ML19310A322
| ML19310A322 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/30/1980 |
| From: | Baker H BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19310A318 | List: |
| References | |
| BAW-1067, NUDOCS 8006110085 | |
| Download: ML19310A322 (11) | |
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- THE BABC0CK & WILCOX COMPANY POWER GENERATION GROUP To l R. J. Trost, Project Management I A i H. A. Baker, Project Engineering sos m.5 Cust.
Florida Power Corp. /' File No.
or Ref. FPC- Reload Subj. Date CR-3, Cycle 3 Reload Report Amendment May 30, 1980 l m . . .. . . , . . . . . . , . . . . . . . . . . , . . . . . . . w . ;r . . %
During the recent refueling at Crystal River 3, a broken holddown spring was discovered in assembly NJ018E, a batch 4 assembly that had been in core location N-14 during cycle 2. This assembly and three symmetric assemblies have been replaced with batch 5 fuel. Thus, fifty-two batch 2 assemblies I and four batch 4 assemblies have been discharged and fifty-six batch 5 assemblies will be loaded for cycle 3. The necessary changes to the Crystal River Unit 3 Cycle 3 Reload Report BAW-1607, Rev.1, April,1980 are attached.
Figures 3-1, 3-2, Table 5-2, and Figure 5-1, have been updated to reflect the changes in the core loading. With the exception of the stuck rod worth at BOC and E0C, as shown in the revised Table 5-2, all values quoted in Table I 5-1 of Reference 1 change by < 1%.
insignificant changes.
It was not deemed necessary to make these l The safetv, control and pwr nea. king analyses performed previously, with and without pump monitors, at power levels of 2452 MWt and 2544 MWt and forwarded to FPC by letter on May 6,1980 have been evaluated and remain valid. No changes to the technical specifications are required by the modified core loading.
In addition, two corrections have been included as the result of discussions with Florida Power. In paragraph 7.12, page 7-8, the cycle 3 predicted i value of maximum rod worth was corrected to 0.49%Ak/k from the 0.59%Ak/k l value. In Table 7-1, on page 7-13, the Dropped Rod Worth was corrected l from 0.65 to 0.40.
Due to the small number of pages effected by this change, we have elected to revise only the necessary pages and not rev the entire report. The revised pages have been noted. Those on distribution of this memo should incorporate the attached pages in their copy of the reload report.
- 1. d A &_ J Safety Analysis l
FRhj '/ Thermo-Hydraulic Engineering
&V,:fW M ECCS W kZ//21K Mar.,J Assato s
- *A FA L wi Nuclear Operations Analysis l M AA BI 2., mac Fuel Management and Development I 9A. L.'& L FMD _
'fHw) MM' ws t 7v HAB:dcr l 8006110075 l
The current and potential transformers are not seismically qualified. However, separation of the cables carrying redundant transformer outputs to the RCPPM cabinets is provided in accordance with the separation criteria stated above.
The current and potential transformers are not seismically qualified because they are not required to safely shutdown the reactor. The loss of the current or potential transformers would result in a " pump inoperable" signal to the RPS. Upon receipt of two such signals, whatever the cause, the REs trips the reactor.
3.2. Core Description The CR-3 reactor core is described in detail in Chapter 3 of the Final Safety Analysis Report for the unit.1 The cycle 3 core consists of 177 fuel assem-blies (FAs), each of which is a 15-by-15 array containing 208 fuel rods; 16 control rod guide tubes; and one incore instrument guide tube. The fuel as-semblies in batches 2, 3, and 5 have an average nominal fuel loading of 463.6 kg of uranium, whereas the batch 4 assemblies maintain an average nominal fuel loading of 468.6 kg of uranium. The cladding is cold-worked Zircaloy-4 with an OD of 0.430 inch and a wall thickness of 0.0265 inch. The fuel consists of dished-end, cylindrical pellets of uranium dioxide (see Table 4-2 for data).
Figure 3-1 is the core loading diagram for cycle 3 of Crystal River 3, The g initial enrichments of batches 2, 3, and 4 were 2.54, 2.83, and 2.64 wt %
uranium-235, respectively. Fifty-two batch 2 and four batch 4 assemblies will be discharged at the end of cycle 2. The batch 5 design enrichment is 2.62 wt % uranium-235. Batches 3 and 4 and the remaining batch 2 assemblies will be shuffled to new locations. The batch 5 assemblies will occupy the periph-ery of the core. Figure 3-2 is an eighth-core map showing the burnup of each assembly at the beginning of cycle 3 and its initial enrichment.
Care reactivity will be controlled by 61 full-length Ag-In-Cd control rod as-semblies (CRAs) and soluble boron shim. In addition to the full-length CRAs, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The cycle 3 locations of the 69 control rods and the group designations are unchanged from cycle 2 and are shown in Figure 3-3. Control rod group 7 will be withdrawn at 250 10 EFPD of operation.
Babcock & Wilcox 3-3 Revised 5/30/80
l l
Figure 3-1. Core Londing Diagram for Crysta.' River 3, Cycle 3 A 5 5 5 5 5 F7 C9 F9 XX Cycle 2 Location 5 5 5 3 3 3 Y Batch Nuznber c D7 N3 L1 P8 L15 N13 D9 5 5 5 5 3 4 4 2 4 4 3 D 5 5 C7 03 M2 D5 R8 D11 M14 Cl3 G13 4 5 5 3 4 3 4 3 4 4 3 G4 B12 Fo K5 K1 L8 K15 K11 F10 B4 G12 5 E 5 3 4 2 3 4 3 4 3 2 4 3 F C12 Bll E9 E5 D6 B10 D10 E11 E7 B5 C4 5 5 5 5 4 4 3 3 3 4 3 3 3 4 4 G C6 A10 E4 A9 F4 B6 D8 F14 F12 A7 E12 A6 G10 5 5 3 4 3 4 3 4 3 4 3 4 3 4 3 G3 H14 H15 H10 F2 H4 H8 H12 L14 H6 H1 H2 K13 H 5 5 3 2 4 3 4 3 2 3 4 3 4 2 3 K6 R10 M4 R9 L4 L2 N8 P10 L12 R7 M12 R6 K10 K 5 5 3 4 3 4 3 4 3 4 3 4 3 4 3 L 012 P11 M9 M5 N6 P6 N10 M11 M7 PS 04 5 5 5 5 4 4 3 3 3 4 3 3 3 4 4 K4 P12 L6 G5 G1 F8 GIS Gil L10 P4 K12 M 5 5 3 4 2 3 4 3 4 3 2 4 3 N K3 C3 E2 N5 A8 N11 E14 Cl3 09 5 5 5 5 3 4 4 3 4 3 4 4 3 0 5 5 N7 D3 F1 B8 FIS D13 N9 3 4 4 2 4 4 5 5 3
L7 07 L9 P S 5 5 5 5 5 3 3 3 R S 5 5 5 5 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 I
I 3-4 Babcock & Wilcox Revised 5/30/80
I Figure 3-2. Enrichment and Burnup Distribution for Crystal River 3, Cycle 3
- 8 9 10 11 12 13 14 15 y 2.54 2.83 2.64 2.83 2.64 2.54 2.83 2.62 17,015 16,512 5,691 19,762 3,085 16,741 15,949 0 2.64 2.83 2.64 2.83 2.64 2.83 2.62 K
5,690 12,950 3,452 17,364 0 3,206 15,308 2.83 2.83 2.64 2.64 2.62 2.62 L
14,095 13,590 4,923 6,051 0 0 2.54 *2.64 2.83 2.62 1 g
'g <
17,460 3,639 17,466 0
- 2.64 2.83 2.62 2.62 15,950 0 0 2.64 0
0 P
R x.xx Initial enrichment xx,xxx BOC burnup, mwd /mtU 1
- Signifies asymmetric (1/4 core) shuffled assemblies.
l I
Babcock a.Wilcox 3-5 Revised 5/30/80 l
Figure 3-3. Control Rod Locations I
@ @ @ @ I
@ @ @ @ @ g
@ l
@ @ @ @ @ @ @ l
@ Q @ @ @ @ Q I O @ @ @ @ @ l O @ O O l l
GROUP NUMBER OF RODS FUNCTION l 8 SAFETY 2 8 SAFETY 3 12 SAFETY 4 9 SAFETY g 5 8 CONTROL g 6 8 CONTROL 7 8 CONTROL E
8 8 APSRs u TOTAL 69 I
I 3-6 Babcock & Wilcox
Table 5-2. Shutdown Margin Calculation for Crystal River 3, Cycle 3 BOC, %Ak/k EOC(* , %Ak/k Available Rod Worth Total rod worth, HZP(b) 9.25 9.21 Worth reduction due to burnup of poison material -0.37 -0.42 Maximum stuck rod worth, HZP -1.82 -1.79 Net worth 7.06 7.00 Less 10% uncertainty -0.71 -0.70 Total available worth 6.35 6.30 Required Rod Worth I Power deficit, HFP to HZP 1.30 2.08 1
Max allowable inserted rsd worth 1.06 1.36 Flux redistribution 0.53 1.02 Total required worth 2.89 4.46 Shutdown Margin Total available minus total required 3.46 1.84 Note: Required shutdown margin is 1.00% Ak/k.
(#
I For shutdown margin calculations, this is defined as N250 EFPD, the latest time in core life in which the transient bank is nearly full-in.
(b)HZP: hot zero power, HFP: hot full power.
t 5-5 Bat ' Wilcox Rev: n30/80
i
'a Figure 5-1. BOC (4 EFPD), Cycle 3 Two-Dimensional Relative Power Distribution - HFP, Equilibrium Xenon, J i
Banks 7 and 8 Inserted I 8 9 10 11 '~_ 13 14 15 l 7 l H 1.08 1.14 1.30 1.17 ,
1.34 0.94 0.46 0.50 )
K 1.27 1.14 1.31 1.16 1.22 0.82 0.57 I' 7 8 L 0.67 1.03 1.15 1.24 1.08 0.54
- l 1
1 M 1.01 *l.28 1.04 0.89 N
- 1.25 1.11 1.10 0.62 i
O i 0.73 '
P l
R I
Inserted rod group No.
I x.xx Relative power density
- Denotes only impact of 1/4-core symmetry greater than 0.01.
l 5-6 Babcock & Wilcox Revised 5/30/80
i l :
The FSAR has identified a double-ended rupture of the steam line between the steam generator and steam stop valve as the w)rst-ca.e situation at end-of-life conditions.
The key parameter for the core response is the moderator temperature coeffi-cient which in the FSAR was assumed to be -3.0 x 10 4 Ak/k/*F. The cycle 3 predicted value of moderator temperature coefficient is -2.63 x 10-4 Ak/k/*F.
This value is bounded by that used in the FSAR analysis; hence, the results in the FSAR represent the worst situation.
The FSAR used an initial power level of 2568 Wt for these accidents. This is more conservative than running the accident at 102% of 2544 wt and tripping the reactor at 110% versus the current 112% setpoint since more energy is added to the system for the FSAR analysis.
7.10. Steam Generator Tube Failure A rupture or leak in a ; team generator tube allows reactor coolant and associ-ated activity to pass to the secondary system. The FSAR analysis is based on complete severance of a steam generator tube. The primary concern for this incident is the potential radiological release. The environmental dose assess-ment is presented in section 7.18.
7.11. Fuel Handling Accident The mechanical damage type of accident is considered the maximum potential source of activity release during fuel handling activity. The primary con-cern is over radiological releases. The environmental dose assessment is pre-sented in section 7.18.
7.12. Rod Ejection Accident For reactivity to be added to the core at a more rapid rate than by uncontrolled rod withdrawal, physical failure of a pressure barrier component in the CRDA must occur. Such a failure could cause a pressure differential to act on a CRA and rapidly eject the ' assembly frc,m the core. This incident represents the most rapid reactivity insertion that can be reasonably postulated. The I values used in the FSAR and densification report at BOL conditions of -1.17 x 10-5 Ak/k/*F Doppler coefficient, 0.0 Ah/k/*F moderator temperature coe?ficient, and ejected rod worth of 0.65% Ak/k represented the maximum possible transient.
7-7 Babcock & Wilcox
I The use of a 0.65% Ak/k maximum rod worth is conservative in comparison to the cycle 3 predicted value of 0.49% Ak/k. Furthermore, the cycle 3 predicted 1
values of -1.52 x 10-5 Ak/k/*F Doppler and -0.30 x 10-5 Ak/k/*F moderator tem-perature coefficient are both more negative than used in the FSAR analysis.
The FSAR used an initial rated power level of 2568 MWt for this accident. This is more conservative than initializing the accident at 102% of 2544 MWt and tripping the reactor at 110% versus the current 112% setpoint since more energy is added to the system for the FSAR analysis. For the accident which trip on high pressure, the effect of higher initial power level (i.e., 102% of 2544 MWt) is to cause the pressure trip to occur slightly sooner. Since the FSAR input bound the cycle 3 predicted values, the results in the FSAR and densifi-cation report are applicable to this reload.
7.13. Maximum Hypothetical Accident There is no postulated mechanism whereby this accident can occur since this I would require a multitude of failures in the engineered safeguards. The hypo- g thetical accident is based solely on a gross release of radioactivity to the E reactor building. The environmental dose assessment is presented in section 7.18.
7.14. Waste Gas Tank Rupture The waste gas tank was assumed to contain the gaseous activity evolved from I degassing all the reactor coolant following operation with 1% defective fuel.
g Rupture of the tank would result in the release of its radioactive contents E to the plant ventilation system and to the atmosphere through the unit vent.
The environmental dose assessment is presented in section 7.18.
7.15. LOCA Analysis Generic LOCA analyses for B&W 177-FA lowered-loop NSSs have been performed I; using the Final Acceptance Criteria ECCS Evaluation Model. The large-break l analysis is presented in~a topical report 13, and is further substantiated in a letter reportl . Tha small break analysis is presented in a letter report 15, l These analyses used the limiting values of key parameters for all plants in the l category. Furthermore, the average fuel temperature as a function of linear heat rate and lifetime pin pressure data used in the LOCA limits analysis 13 are conservative compared to those calculated for this reload. Thus, these j 2
7-8 Dabcock & Wilcox ,
Revised 5/30/80 l
Table 7-1. Comparison of Key Parameters for Accident Analysis FSARI ,
densif'n Cycle 3 Parameter value7 Cycle I II value I BOL Doppler coeff, 10-5 Ak/k/*F -1.17 -1.47 (268 EFPD)
-1.52 EOL Doppler coeff, 10-5 Ak/k/*F -1.30 -1.66 -1.61 (510 EFPD)
BOL moderator coeff, 10-4 Ak/k/*F 0(*) -0.75 -0.30 (268 EFPD)
EOL moderator coeff, 10-4 Ak/k/*F -4.0(b) -2.42 -2.63 (510 EFPD)
All-rod bank worth at BOL, HZP, 12.9 9.12 9.37 1
% Ak/k (268 EFPD)
Boron reactivity worth (HFP), 100 101 108 ppm /1% Ak/k I Max ejected rod worth (HFP), % Ak/k 0.65 0.55 0.49 Dropped rod worth (HFP), % Ak/k 0.40 0.20 0.20 Initial boron conc'n (HFP), ppm 1150 795 1185
(*}+0.50 x 10-4 Ak/k/*F was used for the moderator dilution accident.
} 3.0 x 10-4 Ak/k/*F was used for the steam line failure analysis and dropped rod accident analysis.
Table 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates I Core Allowable elevation, peak LHR, ft kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 I
7-13 Babcock & Wilcox Revised 5/30/80
~
Table 7-3. Input Parameters to Loss-of-Coolant-Flow Transients ;
Cycle 3 value Value used in analysis Initial flow rate, % >109.5 106.5 of 352,000 gpm ,
Flow rate Vs time > Fig. 14-17, FSAR Fig. 14-17, FSAR (4PCD)
Fig.14-19a, FSAR Fig.14-19a, FSAR (LR)
Initial power level, 2544 102% of 2568 MW Doppler coeff, Ak/k/ F -1.52 x 10-5 -1.27 x 10-5 Moderator temp coeff, -0.30 x 10-4 0 ok/k/'F FAH 1.47 1.71 I
Table 7-4. Summary of Minimum DNBR Results for Limitir.g Loss-of-Coolant-Flow Transients Cycle 1 Densif'n FSARI report Cycle 2 Cycle 3 Transient (W-3) (W-3) (
_B&W-2) (B&W-2)
One-pump coastdown (flux / flow NR
- NR 1.75 1.75 trip)
Four-pump coastdown (flux / flow 1.45 1.39 2.10 2.10 trip, cycle 1; pump monitor trip, cycles 2 and 3)
NR: not reported.
4 l 7-14 Babcock & Wilcox