ML20069A678

Proposed Tech Specs,Supporting Cycle 10 Control Rod Trip Insertion Time Testing

ML20069A678
Person / Time
Site:	Three Mile Island
Issue date:	05/20/1994
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20069A671	List: ML20069A666 ML20069A673 ML20069A678
References
NUDOCS 9405260105
Download: ML20069A678 (11)

Text

. , .

$ Table 2.3-1 9 .

REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS (5)

Four Reactor Coolant Three Reactor Coolant One Reactor Coolant 5

~

Pumps Operating Pumps Operating Pump Operating in (Nominal Operating (Nominal Operating Each loop (Nominal Shutdown

% Power - 100%) Power - 75%) Ooerating Power - 49%) Bypass

?

N 1. Nuclear power, max. 105.1 105.1 105.1 5.0(2) l

?  % of rated power (6)

2. Nuclear power based on 1.08 times flow 1.08 times flow 1.08 times flow minus Bypassed w flow (1) and imbalance minus reduction due minus reduction due reduction due to

% max. of rated power to imbalance to imbalance imbalance

3. Nuclear power based Bypassed

- (4) on pump monitors NA NA 55%

g max. % of rated power

~

4. High reactor coolant 2355 2355 2355 1720(3) y system pressure, f, psig max.

o

5. Low reactor coolant 1900 1900 1900 Bypassed system pressure, psig min.

6. Reactor coolant temp. 618.8 618.8 618.8 618.8 F., max.

7. High Reactor Building 4 4 4 4 pressure, psig max.

(1) Reactor coolant system flow, %.

(2) Administratively controlled reduction set during reactor shutdown.

(3) Automatically set when other segments of the RPS (as specified) are bypassed.

(4) The pump monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.

(5) Trip settings limits are limits on the setpoint side of the protection system bistable connectors.

(6) During plant startup from 0% to 47% power, this setpoint shall be lowered to 63% Full Power. During plant shutdown, the high flux trip setpoint change to 63% power shall be initiated within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of reaching a power level at or below 47% full power.

9405260105 940520 PDR ADOCK 05000289 P PDR

~

4.7- REACTOR CONTROL R0D SYSTEM TESTS 4.7.1 CONTROL R0D DRIVE SYSTEM FUNCTIONAL TESTS Applicability Applies to the surveillance of the control rod system.

Objective To assure operability of the control rod system, u

Speci fication j 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions following each l refueling outage prior to return to power. The maximum control rod i trip insertion time for an operable control rod drive mechanism, j except for the axial power shaping rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66* seconds at hot reactor coolant full flow conditions or - l 1.40 seconds for the hot no flow conditions (Reference 1). For the APSRs it shall be demonstrated that loss of-power will not cause rod movement. If the trip insertion time above is not met, the rod shall be declared inoperable.

I 4.7.1.2 If a control rod is misaligned with its group average by more than . ~

an indicated nine inches, the rod shall be_ declared inoperable and the limits of Specification 3.5.2.2 shall apply. The rod with the greatest misalignment shall be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing the average position of the group for. determining the i operability of rods with lesser misalignments.

4.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.

'For the remainder of Cycle 10, the following control rods will be considered operable if the maximum trip insertion time from the fully withdrawn position to 3/4 insertion does not exceed 2.11 seconds at hot coolant full flow d

conditions: Control Rods 1-1, 1-2, 1-3, 3-3, 3-4, 3-5, 3-6, 4-5, 5-4, 5-7, 5-9, and 6-5. The optional hot no flow test and its 1.40 second acceptance value is unchanged.

Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has actuated the 25% withdrawn reference switch during insertion from the fully withdrawn position. The specified trip time is based upon the safety analysis

, in UFSAR, Chapter 14 and the Accident- Parameters as specified therein. The specified trip time of 2.11 seconds for Cycle 10 is based upon reanalysis of the limiting safety analyses using a bounding trip time of 3.0 seconds for all control rods at hot reactor coolant full flow conditions.

4-48 Amendment No. JE7

- - . . .- . ~ ... . . - . . - . -. . _- ...

Each' control rod' drive mechanism shall be exercised by a movement of at least l.

, two~1nches of travel every two weeks. This requirement shall apply to either a partial or. fully withdrawn control rod at reactor operating conditions. .

Exercising the drive mechanisms in this manner provides assurance of reliability of the mechanisms.

A rod is considered inoperable if it cannot be exercised, if the trip insertion time is greater than the specified allowable time, or if the rod deviates from its group average position by more than nine inches. Conditions for operation with an inoperable rod are specified in Technical Specification 3.5.2.

REFERENCE (1) UFSAR, Section 3.1.2.4.3 " Control Rod Drive Mechanism" 4-49 i

j Amendment No. JE7

,y5t wy-* -m

A TABLE 1 TMI-l FSAR TRANSIENTLACCIDENT REVIEW EVENT COMMENT

1. Uncompensated Operating Reactivity Not Affected Changes

2. Startup Accident Reanalyzed

3. Rod Withdrawal Accident at Rated Power Reanalyzed Operation

4. Moderator Dilution Accident Not Limiting

5. Cold Water Accident Not Affected

6. Loss-of-Coolant Flow Reanalyzed

7. Stuck-out, Stuck-in, or Dropped Control Not Affected Rod Accident

8. Loss of Electric load Not Limiting

9. Steam Line Break Not Limiting

10. Steam Generator Tube Rupture Not Affected

11. Fuel Handling Accident, Waste Gas Tank Not Affected Rupture, Fuel Cask Drop Accident

12. Rod Ejection Accident Reanalyzed

13. Large Break Loss of Coolant Accident / Not Affected Maximum Hypothetical Accident

14. Small Break LOCA Evaluated (Bounded by #13)

15. Loss of Feedwater Accident Reanalyzed I

1 l

s

~

Table 2 Summaly of Startuo Accident Reanalyjiis 1 Acceptance 2765 psia Criteria FSAR KAPPB Method FSAR 2.15E-4AK/K/sec reactivity insertion rate Assumptions MTC of + 0.9E-4AK/K/F FSAR 2653.4 psia Results TSCR KAPPB Analysis Method TSCR Original FSAR assun'ptions.

Assumptions 1.34 second scram delay.

TSCR 2544 psia (includes 1% valve accumulation)

Analysis High flux trip setpoint reduced to 63% when power is _< 47%

Results Why is It Establishing the high flux trip setpoint at 63% for power levels below Conservative 47%FP limits the pressure increase to well below acceptance criteria.

Fixed trip delay instead of slowed reactivity insertion.

The rod withdrawal rate is conservative and cannot be achieved without failure of interlocks and operator action.

The MTC for the balance of Cycle 10 is increasingly negative.

Expected Lower pressure than TSCR method because:

FSAR Method 1. Rods start moving immediately at time of trip. l Result 2. The MTC for the balance of Cycle 10 is increasingly negative.

l l

I

Table 3 Summarv of Rod Withdrawal at Raled Power

• A_qcident Roanalysiji Acceptance 2765 psia 112% Thermal Power Criteria FSAR KAPPB Method FSAR 2568 MWt Assumptions 5.0E-5AK/K/sec reactivity insertion rate MTC of 0.0AK/K/F FSAR 2479 psia 110.0% Thermal Power Results TSCR Assumed the pressurizer safety valves (PSVs) would lift.

Analysis Original KAPPB data was obtained and the maximum pressurizer surge line Method flowrate was used to calculate the maximum volumetric insurge rate.

Compared the insurge rate to the PSV capacity at lift setpoint pressure.

Insurge rate was found to be within the capacity of the PSVs.

Peak pressurizer pressure would be limited to the PSV lift setpoint.

The rate of change in thermal power was obtained from the KAPPB data at the time just prior to reactor trip and multiplied by the additional 1.34 second delay to determine the increase in thermal power from the original analysis result.

TSCR Original FSAR assumptions.

Assumptions PSVs assumed to lift.

TSCR 2540 psia 110.5% Thermal Power Analysis Results l

Why is it Assumes PSVs lift. In actuality, peak pressure may remain below the PSV i Conservative lift setpoint. Increased lift setpoint due to valve accumulation (1%) was used and results in higher peak pressure prediction, j Fixed trip delay instead of slowed reactivity insertion.

The rod withdrawal rate is conservative and cannot be achieved without j failure of interlocks and operator action. l I

The MTC for the balance of Cycle 10 is increasingly negative.

Expected Pressurizer pressure may remain less than 2500 psia. The maximum FSAR Method pressure expected is 2540 psia.

Result

l s

Table 4 l Summaty of Qne RCP Cmtsidown Reanalysis Acceptance 1.18 MDNBR using BWC correlation Criteria FSAR LYNXT Method FSAR 2568 MWt Assumptions 108% rated power 106.5% rated flow 2135 psia 555.9 F T.

Flow coastdown to 75% rated flow FSAR 1.6 MDNBR Results TSCR LYNXT with delayed initiation of the transient heat flux profile.

Analysis Method TSCR Original FSAR Assumptions. ,

Assumptions 1.4 seconds trip delay.

TSCR 1.58 MDNBR Analysis Results Why is it 1.4 instead of 1.34 second rod trip delay.

Conservative Fixed trip delay instead of slowed reactivity insertion.

Actual 3 RCP flow is greater than 75%.

Expected Same.

FSAR Method Result i

l

s.

Table 5 Summary of Four Pump Coasjdown Reanalysis Acceptance 1.18 MDNBR using BWC correlation Criteria 1.32 MDNBR using BAW-2 correlation FSAR LYNXT Method 2 second transient FSAR 2568 MWt Assumptions 102% rated power 103.5% rated flow 2135 psia 557.3 F T, Trip of RCP Status FSAR 1.75 MDNBR using BAW-2 correlation Results TSCR LYNXT with delayed initiation of the transient heat flux profile.

Analysis Linear extension of flow coastdown curve to 3.64 acconds based on Method coastdown rate from 1.9 - 2.0 seconds.

TSCR Original FSAR Assumptions.

Assumptions Linear extrapolation of flow coastdown curve.

TSCR 1.55 MDNBR using BWC correlation Analysis Results Why is it Actual flow coastdown has a decreasing slope.

Conservative Fixed trip delay instead of slowed reactivity insertion.

Expected Same.

FSAR Method Result

t Table 6 Summary of Locked Rotor Event Reanalysis Acceptance 1.18 MDNBR using BWC correlation Criteria FSAR Statepoint calculation using BWC correlation and LYNXT.

Method No power reduction is credited.

FSAR 2568 MWt Assumptions 102% rated power 106.5% rated flow 2135 psia 557.3 F T, MTC of 0.0%AK/K/F State points: 75% rated flow, initial power.

FSAR 1.43 MDNBR Results TSCR LYNXT Analysis Method TSCR Same as FSAR.

Assumptions TSCR 1.43 MDNBR (the result of the analysis of record)

Analysis Results Why is it Actual 3 RCP flow is greater than 75%.

Conservative Pressure increase is conservatively neglected.

The MTC for the balance of Cycle 10 is increasingly negative.

Expected Same - not affected by trip delay time.

FSAR Method Result

e Table 7 Summary of Rod Ejection Accident Reanalysis Acceptance 200 cal /gm 10CFR100 Dose Limits Criteria FSAR KAPPB Method FSAR BOL Assumptions 100%FP 0.65%AK/K ejected rod worth

-0.9E-5AK/K/F Doppler 0.0%AK/K/F MTC Adiabatic heatup determine fuel enthalpy.

FSAR 180 cal /gm 17.5% pins in DNB Results 4.43 rem-2hr Thyroid dose TSCR Original KAPPB power vs time results for 0.40AK/K ejected rod worth.

Analysis Extrapolated pre-trip power ramp for 1.34 sec.

Method Normalized power response to the KAPPB power response at time of scram.

Integrated power using trapezoidal rule for delayed scram.

Aver ga fuel temperature from TACO 3 results at maximum LOCA LHR with 12% uncertainty factor.

Initial fuel enthalpy from Bureau of Mines enthalpy as a function of fuel temperature equation.

Determined total peaking factor based on the maximum LOCA LHR and 102% FP.

Calculated peak fuel enthalpy from initial enthalpy and integrated power from delayed scram increased by the total peaking factor.

TSCR Original SAR KAPPB run results used.

Assumptions TACO 3 results of fuel temperature based on maximum LOCALHR.

Adiabatic heatup determines fuel enthalpy.

Maximum Cycle 10 ejected rod worth at HFP is 0.24%AX/K.

Least negative Doppler for Cycle 10 is -1.47 E SAK/K/F.

0.40Ak/k ejected rod worth conservatively bounds Cycle 10 values.

TSCR 193 cal /gm 19.3% pins in DNB Analysis 4.43 rem + 10% -2hr Thyroid dose Results Why is it MTC for the balance of Cycle 10 is increasingly negative.

Conservative Non-trip enthalpy deposition for 1.34 sec beyond FSAR trip.

Expected KAPPB: <193 callgm KAPPB: < 19.3% pins in DNB FSAR Method BWKIN: <193 cal /gm BWKIN: <10% pins in DNB Result Lower dose consequences

e

/ '

Tabin 8 Summary of Loss of Feidwat~r Accidsnt Reanalysis Acceptance Pressurizer does not go water solid.

Criteria 2765 psia FSAR RELAPS-MOD 2 Analys;s Method FSAR 1.02 x 2568MWt Assumptions 220" pressurizer level on 400" range FSAR 2644.6 psia Results TSCR RELAPS/ MOD 2 analysis.

Analysis Extrapolation of pressurization results for 1.34 sec rod delay.

Method TSCR 1.02 x 2568 Mw(t)

Assumptions FW coastdown of 7 sec RPS trip at 2410 psia (14.3 sec)

PSV 70% open at 1% accumulation,100% open at full accumulation (3%)

300,000lbrn/hr per PSV TSCR 2720.1 psia Analysis Results Why is it Rated capacity of TMI-1 safety valves is 623,400 lbm/hr - total for 2 Conservative valves.

Fixed trip delay instead of slowed reactivity insertion.

The large PSV accumulation produces the highest RCS pressure response.

Expected Same FSAR Method Result I

i l

1 l