ML20064H440
| ML20064H440 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/15/1978 |
| From: | Turbak M COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7812200219 | |
| Download: ML20064H440 (9) | |
Text
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O One First Nabonal Plaza. Chicago Ilhnoi Commonwealth Edison Address Reply to: Post Office Box 767 Chicago. Ittinois 60690 December 15, 1978 e
Director of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission 1
Washington, DC 20555
Subject:
Quad-Cities Station Unit 1
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Information Regarding Core Loading Pattern for Reload No. 4 NRC Docket No. 50-254
Reference:
(a) Cordell Reed letter to I
Director of Nucleau Reactor Regulation dated Novenber 20, 1978
Dear Sir:
Reference (a) transmitted our proposed amendment to the License and Appendix A, Technical Specifications, to Facility Operating License No. DPR-29 to support core reload No. 4 at Quad-Cities Station Unit 1.
Subsequent to that transmittal we have revised the core loading pattern.
The enclosed infomation is General Electric documentation for conclusions regarding revisions to the Quad-Cities Unit 1, Relcad 4, Cycle 5 licensing submittal, NEDO-24145 (Reference (a), Enclosure II).
The enclosed information indicates the rr,2ctor shutdown margin, loading error, and
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rod withdrawal error were recalculated.
The new shutdown margin is above the minimum required value, while the l
loading error and rod withdrawal error results are conservatively bounded by NEDO-24145.
The results of gross core transients, control rod j
drop accident, loss of coolant accident, and the GETAB analysis at NEDo-24145 remain applicable.
Thererore, no changes are warranted to the limits presented in NEDO-24145.
One (1) original and thirty-nine (39) copies of this transmittal are provideu tor your use.
Very truly yours,
)g u,lo.
hs Michael S. Turbak
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Nu 1 r Li nsing Admini trat r
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j QUA0 CITIES 1 RELOAD 4 NEW CORE REFERENCE LOADING PATTERN SAFETY EVALUATION
- 1. 0 DESCRIPTION The reference core :oading pattern shown in Reference I has been changed at the request of Commonwealth Edison Company.
was requested in order to improve operational flexibility of theThe change plant.
located in the central region of the core, to peripheral o peripheral locations in the core.
for the full core loading did not change.The number and type of assemblies t
- 2. 0 SAFETY ANALYSIS REVIEW A safety review of the new reference core loading patter'n was performed to determine if the alterations involved any unreviewed safety questions as defined in 10CFR50.59.
deemed to involve an unreviewed safety question:A proposed change is 1.
If the probability of occurrence or the consequences of an accioent or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2.
if a possibility for an accident or malfunction of 5 different type than any evaluated previously in the safety analysis report may be created; or 3.
If the margin of safety as defir.ed in the basis for any technical specification is reduced.
The core modifications described in section 1.0 do not involve loading of a new fuel type.
of an accident is not increased nor is the possibility for a newTher i
type of accident created.
To determine if any margin of safety is reduced or if the conse-quences of any accident are increased, the effect of the loading changes on all the transients and accidents reported in Reference 1 were examined, either by reanalysis or by comparing core character-istics of the reference cores.
It,s concluded that the safety analysis limits reported in Reference 1 for Cycle 5 remain applicable.
2.1 Reactor Shutdown Margin The minimum shutdown cargin for the new reference core is 0.013 aK vs 0.020 aK for the previous reference core.
shutdown margin is still well above the minimum requird value The of.0025 aK.
Therefore, the safety margin is not reduced.
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- 2. 2 Accident Analysis l
'.11 2.2.1 Control Rod Drop Accident i
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The rod drop accident was reviewed and the results will
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remain approximately the same as given in Reference 1.
The incremental rod worth will remain less than 1.0% A 2.2.2 Loss of Coolant Accident The loss of Coolant accident (LOCA) presented by reference in Reference 1 is unaffected by the core changes.
2.2.3 Loading Error Accident The loading error was reanalyzed and the results showed that the analysis presented in Reference 1 is unchanged.
2.3 Abnormal Operating Transients
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2.3.1 Gross Core Transients Based on an examination of the core characteristics it is i
concluded that the transient analyses given in Reference 1 are appitcable to the new core.
2.3.2 Local Transients
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The Rod Withdrawal Error was reanalyzed and the results showed that the analysis presented in Reference 1 is
/'l conservative for the new core pattern.
At the 107% RBM setpoint the ACPR is less than the previous value.
v 2.4 GETAB Analysis No new fuel characteristics were introduced and all the pertinent transient and accident analyses remain applicable; therefore the GETA8 analysis presented in Reference 1 remain applicable.
2.5 Stability Analyses e
No new fuel characteristics were introduced and the important nuclear dynamic paramoters either improved or did not change
'significantly.
remains applicable.Therefore, the stability analysis in Reference 1 i
- 3. 0 CONCLtJSIONS The new reference core loading pattern does not introduce any unreviewed safety questions as defined in 10CFR 50.59 when compared to the results presented in Reference 1.
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4.0 REFERENCES
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" Supplemental Reload Licensing Subm1ttal for Quad Cities 1.
Nuclear Power Station Unit 1 Reload 4." NEDO-24145, dated 3
September 1978.
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NUCLE AR ENERGY BUSINESS GROUP
- GENER AL E LECTRIC COMP ANY SAN JOSE, CALIFORNI A 95125 i
G EN ER AL $ ELECTRIC
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APPLICABLE TO:
J NEM-24145 PusaC ATiON NO ERRATA And ADDENDA 78hTD283 T. i. E. NO.
SUPPLEMENTAL RELOAD TIT LE y
go, LICENSING SUBMITTAL FOR OUAD November 1978 OATE CIIIES NPS UNIT 1 RELOAD 4 NOTE: Correct allcopies of the applicable September 1978 ISSUE OATE publication as specified below.
e REFERENCES g
P RAG APH L NE)
(CO R R E C NS AN ADDITIONS:
1 Page 1 Replace with new page 1.
s 2
Page 3 Replace with new page 3.
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3 Page 6 Replace with new page 6.
4 Page 14 Replace with new page 14.
NOTE: Revisions are indicated by change bar in right-hand margin.
Y 1 Of I PAGE
't NEDO-24145-1
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PLANT UNIQUE ITEMS (1.0)*
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[lantparameterchanges See Appendix A a.
- b. ' Loading Error See Appendix A i
c.
Loss-of-Coolant Accident Analysis See Reference 1 (pg 5) d.
Barrier Lead Test Assembly (BLTA)
See Reference 2 (pg 5) e.
R (item 4)
Value shown includes effect of B C 4
settling (0.0004 Ak) 2.
RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0) 4 Fuel Type Number Number Drilled 0
Irradiated Initial (7DB212) 128 Reload-l (7DB230) 22 (7DB230-STR) 1 (7DB230-Pu) 5 (8DB250) 36 Reload-2 (8DB250) 104 (8DB262) 52 Reload-3 (8DB250) 184 New Reload-4 (8DRB265L) 192 192 Total 724 192 3.
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 15,695 mwd /t.
Assumed reload cycle exposure:
16,100 1Md/t.
f Core loading peittern:
Figure 1.
t 4.
CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM UORTH -
NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)
BOC k,gg Uncontrolled 1.106 Fully Controlled O.953 Strongest Control Rod Out 0.987 R, Maximum Increase in Cold Core Reactivity 0.0004 with Exposure Into Cycle, ak
- ( ) ref ers to areas of discussion in " Generic Reload' Fuel Application,"
NEDE-240ll-P-A, August 1978.
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NED0-24145-1
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SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
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CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) 4 Power Flow e
Q/A s1
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Plant Transient Exposure
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(psig)
(psig) 7xf 8x8R
Response
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Lead Rejection EOC 99 100 264.4 110.7 1213 1248 0.16 0.22 Figure 3a without Bypass Turbine Trip EOC -
100 100 225.0 108.8 1214 1248 Figure 36
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without Bypass 1000 mwd /t Loss of 145'F 100 100 120.7 119.4 993 1045 0.15 0.18 Figure 4 Feedwater Me e t i n,.,
Feedwater EOC 100 100 155.3 107.5 1123 1158 0.08 0.11 Figure 5a Controller Failure Feedwater EOC -
100 100 131.2 105.7 1117 1153 Figure $b Controller 1000 mwd /t Failure 10.
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRLHENT FAILURE) TRANSIST SLYMARY (5.2.1)
Rod Position MLHCR
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Rod Block (Feet ACPR (kW/ft)
Limiting Reading _
Withdrawn) 8x8/8x8R 8x8/8x8R Rod Pattern
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104 3.5 0.13 14.58 Figure 6 105 3.5 0.13 14.58 Figure 6 l
106 4.0 0.15 15.75 Figure 6
- 107 4.5 0.17 16.02 Figure 6 108 6.5 0.22 16.11 Figure 6 l
109 7.5 0.23 16.25 Figure 6
- Indicates setpoint selected l
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NEDO-24145-1 s H H it! H M s sHMMMMMMMMs sMMMMMMMMMMMs
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MMMMMMMMMMMMM sMMMMMMMMMMMME
, :::MMMMMMMMMMMMMMM lrMMMMMMMMMMMMMMM
":M M M M M M M M M M M M M M M
- MMMMMMMMMMMMMMM
- MMMMMMMMMMMMMMM
- MMMMMMMMMMMMM" HMMMMMMMMMMMM L c eMMMMMMMMMMMs "MMMMMMMMM*
"MMMMM*
01 D3 05 07 09 11 13 15 17 19 21 25 27 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 i
FUEL TYPE A
- INITI AL FUEL 0 = RELOAD 2 (809250) 8 = RELOAD 1 (T082301 E = RELOAD 2 (808262)
(S = SE MENTED F = RELOAD 3 (808250)
C = RELOAD 1(808250)
Figure 1.
Quad Cities Unit 1 Reload 4 Design Reference Core Loading 6
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lj-02 06 10 14 18 22 26 30 t
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8 8
56 24 8
$1 10 10 18 i
47 24 10 4
41 12 12 38 34 39 16 6
8 0
35 4
10 36 34 31 32 24 0
14
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NOTES: '1. ROD PATTERN ($ 1/4 CORE MIRROR SYMMETRIC UPPER LEFT QUADRINT SHOWN ON MAP C"-
- 2. NUMBERS INDICATE NUM8ER OF NOTCHES WITHDR AWN OUT OF 48.
8 LANK IS A WITHDRAWN ROO
- 3. ERROR ROD IS (26. 39)
Figure 6.
Limiting Rod Pattern For Rod Withdrawal Error 14