ML20079G531

Proposed Tech Specs Revising pressure-temp Limits & Max Heatup Rate for RCS

ML20079G531
Person / Time
Site:	Millstone
Issue date:	01/04/1984
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20079G528	List: B10990, Application for Amend to License DPR-65,revising pressure- Temp Limits & Max Heatup Rate for RCS ML20079G531 ML20079G536
References
TAC-53512, NUDOCS 8401200165
Download: ML20079G531 (8)

Text

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Docket No. 50-336 i

Attachment 1 ,

I Millstone Nuclear Power Station, Unit No. 2 Proposed Revisions to Technical Specifications Pressure-Temperature Limits  ;

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Revise pages: 3/4 4-17 3/4 4-19b 3/4 4-20 B 3/4 4-6 Delete pages: 3/4 4-19c B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING 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, cocidown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 400F in any one hour period with Tavg at or below 2000F,500F in any one hour period with Tavg at or below 3000F and above 2000F, and 1000F in any one hour period with Tavg above 3000F.

b. A maximum cooldown of 1000F in any one hour period with Tavg above 3000F and a maximum cooldown of 200F in any one hour period with Tavg below 3000F.

c. A maximum temperature change of SoF in any one hour period, during hydrostatic testing operations above system design 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 limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations 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 av and pressure to less than 2000F and 500 psia, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• Sae Special Test Exception 3.10.3.

MILLSTONE - UNIT 2 3/4 4-17

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0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE T ' I c

Figure 3.4-2 Reactor Coolant System Pressure Temperature Limitations for 7 Full Power Years MILLSTONE - UNIT 2 3/4 4-19

TABLE 4.4-3 Reactor Vessel Material Irradiation Surveillance Schedule CAPSULE SCHEDULE (EFPY)

W-97 3.0 W-104 10.0 W-284 17.0 W-263 24.0 W-277 32.0

, W-83 Spare

-W-97 (Flux Monitor) 10.0 i

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i MILLSTONE - UNIT 2 3/4 4-20

REACTOR COOLANT SYSTEM B_ASES The heatup and cooldown limit curves (Figtre 3.4-2) are composite curves which were prepared by determining the most conservative case, with eitner the inside or outside wall controlling, f or any heatup or cooldown rates of up to 1000F 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 initia'.

RTNDT; the results of these test are shown in Table 4.6-1 of the Final Saf ety Analysis Report. Reactor operation and resultant iast neutron (E>l Mev) irradiation will cause an increase in the RTNDT. Theref ore, an adjusted reference temperattre, based upon the fluence, can be predicted using the methods described in SECY-82-465 "NRC Staf f Evaluation of Pressurized Thermal Shock", November,1982.

The heatup and cooldown limit curves shown on Figure 3.4-2 include predicted adjustments f or this shif t in RTNDT at the end of the applicable service period, as well as adjustments ior possible errors in the pressure and temperature sensing instr uments.

The actual shif t in RTNDT of the vessel material will be estaolished periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens 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 meastred transition shif t for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves mtst be recalculated when the RTNDT determined from the surveillance capsule is dif ferent from the calculated RT NDT for the equivalent capsule radiation exposure.

The r,ressure-temperature limit lines shown on Figure 3.4-2 ior 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 and hydrostatic testing.

The maximum RT NDT for all reactor coolant sptem pressure-retaining materials, with the exception of the reactor presstre vessel, has been determined to be 500F. The Lowest Service Temperature limit line shown on Figtre 3.4-2 is based upon this RTNDT since Article NB-2332 (Summer Addenda of 1972) of Section ill of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperattre to be RTNDT + 1000F MILLSTONE - UNIT 2 B 3/4 4-6

REACTOR COOLANT SYSTD4 .

i, SA5ES -

i for piping, pumps and valves. Below this temperature, the system pressure must be l'atted to a maxiom of 205 of the system's hydrostatic test pressure of 8125 psia.

Included in this evaluation is consideration of flange protection in accordance with 10CFR$0, Appendix G. TherequirementmakesgheminimumtemperatureRTNDT plus 900F for hydrostatic test and RTNDT plus 120 for normal operation when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure.

l 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 assure compliance with the requirements of Appendix H to 10 CFR Part 50.

ThelimitationsimposedonthepressurizerIEstupandcooldownrates

• and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fati-gue analysis perfomed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs or an RCS vent opening of greab'than 1.3 square inches ensures that the RCS will be prc,tacted from mss,ure

• transients which could exceed the limits of Appendix 5 to 10 CPR Part 50 when one or more of the RCS cold legs era < 275'F. Either PORV has  !

adequate relieving capability to protect tKe RCS from overpressurization  :

when the transient is limited to either (1) the start of an idle RCP 1 with the secondary water temperature of the steam generator < 43*F (31'F when measured by a surface contact instrument) above tKe coolant tenperature in the reactor vessel or (2) the start of a HPSI pump and its injection into a water solid RCS. .

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3/4.4.10 STRUCTURAL INTEGRITY l

The inservice inspection and testing programs foe ASME Code Class 1, 2 l

and 3 components ensure that the structural integrity ans 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 -

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ASME Boiler and Pressure Vessel Code and applicable Adde'nda as required by l 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).

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MILLSTONE - UNIT 2 B 3/4 4-7

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Docket No. 50-336 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Surveillance Capsule W-97 Analysis Results January,198te