ML19312C415
| ML19312C415 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/04/1977 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C411 | List: |
| References | |
| NUDOCS 7912130937 | |
| Download: ML19312C415 (41) | |
Text
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%,4 UNITED STATES NUCLEAR REGULATORY COMMISSION Q }...
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WASHINGTON. D. C. 20655 y
a DUKE POWER COMPANY 00CKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO.1 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No.47 License No. DPR-38 1.
The U.S. Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated March 30, 1977, as supplemented June 21, August 23, September 8, and 14,1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Cor: mission'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 Comission; C.
There is reascnable assurance (i) that the activities authorized by this amendnent can be conducted without endangering the health and safety of the public,and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of tMs amendment is in accordance with 10 CFR Part 51 of the Cemission's regulations and all applicable require-ments have been satisfied.
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 Facility License No. DPR-38 is hereby amended to read as follows:
99/2.13df 5 7
)
-2
"(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 4; 3re hereby incorporated in the license.
The lices.see shall operate the facility in accordance with the Technical Specifications."
3.
This license amendment is effective as of its date of issuance.
FOR THE NU LEAR REGULATORY COMMISSION
,s'b'!NWCQ__
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}.
A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: October 4, 1977 i
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j UNITED STATES
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NUCt. EAR REGULATORY COMMISSION y^
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W ASHINGTON, D. C. 20666
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DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.47 License No. DPR-47 1.
The U.S. Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated March 30, 1977, as supplemented June 21, August 23, September 8, and 14,1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the prm isions of the Act, and the rules and regulations of the Corcission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can b.' conducted without endangering the health and safety of the public,and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable require-ments have been satisfied.
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 Facility License No. OPR-47 is hereby amended to read as follows:
?
e 2-
"(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 47, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications."
2.
This license amendment is effective as of its date of issuance.
FOR THE NUC EAR REGULATORY COPHISSION k'lt#ddQ
.'s. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 4, 1977 1
M 4
UNITE 3 STATES
/
'f NUCLEAR REGULATORY COMMISSION k
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o WASHINGTON, D. C. 20565
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OUKE POWER COMPANY 00CKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENCMENT TO FACILITY OPERATING LICENSE J
Amendment No. 44 License No. DPR-55 1.
The U.S. Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated March 30, 1977, as supplemented June 21, August 23, September 8, and 14, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable require-ments have been satisfied.
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 Facility License No. DPR-55 is hereby amended to read as follows:
'}
}
. "(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 44, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications."
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
(l
'kild$$L A. Schwence, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 4, 1977
9 ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 47 TO OPR-38 AMENDMENT NO. 47 TO OPR-47 NMENDMENT NO. 44 TO DPR-55 00CKET NOS. 50-269, 50-270 and 50-287 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages.
2.1-1 3.5-8 5.3-1 2.1-2 3.5-9 2.1 -3 3.5-10 2.1-4 3.5-11 2.1-7 3.5-12 2.1-10 3.5-13 2.3-1 3.5-18 2.3-2 3.5-18a 2.3-3 3.5-21 2.3-4 3.5-21a 2.3-11 3.5-24 2.3-12 4.1-9 3.1-14 3.1-15 2.
Add pages:
3.5-13a 3.5-18b 3.5-21b 3.5-23c 3.5-23d 3.5-23e l
e r,
2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant syste= pressure, coolant temperature, and coolant flow during power operation of the plant.
Objective To maintain the integrity of the fuel cladding.
Specification The combination of the reactor system pressure and coolant temperature shall exceed the safety limit as defined by the locus of points established in not Figure 2.1-1A-Unit 1.
If the actual pressure / temperature point is below 2.1-13-Unit 2 2.1-lC-Unit 3 and to the right of the line, the safety limit is exceeded.
The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2A-Unit 1.
If the actual reactor-thermal-power / power 2.1-23-Unit 2 2.1-2C-Unit 3 imbalance point is above the line for the specified flow, the safety limit is exceeded.
)
Bases - Unit 1 The safety it=1ts presanted for Oconee Unit I have been generated using BA'4-2 critical heat flux (CHF) correlation (1)
The reactor coolant syste= flow rate utilized is 10o.5 percent of the design flow (131.32 x 10" lbs/hr) based on four-pump operation.(2) l To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DN3).
At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure.
Although DN3 is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure 2.1-1 Amendments 47, 47 & 44
r
)
J The BAW-2 can be related to DN3 through the use of the BAW-2 correlation (1).
correlation has been developed to predict DN3 and the location of DN3 for flux distributions.
The local DN3 exially uaiform and non-uniform heat flux that would cause DN3 at a ratio (DNBR), defined as the ratio of the heatflux, is indicative of the margin particular core location to the actual heatThe mini =um value of the DN3R, durin to DN3.
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 this is considered a conservative margin to level that DN3 will not occur; The difference between the actual core DN3 for all operating conditions.
pressure and the indicated reac:or coolant system pressure has been outlet The difference considered in determining the core protection safety lisi:s.
in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed.4 n reducing the pressure trip setponts to correspond to the elevated location where the pressure is actually measured.
The curve presen:ed in Figure 2.1-1A represents the conditions at which a mini =u= DN3R of 1.30 is predic:ed for the maxi =u= possible ther=al power coolant pu=es are operating (mini =um reactor (112 percent) when four reac:ct c oolant flew is 106.5 percen: of 131.3 x 10' lbs/hr.).
This curve is based en l the ccebination of nuclear power peaking f actors, with potential ef fects of fuel in a more conservative DNBR than any densification and rod bewing, which result other shape that exists during normal operation.
The curves of Figure 2.1-2A are based on the = ore res:rictive of two thermal and rod bowing:
li=1ts cnd include the ef f ects of potential fuel densification The 1.30 DNBR limit produced by the co=bination of the radial peak, axial 1.
1.30 Ds3g.
peak and position of the axial peak that yields no less than a The cembination of radial and axial peak that causes central fuel melting 2.
at the hot spot.
The lini: is 20.15 kw/ft for Unit 1.
a directly observable quantity and therefore limits have Power peaking is not been established on the bases of the reactor power imbalance produced by :he power peaking.
The specified fiev ra:es for Curves 1, 2, 3 and 4 of Figure 2.1-2A correspond to the expec:ed =inimum flow rates wi:h four pumps, :hree pu=ps, one pu=p in each loop and :wo pu=ps in one loop, respectively.
The curve of Figure 2.1-LA is :he mos: restric:ive of all possible rea::c coolant pe=p-=axi=um :hermal power cembinations shown in Figure 2.1-3A.
The :ay.1=u= :her=al ;cwer f or three-pusp opera:1on is 85.3 per:en: due :: a 74.7 percen: ficw x 1.055 =
pcwer level : rip produced by :he flux-flev ra:ic error.
The power plus :he maxi =um :alibra: ion and instrument 78.S percen:
maximum thermal power f:: other coolan: pump conditions are produced in a similar sanner.
2.1-2 Amendments 47, 47 &44
. Tor Figuro 2.1-3A, o pr"r-urc-temp;raturo point cb va cad to th31Cf t of th3 curva would rotult i. a DNBR grecter than 1.30.
References (1)
Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.
(2) Oconee 1, Cycle 4 - Reload Report - BAW-1447, March, 1977.
l 2.1-3 Amendments 47, 47 & 44
7
.s' 2400 2200 2
E a
E 2000
/
E 1800 r
1600 560 580 600 620 640 Reactor Coolant Outlet Temperature,F CORE PROTEC7 0N SAFETY LIMITS. Uf; 7 1 2.1-4
.ncit m o, OCONEE NUCLEAR STATION Figure 2.1-IA
?
THERalAL POWER LEVEL. 5
_ _ 120
( 37 2.337)
(22.1.112)
,,g ACCEPTA&E 4 100 p,
( 40.98)
OPERATION 90 (50.80)
__ 80 ACCEPTABLE 2
i 40.72) 70 3g4pp OPERATION 60 3
50 ACCEPTABLE t 40.44 21 2.3 & 4 40 P W OPERATION 30 (50.25 0) 20
-- 10 1
i 60 40 20 0
+20
+40
+60 Reactor Poner imoalance. >
C'JRV E REACTOR COOL ANT FLOW ( GPw n 1
374880 2
230035 3
183650 204310 4
CORE PROTECTIO. SAFETY N
LIMITS, UNIT i
,'8 et* o OCONEE NUOLEAR STATION l
%f 2.1,/
'~
Figure 2.1-2A Amendments 47, 47 & 44
,e,
2400 ACCEPTABLE
~
OPERAft0N 4
3
- 2 2 2000
/
3 1
2 3
~
0 1800 r
r 1500 560 560 500 620 640 4es: tor Coolant Outlet Temperature F CURVE RE ACTOR COOL ANT FLOW (GP4)
POWER PUNPS OPER ATING < TYPE OF LIMIT) 1 374880 (100 0 '
112i 4
(DNBR) 2 290035 (T4 7'.)
SS 7'.
3 (DNBR) 3 183690 (43 05) 59 03 2
(00ALITT)
' 106 5', OF FIRST CORE DESIGN FLOW CORE PROTECTION SAFETY LIMITS, UN F i b
botnaal OCONEE NUCLEAR STATION
- . 10 M
Figure 2.1-3A Amendments 47, 47 & 44 l
l l
l
LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 2.3 A g icability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolsnt outlet temperature, flow, nu=ber of pumps in operation, and high reactor building pressure.
Objective To provide automatic protective action to prevent any combination of precess variables from exceeding a safety limit.
Specification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1A - Unit 1 and 2.3-1B - Unit 2 2.3-1C - Unit 3 Figure 2.3-2A - Unit 1 2.3 Unit 2 2.3-2C - Unit 3 Loss of one pump during four-pump operation if power level is a.
greater than 80*.' of rated power.
5.
Loss of two pumps and reactor power level is greater than 55% of rated power. (Power /RC pump trip setpoint is reset to 55% for operation with one pump in each loop).
Loss of two pumps in one reactor coolant loop and reactor power level i c.
greater than 0.0% of rated power.
d.
Loss of one or two pumps during two-pump operation.
Bases The reactor protective system consists of four instru=ent channels to monitor gach of several selected plant conditions.which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a saf ety limit may be reached.
The trip setting limits for protective system instrumentation are listed in Table 2.3-1A - Unit 1.
The safety analysis has been based upon these protective 2.3 Unit 2 2.3-1C - Unit 3 system instrumentation trip set points plus calibration and instrumentation errors.
Nuclear Overpcuer A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
2.3-1 Af9/20sR 014n4;6 Amendments 47, 47 & 44 D* FADTD Nh M=
e l
D D ~3 Th aMo a Enb During normal plant operation with all t eactor coolant pumps operating, reactor trip is initiated when the reae:or power level reaches 105.5% of rated power.
Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is more conserva:ive than the value used in the saf ety analysis. (4)
Overpower Trip 3ased on Ficw and Imbalance The power level trip se: poin produced by the reae:or coolan: system flew is based on a power-to-flow ratio which has been established :o acco=medate the most severe thermal ::ansien: considered in :he design, the loss-of-coolan:
flow accident from high power.
Analysis has demonstrated tha: the specified power-:o-flow ratio is adequate to prevent a DN3R ef less than L.3 should a low flow condicion exist due to any electrical malfunc: ion.
The pcwer level trip set point produced by the power-to-flow ra:1o provides both high power level and low flow protec: ion in the event the reactor pcwer level increases or the reactor coolant flew rate decrease:.
The power level trip set poin: produced by the power-:o-flow ra:io prevides overpcwer DN3 pro-taction for all modes of pu=p operation.
For every flow ra:e :here is a =axi-mum permissible pcwer level, and for every pcwer level :here is a minimu=
permissible low flow ra:e.
Typical pcwer level and low flow ra:e combinations for :he pump si:u:2:1:ns of Table 2.3-1A are as follcws:
1.
Trip would occur when four reac:or coolant pumps are cpera:ing if power is 105.5% and reacco flow race is 100%, or flow ra:e is 94.8% and powcr l
level is 100%.
2.
Trip would occur when three reae:c: coolant pumps are opera:ing if pcwer is 78.8% and reacect flow rate is 74.7% or flew rate is 71.1% and power l
level is 75%.
3.
Trip would occur when two reactor coolant pumps are opera:ing in a single loop if power is 51.7% and the opera:ing loop flow ra:e is 54.5" or fles rate is 48.5% and power level is 46%.
4 Trip would occur when one reactor coolant pump is operating in each icop (total of two pumps opera:ing) if the power is 51.7% and reae:ct flow rate is 49.0% or flow race is 46.4% and the pcwer level is 49%.
The flux-to-flow ratios account for the maximum calibration and instrument errors and the max 1=um variation from the average value of the RC flew signal in such a manner tha: the reactor protective system receives a conservative indication of the RC flow.
For safety calcula: ions the maximum calibration and instru=enta: ion errors for the power level trip were used.
The power-i= balance boundaries are established in order to preven: reactor thermal limits f rom being exceeded.
These thermal limics are ei:her pover peaking kw/f: limi:s or DN3R 11=1:s.
The reac:or pcwer i= balance (pewer in
- he top half of core =inus power in the bo::cs half of core) reduces the power level trip produced by :he power-to-flow ra:io such that the boundaries of Uni: 1 are p;cduced.
The pcwer-to-flew ratio reduces the pcwer Figure 2.3-1A 2.3 Unit 2 2.3-2C - Unic 3 l
2.3-2 Amendments 47, 47 & 44
level trip and associated reoctor power / reactor power-imbalance boundaries by 1.055';-Unit 1 for a 1% flow reduction.
/
1.055%-Unit 2 1.07% -Unit 3 For Units 1 and 2 the power-to-flow reduction ratio is 0.949, and for Unit 3, the power-to-flow reduction factor is 0.961 during single loop opera tion.
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 en 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 (1800) psig (11.14 T "'-4706)
(1800) psig (10.79 I ut-4539) setpoin'.s shown in Figure 2.3-1A have been established to maintain 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)
(11.14 T'"ut-4746) o (10.79 Tout -'3/91 Coolanc Outlet Tempe ra t ure 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 tcmperatures in the operating range.
Due to calibration and instrumentation orrors, 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 cystem pressure trip.
2.3-3 Amendments 47, 47 & 44
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.
The reactor protection system segments which can be bypassed are shown in Table 2.3-1A.
Two conditions are imposed when 2.3-1B 2.3-1C the bypass is used:
1.
By administrative control the nuclear overpower trip set point must be reduced to a value < 5.0% of rated power during reactor shutdown.
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 nor=al operation with part of the reactor protection system bypassed.
This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated.
The over power trip set point of < 5.0% prevents any significant reactor power from being produced when performing the physics tests.
Sufficient natural 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.
TVo 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 tripped.
Af ter the pump contact monitor trip has occurred, the following actions will permit single loop operation:
1.
Reset the pump contact monitor power level trip setpoint to 55.0%.
2.
Trip one of the two protective channels receiving outlet temperature information from sensors in the Idle Loop.
3.
Reset flux-flow setpoint to 0.949 (Unit 1) 0.949 (Unit 2) 0.961 (Unit 3)
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 47, 47 & 44
THERMAL POWER LEVEL. S j
120 1
110 i
(105.5) 9 l
100)
+p.
,'3 y
l 90 I
(30.87)
ACCEPTABLE
(.40,84) 4 PUNP l
80 (70 0)
~~
OPERATION l
l 70l I
l CCEPTABLE 60l (30.60.3)
( 40.57 3) 3 & 4 PUMP l
OPERATION l
I(51.7) l 50l l
40l l
ACCEPTABLE l
2,3 & 4 l
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OPERAil0N l
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OR0TECTIVE SYSTEM MAXI"L'M ALLOWASLE SEIDOINTS Mg U*1IT 1 l
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- t OCONEE NUCLEAR STATION Figure 2.3-2A
,,3_3 Amendments 47, 47 & 44 l
r-l i
g Reactor Protective System Trip bet.ing 1.inits TVo Reactor One Reactor Four Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumpe Operating in A Operating in
- Operating Operating Single 1.oop Each Loop (Operating Power (Operating Power (operating Power (Operating Power Shutdown FPS Segment
-1001 Itat ed)
-75Z Rated)
-4f,1 kated)
-491 Rated)
Bypass I3) 1.
Eclear Power Ns.
11;5. 5 105.5 105.5 105.5 5.0 (I kated) 2.
Eclear Power N s.
Samed 1,0% times glow I,0% timen flow o,949 tg e.
glow I.On timen glow typassed on Flow (2) and Imbalance, minus reduction minus reduction minue reduction minus reduction (I Rated) due to imbalance due to imbalance due to imbalance due to imbalance 1.
Nclear Power h u. Based NA M
SSI (5)(6) 551 (5) typassed on Pump Monitorm. (I, Rated) 4.
High Reactor Coolant 2355 2355 2355 2155 1720 'I I
System Pressure, pets, ha.
Fy 5.
Iow Beactor Coolant 1800 1800 1800 1800 sypassed p
System Pressure, pels, Hin.
III (ll.14 T
- 4706 )
(11.14 T
- 4 706 ) ( }
6.
Variable law Beactor
( 11.14 T
_4 70f9
( 11.14 T,,g-4 706 )( '
BYPaesed at out b
Coolant System Pressure M
pulg. Hin.
9k 7.
Reactor Coolant Temp.
619 619 619 (6) 619 619 F., N s.
rt
- 4 4
4 8.
High Reactor Building 4
4 A
Pressure, pelg, Nm.
i 3
u M
(1) T le in degrees Fahrenheit (*F).
(5) Beactor power level trip set point produced by pump contact monitor reset to 55.02.
p (2) Reactor Coolant System Flow, I.
(1) Administratively controlled reduction set two protection channele receiving outlet temper-only during reactor shutdown.
ature information from sensors in the idle loop.
(4) Automatically set when other segmente of the RPS are bypassed.
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e e-2.F12 Amendments M AV & dt 47, 47 & 44 X
n % r a c Ti
~
3.1.6 Leakage Soecification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gps, the reactor shall be shutdown within 24 h'ours of detection.
3.1.6.2 If uniden:ified reactor coolant leakage (excluding normal evapora-tive losses) exceeds 1 gpm or if any reactor coolant leakage is evalua:ed as unsafe, the reactor shall be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.3 If any reactor coolant leakage exists through a non-isolable f ault in a RCS strength boundary (such as the reactor vessel, piping, valve body, etc., except the steam generator tubes), the reactor shall be shutdown, and cooldown to the cold shutdevn condition shall be ini:iated withir. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.4 If at any time, the leakage through the Unit 1 steam generator tubes equals or exceeds 0.3 gpm, a reactor shutdewn shall be initiated within 4 nours and and the reactor shall be in a cold condition within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
If the leakage is less than 0.3 gpm, as assessment shall be made whether operations may be continued safely or the plant should be shutdown.
In either case, the NRC shall be notified in accordance with Section 6.6.21.
3.1.6.5 If reae:or shutdown is required by Specification 3.1.6.1, 3.1.6.2 or 3.1.6.3, the rate of shutdown and the conditions of shutdown shall be determined by the safety evaluation for each case and justified in writing as soon theraafter as practicable.
3.1.6.6 Accion to evalua'te the safety implication of reactor coolant leakage shall be inicia:ed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection.
The nature, as well as :he magni:ude, of the leak shall be considered in this evaluation.
The safety evaluation shall assure that the exposure of offsi:e personnel to radia:1on is wi:hin the guidelines of 10CTR20.
3.1.6.7 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2, 3.1.6.3 or 3.1.6.4, the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected.
3.1.6.8
~4 hen the reac:or is critical and above 2% power, two reactor coolant leak detection sys: ems of different operating principles shall be operable, with one of the two systems sened-ive :o radioactivity.
The systems sensitive to radioactivity any be out-of-service for j
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided two other means to detect leakage are operable.
3.1.6.9 Loss of reactor coolant through reactor coolant pump seals.utd sys:em valves to connec:ing systems which vent :o :he gas v.st header and from which coolant can be returned to the reactor ecol-ant system shall not be censidered as reac::r coolan: leakage and j
shall not be subjec: to the consideration of Specifications 3.1.6.1, 3.1.6.2, 3.1.6.3, 3.1.6.4 3.1.6.5, 3.1.6.6 or 3.1.6.7 except that such losses when added to leakage shall not ex:eed 30 gpm.
3ases Every reasonable effort will be made to reduce reactor coolant leakage in-ciuding avaporative losses (which may be on the order of.5 g;m) to the loves: possible rate und at leas below 1 gpm in order to prevent a large Amendments 47, 47 & 44
u leck freo masking th3 proacnco of a smallcr leak. W2ter i tutory b31cacca,
~ '
rcdictica monitoring cquipment, boric ccid crystallina d: posits, cod physicci inspections can disclose reactor coolant leaks. Any leak of r:diocctiva fluid, whether from the reactor coolant system primary boundary or not can be a to in-plant radioactivity contamination and serious problem with respect could develop into a still more serious problem; and therefore, cleanup or it indications of such leakage will be followed up as soon as practicable.
firs 6 Although some leak rates on the order of GPM may be tolerable from a dose point of view, especially if they are to closed systems, it must be recog-leaks in the order of drops per minute through any of the walls nized that of the primary system could be indicative of materials failure such as by stress corrosion cracking.
If depressurization, isolation and/or other safety measures are not taken promptly, these small breaks could develop into Therefore, the nature much larger leaks, possibly into a gross pipe rupture.
of the leak, as well as the magnitude of the leakage must be considered in the safety evalua:ica.
When the source of leakage has been identified, the situation can be evaluated This evaluation will be per-to determine if operation can safely continue.
formed by :he Operating Staf f and will be documented in writing and approved Under these condicions, an allowable reactor coolan:
by the Superintendent.
This explained leakage system leakage ra:e of 10 gym has been established.
race of 10 gym is also well within the capacity of one high pressure injection pu=p and =akeup would be available even under the less of off-site j
power condition.
If leakage is to the reactor building 1: =ay be identified by one or more of the following methods:
4 The reaccer building air particulate monitor is sensitive to low leak a.
The rates of reactor coolant leakage to which the instrument rates.
is sensi:ive are.10 gpm to grea:er than 30 gym, assuming corrosion ac:ivity and no fuel cladding leakage. Under these conditions, produe an increase in coolant leakage of 1 gpm is detectable within 10 minutes af:er it occurs.
as sensi:1ve The iodine =oni:or, gaseous magiter and area nonitor are not b.
to corrosion product ac:ivity.tl)
It is calculated that the iodine i
monitor is sensitive to an S gym leak and the gaseous monitor is sen-sitive to a 230 gpm leak based on the presence of tramp uranium (no fission products frem tramp uranium are assumed to be present).
- However, any fission produe:s in :he coolant will make these moni: ors more sensi:1ve to coolan: leakage.
In addition to the rad: e. ion moni: ors, leakage is also monitored by a-level indicator in the feactor building normal sump. Changes in normal c.
sump level say be indicative of leakage from any of :he sys: ems located inside :he reaccor building such as reactor coolant systes, low pressure service wa:er sys:em, component cooling system and steam and feedwater lines or condensation of humid 1:7 wi:hin the reactor building atmosphere.
The sump capacity is 15 gallons per inch of height and each gradua:1on on :he level 'ndicates 1/2 inch of sump heigh:.
This indica:or is
)
capable of detec:ing changes on the order of 7.5 gallons of leakage in:o A 1 gym leak would therefore be detec:able within less than 2
4
- he susp.
10 minu:es.
3.1-15 l,
(3) Except as provided in specification 3.5.2.4.b, the reactor shall be brought to the hot shutdown condition within four hours if the quadrant power tilt is not reduced to less than 3.41% Unit I within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.41% Unit 2 3.41% Unit 3 b.
If the quadrant tilt exceeds +3.41% Unit 1 and there is simultaneous 3.41% Unit 2 3.41% Unit 3 indication of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60%
of the thermal power allowable for the reactor coolant pump combination.
c.
Except for physics est, if quadrant tilt exceeds 9.44% Unit 1, 9.44% Unit 2 9.44% Unit 3 a controlled shutdown shall be initiated immediately, and the reactor shall be brought to the hot shutdown condition within four hours.
d.
'4henever the reactor is brought to hot shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each I percent tilt for the maximum tilt observed prior to shutdown.
e.
Quadrant power tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.
3.5.2.5 Control Rod Positions Technical Specification 3.1.3.5 does not prohibit the exercising a.
of individual safety rods as required by Table 4e 1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.
Except f or physics tests, operating rod group overlap shall be 25% + 5%
b.
between two requential groups.
If this limit is exceeded, corrective measures shall be taken immediately to achieve an acceptable overlap.
Acceptable overlap shall be attained within two hours, or the reactor shall be placed in a hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Positica limits are specified for regulating and axial power c.
shaping control rods.
Except for physics tests or exercising control rods, the regulating control rod insertion /withdecwal limits are specified en figures 3.5.2-1A1, 3.5.2-1A2 and 3.5.2-1A3 (Unit 1) ; 3.5.2-131, 3.5.2-132 and 3.5.2-13 3 (Unit 2) ;
3.5.2-1C1, 3.5.2-1C2 and 3.5.2-1C3 (Unit 3) for four pump operation, and on figures 3.5.2-2A1, 3.5.2-2A2 and 3.5.2-2A3 (Unit 1); 3.5.2-231, 3.5.2-232 and 3.5.2-2B3 (Unit 2) ;
3.5.2-201, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for two or three Amendments 47, 47 & 44 w
pump operation.
Also, exceptira physics tes2s or exercising control rods, the axial power shaping control rod insertion /
withdrawal limits are se ffied on figures 3.5.2-4A1, 3.5.2-4A2 and 3.5.2-4A3 (Unit 1) ano on figures 3.5.2-4B1, 3.5.2-4B2, and 3.5.2.483 (Unit 2).
If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position.
An acceptable control rod position shall then be attained within two hours.
The minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all times.
J.
Except for physics tests, power shall not be increased above the power level cutof f as shown on Figures 3.5.2-1A1, 3.5.2-1A2,3.5.2-1 A3, (Unit 1), 3.5.2-1B1, 3.5.2-182, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, 3.5.2-1C3 (Unit 3), unless the following requirements are met.
(1)
The xenon reactivity shall be within 10 percent of the value for operation at steady-state rated power.
(2)
The xenon reactivity shall be asymptotically approaching the value for operation at the power level cutoff.
3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.
Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3. 5.2-3 A3, 3.5.3-3Bl.
l 3.5.2-332, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3.
If the im-balance is not within the envelope defined by these figures, corrective
, measutes shall be taken to achieve an acceptable imbalance.
If an accep-table imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.7 The control red drive patch panels shall be locked at all times with limited access to be authorized by the mar.ager or his designated alt e rna t e.
3.5-9 Amendments 47, 47 & 44
Ba ses The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3A3, 3.5.2-3B1, 3.5.2-3B2, 3.5.2-3B3, 3 5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is based on LOCA analyses which have defined the maximum linear heat rate (See Figure 3.5.2-5) such that the maximum clad temperature will notl exceed the Final Acceptance Criteria.
Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary.
Operation in a. situation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters their limits while (qvadrant tilt, rod position, and imbalance) must be at simultaneously all other engineering and uncertainty factors are also at Conservatism is introduced by application of:
their limits.**
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Fuel densification effects d.
Hot rod =anufacturing tolerance factors e.
Fuel rod bowing effects The 25% 1 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at tha upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
Group _
Function 1
Safety 2
Safety 3
Safety 4
Safety 5
Regulating 6
Regulating Xenon transient override 8
APSR (axial power shaping bank)
The rod position limits are based on the most limiting of the following three c.-it eria :
ECCS power peaking, shutdown margin, and potential ejected red worth.
Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits.
The minimum available red worth, consis-tent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1).
The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65% ak/k at rated power.
These values have been shown to be safe by the safety analysis (2,3,4,5) of the hypothetical rod ejection accident.
A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod positions limits at hot zero power.
A single inserted control rod worth of 1.0% ak/k at beginning-of-life, hot :ero power would result in a lower transient peak thermal power and, therefore, less severe environmental con-l sequences than a 0.65 ak/k ejected rod worth at rated power.
- Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors.
The method used to define the operating limits is defined in plant operating procedures.
3.5-10 Amendments 47, 47 & 44
e Control rod groups are withdrawn in sequence beginning with Group 1.
Groups 5, 6, and 7 are overlapped 25 percent.
The normal position at power is for Groups 6 and 7 to be partially inserted.
The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established with consideration of potential effects of rod bowing and fuel densification to prevent,the linear heat rate peaking increase aisociated with a positive quadrant power tilt during normal power operation from exceeding 5.107. for Unit 1.
The limits shown in Specification 3.5.2.4 5.10* for Unit 2 5.10" for Unit 3 are measurement system independent.
The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.
The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally will be performed in the process computer.
The two-hour frequency for monitoring these quantities will provide adequate surveillance when the cemputer is out of service.
Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation.
Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.
Operating restrictions are inciuded in Technical Specificatien 3.5.2.5d to prevent ex:essive power peaking by transient xenon.
The xenon reactivity must be beyond its final maxi =um or mini =u=
p=ak and approaching its equili-brium value at the power level cutoff, REFERENCES 1FSAR, Section 3.2.2.1.2 2FSAR, Section 14.2.2.2 3FSAd, SUPPLDfENT 9 43&W RIEL DENSIFICATION REPORT 3AW-1409 (UNIT 1) 1AW-1396 (UNIT 2) 1AW-1400 (UNIT 3) 5Oconee 1, Cycle 4 - Reload Report - BAW-1447, March 1977.
3.5-11 Amendments 47, 47 & 44
(94.30 (174.102)
,,(225 9.102)
OPERATION IN o
g POWER LEVEL THIS REGION l$
CUTOFF NOT Att08E0 (114'901 (225 9 90) g RESTRICTED (169.80)
(230 9 80) 50 REGION g,y RESTRICTED REGION
)
usagig IO (164.70)
(235 9.70)
Litti E
(240 1.60)
(159,60) 60 (300.60)
=
50 - (42.50)
=
(80 50)
PERuissigLE OPERAflNG REGION E 4g 2
30 20 h 10.15) e (39.15) 10 (0 0l I
f I
t I
t i
0%
0 100 200
.' ; 0 Roa inces, 5 segnataan 0
25 50 75 100 0
25 50 7'
100 I
I f
f t
t i
OI3dU I Groug 1 0
25 50 75 100 I
t t
t t
Grouu 6 RJO Indet is 19e yttger:tage sum 31 tng eith3r3 sal of Gr3ugt 5,6 and 7 ROD POSITION LIMITS FOR FOUR-PUMP OPERATION FROM 0 TO 100 (:101 EFPD, UNIT 1 at m't OCONEE NUCLEAR STATION
,M 1,#_1'9 Figure 3.5.2-1A1 Amendments 47, 47 & 44
(188.102) o (225 3,102) 100 RESTRICTED CPE Ail 0N IN THl3 POWER LEVEL (225 9.90) 30 (i74 i,30)
REGION 15 MOT CUTOFF REGION ALL0sE0 B0 (169 1.80)
(230 9 80) 3 HUT 00fN BARG,1N Lluli (164.1.70)
(235.9.70) g
~
(240.9.60)
REsini:tEO REGION (159 1.60)
~.
60 (300.60 )
(126.50) g go PEastSSIBLE OPERATING 40 REGION
~
30 20 (84.15) 10 (3 0) o t
i I
t i
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g 100 200 300 Roo inaet. 5 Estnaraon 0
25 50 75 100 0
25 50 75 100 i
i t
t i
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1 GrGup 5 GrouD 7 0
25 50 75 100 l
t i
1 Groua 6 Aca inae is tne aircentage sus of tne sitnarasal of Grauss 5.6 ana 7 ROD POSITION FOR FOUR-PUMP OPERATION FROM 100 (:10) TO 235 (:10) EFPD, UNIT 1 J84,.
OCONEE NUCLEAR STATION 3.S-13 I' W e 3'5 2~1A2 m
Anendments 47, 47 & 44
s (201,102) o (251.1.102)
POWER LEVEL CUTOFF o(251.1,90:
90 OPEIAfl0N IN THl1 REEL 1H 13 NOT ALL0tED RESTRICTED (246 1.80) g REGION (241 S.70) 70 (236 8,60) f 60 (127.50) 50 j
40 SHUT 0094 NARGIN PERul!3IBLE CPERATWG R?>10N Liuti 30 20 (100.15) 10 (0.0) 0 e
t i
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0 100 200 300 ilos inces. 5 fitnaraan 3
25 50 75 100 0
25 50 75 100 1
I i
t i
t Group 7 Group 5 0
25 50 75 100 1
I I
i Group S 15 tne percentage sum of the eitncrasal of Grouut 5.6 and 7 Rc: ince R00 POSITION LIMITS FOR FOUR-l PUMP OPERATION AFTER 235 (:10) f
! g EFPD, UNIT 1 antmwn OCONEE NUCLEAR STATION D
Figure 3.5.2-iA3 3.5-13a Amendments 47, 47 3 44
1 (94,102)
(159.102) i169.102)
(231 102) 100 OPEasil0N IN THl3 2 AND 3 PUur RESTRICTED REGION IS NOT ALL0tto OPERATl0N REGION (164,89) 90 - sITH 2 OR 3 PuuPS RESTRICTED i (236'59)
ERAfl0N THIS RE;lC
=
I 30 RESTRICTED IN (241'761 THIS REGION (159.76) j 70 - dlN1404 SHu100tN Liutt SC M
$0 - #44 50) 40 3
PERutSSl!LE OPERATING REGION 30 a
20 j
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10 0
1 1
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0 100 200 300 Ras inces. " Witnaraan 0
25 50 75 100 0
25 50 75 100 I
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25 50 75 100
?
f
(
GT3uo S R30 indet 15 I:18 Def 0 tnt 3[t $Um Of Int eitndf asal Of Ef 3uus 5.5 ano 7 RCD POSITION LIMITS FOR TWO-AND THREE UMP OPERATION P:.0Y g
0 TO 100 (-10) EFPD, U'4IT 1 pt Postt OCONEE NUCLEAR STATION l
D L 5-13 Amendments 47, 47 jiye 3.5.2-2Al
1 (188.102)
(231.102) 100 OPERAft0N IN THIS REGION RESTRICTED IS NOT ALLONED REGION FOR 3 PUMP
, 90 (236,89) 3 OPERATION 2
30 SHUT 00sN 4ARGIN Ll4lT (241.76)
= 70
( 00.76)
E 50 7:i 3 50 (125.50) 5 3 40 PERMISSISLE OPERATING 3
REGION
, 20 5 20 (84.15)
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25 50 75 100
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e Groua 5 Group 7 0
25 50 75 100 I
t i
I Group 6 ROD POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION FROM 100 (110) TO 235 (+10), EFPD UNIT 1
- " ' ' )
3.5-18a
'~
Figure 3.5.2-2A2 Amendments 47, 47 & 44
)
(236 0.102)
OPER4floh IN TMis REGION (201.102)
,g (246 8.102)
IS NOT ALLOWE0 fliM 2 OR 3 NEG O PuuPS R312 & 3 00
~
=
Puur (241 8.89) 2 T1
~
PERATION gg 3
RESTRICTED (236 8.76) a IN TMi$
73 f
REGION SHUT 00sN u1RGIN Lluit g
60 50 "
(127.50) 3 2
s3
\\
2 0
PERulS$19tE CPERAtlNG REGION g
20
- d i!00 15) 10 (3 0)
O t
f t
i t
t t
t i
0 100 200 300 Roa Inden. 5 titneraen 0
25 53 75 100 0
25 50 75 100 I
f I
i t
i Group 5 Group 7 0
25 50 75 100 l
t i
t Grour i Rac enoes is tne cercentage sum af tne witnerseal at Groups 5.6 ana 7 I
ROD FOSITION LIMITS FOR T'AO-AND THREE-? UMP OPERATION AFTER 235 (:10) EFPO, UNIT 1 OCONEE NUCLEAR STATION 3.5-186
,V Figure 3.5.2-2A3 Amendments 47, 47 & 44
~
I
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l POWER. 5 of 2566 MWt RESTRICTED REGION 110 (9 0.102)
( 17.3.102) 100 (8 8.90)
( 15 3,90) 90 80 70 PERMISSIBLE OPERATING 60 RANGE 50 40 30 20 10 l
1 1
0 i
i
-20
-10 0
+10
+20 Core Imaalance CPERATIONAL POWER IMEALANCE
- NVELOPE FOR OPERATION FROM 3 TO 100 (:10) EFPC, UNII I pujs.te OCONEE NUCLEAR STATION Figure 3.5.2-3Al 3.5-21 Amendments 47, 47 & 44
.y -
POWER, 5 0F 2568 Nft RESTRICTED REGION 110
(-3.3,102) 100
(+3.8,90)
(-17.0,90) 90 80 70 60 50 40 30 20 10 I
I I
i
-20
-10 0
+10
+20 Core imoalance, 5 OPERATIONAL P0'WER IMEALANCE ENVELOPE FOR OPERATION FRCM 100 (:10) TO 235 (:10) EFPO, A, b g UNIT 1 Y't OCONEE NUCLEAR STATIO u mi Ficure 3.5.2-3A2 3.5-21a l
Amendments 47, 47 3 44 l
.)
POWER, 5 0F 2568 Bft 110 RESTRICTED REGION
(-23.1.102) f(15.8,102) 100
(-21.2.90) 90 (14.9,90)
PERWISSIBLE 80 OPERATING REGION 70 60 50 40 30 20 10 I
I I
I l
30
-20
-10 0
+10
+20 Care I.maalance, ",
OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 235 (:10) EFPD, UNIT 1 OCONEE NUCLEAR STATION Figue 3.5.2-3A3 3.5-21b Amendments 47, 47 & 44
m)
(37.0,102) 100 (39.4,90)
RESTRICTED 90 4
REGION (39.4.80)
B0 (46.4,70) 70
=
(90 4.60) 60
~
o (100.60) e 50 if PERWISSIBl.E OPERATING REGION 30 20 10 I
I I
I I
I I
I 0
0 10 20 30 40 50 60 70 80 90 100 APSR, ", Wi tna r awn APSR DOSITIO'i LIMITS FC'-
OPERATION FROM 0 70 100 g (:10) EFPD, UNIT 1 Aitkatt OCONEE NUCLEAR STATION Y
Figure 3.5.2-4A1 5-23c Amendments 47, 17 & 44
II#: 0,102)
(32,102) 100 P
J RESTRICTE REGION
- 4.,
)
RESTRICTED 90
'(11.6,90)
REGION (34.6,80)
B0 (9.2,80) i (41.6,70) 70 (3.6.70)
(85.6,60) g
[
50 x (3.6.60)
(100.60)
(0.60).
E 50 PERMISSIBLE h
40 OPERATING REGION 30 20 10 0
l l
I I
I I
I
- 1 I
O 10 20 30 40 50 60 70 80 90 100 APSR, " Witnarawn APSR POSITION LIMITS FOR OPERATION FROM 100 (:10) g TO 235 (:10) EFPD, UNIT 1 pt mtij OCONEE NUCLEAR STATION Y
i 3.5-23d Figure 3.5.2-4A2 Amendments 47, 47 & 44
1 D
100 RE010N 90 (B.S 90)
,(34.4,90)
(6.2.80) o (34.4,80) 80 o
(34.4,70)
(6.1,70) 70 (85.4.60)
E B0 w
,(6 I.60) a (100.60)
(0.60) 50 g
e PERMISSIBLE d
40 OPERATING u
f REGION 30 20 10 0
I I
I I
I I
I I
O 10 20 30 40 50 60 70 80 90 100 APSR. ', Witnaraan l
i APSR POSITION LIMITS FOR OPERATION AFTER 235 (:10)
EFPD,'JNIT 1 ls$
1 U; OCONEE NUCLEAR STATION ut rms Figure 3.5.2-4A3 3.5-23e Amendments 47, 47 & 44
20 I
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i a
i
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18 s
2 4
=a
>=
16 a
2 1
i
?
14 2
Generic FAC BA'v'-10103 g
a.
x.
I 1;
1C O
2 6
8 10 12 Axial Location of Peak Power From Bottom of Core, ft LOCA-LIMITE] MA.(!vdM ALLO'4A3; LI* EAR HEAT
,b\\
M' OCONEE NUCLEAR STA sos po; r' are 3.: 2-:.
3.:. 24 Amendments 47, 47 & 44 L
Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Frequency Test Item
)
Movement of Each Rod Bi-Weekly 1.
Control Rod Movement 50" Annually 2.
Pressurizer Safety Valves Setpoine 25:: Annually 3.
Main Steam Safety Valves Setpoint Prior to Functional 4.
Refueling System Interlocks Refueling Main Steam Stop Valves (l)
Movement of Each Stop Monthly 5.
Valve Daily
(
Evaluate 6.
Reactor Coolant System Leakage Functional Annually 7.
Condenser Cooling Water System Gravity Flow Test Functional Monthly 8.
High Pressure Service Water Pu=ps and Power Supplies Prior to Functional 9
Spent Fuel Cooling System Refueling Visual Inspection Annually 10.
Hydraulic Snubbers on Safety-Related Syste=s High Pressure and Low ( )
Vent Pump Casings Monthly and Prior to Testing 11.
Pressure Injection System (1)
Applicable only when the reactor is critical is above 200 F and at a steady-(2)
Applicable only when the reactor coolant state te=perature and pressure.
(3)
Operating pumps excluded.
l l
Amendments 47/47/44 A 3.X/ M/ 3X ~ ~
4.1-9 1311ZKXK
5.3 REACTOR Specification 5.3.1 Reactor Core The reactor core contains approximately 93 metric cons of 5.3.1.1 slightly enrich *A uranium dioxide pellets.
The pellets are The encapsulated in Zircaloy-4 tubing to form fuel rods.
reactor core is made up of 177 fuel assemblies, all of which are prepressurized with Helium.
5.3.1.2 The fuel assemblies shall form an essentially cylindrical lattice with an active height of 144 in. and an equivalent diameter of 128.9 in. (2)
There are 61 full-length control red assemblies (CRA) and 8 5.3.1.3 axial power shaping rod assemblies (APSR) distributed in the reactor core as shown in FSAR Figure 3-46.
The full-length CRA and the APSR shall conform to the design described in the FSAR or reload report.
5.3.1.4 Initia.'. core and reload fuel assemblies and rods shall conform to desi3n and evaluation described in FSAR or reload report and shi.1 not exceed an enrichment of 3.5 percent of U-235.
t 5.3.2 Reaccer Coolant Svstem 5.3.2.1 The design of the pressure components in the reactor coolant system shall be in accordance with the code requirements.(3) l 5.3.2.2 The reactor coolant system and any connected auxiliary systems expcsed to the reactor coolant conditions of ta=perature and pressure, shall be designed for a pressure of 2,500 psig and a temperature of 650*F.
The pressurizer and pressurizer surge line snall be designed for a temperature of 670*F.(4) j 5.3.2.3 The maximum reactor coolant system volume shall be 12,200 ft3 REFERENCES (1)
FSAR Section 3.2.1 (2)
FSAR Section 3.2.2 (3)
FSAR Section 4.1.3 (4)
FSAR Section 4.1.2
(
-w.
5.3-1 Amendments 47, 47 & 44