ML19296C052

Amend 28 to License DPR-64,revising Reactor Coolant Sys Pressure & Temp Heatup & Cooldown Curves & Adding Requirement to Conduct Surveillance for Core Reactivity Anomalies

ML19296C052
Person / Time
Site:	Indian Point
Issue date:	01/19/1980
From:	Schwencer A Office of Nuclear Reactor Regulation
To:
Shared Package
ML19296C045	List: ML19296C044 ML19296C052 ML19296C060 ML19296C062
References
NUDOCS 8002250126
Download: ML19296C052 (13)

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UNITED STATES a

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NUCLEAR REGULATORY COMMISSION 2

3.c. (

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,E W ASHINGTON, 0. C. 20555 N;r/

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POWER AUTHORITY OF TPE S ATE OF NEW YORK DOCKET h;0. 50-286 INDIAN POINT NUCLEAR GENE 3ATING UNIT NO. 3 AMENDMENT TO FACILITY 03ERATING LICENSE Amendment No. 28 License No. OPR-64 1.

The Nuclear Regulatory Cormission (tte Comrission) has found that:

a thority of the State A.

The application for amendment :y wer u

of New York (the licensee) dated 2.ugust 24,1979, complies with the standards ana require e~s of the Atoric Energy Art of 1954, as anended (the Act) a n:

ne C -ission's rules and regulations set forth in 10 CF: ' aster :-,

B.

The facility will operate in c:n':mi;;. v.ith the application, the provisions of the Act, and tre 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 ',i') that such activities will be

. conducted in compliance with the "cmission's regulations; D.

The issuance Cf this amendre-t w il nc*. be inirical to the connon defense and security or to the hstith and safety of the public; and E.

The issuance of this amendrent 1: in acccrdance with 10 CFR Part 51 of the Connission's regulat ic s and all applicable requirenents have been satisfied.

25012 '

8002

. 2.

Accordingly, the license is anended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Faci'ity Operating License No. DPR-64 is hereby anended to read as follows:

(B)

Technical Specifications The Technical Specifications cot:tained in Appendices A and B, as revised through Amendment No. 28, are hereby incorporated in the licer.se.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendnent is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

.& I (WNk A. Rhwencer, Chief C:e ating Peactors Erar

1. 1 Civ sion of Operating C

. tors Att ec;, ment:

changes to the Technical Specifications Date of Issuance: January 19, 1980

ATTACPI'- ~~ ' ' ~E' E l.vE!CI': t 0.

23 FACIL*T~ ::E:'T:K5 LICENSE

',0. C0F-6' C0CrlT 'i:. 50-2Ef Revise Appendix A as fc'l awI:

Remove Pages Irsert Pages 3.1-4 3.1-4 3.1 -5 3.1-5 3,1-9 3.1-9 10 3.1-10

.. ~,.1 1 3.1-11 3.10-8 3.10-8 3.10-Ba 3.1:-8a 4.3-1

4. 3-1 4.3-2 4.3-2 4.3-3 t.3-3 O

_B.

HT.A7.'P AND COOUXMN Specifications 1.

The reactor coolant temperature and pressure and system heatup and cooldown rates averaged over one hour (with the exception of the pressurizer) shall be limited in acccrdance with Figure 3.1-1 and Figure 3.1-2 for the service period up to 9.26 effective full-power years (EFPYs).

The heatup or cooldowr rate shall not exceed 1000F/tr.

Allovable combinations or pressure and temperature for specific a.

temperature change rates are below and to the right of the lir.it lines shown.

Limit lines for cooldown rates between those pre-sented may be obtained by interpolation.

Figure 3.1-1 and Figure 3.1-2 define limits to assure prevention b.

of non-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2.

The limit lines shown in Figure 3.1-1 and Figure 3.1-2 shall be recalcu-lated periodically using metheds discussed in the Basis and results of surveillance specimens as covered in 2;scification 4.2.

The crder cf specimen removal may be modified hased cr the results of testing of previously removed specimens.

3.

The secondary side of the steam generator shall not be pressurized 0

above 200 psig if the temperature of the steam generator is below 70 F.

The pressurizer heatup and cooldown rates averaged over one hour shall 4.

not exceed 1000F/hr and 2000T/hr, respectively.

The spray shall not be used if the temperature dif f erence between the pressurizer and tne C

spray fluid is greater than 32C F.

5.

Reactor Coolant System integrity tests shall be performed in accordance with section 4.3.

Basis Fracture Toughness Properties The fracture toughness properties of the ferritic materials in the Amendment No. 28 3.1-4

reactor vessel are determined in accordance with the Summer 1965 Section (6) and ASTM E185 (5) and III of the ASME Boiler and Pressure VesselIode in accordance with additional reactor vessel requirements.

These properties are then evaluated in accordance with Appendix G of the 1972 Summer Addenda to Section III of the ASME Boiler and Pressee Vessel Code (1 ), and the calculation methods described in WCAP-7924 12 ).

The first reactor vessel material surveillarce capsule was re:noved during the 1978 refueling outage. This capsule has been tested by Westinghouse (7).

Based Corporation and the results have been evaluated and reported on the Westinghouse evaluation, heatup and cooldown curves (Figures 3.1-1 and 3.1-2) were developed for up to 9.26 EFFYs of reactor operation.

The maximum shift in RTIOT after 9.26 EFPYs of operation is projected to be 1270F at the 1/4 T and 85or at the 3/4 T vessel wall locations, f or Plate B2Es03-3 the controlling plate.

Plate B2903-3 was also the controlling plate for the first operating period of 2 ITFYs.

Heatup and cooldown limit curves are calculated using the mest limiting l'

value of RTIOT at the end of 9.26 years of service life.

The 9.26 year service life period is chosen such at the li-iting RT 37 at the 1/4 T Iccation in the core region is higher than che RT.

of the li:-iting unirradiated material.

This service periof assures that all co ponents in the Reactor Coolant System will be operated conservatively in accord-ance with Code reco=cendations.

of the core region naterial is detert.ined by adding The highest RTICT for the applicable time period to the original the radiation induced ARTIOT RT;c; shown in Table Q4.2-1 (3).

The fast neutron (E > 1 Mev) fluence at 1/4 thickness and 3/4 thickness vessel loca*. ions is given as a function cf full power service life in Figure A-2 (7).

Using the applicable fluence the end of the 9.26 year period and the capper content of the material at is obtained from Figure 4.4-3 (4 ).

in question, the 4RTg7

=

Amendment No.

28 3.1-5

RITERENCES:

1.

ASME Boiler and Pressure Vessel Code,Section III, 1972 Su=mer Addenda.

2.

WCAP-7924, " Basis for Heatup and Cooldown Limit Curves",

W. S. Hazelton, S. L. Anderson, S. E. Yanichko, July 1972.

3.

FSAR Volume 5, Response to Question Q4.2.

4.

FSAR Section 4.

5.

ASTM E185-70, Surveillance Tests on Structural Materials in Nuclear Reactors.

6.

ASME Boiler and Pressure Vessel Code,Section III, Su==er 1965.

7.

WCAP-9491, "A alysis of Capsule T frc:- the Indian Point Unit !;o.

3 Reactor Vessel Radiation Surveillance Pro:rar",

J.A.

Davidson, S.L. Anderson, W.T.

Kaiser, Apri'. 19 :-

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3.10.9 Rod Position Monitor If the rod position deviation monitor is inoperable, individual rod positions shall be logged once per shift and after a load change greater than 10 percent of rated power.

.3.10.10 neactivity Balance The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1% dk/k at least once per 31 Effective Fuel Power Days (EFPD).

This comparison shall, at least, i

consider reactor coolant system boron concentration, control rod position, reactor coolant system average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, and I.

samarium concentration. The predicted reactivity values shall he adjusted (normalized) to ccrrespen: to the actual ccre conditien i'

prior to exceeding a fuel h_rnup :f E: ETFD af ter ea:h fuel lcading.

3.10.11 Notification

]

Any event requiring plant shutdown or trip setpoint reduction because of Specification 3.10 shall be reported to the !!uclear Regulatory Commission within 30 days.

EAEIS Design criteria have been chosen for normal operations, operati:nal transients and those events analyzed in FSAR Section 14.1 which are consistent with the fuel integrity analyses.

These relate to fission gas release, pellet temperature and cladding mechanical properties.

Also, the minimum DNBR in the core must not be less than 1.30 in normal operation or in short term transients.

In addition to the above conditions, the peak linear power density must not exceed the limiting Kw/f t values which result from the large break loss of coolant Amendment No.

28 3.10-8

0 accident analysis based on the ECCS acceptance criteria limit of 2200 r.

This is required to meet the initial conditions assumed for loss of coolant accident analyses.

To aid in specifying the limits on power distributics, the following hot channel factors are defined.

9 Fg(Z), Heicht Dependent Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elecation divided by the average fuel rod heat flux, allowing for manufacturin: tolerances on fuel pellets and rods.

Amendment No. 28 3.10-8a

4.3 REACTOR COOLANT SYSTEM IhTIGRITY TESTING Applicability Integrity.

Applies to test requirements for Reactor Coelant Syste:

Obiective To specify tests for Reactor Coolant Syster integrity af ter the syster is closed following normal opening, modification er repair.

Specification When the Reactor Coolant Syste= is closed after it has been opened, a) the systec will be leak tested at not less than 2335 psig and in accordance with NDT require:ents for te=perature, b)

When Reactor Coolant Syste: medifications or repairs have beer made which involve new strer.gth velds on coeponents, the new welds vill eeet the require =ents of ASMI Section XI,15400 and IS500.

c)

The reactor coolant syste: leak test te=perature-pressure relation-

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ship shall be in accordance with the li=its of Figure 4.3-1 for

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heatup for the first 9.26 Ir?Y's of operation.

T:;ure 4.3-1 vill be recalculated periodically.

Alleva':le pressures during cooldown from the leak test temperature shall be in accordance with Figure 3.1-2.

Basis For normal opening,the integrity of the syste=, in terns of strength, is unchanged.

If the system does not leak at 2335 psig (Operating pressure + 100 psi: 1100 psi is normal system pressure fluctuation),

it will be leak tight during normal operation.

Amendment No. 28,

4. 3-1

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J For repairs on co:penents, the tic:cu;h r.:n-destru::ive testing gives a very high degree cf confide:.:e i

.e 1::e;;:.:. :f the rys:er, and vill detect any sigr.ificant de f e::s i: atf near thi mew elis.

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2 za; all cases, the leak test will assure leak :ig:::ess dur.::

operation.

4.~-1.

The The inservice leak test te,e:at::es are sh:._ :n yiru :

terperatures are calculated i= a:::ria :e with J.5E o:e Se:: ion III, Appendix G.

This Cede requires -hat a saf ety.f ac::: ef 1.5 tires the stress intensity f actor caused by pressure be appl:ad t: the calculation.

Ter the first 9.2E ef f e::ive f _1*. r:ver year s.1: :s : ti.::ed that the 11 highest F.!,.__

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ter: te ;erature recuire ents for periods 0; :: 13: ef fe::ive full prwe year are shown or. Figure 4.3-1.

The hea:up lirr.its specified c

he hea:c; cu ve, Tig ure 3-1, cust not be exceeded while the reactor c:cLar: syste: is bei=g :estei :c the

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4. 5-3 and 3.1-2 ar e re ca*,:.~.a e: p e ;;di:aZ, us. ; ta : r. : s di s cu s s ed in t'. e Basis f or Spe:ificatie: 2.1.1 and res.~.:s : f su-. e.;;ar:e s p ecimens,

as ccvered in Spe:ificatie:

.2 Referen_ce l.

FSAR, Section 4.

Amendment No.

28 4.3-2

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