ML19312C406
| ML19312C406 | |
| Person / Time | |
|---|---|
| Site: | Oconee  (DPR-38-A-037, DPR-38-A-37, DPR-47-A-037, DPR-47-A-37, DPR-55-A-034, DPR-55-A-34) |
| Issue date: | 02/04/1977 |
| From: | Goller K, Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C397 | List: |
| References | |
| NUDOCS 7912130929 | |
| Download: ML19312C406 (14) | |
Text
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
s O p*s\\
oN.1 > s1 1 a NUCLEAR REGULATORY COMMISSION y
j j
WASHINGTON, D. C. 20666 i
i
%,
- s < * /
DUKE POWER COMPANY e
DOCKET NO. 50-269 y
OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No. DPR-38 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated June 11, 1976, 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 confomity with the application, tne 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 l
l 7912130 N 7
~
o
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of the Facility License No. DP1-38 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Anendment No. 37, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license anendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0 tit 11SSION
/
.jl.
Karl R. Goller, Assistant Director for Operating Reactors Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 4,1977 1
1 ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 37 TO DPR-38 AMENDMENT NO. 37 TO DPR-47 AMENDMENT NO. 34 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:
Remove Pages Insert Pages 2.1-2 2.1-2 2.1-7 2.1-7 2.3-2 2.3-2 2.3-3 2.3-3 2.3-4 2.3-4 2.3-8 2.3-8 2.3-11 2.3-11 l
i
s can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DNB for The local DNB axially uniform and non-uniform heat flux distributions.
ratio (DNBR), defined as the' ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin The minimum value of the DNBR, during steady-stste operation, normal to DNB.
operational transients, and anticipated transients is limited to 1.30.
A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions.
The difference between the actual core pressure and the indicated reactor coolant system pressure has been outlet The difference considered in determining the core protection safety limits.
in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setponts to correspond to the elevated location where the pressure is actually measured.
The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.30 is predicted for the maxi =um possible thermal power (112 percent)whenfourreactorcoolantpumgsareoperating(minimumreactor flow is 107.6 percent of 131.3 x 10 lbs/hr.).
This curve is based on c oolant the combination of nuclear power peaking f actors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.
The curves of Figure 2.1-2A are based on the more restrictive of two thermal and rod bowing:
limits and include the ef fects of potential fuel densification The 1.30 DNBR limit produced by the combination of the radial peak, axial 1.
1.30 DNBR.
peak and position of the axial peak that yields no less than a The combination of radial and axial peak that causes central fuel melting 2.
at the hot spot.
The limit is 20.15 kw/ft for Unic 1.
Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.
The specified flow rates for Curves 1, 2, 3 and 4 of Figure 2.1-2A correspond to the expected minimum flow rates with four pumps, three pumps, one pump in each loop and two pumps in one loop, respectively.
The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.
The maximum thermal power for three-pump operation is 86.4 percent due to a 74.7 percent flow x 1.07 =
power level trip produced by the flux-flow ratio error. The 79.9 percent power plus the maximum calibration and instrument maximum thermal power for other coolant pump conditions are produced in a si=11ar manner.
2.1-2 Amendments Nos. 37, 37 & 34
inercal Po:er Level, %
UNACCEPTA8LE OPERATION
.120
(-2s. ii 2 )
(
2)
(+ 3.
2)
.110 ACCEPTABLE 4 PUNP 100 OPERAT10N 1
I *" ' " I
/
90
( -37. 90 )
( +2 s.86. 4)
(-29 86,43 (86.4) 60 ACCEPTABLE 70 PUNP
( +40.64. 4) 2 OPERAT10N
( -3 7. 64. 4)
( 29 58.9)
. 60 (58.9)
(+28 58.9)
ACCEPTABLE
. 50 2,3 & 4 PUNP OPERATION 40
( -3 7. 36. 9) 3&4
(+40,36.9) 30 20 10 a
I t
-60
-40 20 0
+20
+40
+60 Reactor Power imoalance, 5 CURVE REACTOR COOLANT TLOW (ID/hr) 6 I
141.3 x 10 6
2 105.6 x 10 6
3 69.3 x 10 6
4 64.7 x 10 LlllT 1 l
CORE PROTECTION SAFETY LIMIT l
aim OCONEE NUCLEAR STATION I
Ficas2.1-2A 2.1-7 Amend ents Nos. 37, 37 & 14 l
4 During normal plant opsratinn with all rsector ecolant pu.aps operating, reactor trip is initintcd wh:n th2 rcceter p; war 1sval racchts 105.5% of Adding to this the possible variation in trip setpoints due
' rated power.
to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is more conservative than the value used in the safety analysis. (4)
Overpower Trip Based on Flow'and Imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the severe thermal transient considered in the design, the loss-of-coolant most flow accident from high power.
Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DN3R of less than L.3 should a low flow condition exist due to any electrical malfunction.
The power level trip set point produced by the power-to-flow ratio provides the reactor power both high power level and low flow protection in the event level increases or the reactor coolant flow rate decreases.
The power level trip set point produced by the power-to-flow ratio provides overpower DN3 pro-tection for all modes of pump operation.
For every flow rate there is a maxi-mum permissible power level, and for every power level there is a minimum permissible low flow rate.
Typical power level and low flew rate corbinations for the pump situtations of Table 2.3-1A are as folicws:
Trip would occur when four reactor coolant pumps are operating if power 1.
is 107%
and reactor flow rate is 100%, or flow rate is 93.5% and power l
level is 100".
Trip would occur when three reactor coolant pumps are operating if power l
2.
is 79.9% and reactor flow rate is 74.7% or flow rate is 70.1% and power level is 75%.
Trip would occur when two reactor coolant pumps are operating in a single 3.
loop if power is 52.4% and the operating loop flow rate is 54.5% or flow rate is 47.9% and power level is 46%.
Trip would occur when one reactor coolant pump is operating in each loop 4.
(total of two pumps operating) if the power is 52.4% and reactor flow rate is 49.0* or flow rate is 45.8% and the power level is 49%.
The flux-to-flow ratios account for the maximum calibration and instru-mentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flew.
For safety calculations the maximum calibration and instrumentation errors for the power level trip were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits frca being exceeded.
These thermal limits are either power The reactor power imbalance (power in limits or DN3R li=its.
peaking kw/ft the top half of core minus power in the bottom half of core) reduces the power the boundaries cf level trip produced by the power-to-flow ratio such that Figure 2.3-2A
- Unit 1 are produced.
The power-to-flow ratio reduces the power j
2.3 Unit 2 2.3-2C - Unit 3 i
l 2.3-2 Amendments Nos. 37, 37 & 34
level trip and associated reactor power / reactor power-imbalance boundaries by 1.07%
.for a 1% flow reduction.
The power-to-flow reduction. ratio is 0.961 during single loop operation, j
Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio. The pump monitors also restrict the power level for the number of pumps in operation.
Reactor Coolant System pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear over-power trip set point. The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (1)
The low pressure (1800) psig and variable low pressure (11.14 T
-4706) trip (10. 79 T "U-4539)
(1800) psig (10.79 T "U-4539)
(1800) psig setpoints shown in Figure 2.3-1A have been established to maintaSn the DNB 2.3-1B 2.3-1C ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3)
Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T
-4746) 9 (10.79 T
-4579)
(10,79 T " -4579) u Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-1B 2.3-1C temperatures in the operatinq range.
Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.
Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
2.3-3 Amendments Nos. 37, 37 &34
s Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing.
and startup procedures, there is provision for bypassing certain segments of The reactor protection system segments which the reactor protection system.
Two conditions are imposed when can be bypassed are shown in Table 2.3-1A.
2.3-1B 2.3-1C the bypass is used:
By administrative control the nuclear overpower trip set point must be 1.
reduced to a value < 5.0% of rated power during reactor shutdown.
l 2.
A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal This high operation with part of the reactor protection system bypassed.
pressure trip set point is lower than the normal low pressure trip set point The over so that the reactor must be tripped before the bypass is initiated.
reactor power from power trip set point of < 5.0% prevents any significant Sufficient natural being produced when performing the physics tests.
circulation (5) would be available to remove 5.0% of rated power if none of the reactor coolant pumps were operating.
Two Pump Operation A.
Two Loop Operation Operation with one pump in each loop will be allowed only following reactor shutdown.
After shutdown has occurred, reset the pump contact monitor power level trip setpoint to 55.0%.
B.
Single Loop Operation Single loop operation is permitted only after the reactor has been After the pump contret monitor trip has occurred, the following tripped.
actions will permit single loop operation:
to 55.0%.
the pump contact monitor power level trip setpoint Reset 1.
Trip one of the two protective channels receiving outlet temperature 2.
information from sensors in the Idle Loop.
3.
Reset flux-flow setpoint to 0.961.
t REFERENCES (1) FSAR, Section 14.1.2.2 (4) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.7 (5) FSAR, Section 14.1.2.6 (3) FSAR, Section 14.1.2.8 2.3-4 Amendments Nos. 37, 37 & 34
'S
-s P0cer Level, v 5 120 UmACCEPTABLE OPERATE 0N (107)
+
d y
l 100
)
I l
ACCEPTABLE l
- a
//
4 PUMP OPERAT10N g
80 (79.9)
^
(33'79)
(-28,79) l I
I I
ACCEPTA3LE l
3& 4 PUMP
_g l
OPERAT;0M
('
(3i Si S)
( _2 s. s i. e 1
i I
40 ACCEPTABLE 2.3 & 4 PUMP l
0 RATION
(-28.24.4)
{3i 2# %)
20 l
1 1
e
,r ir l ni s
i
- 1
- 1 I
i i
i i
60
-40 20 0
+20
+40
+60 I
Reactor P0wer imoalance. ",
i UNIT 1 2.38 PROTECTION SYSTEM MAXIMUM (b,d okoh0ggft,{QPpgig FIGURE 2.3-2A l
Amendments Nos. 37, 37 & 34
l l
Table 2.3-1A Unit 1 l
Reactor Protective System Trip Settina Limite i
1%o Beactor One Reactor l
Four Reector Three Reactor Coolant Pumpo Coolant Pump Coolant Pumpe Coolant Fumpa Operating to A Operating in Operating Operating Single Loop Each Loop (Operating Power (Operating Power (Operating Power (Operating Fower Shutdown RFS Seament
-1001 Rated)
-75I Rated) _
-461 Rated)
-491 Rated) synese I3) 1.
Nuclear Power Ham.
105.5 105.5 105.5 105.5 5.0 j
(I mated) 2.
Nuclear Power Max. Based
[1.07 times flow 1.07 times flow 0.961 times flow 1.07 times flow Sypaeeed on Flow (2) and 1mbalance, minue reduction minua reduction minue reduction minue reduction (1 Rated) due to imbalance due to imbalance due to imbalance due to fabalance 3.
Nuclear Power Max. Based NA NA 55Z (5)(6) 551 (5) sypeseed on Pump Honitors, (I, Rated) 4.
High Reactor Coolant 2355 2355 2355 2355 1720(4)
System Pressure, pelg, Max.
p
'y 5.
Iow Reactor Coolant I800 1800 1800 1800 Bypeseed p
System Pressure, pois Kla.
( 11.14 T, _470e)(1) ( l 1.14 T,,g-47M ) W (11.14 T,g-4106)m (Ia.14 T,,g-4706 %sypeesed 6.
Variable im Reactor g
Coolant System Fressure e
pets, Min.
3 7.
Reactor coolent Temp.
619 619 619 (6) 619 619 g
r., u.a.
\\
d,
'l 8.
High Besctor Building 4
4 4
4 4
n (w)
Fressure, peig, Man.
(1) 7,,g to in degrees Fahrenheit (*F).
(5) Reector power level trip est point produced by pump contact monitor reset to 55.01.
g g w
N l
(2) Reactor Coolant System Flow, 1.
(6) Specification 3.1.8 applies. Trip one of the
'M j
w two protection chamaele receiving outlet temper-(3) Administratively controlled reduction set ature informattee'from eenoore la the idle loop.
g G"
only during reactor ebutdown.
i m
i (4) Automatically set when other segments of the RFS are bypaesed.
b 4
l
0 UNITED STATES
/
%g NUCLEAR REGULATORY COMMISSION y
WASHINGTON, D. C. 20666
{
j
\\'1 DUKE POWER COMPANY e
00CXET NO. 50-270 s
OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No. DPR-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated June 11, 1976, 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 puolic; and E.
The issuance of this ameadment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
q
. 2.
Accordingly, the license is amended by changes to the Tech.7' cal Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of the Facility License No. DPR-47 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Aopendices A and B, as revised through Amendment Po. 37, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Soecifications.
3.
This license anendment is effective as of the date of its issuance.
FOR THE MUCLEAR REGULATORY COMMISSION N:wn
(
.r A. Schwencer, Chief Operatina Reactors Branch 41 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: February 4,1977 a
- m.
a.
~
T
./c/
$*t NUCLEAR REGULATORY COMMISSION UNITED STATES l
j e
WASHINGTON, D. C. 20656
\\, *-
/
...+
DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 34 License No. DPR-55 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duke Power Comoany (the licensee) dated June 11, 1976, 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.
l i
1 l
l
']
')
, 2.
Accordinoly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of the Facility License No. OPP-55 is hereby amended to read as follows:
(2) Technical Specific,:tions The Technical Soecifications contained in Appendices A and B, as revised through Amendment No. 34, are hereby incoroorated in the license.
The licensee shall ooerate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE N!JCLEAR PEGtlLAT00.Y COMt1ISSION fWWW
}
A A. Schwencer, Chief Operating Reactors Branch *1 Division of Operating Reactors
Attachment:
Changes to.
rechnical 4
Specifications Date of Issuance: February 4,1977 f
l'
_