ML20067D658

Amend 162 to License DPR-69,revising Heatup & Cooldown Curves Which Will Allow Operation Beyond Current 12 EFPYs to Approx 13.8 EFPYs

ML20067D658
Person / Time
Site:	Calvert Cliffs
Issue date:	03/01/1994
From:	Capra R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20067D661	List: ML20067D658
References
NUDOCS 9403080347
Download: ML20067D658 (9)

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UNITED STATES hj [ j NUCLEAR REGULATORY COMMISSION

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BALTIM0RE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 i

CALVERT CLIFFS NVCLEAR POWER PLANT. UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE bP-69 j

cne o 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Baltimore Gas and Electric Company (the licensee) dated November 1, 1993, as supplemented on February 1, 1994, complies 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 withoui. 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 E.

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-69 is hereby amended to read as follows:

9403090347 9403o1 PDR ADOCK 05000318 P

PDR

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(2) Technica1 'Soeci fications i

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.162, are hereby incorporated in the j

license.

The licensee shall operate the facility in accordance with j

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the Technical Specifications.

l-3.

This license amendment is effective as of the date of its issuance to be i

implemented within 30 days.

-l FOR THE NUCLEAR REGULATORY COMMISSION i

l Robert A. Capra, irector

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Project Directorate I-l l

j Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation f

Attachment:

Changes to the Technical Specifications Date of Issuance: March 1, 1994 i

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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 162 FACILITY OPERATING' LICENSE N0. DPR-69 DOCKET NO. 50-318 Revise Appendix A as follows:

Remove Paaes Insert Paggs 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 3/4 4-32 3/4 4-32 3/4 4-34 3/4 4-34 B3/4 4-6 B3/4 4-6 B3/4 4-7 B3/4 4-7 l

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3/4.4 REACTOR C00LANT SYSTEM j

i 2500

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i HEATUP - __l

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EINSERVICE HYOROSTATIC TEST 8

2000 l

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$ 1500

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' CORE CRITICAL I

C E LOWEST E

5 SERVICE 4

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ETEMPERATURE d

5160*F-l l

p 1000 RCS TEMP.

H/U RATE E

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ALL TEMPS $75T/1HR 6

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t 5

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500 f

1 MIN. 80LTUP TEMP. 70 'F '

MAXIMUM PRESSSURE.

FOR SDC OPERATION ^

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100 200 300 400 500 600 l

INDICATED REACTOR COOLANT TEMPERATURE Tc, F 1

FIGURE 3.4.9-1 CALVERT CLIFFS UNIT 2 HEATUP CURVE. for FLUENCE $ 1.92x10" n/ca" l

REACTOR C0OLANT SYSTEM PRESSURE TEMPERATURE LIMITS 4

i CALVERT CLIFFS - UNIT 2 3/4 4-29 Amendment No. 162 j;

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3/4.4 REACTOR COOLANT SYSTEM 2500

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=lNSERVK:E HYDROSTATIC TESTL S&

2000 glOWEST

= SERVICE, E TEMPER ATURE E

51604 5

E 1500 g

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' COOLDOWN e

2.

g d

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5 M 1000 d

RCS TEMP.

CID RATE M

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> 180* F S100*F/1 HR h

180T TO 140'F

$40aF/1 HR

< 140*F s15'F/1 HR 5

500

- MlH. BOLTUP TEMP. 70 'F MAXIMUM PRESS'SUREi FOR SDC OPERATION i 0

100 200 300 400 500 603 INDICATED REACTOR COOLANT TEMPERATURE Tc, F FIGURE 3.4.9-2 CALVERT CLIFFS UNIT 2 C00LDOWN CURVE, for FLUENCE < 1.92x10" n/cm l

REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 2 3/4 4-30 Amendment No. 162

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Overoressure Protection Systems 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:

l 1.

Two power-operated relief valves (PORVs) with a lift setting i

of 5 430 psia, or l

2.

A single PORV with a lift setting of 5 430 psia and a Reactor Coolant System vent of 21.3 square inches, or 3.

A Reactor Coolant System (RCS) vent 2 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.

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

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

l d.

No more than one OPERABLE high pressure safety injection pump 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 o.ne of the following restrictions shall apply:

1.

The total high pressure safety injection flow shall be limited to < 210 gpm OR 2.

A Reactor Coolant System vent of 3 2.6 square inches shall exist.

e.

When not in use, the above OPERABLE HPSI pump shall have its handswitch in pull-to-lock.

APPLICABILITY: When the RCS temperature is 5 305 F and the RCS is vented to < 8 square inches.

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Except when required for testing.

CALVERT CLIFFS - UNIT 2 3/4 4-32 Amendment No. 162 l

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3/4.4 REACTOR C0OLANT SYSTEM l

LIMITING CONDITION FOR OPERATION (Continued) i 3.

If a pressure limit was exceeded, take action in accordance l

with Specification 3.4.9.1.

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The provisions of Specification 3.0.4 are not applicable.

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l SURVEILLANCE REQUIREMENTS I

i 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Perfonnance of a CHANNEL FUNCTIONAL TEST on the PORY actuation I

channel, but excluding valve operation, within 31 days prior to i

i entering a condition in which the PORY is required 0PERABLE and i

at least once per 31 days thereafter when the PORV is required OPERABLE.

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b.

Performance of a CHANNEL CALIBRATION on che PORV actuation channel at least once per 18 months.

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Verifying the PORV isolation valve is open at least once per i

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.

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

t 4.4.9.3.) hen the vent (s) is being used for overpressure protection.The Rj 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> w 4.4.9.3.3 All'high pressure safety injection pumps, except the above i

OPERABLE 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 j

electrical power supply circuits or by verifying their discharge valves are 1

locked shut. The autorritic opening feature of the high pressure safety l

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 />.

The above OPERABLE pump shall be verified to have its handswitch in pull-l to-lock at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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Except when the vent pathway is locked, sealed, or otherwise secured j

in the open position, then verify these vent pathways open at least once per 31 days.

i CALVERT CLIFFS - UNIT 2 3/4 4-34 Amendment No. 162 i

l 3/4.4 REACTOR COOLANT SYSTEM i

BASES the unit's yearly operating time since the activity levels allowed by l

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 l

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 i

is below the lift pressure of the atmospheric steam relief valves. The l

surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in l

sufficient time to take corrective action.

Infomation obtained on iodine spiking will be used to assess the parameters associated with spiking I

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 Reactor 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 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 current limits Figures 3.4.9-1 and 3.4.9-2, are for up to and including a fluence of 1.92x10 n/cm at the inner surface of the reactor vessel, which corresponds to approximately 13.8 Effective Full Power Years.

l The reactor vessel materials have been tested to determine their initial l

RTm; the results of these tests are shown in Section 4.1.5 of the UFSAR.

l Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RTm. The actual shift in RTm of the vessel l

material will be established periodically during operation by removing and 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 l

Table 4-13 and are approved by the NRC prior to implementation in i

compliance with the requirements of 10 CFR Part 50, Appendix H.

CALVERT CLIFFS - UNIT 2 B 3/4 4-6 Amendment No. 162 l

3/4.4 REACTOR COOLANT SYSTEM 4

BASES The shift in the material fracture toughness, as represented by RTer,f is calculated using Regulatory Guide 1.99, Revision 2.

For a fluence o 1.92x10" n/cm, at the 1/4 T position, the adjusted reference temperature (ART) value is less than 171 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 i

resulting in a composite limit curve for the reactor vessel beltline for the heatup event.

To develop a composite pressure-temperature limit for the cooldown event, the isothemal pressure-teraperature limit must be calculated. The isothemal 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 inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be detemined by entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for a fluence of 2

1.92x10" n/cm on Figures 3.4.9-1 and 3.4.9-2.

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 which the core can be brought critical.

q 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 RTur 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 pemitted to operate during a cooldown below 150 F.

CALVERT CLIFFS - UNIT 2 B 3/4 4-7 Amendment No. 162

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