ML19319D763
| ML19319D763 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Rancho Seco  |
| Issue date: | 10/23/1974 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19319D764 | List: |
| References | |
| RJR-74-401, NUDOCS 8003250765 | |
| Download: ML19319D763 (7) | |
Text
'
AEC DISTRIBUTION FOR PART 50 DOCKET MATERIAL (TEMPORARY FORM) 110h2 CONTROL NO:
FILE:
FROM: SMUD DATE OF DOC DATE REC'D LTR TWX RPT OTHER 8*C %$ $ Calif.
95813 lo-a-74 10-29-74 XX 7r TO:
DL i signed SENT LOCAL PDR H
CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
g 1
50-312 XXX DESCRIPTION:
Ltr re their 10-18-Th 1tr...
ENCLOSURES: Corrections & additions to report trans the follcWing:
entitled " Ejected Rod Worth Measurement"....
Proposed Change to the Tech Specs & the FSAR.
w i m giv= 5
%Jci,J Mdg (1 cy ea encl rec'd) f t90fil.0V8 PLANT N AME: Rancho Seco Station Unit r FOR ACTION /INFORMATION DHL 10-30-74 BUTLER (L)
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$SMUD l
SACRAMENTO MUNICIPAL UTILITY DISTRICT C 6201 S Street, Box 15830. Sacrarnento. California 95813; (916) 452-3211 RJR 74-401 October 23, 1974 Fegu!aury gcg Director ys Directorate of Licensing jg ',
7 Office of Regulation p
N United States Atomic Energy Comission p
Washington, D. C.
20545
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AEC Docket No. 50-312
',,h,9d Non-Routine Report No.1, Revision 1
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j Rancho Seco Nuclear Generating
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Station, Unit No.1
/
N:y y y y f /
Dear Sir:
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This letter supersedes the District's letter of October 18, 1974 and includes corrections and additions to the attached report.
The Rancho Seco Technical Specifications, Section 6.12.2B.1, require a written report to you within thirty (30) days upon discovery of any substantial errors in the transient or accident analyses, or in the methods used for such analyses, as described in the Final Safety Analysis Report or in the Technical Specifications.
In compliance with these specifications, we are enclosing a report titled " Ejected Rod Worth Measurement" and a proposed change to the Technical Specifications and the FSAR. A fomal change will be submitted at a later date.
Sincerely yours,
. bt$
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. J. Mattimoe Assistant General Manager, ({
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and Chief Engineer 5
1x Enclosures S
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R. H. Engelken 6
g Director of Regulatory Operations e
y Regfon V 11042
.h4rv Elected Rod Worth Measurement
_j()-;gq; 7 p On September 21, 1974, during the Rancho Seco zero power startup testing, an ejected rod worth of 1.247. 4c was measured with the control rods positioned near the technical specification rod insertion ihnit for zero power. This value is in excess of the allowable limit of 1% Ap for normal operation (not applicable during testing) and disagreed with a prediction that the ejected rod worth would be approximately.97. Ao.
Following this measurement the Babcock and Wilcox Nuclear Engineering Group re-evaluated the prediction by performing additional calculations and developing the correlation shown in Figure I which shows the relaticnship between ejected rod worth and inserted rod worth, where the data points include all available calculations and measurements, including measurements of stuck rod worth.
The results of this correlation indicate excellent agreement between measure-ments and calculations.
It is also apparent from the shape of a curve that a prediction of ejected rod worth based on a linear fit of the data available for inserted rod worth up to approximately 3% de would lead to an underesti-mate of ejected rod worth for greater amounts of inserted worth.
As a result of these measurements, it is necessary to modify the existing rod insertion limits to insure that the maximum potential ejected rod worth will be less than the 1% de required for normal operation at 7525*F and zero power.
Based on the measured and predicted data shown in Figure 1. it has been decided to conservatively thnit the maximum potential ejected rod worth to.857. 49.
This corresponds to a maximum inserted rod worth of 3.87 4p which would occur at a control rod insertion limit of 49%
withdrawn on Group 5 at 7525'F and zero power.
Following this measurement in subsequent operations the maximum insertion of Group 5 control rods has been limited to 497. whenever the reactor is critical.
Consistent with this limitation, the attached pages contain recommended changes to the Final Safety Analysis Report and the Technical Specifications. The changes to Pages 14.2-20 and 14.2-21 of the Final Safety Analysis Report show that the thermal power transient resulting from an ejected rod worth of 1% ak/k at zero power is less severe than the transient caused by an ejected rod worth.65%-4k/k at full power.
Similarly, Page 3-33a of the Technical Specifications has been revised to clarify the basis for an ejected rod worth of 17. ok/k at hot zero power. Figure 3.5.2-1 of the Technical Specifications has been modified to restrict rod postion limits on control rod Group 5 to greater than 497, at zero power.
An additional restric-tion of a minimum rod index of 80 at 15% of raced power has been added to insure a minimum shutdown margin of 17. ok/k assuming the highest worth control rod is stuck fully withdrawn, even though it is inserted.
The FSAR changes will be submitted formally as Amendment No. 29 and the changes to the Technical Specifications will be submitted formally within sixty days.
~
c In response to a request by the AEC that B&W provide measurement' uncertainties associated with the ejected rod worth measurement, the following information is provided:
The ejected rod worth was measured by two techniques.
A.
Boron' Swap Measurement Uncertainty:
5%
1.
Contributions:
a.
Reactimeter errors 1.
Delayed neutron parameters 2.
Ranger non-linearity 3.
Algorithm approximation, non-ideal sampling, limited precision in numbers.
b.
Errors in determining reactivity step changes l.
Random errors 2.
Systematic errors due to spatial flux redistribution B.
Rod Drop Measurement Uncertainty:
Drops (0 to 17. ak/k) +/- 10%
Drops (17. to 27. Ak/k) +20, -10%
1.
Contributions a.
Reactimeter errors 1.
Delayed neutron parameters 2
Ranger non-linearities 3.
Algorithn approximations, non-ideal sampling, limited precision in numbers.
b.
Spatial flux redistribution l
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dby Safeguards Analysis
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"As a check on the point kinetics c'alculation,. the rod ejection accident also analy cd for a limited number of cases in support of the technical speci-
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was fication rod worth using the exact TUICL digital cc=puter program. (ll$
wo di=cnsional, space-and-time dependent t
The point flux shape rc=ains constant during a transient. kinetics model assumes that the This flux shape contains peaking factors which reflect unusual rod patterns such as th6 flux adjacent to a position where a high worth rod has been removed.
Therefore, these poinc
~kinctics peaking factors are much higher than any that would actually occur in the core during normal operation.
calculation is to find the flux shape during a transient.The purposc of using an exact spac But to have a transient where a rod is ejected from the core,'one =ust start with a flux shape that is necessarily depressed in the region of the ejected rod.
- fact, In the higher the worth of the rod, the more severe becomes the depression This flux depression also causes a fuel te=perature depression.
is ejected from this position, Uhen the rod 6
so=e local peaking.
the flux quickly assumes a shape that shows However, when this " exact" peaking is applied to a region initially at depressed fuel te=per rurcs, as it is in the case of the regions adjacent the ejected rod, the resultant energy deposited in' these regions causes a to lower peak te=perature and peak ther=al power than does applying an arbitrary maximum peaking factor to an undepressed peak power region.
The results from TUICL were used to calculate the maximum total energy deposited in each region of the core following a rod ejection; the highest energy is reported in Table 14.2-10.
The result is that the hotte'st TUICL cade actually undergoes a less severe transientregion simulated in the than the hottest fuel rod assumed in the point kinetics model.
result is unifor=1y true for all red worths.-As scen in table 14.2-10, this For certain cases where the ejected rod has a lcw wor.th, or where at least cne reactivity coefficient is very negative, or the initial power level is low, there' is considerable pressure buildup in the reactor coolant system
.bccause of the increased heat being added to the coolant with no increase in h2at de=and.
}bny of these transients never reach the overpower trip point.
Far this class of possibility, the;high pressure trip must be ~ relied on, and this is incorporated in the calculation.
p 14,'. 2. 2. 4. 4 Results of Analysis
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A.
Zer,o Power. Level, C
The nominal BOL and EOL rod ejection analysis' vas performed at 10
~3 of rated power, and the results can be see.n. in tabic 14.2-11.
No DNB and no fuel damage would result from the transient caused by the ejection of a rod worth.657. ak/k.
The percent of fuel rods 20 in DNB for an ejected rod worth 17. dk/k would be less than the number of rods in DNS with an ejected rod worth.657. 4k/k at rated power.
A sensitivity analysis has been performed around these two cases in which'the Deppler and nederator coefficients, trip del 3Y ticc 8'd e
fod worth ucre varied.
Figure 14.2-2 shows the peak neutron power as a function of 'cjected rod worda from 0.2 to 0.7 percent A k/k.
The 14,2 20 Amend.nent 29
,i
-dby Safeguards Analysis
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TABLE '14.2 :lCT
- l COMPAP,ISON OF SPACE-DE?ENDENT.AND POINT KINETICS
..RESULTS ON FUEL ENTHALPY
' Ejected Peak-to-Average V alues Fuci Enthalpy, cal /gm Rod Worth.
%Ak/k TUIGL Point Kinefics
".TWIGL Point Kinetics s
BOL Rated Pouer
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0.38
'3. 04.
3.24
'125 150 O.83 2.67 3.24 174 225 BOL Zero Pouer 60 L-O.56 '
4,.1
, 3,24 38 48 71 0.83 4.4 3.24 TABLE 14.2-11 SUlt!ARY OF CONTROL ROD EJECTION ACCIDENT ANALYSIS ea
- wer, ated po,rer
. Initial Power Level, Ejected Rod Worth,
% rated power Z Ak/k Neutron Thermal
~
O.1 (POL) 0.65 94 70 0.1 (EOL) 0.65 1,160 '
32 0.1 (BOL) 1.0 8,417 132 16,302 102 25
,0.1 (EOL) 1.0 100.0 (BOL) 0.65
.', 700 158 0.65 1,600 138 100.0 (EOL) f curve shows two distinct parts corresponding to worths less than and v'ilues near to and greater than 3.
Figure 14.2-3 shows the corres-ponding results for the peak thermal power.
It is seen that for rod worth values near pronpt critical, the period is small enough to carry the t,rinsient through the high neutron flux trip.
For lower values the pressure trip is relicd on.
No DNS occurs for any of these parameter variations.
~
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14.2-21 Amendment 29 w
.