the core redesign and safety analyses performed for Browns Ferry Nuclear Plant (BFN) unit 2, cycle 6 operation.
The current licensed design is documented in references 1 and 2.
The methodology and technical bases employed in the performance of these analyses are discussed in references 3-8.
Items specifically addressed here include the nuclear fuel assemblies and core loading to be used in cycle 6, the reload-core nuclear design characteristics, the transient and accident, safety analysis results, and the proposed operating thermal limits.
The cycle 6 reload core will include four Westinghouse QUAD+
demonstration assemblies located in nonlimiting core peripheral locations.
A complete description of the demonstration assemblies is contained in Westinghouse Report WCAP-10507 (reference 9).
The cycle 6 core'oading has been changed based on results of inspection and reconstitution of the fuel available for use in cycle 6
~
The unit 1 once-burned fuel will replace the unit 2 once-burned for unit 2, cycle 6.
- Also, 212 twice-and thrice-burned bundles to be loaded were inspected and reconstituted as needed.
II.
Reload C cle Information A.
Design Basis Exposures 1.
Actual cycle 5 core average exposure at end of cycle:
20.8 GWd/ST 2.
Hinimum cycle 5 core average exposure at end of cycle from cold shutdown considerations:
- 20. 8 GWd/ST 3.
Assumed cycle 6 core average exposure at depletion of reactivity (DOR)*'7.9 GMd/ST B.
Reload Fuel Assemblies
~Fuel T e
C cle Loaded Number Irradiated 8DRB284L,U2R2 P8DRB284L,U2R3 P8DRB265H,U1R5 PBDRB284L,U1R5 PBDRB284Z,U1R5 V2CY3 U2CY4 U1CY6 U1CY6 V1CY6 53 159 160 80 8
New P8DRB284L,U2R5 QUAD+ Demo U2CY6 U2CY6 300 4
TOTAL 764 WOR=~nd of full power capability 5503B
2 Revision 2
July 1988 Descriptions of the nuclear and mechanical design of the General Electric irradiated and new fuel assemblies to be loaded in cycle 6
are contained in reference 10.
The nuclear, mechanical, and thermal-hydraulic design descriptions for the Westinghouse demonstration assemblies are contained in reference 9.
C.
Reference Core Loading Pattern The reference loading pattern is the basis.for all reload licensing and operational planning and is comprised of the fuel assemblies designated in item II.B of this report. It is based on the cor.e condition at the end of the previous cycle, the number and type of fuel assemblies suitable for use, and on the desired core energy capability for the reload cycle.
The reference loading pattern is designed with the intent that it will represent, as closely as
- possible, the actual core loading pattern.
Figure 1 shows the reference core loading pattern for cycle 6.
The reference loading pattern includes four westinghouse QUAD+
demonstration assemblies loaded in peripheral locations.
These locations satisfy the criteria specified in references 2 and 9.
Evaluations performed by Westinghouse (reference
- 9) show that the results of licensing analyses for the lead P8x8R fuel assembly bound those for the QUAD+ demonstration assemblies.
Cycle specific analyses performed by TVA confirm this conclusion.
A total of 212 twice-and thrice-burned assemblies were inspected and reconstituted for use in cycle 6.
Prior to the reconstitution project, guidelines were implemented to ensure that performance of the reconstituted assemblies would not differ significant.ly from the original assemblies.
Consequently, the safety analysis results reported in this document were generated with the reconstituted assemblies modeled as original assemblies.
Following completion of the reconstitution work, this modeling assumption was verified by individually analyzing each reconstituted assembly and by performing, core-wide analyses to specifically address the effects of reconstitution.
These analyses confirmed that all design criteria are satisfied and that operating limits reported in this document remain valid.
D.
Special Conditions The use of increased core flow (ICF) is planned for cycle 6
operation.
Safety analyses were performed for both 100 percent and 105 percent of rated core flow with the most conservative results used for determining the operate.ng limits.
The conclusions regarding LOCA analysis, reactor internals pressure
- drop, and flow-induced vibration as discussed in reference 11 are applicable to cycle 6.
The flow-biased instrumentation for the rod block monitor will be signal clipped For a setpoint of 106 percent since flow rates higher than rated would otherwise result in a ACPR higher than reported for the rod withdrawal error.
5503B
3 Revision 2
July 1988 III. Nuclear Desi n Characteristics A.
Shutdown Margin The reference core is analyzed in detail to ensure that adequate shutdown margin exists.
This section discusses the results of core calculations for shutdown margin (including the standby liquid control system).
1.
Core Effective Multiplication and Control Rod Worth Core effective multiplication and control rod worths were calculated using the TVA BWR simulator code (references 4 and 6) in conjunction with the TVA lattice physics data generation code (references 5 and 6) to determine the core reactivity with all rods withdrawn and with all rods inserted.
A tabulation of the results is provided in table 1.
These three eigenvalues (effective multiplication of the core:
uncontrolled, fully controlled, and with the strongest rod out) were calculated at the beginning-of-cycle 6 core average exposure corresponding to the actual end-of-previous-cycle core average exposure.
The core was assumed to be in a xenon-free condition.
Cold keff was calculated with the strongest control rod out at various exposures through the cycle.
The value R is the difference between the strongest rod out keff at BOC and the maximum calculated strongest rod out keff at any exposure point.
The maximum strongest rod out keff at any exposure point is equal to or less than:
SRO SRO Maximum keff = keff (BOC) + R 2.
Reactor Shutdown Margin Technical Specifications require that the refueled core must be capable of being made subcritical with 0.38-percent Ak margin in the most reactive condition throughout the subsequent operating cycle with the most reactive control rod in its full out position and all other rods fully inserted.
The shutdown margin is determined by using the BWR simulator code to calculate the core multiplication at selected exposure points with the strongest rod fully withdrawn.
The shutdown margin for SRO the reloaded core is obtained by subtracting the maximum keff from the critical keff of 1.0, resulting in a calculated minimum cold shutdown margin of 1.0-percent Ak for BFN unit 2, cycle 6.
5503B
4 Revision 2
July 1988 Table 1
CALCULATED CORE EFFECTIVE MULTIPLICATION NO VOIDS, NO XENON, 20 C
CON Fully Controlled, Keff (BOC)
SRO Strongest Control Rod Out, Keff (BOC)
R, Maximum Increase in Cold Core Reactivity Mith Exposure Into Cycle, 4k 1:120
- 0. 956.
0.985 0 '05 3.
Standby Liquid Control System The standby liquid control system (SLCS) is designed to provide the capability of bringing the reactor, at any time in a cycle, from full power and a minimum control rod inventory (which is defined to be at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon-free state, The SLCS shutdown marg,in is determined by using, the BMR simulator code to calculate the core multiplication for the cold, xenon-free, all-rods-out condition at the exposure point of maximum cold reactivity with the soluble boron concentration given in the Technical Specifications.
The resulting k-effective is subtracted from the critical k-effective of 1.0 to obtain the SLCS shutdown margin.
The results of the SLCS evaluation are g,iven in table 2.
Table 2
STANDBY LIQUID CONTROL SYSTEM CAPABILITY PPM Shutdown Margin (4k) 20 C
Xenon Free 660 0.029 B.
Reactivity Coefficients The reactivity coefficients associated with the nuclear design of BFN unit 2, cycle 6 are implicit in the 1-D cross sections used for the safety analyses.
As such, reactivity coefficients are not separately calculated for input to the transient analyses.
- However, a void coefficient is generated in the 3-D to 1-D cross section collapsing process and is used as a verification check, For BFN unit 2, cycle 6 the following results were obtained:
100'7. core flow, DOR 1051'ore flow, EDOR~
-0. 0734
-0. 0745
%4k/%void 14k/%void
~
EDOR extended depletion of reactivity resulting from increased core flow.
5 Revision 2
July 1988 C.
Fuel Performance The BFN unit 2, cycle 6 fuel performance is predicted by projecting the fuel burnup to the end of cycle with the 3-D simulator code.
The calculated peak pellet exposures for the various fuel types are less than the limits specified in references 9 and 10.
Furthermore, peak linear. heat rates satisfy the assumptions made in the fuel vendors'hermal-mechanical integrity analyses (references 9 and 10).
All fuel types loaded in cycle 6 are predicted to operate within these bounding, assumptions, Additionally, the QUAD+
demonstration assemblies are predicted to have substantial margin to the lead P8x8R assembly in steady-state bundle power and thermal limits throughout cycle 6 (figures 20-22).
The minimum margin for bundle power is 27 percent which satisfies the requirement for at least a 20-percent margin specified by NRC (reference 2).
For MCPR the minimum margin is 43 percent and for LHGR it is 32 percent.
IV.
Transient Anal ses A.
Pressurization Events The RETRAN computer code (reference
- 12) is used to analyze both the reactor system and hot channel responses during core-wide pressurization transients.
The analytic models used in these analyses are described in reference 7.
A description of the CPR correlation.and its application to Browns Ferry is contained in reference 13.
Analyses are performed for the potentially limiting events at the most adverse initial conditions expected during the cycle.
Reload unique initial conditions and transient analyses results are summarized in the following tables.
NSSS Initial Conditions ExXleaeee Steam Flow 1o Rated Core Flow
~LRated Gap Conductance BTU/ft~-hr-4F EDOR 105 105 674 I
Hot Channel Initial Conditions Limitin Event Fuel
~Te ICPR P8X8R 1.295 Bundle P~owev RW 6.416 Bundle Flow Klb/hr 123.7 R-Factor 1.051 Gap Conductance BTU/ft~-hr-F 1287 5503B
6 Revision 2
July 1988 Pressurization Event Anal sis Results Peak Power Peak Heat Peak Vessel ACPRi System Load Rejection w/o Bypass 403.4 121.6 1235.3 0.225 Figures 2-5 Feedwater Controller Failure 234.8 115.5 1215.1 0.149 Figures 6-9 B.
Honpressurization Events The nonpressurization events analyzed for reload licensing are either steady-state events or relatively slow transients that can be analyzed in a quasi-static manner using a 3-D BWR simulator (reference 4).
The methods used to analyze these events are described in reference 3.
Results are summarized below.
'on ressurization Event Anal sis Results Event ACPR4 Peak LHGR kW/ft 4 P8xSR/SxSR/
UAD+
PSx8R/8x8R/
UAD+
Loss of Feedwater Heating (100 F) 0.18 17.5 Rod Withdrawal Error Rotated Bundle Error 0.20~
0.193 20.8 15.3 Hislocated Bundle Error 0.13 14.4 i Results presented were calculated for P8x8R fuel and will be conservatively applied to Sx8R.
~ For increased core flow based on a signal clipped rod block setpoint of 106 percent.
3 Includes 0.02 penalty required when using the variable water gap method (reference 10).
4 Results presented were calculated for the P8x8R fuel type and conservatively bound the results calculated for the 8x8R fuel type.
The results are also bounding for the QUAD+ demonstration assemblies which will be loaded into nonlimiting, peripheral core locations.
5503B
7 Revision 2
July 1988 C.
Overpressure Protection The main steamline isolation valve closure with failure of direct scram is analyzed to demonstrate sufficient overpressure protection (peak vessel pressure must be less than 110 percent of design pressure 1390'psia).
The event is analyzed using the models and methods described in reference 7.
Results are summarized below.
MSIV Closure Flux Scram Results Peak Vessel Pressure sia Peak Steamline Pressure sia System R~es ense 1281.0 1242.5 Figures 10-13 MCPR 0 eratin Limit Summar The methods used to determine the required OLMCPR values for each event analyzed are described in references 3 and 7.
The application of Options A
and B limits in determining the cycle OLMCPR is described in the unit Technical Specifications, Results are summarized below and in figure 14.
OLMCPR for Pressurization Events BOC6-EOC6 P8x8R/8x8R/
UAD+
P8x8R/8x8R/
UAD+
Load Rejection Without Bypass (GLRWOB) 1.35 1.26 Feedwater Controller Failure (FWCF) 1.27 1.23 OLMCPR for Non ressurization Events BOC6-EOC6 P8x8R/8x8R/
UAD+i Loss of Feedwater Heaters (LFWH)
Rod Withdrawal Error (RWE)
Rotated Bundle Error (RBE)
Mislocated Bundle Error (MBE) 1.25
, 1.27 1.26 1.20 x Results presented were calculated for the P8x8R fuel type and conservatively bound the results calculated for the 8x8R fuel type.
The QUAD+ demonstration assemblies will be loaded into nonlimiting core locations and monitored to the same OLMCPR.
5503B
VI.
Accident Anal ses 8
Revision 2
July 1988 A.
Loss of Coolant Accident (LOCA)
MAPLHGR limits for the unit 1 PSDRB284Z fuel type (from reference
- 14) still apply for fuel being transferred to unit 2 since the LOCA responses for the two units are identical (reference 15),
The limits for remaining fuel types are taken from reference 16.
Reference 9 indicates that the MAPLHGR limits for fuel type P8DRB284L can be conservatively applied to QUAD+ demonstration assemblies.
Tables of MAPLHGR limits for all fuel types in unit 2, cycle 6 are presented below.
1 LOCA Limits for UAD+ Demonstration Assemblies Average Planar E
osure MWd/t MAPLHGR
~kM/ft 200 1,000 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 11.2 11.3 11.8 12.0 12.0 11.8 11.2 10.8 10.2 9.5 8.8 LOCA Limits for GE Fuel T e P8DRB284Z Average Planar E
osure HMd/t MAPLHGR
~kM/ft 200 1,000 5,000 10,000 15,000 20,000 25>000 30,000 35,000 40,000 45,000 11.2 11.2 11.7 12.0 12.0 11.8 11.1
-10.4 9.8 9.1 8.5 5503B
LOCA Limits for GE Fuel T e P8DRB265H 9
Revision 2
July 1988 Average Planar E
osure MMd/t MAPLHGR
~kM/ft 200 1,000 5,000 10,000 15,000 20)000 25,000 30,000 35,000 40,000 11.5 11.6 11.9 12.1 12.1.
11.9 11.3 10.7 10.2 9.6 LOCA Limits for GE Fuel T e 8DRB284L Average Planar E
osure MMd/t MAPLHGR
~kM/ft 200 1,000 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 11.2 11.3 11.8 12.0 12.0 11.8 11.2 10.8 10.2 9.5 LOCA Limits for GE Fuel T e P8DRB284L Average Planar E
osure HMd/t MAPLHGR
~kW/ft 200 1,000 5,000 10)000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 11.2 11.3 11.8 12.0 12.0 11.8 11.2 10.8 10.2 9.5 8.8 5503B
B.
Rod Drop Accident (RDA) 10 Revision 2
July 1988 The methodology used to analyze the rod drop accident is described in appendix A of reference 8.
Results for unit 2, cycle 6 are summarized below.
Results for the Limitin RDA Condition:
375 F, MOC Exposure Rod Worth:
1.05 percent hk Rod Position:
38-15 Peak Fuel Enthalpy:
194.5 cal/gm Core Response:
Figures 15-18 VII. Stabilit Anal ses The methodology used to analyze core and channel stability is described in appendix B of reference 8.
The minimum stability margin occurs at the intersection of the natural circulation line and the 105-percent rod line (the flow biased scram line also passes through this point.).
Results for BFN unit 2, cycle 6 are summarized below and in figure 19.
Stabilit Anal sis Results at Limitin Initial Conditions
~Anal ala Maximum Deca Ratio Core Stability Channel Stability P8x8R/8x8R/QUAD+
0.84~
0.59~
~ Includes 0.14 uncertainty adder as described in appendix B of reference 8.
~ Results presented are for the P8x8R fuel type and conservatively bound the 8x8R fuel type and the QUAD+ demonstration assemblies 5503B
References 11 Revision 2
July 1988 1.
TVA-RLR-002, Rev.
1 dated April 1985, "Reload Licensing Report for Browns Ferry Unit 2, Cycle 6," TVA.
2.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 125 to Facility Operating License No. DPR-52, Tennessee Valley Authority, Browns Ferry Huclear Power Plant, Unit 2, Docket Ho. 50-260.
3.
TVA-EG-047 dated January
4.
TVA-TR78-03A dated January 1979, "Three-Dimensional LWR Core Simulation Methods,"
TVA.
5.
TVA-TR78-02A dated April 1978, "Methods for the Lattice Physics Analysis of LWRs," TVA.
6.
TVA-TR79-01A dated January 1979, "Verification of TVA Steady-State BWR Physics Methods,"
TVA.
7.
TVA-TR81-01A dated December
TVA.
8, TVA-RLR-001 dated January
- 1984, "Reload Licensing Report for Browns Ferry Unit 3, Cycle 6," TVA.
9.
WCAP-10507 dated March 1984, "QUAD+ Demonstration Assembly Report,"
Westinghouse Electric Corporation.
10.
HEDE-24011-P-A-8 dated May 1986, "General Electric Standard Application for Reactor Fuel," General Electric.
11.
NEDO-22245 dated October 1982, "Safety Review of Browns Ferry Nuclear Plant Unit No.
2 at Core Flow Conditions Above Rated Core Flow During Cycle 5," General Electric.
12.
EPRI HP-1850-CCM dated May 1981, "RETRAN02 A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," Electric Power Research Institute.
13.
HEDE-24273, "GEXL Correlation Application to TVA Browns Ferry Nuclear Power Station," General Electric.
14.
NED0-24056, Rev. 1, dated May 1983, "Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 1," General Electric.
15.
DGC:88-146, Letter from D.
G. Churlik to J.
D. Robertson dated July 13,
- 1988, "Telecon of 7/13/88," General Electric.
16.
NEDO-24088-2 (as amended) dated May 1985, "Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2," General Electric.
5503B
FIGURE 1 REFERENCE LOADING PATTERN BROWNS FERRY UNIT 2 - CYCLE 6 12 Revision 2
July 1988 60 68 66 64 62 60 48 46 44.
42 40 38 36 32 30 28 2B 22 20 18 18 14 12 10 8
8 2
B B F A A F F
C F
C F A B B B A BBBF BBFC BF DF QCF C
EF BF EF BF F DF D
CF DA FCFD CF DB FCFD CF CF QCF C
BF DF BBF C
BBBF B A B
B A F F
B C F F
D D F B D C F B
C D F F A A F F
C F
C A F F A B C B D D F F
D C F F
B C F F
C A F B
B B B F A C F B F F
C D F F
D D B C B F D A F F A C F A F F
D C
B D B F
D D F F
C C F F A B
B BBB F AF CFA FCF C F C
F CF CF C
F DF DBC BFC BFC DBC F DF DF D F DF C
F CF CF C
CFA B8B BBB FBB DF BB FCQ B
FCF CF DFCFC FDF BF BFEFC BFEFC FDFBF DFCFC F CF CF DFCFC F CF CF DFCFC FCF DF CF CF B
FCQ B
DF BB FBB ABB B B FBB FAB CF B
FAB CF B
FFB F
F CF B
F AB C F B
FAB F
B FBB B
B 3
7 9
13 17 21 26 29 33 37 41 46 49 63 67 39 43 47 6 1 66 6
11 16 19 23 27 31 36 A~ 8DRB284L,U2R2 C= PSDRB286H,U1R6 E= PSDRB284Z,U1R6 Q= QUAD+DEMO,U2R5 B~ P8DRB284L,U2R3 D= PSDRB284L,U1R6 F= P8DRB284L,U2R6
FIGURE 2
- EIF2CY6, GLRWOB ICF 13 Revision 2
July 1988 500 400 300 Legend TOTAL POWER 7.
AYE SURFACE HEAT FLUX {X$
CORE INLET FLOW (Xg CORE/EEET RU8~COOLIRO 8) 200 100
~
~ 8
~ 8
~ ~
8
~ ~
~
~
~
~ 8 ~
~
~
~ ~ ~ ~ ~
~
~ ~
~
~
~ ~ ~ 8
~ ~
~
~
0 0
3 4
TIME (SEC)
FIGURE 3
- BFZCY6, GLRWOB ICF 200 150 100 50 0
0 I
I II I
IIIIIII 3
4 TIME (SEC) 5 6 7 Legend VESSEL PRESS RISE PSI TOTAL SgR VALVEFLOW {g BYPASS VALVEFLOW {g
FIGURE 4
- BF2CY6, GLRWOB ICF 14 Revision 2
July 1988 150 100 50 l
I e
I I
I I
II I
e I I I
0
-50
-100 0
~
I I
II I I I I II II I
I
~
I I
I I
TI
~
~
~
~
Legend LEVEL INCH-REF-SEP SKIR VESSEL STEAM FlQW (5)
TURBINE STEAM FLOW h}
FEED WATER FLOWER 3
4 TIME (SEC)
FIGURE 5
- BF2CY6, GLRWOB ICF Leg e'nd TOTAL REACllVITY SCRAM REACTIVITY$Q
~ ~ ~e e~
0.5 1.5 TIME (SEC) 2.5
FIGURE 6
- BF2CY6, FWCF ICF 15 Revision 2
July 1988 250 200 150 100 50 0
Legend TOTAL POWER X
AVE SURFACE HEAT FLUX (XQ CORE INLET FLOW fX) coco eeoc eue~cooueo e}
~ ~
~ 0 ~ ~r 5'0 15 20 25 TIME (SEC)
FIGURE 7
BYPASS VALVE FLOW (g 50 0
0 10 15 l
Il I
II 20 II I
ItIII I
~ 'I
~ a\\re I
I I
I I
I 25 TlME (SEC)
FIGURE 8
- BF2CY6, FVlCF ICF 16 Revision 2
July 1988 150 100 50 0
-50 Legend LEVEL INCH-REF-SEP SKI VESSEL STEAM FLOW (X)
TURBINE STEAM FNW (X)
FEEDwATER FLO'lfJÃX Il II I
I I
I I
III(I II II II III
~
~
~ e ~
IllI 1
I
-100 0
10 TIME (SEC) 20 25 RGURE 9
- BF2CY6, FWGF ICF 0
Legend TOTAL REACTIVITY SCRAM REACTIVITY(0(0) t l
\\'I 1
-6 15 17 vMc (sEc)
FIGURE 10
- BF2CY6, MSIVC ICF 17 Revision 2
July 1988 600 400 Legend TOTAL POWER AVE SURFACE HEAT FLUX {XQ CORE INLET FLOW (X)
CORE INLET SUBCOOLING (Xg 200
~
~
~
so~0
~
~
~ \\ ~ ~
~
~ OO ~
~
~
0 0
2 TIME (SEC)
RGURE 'l1
- BF2CY6, MSIVC ICF 250 200 150 100 50 0
0 Itt I
II I
II Legend VESSEL PRESS RISE PSI TOTAL SIR VALVEFLOW 5X)
BYPASS VALVEFLOW (Fi)
TIME (SEC)
FIGURE 12
- BF2CY6, MSIVC ICF 18 Revision 2
July 1988 150 100 50 0
I I
1 1 ri rl'lii " l il li il i ir ii ii n
I I ill~i~~
v I
I I
I I
Legend LEVEL INCH-REF-SEP SKIR VESSEL STEAM FLOW (X)
FEEOWAIER FLOWER a/
~ ~
~ ~ ~ ~ ~
~
~
~Ftt ~
~ ~
-50 0
4 TIME (SEC)
RGURE 13
- BF2CY6, MSIVC ICF Legend TOTAL REACTIVITY SCRAM REACTIVITY(t(t)
~F~
Rll
\\l l
1 0
2.
3 TIME (SEC)
Figure 14 OLMCPR for PBXBR/BXBR/QUAD+
19 Revision 2
July 1988 GLRW08 RWE R8E LFWH 1,27 FWCF 0.2 0.4 TAU+
0.6 O.S
+SCRAM Speed Interpolation Parmater as Defined in the Technical Specifications
Figure 15 BF2CY6 Rod Drop Accident 10000 8000 6000 0
4000 Legend 0
Power 2000 0.5 2.5 Time '(sec) 3.5 (D
Y MO 00 D (0
120 Figure 16 BF2CY6 Rod Drop Accident 100 80 60 CL
~ E 40 Legend Q
Core Avera e Tem erature Rise 20 0.5 1.5 2.5 3
Time (sec) 4.5
Figure 17 BF2CY6 Rod Drop Accident Legend 0.5 1.5 2.5 3
Time (sec)
4.5 4+M 0
W C
'C MO CO 8 CO
200 Figure 18 BF2CY6 Rod Drop Accident 150 EQl 0
~o 100 CL O
50 Legend 0
Maximum Pin Enthal 0.5 1.5 2.5 3
Time (sec) gl N C I v
m 0 CO P CO
Figure 19 Decay Ratio vs. Power 0.8 Natural Circulation With Conservative Adder 105% Rod Line 0.6 0
lY D
O 0)
OA 0.2 0
0 20 40 60 80 Power (percent of 'rated) 100 120 M
C (D
U)
WO 00 g 00
Figure 20 Bundle Power Comparison: QUAD+ vs Lead Bundle 1.6 O~O~e~
1.3 1 02 0
0-g) 1.1
~~
CO 1
0.9 0.8 0.7 0.8 0
DEMO BUNDLE O
LEAD BUNDLE 3,
4 B
Cycle Exposurs (GWD/MT) 4
$0 h)
C lD Ln re Y
W 0 Co 5 CO
Figure 21 MCPR Comparison: QUAD+ vs Lead Bundle 2.8 2.8 0
DEMO BUNDLE O
LEAD BUNDLE 2.4 2.2 2
1.8 1.B 1.2 O
5=
e O
e~e 0 e
tBN BORL OPERRTINB LIMIT 1
0 1
3 4
6 B
Cycle Exposure (GWD/MT) hJ C
lb M 0 Co 5 00
Figure 22 MLHGR Comparison: QUAD+ vs Lead Bundle OPERRTING LIMIT O~~
O DESIGN GQRL
~
~ ~ ~
~
~
10 CC C9 8
0-0 DEMO BUNDLE 0
LEAD BUNDLE 3
4 6
8 Cycle Exposure (GWD/MT) 4+M 4
M O 00 5 CO
ENCLOSURE 1 PROPOSED TECHNICAL'PECIFICATIOH CHANGES BROWHS FERRY UNIT 2, CYCLE 6 BASED OH BMUS FERRY NUCLEAR PLANT RELOAD LICEHSIHG REPORT UHIT 2, CYCLE 6 TVA-RLR-002 REVISION 2
COO 3.5.I ear eat LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.5.I During steady-atate power operation, the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for each type of fuel as a function of average planar exposure shall not exceed the limiting value shown In Tables 3.5.I-l, 2, 3, and 4. If at any time during operation it ia determined by normal surveillance that the limiting value for APLHGR ia being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2)
- hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the preacribed limits.
The MAPLHGR for each type of fuel, as a function of average planar expoaure shall be determined daily during reactor operation at g 25A rated thermal power.
During steady-atate power operation, the linear heat generation rate (LHGR) of any rod in any fuel assembly at any axial location shall not exceed 13.4 kM/ft. If at any time during operation it ia determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
Zf the LHGR ia not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
The LHGR shall be checked daily during reactor fuel operation at g 25K rated thermal power.
BFN Unit 2 3.5/4.5-18
4
~ 4
~
~
Table 3.5.I-l
~ ~
hvec aBe Planar u a
/
MAPLHCR VERSUS AVERAGE PLANAR. EXPOSURE
~uel "Type~BDRB284L/~AD+:
r MAPLHGR
~W/~ft 200 1,000 5,000 10,000 15, 000 20,000 25,000 30,000 35e000 40,000 45,000
'1.2 11.3
'l1 8 12.0 pe.6;-$ '.
11.2
'b.B 10.2 9.5 8.8 Table 3.5,I-2
~ r HJQ'LHGR VERSUS hVERACE PLANAR EXII'bSURE Fuel Type:
P8DRB265H
~ errr~
r
'Lh% & A hvarage Planar 200 1,000 5,000 10,000 15,000 20,000 25,000 30,000 35,00d 40,000 MAPLHCR
~M/~t 11.5
- 11. 6 11.9 12.1 12.1
- 11. 9 11.3 10.7 10-2 9.6
'~i~ji9i
'3<YA;.
3.S/4.5 21 ILN'%N'
~ KNISH%)11 3LIS
,.8 C
0E: TT E66TW~B
MAPLHCR VERSUS AVERACE PLANAR EXPOSURE Fuel Type:
P8DRB284Z Average Planar HAPLHGR
~1cW/f
~ 4
~
200 1,000 S,OOO 10,000 15, 000 20,000 25,000 30,000 35>000 a0,000 a$,000 11.2 ll~ 2 11 '
12,0 12,0 11 '
- 11. 1 10.4 9.8 9,1 8.5 Table 3,5.'E-4 HAPLHGR VERSUS AVERACc. PLANAR EXPOSURE Fuel Type:
8DRB284L Average Planar E
assure MMd/
200 1,000 5 )000 10,000 15,000 20,000 25i000 30,000 35,000 40,000
~ I I
MAPLMCR
~lW/!t 11,2 1!.3 11.8 12.0 12.0
- 11. B 11.2 10.8 lb.2 9.$
Table 3.5.I-1 HAPLHCR VERSUS AVERAGE Pl ANAR EXPOSURE Fuel Type:
PSORB284L/QUAD+
Average Planar osure HMd/T HAPLHCR
~kM/fh 200 1,000 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40)000 45)000 ll.2 ll.3 11.8 12.0 12.0 11.8 11.2 10.8 10.2 9 '
8.8 Table 3.5.I-2 HAPL}{CR VERSUS AVERACE PLANAR EXPOSURE Fuel Type:
PBDRB265H Average Planar E
osut e MWd/T HAP( HGR
~kMr ft 200 1,000 5,000 10,000 15,000 20,000 25,000 30)000 35,000 40,000 11.5 11
~ 6 11.9
- 12. 1 12.1 11.9 11.3 10.7 10.2 9.6 3.5/4,5 21
(1.35'.0 o27 0.
)
O.1 02 0.3 0.4 O.S 0.6 0.7 TAU 0,8 0.9 Fiaure 3.5.K-t '~
MCPR Limtts for V'8 X 8+8 )(8+ QUAD+
3.5/4.S-22 gE6MRW8
The peak cladding temperature following a postulated loss-of-coolant accident ie primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly.
Since expected local variations in power distribution vithin a fuel assembly affect the calculated peak clad temperature by less than g 20'F relative to the peak temperature fox' typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit.
The limiting value for MAPLHGR is shown in Tables 3.5.I-1, 2, 3,
and 4.
The analysea supporting these limiting values are presented in Reference 1.
3.5.J
~
ar This specification assures that, the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet, densification is postulated.
The LHGR shall be checked daily during reactor operation at g 25 percent power to determine if xuel buxnup, ol'ontrol rod movement has caused changes in power distz'ibution.
For LHGR to be a limiting
'alue below 25 percent rated thermal power, the R factor vould have to be less than 0.241 which is precluded by a considerable margin vhen employing any permissible control rad pattern, 3.5.K.
At coze thermal pover levels less than or equal to 25 percent, the reactor vill be operating at minimum recirculation pump speed and the maderatox void content vill be very small.
For all designated control rod patterns vhich may be employed at this point, operating plant experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin.
With this low void content, any inadvertent core flov increase would only place operation in a more conservative mode relative to MCPR, The daily requirement for calculating MCPR above 25 percent rated thermal povex is sufficient since power distribution shifts are very slow when there have not, been significant pover or control rod changes.
The requirement for calculating MCPR when a limiting control xod pattern is approached ensures that MCPR will be known following a change in paver or pover shape (regardless of magnitude) that could place operation at a thermal limit.
3.S.L.
Operation is constrained to a maximum LHGR of 13.4 kW/ft for 8z8 fuel.
This limit is reached when core maximum fraction of limiting paver density (CMFLPD) equals 1.0.
For the case where CMFLPD exceeds the fraction of rated thermal pover, operation is permitted only at less than 100-percent rated power and only with APRM scram settings as required by Specification 3.5.L.l.
The scram trip setting and rod block txip setting are adjusted to ensure that no combination of CMFLPD and FRP will increase the LHQR transient peak beyond that alloved by the 1-percent plastic strain limit.
A 6-haur time period to achieve this condition is justified since the additional margin gained by the setdawn adjustment is above and beyond that ensured by the safety analysis.
BFN Unit 2 3.5/4.5-31
ENCLOSURE 2
DESCRIPTION, REASON AND JUSTIFICATION FOR CHANGE BROWHS FERRY NUCLEAR PLANT (BFN)
UNIT 2 Descri tion Of Chan a
c The BFN Unit 2 Technical Specifications are being updated to raElact the limits for cycle 6 operations.
The cycle 6 core loading has been changed because of the results of inspection and reconstitution of the fuel completed in July 1988.
The actual changes aee a slight adjustment in the Minimum Ceitical power Ratio (HCPR) and the addition of two Tables of Maximum Average Planar Linear Heat Generation Rata (HAPLHGR) versus average planar exposuee.
Reason For Chan a
The Hinimum Ceitical Powee Ratio as a Eunction oE scram time (figure 3.5.K-1) has changed because of the reanalysis performed to include BFN Unit 1 fuel in the Unit. 2 reload.
two HAPLHGR Tables (Tabl I
Justification Foe Chan e
The HAPLHGR for each type of fuel as a function of aveeage planae exposure is presented in tables 3.5.I-1, 2, 3, and 4.
These tables have changed because of the inclusion of a different. fuel type from BFN Unit, 1, and the pressurized and unpeessurized HAPLHGR have been separated into two tables.
Technical specification 3.5.I and the bases are changed to reflect the addition of the e 3.5.I-3 and 4).
The initial cycle 6 eeload was submitted to NRC by lattae dated August 23,
- 1984, and was approved by the issueance of BFN Technical Specification 199 dated August l9p 1986.
The cycle 6 coca loading has'changed as a result of the fuel inspection and reconstitution progeam completed in July 19S8.
The justification and safety analysis results Eor the changes ara pr'esented in TVA-RLR-002 Revision 2, July 1988, "Reload Licensing Report for Browns Peery Unit 2 Cycle 6."
A summary is presented below.
Figure 3.5;K-1 HCPR vs TAU is changed because of the eeanalysis.
The reanalysis indicated the bounding accidents are rod withdrawal aeror and generator load reject without bypass, All of the accidents and the bounding envelope aee shown in fi8uee 14 of the Reload Licensing Report.
Pour HAPLHCR figures ara required tc define tbe limits for aii fuel to be
'toiled tqt"c~c'e d.
Tba currei~s tedbnical speciPi'citi'on b'ave two oF tbeia
- Eiguees, Table 3.5.I-3 is specific to fuel type PSDRB284Z.
This fuel type was not in the initial cycle 6 fuel load but was added as a result oE the fuel inspection and reconstitution program.
Table 3.5.I-4 was added to sepaeate the peessueized (PSDRB284L/QUAD + shown in Table 3.5.I-1) and the non-peessurized fuel (SDRB284L).
The pressurized fuel allows highae axposuees.
The changes to specification 3.5.I and the Bases ara administrative in nature to refeeence the additional MAPLHGR tables,
ENCLOSURE 3
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 Desert tioa of Pc'o osad Amendment The BFN Uni.t.2 Technical specifications aee being updated to eef lect, the limits for cycle 6 operations.
The cha<<ges consist of a slight revi.sion to the Minimum Ceitical Power Ratio (MCPR) and the 'addi.tion of two Maximum Average Planar. Lineac'eat Ge<<eration Bate (MAPf HGR) versus aveeage planae exposuee tables.
Basis foe Pro osed No Si nificant Hazards Considec'ation Determination NRC has provided standards foc'etermining whether a signi.ficant, hazards consideration exi.sts as stated in 10 CFR 50.92 (c).
A peoposed amendment.
to an opeeating license involves no significant hazards considerat,iona if operation of the feei.lity in accordance with the pcoposed amendmant would not (1) involve a significant inceease in the pcobabili.ty oc consequences of an acci.dent. pceviously evaluated, oe (2) create the possi.bi.li.ty of a nev oc different lcind of accident from an accident previously evaluated, oe (3) involve a significant reduction in a margin of safety.
1.
The peoposed amendment does not involve a significant inceease in the peobsbi.lity oe consequences of an accident. peeviously evaluated.
operational transients analyzed in the Final Safety Analysis Repoct, have been reevaluated in detai.l.
The Reload Licensing, Report. for Bcowns Feeiy Unit 2, Cycle 6, Revision 2, peovi,des a summary of the limiting operating transient, stabi,lity, and selected
- accident, analyses foe the peoposed core arrangement..
The 8x8 fuel assemblies to be installed i.n the cora are <<ot, significantly different from the 8x8 fuel assemblies they are replacing.
The
'NRC staff has peeviously appc'oved the design pf the GE P8x8R assemblies as described in the GESTAR document.
(NEDO-24011-P-A-8).
The NRC staff has peeviously evaluated and approved the use of four Mestinghouse designed QUAD + demonsteation assemblies in the lov powee cegion of the coce.
The NRC staff has also appeoved the analysis methods used by TVA.
2.
The proposed amendment does not create the possibility of a nav oe different accident..
This c'eload changes the ini.ti.al conditions and/oe final co<<dition used irc the existing analyses and does not create any new accident mode.
margin of safety because the plant vill be opeeated undec the same safety limits with HCPR and HAPLHGR operating limits comparable to those cueeently established.
The Reload Li.censi<<g Repoet. provt.des a sumnmry of the limiting operating transient., stabi.lity, and selected accident analyses foc the peoposed core areangement,.
The HCPR a<<d HAPLHCR limits have been eevi.sed to assure the mac'gin of safety is maintained as demonstrated in the Reload Li.cen ing Report for Bcovns Ferry Unit 2.
Cycle 6, Revision 2.
Based on the above eeasoning, TVA has determined that the proposed amendment does not involve a significant hazards consideration.