ML18139C132

Proposed Tech Specs Re Full Flow Thermal Limits

ML18139C132
Person / Time
Site:	Surry
Issue date:	11/22/1982
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18139C131	List: ML18139C130 ML18139C132
References
NUDOCS 8211290737
Download: ML18139C132 (8)

Text

e Attachment 2 Technical Specification Changes for Surry Units 1 and 2 Full Flow Thermal Limits

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e TS 2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the integrity of the fuel cladding.

Specification A. The combination of reactor thermal power level, coolant pressure, and coolant temperature shall not:

1. Exceed the limits shown in TS Figure 2.1-1 when full flow from three reactor coolant pumps exist.

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2. Exceed the limits shown in TS Figure 2.1-2 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are open *.

3. Exceed the limits shown in TS Figure 2.1-3 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are closed.

e e TS 2.1-3 uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNB ratio (DNBR) during steady state operation, normal operational transients and anticipated transients, is limited to 1.30. A DNBR of 1. 30 corresponds to a 95%

probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operation conditions. (l)

The curves of TS Figure 2 .1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent limits equal to, or more conservative than, the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which the DNB ratio is equal to 1.30 or the average enthalpy at the exit of the core is equal to the saturation value. The area where clad integrity is assured is below these lines. The temperature limits are considerably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30 alone but are such that the plant conditions required to violate the limits are precluded by the self-actuated safety valves on the steam generators. The three loop operation safety limit curve has been ~evised to allow for heat flux peaking effects due to fuel densification and to apply to 100% of design flow. The effects of rod bowing are also considered in the DNBR analyses.

The curves of TS Figures 2.1-2 and 2.1-3 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation), represent limits equal to, or more conservative,

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e e TS Figure 2 .1-1 670 660 650

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Cl Q

640

.J 0 630 u

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+ 620 I-0

c:

-I-N 610.

j

'~-... 600

• {.

u 0:::

~

I-

< 5*90 c:::

w ci.

~

u 580 I-w

(.:,

< 570 c:::

w

< 560 550 0 10 20 30 40 50 60 70 80 90 100 11 C 1 20*

POWER (PERCENT OF RATED)

FIGURE 2.1-1 REACTOR CORE THERMAL & HYDRAULIC. SAFETY LIMITS-THREE LOOE OPERATION, 100%. FLOW

(b) e High pressurizer pressure - S 2385 psig.

e TS 2.3-2 (c) Low pressurizer pressure - ~ 1860 psig.

(d) Overtemperature AT AT~AT [Kl - K (1 + ,cl S 0 2 l ) (T - T') + K (P - P') - f(AI)]

+ ,c2S 3 where AT = Indicated AT at rated thermal power, °F 0

T = Average coolant temperature, °F T'= 574.4°F p = Pressurizer pressure, psig P' = 2235 psig Kl = 1.12 K2 = 0.01012 K3 = 0.000554 for 3-loop operation Kl = 0.951 K2 = 0. 01012 for 2-loop operation with loop stop K3 = 0.000554 valves open in inoperable loop Kl = 1.026 K2 = 0.01012 for 2-loop operation with loop stop K3 = 0.000554 valves closed in inoperable loop AI= qt - qb' where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt+ qb is total core power in percent of rated power f(AI) = function ofAI, percent of rated core power as shown in Figure 2.3-1 ttl = 25 seconds 11:2 = 3 seconds (e) Overpower ll.T b,.Ts_/.if [K - K ( ,c3S ) T - K (T - T') - f (AI)]

0 4 5 6 1 + ,c S 3

'* * . ~!

e TS 2.3-3 where AT = Indicated AT at rated thermal power, °F 0

T = Average coolant temperature, °F T' = Average coolant temperature measured at nominal conditions and rated power, °F K4 = A constant= 1.09 KS = 0 for decreasing average temperature A constant, for increasing average temperature 0.02/°F K6 = 0 for T ~T'

= 0.00108 for T>T' f(AI) as defined in (d) above,

,c = 10 seconds 3

(f) Low reactor coolant loop flow -~90% of normal indicated loop flow as measured at elbow taps in each loop (g) Low reactor coolant pump motor frequency -~57.5 Hz (h) Reactor coolant pump under voltage - ~70% of normal voltage

3. Other reactor trip settings (a) High pressurizer water level - $92% of span (b) Low-low steam generator water level - ~ 5% of narrow range instrument span (c) Low steam generator water level -~15% of narrow range instrument span in coincidence with steam/feedwater 6

mismatch flow - S l. OxlO lbs/hr (d) Turbine trip (e) Safety injection - Trip settings for Safety Injection are detailed in TS Section 3.7.

e e

• • TS 2.3-5 and source range high flux, high setpoint trips provide additional protection against uncontrolled startup excursions. As power level increases, during startup, these trips are blocked to prevent unnecessary plant trips.

The high and low pressurizer pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig) . The low pressurizer pressure reactor trip also trips the reactor in the unlikely event of 3

a loss-of-coolant accident. ( )

The overtemperature aT reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 3 seconds), and pressure is within the range between high and low pressure reactor trips. With normal axial power distri-(2) bution, the reactor trip limit, with allowance for errors, is always below the core safety limit as shown on TS Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor (4)(5) limit is. automatically reduced.

The overpower and overtemperature protection system setpoints have been revised to include effects of fuel densification on core safety limits and to apply to 100% of design flow.

the Technical Specifications will ------

ensure Th~_revised setpoints in that the power, temperature, and pressure will not exceed the revised combination of

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• e COMMONWEALTH OF VIRGINIA)

)

CITY OF RICHMOND )

The foregoing document was acknowledged before me, in and for the City and Commonwealth aforesaidj today by W. L. Stewart, who is Vice President-Nuclear Operations, of the Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this,:;,(~,,,,_( day of 71c:rv-.o~ ~. - , 19 )cl.

My Commission expires:

_________ , 19 ~s--

Notary Public (SEAL)

M2/004