ML20149F503
| ML20149F503 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/22/1988 |
| From: | Jabbour K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149F507 | List: |
| References | |
| TAC-66524, TAC-66525, NUDOCS 8802170206 | |
| Download: ML20149F503 (19) | |
Text
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'o, UNITED STATES i
NUCLEAR REGULATORY COMMISSION 5
E W ASHINGTON. D. C. 20066
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GEORGIA POWER COMDANY OGLETHORPE POWER CORPORATION PUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-3?1 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 1 APENDMENT TO FACILITY OPERATING LICENSE Amendment No.151 License No. DPR-57 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant.
Unit 1 (the facility) Facility Operating License No. DPR-57 filed by Georgia Power Comoany acting for itself. Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated October 21, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regu-lations 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 reasonable assurance (1) that the activities authorized by i
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 set forth in 10 CFR Chapter I; 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 requirements have been satisfied.
8802170206 880122 PDR ADOCK 05000321 P
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-57 is hereby.
amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained'in Appendices A and B, as revised through Amendment No 151, 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.
FOR THE NUCLEAR REGULATORY COMMISSION Kahtan N. Jabbour, Acting Director Project Directorate 11-3 Division of Reactor Projects-I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 22, 1988 l
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I ATTACHMENT TO LICENSE AMENDMENT NO.151 o.
FACILITY OPERATING LICENSE NO. OPR-57
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DOCKET NO. 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 vertical lines indicating the areas of change.
Remove Insert Page Page 3.10-2 3.10-2 3.10-7 3.10-7 5.0-1 5.0-1 i
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LIMITlhG C040111045 FOR OPERATION SURVEILLANCE RE00lREMENT5 3.10.C.
Core monitorino Ourino Core 4.10.C.
Core Monitorino Durino core Alterations Alterations 1.
During norm) core alterations, two Prior to making normal alterations SRMs shall be operable; one in the I to the core the SRMs shall be i
core quadrant where fuel or control functionally tested and checked for rods are being moved and one in an neutron response. Thereafter, adjacent Quadrant, except as specified while required to be operable, the in 2 and 3 below.
SRM's will be checked daily for response.
For an SRM to be considered operable, it shall be inserted to the normal Use of special movable, dunking operating level and shall have a type detectors during initial fuel minimum of 3 cps with all rods capable loading and major core alterations of normal insertion fully inserted.
in place of normal detectors is permissible 45 long as the detector 2.
Prior to spiral unloading the SRMs
/
is connected to the normal SRM thall be proven operable as stated circuit.
above, however, during spiral unloading the count rate may drop Prior to spiral unloading or below 3 cps.
reloading the SRMS thall be l
functionally tested. Prior to 3.
Prior to spiral reload, up to four (4) spiral unloading the SRMs should j
fuel assemblies will be loaded into also be checked for neutron core positions next to response.
each of the 4 SRMs to obtain the required 3 cps. These assemblies may be any which have been shown to meet the criteria for storage in the spent fuel pool. Until these assemblies have been loaded, the 3 cps requirement is not necessary.
D.
Spent Fuel Pool Water level D.
Scent Fuel Poc1 Water Level Whenever irradiated fuel is stored in Whenever irradiated fuel is stored the spent fuel pool, the pool water in the spent fuel pool, the water level shall be maintained at or above level shall be checked and recorded 8.5 feet ab)ve the top of the active daily.
fuel.
C.
Control Rod Drive Maintenance E.
Control Rod Drive Maintenarce 1.
Recuirements for Withdrawal 1.
Reovirements for Withdrawal of 1 or 2 Control sods of 1 or 2 Control Rods A maximum of two control rods separated by at least two control cells in all directions may be withdrawn or removed from the core for the purpose of performing control rod det"e maintenance provided that:
a.
The Mode Switch is locked in the REFUEL position. The refueling a.
This surveillance requirement is interlock which prevents more than the same as given in 4.10. A.
one control rod from being withdrawn may be bypassed for one of the control rods on which maintenince is being HATCH - UNIT 1 3.10-2 Amendment No. 151
1 Ra5ES FOR LIMITlWG Conolfl0NS FOR OPEtaTION s
3.10.A.2.
Fuel Grancle Noist load Settine Interlocks Fuel handling is normally conducted with the fuel grapple hoist. The total lead on this hoist wnen the interlock is required consists of the weight of the fuel grapple and the f uel assembly. This total is approximately 1500 lbs. in comparison to the load setting of 485 1 30 lbs.
3.
Auxiliary 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 being handled.
8.
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 l,
refueling interlocks, as designed, will prevent inadvertent criticality.
C.
Core Monitorino Durine Core Alterations TPe SRNs are provided to monitor the core during periods of Unit shutdown and to l guide the operator during refueling operations and Unit startup. Requiring two operable SRMs in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. l The requirements of 3 counts per second provides assurance that neutron flus 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 SRMs before attaining the 3 cps is permissible because these bundles form a subtritical configuration.
D.
Spent Fuel Pool W4ter level i
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 tut 4 water level at or in excess of eight and one-half feet over the top of the ara., fuel will provide shielding such that the maximum calculated radiolo, scal doses do not exceed the Itaits of 10 CFR 20. The normal water level providrs 14-1/2 feet of additional water shielding. All penetrations of the fuel peel 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 1. Hatch Nuclear Plar.t spent fuel pool shall have an uncontrolled lattice Ka less than or equal to the liett for high-density fuel racks described in the l
'6eneral Electric Standard Application for Reactor Fuel' (H5 TAR !!).
NEDE-24011 P-A-8.
Alternatively, fuel not described in M5fAR 11 shall have been analyzed with another NRC-approved methodology to ensure conformity to the F5At design basis for fuel in the spent fuel racks.
E.
Control Rod Drive Maintenance Ouring certain periods, it is desirable to perform maintenance on two control rod drives at the same time.
i HATCH - UNIT 1 3.10-1 Amendment h'o 151
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5.0.
MAJOR DESIGN FEATUaES A.
}itt Edwin 1. Hatch Nuclear Plant Uait 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 Mertator Coordinates of the center of the reactor building are: tone 17R LF 372.935.2m E and 3.533.765.2m N.
B.
Reactor rdDt 1.
Fuel Assemblies 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 (F5AR).
2.
Control Rods The reactor shall contain 137 cruciforn shaped control rods.
C.
Reactor Yessel The reactor vessel is described in Table 4.2-2 of the F5AR. The applicable design specifications shall be as listed in Table 4.2-1 of the FSAR.
D.
Containment 1.
Primary Containment The principal design parameters are characteristics of the primary containment shall be as given in Table 5.2-1 of the FSAR.
2.
Secondary containment * (See Page 5.0-1a)
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.
3.
Primary Containment Penetrations Penetrations to the primary containment and piping passing through such penetratioos shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.
E.
Fuel Storane 1.
Soent Fuel All arrangeeents of fuel in the spent fuel storage racks and in other credible configurations in the spent fuel pool outside the racks shall be evaluated and shown to have a Keff not greater than 0.95.
2.
NewFM The new fuel storage vault shall be such that the kef t dry shall not be greater than 0.90 and the kort flooded shall not be greater than 0.95.
HATCH - UNIT 1 5.0-1 Amendment No. 151
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UNITED STATES '
NUCLEAR REGULATORY COMMISSION 2
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORAT!ON MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE knendment No. 89 License No. NPF-5 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No NPF-5 filed by Georgia Power Company, acting for itself Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated October 21, 1987, complies with the standards and recuiretrents 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 confonnity with the application, the provisions of the Act, and the rules and regulations of the i
Comission; C.
There is reasonable assurance (1) that the activities authorized by this anendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Cocr11ssion's regulations set forth in 10 CFR
- ~
Chapter I; 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 requirements have been satisfied.
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- 2. Accordingly. the license is amended by changes to the Technical Specifi-l cations as indicated in the attachment to this license amendtnent, and paragraph 2.C.(2) of Facility Operating License No. NPF-5_is hereby amended to read as follows
.(2)
' Technical Specifications
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'The Technical Specifications contained in Appendices A and B, as revised through knendment No. 89, are hereby incorport ted 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.
FOR THE NUCLEAR REGULATORY COMMISSION 4
Kahtan N. Jabbour Acting Director Project Directorate 11-3
[
Division of Reactor Projects-l/II
Attachment:
Changes to the Technical i
Specifications Date of Issuance: January 22, 1988 P
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ATTACHMENT TO LICENSE AMEN 0 MENT NO. 89 FACILITY OPERATING LICENSE NO. NPF 5 DOCKET NO. 50-366 Replace the following pages of the Appendix A Technical Specifications with the enclosed paces.
The revised Dages are identified by amendwent number and contain vertical lines indicating the areas of change. Corresponding overleaf pages are provided to maintain document completeness.
Remove Insert Page Page 3/42-3 3/42-3 3/4 2-4j 3/4 2-4j 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 9-4 3/4 9-4 B 3/4 2-2 8 3/4 2-2 B 3/4 2-4 8 3/4 2-4 8 3/4 9-1 B 3/4 9-1 5-3 5-3
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NERAGE PLANAR LINEAR HEAT GENERATION RATE vs c*
NERAGE PLANAR EXPOSURE
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NERAGE PLANAR EXPOSURE (GWd/st) h FIG AE 3.2.5-2 o_m,.um -.x.uu
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- MAPU40Rgg MAPLHGngg - STANDARD ASAPLHOR LIMITS 4
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MAPRSULT - 1.0 FOR FLOW > 81%
- O se FOR FLOW 581%
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c FOR 7X7 FOft PeMeR Maxmuss exe exer AND 9XS LFAs r
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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 ALL MINIMUM CRITICAL POWER RATIOS (MCPRs) for two-loop operation, shall be equal to or greater than the MCPR operating limit (OLMCPR), which is a function of average scram time, core flow, and core power.
For 25%
5 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 applicable limit determined from Figure 3.2.3-3, or b.
The appropriate Kp given by Figure 3.2.3-4, multiplied by the appropriate limit from Figure 3.2.3-1 or 3.2.3-2, where t is the relative measured scram speed with respect to Option A and Option B scram speeds.* If t is determined to be less than zero, then the OLMCPR is evaluated at t = 0.
For single-loop operation, the MCPR limit is increased by 0.01 over the comparable two-loop value.
APPLICABILITY: CONDITION 1, when THEMAL POWER 2 25% RATED THERMAL POWER
- The specific formula for determining t is provided in plant procedures.
HATCH - UNIT 2 3/4 2-6 Amendment No. 89
3/4.2.3 MINIMUM CRITICAL POWER RATIO (CONTINUED) l ACTION:
With MCPR less than the applicable limit determined from Specification 3.2.3.a, or 3.2.3.b for two-loop or single-loop operation, 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 />.
SURVEILLANCE REQUIREMENTS l
4.2.3 The MCPR limit at rated flow and rated power shall be determined for each type of fuel (8X8R, P8X8R, BP8X8R, 9X9 LFA, and 7X7) from Figures 3.2.3-1 l
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 is determined from scram time measurements performed in accordance with Specification 4.1.3.2.
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, 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 MCPR.
Amendment No. 89 HATCH - UNIT 2 3/4 2-7
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ALL 8 X 8R AND 9 X 9 LFA FUEL TYPES FIGURE 3.2.31 l
HATCH - UNIT 2 3/4 2-7a Amendment No. 89
INSTRUMENTATION SURVEILLANCE REQUIREMENTS CONTINUE 0 b.
Performance of a CHANNEL FUNCTIONAL TEST:
1.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.
At least once per 7 days.
Verify that the channel count rate is at least 3 cps at least once c.
per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except:
1.
The 3 cps is rot required during core alterations involving only fuel unloading provided the SRMs were confirmed to read at least 3 cps initially and were checked for neutron response.
2.
The 3 cps is not required initially on a full core reload.
prior to the reload, up to four fuel assemblies will be loaded into core positions next to eacn of the 4 SRMs to obtain the required count rate.
These assemblies may be any which have been shown to meet the criteria given in Section 5.6.1 of these Technical Specifications for storage in the spent fuel pool.
d.
Verifying that the RPS circuitry "shorting links" have been removed and that the RPS circuitry is in a non-coincidence trip mode within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to starting CORE ALTERATIONS or sht?iown me.rgin demonstrations.
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Amendment No. 89 HATCH - UNIT 2 3/4 9 4
Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS FOR HATCH-UNIT 2 #
Plant Parameters:
Core Thermal Power.....................
2531 Mwt which corresponds to 105% of license core power
- Vessel Steam Output..................... 10.96 x 10' lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.............
1055 psia Design Basis Recirculation Line Break Area For:
a.
Large Breaks.................. 4.0, 2.4, 2.0, 2.1 and 1.0 ft 8 b.
Small Breaks................... 1.0, 0.9, 0.4 and 0.07 ft*
Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft)
FACTOR RATIO initial Core 8x8 13.4 1.4 1.18 A more detailed list of input to each model an'd its source is presented in subsection 6.3.3 of the FSAR.
For convenience, the APLHGR limits are reported in the units of kW/ft, which is the bundle planar power normalized to the number of fueled rods.
Figure 3.2.1-2 shows that the 9x9 LFAs have the same planar power limits as the GE P80RB284H fuel; however, on a kW/ft basis, the APLHGR limits for the LFAs are 62/79 times the P80RB284H limits.
- 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.
- These are the initial core input parameters.
For the updated Loss-of-Coolant Accident Analysis using SAFER /GESTR-LOCA, see Reference 4.
HATCH - UNIT 2 8 3/4 2-2 Amendment No. 89
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
As depicted on. Figure 3.2.3-1 or 3.2.3-2 the 100% power,100% flow operating limit MCPR (OLMCPR) depends on the average scram time, t, of the control rods, where:
= 0 or ' ave
'9, whichever is greater t
tg, tB t
where:
A = 1.096 sec (Specification 3.1.3.3, scram time limit to notch 36)
'8 = 9 + 1.65 3
1/2, N
n IN I 1=1,
where:
u = 0.822 see (mean scram time used in the transient analysis) o =.018 see (standard deviation of 9) n intgg ave =
i=1 inIN 1
i=1 where:
n = number of surveillance tests performed to date in the cycle i = number of active control rods measured in the ith surveillance test T 1 = average scram time to notch 36 of all rods in the ith surveillance test N 1 = total number of active rods measured in 4.1.3.2.a.
The purpose of the MCPR, and the Kp of Figures 3.2.3-3 and 3.2.3-4, f
respectively is 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 MCPRo at the existing core flow and power state. The MCPRgs are established to protect the core from inadvertent core 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 f
the corresponding 7HERMAL 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 Safsty 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 rated power p
and rated flow MCPR operating limit multiplied by the Kp factor given in Figure 3.2.3-4.
i The Kps are established to protect the core from transients other than core l
flow increases, including the localized event such as rod withdrawal error. The Kps were determined based upon the most limiting transient at the given core power level. For further information on MCPR operating limits for off-rated conditions, see NEDC-30474 P (Reference 2).
HATCH - UNIT 2 B 3/4 2-4 Amendment No. 89
3/4.9 ' REFUELING OPERATIONS BASES 3/4.9.1 REACTOR MODE SWITCH Locking the OPERABLE reactor mode switch in the refuel position ensures that the restrictions on rod withdrawal and refueling platform movement during the refueling operations are properly activated.
These conditions reinforce the refueling p*ocedures and reduce the probability of inadvertent criticality, damage the reactor internals or fuel assemblies, and exposure of personnel to excessive radioactivity.
3/4.9.2 INSTRUMENTATION The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. During the 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 bundles around the SRMs before attaining the 3 cps is permissible because these bundles form a subcritical configuration.
l 3/4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control red and prevents two positive reactivity changes from occurring simultaneously.
3/4.9.4 DECAY TIME The minimum requirement for reactor suberiticality prior to fuel movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.
This decay time is consistent with the assumptions used in the accident analyses.
3/4.9.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building provides secondary containment during normal operation when the drywell is sealed and in service. When the reactor is shutdown or during refueling, the drywell may be open and the reactor building then becomes the primary containment.
The refueling floor is maintained under the secondary containment integrity of Hatch-Unit 1.
Establishing and maintaining a vacuum in the building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches and dampers, is adequate to ensure that there are no violations of the integrity of the secondary containment.
Only one closed damper in each penetration line is required to maintain the integrity of the secondary containment.
HATCH - UNIT 2 B 3/4 9-1 Amendment No. 89
DESIGN FEATURES -
CONTROL ROD ASSEMBLIES 15.3.2f The reactor core shall contain 137 cruciform-shaped control rod assemblies.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The _ reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure ofl1250 psig, and c.
For a temperature of 575'F VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 17,050 cubic feet at a nominal Tave of 540'f.
5.,5 METEOROLOGICAL TOWER LOCATION 5.5.1 The primary and backup meteorological towers shall be located as shown on Figure 3.11-1.
5.6 FUEL STORAGE CRITICALITY l
5.6.1 The new and spent fuel storage racks are designed and shall be maintained with sufficient center-to-center distance between fuel assemblies placed in the storage racks to ensure a keff equivalent'to s 0.95 when flooded with unborated water. The keff of s 0.95 includes conservative allowances for uncertainties in calculations of both normal and abnormal storage conditions as specified in the FSAR.
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HAT H - UN:T 2 5-3 Amendment No. 89 S