ML20236U519
| ML20236U519 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/23/1987 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236U506 | List: |
| References | |
| NUDOCS 8712030111 | |
| Download: ML20236U519 (24) | |
Text
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UNITED STATES I
NUCLEAR REGULATORY COMMISSION n
,j.:
WASHINGTON, D. C. 20666
%,...../
PUBLIC SERVICE CCHPANY OF COLORADO DOCKET NO. 50-267 FORT ST. VRAIN NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 57 License No. DPR-34 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Public Service Company of Colorado (the licensee) dated June 25, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
i 8712030111 871123 PDR ADOCK 05000267 P
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2.
Accordingly, the license is amended by changes to the Technical l
Specifications as indicated in the attachment to this license amendment, and paragraph 2.D(2) of Facility Operating License No.
DPR-34 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 57, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 30 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION O w Jose A. Calvo, Director Project Directorate - IV Division of Reactor Projects --III, IV, V and Special Projects Office of Nuclear Reactor Regulat' ion
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 23, 1987 4
O ATTACHMENT TO LICENSE AMENDMENT NO. 57 TO FACILITY OPERATING LICENSE NO. OPR-34 DOCKET NO. 50-267 Replace the following pages of the Appendix A Technical Specifications with the attached pages as indicated.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 2-9 4.0-1 4.0-1 4.0-2 4.0-3 4.0-4 4.0-5 4.0-6 4.0-7 4.1-14 4:1-14 4.1-15 4.1-15 4.1-16 4.1-16 4.1-17 4.1-18 4.1-19 4.1-20 4,1-21 4.1-22 4.1-23 4.1-24 4.1-25 5.1-16 1
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L_______--___--
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Fort St. Vrain #1 Tcchnical Specifications Amendment No. 57 Page 2-9 2.23 CALCULATED BULK CORE TEMPERATURE The CALCULATED BULK CORE TEMPERATURE shall be the calculated average temperature of the core, including graphite and
. fuel, but not the reflector, assuming a loss of all forced circulation of primary coolant flow.
2.24 CORE AVERAGE INLET TEMPERATURE The CORE AVERAGE INLET TEMPERATURE shall be the arithmetic average of the operating circulator inlet temperatures, adjusted for circulator power input, steam generator regenerative heat loads, and PCRV. liner cooling system heat losses.
I
Fort St. Vrain #1 fechnical Spec k3endmentNo.gications Page 4.0-1 4.0 LIMITING CONDITIONS FOR OPERATION 4. 0 '.1
.The Limiting Conditions for Operation, specified
~
In this
- section, define the lowest functional capability or performance levels necessary to assure safe operation of the facility. These Limiting Conditions for Operation ~prov'ide for operation with sufficient redundancy so that
- further, but
- limited, degradation of equipment capability or performance, or the' occurrence of a postulated
. incident will not prevent a safe reactor shutdown.
4.0.2 These Limiting Conditions for Operation do not replace plant operating procedures.
Plant operating procedures establish. plant operating
. conditions with at least the capability and performance specified in these Limiting Conditions for Operation.
4.0.3 Violation of a. Limiting Condition for Operation shall be corrected as soon as practicable. Unless otherwise stated in these specifications, the condition would be corrected or the reactor shall be shutdown in an orderly manner within a 24-hour period.
Fort St. Vrain #1 Technical Specifications Amendment No. 57 Page 4.0-2 4.0.4 Where the Applicability of a-FSV Technical Specification is defined in terms of the CALCULATED BULK CORE TEMPERATURE, the time at which this temperature reaches 760 degrees F
fcilowing an interruption of all primary coolant flow is the time after which specification requirements are applicable,.
The time for the CALCULATED BULX CORE TEMPERATURE to reach 760 degrees F following an interruption of all primary coolant flow is determined as follows:
a.
Using the applicable operating power history prior to interruption of primary coolant flow, determine the decay heat power from Figure 4.0-1.
b.
Using this decay heat power and the average core temperature prior to the primary coolant flow interruption, determine the time required to reach 760 degrees F from Figure 4.0-2.
c.
The maximum time for which primary coolant flow can be interrupted is the time interval determined in Specification 4.0.4.b, not to exceed 21 days.
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c-7 Fort St. Vrain #1 Technical Specifications Amendment No. 57 Page 4.0-5 BASIS for SPECIFICATION 4.0.4 The CALCULATED BULK CORE TEMPERATURE is the calculated,
}
j
' time dependent, average tempe,rature of the' core, including graphite and fuel, but not the reflector, assuming a loss t
of all forced circulation of primary coolant flow.-
The 4
calculation uses several conservative assumptions:
- 1) The decay heat power at the start of the core heatup l
has been conservatively selected using Figure 4.0-1 and is I
assumed to remain constant for the total interval; 2) All decay heat power generated is assumed to be retained in the active core with no heat transfer to the reflect' r, o
PCRV internals or primary coolant; and 3) A 10 percent margin has been included on the core heatup time given in Figure 4.0-2.
If the active core remains below 760 degrees F, which corresponds to the design maximum core it et temrirature, then there can be no damage to fuel or PCRV internal components, even in the absence of forced circulation of primary cociant helium flew.
i
1 Fort St. Vrain #1 TechnicalSpecigfcations Amendment No.
Page 4.0-6 The time required to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F is primarily dependent upon
- the decay heat power and the current average core temperature.
This time is conservatively estimated
~
using the data in Figures 4.0-1 and 4.0-2.
The decay heat power data in Figure 4.0-1 was explicitly calculated for the Fort St. Vrain core and is derived from Appendix D.1 of the FSAR, Figure D.1-9, revision 2.
The decay heat power resulting from a varying power history can be conservatively calculated by representing the actual power history by a series of constant power steps and'then summing the individual decay heat power cor,tribution' from each power step.
The decay heat power due to operation during the last 1000 days can be determined in this manner.
Residual decay heat power from earlier operation is conservatively estimated by assuming that this was full power continuous operation, and by then adding this decay heat power component to the calculated decay heat power value.
c-
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~
Fort St. Vrain #1
.TechnicalSpecgications Amendment No.
Page 4.0-7 Knowing the decay heat power and the current average core temperature, the time for the core to heat' up from its i
. current temperature to 760 degrees F can be obtained from Figure 4.0-2, which has been generated'using the adiabatic heat transfer-model and a heat capacity for. composite graphite as given in Appendix 0.1 of the FSAR, Figure D.1-3, revision 2.
To allow for uncertainties associated with determinir.; the time to reach a Ct.LCULATED BULK CORE TEMPERATURE of 760 degrees F,
an additional 10 percent has been included in the decay heat energy given in Figure 4.0-2.
In addition, it has been specified that any time. interval for which the primary coolant flow is interrupted shall not exceed 21 days.
This ensures a restoration of forced circulation of L
primary coolant flow to confirm core average t emperature on a periodic basis.
Although much longer intervals can be determined from Figure 4.0-2, 21 days is an adequate time to conduct operations requiring flow interruption, such as maintenance or circulator changeout.
Operating experience at Fort St. Vrain has shown that the calculated core heatup rate has always been higher than the actual core heatup rate.
Fort St. Vrain #1 Technical Specifications Amendment # J3, 57 Page 4.1-14 Basis for Specification LC0 4.1.8 An unexpected and/or unexplained change in the observed core react 1vity could be indicative of, the existence of potential safety problems or of operational problems.
A reactivity an'omally greater than 0.01 delta k would be unexpected, and 'its occurrence would be throughly investigated and evaluated.
The value of 0.01 delta k is considered to be a safe limit since a shutdown margin of at least 0.01 delta k with the highest worth rod pair fully withdrawn is always maintained (see LCO 4.1.2).
r Fort St..Vrain #1 Technical Specif Amendment # J3, gpatter.s Page 4.1-15 3
LCO 4.1.9 CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW and MAXIMUM CORE-REGION TEMPERATURE RISE
- LIMITING CONDITION FOR OPERATION l
The total helium circulator flow or the helium cociant temperature rise for all core regions shall be maintained within the limits given in Table 4.1.9-1.
-APPLICABILITY: Whenever the reactor is. operated at POWER *, in' LOW POWER OPERATION, or with the REACTOR SHUTDOWN **.#
ACTION:
a.
In POWER or LOW POWER, with any of the above limits exceeded, either:
1.
Correct the out-of-limit condition within 15 minutes, or 2.
Be in at least REACTOR SHUTDOWN within I hour with the inlet orifice valves adjusted for equal region coolant flows within the following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and, if the applicable limits are still
- exceeded, initiate PCRV depressurization within 'the following I hour.
b.
In REACTOR SHUTDOWN with the inlet orifice valves adjusted for equal region flows, with any of the above limits exceeded, either:
1.
Correct the out-of-limit condition within 15 minutes, or 2.
Initiate PCRV depressurization within the time limits of Figure 4.2.18-I of LCO 4.2.18.
Up to the power levels for which limits are shown in Figures 4~.1.9-1, -3, and -5.
With the CALCULATED BULK CORE TEMPERATURE greater than 760 degrees F as provided in Specification 4.0.4.
POWER includes operation greater than 2% RATED THERMAL POWER per Definition 2.10, LOW POWER OPERATION is per Definition 2.5, REACTOR SHUTDOWN is per Definition 2.14.
. _. _. ~. _ _. -
ji' Fort St. Vrain #1 j
Technical Specifications j
Amendment # J3, 57 Page 4.1-16 c.
In REACTOR SHUTDOWN with the inlet orifice valves adjusted at any other position, with any of the above limits exceeded, either:
1.
Adjust the inlet o'rifice valves to equal region coolant flows and be within the above limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 2.
Initiate.PCRV depressurization within the time limits of Figure 4.2.18-1 of LCO 4.2.18..
ASSOCIATED SURVEILLANCE REQUIREMENT:
SR 5.1.8 l
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Fort St, Vrain (1 Technical Spec 4 ficati:rs Amendment #
57 Page 4.1-17 Table 4.1.9-1 Region Orifice l Reactor Pressure l Lietting Condition Position l Helium Density l
for Operation l
- 1
- All regions set l Greater than 50 i The total heltum circulator for equal region l psia, with l flow shall be greater than coolant flow'**
l helium density" l or equal to the minimum EXCEPT l greater than l allowable value shown in Up to 10 regions l 60%, but less l Figures 4.1.9-1 or 4.1.9-2.
may have their l than, or equal l
orifices further l to,107.55.
l open.
l l
l l
As above.
l Greater than 50 l The total helium circulator l psia with helium l flow shall be greater i density
- 1ess l than, or equal to, the 1
l than, or equal l minimum allowable value shown l to, 60%.
l in Figures 4.1.9-3 or 4.1.9-4 I
I All regions set l Less than or l The helium coolant temperature for equal region l equal to 50 psia.l rise ** through any core region coolant flow.***l l shall not exceed 600 degrees l
l F.
I I
Orifice valves ( Greater tnan l The helius coolant temperature at any position l 50 psia.
l rise ** through any core region (Adjusted for l
l shall not exceed the limit nominal equal l
l shown in Figure 4.1.9-5.
region outlet l
l temperature).
l l
l l
Orifice valves l Less than or l The helium coolant temperature at any position l equal to 50 psia.1 rise ** through any core region (Adjusted for l
l shall not exceed 350 degrees nominal equal l
l F.
region outlet l*
l temperature).
l l
I i
Percent helium density equals:
175.12 x Reactor Pressure (psia)
(Circulator inlet temperature (degrees F) plus 460)
Helium coolant temperature rise equals IN0!VIDUAL REFUELING REGION OUTLET TEMPERATURE minus CORE AVERAGE INLET TEMPERATURE.
- Equal region coolant flow with 7 column region orifice valves set between 8% and 20% open (or the corresponding
~
position for 5 column regions).
_-____A----
1 i
Fort St. Vrain #1-Technical Specifications o
Amendment f A, 57 Page 4.1-18 l
ORIFIG VALVES ADJUSTED FOR EQUAI. REGION COOLA 1
60s 701107.31 HE!!UM DENSITY
> 50 PSIA REACTOR PRESSURE 35 35
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l l
Fort St. Vrain el Technical Specifications
~
Amendment # #, 57 l
l-Page 4.1-19 t
ORIFICE VALVES ADJUSTED FOR EQUAL REGION C001. ANT FI.07
> 60% To s 107.5% HE!. LUM DENSITY
> $0 PSIA REACTOR PRESSURE 18 10
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Fort St. Vrain #1 Technical Specifications Amendment # 57 Page 4.1-20 m
ORIFICE VALVES ADJUSTED FOR EQUAL REGION COOLANT FLOW s 60s HEllUM DENSITY 50 PSIA REACTOR PRESSURE 3S 35
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( %, HEAT BALANCE) t FIGURE 4.1.9-3 MINIMUM ALLOWABLE PRIMARY COOLANT FLOW
Fort St. Vrain #1 Technical Specifications i.
Amendment # 57 Page 4.1-21 ORIFICE V ALVES ADJUSTED FOR EQUAL REGION COOLANT FLOW s 60s HELIUM DENSITY 50 PSIA REACTOR PRISSURE 16 16 i
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Fort St. Vrain #1 Technical Specifications Amendment # 57 Page 4.1-22 ORIFICE VALVES ADJUSTE FOR NOMINAL EQUAL REGION OUTLET TEMPERATURES 1107.5% HELIUM DENSITY.
50 PSIA REACTOR PRESSURE 1000 10 4 : :
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1
i Fort St..Vrain #1 Technical Specifications Amendment # 57 Page 4.1-23 BASIS FOR SPECIFICATION _t.C0 4.1.9 The minimum. helium circulator flow or the maximum core region helium coolant' temperature rise as a
function of calculated riactor THERMAL. POWER (including power from decay heat) have been specified to prevent very low helium coolant flow rates through any coolant channel.
Very low helium coolant flow
.. rates may result in laminar flow conditions with resultant high friction factors and low heat transfer film coefficients and
~
potential for possible local helium flow stagnation'or reverse flow, which could result in excessive fuel temperatures.
This specification addresses minimum flow requirements for all coolant channels.
Since low coolant flows exist at lower reactor powers, its applicability is limited to less than approximately 25% RATED THERMAL POWER.-
Specific Fower level end points for given. conditions are as shown on the Figures.
Since THERMAL p0WER is continuously generated by decay heat-even after the reactor is shutdown, the flow requirements are also applicable in the REACTOR SHUTOOWN mode.
This specification is not applicable during SHUTDOWN when the CALCULATED BULK CORE TEMPERATURE is less than 760 degrees F.
Specification 4.0.4 provides the methodology and necessary data to determine the appropriate time interval to reach a
CALCULATED BULK ' CORE TEMPERATURE of 760 degrees F.
If the active core remains below this temperature, which corresponds to the design maximum core inlet temperature, then the design core inlet temperature can not be exceeded and there can be no damage to fuel or PCRV internal components regardless of the
- amount, including total
- absence, or
- reversal, of primary coolant helium flow.
The applicability of this Specification is also limited to the range of power level indicated in Figures 4.1.9-1, 4.1. 9-3, and 4.1.9-5.
Above the power levels for which limits are shown in these Figures, the Reactor Core Safety Limit, Specification 3.1, governs.
In addition to this Specification, fuel i -
integrity is ensured for power levels from 0 to 100% by limiting the INDIVIDUAL REFUELING REGION OUTLET TEMPERATURES to values given in Specification 4.1.7.
The core flow fraction I
limits shown in Figures 4.1. 9-1, 4.1.9-2, 4.1.9-3, and 4.1. 9-4 are based on, and thus valid for, equal region coolant flow orifice settings within the range of 8% to 20% open for seven column regions and the corresponding settings for five column
- regions, i.e.; within the range of 4.4% to 13.4% open for five J
column regions.
Equal region outlet temperature orificing is precluded below about 3% power by Figure 4.1.9-5 because uncertainties in instrumentation exceed the allowable temperature rise.
Fort St. Vrain #1 1
Technical Specifications Amendment # 57 Page 4.1-24 The limits have been developed based upon a number of conservative assumptions.
For the limits in Figures 4.1.9-1, 4.1.9-2. and.4.1.9-5, it was assumed that the primary. system was
. pressurized to.107.5 percent' of design helium density.
At lower ' densities higher region temperature rises and lower primary coolaot flow are acceptable.
Since startup operations can proceed with lower helium densities, after the reactor has been pressurized to creater.than 100 psia, which corresponds to i
about 30 percent helium inventory at 200 degrees F, flow requirements were calculated for 60% helium. density and are given in Figures 4.1.9-3 and 4.1.9-4.
Percent helium density i
equals:
175.12 x Reactor Pressure (psia)
(Circulator inlet temperature (degrees F).plus 460)
The core inlet helium temperature used in the analysis covers i
the range of 100-400 degrees F between 0 and 5% RATED THERMAL l
POWER and 100-700 degrees F above 5% RATED THERMAL POWER.
These are reasonable assumptions for low power operation.
1 The analysis is based on operation of two circulators between 0 and 5% RATED THERMAL POWER and four circulators above 5% RATED THERMAL POWER.
This is consistent with plant cperation.
j In the.analysi s to determine the limits, the effects of heat conductio:n between columns in a region, or between regions, were conservatively neglected.
Envelope values of RPF/ Intra Region Peaking (3.0/1.25 and 1.6/1.61) were used to anticipate worst-case conditions considering all future fuel cycles.
Consistently conservative nominal values and uncertainties were used for bypass flows and measured parameters throughout the analysis.
For the condition with -orifice valves at any
- position, the allowable region delta T is based upon a region peaking factor equal to 0.4.
For regions with higher power densities, higher region delta T's are acceptable.
The circulator flow determination is normally based on the empirical relationship between flow and circulator inlet nozzle delta P,
local temperature,, and local pressure.
The uncertainties associated with control room indication of these parameters were accounted for in the analysis.
Other flow' l
determination methods are acceptable providea the associated l
uncertainties are accounted for and the calculated circulator I
flow is adjusted accordingly.
b
Fort St. Vrain #1
-Technical Specifications
~
Amendment #
57 Page 4.1-25 Besides the minimum flow requirement curves with the orifices set for equal ragion flows in Figures 4.1.9-1, 4.1.9-2, 4.1.9-3, and 4.1.9-4, flow requirements are provided with up to 10 orifice valves positioned further open.
These curves assist i'n the transition between equal region flows and equal region outlet gas temperatures.
By monitoring the total circulator flow when the orifices are adjusted for equal region coolant flows, minimum flow through each region at the appropriate power can be ensured. When the orifice valves are adjusted to different positions, minimum coolant flows can be ensured for each region by monitoring the helium coolant temperature rise in that region. Maximum temperature rise requirement curves are presented for the case where the inlet orifice valves are adjusted to any position as well as the cases where no seven column orifice valve is closed to less than 8% or less than 6%
open (or the corresponding position for five column regions).
For depressurized operations, helium coolant temperature rise limits are also specified to prevent very low helium coolant flow rates through any coolant channel.
These limits haye been established based upon a 50 psia reactor pressure.
To ensure that flow stagnation in a fuel column or region does not persist, an ACTION time of only 15 minutes is allowed to correct the out of limit condition.
The requirement to be in REACTOR SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with the orifices set for equal flows in an additional 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is realistic because it takes approximately 4 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to set the orifice valves from equal temperatures to equal region flows.
This is considered acceptable since there is sufficient primary coolant flow from the circulators which are driven by steam generated from residual heat in the system following REACTOR SHUTDOWN.
If, aft'er this action, there is still inadequate flow, depressuri' zing the PCRV further reduces the tendency toward stagnation and reverse flow.
r-Fort St. Vrain #1 Technical Specification s
]
Amendment # 57 Page 5.1-16 I
h SPECIFICATION SR 5.1.8 - MINIMUM HELIUM FLOW / MAXIMUM CORE REGION TEMPERATURE RISE SURVEILLANCE REQUIREMENT l
The total helium circulator flow or the helium coolant temperature i
rise through each core region shall be' determined te be within the i
limits of LCO 4.1.9 at least once per 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.
BASIS for SPECIFICATION SR 5.1.8 Surveillance.
of the helium circulator flow or helium ' coolant temperature rise once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that the requirements of LCO 4.1.9 are met.
In addition, plant procedures require that the flow rate, core outlet temperatures, and power level be monitored l
continuously whenever the power level is being changed or orifice l
valves are being adjusted.
In performance of the surveillance, the i
total reactor helium coolant flow is determined by calculation l
consistent with the method used to determine the required flow for the analysis.
ASSOCIATED LCO:
m
_ - _ _ _ _ _ _. - - - - -