ML19350A551
| ML19350A551 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/06/1981 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19350A548 | List: |
| References | |
| TAC-43589, NUDOCS 8103160466 | |
| Download: ML19350A551 (7) | |
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4 ATTACHMENT 2 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS The proposed change to Technical Specifications (Appendix A to Operating License DPR-57 would be incorporated as follows:
Remove Page Insert Page 3.11.2 3.11.2 3.11.2a 3.11.2a Figure 3.11.4 Figure 3.11.4 Figure 3.11.5 Figure 3.11.5 Figure 3.11.6 Figure 3.11.6 3.11.6 3.11.6 6
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEtiTS 3.11.8 Linear Heat Generation Rate (LHGR)
(Continued)
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 re-quired.
C. Minimum Critical Power Ratio (MCPR) 4.11.C.1. Minimum Critical Power Ratio (MCPR)-
The minimum crit'ical power ratio
'MCPRshallbedeterminedtobb (MCPR) as a function of scram time equal to or greater than the and core flow, shall be equal to or applicable limit, daily during greater than shown in Figures 3'.11.4, reactor power operation at > 25%
3.11.5, or 3.11.6 multiplied by the rated thermal power and.folTow!ng K shown in Figure 3.11.3, where:
any change in power. level or dis-f tribution that would cause opera-
, ave.
tion with a limiting control rod T = 0 or [
'B], whichever is Pattern as described in the bases TATB greater for Specification 3.3.F.
T =0.90 sec (Specifications 3.3.C.2.a 4.11.C.2 Minimum Critical Power' Ratio Limit A scram time limit to 20% inse tion from fully w % drawn)
The MCPR limit at rated flow shall
,N be determined for each fuel type, 1
8X8R, P8X3R, 7X7 from figures r =0.710+1.65[n
](0.053)[Ref.101 3.11.4, 3.11.5, and 3.11.6 re-B I Nj spectively using:
i=1
- a. t=1.0 prior to initial scram i
n k=1"i i time measurements for the m
t cycle, performed in accordance t
=
l ave n
with specifications 4.3.C.2.a r
N.
or 3
i=1
- b. T as defined in specif,ication n = number of surveillance tests performed to date in cycle ne hh of & lW N = number or; active control rods must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> g
measured in the ith surveillance of the conclusion of each scram test time surveillance test required l
tj= Average scram time to 20% in-by specification 4.3.C.2.
sertion from fully withdrawn of all rods measured in the ith surveillance test, and, 1
N = total number of active rods y
l measured in 4.3.C.2.a.
If at any time during operaticn it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be
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( LIMITIfdG C0fiDIT10!45 FOR OPERAT10t4 SURVEILLAfiCE REQU$RETdiTS
.-5
- N;kl.C ;' Minimum Critical Pow:r Rat'io (MCPR) 21.
(Continued)
L; wi t;
. initiated within 15 minutes to era.
restore operation to within the Eli.
prescribed limits. If the steady is state MCPR is not returned to within 35c.
the prescribed limits within two (2) -
hours, then reduce reactor power to r:
5%
less than 25% of rated thermal power 1c, within the next four (4) hours, If.
the Limiting Condition for Operation p-is restored prior to expiration of 30 the specified time interval, then Jrs further progression to-less than 1
25% of rated thermal power is not
' ~,."
required.
- g; D. Reportino Requirements u gel?'
he If any of the limiting values iden-
~
if pe:
tified in Specifications 3.11. A.,
" ~ '
tre B., or C. are exceeded, a Reportable Z,'
re..
Occurrence report shall be submitted.
" - ~
>ec If the corrective action is taken, 27
!d, as described, a thirty-day written m.
report will meet the requirements
'~ :I; ci of this specification.
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' 3.1'l.E. Re ferences 3
..; ), Cencral Electric Cocpany Analytical Model fcr loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NE."I-20566-?, Novceber,1975.
"; 2. Cencral Electric Refill Reflood Calculation (Supplement to SAFE Code 5.. Description) c cas=itted to USAEC by letter, G. L. Cyorcy to V. Scello,
.Jr., dated Decc=ber 20, 1974.
.E 3. Edwin I. Hatch Nuclear Plant Unit 1 E=crgency C m Cooling System
- 9__.. Anal) sis - APc. endix K Rec..tirement.With Modified a Pressure Coolant 2.-
Injection Systes, NEDO-21187, Supple =ent 1, April, 1976.
. ?
M 4. "Fuc1 Densification Effects on General Electric 3ci31n:; Water Reactor Tucl", Supple =ents 6, 7, and 8, NEDM-10735, August, 1973.
'=-
. 2 5. Supplement.1 to Technical Report on Densification of General Electric 1
Reactor Fuels, Dece=ber 16, 1974 (USA Regulatorf Staff).
~+-
=
i, 6. Coc=cnication:
- 7. A. Muore to I. S. Mitchell, " Modified GE Medel:for j-
]~
Fuel Densificatica", Docket 50-321,. March 27,1974.
=-:
E. 7. "Edwin I. Estch Nuclear Plant Unit 1 Channel Ir.spection and Safety j
g Analysis with 3ypass Floa coles Flug;;ed", NEDO-21124-1, July,1976.
~
~-
F 8. R. 3. Li= ford,. A=alytical Methods of Plant Transient Evaluations for i_.
+ CE 3WI., To-y, 1973 C:EEO-10S02).
T
=
- 9. cencral Electric 3oille; Water Reacte-Reload No. 1 Licensing A=end=cnt
=
IE for the Edwin I. Estch Nuclear Plant Unit 1 Full Core Drilled Cc:ditions,
@~
NEDO-21580, February, 1977.
=.
.E
_- 10. Letter from R. H. Buchholz (G;E.) to P.S. Check (NRC),". Response to E
g NRC request for information on ODY.N compdter 'model," Septemoer 5,1980.
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