ML20154E290
| ML20154E290 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/12/1988 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20154E292 | List: |
| References | |
| TAC-68688, NUDOCS 8809160283 | |
| Download: ML20154E290 (17) | |
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UNITED STATES y
g NUCLEAR REGULATORY COMMISSION n
l WASHINGTON, D. C. 20666
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGM DOCKET NO. 50-321 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No.
157 License No. DPR-57 1.
The Nuclear Regulatory Comission (the Comission) has found that:
The app (lication for amendment to the Edwin I. Hatch Nuclear Plant, A.
Unit 1 the facility) Facility Operating License No. DPR-57 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (tle licensee) dated June 20, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; l
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 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 l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have l
been satisfied.
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Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachrent to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby arnended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.157, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license arnendment is effective as of its date of issuance and shall be iroplemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate II-3 Division of Reactor Projects-I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: September 12, 1988 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 157 FACILITY OPERATING LICENSE NO. OPR-57
, DOCKET NO. 50-321 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised page is identified by amendment number and contains a vertical line indicating the area of change.
Reinove Page Insert Page X
X Xi Xi 3.11-la 3.11-la 3.11-2 3.11-2 3.11-3 3.11-3 3.11-4 3.11-4 3,11-4a 3.11-4a 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 8)
Figure 3.11-1(Sheet 8)
Figure 3.11-2 Figure 3.11-2 4
Figure 3.11-4 Figure 3.11-4 Figure 3.11-5 Figure 3.11-5 i
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LIST OF FIGUets Eiggt Title
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1.1 -1 (Deleted) l 2.1 -1 Reactor Yessel Water Levels i
4.1 -1 Graphical Aid for the Selection of an Adequate Interval Between Tests
- 4. 2 -1 System Unavailability
- 3. 4 -1 Sodium Pentaborate Solution Volume Versus Concentration Requirements 3.4-2 Sodium Pentaborate Solution Temperature Versus Concentration Requirements 3.6-1 Pressure versus Minimum Temperature for Pressure Tests i
Based on Surveillance Test Results 3.6-2 Pressure versus Minimum Tamperature for Non-nuclear Heatup/Cooldown and Low-Power Physics Test 3.6-3 Pressure versus Minimum Temperature for Core Critical Operation other than Lw-Power Physics Test (includes 40'F Margin Required i
by 10 CFR 50 Appendix G) 3.6-4 Deleted 3.5-5 Power-Flow operating Map with One Reactor Coolant System Recirculation 1.000 in Operation 3.11-1 (Sheet 1) Limiting Value for APLHGR (Fuel Types OP80RB265H.
P80R8265H, SP80R284H, and P80RS284H) 3.11-1 (Sheet 2) Limiting Value for APLHGR (Fuel Types 8P80R8283, P80RB283, SP80RS299, and P90Td299) 3.11-1 (Sheet 3) Limiting Value for APLHGR (Fuel Types 8P80R8301L.
P80R8301L and 1987 Hatch LTAs) 3.11-1 (Sheet 4) Limiting Value for APLHGR (Fuel Type 9x9 LFA) l 3.11-1 (Sheet 5) Limiting Value for APLHGR (Fuel Type 80296A) l 3.11-1 (Sheet 6) Deleted.
3.11-1 (Sheet 7) MAPFACp (Power Dependent Adjustment Factors to MAPLHGRs) j 3.11-1 (DJet 8) MAPFACF (Flow Dependent Adjustment Factors to MAPLHGRs) 3.11-2 (Deleted) l 3.11-3 MCPRp (Flow Dependent Adjustment Factors for MCPRs) 3.11-4 MCPR Limit for All 8 8 and 9:9 Fuel Types for Rated Power and Rated Flow l
HATCH - UNIT 1 Amendment No. 157 l
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LIS? 0F FfGURES
. M Title 3.11-5 (Deleted) l 3.11-6 Kp (Power Dependent Adjustment Factors for MCPRs) 3.15-1 Unrestricted Area Boundary l
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HATCH - UNIT 1 at Amendment No. 157
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LINITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS 3.11.8.
Linear Heat Generation Rate (LH&R) 4.11.8.
Linear Heat Generatica Rate (LMGR)
Ouring power operation, the LHGR The LHGR shall be checked daily l
Shall not exceed the limiting during reactor operation at 1 255 value of 14.4 kW/f t for Gt8x8(8 rated thermal power.
fuel or the limiting value of 13.4 kW/f t for any other 8 x 8 fuel. If at any time during HATCH = UNIT I 3.11 14 iU*.endment No. 137
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j LIMITING CONDITIONS FOR OPERATION SURVE!LLANCE AEOUIREMENTS 3.11.8.
Linear Heat Generation Rate (LH6R)
(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 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then reduce reactor power to less than 25 percent 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 />.
If the limiting condition for opera-tion is restored prior to expiration of the specified time interval, then further progression to less than 25 percent of rated thermal power is not required.
C.
Minieue Critical Power Ratio (MCPR) 4.11.C.1.
Minimus Critical Power Ratig,1M_CF The minimum critical power ratio MCPR shall be determined to be (MCPR) for two-loop operation shall equal to or greater than the be equal to or greater than the applicable limit, daily during operating limit MCPR (CLMCPR), which reactor power operation at 125-is a function of scram time, core percent rated themel power and power, and core flow. For 25 percent following any change in power leve.
$ power < 30 percent, the OLMCPR is or distribution that v uld cause given in Figure 3.11-6.
For power l
operation with a limiting control 1 30 percent, the OLMCPR is the rod pattern as described in the greater of either:
bases for Specification 3.3.F.
1.
The applicable limit determined from Figure 3.11-3, or l 4.11.C. 2.
Minimum Critical Power eatio Lhy 2.
The applicable limit from The MCPR limit at rated flow and Figure 3.11-4 multiplied l
rated power shall be determined foi by the K, f actor determined each fuel type, as appropriate f rom Figure 3.11-4, where e I
from Figure 3.11-4, using:
l is the relative measured scram speed with respect to Option A a.
t=1.0 prior to initial scram and Option 8 scraa speeds. If time measurements for the t is determined to be less cycle. performed in accordant (
than zero, then the OLMCPR with Specification 4.3.C.2.a.
is evaluated at t=0.
or i
b.
t is determined from scram time measurements performed in accordance with Specifica-tion 4.3.C.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 4.3.C.2.
l HATCH - LIMIT 1 3.11-2 Amendment No. 157 i
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BASES FOR LIMITING CON 0!T!0=$ FOR OPERATION AND SUAVEILLANCE 8E00!REMENTS 3.11.
FUEL e005 A.
Averaee Planar Linear West Generation eate f APLHGR1 This specification assures that the peak cladding temocrature following the l
postulated design basis loss-of-coolant accident (LOCA) will not exceed the l
limit specified in 10 CFR 50.46 even considering the postulated effects of fuel pellet dentification.
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 af fect the calculated peak clad temperature by less than i 20'F relative to the 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 APLMGR at rated conditions is shown in figures 3.11-1, sheets 1 thru 6.
l For convenience, the APLHG4 timits are reported in the units of kW/f t, I
which is the bundle planar power nonealized to the num6er of fueled rods.
Figure 3.11-1 (Sheet a) shows that the 9:9 LFAs have the same planar power l
limits as the 6C t/P90RB20aN fuel; however, on a kW/f t basis, the APLHG4 limits for the LFAs are 62/79 times the B/P90RB204H limits.
l The actual APLHG4 limits for G(0:0[B fuel are lattice-type dependent and are explicitly modeled in the process computer. At each esposure, the l
Technical Specifications APLHGR limit is defined 45 the most limiting
)
value of all the enriched lattices. The Technical Specifications APLH64 limits will be used for manual calculations.
The calcu14tional procedure used to establish the APLHGR shown in figures 3.11-1, sheets 1 thru 6, is based on a LOCA analysis. The analysis was l
perfonned using General Electric (GE) calculation 41 models which are consistent with the requirements of Appendia K to 10 CFR 50. The LOCA l
analysis sas performed utilizing the new imoroved calculational model,
$AF(R/GESTR-LOCA. The analysis demonstrated that loss-of-coolant concerns do not limit the opar& tion of the fuel tmce margin to the 2200'F limit was demonstrated (Reference 9).
Therefore, the API,HGR limits for the fuel types shown in figure 3.11 1 are derived to assure that the fuel thermal-mechanical design criteria are met.
A list of the significant plant input paranetrr3 to the LOCA analysis is l
presented in tables 4-1 and 4-2 of Ref erence 9.
Further discussion of the APLH64 bases is found in NC0C-30474 P(*).
A flow dependent correction f actor incorporated into figure 3.11-1 (sheet 8) is apl11ed to the rated conditto") APLH64 to assure that the 2200'F PCT limit is comolted with during LOCA in' tied f rom less than rated core flow. In addition, other power and fit o edent corrections given in figure 3.11-1 (sheets 7 and 8) are applied to the rated conditions APLH64 limits to assure that the fuel thermel-eechanical design criteria are met during abnorme1 transients initiated f rom of f rated conditions for two-loop and single-loop operations, References 2 and 8.
For single-loop operation 4 0.75 multipItca-tion f actor to APLH64 limits for all fuel bundle types Conlervatively bounds that required by Reference 2.
For single-loop operation (SLO), the most restrictive of the SLO and ARTS (*) MAPLHGas will define the Limiting Conditon for Operation.
MATCH - UNIT 1 3.11 3 Amendment No. 157
8ASESFORLIMITINGCONDITIONSFOROPERAT!cmA10syJVEIMANCEeEQUleENEnf5 3.11.0.
Linear Heat Generation eate (LH64)
This specification assures that the LHG4 in any rod is less than the design linear heat generation if fuel pellet densification is postulated, for LNGR to be a limiting value below 25-percent rated thermal power, the ratio of peat LMAP to core average LHGR would have to be greater than g.6, which is precluded by a considerable margin when teoloying any permissible control rod pattern.
C.
Minimum Critical Power eatio (MCPR)
The required operating limit MCPR 45 specified in Specification 3.11.C. is derived f rom the established fuel cladding integrity Saf ety Limit MCPR and an analysis of abnormal operational transients presented in References 1, 2, and 8.
Various transient events will reduce the MCPR below the operating MCPR.
To assure that the fuel cladding integrity safety limit is not violated during anticipated abnormal operational transients, the most limiting transients have been analyzed to determine which one retWIts in the largest reduction in critical power ratio (a MCPR). Addition of the largest
& MCPR to the safety limit MCPR gives the minimum operating limit MCPR to avoid violation of the safety limit should the most limiting transient occur.
The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
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t MATCH - UMlf 1 3.11 4 Arenhent No. 157 l
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BASES FON LIMITING CONDITICMS FCs OPEeATION AND SUQy([(LANC[ QEQUIREMENTS 3.11. C. Minimum critical Powe satio (MCPel (Continued)
According to Figure 3.11-4 the 100-percent power,100-percent flow l
operating limit MCPR (OLMCPR) depends on the average scram time, e, of the control rods, where:
t = 0 or ' ave
'8, whichever is greater SA ~'8 where:
'A = 1.016 sec ($pecification 3.3.C.2.a. scram time limit to notch 3,6)
'8 = u + 1.65 "L 2e (Reference 7]
n
[N 1=1 where: u = 0.822 see (mean scram time used in the transient analysis) v =.C18 sec (standard deviation of u)
IMt gI
' ave =
i=1 n
IN
. i=1,
where: n = number of surveillance tests perfomed to date in the cycle N1 = number of active control rods measured in the ith surve1116Mu test
'i = average scram time to notch 36 of all rods in the ith surveillance test total number of active rods measured in 4.3.C.2.4 N =
g The purpose of the MCPRp, and the Ep of Figures 3.11-3 and 3.11-6, respectively, n to define coerating limits at other than rated core flow and power conditions. At less than 100 percent of rated f1w and powr, the required MCPR is the larger valm of the MCPRp and MCPRg at the existing core flow and power state. The MCPips are established to protect the core from inadvertent core flow increases such that the 91.g-percen. MCPR limit requirement can be assured.
The MCPRps w re calculated such that for the masimum core flow rate and the corrc-pending TM(RMAL POWtt along the 105 percent of rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above th:
$4fety Limit. Using this relative bundle power, the MCPts wre calculated at diff erent points along the 10$ percent of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point ef core flow is defined as MCPip.
The core power dependent MCPR operating limit MCPRg is the power rated flow MCPR operating limit multiplied by the cp factor given in Figure 3.11 6.
The c s are established to protect the core f rom transients other than core flow p
increases, including the localized event such as rod withdrawal error. The Eps were determined based upon the mest limiting transient at the given core power level. (For further information on MCPR operating limits for of f-rated conditient,
reference N(DC 30474-P.(*))
When operating with a single-recirculation pumo, the MCPR Safety and Operating Limits are increased by an amount of 0.01 over the comparatie values f or two-retirculation pump operation.(s)
MATCM - U4171 3.11-aa Amendment No. 157
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9 NERAGE PLANAR LfEAR WAT GEERATION RATE vs c
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NERAGE PLANAR EXPOSURE
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AVERAE PLANAR LINEAR HEAT GEERATION RATE LIMIT vs 2
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FIGLEE 3.11-1 (SEET 5) v.
esOTE-THe8 8E THE ALFtrtA Lae8T FOR THE esOST Lasesiasse LATTICE As A FuesCTeOes OF AvinAGE PLAanMB EMPOSustE.
.o Figure 3.11 1 (Sheet 4) (Deleted)
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l MATCH - UNIT 1 A: xndment No. 157 I
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bendment No. 157 t
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FIGURE 3.114 MCPR LIMIT ALL SX8 AND SX9 FUEL TYPES
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HATCH - UNIT 1 Arenament No. 157 f
FIGURE 3.11-5 (Deleted) f i
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NATCH - UN(f 1 Amendeent No. 157
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