ML20052F755
| ML20052F755 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/06/1982 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20052F756 | List: |
| References | |
| TAC 48110, NPF-05-A-028 NUDOCS 8205130577 | |
| Download: ML20052F755 (12) | |
Text
sb.Le G%
'o UNITED STATES
- e 8 y, **,,,,g NUCLE AR REGULATORY COMMISSION t
W ASHINGTON. D. C. 20555
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GEORGIA POWER CGMPANY 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 OPERATING LICENSE Amendment No. 28 License No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated March'll,1982, complies with the standards and requirements of the Atonic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) 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 l
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 te the Technical Spec-ifications as indicated in the attachment to thi. license arndment and paragraph 2.C.(2) of Facility Operating License No. NPF '. is i
hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and.
B, as revised through Amendment No. 28, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
820S130577 820506 PDR ADOCK 05000366 P
L e.
3.
This license amend.nent is effective as of the date of its issuance...
FOR THE NUCLEAR REGULATORY COMMISSION (44 Jo rF. Stolz, Chief
/
Op rating Reactors Branch #4 ivision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 6,1982 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 28 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 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 vertical lines indicating the area of change. The overleaf pages are provided to maintain document completeness.
Insert Remove 3/4 2-1 3/4 2-1 3/4 2-4B 3/4 2-7a 3/4 2-7a B3/4 2-1
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'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) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5.
APPLICABILITY: CONDITION 1, when THERMAL POWER 3,25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the limits of Figures 3.C.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5, initiate corrective action within 15 minutes and continue corrective action so that APLHGR 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 THEP. MAL 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, 3.2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5:
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 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 c.
operating with a LIMITING CONTROL R0D PATTERN for APLHGR.
HATCH - UNIT 2 3/4 2-1 Amendment No. ST, 28.
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FUEL TYPE P8DRB283 MAXIMUM AVERAGE PLANAR LINEAR HEAT s
GENERATION RATE (MAPLHGR)
VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-5
'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) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5.
APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the limits of Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5, initiate corrective action within 15 minutes and continue corrective action so that APLHGR 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 />.
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, 3. 2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5:
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 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 c.
operating with a LIMITING CONTROL R00 PATTERN for APLHGR.
HATCH - UNIT 2 3/4 2-1 Amendment No.,24', 28.
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FUEL TYPE P80R8283 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLilGR)
VERSUS AVERAGE PLANAR EXPOSURE FIGURE 3.2.1-5
3/4.2.3 MINIMUM CRITICAL POWER RATIO SURVEILLANCE REQUIREMENTS (Continued) 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 requir'ed by Specification 4.1.3.2.
MCPR shall be determined to be equal to or greater than the applicable f
limit:
a.
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Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state oparating conditions have been established, and Initially and at least once per 12' hours when the reactor is c.
operating with a LIMITING CONTROL R0D PATTERN for MCPR.
f IHATCM-UNIT 2 3/4 2-7 Amendment No. 21
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MCPR LIMIT FOR P8X8R FUEL AT RATED FLOW FIGURE 3.2.3-2 HATCH - UNIT 2 3/4 2-7a Amendment No.X,28
3/4.2 POWER DISTRIBUTION LItilTS BASES 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 0
(FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature follow-ing the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K.
The peak cladding temperature (PCT) following a postulated loss-of-coolant 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 peak 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 dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification APLHGR is this LHGR of the highest powered rod divided by its local peaking factor.
The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5.
l The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5, is based on a loss-l of-coolant accident analysis. 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 compared l
to previous analyses performed with Reference 1 are: (1) the analysis assumes a fuel assembly planar power consistent'with 102% of the MAPi.HGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5; (2) fission product
[
decay is computed assuming an energy release, rate of 200 MEV/ fission; (3) pool boiling is assumed after nucleate boiling is lost.during the flow stagnation period; and (4) the effects of core spray entrainment and counter-current flow limitation as described in Reference 2, are' included in the reflooding calculations.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.
l HATCH - UNIT 2 B 3/4 2-1 Amendment No.,2f, 28 l
Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE -
LOSS-OF-COOLANT ACCIDENT ANALYSIS F10R HATCH-UNIT 2 Plant Parameters:
Core Thermal Power...............
2531 Mwt which corresponds to 105% of license core power
- 6 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:
2 a.
Large Breaks............ 4.0, 2.4, 2.0, 2.1 and 1.0 f t
?
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 GE0 METRY (kw/ft)
FACTOR RATI0 '
Initial Core 8x8 13.4 1.4 1.18 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.
/
HATCH - UNIT 2 B 3/4 2-2 t
e-.---
9 --.
~.
TABLE 3.3.5-1 (Continued)
CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION NOTE When THERMAL POWER exceeds the preset power level of the RWM and a.
RSCS.
b.
This function is bypassed if detector is reading > 100 cps or the IRM channels are on range 3 or higher.
This function is bypassed when the associated IRM channels are on c.
range 8 or higher.
Atotalof6IRMinstrumentsmustbeOP5RABLE.
d.
This function is bypassed when the IRM channels are on range 1.
e.
f.
With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.
l r
HATCN'- UNIT 2 3/4 3-39
l TABLE 3.3.5-2 E
-CONTROL R00 WITil0 RAMAL BLOCK INSTRUMEllTATION SETPOINTS pl f
TRIP FUf!CTI0ff TRIP SETP0_ItiT, ALLOHA01.E VALUE T
E 1.
APRf1 m
m a.
Flow Referenced. Simulated lhennal Powcr;- lipscale 1 (0.66 U + 427)*
1 (0.66 W + 42%)*-
l b.
Inopera t Ive I;A IIA l
c.
pounsca le 7 12/125 of full scale 7 12/125 of full scale
> 3/125 of full scale
> 3/125 of full scale d.
IIeutron flux - lligh,12%
2.
R00 DLOCK tt0filTOR a.
U;rcale
< (0.66W + 41%)Not to exceed 107%
(0.66 W + 41%)
[
j b.
Inupera tive ilA fiA c.
Downscale 1 3/125 of full scale
> 3/125 of full scale w
i 2
3.
50tIRCE RANGE MONITORS _
l w
l E
a.
Detector not ' full in flA NA 5
5 b.
Upscale i 1 x 10 cps 1 1 x 10 cp, c.
Inoperative NA flA d.
Downscale 2,3 cps
> 3 cps 4.
tilTERMEDIATE RANGE M0filTORS g
a.
Detector not full in NA NA 108/125 of full scale m
b.
Ilpsca le
< 108/125 of full scale 8
c.
Inoperative fiA liA y
5/125 of full scale 3,5/125 of full scale d.
Downsca le o
g 5.
SCRAM DISCllARGE VOLUME
.)
\\
a.
Water level-High
< 36.2 gallons
< 36.2 gallons y
- The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).
The trip setting of this function must be maintained in accordance with Sp~cci fica tion 3.2.2.
__