ML20136G566
| ML20136G566 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/14/1997 |
| From: | Kalman G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20136G571 | List: |
| References | |
| NUDOCS 9703180097 | |
| Download: ML20136G566 (9) | |
Text
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4 UNITED STATES i
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,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001 1
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l ENTERGY OPERATIONS INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 188 License No. DPR-51 1.
The Nuclear Regulatory Comission (the Comission) has found that-A.
The application for amendment by Entergy Operations, Inc. (the 1
licensee) dated November 26, 1996 as supplemented on December 17, i
1996, March 4, 1997 and March 10, 1997, 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 regelations 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 Comission's regulations; D.
The issuance of this license 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 appl.icable requirements have been satisfied.
9703180097 970314 PDR ADOCK 05000313 P
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2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-51 is hereby amended to read as follows:
'f 2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.188, are hereby incorporated in the license, i
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION N
i George Kalman, Senior Project Manager 1
Project Directorate IV-1 i
Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications l
Date of Issuance:
March 14, 1997 5
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ATTACHMENT TO LICENSE AMENDMENT NO.188 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 l
Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES l
18a 18a 19 19 20 20 20a 20a 20b 20b 20c 20c l
l
.-.-- _ ~ -. - ~. - -. -. -.. - -...
3.1.2.7 Prior to roaching thirty cn3 offcctiv3 full pcwer y3ers of l
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cparatien, Figurss 3.1.2-1, 3.1.2-2 cnd 3.1.2-3 shs11 b3 updated for the next service period in accordance with 10CFR50, Appendix G.
The service. period shall be of sufficient duration l
to permit the scheduled evaluation of a portion of the surveillance data scheduled in accordance with the latest revision of Topical Report BAN-1543(5). The highest predicted adjusted reference temperature of all the beltline region l
unterials shall be used to determine the adjusted' reference temperature at the.end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8.
The provisions of Specification 3.0.3 are not applicable.
I 3.1.2.8 The updated proposed technical specifications referred to in 3.1.2.7 shall be submitted for NRC review at least 90 days prior to the end of the service period.
l 3.1.2.9 With the exception of ASME Section XI testing and when the core flood tank is depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit breakers for the motor operators opened before depressurizing the reactor coolant system below 600 psig.
3.1.2.10 With.the exception of ASME Section XI testing, fill and vent of the reactor coolant system, emergency RCS makeup and to allow I
maintenance of the valves, when the reactor coolant temperature is less than 262*F, the High Pressure Injection motor operated l
valves shall be closed with their opening control circuits for the motor operators disabled.
3.1.2.11 The plant shall not be operated in a water solid condition when the RCS pressure boundary is intact except as allowed by Emergency operating Procedures and during System Hydrotest.
Amendment No. 64,44,96,444,444, 18a ARKANSAS - UNIT 1 444, ML, 188
. BASES Al1~ reactor. coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. (*)
These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR. The maximum unit heatup and cooldown rates satisfy stress limits for cyclic operation. (2)
The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100*F satisfies stress levels for temperatures below the DTT. (3)
The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10CFR50.
Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in FTI Document 77-1258569-01 (4). The limiting weld l
material is being irradiated as part of the B&W owners Group Integrated Reactor Vessel Material Surveillance Program and the identification and locations of the capsules containing the limiting weld material is discussed in the latest revision to B&W report, BAW-1543. (5)
The chemical composition of the limiting weld material is reported in the B&W Report, RAN-2121P (6).
The effect of neutron irradiation on the RTNDT of the limiting weld material is reported in FTI calculations 32-1245917-00 and 32-1257716-00 (1).
Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 present the pressure-temperature limit curves for hydrostatic test, normal heatup, and nornal cooldown respectively.
1 The limit curves are applicable through the thirty first effective full power year of operation. The service period was reduced by one effective full power year from that assumed in FTI Document 77-1258569-01 to be conservative with respect to independent calculations performed by the NRC staff. The pressure limit is also adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all allowed operating reactor l coolant pump combinations, i
The pressure-temperature limit lines shown on Figure 3.1.2-2 for reactor criticality and on Figure 3.1.2-1 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing.
The actual shift in RTNDT of the beltline region material will be established periodically during operation by removing and evaluating, in accordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region.
j The spray temperature difference restriction based on a stress analysis of the l
spray line nozzle is-imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
Temperature requirements for the steam generator correspond with the measured NDTT for the shell.
I Amendment No. M,M,M,M,M4,188 19
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Th3 h00 tup cnd socid:wn ratos stated in thic cpscificaticn arc intcnd:d cs the maximum chtng:s in temparcturo in cno directicn in o cna hsur paried. Tho cetus 1 temp 3rsturo linter ramp rato may cxesed the stated limits for a time period provided that the maximum total temperature difference does not exceed the limit and that a temperature hold is observed to prevent the total j
temperature difference from exceeding the limit for the one hour period.
specification 3.1.2.9 is to ensure that the core flood tanks are not the source d
for pressurizing the reactor coolant system when in cold shutdown.
specification 3.1.2.10 is to ensure that high pressure injection is not the source of pressurizing the reactor coolant system when in cold shutdown. The LTOP enable temperature has been calculated in accordance with code case N-514.
Instrument error is not included in the reactor coolant temperature of 262*F.
Specification 3.1.2.11 is to ensure that the reactor coolant system is not operated in a manner which would allow overpressurization due to a temperature transient.
REf7AENCES (1)
FSAR, section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3) FSAR, section 4.3.11.5 (4)
FTI Document Number 77-1258569-01 l-(5) BAW-1543, latest revision (6)
BAW-2121P l
(7)
FTI Calculation Numbers 32-1245917-00 and 32-1257716-00 l
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O Amendment No. 4,43,44,M,M,M4,188 20
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FIGURE 3.1.2-1 d
R OS INSERVICE HYDROSTATIC TEST H/U & C/D LIMITS TO 31 EFPY 2400 I
2200 I
j 4
4 2000
@1800 w
6 1600 1
g i
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ew 1400 A
H o 1200 ACCEPTABLE PEGION
< 1000 g
O 800
{
b i
g 600
-L-l 400 200 s
J j
0 1
0 50 100 150 200 250 300 350 400 450 500 550 600 RCS COID LEG TEMPERATURE (*F) l Notes:
1.
This curve is not adjusted for instrument error and shall not be used i
for operation.
2.
All Notee on Figure 3.1.2-2 are applicable for heatups. This curve is lased on a heatup rate of < 90*F/HR.
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3.
All Notes on Figure 3.1.2-3 are applicable for cooldowns.
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Amendment No. 44,84,M4,188 20a I
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e FIGURE 3.1.2-2 RCS HEATUP LIMITATIONS TO 31 EFPY 2400 2200
- J 2000
< 50'F/HR Non-Critical
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2 1800 Curve l
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< 90*F/HR I:
Critical l
1600 Curve d7i l J
j 1400 v
b O
l E 1200 7
H 4
g1000 i
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800 ACCEPTABLE I
s
< *10*F/HR
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REGION Non-Critical y j
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Curve w
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< 90 F/HR 200 Non-Critical j
Curve 0
0 50 100 150 200 250 300 350 400 450 500 550 600 RCS COLD LEG TEMPERATURE ('F)
Notes:
1.
These curves are not adjusted for instrument error and shall not be used for operation.
2.
When DHR is in operation with no RCPs operating, the DHR system return temperature shall be used.
3.
RCP Operating Restrictions:
RCS TEMP RCP RESTRICTIONS T > 300*F None 300*F 2 T 2 225*F
$3 225'F > T 2 84'F s2 T < 84*F No RCPs operating 4.
Allowable Heatup Rates:
RCS TEMP H/U RATE 60*F < T s 84*F 5 15*F/HR T > 84*F As allowed by applicable curve Amendment No. 48,44,M4,188 20b
FIGURE 3.1.2-3 RCS COOLDOWN LIMITS TO 31 EFPY 0
2400 i
l 2200 2000
/
e g1800
/
%1600 e
U1400 f=0 ACCEPTABLE 1200 agazog
$1000 3
800 b
0 600 4*
200 i
0 0
50 100 150 200 250 300 350 400 450 500 550 600 RCS COIE LEG TEMPERATURE (*F) tiotes:
- 1. This curve is not adjusted for instrument error and shall not be used for operation.
- 2. A maximum step temperature change of 25*F is allowable when securing all RCPs with the DHR system in operation. This change is defined as the RCS temperature prior to securing all the RCPs minus the DHR return temperature after the RCPs are secured. When DHR is in operation with no RCPs operating, the DHR system return temperature shall be used.
- 3. RCP Operating Restrictions:
RCS TEMP RCP RESTRICTIONS T > 255*F None 150*F s T s 255'F s2 (See Note 5)
T < 150*F No RCPs operating
- 4. Allowable Cooldown Rates:
RCS TEMP C/D RATE STEP CFANGE T 2 280*F 100*F/HR s 50*F in any 1/2 HR 280*F > T 2 150*F 50*F/HR (See Note 5) s 25*F in any 1/2 HR T < 150*F 25*F/HR s 25*F in any 1 HR
- 5. If RCPs are operated < 200*F, then the RCS cooldown rate from 150*F 5 T s 180*F is reduced to 30*F in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
Amendment No. 48,43,M4,188 20c