ML20238C957
| ML20238C957 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/21/1987 |
| From: | Jabbour K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20238C960 | List: |
| References | |
| TAC-66472, TAC-66473, NUDOCS 8801040015 | |
| Download: ML20238C957 (28) | |
Text
.
! l N*
o UNITED STATES
~g NUCLEAR REGULATORY COMMISSION n
s,E WASHINGTON, D. C. 20665
- %,.....f GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO.-1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 150 License No. DPR-57 1.
The Nuclear Regulatory Conmission (the Commission) 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 Company, acting for itself, Oglethorpe' Power Corporation, Municipal Electric Authority of Georgia..and City of Dalton, Georgia, (the licensee) dated October 8,.1987, complies with the standards and req)uirements of the Atomic Energy Act'of 1954, as amended (the Act, and the Commission's. rules and regu-lations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the '
orovisions 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 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 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 Comission's regulations and all applicable requirements have been satisfied.
f P
l
1
..- 2.
Accordingly, the license is amended by changes to the Technical Specifi-I cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of f acility Operating License No. DPR-57 is hereby anended to read as follows:
(2)
Technical Specifications l
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.150, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with j
the Technical Specifications.
I 3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
1 FOR THE NUCLEAR REGULATORY COMMISSION
\\)\\
S Kahtan N. Jabbour, Acting Director Project Directorate II-3 Division of Reactor Projects-I/II l
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 21, 1987 4
l l
l[fd P,D #,1
/DRP-1/II PD#II-3/DRP-I/II OC da PD#II-3/DRP-I/II MRob /mac LCrocker KJabbour _ Acting PD 12/ / /87 124 ) /87 12/7/87 12/1l /87 NRR/SRXBp AD PII WHodges Glailnes 12/ } /87 12/l
/87 l
}
ATTACHMENT TO LICENSE AMENDMENT NO. 150 FACILITY OPERATING LICENSE N0. OPR-57 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 l
i X
X 3.5-1 3.5-1 3.5-14 3.5-14 3.5-21 3.5-21 3.11-3 3.11-3 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 4-6)
Figure 3.11-1(Sheet 5 Figure 3.11-1 (Sheet 6) 3.11-6 3.11-6 l
r LI,,ST 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 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. Such as Required by ASME Section XI 3.6-2 Pressure versus Mir.imum Temperature for Non-nuclear Heatup/Cooldown and Low Power Physics Test 3.6-3 Pressure versus Minimum Temperature for Core Critical Operation other than Low Power Physics Test (includes 40'F Margin Required by 10CFR50 Appendix G) 3.6-4 Deleted 3.6-5 Thermal Power Limitations During Operation with Less Than Two Reactor Coolant System Recirculation Loops in Operation 3.11-1 (Sheet 1) Limiting Value for APLHGR (Fuel Types 8P80RB265H.
P80RB265H, BP80R284H, and P80RB284H) 3.11-1 (3heet 2) Limiting Value for APLHGR (Fuel Types BP80RB283, P80RB283, BP80RB299, end P804B299) 3.11-1 (Sheet 3) Limiting Value for APLHGR (Fuel Types BP80R8301L.
P80RB301L, and 1987 Hatch LTAs) 3.11-1 (Sheet 4) Deleted.
3.11-1 (Sheet 5) Deleted.
3.11-1 (Sheet 6) Deleted.
3.11-1 (Sheet 7) MAPFACp (Power Dependent Adjustment Factors to MAPLHGRs) 3.11-1 (Sheet B) MAPFACp (Flow Dependent Adjustment Factors to MAPLHCRs) 3.11-2 Limiting Value for LHGR (Fuel Type 7 x 7) 3.11-3 MCPRr (Flow Dependent Adjustment Factors for MCPR$)
3.11-4 MCPR Limit for All 8 x 8 Fuel Types for Rated Power and Rated Flow HATCH - UNIT 1 x
Amendment No. 150 I
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS j
3.5.
CORE AND CONTAINMENT COOLING 4.5.
CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Applicability Applicability The Limiting Conditions for The Surveillance Requirements Operation apply to the apply to the core and containment operational status of the core cooling systems when the corres-and containment cooling systems.
ponding limiting conditions for operation are in ef fect.
Objective Obiective The objective of the Limiting The objective of the Surveillance Conditions for Operation is to Requirements is to verify the assure the operability of the operability of the core and con-core and containment cooling tainment cooling systems under all systems under all conditions conditions for which this cooling for which this cooling capa-capability is an essential response bility is an essential to plant abnormalities.
response to plant abnor-malities.
Specifications Specifications A.
Core Sorav (CS) System A.
Core Sorav (CS) System 1.
Normal System Availability 1.
Normal Operational Tests a.
The CS System shall be operable:
CS system testing shall be performed as follows:
)
(1) Prior to reactor startup f rom a cold condition, or item Frecuency a.
Simulated Once/ Operating (2) When irradiated fuel is in the Automatic Cycle reactor vessel and the reactor Actuation pressure is greater than Test atmospheric pressure, except as stated in Specification 3.5.A.2.
b.
System flow Once/3 mor.ths rate:
Each loop shall deliver at least 4250 l
gpm against a system head corresponding to a reactor vessel pres-sure of at least 113
- psig, c.
Pump Opera-Once/ month bility d.
Motor Once/ month Operated Valve Operability HATCH - UNIT 1 3.5-1 Amendment No. 150
y,.
/
l-
.y BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.5.
CORE AND CONTAINMENT COOLING SYSTEMS A.
Core Spray (CS) System
- 1. Nor=al System Availability Analyses presented in Reference 1 elemonstrated that the core spray system provides adequate cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature to below 2200"F which assures l
that core geometry remains intact and to limit any clad metal-water reaction to less than one percent. Core spray distribution has been shown in tests I
of systems similar in design to HNP-1 to exceed the minimum requirements.
In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel.
The intent of the CS sy D m specifications is to prevent operation above j
atmospheric pressure without all associated equipment being operable.
However, during operation, certain components may be out of service for the specified allowable repair times. The allowable repair times have been selected using engineering judgment based on experiences and supported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstrating operability immediately and by requiring selected testing during the outage period.
When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water to the core. Requiring two operable RHR pumps and one CS pump provides redundancy to ensure makeup water availability.
- 2. Operation with Inoperable Components Should one core spray loop become inoperable: the remaining core spray loop and the RHR system are demonstrated to'be operable to ensure their availability should the need for core cooling arise. These provide extensive margin over the operable equipment needed for adequate core cooling. With due regard for this margin, the allowable repair time j
of 7 days was chosen.
B.
Residual Heat Removal (RHR) System (LPCI and Containme: Cooling Mode)
- 1. Normal System Availability The RER system LPCI mode is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system is completely independent of the core spray system; however, it does function in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI mode of the RHR system and the core spray system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high-pressure emergency core cooling systems.
HATCH - UNIT 1 3.5-14 Ame dment No. 150 n
....a-....--
- _c
r BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS h
3.5.J/4.5.J Plant Service Water System The Plant Service Water (PSW) system consists of two subsystems (divisions) of two pumps each and a separate Sttndby service water pump system for diesel generator 18.
During I
norwel full power operation the two subsystems function as a 3 out of 4 pump cross connected system supplying cooling water to the turbine and reactor building cooling systems. In the event of an accident signal, non safety-related cooling loads are iso-lated and the PSW pumps in the two subsystems supply cooling water to diesel generators l A and IC, the reactor building cooling system and the control room air conditioners, while the standby service water pump is available to automatically supply cooling water to diesel generator 18 should it be needed. Additionally, diesel IB has a manual back-up water supply available from tne Unit 1 Division 1 or Division 2 PSW subsystems so that during maintenance on the standby diesel service water pump, either division of the PSW system can manually be aligned to supply cooling water to the 1B diesel. The two subsystems and the standby service water pump system are split in the accident mode for greater reliability with one pump in each of the two subsystems automatically starting while a start signal from diesel generator 18 initiates standby service water pump operation. Only one of the Division 1 PSW pumps and one of the Division 2 PSW pumps are required for cooling diesel generators lA and IC, respectively, while the standby ser-i vice water pump provides adequate cooling water to diesel generator 18.
In the event that the standby service water pump is inoperable, the HNP-1 Division 1-Division 2 intertie supply piping can be aligned to cool the IB diesel. In this condition, one PSW pump is capable of supplying the cooling requirements for the reactor buildi,ng cooling system, the control room air conditioners, and the IA, 18 and IC diesel generators.
l l
The PSW system can supply all power generation systems at full load and the diesel l
generators with redundancy if one PSW pump and/or the standby service water pump are inoperable. Hence, a 60-day outage time is justified if the standby service water pump is inoperable since all four PSW pumps are available (divisional intertie to 1B diesel required). In addition, a 30-day outage is justified if one PSW pump is inoperable, or if one PSW pump and the standby service water pump are inoperable (divisional intertie to 18 diesel required). Should two PSW pumps (or one subsystem) become inoperable, or should two PSW pumps (or one subsystem) and the standby service water pump become inoperable (division intertie to IB diesel required) plant operation will probably only continue at less than full power. However, safety-related loads are still adequately powered for these conditions. Therafore, a 7 day outage time is justified for such events.
I K.
Enqineerino Safety Features Ecu1oment Area Coolers
~
The equipment area cooler in each pump compartment is capable of providing adequate ventilation flow and cooling. Engineering analyses indicate that the temperature rise in safeguard compartments without adequate ventilation flow or cooling is such that continued operation of the safeguard equipment or associated auxiliary equip-ment cannot be assured.
The surveillance and testing of the equipment area coolers in each of their various modes is accomplished during the testing of the equipment served by these coolers.
The testing is adequate to assure the operability of the equipment area coolers.
L.
References 1.
'Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NE0C-31376-P, December 1986.
i 1
3.5-21 Junendment No. 150 l
HATCH -
=
- _ _ - _ ~
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS 3.11.
FUEL RODS A.
Averace 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 10 CFR 50.46 even considering the postulated l
effects of fuel pellet densification.
The peak cladding temperature following a postolated 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 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 APLHGR at rated conditions is shown in Figures 3.11-1, sheets 1 thru 6.
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.
The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.
The Loss-of-Coolant Accident (LOCA) analysis was performed utilizing the new improved calculational model, SAFER /GESTR-LOCA. The analysis demonstrated that loss-of-coolant concerns do not limit the operation of the fuel since margin to the 2200*F limit was demonstrated (Reference 9).
Therefore, the APLHGR limits for the fuel types shown in Figure 3.11-1 are derived to assure that the fuel thermal-mechanical design criteria are net.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Tables 4-1 and 4-2 of Reference 9.
Further discussion of the APLHGR bases is found in NEDC-30474-P(*).
A flow dependent correction f actor incorporated into Figure 3.11-1 (sheet 8) is 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 and flow dependent corrections given in Figure 3.11-1
]
(sheets 7 and 8) are applied to the rated conditions APLHGR limits to assure that the fuel thennal-mechanical design criteria are met during abnornal transients initiated from off-rated conditions for two-loop and single-loop operations, References 2 and 8.
For single-loop operation, a 0.75 multiplica-tion factor to APLHGR limits for all fuel bundle types conservatively bounds that required by Reference 2.
For single-loop operation (SLO), the most restrictive of the SLO and ARTS (e) MAPLHGRS will define the Limiting Conditon for Operation.
1 HATCH - UNIT 1 3.11-3 Amendment No. 150
AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMIT vs l
AVERAGE PLANAR EXPOSURE f ll ll l
/
/'
h* /jl//
[j (p/// /3,s' /'
4fNACC ' TABLE OPERATI'ON '/
) IE* 0 /.
13 9 '
/// //l //[
//
6 ss.Y
'f,/j//'/
~//
' /,
,','/
i
- E0 t; 11.0
//
c00n 2
y y
, 7f7 3
- 0. 0
- 11. 5 ACCEPTABLE OPERATION
'f
/,/
is
- 0. 2
- 11. 5 10.0
/
/
/
i c, u u,a
,,/
- 4. 0
- 11. 9 FLEL TYPE:BP80R826SH AND P90RB265H k'//
g go Ig 1 80-MIL CHANNELS
/y 16.'6 IE5 12.5 12.5 se 0
- 5. 0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 AVERAGE PLANAR EXPOSURE (GWd/st) f,
/
n ' 3* O /
./
c'
,/~
f s' s' s'77 *7~/l, '" ~Y'?
s'//'
////
/
i' i
~
/
s
/
j ll,!ll,' ff,, '/
^
' UNACCEPTABLE
)
p/
S3.4 $5,4 83 4
OPERATION /
7 If,/
2
/'
_/,','3 * [,/ j 5
A. =
Q
- 11. 0; gg 7,
31 NATES AIEPTAB E (PERATION sa l'" il li M
na Tree, sea s284 u ND
- 4. 0 11.5
_ _ _ P83R8284H g g,
/
Eu 31.3 iIU-mil L.N/UTR.Lb 1
L5 12.5 12.4
- 8. 0
- 0. 0
- 5. 0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 AVERAGE PLANAR EXPOSURE (GWd/st)
FIGl5tE 3.11-1 (SEET 1) i Amendment No. 150
AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMIT vs AVERAGE PLANAR EXPOSURE
/ /
/
/
f/
NACCb) TABLE OPE TION /,/
A 4,-
N f,
a
,//// /// // / / / /
n_
"na.s si.s
- XPOSglE ACCEPTABLE OPERATION
/
'f f
ko
!!,'l FIEL TYPE:OP90RB2il3 AND F'80RB281
'.//
k
- 6. 0 12.0 90-MIL OR 100-HIL CHANNELS Vf/
0 B. O 12.3 X/
10.0 12.3 a.h 12.5 12.4
- 0. 0
- 5. 0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 AVERAGE PLANAR EXPOSURE (GWd/st)
,l,
/
9 //
//
//
///
/ '/ (
//'
^
/'/
g
/
/
t 2 -- / 17 4,
ACCEPTABLE OPERATION'
. /
5
.0 12
-/
k" '
,f f 11.0 i rw rrpmist
//
./
7 7 7 s
y ACCEPTABLE OPERATION 3"
,/
0.0 11L7
/ // /
$ 0
.0 11EL TYPE: BP8DPB299 AFD P80RE299
'//
d 4,0 1 1.3 80-MIL CHA E.S
- 9. 0 -
{ }
j 10.0 I L2 to 12,5 I L4
- 0. 0
- 5. 0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 AVERAGE PLANAR EXPOSURE (GWd/st)
FIGL5E 3. !!-! (SEET 2)
Amendment No. 150
AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMIT vs AVERAGE PLANAR EXPOSURE
/
/
p h
/,/
//h) / ', / ' NACC PTAB[E P RAT 10N'
,I/,/j 3
nt s "
A '/./,,/l ,,' l
> 12.0
', I
/
P 5
,u g
?
DW IXPOSLRE
/
"I"
- 3. !!. 0 X
Y 10h
'/.
J
- 0. 0 11.3 ACCEPTABLE OPERATION
% 10.0
- 0. 2
- 11. 3
// /
$'O ll,'h FLEL TYPE:BP80R830: L AND F 80RB301 L
/
O
- 6. 0 ll.g 80-MIL CHMNELS
.'/
< 9. 0 -
go go 10.0 12.2 y
12.5 12.3
- a. 7
- 8. 0
- 0. 0
- 5. 0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 41 0 AVERAGE PLANAR EXPOSURE (GWd/st) 14.0 13.0 b12.0i
/
/
&lj ',/,/,!,l, I } '/,/
ss.e
/j j//j[j g
ACCEPTABLE OPERATION
/'
/j/
/
j, b
l/
/
- s. 0 -
FLEL TYPE:1987 HA ICH 1 LTAs 7-
- 6. 0 a.
t
- 0. 0
- 5. 0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 i
AVERAGE PLANAR EXPOSURE (GWd/st)
FIGLRE 3.11-1 (SHEET 3) i HATCH - UNIT 1 Amendment No. 150 l
i
- i e.
.l Figure 3.11-1 (Sheet 4)- (Deleted)
Figure 3.11-1-(Sheet 5) (Deleted)
Figure 3.11-1 (Sheet 6) (Deleted) i 1
j
-i i
HATCH - UNIT 1 Amendment No. 150
i BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS 3.11.E.
References 1.
" General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-240ll-P-A.
2.
"Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NEDO-24205, August 1979.
l 3.
" Loss-of-Coolant Analysis for Edwin I. Hatch Nuclear Plant Unit 1,"
NE00-24086, December 1977.
4
" Fuel Densification Ef fects on General Electric Boiling Water Reactor Fuel', Supplements 6, 7, and 8, NEDM-10735, August, 1973.
5.
Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 16, 1974 (USA Regulatory Staff).
6.
Communication:
V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27,1974.
7.
Letter from R. H. Buchholz (G. E.) to P. S. Check (NRC), " Response to NRC request for information on ODYN computer model", September 5,1980.
8.
" Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2,"
NEDC-30474-P, December 1983.
9.
'Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coola Accident Analysis," NEDC-31376-P, December 1986.
~
l I
j HATCH - UNIT 1 3.11-6 Amendment No. 150 l
l
b UNITED STATES
~,,
NUCLEAR REGULATORY COMMISSION o
5 "E
W ASHINGTOE, D. C. 20655
%...../
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION 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 Amendment No. 87 License No. NPF-5 1.
The Nuclear Regulatory Comission (the Comissioni 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 8, 1987, complies with the standards and req)uirements 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; 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 (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 set forth in 10 CFR Chapter I; 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, t
-.-_m.____
i l
l l l
1 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and i
paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 87, 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 cf issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
\\V Kahtan N. Jabbour, Acting Director Project Directorate II-3 Division of Reactor Projects-I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: December 21, 1987 l
l
[
PD#1 N /DRP-I/II PD#11-3/DRP-I/II O
e9f da PD#11-3/DRP-I/II MR66Bh6ac LCrocker V
KJabbour, Acting PD 12// /87 12/ l /87 12/f /87 12/))/87 40 NRR/SRXB AD/ PII WHodges Gla as 12/j/87 12/
/87
ATTACHMENT TO LICENSE AMENDMENT N0.87 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 areas of change.
Corresponding overleaf l
pages are provided to maintain document completeness..
Remove Insert Page Page 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 i
3/4 2-4d 3/4 2-4d 3/4 2-4e 3/4 2-4e - 3/4 2-41 3/4 2-4f 3/4 2-49 3/4 2-4h 3/4 2-41 3/4 3-30 3/4 3-30 3/4 5-5 3/4 5-5 B3/4 2-1 B3/4 2-1 B3/4 2-2 B3/4 2-2 B3/4 2-6 B3/4 2-6
\\
\\
0
}
,/
/ ~/./^/
5
/
/ /
/-
4 4
~/
r o
L / s,,/'/'/'/ - / /,f '
/,,
N
,O' 0
s I
T/b
//,.
0 v
4 A
/,
T R
/,P'I? //
/,/
E I
H O
f 5
0 M
I 1
L
/
/ / /,' e,
,L((
C2 5
E 4
/
B t
3
)
E
/
Al/
0 R
s B
l;,T /.A T
A
!,,,/,
1 0
/
A 8
0 d R
3 W
fr F' E
l S
0 G NE C
D OR C
N N L 3
(
N(('
U E
I TS
/
A O
A E
N I
R AO
,J EX l
' l T
H N 0 U RP A
A 5
'\\8 S
H R
6 5 O NE E
E 2 C 2 P 1
GR P
B X
1 1
A O
R L 1
I E
2 TN 0
AA E
8-M 0
3 R
EL
/,,5 L
F -
B B 0 0 A E
8 2 N R
HP U
2 A
RE T
E A
G L
I AG P
P P
F EA O
E Y
0 C
f NR E
J3 2
C 5 G I E 5
A L
LV A l,,L.
1 A
1 E
R 2
U E
R 2
F V
A 71 N
0 A
/_'
! /-,2 A
E 45 0
s.58 91..5 -
S 4
R L
E Y
r U
1 P
82222 T
11 1 S
A 11 111 1 1 1 V/1/
O N E
P I
X G
Eo-0 R
A w0 -
R o oX 0200oo05
'j/
5 C-E L
0o246s02....... -
V I,,,/ p /.
11 A
//'
0 0
0 0
0 0
00 3
2' 0
9 8
1 1
1 1
1 v bE 1
gL3n<
3 Ed. iZ i
!is g g I Ra mA, I
1
0 y/!} / / 7 / ' '/ '-
5 s
4 hO
/
/
8
/
7 /.
0
~
s I
,T
/
0 7E/l/,l/ /
v 4
,A,R
/
T
/ /[
I
.,Pf /
,f 0
M I
.f o,
L 5
N E
,L
/
O 3
)
i t
E
,Bf,
/
A s
T A
/
A T
H 0 d R
W F'
t H
,Cl
,E 84 0 G NE
,C N
28 3
(
OR 7Nll O
B2 U
I RB E
,A TS I
R AO T
OR 0 U U
A B0 RP S
EX R
P8 5 O 2
NE E
BP 2 P ER P
1 O
E X
A P
E 2
4 E
Y 0
3 TN AA L
T R
EL B
0 A E
HP A
L 2 N R
1 L
T E
A C
L I
RE P
U P
F E
F A
4 EA C
0 E
NR C
/, 4 5
x A
IE 1
A LV R
A 5
,$ 2j3 s
3 E
R eS y$i V
A sE j22 0
A N
nT lt j1 1
/
pW 0
e A
P 8
L I
x 1
eD 1
R w0
$0@o0 1
A f), {/
E o0 x$2[e 0 t C-1 0
E
/
,f?
R 5
A
/ j '/, '-
V
/
0 r
~
0 0
0 0
00 3
1 0
9 8
1 1
1
?}" tx jo$%
5N. c6 ea g
[8EE E $
j
,o ll)1l i
=
~
0 I
5
, ~,
5 4
/
- - /
/O.
/
- /
y 8
j. /, ' -
J P
/
0 s
I 4
/ I,T/,, -
'/,
!/R'.v
, k%
0 v'
i 4
/
A T
/
E f.
/ P, ~ 'y -
1 e
0
'r J
I 3
L
/,
/
/ / rl
/
8 5
E 2
2 3
)
/
. //L, 2
E t
BS B,
0 T
RL s
d,A
'fr// n
'."1 T '
- 3. - 0E
/
A
?
9N 0 d I
P N' W
r; t
t NE
/
' fC, f E
A 0 G OR N
J H i
3
(
/C, O'
- N E
I U C
'/b' /N lj,h A
TS
/
I AO L
R T
.E I 0 U s
U.
A RP EX S
/
'fl R
c gF E
2-5 NE O
3-2 P E
P B0
/
/. &
GR O
R0 X1 s
=
A E
01 2
TN I,/
E 8 L' 0
/
fl 2
R3 H P l'l/
/
s.
i EL B
8 D' 0 A E 4
2 N R I /2 A
T EL_
A l AG l
'/A G
RE P
E YM PF
/
C T-
- 0 s EA j
, ' ' :/
NR E
h
/
5 1
C 0
5 G IE
~
L8_<
A S
LV 1
A E
/
/
A R
U 2
R
/
/ '.y'1 F
E t
- 0 V A
~
N A
i, j/l<1 A
g 0
6 0 0 37 j3 4 L
1 P
11; $22 e
/,
111il11 f ' i l /3' E
2 '-
G A
l
- 0 l 1' Y
//
/
/
'//
R 20a @0s m5 E
J 2a[02
/,/ /'. l
- 0 V
J
~
n i
11 A
0 0
0 0
0 00 s.
=
3 2
1 0
9 8
J 1
1 1
1 e.$
J n:$,_%
h a
4
=
x D cx
- c 6 ~
z m
3g gs" 2 5
4,,
2 Il
({
ll 1
o
0 f/'.;/,#/,V 5
,/
4
/
e
/,-
//
0 s
/ N Y' / 6/
9 0
v O
/,///,
9 4
T
/I 2
T 8
M
/A l I
/
R R
0 O
I L
/E 4g E
P 5
P E
/O 3
)
t
/7-D S T
s
/E N
N L A
/
L E
0 A
R A
~ N 0 d
/
3 W
1 N
NE T
/~
T 9
0 G OR
/
A A
9 3
(
P R
2 H IU TS
/E C
E B
E C
~
/ C P
R R
O O L 0 U RP v.
A I
H EX NE
/
/
S E
P M 5 O 4 L
B-2 P ER B
0 s
1
- 8 X
A E
E 2 A
/
T P
0 3
TN AA P
Y R
/
/4 EP E
1 0 A L
E C
2 N C
L A G
~
RE A
L
/
E P F I
A EA F
0 NR
/
E
/4.
w, s
E 7
- 81 1 24 5
IE g
LV 1
A
! 11 22 R
A
/
2
! 11 11 R
1 A
/ /
V 0
mg N
A
/
/ j2 A
0 0 0. cs.
0 L
'y 2 68 s2 1
P
/ f2 i1 E
/fl/
0 A
R
, /7O7
.I
/
/
5 EV
',/
/
A 0
0 0
00 3
9 8
1
$} e5 h$
b9.gZ
=
gg$n E $
0 e4
/
/ g 5
4
.e
/
V pL b
s m7 0
v 4
1
%y f
0 T
3 I
M W
0 I
//
/
0 5
L 8
3
)
E P
t T
s S
a%y D
/
A N L 0 d R
AE W
h 0
G.
NE N
OR L
3
(
N A
a,4 U
I 1
H E
O 0
TS 3 C R
I AO T
8 0 U RP h
L EX S 5 N0 E2 5 O -
R I
NE M
2 P1 E
E GR P
~
X A
O w
8 TN b
3 E
E R
L P
0 A E B
Y 2 NR A
T A U n "-
T LG RE P
L P
I AG T
E 0
F EA NR E
y" l
C F
5 G I E A
1 A
LV A
E R
R Up R
3$ e o, E
S a, 1'[0gla Y
A V
b N
7,d,u>cLu X A
1
'i A
Pn 0
L X
P E
1 W
O "ij:g u-0$
8 a
O E
G 0
A
- yA7, R
5 E
V A
/
d o
0 0
0 2
1 0
9 1
1
? ; l28 D' $ %
59,gM
[g
$ i $
f
i l
1 Figure 3.2.1-6 (Deleted)
Figure 3.2.1-7 (Deleted)
Figure 3.2.1-8 (Deleted)
Figure 3.2.1-9 (Deleted)
Figure 3.2.1-10 (Deleted)
Figure 3.2.1-11 (Deleted) i HATCH - UNIT 2 3/4 2-4d Amendment No. 87
4
.i
- j 1
l 1
l
'l l-l l
b (THESE PAGES ARE INTENTIONALLY LEFT BLANK.)
l l
l I
l l
\\
1 I
l 4
HATCH - UNIT 2 3/4'2-4e-- 3/4 2-41 Amendment No. 87 i
~.
i
4 TABLE'3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds) 1.
CORE SPRAY SYSTEM s 34 l
2.
LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM s 64 l
3.
HIGH PRESSURE COOLANT INJECTION SYSTEM s 30 4.
AUTOMATIC DEPRESSURIZATION SYSTEM NA 5.
)
HATCH - UNIT 2 3/4 3-30 Amendment No. 87 i
EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Centinued) b.
In CONDITION 4 or 5*;
1.
With one CSS subsystem inoperable, operation may continue i
provided that at least one LPCI subsystem. is OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, suspend all operations that have a potential for draining the reactor vessel.
2.
With both CSS subsystems inoperable, operation may continue provided that atsleast one LPCI subsystem is OPERABLE and both LPCI subsystems are OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise, suspend all operations that have a potential for draining the reactor vessel and verify that at least one LPCI subsystem is OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
~
3.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.5.3.1 Each CSS subsystem shall be demonstrated OPERABLE:
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the condensate storage a.
tank minimum required volume when the condensate storage tank is required to be OPERABLE.
.t b.
At least once per 31 days by:
1.
Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water, and 2.
Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct
{
position.
At least once per 92 days by:
c.
1.
Verifying that each CSS pump can be started from the
)
control room and develops a flow of at least 4250 gpm on recirculation flow against a system head corresponding to l
a reactor vessel pressure of ;t 113 psig, and j
i HATCH - UNIT 2 3/4 5-5 Amendment No. 87
- - - - - - - - - - - - - - - ~
3/4.2 P0tfER DTSTRIBUTTON LIMITS BASES temperature following the postulated design basis lo will not exceed the 2200*F limit specified in the Final Acceptance C oolant accident (FAC) issued in June 1971 considering the postulated effects of f r teria densification.
maintained during abnormal transients.These specifications also assure uel pellet e
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE the postulated design basis loss of coolant accident w o owing specified in 10 CFR 50.46.
not exceed the limit l
accident is primarily a function of the average he of-coolant the rods of a fuel assembly at any axial location and is d e of all secondarily on the rod-to rod power distribution within an assembly ependent only clad temperature is calculated assuming an LHGR for the highes The peak which is equal to the design LHGR for that fuel type.
rod l
figures in Technical Specification 3/4.2.1 is based on n the accident analysis.
The analysis was performed using General Electric (G s of-coolant calculational models which are consistent with the requiremen to 10 CFR 50.
The Loss of-Coolant Accident (LOCA) analysis was perform pendix K utilizing the new improved calculational model, SAFER /GESTR-LOCA of the fuel since margin to the 2200*F limit was de The fuel types (Reference 4).
e operation shown in the figures for Technical Specification 3/4.2.1 are all of these that the fuel thermal-mechanical design criteria are met o assure A flow dependent correction factor incorporated into Figure 3 21 applied to the rated conditions APLHGR to assure that the 2200
.. -12 is complied with during a LOCA initiated from less than rated core fl addition, other power and flow dependent corrections given in Fig m t is ow.
In and 3.2.1-13 are applied to the rated conditions' to assure that the f res 3.2.1-12 thermal-mechanical design criteria are preserved during ab uel initiated from off rated conditions.
normal transients requires a 0.75 multiplication factor for all fuel bundlesFor single-loop operat{
3 accident analysis is presented in bases Table B 3.2.1-1A s of coolant of the ApLHGR limits is given in Reference 2.
Further discussion HATCH - UNIT 2 B 3/4 2-1 Amendment No. 87
4 Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS l
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 106 lbm/h which corresponds to 105% of rated steam flow Vessel Steam Oome Pressure.............
1055 psia Oesign Basis Recirculation Line Break Area For:
l
)
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 l
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 and its source is presented in subsection 6.3.3 of the FSAR.
l "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 B 3/4 2-2 Amendment No. 87
POWER DISTRIBUTION LIMITS BASES
References:
1.
" General Electric Standard Application for Reactor Fuel (Supplement for. United States)," NEDE-24011-P-A.
2.
" Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2," NEDC-30474-P, December 1983.
3.
"Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NED0-24205, August 1979.
4.
"Edwin I. Hatch Nuclear Power Plant, SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31376-P, December 1986.
l 1
j t
HATCH - UNIT 2 B 3/4.2-6 Amendment No. 87 i
l
)
_ ___ _ _ _ _ _ _ i