ML19343C555
| ML19343C555 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/16/1981 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML19343C552 | List: |
| References | |
| NUDOCS 8103240480 | |
| Download: ML19343C555 (9) | |
Text
Fort St Vrain <!1 Technical Specifications Amendment Page 3.1-1 O
3.1 Reactor Core - Safetv Limit 3.1.1 Acolicabiliev Applies to the limiting combinations of core ther=al power and core helium flow rate.
3.1.2 Objective To maintain the integrity of the fuel particle coatings.
3.1.3 Soecification SL 3.1 - Reactor Core Safety Limit The combination of the reactor core power-to-flow ratio and the tota'.
integrated operating time at the power-to-flow ratio during the lifetime of any segment sh .1 not exceed the following limits.
3.1. 3. .i. Power-to-Flow Ratio Between 1.17 and 2.5 The co :bination of the reactor core power-to-flow ratio and the total integrated operating time at this power-to-flow ratio during the lifetime of any segment shall not exceed the limit given in Figure 3.1-1. This safety limit is exceeded when the combination of operating parameters (power, flow, md time) lies above or to the right of the line given in Figure 3.1-1.
For the purpose of obtaining the total eff ective integrated operating time f or Figure 3.1-1, only. transients resulting in a pownr-to-flow ratio above
- the curve of Figure 3.1-2, at the appropriate core power level shall be used.
3108240480
Fart St. Vrein #1 Technical Specifications Amends:nt Pags 3.1-2 For transients which result in power-to-flow ratios between 1.17 and 2.5, the operator shall immediately reduce power to lower the power-to-flow ratio to less than 1.17. If this corrective action is not successful within two minutes, an immediate shutdown shall be in1:iated. .
3.1.3.2 Power-to-Flow Ratio Greater than 2.5 and' Less Than or Ecual to 15 The time interval (t) from the start of the transient in power-to-flow ratio above Figure 3.1-2 to the time at which the power-to-flow ratio goes below a value of 2.5 shall be reduced by 100 seconds and the remaining time shall be limited to a total allowable time of 2 minutes. The allowable time for power-to-flow ratios less than 2.5 at times larger than (t) are given in 3.1.3.1.
3.1.3.3 Power-to-Flow Ratio Greater than 15 The time interval (c) from the start of the transient in power-to-flow ratio above Figure 3.1-2 to the time at which the power-to-flow ratio goes below a value of 2.5 shall be reduced by 60 seconds and the remaining time shall be limited to a total allowable time of 2 minutes. The allowable time for power-to-
~
flow ratios less than 2.5 at times larger than (t) are given in 3.1.3.1.
3.1.3.4 Power-to-Flow Ratio Less Than l'.17 For power-to-flow ratios exceeding values of Figure 3.1-2 but less than 1.17, an operating time limit of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> has been established to insure fuel integrity. If the combination of power-to-flow ratio and percentage of design
For: S:. Vrain EI.
Iachnical Specifica:1ons Ansndment Pags 3.1-3 h
core :her 41 power exceeds the curve of Figure 3.1-2, the operator vill take prompt ac:fon to bring the co=bination of power-:o-flow and percentage of de-sign core ther=al power under the curve of Figure 3.1-2. If this cannot be accomplished in 30 =inutes, an orderly shutdown shall be initiated.
3.1.4 3 asis for Specification Si, 3.1 Ia order to assure integrity of the fuel par:1cles as a fission pro-duct barrier, it is necessary to prevent the failure of significant quantities of fuel particle coatings. Failure of fuel particle coatings can result fro =
the =1gration of the fuel kernels through their coatings. The dependence of the rate of =igration of the particle kernel upon ta=perature and te=pera:ure difference across the particle kernel using 95% confidence levels on the ex-perl= ental data was used. During power operation, there is a temperature gradient across each fuel rod, the higher temperature being at the center of the fuel rod and the lower temperature a: the outer ~dge of the fuel. In an overtemperature condition, fuel kernels can =ove through their coatings -
in this temperature gradient, in the direction of the higher te=perature.
The Core Safety Li=1t has been constructed to assure that a fuel ker-nel migrating at the highest rate in the core vill penetrate a distance less than the combined thickness of the buffer coating plus the inner isotropic coating on the particle.
The quantity of failed particle ceatings in the core at all ti=es is determinable by =easurement of gaseous ?ission product activity in the pri-
=ary loop.
Fort St. Vrain #1 Technical Spccificstions
. . Amsndment Page.3.1-4 .
J In Figure 3.1-1, the quantity ? is the fraction of design core ther-mal power, i.e., core thermal power (>M) divided by 842. The quantity F is the fraction of design core coolant flows at the circulators, i.e. , the total coolan: flow at the circulators in (lb/hr) divided by 3.5 x 106 lb/hr.
The limiting combinations of core thermal power and core coolant flow rate are established using a series of short time conservative assumptions.
All hot channel factors discussed in Section 3.6 and all power peaking fac-tors discussed in Section 3.5.4 of the FSAR were applied in deter =ining this limiting curve. The range of region radial power peaking factors (average pesar density in any refueling region, E=,S, Med W amage pwer densW in the core, Icore) was assumed to be less than or equal to 1.83 and guater than or equal to 0.4. The *== intra-region power peaking factor (average power density in a fuel column, Ig, divided by the average power density in a fuel region, Yreg) used was 1.46 1 0.2 for regions with control rods inserted and 1.34 1 0.2 for all unrodded regions. A conservative estimate of the most unfavorable axial power distribution was also used. That is, the ratio of power density in the bottom layer of fuel elements of a core region, j P ; the average power density of the region, P is less than lmrlp- ,
or equal to 0.90 1 0.09 for regions with control rods fully inserted or with-drawn, and 1.23 1 0.12 for regions with control rods inserted more than two feet. The measured region coolant outlet temperature for the nine regions with their orifice valves most fully closed and all regions with control rods inserted more than two feet, was assumed to be not more than 50*F greater than the core average outlet temperature. The measured region coolant out-let temperature for the remaining core regions was assumed to be not more than il
Fort St. Vrain dl Technical Sptcificationa A=tndment Pags 3.1-5 200*F greater than the core average outlet temperature. Durir g nor=al full power cperation, a condition with any =easured region outlet te=perature more than 50*F above average should not persist for longer than a few hours. A measurement uncertainty for the cara region outlet temperature of i 50'? was assumed. A 5% uncertainty in flow measurement and a 5: uncertainty in reactor thermal power measui.euent was assumed in establishing the li=it.
For the total fuel lifetime in the core, based on calculations incor-porating plant para =eters and uncertainties appropriate for longer ti=es, mi-gration of the fuel particle kernel through its coating would be less than 20 microns for the fuel with the most damaging temperature history and with the core operated constantly at any of the power-to-flow ratios and power com-binations shown on the curve of Figure 3.1-2. Out of a total inner coating thickness of 70 =1erons, only 50 microns have been used for the deter-Sation of fuel particle failure in setting the limit curve in Figure 3.1-1.
As can been seen from Fige.re 3.1-1, sufficient time (at least nine minutes) is available for the operator to take corrective action to prevent the core safety limit from being exceeded for power-to-flow ratios less than or equal to 2.0. In order to reach a power-to-flow ratio of this magnitude
' through an increase in core power, significant equipment malfunction, or failure, and/or one or more significant deviations from operating procedures i
would have to occur.
However, high core power-to-flow ratios can also be obtained as a result of a reduction or loss of primary coolant circulation. The core nega-tive coefficient of reactivity provides an intrinsic means to reduce the core
Fort St. Vrain #1
- Tcchnical Specifications
! Amendment Pagn 3.1-6 power and the power-to-flow ratio, and the plant control system will usually initiate scram sequences in such cases. Nevertheless, for brief periods of time prior to or during the scram, high power-to-flow ratios can exist. Due to the slow ther=al response of the core as a result of its high heat capacity, these power-to-flow ratios can exist for short periods of time without signifi-cantly increasing fuel te=peratures and fuel kernel migration distances.
The behavior of the core during numerous transients has been dis-cussed in the FSAR. The slow thermal response of the core is evident from I the analysis results shown in Chapter 14 and Appendix D. For example, the Loss of Forced Circulation (IDFC) accident analysis presented in FSAR Appendix D shows that the =+= core temperature rises at a rate of only 6*F/ minute for the first two hours following transient initiation. During that time, however, the primary flow rate is zero, while due to fission product decay heat the effective core power is as high as 3". Thus , the power-to-flow natio is far above the highest value shown in Figure 3.1-1.
1 Under transient conditions, either abnormal rapid power increases or sudden flow decreases, the allowable time in Figure 3.1-1 and 3.1-2, which was derived fron steady state calculations, is not a meaningful indicator of kernel migration and fuel integrity. Accordingly. a delay period is appropriate for transients entailing either a sudden decrease in primary coolant flow with a consequent decrease in reactor power or an abnormal rapid power increase.
This delay period represents the time required for the fuel to heat up from normal operating temperatures to the steady state temper!. cures at higher power-to-flow. ratios. represented by f:e Core Safety Limit Curve. Therefore,
Fort St. Vrain di i Technien1 Spccifications ;
. Am:ndnint ;
Pags 3.1-7 '
this delay period can be allowed without co= promising the integrity of the fuel. As a result of =any transient analyses, the delay period has been con-servatively set at 100 seconds for transients resulting in a power-to-flow ratio above 2.5 but less than or equal to 15 and 60 seconds if the power-to-flow ratio is greater than 15.
The allowable time, af ter the delay time, for all transients which lead to a power-to-flow ratio in excess of 2.5 is set at 2 minutes which is also the allowable time for a power-to-flow ratio of 2.5 given by Figure 3.1-1.
The limitation of allowable operating time to a value of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> l
for all operations with a power-to-flow ratio above the curve of Figure 3.1-2 and below a value of 1.17 provides a conservative limit since this is the allowable time for a power-to-flow ratio of 1.17 given by Figure 3.1-1.
Based on the analyses, a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> continuous operating time in this range of power-to-flow ratios would be conservative, with reference to possible fuel damage. However, from an operating viewpoint, a 30 ninute time limit has been established for operator action which adds sufficient conservatism.
g ?
Technical Specifications e
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