ML18227D268
| ML18227D268 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/05/1977 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18227D268 (35) | |
Text
i1 NR+ FoAM 196 12 76)
U.S. NUCLEAR REGULATORY Co SSION NRC DISTRIBUTION FoR P."~RT 50 DOCKET MATERIAL FILE NUMBEA TOr Mr. Victor Stello FROM:
Florida Power 8 Light Company Miami, Florida Robert ED Uhri DATE OF DOCUMENT 5/5/77 DATE RECEIVED 5/lo/77 i&.ETTER PX.ORIGINAL CI COP Y DESCRIPTION QAOTO AIZ E D
~ NC LASS IF I E D PROP INPUT FOAM ENCLOSURE NUMBER OF COPIES RECEIVED t
!3., K/5"W4W Ltr, trans the following:,
Consists of the Safety Evaluation'performed for the reload of Unit-No.: 4'and the subsequent return to opergtion
~ ~ ~ ~
I PLANT NAME:
Turkey Point Unit No.
RJL (l-P)
(22 P)
O~(g PpgOVS
~~D~PDGE>
SAFETY ASSIGNED AD!
FOR ACTION/INFORMATION PR CT MANA L C--ASST PROJECT MANAGER LIC ASST INTERNALDISTR I BUTION REG FETE
'NRI PDR I&E OELD GOSSXCK & STAFF MIPC HARLESS PROJECT MANAGEMENT BOYD P ~ COLLXNS HOUSTON PETERSON MELTZ HELTEMES SKOVHOLT LPDR0 TIC:
NSIC:
ASLB0 ACRS CYS HERBXNC/ E SYSTEMS SAFETY HEINEMAN SCHROEDER ENGINEERING MACARRY BOSNAK SIHWEIL PAWL CK REACTOR SAFE ROSS NOVAK ROSZTOCZY CHECK AT& I SALTZMAN RUTBERG EXTERNAL DISTRIBUTION NAT LAB ~
PLANT SYSTEMS TEDESCO N
XPPOLITO OPERATING REACTORS STELLO OPERATING TECH EISENHUT "B
E 00 ULR KSON QR S
SA ERNST BALLARD YOUNGBLOOD S'XTE TECH GAMMXLL STEPP HULMAN SITE ANALYSXS VOLLMER BUNCH J
COLLINS KREGER CONTROL NUMBER 77l3i0038 NAC FORM 195 (2-7SI
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FLORIDAPOWER 8( LIGHT COMPANY May 5, 1977 L-77-139 8'egu1atory Doc)<et File Office of Nuclear Reactor Regulation Attention:
Mr. Victor Stello, Director Division'of Operating Reactors U.
S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Stello:
Re:
Turkey Point Unit 4 Docket No. 50-251 Cycle 4 Information 9~SO QEovEO g~ZO$77 5 ggggAIOXfj
~1~~1~e~
gg~
Attached herewith is the Safety Evaluation performed for the reload of Turkey Point Unit 4 and the subsequent return to operation.
This report is being forwarded to you for your information.
Very truly yours, Robert
. Uhrig Vice President REU/MAS/cpc Attachment cc:
Mr. Norman C. Moseley, Region II Robert Lowenstein, Esquire PEOPLE... SERVING PEOPLE
l 1
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I '
Turke Point Unit 4
CYCLE 4
RELOAD SAFETY EVALUATION
TABLE OF CONTENTS Title
~Pa e
1.0 INTRODUCTION
AND SUt"t@RY 2.0 REACTOR DESIGN, 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Hydraulic Design 2
2 3
3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Capability 3.2 Accident Evaluation 3.3 Incidents Reanalyzed 3.4 ECCS Evaluation 7
8
- 4. 0 I REFERENCES io[
l t
LIST OF TASLES Table Title Fuel Assembly Design Parameters Kinetics Characteristics Shutdown Requirements and Margins Rod Ejection Parameters Results of Rod Ejection Analyses Hot Spot Clad and Fuel Temperatures
~
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11
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LIST OF FIGURES 0
Title Loading Pattern Source and Burnab1e Poison Locations Trip Reactivity versus Rod Position
.0 INTRODUCTION AND
SUMMARY
Turkey Point Unit 4 is in its third cycle of operation.
The unit will refuel and be ready for Cycle 4 startup in June, 1977.
The Turkey Point Unit 4 Cycle 4 core loading pattern is shown as Figure 1.
Fifty-nine fuel assemblies will be discharged from Cycle 3, and replaced by forty fresh assemblies and nineteen previously burned assemblies.
The two Region 4 removable rod assemblies which are in Cycle 3 will remain in the core unaltered for Cycle 4.
The Cycle 4
fuel inventory is given in Table 1.
Depleted borosilicate burnable poison rods will be used in Cycle 4.
The location of these. rods is shown in Figure 2.
This report presents an evaluation for Cycle 4 which demonstrates that the core reload will not adversely affect the safety of the plant.
It is not the purpose of this report to present a reanalysis of all poten-
. tial incidents.
Those'ncidents analyzed and reported in the FSAR which could potentially be affected by fuel reload have been reviewed for the Cycle 4 design described herein.
The results of new analyses have been
- included, and the justification for the applicability of previous re-sults from the remaining analyses is presented.
It has been concluded that the Cycle 4 design does not cause the previously acceptable safety limits for any incident to be exceeded.
This conclusion is based on the assumption that:
(1) Cycle 3 operation is terminated after 7750 +
1 ppp HWD/NTU (2) Cycle 4 burnup is 1 imited to the end-of-ful 1 300 power capability*, and (3) there is adherence to plant operating limita-tions contained in the Technical Specifications.
Nominal design parameters for Cycle 4 are 2200 NMt core power, 2250 psia system pressure, 546.2'F core inlet temperature, and 5.58 kw/ft average linear fuel power density (based on 144" active fuel length).
Definition:
Full rated power and temperature (approximately 575'p Tarp)>>
control rods fully withdrawn, and zero ppm residual boron.
2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of Region 6 fuel assemblies is the same as Region 5
except for hold down provisions.
The fuel assembly hold down. sprincs and top nozzle were modified to provide additional hold down force, and to standardize on the same design as used in larger Westinghouse cores.
The fuel rod design'for Region 6 is mechanically the same as for Region 5.
Differhnces exist in enrichment between Region 5 and Region 6, as shown in Table 1-~
R Clad flattening time is predicted to be 31,200 EFPH for the limiting region (Region 3) using the current Westinghouse Evaluation Yodel Therefore, Region 3 has a nominal allowed Cycle 4 residence time of 8,500 EFPH.
This is based on a cumulative residence time for Cycles 1, 2, and 3 of 22,700 EFPH.
Expected Cycle 4 lifetime is 7000 EFPH.
Westinghouse has had considerable experience with Zircaloy clad fuel.
This experience is extensively described in WCAP-8183, "Operational Experience with Westinghouse Cores This report is updated about every six months.
- 2. 2 NUCLEAR. DESIGN The Cycle 4 loading pattern results in a maximum F
less than 2.22 under nor-mal operating conditions.
Table 2 provides a comparison of the range of values encompassing the Cycle 4 core kinetics parameters with the current limit based on previously submitted accident analyses.
It can be seen from the table that most of the Cycle 4 range of values fall within the current limits.
These parameter s are evaluated in Section 3.0.
Table 3 provides the control rod worths and requirements.
The required shutdown margin is based on previously submitted accident analysis The available shutdown margin exceeds the minimum required.
The trip reactivity insertion rate for Cycle 4 is slower than the one t
used in previous cycles (see Section 3.3 and Figure 3}.
The reactivity insertion rate is different because the combined bank worth as a func-tion of time (axial location) has changed.
The reactivity insertion rate for Cycle 4 was calculated by a very conservative method that produces a flux distribution skewed towards the bottom of the core.
This reduces the reactivity worth of the banks at the top of the core
'elative to the total worth.
Such a calculation provides a conserva-tive trip reactivity shape for accident analysis sin'ce the axial flux distribution is normally distributed evenly with constant axial 'offset control.
2.3 THERMAL AND HYDRAULIC DESIGN No significant variations in thermal margins will result from the Cycle 4 reload.
The present DNB core limits have been found to be conserva-tive.
3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY This section reviews the plant power capabil ity considering the consequences of those incidents examined in the FSAR using the previou'sly accepted design bases.
It is concluded that the core reload will not adversely affect the ability to safely operate at 100/ rated power during Cycle 4.
For the overpower transient the fuel centerline temperature limit of 4700 F can be accommodated with margin in the Cycle 4 core.
The time dependent densification model was used for this evaluation.
The LOCA limit is met by maintaining F~ at or below 2.22 with less than 10/ steam generator tube plugging'.
. 3.2 ACCIDENT EVALuATION The effects of the reload on the design basis and postulated iiicidents analyzed in the FSAR have been examined.
In most cases it was found that the effects can be accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis.
For those incidents which were reanalyzed, it was determined that the applicable design basis limits are not exceeded, and therefore, the conclusions presented in the FSAR are still valid.
A reload can typically affect accident analysis input parameters in three major areas:
kinetics character istics, control rod worths, and core peaking factors.
Cycle 4 parameters in each of these three areas were examined as discussed below to ascertain whether new accident analyses are required.
I
Kinetics Parameters A comparison of tHe range of values 'encompassing the Cycle 4 kinetics parameters with the current limits is given in Table 2.
Most of the range of values remain within the bounds of the current limits.
The moderator temperature coefficient will be zero or negative during normal operation.
With the exception of delayed neutron fractions, the small changes in core physics parameters have a negligible effect on transient analysis.
The minimum delayed neutron fractions, 8, for the beginning and end of life Cycle 4 are outside the current limits.
This will significantly affect only the rod ejection transients.
This is discussed in Section 3.3.
Control Rod Morths Changes in control rod worths may affect shutdown margin, differential rod worths, ejected rod worths, and trip reactivity.
Table 3 shows that the Cycle 4 shutdown 'margin requirements are satisfied.
As shown in Table 2, the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 4 is less than or equal to the current limit.-
Cycle 4 ejected rod worths are given in Table 4.
Since some ejected rod worths are greater than the corresponding current limits, these cases were reanalyzed as discussed in Section 3.3.
Cycle 4 has a slower trip reactivity insertion rate than Cycle 1;
- however, the total trip reactivity is significantly greater than the value. assumed in Cycle 1 (Figure 3).
The effects of this reduced reac-tivity trip rate have been evaluated for those accidents affected, and compared with the Cycle 1 analyses.
Slow transients are relatively insensitive to chanqes in trip reactivity insertion rate, and therefore need not be reanalyzed.due to the change in trip reactivity versus rod position.
Fast transients such as rod ejection and rod withdrawal from 0
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.subcritical, in which negative reactivity inset tion is due primarily to Doppler feedback, will be unaffected by the change in trip reactivity since the transient is essentially turned around before rod insertion starts.
The effect of variations in trip reactivity insertion rates for the rod withdrawal at power incident has been investigated.
The results of this. analysis show that the minimum ONBR is unaffected since the minimum DNBR for the transient occurs at relatively low reactivity insertion rates.
For the loss of flow and locked rotor transients the change in trip reactivity versus rod insertion will result in a slightly higl ~r.tran-sient heat flux.
Since the minimum DHB ratios for these transients are
= sensitive to heat flux relative to flow., these incidents were reanalyzed.
The results of these calculations are discussed. in Section 3.3.
Core Peakin Factors Peaking factors following control rod ejection are within the bounds of the current limits.
Evaluation of peaking factors for the rod out of position and dropped RCCA incidents shows that DNBR is maintained above 1.30.
For the dropped bank incident the turbine runback setpoint is sufficient to prevent a
DHBR lass than 1.30.
The steamline rupture analysis has been examined for Cycle 4.
A review of the hypothetical break cases with and without a loss of offsite power demonstrated that the cases with offsite power available were still limiting.
Statepoints for the most limiting transient (the hypothetical break inside containment with offsite power available) were analyzed, and the minimum DNBR was found to remain above 1.30.
For the case with a break area equivalent to a steam generator safety or dump valve (cred-ible break) it was confirmed that the core did not return to criticality.
- 3. 3 INCIDENTS REANALYZED Rod Ejection For the beginning and end of life hot full power rod ejection cases, maximum ejected rod worths are greater and minimum delayed neutron fractions are less than the current limits.
These cases were reanalyzed to ensure that the fuel and clad'imits were not exceeded.
In addition, the beginning and end of life hot zero power cases were reanalyzed with a reduced delay neutron fraction as shown in Table 4.
The analysis was performed using the same methods as described in Reference 5.
The results are given in Table 5.
The results show that the fuel rod at the hot spot does not exceed the limiting criteria; thus the conclusions presented in Reference 5 are still valid.
Loss of Flow and Locked Rotor The complete loss of flow and locked rotor transients were reanalyzed due to a slower trip reactivity insertion rate.
A comparison of the FSAR and Cycle 4 tr ip'eactivity versus rod position values are shown in Figure 3
along with the conservative values assumed in the reanalysis.
As noted in Section 3.2, the assumed total trip worth has been increased from a value of 2.84 to 4.05 hk/k which is still conservative for Cycle 4.
The calculations were performed using the same methods used for Cycle 1
For the complete 3/3 pump loss of flow incident, the minimum DNB ratio does not fall below the value of 1.30.
The~other cases described in the FSAR incident show larger DHB ratios than the 3/3 pump incident.
Thus it is concluded that all cases would remain above 1.30 and the conclusions as presented in the FSAR for this incident are still valid.
For the two and three loop locked rotor cases, no rods are expected to experience DAB; however, the hot channel quality at the location of the minimum DNB ratio exceeds the range of quality over which the DNB cor-
"relation was derived.
For this reason, DNB was conservatively assumed to occur.
The three loop case is most limiting from a peak clad tem-perature standpoint:
The results show a peak clad temperatur e of 1500'F which is well below the limiting value of 2700'F.
The amount of Zr-H20 reaction is small (less than 1X by weight) and contributes less than 10'F to the peak cladding temperature.
The two loop case is most limiting from a peak pressure stand~nint.
The results show a peak system pressure of less than 2?20 psia which is below the value which would cause stresses to exceed the faulted con-dition stress limits of the primary coolant system.
Thus, the conclu-sions as presented in the FSAR are still valid for Cycle 4.
3.4 ECCS Evaluation The LOCA limit is met by maintaining F~xPREL at or below 2.22.
The fuel in Turkey Point Unit 4, Cycle 4 core has a 2.22 LOCA limit, with the exception of three Region 3 assemblies located in the posi-tions H-07, H-08 and H-09 (see Figure 1).
These Region 3 assemblies were discharged from Cycle 1 with low burnup rods and are more limiting due to higher fuel temperatures than the other regions of fuel.
A LOCA limit of 2.13 was determined for these three Region 3 assemblies.
These Region 3 assemblies will meet their LOCA limit if the remainder of the core meets the 2.22 LOCA limit because of the following reasons:
1.
The limiting elevation dependent peak-to-average powers, Fry(z),
do not occur in these Region 3 assemblies during Cycle 4 operation.
It has been determined from 3D calculations that at each elevation, the ratio:
Pxy(z)/Fxy(z)
< 2.13/2 22 is met where Pxy(z) is defined as:
Pxy(z)
= the peak power in the three Region 3 assemblies at elevation z to the core average power at elevation z.
2.
The analysis showed about 8Ã margin to the 2.13 LOCA limit for these Region 3 assemblies using Pxy(z) from the 3D calculations.
3.
Further note that the three Region 3 assemblies are adjacent to or under the 0 bank control rod at H-08; hence, the most limiting Pxy(z) for these assemblies is with the control rods fully withdrawn.
Therefore, full power capability based on the F *P L envelope of 2.22 g* REL is possible throughout Cycle 4 operation of Turkey Point Unit 4.
5.0 REFERENCES 1.
George, R.'.,'et.
al "Revised Clad Flattening Model," >ICAP-8377
~ (Proprietary) and tFlCAP-8381 (Non-Proprietary), July 1974.
2.
Schreiber, R. E., Iorii, J. A., "Operational Experience with Westinghouse Cores,"
WCAP-8183, Revision 5, September, 1976.
3.
Final Safety Analysis Report, Turkey Point Units Nos.
3 and 4..
4.
Hellman, J.
M. (Ed.),
"Fuel Densification Experimental Results and Model for Reactor Operation,"
MCAP-8218-P-A, March 1975 (Proprietary) and MCAP-8219-A, March 1975 (Non-Proprietary).
5.
"Fuel Densification, Turkey Point Plant Unit No. 3," WCAP-8074 (Proprietary) and 1JCAP-8075 (Non-Proprietary),
February 1973.
6.
- Risher, D. H., Jr.
"An Evaluation of the Rod Ejection Accident in W PNRs using Spatial Kinetics Methods",
NCAP-7588, Rev-lA (Non-Proprietary),
- January, 1975.
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TABLE 1
'urke Point 4 - " cle 4 Fuel Assembl Desi n Parameters
~Re ion 6-1 6-2 Enrichment '(w/o U-235) 2,56.
3.11
- 2. 55 3.00
- 2. 90 3.10 Density (5 Theoretical)*
92.9 92.8 94.3 94.7 96.0 95.0 Number of Assemblies 16 52 40 16 24
'pproximate Burnup at Beginning of Cycle 4 (WD/NTu) 24000 20400 15500 6600 0
0
- All regions except Region 6 are as-built values; Region 6 is the nominal
- value, however, an average density of 94.5X theoretical was used in thermal evaluations.
TABLE 2 Turke Point Unit 4 Kinetics Characteristics Current Limit'
'Cele 3
(3 5)
~Cele e
Moderator Temperature Coefficient, (hp/'F) x 10
-3.5 to 0.3*
-3.5 to 0
-3.5 to 0.2*
Doppler Coefficient (hp/ F) x 10
-1.6 to -1.0
-2.6 to -1.0
-2.6 to -1.0 Delayed Neutron Fraction jeff (X)
.50 to.72
.50 to.72
.49 to
. 72 Prompt Neutron Lifetime (pSec) 14 to 18 20 20.1 Maximum Differential Rod Morth of Two Banks Moving Together at HZP (pcm/in)**
80 80 80 The positive coefficient does not occur at operating conditions.
pcm = 10 hp
TABLE 3 Turke Point 4 -
C cles 3 'and 4
Shutdown Requiremenrs and Mar ins C~c1e 3
~Cele 4
BOC EOC Control Rod Worth
',~ ap All Rods inserted less Worst Stuck Rod 5.86 5.84 (1) Less 10/
- 5. 27 5.25
- 5. 64 5.89
- 5. OB 5.30 Control Rod Re uirements 5 zp Reactivity Defects (Doppler: Tavg, Void, Redistribution)
Rod Insertion Allowance (2) Total Requirements
.50 50'.
23
- 3. 21
- 1. 73
- 2. 71
- 1. 76
- 2. 69
.70
.70 2.46 3.39
~))td M
)
))-2)
)) j 3.04 2.04 2.62 1.91 Required Shutdown Margin (i! Lo)
- 1. 00 1. 77 1
00
'77
TABLE 4 Rod E'ection Parameters Current Cycle Used In HZP -
BOC Max. Ejected Rod Worth, Khp Max.
FN jeff
.71
.48
- 7. 0 6.1
.0070
.0056
.71 7.0
..0050 HFP -
BOC Max. Ejected Rod Worth, Alp Max.
Fq eff Initial Fuel Avg. TemPerature,
'F
.27
. 31 5.48 4.66
.0070
.0056 3168 2625
.35 5.48
.0050 2660
.HZP -
EOC Max. Ejected Rod Worth, Ahp Max.
FN eff
.84
.37 14.3 5.48
.0050
.0049
.84
- 14. 3
.0044 HFP -
EOC Max. Ejected Rod Worth, Xhp Max.
FN jeff Initial Fuel Avg. Temperature,
'F
.26
.27 5.52 4.42
.0050 -'0049 2473 2450
.30 5.52
.0044 2475
TABLE 5 Results of Rod Ejection Anal sis Hot S ot Clad and Fuel Tem eratures Limiting Value BOC BOC EOC EOC (6)
Initial Power, I 102 0
102 Fuel Pellet Melting,.%
10
<10 0
<10 Maximum Fuel Average TemperatureF)
Maximum Fuel Enthalpy (Cal/gm) 200 Maximum Clad Average Temperatur e ('F) 2700 2040 1580 80 3990
.3080 3490 2370 2270 2125
'73 128 148
FIGVRE 1
TURKEY POINT UNIT 4 CYCLE 4 LOADING PATTERN R
P N
H L
K J
H G
F E
0 C
B A
6-2 6-2 6-6-1 4
6-1 6-2 2*
6-2 6-1 4
6-1 6-2 4***
4 5
4 5
4 6-2 6-1 5
5 2*
5 4
5 6-1 6-2 4
4 4
5 4
5 6-2 6-2 6-1 3
4 4
5 5
4 3**
3 5
2*
4 5
5 4
5 6-1 6-2 6-2 6-1 4
5 3**
4 5
2*
6-1 6-2 6-2 6-1 5
4 5
5 2*
5 4
5 6-1 6-2 4***
6-2
'-'2 6-1 5
4 4
6-1 6-2 6-2 6-1 4
6-1 6-2 2*
6-2 3
6-2 FUEL REGION Region 2 From Cycle 2
Region 3 From Cycle 1
Removable f!od 4ssembly
I
~ t i
- ~
1
~
FIGURE 2 TURKEY POINT UNIT 4 CYCLE 4
. SOURCE AND BURNABLE POISON LOCATIONS P,
P N
M L
K 0
H 6
F E
0 C
B aA 12 BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 I
12 BP's 12
'P's 12 BP's 12 BP's S
- Source Location BP's
- Depleted Burnable Poisons
~ '4 "[
r
~
~
~-
~ '
FIGURE 3 TURKEY POINT UNIT 4 CYCLE 4 TRIP REACTIVITY VERSUS'~ROD POSITION 5.0 4.0 3.0 2.0 1.0 I-
'0
[0
"'0 fO 40 60 60 70 80 90 100
.X.Rod:Inser tion Cur rent'lmi.t Cycl'e 4l'Value R
1y. i If 1
~
~
0