ML20083Q945
| ML20083Q945 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/10/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20083Q947 | List: |
| References | |
| TAC-53512, NUDOCS 8404230397 | |
| Download: ML20083Q945 (10) | |
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<ja34%,jo, UNITED STATES NUCLEAR REGULATORY COMMISSION r,
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NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHtiSETTS ELECTRIC COMPAflY C0CKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION,llNIT fl0 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. DPR-65 1.
The Nuclear Regulatory Comission (the Comissioni has found that:
A.
The application for anendment by Northeast Nuclear Energy Comoany, et al. (the licensee) dated January 4,1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as arended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chauter I:
B.
The facility will operate in confornity with the application, the provisiers of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized hv this amendnent can ha conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in corpliance with the Comission's reculations; D.
Tha issuance o' this amendment will not be ininical to tha conncn 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 renulations and all applicable recuirements have been satisfied.
0404230397 040410 PDR ADOCK 05000336 P
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Accordingly, the license is amended by changes to the Technical i
i Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DpR-65 is hereby amended to read as follows:
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i (2) Technical Specifications I
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 94, are hereby incorporated in the license. The licensee shall operate 1
j the facility in accordance with the Technical Specifica-tions.
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This license amendment is effective on the date of issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION W
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i James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical i
Specifications i
Date of Issuance:
April 10, 1984 i
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t ATTACHMENT TO LICENSE AMEN 0 MENT NO. 94 j
FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336
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Remove and replace the following pages of the Appendix A Technical Specifica-tions with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corre-i sponding overleaf pages are provided to maintain document completeness.
Remove Insert 3/4 4-17 3/4 4-17 i
3/4 4-19a 3/4 4-19 3/4 4-19 3/4 4-19c 3/4 4-20 3/4 4-20 8 3/4 4-6 R 3/4 4-6 R 3/4 4-7 B 3/4 4-8 R 3/4 4-9 B 3/4 4-10 R 3/4 4-11 8 3/4 4-7 8 3/4 4-12 l
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3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM UMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
A maximum heatup of 40*F in any one hour period with T,y9 a.
at or below 200*F, 50*F in any one hour period with T atorbelow300*Fandabove200*F,and100'FinanyoM9 hour period with T above 300*F.
avg b.
A maximta cooldown of 100*F in any one hour period with T3yg above 300*F and a maximum cooldown of 20*F in any one hour period 'ith T,yg below 300'F.
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A maximum temperature change of 5*F in any one hour period, during hydrostatic testing operations above system design d
pressure.
APPLICABILITY: MODES 1, 2*, 3, 4 and 5.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limf t within 3'..inutes; perform an engineering evaluation to determine the effects or the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; detennine
. that the Reactor Coolant System remains acceptable for continued opera-tions or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200'F and 500 psia, respectively, withintheiollowing30 hours.
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- See Special Test Exception 3.10.3.
.f MILLSTONE - UNIT 2 3/4 4-17 Amendment No. M, 94 3
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SURVEILLANCE REQUIREMENTS 4.4.9.1 a.
The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
t b.
The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to making the reactor critical.
c.
The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in materiit!
properties ~, at the intervals shown in Table 4.4-3.
The results of these examinations shall be used to update Figure 3.4-2.
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0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE T. F g
t Figure 3.4-2 Reactor Coolant System Pressure Temperature Limitations fer 7 Full Power Years MILLSTONE - UNIT 2 3/4 4-19 Amendment No. 29, 94
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TABLE 4.4-3 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE CAPSULE
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W-97 3.0 W-104 10.0 W-284 17.0 i
W-263 24.0 W-277 32.0 W-83 Spare W-97 (Flux Monitor) 10.0 4
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REACTOR COOLANT SYSTEM BASES Reducing T to < 515'F prevents the release of activity should a steam generator dbe rupture since the saturation pressure of the a
primary coolant is below the lif t pressure of the atmospheric steam The surveillance requirements provide adequate assurance relief valves.
that excessive specific activity levels in the primary coolant will be Information detected in sufficient time to take corrective action.
obtained on iodine. spiking will be used to assess the parameters associated with iodine spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSUR'E/ TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.0 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients estab-lished during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which ~the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
MILLSTONE - UNIT 2 B 3/4 4-5
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REACTOR COOLANT SYSTEM 8ASES The heatup and cooldown limit curves (Figure 3.4-2) are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to 100*p per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figure 3.4-2.
The reactor vessel materials have been tested to determine their initial RT
- the results of these tests are shown in Table 4.6-1 of the FinalSafekfAnalysisReport. Reactor operation and resultant fast neutron Therefore, an (E>l Mev) irradiation will cause an increase in the RTN adjusted reference temperature, based upon the fluence,DT.can be predicted using the methods described in SECY-82-465 "NRC Staff Evaluation of Pressurized Thennal-Shock", November,1982.
The heatup and cooldown limit curves shown on Figure 3.4-2 include l
at the end of the applicable predicted adjustments for this shif t in RT@possible errors in the pressure T
service period, as well as adjustments for and temperature sensing instruments.
The actual shift in RT of the vessel material will be established periodically during operatibb by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance speci-mens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different from the calculaENaRT for the equivalent capsule radiation exposure.
NDT The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak l
and hydrostatic testing.
l for all reactor coolant system pressure-retaining The maximum RTNDT materials, with the exception of the reactor pressure vessel, has been determined to be 50*F.
The Lowest Service Temperature limit line since Article NB-2332 shown on Figure 3.4-2 is based upon this RT (Summer Addenda of 1972) of Section III of $N ASME Boiler and Pressur Vessel Code requires the Lowest Service Temperature to the RTNOT + 100*F MILLSTONE - UNIT 2 B 3/4 4-6 Amendment No. 94
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O REACTOR COOLANT SYSTEM 3ASES l
for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test 3ressure of 3125 psia.
Included in this evaluation is consideration of flange protection in accordance with 10 CFR 50, Appendix G.
The requirement makes the minimum temperature RTNOT plus 90*F for hydrostatic test and RTNDT plus 120*F for normal operation when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to a.ssure compliance with the requirements of Appendix H to 10 CFR Part 50.
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERA 8ILITY of two PORVs or an RCS vent opening of greater than 1.3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are < 275*F.
Either PORV has hdequate relieving capability to protect tee RCS from overpressurization Lhen the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator < 43*F (31*F when measured by a surface contact instrument) above the cooTant temperature in the reactor vessel or (2) the start of a HPSI pump and its injection into a water solid RCS.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable-Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(1).
MILLSTONE - UNIT 2 B 3/4 4-7 Amendment No. 50, 70, 94 1
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