ML19296B439

Tech Specs 1 & 1.3 Re Safety Limits & Limiting Safety Sys Settings of Reactor Protective Sys

ML19296B439
Person / Time
Site:	Fort Calhoun
Issue date:	02/12/1980
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19296B432	List: ML19296B430 ML19296B436 ML19296B439 ML19296B448
References
NUDOCS 8002200569
Download: ML19296B439 (2)

Text

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1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 System Settincs - Eeactor Protective System (Continued)

(3) Hich Pressuriner Pressure - A reactor trip for high pres-surizer pressure is provided in conjunction with the re-actor and steam system safety valves to prevent reactor coolant system overpressure (Specification 2.1.6). In the event of loss of load without reactor trip, the tem-perature and pressure of the reactor coolant system would increase due to the reduction in the heat removed from the coolant via the steam generators. The power-operated relief valves are set to operate concurrently with the high precourizer pressure reactor trip. This setting is 100 psi below the nominal safety valve cetting (2500 psia) to avoid unnecessary operation of the safety valves.

This setting is consisten+y with the trip point assumed in the accident analysis. (1)

(L) Thermal Margin / Low Pressure Trip - The thermal margin /

low pressure trip is provided to prevent operation when the DNBR is less than 1.3, including allowance for measure-ment error. The thermal and hydraulic limits shown on Figure 1-3 define the limiting values of reactor coolant pressure, reactor inlet temperature, and reactor power level which ensure that the thermal criteria (8) are not exceeded. The low set point of 1750 psia trips the re-actor in the unlikely event of a loss-of-coolant accident.

The thermal margin / low pressure trip set points shall be set according to the formula given on Figure 1-3 The variables in the formula are defined as:

B = High auctioneered thermal ( AT) or nuclear power in % of rated power.

T 13 = Core inlet temperature, OF.

PVAR = Reactor pressure, psia.

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DISCUSSION Su=ary The following document reports the revised TI!/LP trip equation for Fort Calhoun Cycle 6 stretch power operation.

The transient events which were reported in the Reference to have tripped on TM/LP have been reanalyzed to verify the adequacy of the final TM/LP equaticn. Results of the reanalysis are sun-marized:

(1) No event considered causes a TM/LP trip.

(2) Adequate nargin to DNB is maintained for the duration of the CEA withdrawal events by the high power and high pressure cetpoints.

(3) The most limiting transient considered, the fast CEA withdrawal event, trips on high power with MDNBR equal to 1.36.

(h) Excessive load increase incidents trip on high power or establish a new steady-state with substantial margin to DNB.

The TM/LP Trip Function The following function is the revised version of the Fort Cal-houn TM/LP LSSS:

P var = 5.5h PF(B)B + 22.h8 Tin - 10801. (1) 1.0 B > 100 PF(B) = -0.01B+2.0 50 < B < 100 15 B < 50 This function is based on final Cycle 6 neutronics calculations.

It protects core saturation limits and the SAFDL on DUB during those anticipated operational occurrences listed in category A of Table 1.1, Reference. In addition, it should prove amenable to existing Fort Calhoun analog control systems. Isobars from equation (1) are plotted in Figure 1.

Those category A transients (see above) for which TM/LP trips are reported have been reanalyzed with the new TM/LP function to assure adherence to the SAFDL on DNB. Analytical methodology for the reanalysis is as described in the Reference. The MDNBR'r and

Reference:

XN-NF-79-79, " Fort Calhoun Cycle 6 Reload Plant Transient Analysis Report," October 1979 ATTACIU4ENT B

core ecnditions calculated in the reanalysis are summarised in the attached Table 1 and indicate that the RPS as modified by inclusion of the new TM/LP trip equation adequately protects SAFDL's. The individual analyses are summarised below.

CEA Withdrawal The CEA withdrawal event is described in Section 3.1 of the Reference. The analysis was performed for a wide rance of reacti-vity additicn rates at both EOC and EOC conditions. All cases re-sulted in reactor trips on variable high power except the three LOC cases at the lower end of the reactivity insertion spectrum, which resulted in reactor trips on high pressure. Calculated MDNER's appear in Figure 2 as a function of reactivity insertion rate, and indicate that adequate thermal margin is maintained for all the CEA withdrawal incidents considered.

Excessive Load Increase Incident Excessive load increase incidents and analyces are as described in Section 3.h of the Reference. The cases considered are: (1) rapid opening of the turbine control valves, and (2) sudden opening of steam dump and steam bypass valves.

A new steady-state is attained in case (1) without initiating a reactor trip. Marginal cooldown and pressure decrease occur, accompanied by a small increase in power. The MDNER is 1.53.

For case (2), the reactor scrams on high power at about 9 seconds with a MDNER of 1.LT. Peak power level is 1702 MW. Cooldown and depressurisation are more rapid and extensive than for case (1).

Loss of t'eedwater Heating The loss of feedwater heating event and analysis are described in Section 3 5 of the Reference. Reanalysis results in a variable high power trip at about 25 seconds. Calculated MDNER is 1.43 at about 27 seconds. Adequate margin to DNB is maintained for the duration of the transient.

Conclusion The results of the transient analyses reported above indicate that the fast CEA withdrawal transient is the most limiting of those previously reported to have tripped on TM/LP. More than adequate margin to DNB is maintained for the duration of that limiting tran-sient. It may be concluded that modification of the RPS by sub-stitution of the revised TM/LP equation doer not impinge on the integrity of the RPS.

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-6 Positive Reactivity Insertion Rate in (SCC.)

FICRE 2 6;DNBR AS A FUNCTION OF WIT 1tDRANAL RATE F0il CEA 11ITi!!?RANAL INCIDENT