ML20211J025
| ML20211J025 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/31/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | City of Dalton, GA, Georgia Power Co, Municipal Electric Authority of Georgia, Oglethorpe Power Corp |
| Shared Package | |
| ML20211J030 | List: |
| References | |
| DPR-57-A-132, NPF-05-A-066, TAC 61283, TAC 61284 NUDOCS 8611100170 | |
| Download: ML20211J025 (70) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION Q
j WASHINGTON, D. C. 2f"555
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. I 1
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.132 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated April 15, 1986, as supplemented July 25 and September 22, 1986 complies with the standards and requiren.ents 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; 8.
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 l
safety of the public, and (ii) that such activities will be conducted l
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 i
l 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 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:
8611100170 861031 PDR ADOCK 05000321 l
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< (2) Technical Sy_ecifications The Technical Specifications contained in Appendices
.A and B, as revised through Amendment No. 132, 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 and shall be. implemented within 60. days of issuance.
F THE NUCLEAR REGUl.ATORY COMMISSION Y4W*
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v Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 31, 1986 e
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a ATTACHMENT TO LICENSE AMENDMENT N0. 132 FACILITY OPERATING LICENSE N0..DPR-57 DOCKET N0. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
Remove Insert x
x xi 3.3-5 3.3-5 c
3.3-6 3.3-6 3.3-7 3.3-7 3.3-16 3.3-16 3.3-17 3.3-17 3.10-7 3.10-7 3.11-1 3.11-1 3.11-2 3.11-2 3.11-3 3.11 3.11-4a 3.11-4a Figure 3.11-1 Sheet 1 Figure 3.11-1 (Sheet 1)
Figure 3.11-1 Sheet 2 Figure 3.11-1(Sheet 2)
Figure 3.11-1 Sheet 3 Figure 3.11-1 (Sheet 3)
Figure.3.11-1 Sheet 4 Figure 3.11-1 Sheet 4 Figure 3.11-1 Sheet 5 Figure 3.11-1 Sheet 5 Figure 3.11-1 Sheet 6 Figure 3.11-1 Sheet 6 Figure 3.11-1(Sheet 7)
Figure 3.11-1 (Sheet 7)
Figure 3.11-1(Sheet 8)
Figure 3.11-2(Sheet 1)
Figure 3.11-2 Figure 3.11-4 Figure 3.11-4 Figure 3.11-5 Figure 3.11-5 Figure 3.11-6 Figure 3.11-6 Figure 3.11-7 5.0-1 5.0-1 5.0-2 5.0-2 I
es
'w-e LIST OF FIGURES Fioure Title 1.1 -1 Core Thermal Power Safety Limit Versus Core Flow Rate 2.1 -1 Reactor vessel Water Levels 4.1 -1 Graphical Aid for the Selection of an Adequate Interval Between Tests
- 4. 2-1 System Unavailability 4
- 3. 4-1 Sodium Pentaborate Solution Volume Versus Concentration Requirements 3.4-2 Sodi.um Pentaborate Solution Temperature Versus Concentration Requirements
- 3. 6-1 Change in Charpy V Transition Temperature Versus Neutron Exposure 4
3.6-2 Minimum Temperature for Inservice Hydrostatic and Leak Test 3.6-3 Minimum Temperature for Mechanical Heatup or Cooldown Following Nuclear Shutdown
- 3. 6-4 Minimum Temperature for Core Operation (Criticality) 3.11-1 (Sheet 1) Limiting Value~ for APLHGR (Fuel Type IC Types 1, 2, and 3) 4 3.11-1 (Sheet 2) Limiting Value for APLHGR (Fuel Types 80250, 8DR8265H, P8DRB265H, and BP8DR8265H) 3.11-1 (Sheet 3) Limiting Value for APLHGR (Fuel Types P80RB284H, BP80RB284, and 80R183) 3.11-1 (Sheet 4) Limiting Value for APLHGR (Fuel Types 80R233, P8DR8284LA, and BP80RB284LA) 3.11-1 (Shret 5) Limiting Value for APLHGR (Fuel Types P8DR8283 and BP80RB283) 3.11-1 (Sheet 6) Limiting Value for APLHGR (Fuel Type BP80R8299) 3.11-1 (Sheet 7) MAPFACp (Power Dependent Adjustment Factors to MAPLHGRs) 3.11-1 (Sheet 8) MAPFACp (Flow Dependent Adjustment Factors to MAPLHGRs) 3.11-2 Limiting Value for LHGR (Fuel Type 7 x 7) 3.11-3 MCPRy (Flow Dependent Adjustment Factors for MCPRs) 3.11-4 MCPR Limit for All 8 x 8 Fuel Types for Rated Power and Rated Flow HATCH - UNIT 1
'x Amendment No. Il0, 126, 132
LIST OF FIGURES (Continued) i Fiaure Title 3.11-5 MCPR Limit for 7 x 7 Fuel for Rated Power and Rated Flow 3.11-6 Kp (Power Dependent Adjustment Factors for MCPRs) 3.15-6 Unrestricted Area Boundary
- 6. 2.1 -1 Offsite Organization
- 6. 2. 2-1 Unit Organization i/TI:, -
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HATCH - UNIT 1 xi Amendment No. 132
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.F.
gperation with a limitino Control 4.3.F.
Operation with a limitino Control Rod Pattern (for Rod Withdrawal Rod Pattern (for Rod Wittdrawal Error. RWE)
Error. RWE)
A Limiting Rod Pattern for RWE exists During operation when a Limiting when:
Control Rod Pattern for RWE exists and only one RBM channel is 1.
Thermal power is below 90%
operable, an instrument functional of rated and the MCPR is less test of the RBM shall be performed than 1.70, or prior to withdrawal of the control rod (s). A Limiting Rod Pattern for 2.
Thennal power is 90% of rated RWE is defined by 3.3 F.
or above and the MCPR is less than 1.40.
During operation with a Limiting Control Rod Pattern for RWE and when core thermal power is > 30%,
either:
1.
Both RBM channels shall be oper-G.
Limiting the Worth of a Control Rod able, or Below 20% Rated Thermal Power 2.
If only one RBM channel is oper-1.
Rod Worth Minimizer (RWM) able, control rod withdrawal shall be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or Prior to the start of control rod withdrawal at startup, and as soon 3.
If neither RBM channel is oper-as automatic initiation of the RWM able, control rod withdrawal shall occurs during rod insertion while be blocked.
shutting down, tht capability of the Rod Worth Minimizer to properly G.
Limiting the Worth of a Control Rod fulfill its function shall be veri-Below 20% Rated Thermal Power fled by the following checks.
1.
Rod Wortn Minimizer (RWM) a.
The correctness of the Banked
{
Position Withdrawal Sequence l
l Whenever the reactor is in the Start input to the RWM computer
& Hot Standby or Run Mode below 20%
shall be verified.
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" f:'5.p p.
rated thermal power, the Rod Worth
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Minimizer shall be operable or.a..
.b.
The RWM computer on line diag-second licensed operator shall nostic test shall be successfully
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verify that the operator at the performed.
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reactor console is following the
@ ir-control rod program.
c.
Proper annunciation of the selec-E h'F ' -
9I' tion error of at least one out-
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of-sequence control rod in each fully inserted group shall be verified.
d.
The rod block function of the RWM shall be verified by withdrawing or inserting an out-of-sequence control rod no more than to the block point.
HATCH - UNIT 1 3.3-5 Amendment No. 27, 38, 42, 52, 105, 132 em
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- -
-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUlfEMENTS 3.3.G.2.
Rod Secuence Control System (RSCS)
- 2. Rod Secuence Control System (RSCSI a.
Operability a.
Doerability When the reactor is in the Start As soon as the group notch node and Hot Standby or Run Mode below is entered during each reactor 20% rated thermal power and control startup and as soon as automatic rod movement is within the group initiation of the RSCS occurs notch mode af ter 50% of the during rod insertion while control rods have been withdrawn, shutting down, the capabil-the Rod Sequence Control System ity of the Rod Sequence Control shall be operable except when System to properly fulfill its performing the RWM surveillance function shall be verified by at-
- tests, tempting to select and move a rod in each of the out-of-sequence groups.
When the control rod movement is within the group notch mode and as soon as automatic initiation of the RSCS occurs during rod insertion while shutting down, the operability of the notch'ing restriction shall be demonstrated by attempting to move a control rod more than one notch in the first programmed rod group.
b.
Failed Position Switch b.
_Faileo Position Switch l
Control rods with a failed " Full-A second licensed operator shall in" or " Full-out' position switch verify the conformance to Speci-may be bypassed in the Rod Se-fication 3.3.G.2.b before a rod quence Control System if the ac-may be bypassed in the Rod Se-
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tual rod position is known. These quence Control System.
rods shall be moved in sequence to
.their correct positions (full in on
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insertion or full out on withdrawal).
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l HATCH - UNIT 1 3.3-6 Amendment No. 132
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.G.2.c.
Shutdown Marcin/ Scram 4.3.G.2.c.
Shutdown Marcin/ Scram Time Time Testina Testino in order to perform the Prior to control rod with-required shutdown margin drawal for startup, verify demonstrations subsequent the conformance to Speci-to any fuel loading opera-fication 3.3.G.2.b. before tions, or to perform con-a rod may be bypassed in trol rod drive scram and/or the RSCS. The requirements friction testing as specified to allow use of the indi-in Surveillance Requirement vidual rod position bypass 4.3.C.2 and the initial start-switches within rod groups up test program, the relaxa-A12, A34, 812. OF B34 Of tion of the following RSCS the RSCS during shutdown restraints is permitted. The margin, scram time or fric-sequence restraints imposed tion testing are:
on control rod groups A12 A34, 812, or B34 af ter 50%
(1) RWM operable as per Speci-of the control rods have been fication 3.3.G.I.
withdrawn may be removed for the test period by means of the (2) Af ter the bypassing of individual rod position bypass the rods in the RSCS groups switches.
A12. A34. B12, or 834 for test purposes, it shall be demonstrated that movement of the rods in the 50% dens-ity to the preset power level range is blocked or limited to the single notch mode of withdrawal.
(3) A second licensed operator shall verify the conformance to procedures and this Specification.
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H.
Shutdown Requirements If Specifications 3.3.A through 3.3.G are not met, an orderly shutdown shall be initiated and I
the reactor placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
HATCH - UNIT 1 3.3-7 Amendment No. 132
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8ASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS i
3.3.6.1.
Rod Worth Minimizer (RWM) -(Continued) i.
l In perfor.aing the function described above, the RWM and RSCS are not re-quired to impose any restrictions at core power levels in excess of 20%
of rated. Material in the cited references shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occur-ring at power greater than 20%, regardless of the rod pattern. This is true for all normal and abnormal patterns including those which maximize j
the individual control rod worth.
At power levels below 20% of rated, abnormal coatrol rod patterns could produce rod worths high enough to be of concern relative to the 280 cal-orie per gram rod drop limit. In this range of RWM and the RSCS con-j strain the control rod sequences and patterns to those which involve only I
acceptable rod worths.
I The Rod Worth Minimizer and the Rod Sequence Control System provide auto-matic supervision to assure that out of sequence control rods will not
, be withdrawn or inserted; i.e., it limits operator deviations from plan-ned withdrawal sequences. They serve as a backup to procedural control of control rod sequences, which limit the maximum reactivity worth of I
control rods. In the event that the Rod Worth Minimizer is out of ser-vice, when required, a second licensed operator or other qualified tech-nical plant employee whose qualifications have been reviewed by the AEC can manually fulfill the control rod pattern conformar.ce functions of this system.
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The functions of the RWM and RSCS make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 20%, these devices force ad-herence to acceptable rod patterns. Above 20% of rated power, no con-l sequences are acceptable. Control rod pattern constraints above 20% of rated power are imposed by power distribution requirements as defined in
.Section 3.11 and 4.11 of these Technical Specifications. Power level for automatic cutout of the RSCS function in sensed by first stage turbine pressure. Because the instrument has an instrument error of i 10% of
~ full power the nominal instrument setting is 30% of rated power.
Power level for automatic cutout of the RWM function is sensed by feedwater
(
and steam flow and is set nominally at 30% of rated power to be consistent
- gg with the RSCS setting.
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Surveillance Requirements:
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Functional testing of the RWM prior to the start of control rod withdrawal at startup, and prior to attaining 20% of rated thermal power during rod in-sertion while shutting down, will ensure reliable operation and minimize the probability of the rod drop a cident.
2.
Rod Seauence Control System (RSCS) a.
Coerability Limiting Conditions for Operation:
See bases for Technical Specification 3.3.G.I. Rod Worth Minimizer.
HATCH - UNIT 1 3.3-16 Amendment No. 132
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8ASES FOR LIMITING CONDIT!ONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.3.G.2.a.
Operability Surveillance Requirements:
The RSCS can be functionally tested after 50% of the control rods have been withdrawn, by demonstrating that the continuous withdrawal mode for the control drives is inhibited.
This demonstration is made by attempting to withdraw a control rod more than one notch in the first programmed rod group subsequent to reaching the 50% rod density point.
a This restriction to the notching mode of op-eration for control rod withdrawal is automatically removed when the re-actor reaches the automatic initiation setpoint.
During reactor shutdown, similar surveillance checks shall be made with regard to rod group availability as soon as automatic initiation of the RSCS occurs and subsequently at appropriate stages of the control rod insertion.
b.
Failed Position Switch Limiting Conditions for Operation:
In the event that a control rod has a failed " Full-in" or " Full-out" position switch, it may be bypassed in the Rod Sequence Control System if its position is otherwise known. It is a safer and more desirable condition for such rods to occupy their proper positions in the control rod patterns during reactor startup or shutdown.
Surveillance Requirements:
Having a second licensed operator verify the actual rod position prior to bypassing a rod in the Rod Sequence Control System provides assurance that Specification 3.3.G.2.b. is met.
c.
Shutdown Marcin/ Scram Time Testino After initial fuel loading and subsequent refuelings when operating above 950 psig all control rods shall be scram tested within the constraints imposed by the RSCS and before the 40%. power. level is reached. To main-tain the required reacter pressure conditions the individually scrammed-or inserted rod should be withdrawn to its original position immediately following testing of each rod. In order. to select and withdraw the scram-med or inserted insequence control rod (also to select and insert a fully withdrawn insequence rod in case of friction testing) it will be neces-sary to simulate all the insequence withdrawn rods of the succeeding RSCS groups as being at full,in position by utilizing the individual rod posi-HATCH - UNIT 1 3.3-17 Amendment No. 132
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BASES FOR LIMITING CONDITIONS FOR OPERATION 3.10.A.2.
Fuel Granole Holst Load Settino Interlocks Fuel bandling is normally conducted with the fuel grapple hoist. The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly. This total is approximatelm 1500 lbs. in comparison to the load setting of 485 1 30 lbs.
3.
Auxiliarv Hoists load Settine Interlock Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks. The 485 1 30 lb load setting of these hoists is adequate to trip the interlock when a fuel bundle is bothg handled.
B.
Fuel Loadine To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality.
C.
Core Monitorine Durine Core Alterations The SRM's are provided to monitor the core during periods of Unit shutdown and to guide the operator during refueling operations and Unit startup. Requiring two operable SRM's in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations.
The requirements of 3 counts per second provides assurance that neutron flux is being monitored.
During spiral unloading, it is not necessary to maintain 3 cps because core alterations will involve only reactivity removal and will not result in criticality.
The loading of up to four fuel bundles around the SRM's before attaining the 3 cps is permissible because these bundles were in a subcritical configuration when they were removed and therefore they will remain subcritical when placed back in their previous positions.
D.
Scent Fuel Pool Water Level The design of the spent fuel storage pool provides a storage location for 3181 fuel assemblies in the reactor building which ensures adequate shielding, cooling, and the reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the active fuel will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20. The normal water level provides 14-1/2 feet of additional water shielding. All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the water level to less than 10 feet above the top of the active fuel. Lines extending below this level are equipped with two check valves in series to prevent inadvertent pool drainage. All fuel loaded into the Edwin I. Hatch Nuclear Plant spent fuel pool shall have an uncontrolled lattice ke less than or equal to the limit for high density fuel racks described in the
" General Electric Standard Application for Reactor Fuel' (GESTAR II),
NEDE-24011-P-A-8. Alternatively, fuel not described in GESTAR II shall have been analyzed with another NRC approved methodology to ensure conformity to the FSAR design basis for fuel in the spent fuel racks.
E.
Control Rod Drive Maintenance During certain periods, it is desirable to perform maintenance on two control rod drives at the same time.
HATCH - UNIT 1 3.10-7 Amendment No. 74, 102, 132
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
(
3.11.
FUEL ROOS 4.11.
FUEL R005 Aco11cability Applicability The Limitt'ng Conditions for Operation The Surveillance Requirements apply associated with the fuel rods apply to to the parameters which monitor the those parameters which monitor the fuel rod operating conditions.
fuel rod operating conditions.
Obiective Obiective The Objective of the Limiting Condi-The Objective of the Surveillance tions for Operation is to assure the Requirements is to specify the type performance of the fuel rods.
and frequency of surveillence to be applied to the fuel rods.
Specifications Specifications A.
Average Planar Linear Heat Genera-A.
Averace Planar Linear Heat Genera-tion Rate (APLHGR) tion Rate ( APLHGR)
During power. operation, the APLHGR The APLHGR for each type of fuel as for all core locations shall not a function of average planar exceed the appropriate APLHGR limit exposure shall be determined daily for those core locations. The APLHGR during reactor operation at 125%
limit, which is a function of average rated thermal power.
planar exposure and fuel type, is the appropriate value from Figure 3.11-1, sheets 1 through 6 multiplied by the I smaller of the two MAPFAC factors de-termined from Figure 3.11-1, sheets 7 and 8.
If at any time.during oper-l ation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the pre-scribed limits. If the APLHGR is not returned to within the pre-scribed limits within two (2) hours.
then reduce reactor power to less than 25% of rated thermal power with-in the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not required.
8.
Linear Heat Generation Rate (LHGR) 8.
Linear Heat Generation Rate figiil During power operation, the LHGR as The LHGR as function of core a function of core height shall not height shall be checked daily dur-exceed the limiting value shown in ing reactor operation at 125%
g Figure 3.11-2 for 7 x 7 fuel or the rated thernal power.
limiting value of 13.4 kw/ft for any 8 x 8 fuel. If at any time during HATCH - UNIT 1 3.11-1 Amendment No. 51, 52, 69, 87, 96, 105, 132 S
M
d LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.11.8.
Linear Heat Generation Rate (LHGR)
(Continued) operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25%
of rated thermal power is not required.
C.
Minimum Critical Power Ratio (MCPR)4.ll.C.l. Minimum Critical Power Ratio (MCPR)
The minimum critical power ratio (MCPR)
MCPR shall be determined to be shall be equal to or greater than the equal to or greater than the operating limit MCPR (0LMCPR), which applicable limit, daily during is a function of scram time, core reactor power operation at 125%
power, and core flow. For 25% 5 rated thermal power and following power < 30%, the OLMCPR is given in any change in power level or dis-Figure 3.11.6.
For power 1 30%,
l tribution that would cause opera-the CLMCPR is the greater of either:
tion with a limiting control rod pattern as described in the bases 1,
The applicable limit determincd for Specification 3.3.F.
[
- f rom Figure 3.11.3, or 4.ll.C.2.
Minimum Critical Power Ratio Limit 2.
The. applicable limit from either Figures 3.11.4 or 3.11.5, l
The MCPR limit at rated flow and multiplied by the Kp factor I
rated power shall be determined for determined from Figure 3.11.6, l
each fuel type, 8X8R, P8X8R, BP8x8R or where:
7X7 from figures 3.11.4 and 3.11.5 m
respectively using:
't = 0 or 'tave 18
, whichever is
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a.
t=1.0 prior to initial scram vg.vB _
greater time measurements for the.
cycle, performed in accordance tg = 0.g0 sec (Specifications 3.3.C.2.a.
with specifications 4.3.C.2.a.
scram time limit to 20% insertion f rom fully withdrawn) or s/s b.
t as defined in specification TB = 0.710&l.65-N1 (0.053) Dief.103 3.ll.C.
I Ng The determination of the limit i=1 must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
. of the conclusion of each scram tithe surveillance test required by specification 4.3.C.2.
HATCH - UNIT 1 3.11-2 Amendment No. 51, 52, 69, 76, SS, 105, 132
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BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS 3.11.
FUEL R005 A.
Average Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K, even considering the postulated effects of fuel pellet densification.
The peak cladding temperature following a postulated loss-of-coolant acci-dent is primarily a function of the' average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent second-arily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power distribution within a fuel assembly affect-the calculated peak clad temperature by less than ! 20'F relative to the i
peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures conform to 10 CFR 50.46. The limiting value for APLHGR at rated conditions is shown in Figures 3.11.1. sheets 1 thru 6.
l t
A flow dependent correction factor incorporated in to Figure 3.11-1 (sheet 8) is-l applied to the rated conditions APLHGR to assure that the 2200*F PCT limit is complied with during LOCA initiated from less than rated core flow. In addition, other power anc. flow dependent corrections given in Figure 3.11-1 (sheets 7 and 8) are applied to the rated conditions APLHGR limits to assure l
that the fuel thermal-mechanical design criteria are met during' abnormal transients initiated f rom of f-rated conditions.
The calculational procedure used to establish the APLHGR shown in Figures 3.11.1, sheets 1 thru 6, is based on a loss-of-coolant accident analysis.
l i
The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
Differences in this analysis as compared to previous analyses performed with Reference 1 are: (1) The analyses assume a fuel assembly l
planar power consistent with 102% of the MAPLHGR shown in Figure 3.11.1; (2) Fission product decay is computed assuming an energy release rate of 200 HEV/ Fission; (3) Pool boiling is assumed after nucleate boiling is lost i
during the flow stagnation period; (4) The effects of core spray entrainment and counter-current flow limiting es described in Reference 2, are included in the reflooding calculations.
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A list of the significant plant input parameters to the loss-of-coolant
.1 accident analysis is presented in Table 1 of NE00-21187(s). Further discussicn of the APLHGR bases is found in NE0C-30474-p(**).
l HATCH - UNIT 1 3.11-3 Amendment No. 10, 27,~33, 42, 96, JQ3, 132
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o 8ASES FOR LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.ll.C.
Minimum Critical Power Ratio (MCPR) (Continued)
The purpose of the MCPR, and the K of Figures 3.11.3 and 3.11.6, respectively, is f
p to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRf and MCPRp at the existing core flow and power state. The MCPRys are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.
The MCPRys were calculated such that for the maximum core flow rate and the corres-ponding THERMAL POWER along the 105% of rated steam flow control line, the limiting bundle's relative power was adjusted untti the MCPR was slightly above the Safety Limit. -Using this relative bundle power, the MCPRs were calculated at different points along the 105% of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPR.
f The core power dependent MCPR operating limit.MCPR is the power rated flow MCPR p
operating limit multiplied by the K p factor given in Figure 3.11.6.
The K s are established to protect the core from transients other than core flow p
increases, including the localized event such as rod withdrawal errcr. The Kps were determined based upon the most liraiting transient at the given core power level. (For further information on MCPR operating limits for of f-rated conditions, reference NE0C-30474-P.(11))
l l
1 I
HATCH - UNIT 1 3.11 -4a Antendment No. 42, 105, 132
. ~..
16
-FUEL TiPE 7X7 '
IC 1&2 g
80 MIL CHANNELS l
14.5
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FIGURE 3.111 (SHEET 1)
HATCH - UNIT 1 9
Amendment No. 3), #2, 52, 96, 132
t 14 x
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HATCH - UNIT 1 Amendment No. 33, 42, 32, 96, 132 21.
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FIGURE 3.11 1 (SHEET 3)
HATCH - UNIT 1 Amendment'No. 96, 132
16 FUEL TYPE: 8DR233(8DRB221) 100 MIL CHANNELS l
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Amendment No. 96, 132
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FIGURE 3.11 1 (SHEET 5)
HATCH - UNIT 1 Amendment No. 96, 132
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HATCH - UNIT 1 Amendment No. 195, 132'
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MAPLHGRp = MAPFACp + MAPLHGRSTD y
MAPLHGRSTD = STANDARD MAPLHGR LIMITS 8 0.7 g
FOR 25%>P:
NO THERMAL LIMITS MONITORING REQUIREO 2
NO LIMITS SPECIFIED
~ 50% CORE FLOW w
m FOR 25% $P 430%:
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w FOR $ 50% CORE FLOW E
0.6 -
MAPFACp = 0.433 + 0.005224 (P - 30%)
e FOR > 50% CORE FLOW w
l FOR 30% $P:
g MAPFACp = 1.0 + 0.005224 (P - 100%)
> 50% CORE FLOW I
05 I
1ii I
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20 25 30 40 50 60 70 80 90 100 POWER (% R ATEb)
FIGURE 3.11-1 (SHEET 7) MAPFACp
4 i
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- STD,
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u 4
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Og F = FRACTION OF RATEO CORE FLOW, a
ANO A.8, ARE FUEL TYPE DEPENDENT p
o CONSTANTS GIVEN SELOW:
1 4
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(% R ATED)
F F
F F
A E
A E
j 102.5 0.4400 0SS57 0.4001 0.5784 107.0 0.4421 0.4633 0.4674 0.6758 112.0 0.4074 0.8581 0.4214 0.0007 117.0 0.3701 0.M64 0.3824 0.4088 6.50 f
l f
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CORE FLOW (% RATED)
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FIGURE 3.111 (SHEET 8) MAPFACp
T6P 12 LIMITING VALUE FOR LHGR AS A FUNCTION OF CORE 11 HEIGHT
= ~ 18.5 [1 -~.026 (f2II 10 9
WHERE: Z = HEIGHTIN FEET 8
ABOVE BOTTOM OF THE CORE I
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PERMISSIBLE REGION 3
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F!GURE 3.112 LIMITING VALUE FOR LHGR l
HATCH - UNIT 1 FUEL TYPE 7X7 i
l Amendment No. 27, #2, 132 l
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a
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FIGURE 3.11.4 MCPR LIMIT FOR ALL 8X8 FUEL TYPES FOR RATED POWER AND RATED FLOW I
HATCH - UNIT 1 Amendment No. $$, $7, 95, 195, 132-
e 1.31 1.30 1.29 a
1.28 1.27 t:23 1.26 e
Ba 1.2s 1.24 1.23 1.22 0.0 33 FIGURE 3.11-5 MCPR LIMIT FOR 7X7 FUEL FOR RATED POWER AND RATED FLOW i
HATCH - UNIT 1 Amendment No. 59, 57, 96, 105, 132
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%0C1tdWOd(SMI WElldfilf)W Ud3W 031VW SECC) d5 %E WOd W3dW10 HATCH - UNIT 1
- ne
- 5.0.
- MAJOR DESIGN FEATURES A.
$ji.t_t Edwin I. Hatch Nuclear Plant Unit No.1 is located on a site of about 2244 acres, which is owned by Georgia Power Company, on the south side of the Altamaha River in Appling County near Baxley, Georgia. The Universal Transverse Mercator Coordinates of the center of the reactor building are: Zone 17R LF 372,935.2m E and 3,533,765.2m N.
B.
Reactor Core 1.
Fuel Assemblies P
The core shall consist of not more than 560 fuel assemblies and shall be limited to those fuel assemblies which have been analyzed with NRC approved codes and methods
- and have been shown to comply with all Safety Design Bases in the Final Safety Analysis Report (FSAR).
2.
Control Rods The reactor shall contain 137 cruciform-shaped control rods.
C.
Reactor vessel' The reactor vessel is described in Table 4.2-2 of the FSAR. The applicable design specifications shall be as listed in Table 4.2-1 of the FSAR.
0.
Containment l
1.
I The principal design parameters are characteristics of the primary containment shall be as given in Table 5.2-1 of the FSAR.
j 2.
Secondary Containment * (See Page 5.0-la)
The secondary containment shall be as described in Section 5.3.3.1 of the FSAR and the applicable codes shall be as given in Section 12.4.4 of the FSAR.
l 3.
Primary Containment Penetrations Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.
E.
Fuel Storaae 1.
Spent Fuel All arrangement of fuel in the spent fuel storage racks shall be maintained in a subcritical configuration having a k,f f not greater than 0.95.
(
2.
New Fuel The new fuel storage vault shall be such that the k rt dry shall not be greater e
than 0.90 and the k rr flooded shall not be greater than 0.95.
e HATCH - UNIT 1 5.0-1 i
i Amendment No. 74, 91, 102, 132 I,
5.0.F.
Seismic Desian The reactor building and all engineered safeguard systems are designed for the design basis earthquake with a horizontal ground acceleration of 0.15 g.
The operating basis earthquake has a horizontal ground acceleration of 0.06 g.
G.
Component Cyclic or Transi_ent Limit The Reactor Pressure Vessel is designed for and shall be maintained within the cyclic or transient limits of Table 5.0.G-1.
H.
References 1.
FSAR Section 4.2, Reactor Vessel and Appurtenances Mechanical Design 2.
FSAR Section 5.2, Primary Containment System 3.
FSAR Section 5.3, Secondary Containment System 4.
FSAR Section 12.4.4, Governing Codes and Regulations 5.
FSAR Section 10.3, Spent Fuel Storage 6.
FSAR Section 10.2, New Fuel Storage HATCH - UNIT 1 5.0.2 Amendment No. 74, US, 132
- pa ncy[o,,
UNITED STATES E
NUCLEAR REGULATORY COMMISSION g
3 E
WACHINGToN, D. C. 20555
\\*****/
GECRGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY ODERATING LICENSE Amendment No. 66 License No. NPF {
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated April 15, 1986, as supplemented July 25 and September 22, 1986 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; 8.
The facility will operate in conformity with the application, the provisions of'the Act, and the rules and regulations of the Commissinn; 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.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. N?F-5 is hereby amended to read as follows:
. (2) - Technical Specifications The. Technical Specifications contained in Appendices A and B, as revised through Araendment No. 66, are hereby' incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This ifcense amendment is~ effective as of its date of issuance and shall be implemented within 60 days of issuance.
F THE NUCLEAR REGULATORY COMMISSION M
g/
As Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications:
Date of Issuance: October 31, 1986 I
l l
l l
n.
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.,. - -m,--
-4
O ATTACHMENT TO LICENSE AMENDMENT N0. 66 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET N0. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change. The overleaf pages are provided for convenience.
Remove Insert 3/4 1-14 3/4 1-14 3/4 1-15 3/4 1-15 3/4 1-16' 3/4 1-16' 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4a
'3/4 2-4a 3/4 2-4b 3/4 2-4b 3/4 2-4c 3/4 2-4c 3/4 2-4d 3/4 2-4d 3/4 2-4e 3/4 2-4e 3/4 2-4f 3/4 2-4f 3/4 2-49 3/4 2-4g 3/4 2-4h 3/4 2-4h 3/4 2-41 3/4 2-41 3/4 2-4j 3/4 2-4k 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-7a 3/4 2-7a 3/4 2-7b 3/4 2-7b 3/4 2-7c 3/4 2-7c 3/4 2-7d 3/4 2-7d 3/4 2-8 3/4 2-8 8 3/4 1-3 B 3/4 1-3 8 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 B 3/4-2-4 B 3/4 9-2 B 3/4 9-2 5-1 5-1 5-4 5-4
REACTIVITY CONTR0l. SYSTEMS CONTROL ROD DRIVE HOUSING SUPPORT LIMITING CONDITION FOR OPERATION I-3.1.3.8 The control rod drive housing support shall be in place.
APPLICABILITY: CONDITIONS 1, 2 and 3.
ACTION:
.l With the control' rod drive housing support not in place, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.8 The control rod drive housing support shall be inspected after reassembly and verified to be in place, prior to startup, any time it H
has been disassembled or when maintenance has been performed in the i
,,,,'4,h control rod drive housing support area.
i I
d HATCH - UNIT 2 3/4 1-13
-- - +
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REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS R00 WORTH MINIMIZER
~
LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod Worth Minimizer (RWM) shall be OPERABLE.
APPLICABILITY: CONDITIONS I and 2*, when THERMAL POWER is less than 20%
of RATED THERMAL POWER.
ACTION:
With the RWM inoperable, the provisions of Specification 3.0.4 are not l
applicable, operation may continue and control rod movement is permitted l
provided that a second licensed operator or other qualified member of j
the technical staff is present at the reactor control console and
, verifies compliance with the prescribed control rod pattern.
t SURVEILLANCE REQUIREMENTS i
l 4.1.4.1 1he RWM shall be demonstrated OPERABLE:
In CONDITION 2 prior to withdrawal of control rods for the purpose a.
of making.the reactor critical, and in CONDITION 1 when the RWM is initiated during contrcl rod insertion when reducing THERMAL POWER, i
by:
1.
Verifying proper annunciation of the selection error of at least one out-of-sequence control rod, and 2.
Verifying the rod block function of the RWM by moving an out-of-sequence control rod.
b.
By verifying that the Banked Position Withdrawal Sequence input to the l
RWM computer is correct following any loading of the sequence program into the computer.
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
HATCH - UNIT 2-3/4 1-14 Amendment No. 66
REACTIVITY CONTROL SYSTEMS ROD SEQUENCE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.4.2 The Rod Sequence Control System'(RSCS) shall be OPERABLE.
APPLICABILITY: CONDITIONS 1* and 2*#, when THERMAL POWER is less than 20%
of RATED THERMAL POWER and control rod movement is within the group notch mode after 50% of the control rods have been withdrawn.
ACTION:
With the RSCS inoperable control rod movement shall not be permitted, except by a scram.
SURVEILLANCE REQUIREMENTS 4.l.4.2 The RSCS shall be demonstrated OPERABLE by:
a.
Selecting and attempting to move an inhibited control rod:
1.
As soon as the group notch mode is entered during each reactor l
'startup, and i
l 2.
As soon as the rod inhibit mode is automatically initiated during control rod insertion.
i i
9 l
i
- See Special Test Exception 3.10.2.
- Entry into CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RSCS prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
HATCH - UNIT 2 3/4 1-15 Amendment No. 66 l
REACTIVITY CONTROL SYSTEMS
~
4
$URVEILLANCE REQUIREMENTS (Continued) b.
Attempting to move a control rod more than one notch as soon as the group notch mode is automatically initiated during control rod:
1.
Withdrawal each reactor startup, and i
2.
Insertion.
Performance of the comparator check of the group notch circuits c.
prior to control rod; 1.
Movement within the group notch mode during each reactor startup, and 2.
Insertion to reduce THERMAL POWER to less than 20% of RATED THERMAL POWER.
i
? ;.
~
HATCH - UNIT 2 3/4 1-16 z
Amendment No. 66
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 ALL AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall be equal to or less than the applicable APLHGR limit, which is a function of fuel type and AVERAGE PLANAR EXPOSURE. The APLHGR limit is given by the applicable rated power, rated-flow limit taken from Figures 3.2.1-1 through 3.2.1-11, l
multiplied by the smaller of either:
a.
The factor given by Figure 3.2.1-12, or.
l b.
The factor given by Figure 3.2.1-13.
l APPLICABILITY-CONDITION 1, when THERMAL POWER 2 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the limits of Figures 3.2.1-1 through 3.2.1-11, as adjusted per Figures 3.2.1-12 and 3.2.1-13, initiate corrective action within 15 minutes and continue corrective action so that the APLHGR meets 3.2.1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less thar. 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE ' REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limit determined from Figures 3.2.1-1 through 3.2.1-11, as adjusted per Figure 3.2.1-12 and 3.2.1-13:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been estabitshed, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is
(
operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
.i HATCH - UNIT 2 3/4 2-1 Amendment No. 71, 75,.77, 77, 66
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),
FUEL TYPE 8D18175 (8DRL183) 100 MIL CHANNELS l
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION i
RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1 1 i
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E UNACCEPTABLE OPERATION I
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AVERAGE PLANAR EXPOSURE (GWd/t) 1
=
FUEL TYPE'8D18221 (8DRL233) 100 MIL CHANNELS l
MAXIMUM' AVERAGE PLANAR LINEAR HEAT GENERATION
~
RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2-1-2
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RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-3 e
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I FUEL TYPES P8DRB284LA AND BP8DRB284LA 100 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION l
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RATE (MAPLHGR) VERSUS PLANAR EXPOSURE I
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i AVERAGE PLANAR EXPOSURE (GWd/t) 4 i
FUEL TYPES P8DRB283 AND BP8DRB283100 MIL CHANNELS 1
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION i
RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE L
FIGURE 3.2.1-5
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AVERAGE PLANAR EXPOSURE (GWd/t) 1 FUEL TYPES HATCH 1 f.C.1,2,3 (7X7) 80 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT. GENERATION l
RATE (MPALHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.16 s
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AVERAGE PLANAR EXPOSURE (GWdh) i
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, FUEL TYPES P8DRB26SH AND BP8DRB265H 80 AND 100 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEATGENERATION j
RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE
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FIGURE 3.2.1-7 l
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i MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION i
RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-8
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40 45 50 AVERACE PLANAR EXPOSURE (Guud/tl FUEL TYPES P8DRB284H AND BP8DRB284H 80 AND 100 MIL CHANNELS l
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-9
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AVERAGE PLANAR EXPOSURE (Gwd/t) '
l FUEL TYPES bR8283 ANObDRB283 80 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION 1
RATE (MAPLHOR) VERSUS AVERAGE PLANAR EXPOSURE l
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HATCH - UNIT 2 3/4 2-41 Amendment No. 39, 66 t
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[u. ese asAPLHGR, = asAPFAC, *MAPLHGA STO g
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c Aeso A,3, ARE FUEL TVPE DEPEe00ENT O
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FOR 25%>P:
ISO THERAAAL LlastTS MO88tTORING REOutRED j
A NO LiteITS SPECIFIED
"" eSIL CORE FLOUT e
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2 FOR 2WILIP 43915:
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FORS 5095 CORE FLOWU
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FOR > Seit CORE FLOct FOR N'IL$P:
.C. =,.. e 41.-,se,,e
> 3915 CORE PLOUU i
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3/4.2.2 APRM SETPO!hT5
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Amendment No. JA, 35, 39
POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 ALL MINIMUM' CRITICAL POWER RATIOS (MCPRs), shall be equal to or greater than the MCPR operating Ilmit (OLMCPR), which is a function of average scram time, core flow, and core power. For 25% s Power < 30%, the OLMCPR is given in Figure 3.2.3-4.
For Power 2 30%, the OLMCPR is the greater of either:
a.
The appitcable limit determined from Figure 3.2.3-3, or b.
The appropriate Kp given by Figure 3.2.3-4, multipiled by the appropriate limit from Figure 3.2.3-1 or 3.2.3-2 where:
t = 0'er
' ave
'8, whichever is greater, tg tB.
tg = 1.096 sec (Specification 3.1.3.3 scram time limit to notch 36),
~
N 1/2 B = 0.834 + 1.65 g
(0.059),
t nE Nj-t=1 n
E Njrj
= i=1 t,y, N'
n E
i=1
(
n=
number of surveillance tests performed to date in cycle, I
th Ng = number of active control rods measured in the i surveillance
- test, tg = average scram time to notch 36 of all rods measured in the th i
surveillance test, and Ng = total number of active rods measured in 4.1.3.2.a.
APPLICABILITY: CONDITION 1, wt en THERMAL POWER 2 25% PATED THERMAL POWER ACTION:
With MCPR less than the applicable limit determined from Specification 3.2.3.a. or 3.2.3.b, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
HATCH - UNIT 2 3/4 2-6 I
AmendmentNo.71,$3,Jf,_66
O 3/4.2.3 MINIMUM CRITICAL POWER RATIO (CONTINUED)
SURVEILLANCE REQUIREMENTS 4.2.3 The MCPR limit at rated flow and rated power shall be determined for each type of fuel (8X8R, P8X8R, BP8X8R, and 7X7) from Figures 3.2.3-1 and 3.2.3-2 using a.
t = 1.0 prior to the initial scram time measurements for the cycle performed in accordance with Specification 4.1.3.2.a. or
^
b.
t as defined in Specification 3.2.3; the determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2.
MCPR shall be determined to be equal to or greater than the applicable limit:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, rnd c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.
9 HATCH - UNIT 2 3/4 2-7 I
Amendment No. 2I, 39, 66
O 1.3B 1J7 g'
g 1.30 0
e 1JE i
/
1.u l 1.as g
a
/
t.s t.at g
b
/
1.30 tJe i.as o.o o.
o.4 o.s os 1.o.
T ALL 8X8 FUEL TYPES l
FIGURE 3.2.31 l
I i
i HATCH - UNIT 2 3'4 2-7a
/
l Amendment No. 21, 78, 39, 66
, -,., _, - - - ~ _ _.,. - -, - -, -
,_~-..n
u 1.33 132 1 31 ACCEPTABLE OPER ATION 1.30 1.29 1.28 1.27 1.26 UNACCEPTABLE OPER ATION 1.25 1J4 4
1.23
~
o.o 0.2 c.4 0.s o.s 1.o T
FIGURE 3.2.3 2 l
MCPR LIMIT FOR 7X7 FUEL AT RATED FLOW AND RATED POWER HATCH - UNIT 2 3/4 2-7b Amendment No. 33, 39, 66
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~ -
Amendment No. 21, 33, 39, 66 1
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=-
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POR P< W E: 800 THEftAAAL LIBAffS 98000070#00e818E0U080E0 2.4 300 LEASITSSPECSF6ED I
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l FIGURE 3.2.3 4 Kp
]
POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 ALL LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed 13.4 Kw/ft for 8X8R/P8X8R/8P8X8R fuel or 18.0 Kw/ft for 7X7 fuel.
l APPLICABILITY: CONDITION 1, when THERMAL POWER 225% of RATED THERMAL POWER.
ACTION:
With the LHGR of any fuel rod exceeding the limit, initiate corrective action within115 minutes and continue corrective action so that the LHGR is within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
$URVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit; a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
When THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL R00 PATTERN'FOR LHGR.
~
HATCH - UNIT 2 3/4 2-8 Amendment No. 33, 66
4
' REACTIVITY CONTROL SYSTEMS BASES l
CONTROL RODS (Continued) than has been analyzed even though control rods with-inoperable accumulators may still be inserted with normal drive water pressure. Operability of the j
accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactors.
Control rod coupling integrity is required to ensure compliance with i
the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after each refueling. The subsequent check is performed as a backup to the initial demonstration.
i In order to ensure that the control red patterns can be followed and therefore that other parameters are within their limits, the control rod j
position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a
{
control rod to less than (3) inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not i
contribute to any damage to the primary coolant system. The support is not required when there is no pressure to-act as a driving force to rapidly eject a drive housing.
l The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL R00 PROGRAM CONTROLS l
Control rod withdrawal and insertion sequences are established to l
assure that the maximum insequence indiv.idual control rod or control rod I
segments which are withdrawn at any time during the fuel cycle could not
(
be worth enough to cause the peak fuel enthalpy for any postulated control rod accident to exceed 280 cal /ge. The specified sequences are characterized i
by. homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is a 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus, requiring the RWM to l
be OPERABLE below 20% of RATED THERMAL POWER and the RSCS to be OPERABLE from 50% control rod density to 20% of RATED THERMAL POWER provides adequate i
i control.
i I
r l
HATCH - UNIT 2 B 3/4 1-3 i
Amendment No. 66 i
l
~._.
2.: ; : L REACTIVITY CONTROL SYSTEM BASES i
C0h'1"ROL RCD PROGRAM Coh"TROLS (Continued)
The RSCS and RVM provide automatic supervision to assure that cut-of-sequence rods will not be vi:hdrawn or inserted.
s The analysis of the rod drop accident is presented in Sectica 15.1.38 of the ySAR and the techniques of tiie analysis are presented in a topical Reference 1, and two supplements, References 2 and 3.
- repcrt,
~
The R3M is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during j
high power operation.
The RBM is only required to be operable when the Limiting Condition defined in Specification 3.1.4.3 exists. Two channels are provided. Tripping one of the channels will block erroneous rod with-d:swal soon enough to prevent fuel damage. This system backs up the wri::en sequence used by the operator for. withdrawal of control rods, yurther dis-cussion of the RBM system and power dependent setpoints may be fcund in hTDC-30474-P (Ref. 4).
3/4.1.5 STANDBY LIQUID Coh" TROL SYSTEM The standby liquid control system provides a. backup capability for maintaining the reactor suberitical in the event that insufficient rods inserted in the core when a scram is called for. The volume of the art
' poison solution and weight percent of poison material in solution is based on being able to bring the reactor to the subcritical condition as i
'the plant cools to ambient condition. The temperature requigament is necessary to keep the sodium pentaborate in solution. Checking the volume and temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Vith redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is
. permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant c,omponents inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the systes. Once the solution is established, boron concentration will not v'ary unless more boron water is added; thus, a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
l i
?
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B 3/4 1 4 Amendment No. 39 i
j' nam m,- - -
m e-a as-,-r-----
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vw -
warm.-w,---
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3/4.2 POWER DISTRIBUTION LIMITS BASES i
The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident
' will not exceed the 2200*F limit specified in the Final Acceptance Criteria (FAC) issued.in June 1971 considering the postulated effects of fuel pellet j
densification. These specifications also assure that fuel design margins are j
' maintained during abnormal transients.
i 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1
This specification assures that the peak cladding temperature following i
' the postulated design basis loss-of-coolant accident will not exceed the limit j
specified in 10 CFR 50, Appendix K.
The peak 'ladding temperature (PCT) following a postulated loss-of-coolant c
1 accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The piak i
clad temperature is' calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.
This LHGR times 1.02 is used in the heatup code along with the exposure i
dependent steady state gap conductance and rod-to-rod local peaking factor.
The Technical Specification APLHGR is this LHGR of the highest powered rod i
divided by its local peaking factor. The limiting value for APLHGR is shown in the figures for in Technical Specification 3/4.2.1.
The calculational procedure used to establish the APLHGR shown in the i
figures in Technical Specification 3/4.2.1 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K l
to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
Differences in this analysis compared to previous analyses performed with Reference 1 are:
(1) the. analysis assumes a fuel t
n assembly planar power consistent with 102% of the MAPLHGR shown in the figures in Technical Specification 3/4.2.1; (2) fission product decay is computed assuming an energy release rate of 200 MEV/ fission; (3) pool boiling is.
j assumed after nucleate boiling is lost during the flow stagnation period; and
'(4) the effects of core spray entrainment and counter-current flow limitation i
as described in Reference 2, are included in the reflooding calculations.
A flow dependent correction factor incorporated into Figure 3.2.1-12 is i
appited to'the rated conditions APLHGR to assure that the 2200 'F* PCT limit is j
complied with during a LOCA initiated from less than rated core flow.
In addition, other power and flow dependent corrections given in Figures 3.2.1-12 and 3.2.1-13 are applied to the rated conditions to assure that the fuel 3
thermal-mechanical design criteria are preserved during abnormal transients 1
initiated from off-rated conditions.
- ~ A list of the significant plant input parameters to the loss-of-coolant l
accident analysis is presented in bases Table B 3.2.1-1.
Further discussion
(
of the APLHGR limits is given in Reference 4.
HATCH - UNIT 2 B 3/4 2-1 Amendment No. 21, 28, 33, 39, 66 I. - - -..
Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS c
FOR HATCH-UNIT 2 Plant Parameters:
Core The rmal Power.....................
2531 Mwt which corresponds i
to 105% of license core power
- I Vessel Steam Output....................
10.96 x 10' l bm/h which l
corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.............
1055 psia f
j
. Design Basis Recirculation Line Break Area For:
i a.
Large Breaks................... 4.0, 2.4, 2.0, 2.1 and 1.0_ft*
b.
Small Breaks................... 1.0, 0.9, 0.4 and 0.07 ft 2
f Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL i
FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft)
FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18 rc
~
A more detailed list of input to each model and its source is presented'in Section II of Reference 1 and subsection 6.3.3 of the FSAR.
- This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification linear heat generation rate limit.
E' HATCH - UNIT 2 8 3/4 2-2
- 2
.i" ' "" "POVER DISTRDLhTOW C' FITS' ~" ~
~ ~ ~ ~
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- ~
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i*
BASES
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3/t. 2.2 APRM SETPOINTS his sortion deleted.
i 3/t. 2.3 MINIML"1 CRITICAL POVER RATIO l
The required operating limit MCPRs at steady state cperating conditicas as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07, and an analysis of abnor=al cperational transients.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification 2.2.1.
T.
To assure that the fuel cladding integrity Safety Limits are not during any anticipated abnormal operational transient, exceeded the most limiting transients have been analyzed to determine which results in the largest reduction.in CRITICAL POWER RATIO (CPR).
Tha type of transients evaluated
% ere loss of f1'ow, increase in pressure and power, positive reactivity
~
insertion, and coolant. temperature decrease.
i
~
l l
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O I (i EA70H-2 3 3/' 2-3 Amendment No JA, gy, 39 l
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.1-6 that are input to a GE-core dynamic behavior transient computer program described in NE00-10802'". Also, the i
void reactivity coefficients that were input to the transient calculational procedure are bas ~ed on a new method of calculation termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR form the input for ferther analyses of the thermally limiting bundle with the single channel transient ~ thermal hydraulic SCAT code described in 4
NED0-20566' ". The principal result of this evaluation is the reduction in MCPR caused by the transient.
The-purpose of the'MCPR, and the Kp of Figures 3.2.3-3 and 3.2.3-4, re-l f
spectively is to de' fine operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRf and MCPRp at the existing core flow and power state. The MCPR s are established to protect the core from inadvertent core f
flow increases such that the 99.9% MCPR limit requirement-can be assured.
The MCPR s were calculated such that for the maximum core flow rate and the corresponding THERMAL POWER along the 105% of rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calculated at different points along the 105% of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPR.
f The core power dependent MCPR operating limit MCPR is the power rated flow MCPR operating limit multiplied by the Kp factor given in Figure 3.2.3-4.
l The Kps are established to protect the core from transients other than core flow increases, including the localized event such as rod withdrawal error. The K s were determined based upon the most limiting transient at the given core power plevel.
For further information on MCPR operating liniits for off-rated conditions, reference NEDC-30474-P.
+-
(
HATCH - UNIT 2 8 3/4 2-4
~
Amendment No. 33, 39, 66
1 i
(
3/4.9 RIr"G CPI?x"C"S 3 ASIS y
3/4.9.1."C-CCR XCCI S;Ci D:c'<i.g the C?I?MI.I reacter rede swicch in the refuel pcsition ensures dat the restrictions en red widder. al and refueling pla:fer n me s r.:
during ene refueling c;eratiens -ere pr:per'.y activa:ed.
These cend.::: s reinforce the refueling precedures and reduce the pretac lity of. inadrertsn:
criticality, damage the reactor internals or fuel asse-blies, and ex;csure of personnel to excessive radioactivity.
3/4.9.2 TNSTRtFENTATICN 2e CPI?MILITY of at least two source range moniters ensures that
~
redundant monitering capability is available to detect changes in tne reactivity conditien of the core. Curing the unicading, it is net necessary to maintain '3 eps because core alteraticns will involve only reactivity rencval and' will not result in criticality.
The loading of up to fcur bundles around. the SD:s before attaining the 3 cps is permissible because these bundles were in suberitical cenfiguration wnen they were reeved and therefere will remain suberitical when placed back in the previous pcsitiens.
3/4.9.3 CON: POL RCD POSITICN
'Ibe requirenant that all ' control rods be inserted during CCRE O.
ACERATICNS ensures that fuel will not be loaded into a cell without a
centrol red and prevents two positive reactivity changes frm occurring simultaneously.
)
3/4.9.4 DECAY TIME
':he minimm requirenant for' reactor suberiticality prior to. fuel movenent ensures that sufficient time has' elapsed to allow the radicactive decay of the short lived fission products.
Bis decay time is consistent.
with the assm ptions used in the accident analyses.
3/4.9.5 SEC2CARY CCNIAIWENT
" secondary contaiment is designed to minimire any ground l i
of radioactive material which may result fra an accident. ~evel release The reacter WWmy provides* secondary contalment during normal cperation when the drywell is sealed and in service. When the reactor is shutdown or during L
refueling, the drywell may be open and the reactor building then becmes the primary contaiment. De refueling floor is maintained under the secondary contaiment integrity of Hatch-Unit 1.
Establishing and maintaining a vacum in the building with the ' standby T.
gas treatment system once per 18 months, along with the surveillance of the doors, hatches and dampers, is ' adeguate to ensure that there are no violations of the integrity of the secondary contaiment. Only one closed danper in each penetration line is require $ to maintain the integrity of the secondary contalment.
HA':S
'OC 2 B 3/4 9-1 Amendment No. 2%,39 l
L mem-..o.-
REFUELING OPERATIONS BASES
)
3/4.9.6 COMMUNICATIONS I
The requirement.for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.
i 3/4.9.7 CRANE AND HOIST OPERABILITY The OPERABILITY requirements of the cranes and hoists used for movement of fuel assemblies ensures that: (1) each has sufficient load capacity to lift a fuel element, and (2) the, sore internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.8 CRANE TRAVEL-SPENT' FUEL ~ STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel element over irradiated fuel assemblies ensures that no more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumeo in the accident analyses. All fuel loaded into the Edwin I.
Hatch Nuclear Plant spent fuel pool shall have an uncontrolled lattice k.
less than or equal to the limit for high density fuel racks describad in the
" General Electric Standard Application for Reactor Fuel" (GESTAR II),
NEDE-24011-P-A-8. Alternatively, fuel not described in GESTAR II shall have been analyzed with another NRC approved methodology to ensure conformity to the FSAR design basis for fuel in the spent fuel racks.
3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND WATER LEVEL-SPENT FUEL STORAGE POOL The restrictions on minimum witer level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum l
water depth is consistent with the assumptions of the accident analysis.
l 3/4.9.11 CONTROL ROD REMOVAL This specification ensures that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN 1
MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.
HATCH - UNIT 2 B 3/4 9-2 4
Amendment No. 66
. m m
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a; DESIGN ~ FEATURES 5.0 5.1 SITE EXCLUSION AREA
~
5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
LOW POPULATION ZONE 5.1.2 The low population zone coincides with the exclusion area and is also shown in Figure 5.1.1-1.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary ~ containment is a steel structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a steel pressure vessel in the shape of a torus. The primary containment has a total minimum free air volume of 255,978 cubic feet.
DESIGN TEMPERATURE AND' PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:
a.
Maximum design internal pressure 56 psig.
b.
Maximum allowable internal pressure 62 psig.
c.
Maximum internal temperature 340*F.
~, " '
d.
Maximum external pressure 2 psig.
5.3 REACTOR CORE FUEL ASSEMBLIES s
5.3.1 The core shall consist of not more than 560 fuel assemblies and shall l
be limited to those fuel assemblies which have been analyzed with NRC approved codes and methods and have been shown to comply with all Safety Design Bases in the Final Safety Analysis Report (FSAR).
l l
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HATCH - UNIT 2 5-1
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Amendment,No. 21, 33, 66 l',,
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EXCt.USION AREA AND LOW POPULATION ZONE FIGURE 5.1.1-1 HATCH - UNIT 2 5-2 u
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= = me - -
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t............ =
h O! SIC; ??A"t?IS C*'CF.C. FCD ASSETL Is 5.3.2 2.e reac cr ccre shall centain 12-' crucifccm-shapei ccncrc' red i
asserclies.
I 5.4 Fr.ACICR CCC'.;20 STS r4 I
- .5IQ! :2255*.?E ;ND TD'_:I?ATUFI 5.4.1 The reacter coolant syste is designed and shall be maintained:
a.
In accordance with the code requirenents specified in Section 5.2 cf the ISAR, with allewance for nc,r=al degradatien pursuant to the applicable Surveillance Re:Jul aments, b.
Fcr a pressure of 1250 psig, and c.
For a tetperature of 575cy itLD2 5.4.2 ne tetal wa c =.r ard steen voltzne of the reacter vessel and recirculatien syste is ap.roximately 17,050 c'. ic feet at a ncminal T ave of 54Ccy, Ii 5.5 he.u.C.<i.CGICAL TCM I.CCATICN 5.S.1 n o meteorological tower shall be located as shown on Figure 5.1.1-1.
l 5.6 FUEL STORAGE l
~
exmCAun 5.6.1 ne new and spent fuel storage racks are designed and shall be maintained with sufficient center-to-center distance bee.eeen fuel assenblies placed in the storage racks to ensure a k gg sluivalent to $ 0.95 when i
flooded with unborated water.
Se kegg of
$ 0.95 includes conservative allowances for uncertainties.
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HATCH e UNIT 2 8
5-3 Amendment No.39 s
4 DESIGN FEATURES
?
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained
-to prevent inadvertent draining of the pool below elevation 185 feet.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2845 fuel assemblies.
5.7 - COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designe'd and shall be maintained within the. cyclic or transient limits of Table 5.7.1-1.
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1 HATCH - UNIT 2 5-4
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Amendment No.,15, 66
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