ML20059H835

Proposed Tech Specs to Extend Use of Current Heatup & Cooldown Curves to Allow Operation Beyond 12 EFPY

ML20059H835
Person / Time
Site:	Calvert Cliffs
Issue date:	11/01/1993
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20059H829	List: ML20059H827 ML20059H835
References
NUDOCS 9311100240
Download: ML20059H835 (7)

Text

.

ATTACIIMENT (1) i UNIT 2 .

i TECIINICAL SPECIFICATION REVISED PAGES 3/4 4-29 3/4 4-30 3/4 4-32 3/4 4-34 B 3/4 4-6 B3/44-7 V

1 4

i 1

l 1

1

)

i l

I 9311100240 931101 cl1 p

PDR ADOCK 05000318 .g p PDR -h ,

I l

3/4.4 REACTOR COOLANT SYSTEM i

i

{ l l

l l

2500 -i-HEATUP - . _ _ , _

EINSERVICE HYDROSTATIC TESTL 2000

$ =

lC 1500 i_

=_,_ LOW EST

' CORE CRITICAL

$ 5 SERVICE g 5 TEMPERATURE N E 160'F 't m

^ 1000 W d RCS TEMP. H/U RATE o.

O '-

ALL TEMPS 5757/1 HR 5

8 ..

3 500 .'

AMIN. BOLTUP TEMP. 70 *F-MAXIMUM PRESSSURE FOR SDC OPERATION 0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE Tc, F l

ge y Laenew & l.9LX/O"nb'm' FIGURE 3.4.9-1 l CALVERT CLIFFS UNIT 2 ilEATUP CURVE, REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS i CALVERT CLIFFS - UNIT 2 3/4 4-29 Amendment No fih l

i

3/4.4 REACTOR COOLANT SYSTEM

/'

2500- .

EINSERVICE HYDROSTATIC TEST:

N

4 _

2000 E LOWEST

~ESERVICE

$ E TEMPER ATURE

$ 5160T [

1500 -

m -

t0 - __

' ' ~COOtoOwn e

LJ d

a

& W 1000

[0 RCS TEMP. C/D RATE .i i

' S1004/1 HR r

> 180* F 1809 TO 140*F $40*F/1 hA h <14 0*F S15'F/1 MR

~

c o z

~

500

- - MIN. BOLTUP TEMP. 70 'F MAX 1 MUM PRESS'SURE; FCR SDC'OPER ATION :

100 200 300 400 500 60C 0

INDICATED REACTOR COOLANT TEMPERATURE Tc, F i

ice Fut'NC6 L l.94%/0 /1lC!!!' I FIGURE 3.4.9-2 CALVERT CLIFFS UNIT 2 C00LDOWN CURVE,M REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 2 3/4 4-30 Amendment No. J49

-i

~

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 f- PRESSURE / TEMPERATURE LIMITS Overpressure Protection Systems l

' LIMITING CONDITION FOR OPERATION 3.4.9.3 The following overpressure protection requirements 'shall be met:

a. One of the following three overpressure protection systems shall be in place.

1. Two power-operated relief valves (PORVs) with a lift setting of 5 430 psia, or  ;

2. A single PORV with a lift setting of 5 430. psia 'and a j

Reactor Coolant System vent of 3 1.3 square inches, or

3. A Reactor Coolant System (RCS) vent 1 2.6 square inches.  ;

b. Two high pressure safety injection (HPSI) pumps' shall be . ,

disabled by either removing (racking out) their motor circuit -

breakers from the electrical power supply circuit,' or by locking .

shut their discharge valves. .

c. The HPSI loop motor operated valves (MOVs)# shall be prevented

1. from automatically aligning HPSI pump flow to the RCS by placing their handswitches in pull-to-override. I 1

d. No more than one OPERABLE high pressure.s'afety injection pump -I with suction aligned to the Refueling Water Tank may be used to  :

inject flow into the RCS and when used, it must be under manual ';

control and one of the following restrictions shall apply: .

1. The total high pressure safety injection flow shall~ be i limited to 5 210 gpm OR l t

2. A Reactor Coolant System vent of 2 2.6 square inches shall ,

exist. j APPLICABILITY: When the RCS temperature is $ 305 F and the RCS~is vented to < 8 square inches.

, e, idhen ned h) ase; he abwe DMKASL6 Mgit pnessMe sdek '

t ht(ecky pantp shntl have Hs hadsuilcJt M pall-Lkek _

N

~

l l

Except when required for testing.

CALVERT CLIFFS - UNIT 2 3/4 4-32 Amendment No. $  ;

.i

3/4.4 REACTOR COOLANT SYSTEM

( LIMITING CONDITION FOR OPERATION (Continued)

3. If a pressure limit was exceeded, take action in accordance with Specification 3.4.9.1.

g. The provisions of Specification 3.0.4 are not applicable. i SURVEILLANCE REQUIREMENTS l

4.4.9.3.1 -Each PORY shall be demonstrated OPERABLE by:  !

a. Performance of a CHANNEL FUNCTIONAL TEST on the PORY actuation i channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORY is required OPERABLE.

b. Perfomance of a CHANNEL CALIBRATION on the PORV actuation '

channel at least once per 18 months.

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORY is being used for overpressure protection, h '

d. Testing in accordance with the inservice test requirements pursuant to Specification 4.0.5.

4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />' when the vent (s) is being used for overpressure protection.

4.4.9.3.3 All high pressure safety injection pumps, except the above  ;

OPERALLE pump, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i by verifying that the motor circuit breakers have been removed from their electrical power supply circuits or by verifying their discharge valves are ]

locked shut. The automatic opening feature of the high pressure safety  :

injection loop MOVs shall be verified disabled at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. I A

A oboa 6PERMu?pwnpsk// be vee %d6 Aue J/s hMsu);kh ,

ln yll-4-/xL d /ad usce pa e hoas . .' ;

^ ~ -

Except when the vent pathway is locked, sealed, or otherwise secured in the open position, then verify these vent pathways open at least once per 31 days.

CALVERT CLIFFS - UNIT 2 3/4 4-34 Amendment No. # 7 ,

3/4.4 REACTOR COOLANT SYSTEM BASES the unit's yearly operating time since the activity levels allowed by Figure 3.4.8-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing T,,, to < 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive -

specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with 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 PRESSURE / TEMPERATURE LIMITS All components in the Rec.ctor Coolant System are designed to withstand 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 operation. The various categories of load cycles used for design purposes are provided in Section 4.1.1 of the UFSAR.

During STARTUP and shutdown, the rates of temperature and pressure changes g 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.

Operation within the appropriate heatup and cooldown curves assures the integrity of the reactor vessel against fracture induced by combinative thermal and pressure stresses. As the vessel is subjected to increasing fluence, the toughness of the limiting material continues to decline, and even more restrictive Pressure / Temperature limits must be observed. The e 4  ? el Sh:c)

The reactor vessel materials have been tested to detemine their initial cugMFM RT,n; the results of these tests are shown in Section 4.1.5 of the UFSAR. -4> apPWI'd Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will /3.8 EM&c cause an increase in the RT,n. The actual shift in RT,n of the vessel ps/ he material will be established periodically during operation by removing and yegs, evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in UFSAR Table 4-13 and are approved by the NRC prior to implementation in compliance with the requirements of 10 CFR Part 50, Appendix H.

CALVERT CLIFFS - UNIT 2 B 3/4 4-6 AmendmentNo.,jV

(..

1 l

. )

3/4.4 REACTOR COOLANT SYSTEM a C/uence s / N Y 0 b "

BASES -

7" The shif t in the material fracture toughness, as represen 4RTm, is calculated using Regulatory Guide 1.99, Revision 2. For ? EFP, at the A 1/4 T position, the adjusted reference temperature (ART) va ue is471*F.

At the 3/4 T position the ART value is 125*F. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G to calculate heatup and cooldown limits in accordance with the requirements of 10 CFR Part 50, Appendix G.

To develop composite pressure-temperature limits for the heatup transient. -

the isothermal,1/4 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given themal rate. Then the most restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event. i To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive allowable pressure-temperature limit is chosen resulting in a composite '

limit curve for the reactor vessel beltline.

Both 10 CFR Part 50, Appendix G and ASME, Code Appendix G require the development of pressure-temperature limits which are applicable to W inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the test pressure (1.1 times nomal operating pressure) a d ocatingph corresponding temperature. This curve is shown for' 2 EF- on Figures 3.4.9-1 and 3.4.9-2. - 4 fj ,ge,gf,9 d ade Similarly,10 CFR Part 50 specifies that core critical limits be

• established based on material considerations. This limit is shown on the heatup curve. Figure 3.4.9-1. Note that this limit does not consider the '

core reactivity safety analyses that actually control the temperature at i which the core can be brought critical. 1 The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure (625 psia). This temperature is defined as equal to the most limiting RTm  !

for the balance of the Reactor Coolant System components plus 100'F, per Article NB 2332 of Section III of the ASME Boiler and Pressure Vessel Code. ,;

The horizontal line between the minimum boltup temperature and the Lowest )

Service Temperature is defined by the ASME Boiler and Pressure Vessel Code '

as 20% of the pre-operational hydrostatic test pressure. The change in the line at 150 F on the cooldown curve is due to a cessation of RCP flow induced pressure deviation, since no RCPs are permitted to operate during a ,

cooldown below 150*F.

CALVERT CLIFFS - UNIT 2 B 3/4 4-7 Amendment No. J5&

_ ,