ML20214P435
| ML20214P435 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/26/1987 |
| From: | Mihalcik J BALTIMORE GAS & ELECTRIC CO. |
| To: | Mcneil S NRC |
| References | |
| FCM-87-117, NUDOCS 8706030378 | |
| Download: ML20214P435 (4) | |
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i bBALTIMORE GASAND ELECTRIC CHARLES CENTER. P.O. BOX 1475 BALTIMORE. MARYLAND 21203 NUCLEAR ENGINEERING SERVICES DEPARTMENT CALVERT CUFFS NUCLEAR POWER PLANT LU$8Y, MARYLAND 20657 May 26,1987 FCM 87-117 6Y SIg U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Mr. Scott A. McNeil
SUBJECT:
Calvert Cliffs Unit 2 Cycle 8 Reload License Submittal
REFERENCES:
1.
Mr. 3. A. Tiernan (BG&E), to NRC, letter dated February 6,1987,
" Request For Amendment Eighth Cycis License Application,"
2.
Mr. S. A. McNeil (NRC), to Mr. 3. A. Tiernan (BG&E), letter dated May 4,1987," Amendment to Facility Operating License No.108"
Dear Mr. McNell:
Combustion Engineering identified two input values to the Uni 2 Cyc198 Reload Design Report (Reference (1)) that were incorrectly reported. The and F Bank 5 inserted input values for Unit 2 Cycle 8, as presented en page 7-3, T ble 7 Nf Reference (1),
should both be 1.904 instead of 1.87 as originally reported. The value of 1.904 was used L
in the performance of the safety analyses, but was not reported on Table 7-2 as such.
l Rather the reference cycle values were reported by mistake. No results of the safety i
analyses are altered since the correct value of 1.904 was used. This error was identified during the quality assurance process at Combustion Engineering.
Two pages of Reference (1) are affected by this error and are included as attachments to this letter.
BG&E supports that no conclusions or results of either Reference (1), the Reload Design Report or Reference (2), the Safety Evaluation Report, are affected by this error.
M
/d.~A. Mihalcik f
j g uel Cycle Management 870603037e 870526 h
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ADOCK 05000318 Attachment
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FCM 87-117-Page 2 ~.
. D. A'. Brune, Esqdire cc
- 3. E. Silberg, Esquire
- R.' A. Capra, NRC 2
- 3. M. Allen, NRC.
T. Foley/D. A. Trimble, NRC '
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i 24-57a(86H2)/cp-21 April 28, 1987 B-87-035 7.0 TRANSIENT ANALYSIS The ' Design Basis Events (DBEs) considered in the Unit 2 Cycle 8 safety-analyses are listed in Table 7-1.
Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to melt fuel design limits are presented in Table 7-2.
As indicated in Table 7-1, the only reanalyses included herein are for the Boron Dilution and CEA Ejection events.
No other reportable reanalyses were performed for any other DBE's since all results for these DBE's lie within the bounds of (or are conservative with respect to) the reference cycle values (Unit 2 Cycle
.7, Reference 1).
The major characteristics of the Unit 2 Cycle 8 transient analyses are the following:
~
a.
With the exception of data for the Boron Dilution, Full Length CEA Drop, CEA Ejection and Steam Line Rupture events and the Bank 5 Radial Peaking Factors, all key input parameters to the transient analyses lie within the bounds of those of the reference cycle, b.
An evaluation of data generated by FATES 3B ~ (Reference 2), which is a revised version of the FATES 3 fuel evaluation model (Reference 3), showed no significant impact.
-FATES 3B has C
received interim NRC approval (Reference 4).
c.
A reanalysis of the Boron Dilution event was performed to
- acconnodate changes in input data due to the implementation.of 24-month cycles.
d.
An evaluation of the Full Length CEA Drop event showed that a revised Doppler curve (changed to acconnodate the implementation of 24-month cycles) does not have a significant impact on the analysis.-
e.
A reanalys'is of the CEA Ejection event was performed to establish generic values for the CEA ejection worth and the post-ejected radial power peak for the Hot Zero Power condition.
~
f.
A reanalysis of the Steam Line Rupture event to accommodate the implementation of 24-month cycles yielded results that were less limiting than those previously reported, g.
An evaluation of the effect of increasing the Bank 5 Radial Peaking Factors showed no significant impact, i
7-1
i 4'24-57a(86H2)/cp #*3* '
f' April'28,'1987 B-87-035 TABLE 7-2 CALVERT CLIFFS UNIT 2 CYCLE 8 CORE PARAMETERS INPUT TO 5AFETY ANALYSES
~
FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS
~
Reference Cycle
- Physics Parameters Units (Unit 2 Cycle 7)
Unit 2 Cycle 8 Radial Peaking Factors For ONB Margin Analyses T
(Fr)
Unrodded Region 1.70**'+
1.70**'+
Bank 5 Inserted 1.87***+
1.904**'+
l For ylanar Radial Component (F
) of 3-0 Peak (CiNLimitAnalyses)-
Unrodded Region-1.70**
1.70**
Bank 5 Inserted 1.87**
1.904**
l Moderator Temperature 10~44/*F '
-2.7 - +.7
-2.7 + +.7
(
Coefficient Shutdown Margin (Value
%ao
-3.5
-4.5 Assumed in Limitin
~ EOC Zero Power SLB Tilt Allowance 3.0 3.0 Power Level MWt 2700**
2700"*
Maximum Steady State
'F 548**
548**
Inlet Temperature Minimum Steady State psia 2200**
2200**
RCS Pressure 1
6
(
Reactor Coolant Flow 10 1bm/hr 138.5**
138.5**
i Negative Axial Shape I
.15**'+
.15**'+
LCO Extreme Assumed P
at Full Power (Ex-Cores) l
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l 7-3
- _ - _ -. _ -. - - _ -. -. - -,,