ML17054A681
| ML17054A681 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/18/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Niagara Mohawk Power Corp |
| Shared Package | |
| ML17054A682 | List: |
| References | |
| DPR-63-A-058 NUDOCS 8405150518 | |
| Download: ML17054A681 (30) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
58 License No.
DPR-63 1.
The Nuclear Regulatory Commission (the Commission}
has found that:
A.
The applications for amendment by Niagara Mohawk Power Corporation (the licensee) dated March 22,
- 1978, and supplemented and clarified by letters dated April 20 and October 26,
- 1983, and September 26, 1983 comply with the standards and requirements 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 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 Commission's regulations; 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 F.
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 to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph
- 2. C.(2) of Facility Operating License No.
DPR-63 is hereby amended to read as follows:
8~0Sl50Sl8 SeOelS PDR ADOCK 05000220 j
P
I
(2) Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 58, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This 1-icense amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 18, 1984 1
P/
Domenic B. Vassallo, Chief Operating Reactors Branch ¹2 Division of Licensing
0
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a ATTACHMENT TO LICENSE AMENDMENT NO.
58 FACILITY OPERATING LICENSE NO.
DPR-63 DOCKET NO. 50-220 Revise the Appendix A Technical Specifications by removing and inserting the following pages:
Existing Pa<ac 77 78 79 80 81 275f Revised
~Pa e
77 78 79 79a 80 80a 81 81a 82 82a 82b 257f The revised areas are indicated by marginal lines.
LIMITING CONDITION FOR OPERATION 3.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESS URI ZATION
~AA i 11 1 1 Applies to the minimun vessel temperature required for vessel pressurization.
Objective:
To assure that no substantial pressure is imposed on the reactor vessel unless its temperature is considerably above its Nil Ductility Transiti m Temperature (NDTI).
~ttft tl SURVEILLANCE REQUIREMENT 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION
~Ati 1111 Applies to the required vessel temperature for pressurization.
Objective:
To assure that the vessel is not subjected to any substantial pressure unless its temperature is greater than its NDTT.
~till ao b.
During reactor vessel heat-up and cooldown when the reactor is, not critical the reactor vessel temperature and pressure shall satisfy the requirements of Figure 3.2.2.a.
During reactor vessel heat-up and cooldown when the reactor is critical the reactor vessel temperature and pressure shall satisfy the requirements of Figure 3.2.2. b, except when performing low power physics testing with the vessel head removed at power levels not to exceed 5 mw(t).
a.
Reactor vessel temperature and pressure shall be monitored and controlled to assure that the pressure and temperature limits are met.
b.
Vessel materi a1 surveil'l ance samples located within the core region to permit peri odic mmitoring of exposure and.
material properties shall be inspected on the follming schedule:
First capsule - one fourth service life Second capsule - three fourth service life Third capsule - standby In the event the surveillance specimens at one quarter of the vessels service life indicate a shift of reference temperature greater than predicted the schedule shall be revised as follows:
Amendment No.
58 Second capsule - one half service life Third capsule - standby 77
1
LIMITING CONDITION FOR OPERATION SURVEILLANCE R E UIR EMENT c.
During hydrostatic testing the reactor vessel temperature and pressure shall satisfy the require-ments of Figure 3.2.2.c if the core is not critical and Figure 3.2.2.d if the core is critical.
d.
The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head are are equal to or greater than 100F.
78 Amendment No.
58
I600 1400 l4I2 l200 LIMIT FOR NON-CRITICAL OPERATION INCLUDING HEATUP/COOLDOWN AT UP TO IOO F/HR Cg M IOOO
~o 0- a QJ CL K D 800 z I-V)
Z'J 600 UJ K m 0 z I-m O g.
400
'4 QJ 387 7I2 200 IOO I60 200 MINIMUM VESSEL TEMPERATURE (F)
FIGURE 3.2.2.a MINIMUM TEMPERATUR E FOR PRESSURIZATION DURING HEATUP OR COOLDOWN (REACTOR NOT CRITICAL)
(HEATING OR COOLING RATE < IOOF/HR)
', FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION
~ ~
LIMIT FOR NON-CRITICAL OPERATION INCLUDING HEAT-UP/COOLDOWN AT UP TO 100F/HR PRESSURE si
)
387 387 712 762 812 862 912 962 1012 1062 1112 1162 1212 1312 1412 TEMPERATURE (F) 100 100-160 160 166 172 177 182 187 192 196 199 203 207 213 21.9 TABLE 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEAT-UP OR COOLDOWN (REACTOR NOT CRITICAL)
(HEATING OR COOLING RATE 100F/HR)
FOR UP TO TEN EFFECTIVE FULL POWER YcARS OF CORE OPERATION Amendment No.
58
1600 l400 l4I2 l200 IOOO M ~
- a. o Q
CX O coo
~ I-M Z lZ 600 hl K K 0 z I-n O g 400 K
200 LIMIT OR POWER OPERATION
( CORE CRITICAL) INCLUDING HEATUP/COOLDOWN AT UP TO IOO F/HR 302 7I2 WATER LEVEL MUST BE IN NORMAL OPERATING BAND FOR CORE TO BE CRIT I CAL AT TEMPERATURES
< 200 F
IOO 200 MINIMUM VESSEL TEMPERATURE (F)
FIGURE 3.2.2.b MINIMUM TEMPE R ATURE FOR PRESSURIZATION DURING HEATUP OR COOLDOWN (REACTOR CRITICAL)
(HEATING OR COOLING RATE < IOOF/HR)
FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION Amendment No.
58
LIMIT FOR POWER OPERATION.(CORE CRITICAL) INCLUDING HEAT-UP/
'OOLDOWN AT UP TO 100F/HR PRESSURE si 302 312 362 387
. 387 712 762 812 862 912 962 1012 1062 1112 1162 1212 1312 1412 TEMPERATURE (F) 100 106 127 136 137-200 200 206 212 217
'22 227 232 236 239 243 247 253 259 TABLE 3.2.2.b MINIMUM TEMPERATURE fOR PRESSURIZATION DURING HEAT-UP OR COOLDOWN (REACTOR CRITICAL)
(HEATING OR COOLING RATE 100F/HR)
FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION Amendment No.
58
1 I
i t
I600 l400 l4I2 I200 C9 M
Q IOOO Q-o M"
M Q
<o 800 Q.I-~
M hl C5 600 Q
0 D I
M 400 o
Q CI 0-z:
200 930 LIMIT FOR INSERVICE TES
( CORE NOT CRI T ICAL, FUEL IN VESSEL)
IOO I30 200 MINllVlUM VESSEL TEMPERATURE (F)
F IGURE 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING (REACTOR NOT CRITICAL)
FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION Amendment No.
58
0
LIMIT FOR IN-SERVICE TEST
'CORE NOT CRITICAL, FUEL "
IN VESSEL)
PRESSURE si
~ 387 930 962 1012 1062 1112 1212 1312 1412 TEMPERATURE F
100-130 130 135 142 148 153 164 173 181 TABLE 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING (REACTOR NOT CRITICAL) fOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION Amendment No.
58
I600 I400 l4I2 l200 C5 V)
Q 1000 Q-o M
CO QO 800 Q.
U)
~Ul ID 600 K
O D M
I- ~
400 OK C5 0-200 I278 LIMIT FOR IN SERVICE TESTS (CORE CRITICAL).
387 100 l56 I70 200 MINIMUM VESSEL TEMPERATURE (F)
FIGURE 3.2.2.d g
MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTI NG (REACTOR CRITICAL)
FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION Amendment No.
58
LIMIT FOR IN-SERVICE TESTS (CORE CRITICAL)
PRESSURE si 387 1278 1312 1412 TEMPERATURE f 156 170 173 181 TABLE 3.2.2.d MINIMUM TEMPERATURE FOR PRESSURIZATION OURING HYOROSTATIC TESTING (REACTOR CRITICAL)
FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION
BASES FOR 3.2.2 AND 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figures 3.2.2.a and 3.2.2.b are plots of pressure versus temperature for a heat-up and cool down rate of 100F/hr.
maximum.
(Specificatim 3.2.1).
Figures 3.2.2.c and 3.2.2.d are plots of pressure versus temperature fa. hydrostatic testing.
These curves are based on calculations of stress intensity factors according te Appendix G cf Section III of the ASME Boiler and Pressure Vessel Code 1980 Editim with Winter 1982 Addenda.
In additi cn, temperature shifts due to integrated neutron flux at ten effective full power years of operation were incorporated into the figures.
These shifts were calculated from the fo.mula presented in Regulatory Guide 1.99, Revisicn I and the copper/phosphorus content of the reactor vessel.
These curves are applicable to the beltline region at low and elevated temperatures and the vessel flange at intermediate temperatures.
Reactor vessel flange/reactor head flange boltup is governed by other criteria as stated in Specification 3.2.2.d.
The pressure readings on the figures have been adjusted to reflect the calculated elevation head difference between the pressure sensing instrument locatims and the pressure sensitive area of the core beltline region.
The reactor vessel head flange and vessel flange in combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted together.
When the vessel head is placed on the reactor vessel, only that portim of the head flange near the inside of the vessel rests on the vessel flange.
As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flange.
Both the )cad and vessel and flange have a NDT temperature of 40F and they are not subject to any appreciable neutron radiati e exposure.
Therefa e, the mi nimun vessel head and head flange temperature for bolting the-head flange and vessel flange is established as 40 + 60F or 100F.
Figures 3.2.2. a, 3.2.2. b, 3.2.2. c and 3.2.2. d have incorporated a temperature shift due to the calculated integrated neutron flux.
Tlute integrated neutron flux at the vessel wall is calculated fry core physics data and has been measured using flux monitors installed inside the vessel.
The curves are applicable fcr up to ten effective full power years of operation.
Vessel material surveillance samples are lmated within the core regim to permit periodic monitoring of exposure and material properties relative to control samples.
The material sample program conforms with ASTM E 185-66 except for the material withdrawal scheduled which is specified in Specificatin 4.2.2.b.
82b
0
(2)
Note:
(3)
(4)
N cte:
Reactor protecti cn system cr engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functinal requirements of affected systems.
Conditions leading to operation in a degraded mode permitted by a.limiting condition for operation or plant shutdown required by a limiting conditi e f0 operaticn.
Routine surveillance testing, instruaent calibration, or preventative maintenance which require system configuratins as described in items 2.b(l) and 2.b(2) need not be reported except where test results themselves reveal a degraded mode as described above.
Observed inadequacies in the implementatisn of administrative or procedura1 contro1s which threaten, to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.
Abnormal degradatim of systems other than those specified in item 2.a(3) above designed to contain radioactive material resulting frcm the fission process; Sealed sources cr calibrati m sources are not included under this item.'eakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.
6.9.3 Unique Reporting Requirements Special reports shall be submitted to the Oirector of Regulatory Operaticns Reginal Office within the time period specified for each report.
These reports shall be submitted covering the activities identified belm pursuant to the requirements of the applicable reference specificaticn:
Reactor Vessel Material Surveillance Specimen Examination, Specification 4.2.2(c)
(12 months) 257f