ML19270H572

Draft Tech Spec Change Request 36 Re Elimination of Reactor Interval Vent Valve Flow Penalty & Incorporation of Correct Fuel Densification Penalty

ML19270H572
Person / Time
Site:	Three Mile Island
Issue date:	06/18/1976
From:	METROPOLITAN EDISON CO.
To:
Shared Package
ML19270H570	List: ML19270H569 ML19270H572
References
NUDOCS 7911110183
Download: ML19270H572 (10)

Text

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Metropolitan Edison Company (Met-Ed)

Three Mile Island Nuclear Staticn Unit 1 (TMI-1)

Docket No. 50-269 Operating License No. DPR-50 Technical Snecification Change Request !!o. 36 The Licensee requests that the attached pages replace the corresponding existing Technical Specifications pages.

Reasons for Proposed Change On June h, 1976, it was discovered that:

a. The fuel densification penalty was not properly incorporated into technical specifications prepared for cycle 2.

b. Proper incorporation of this penalty would affect DNB based pressure-temperature limit curves such that they would be more restrictive.

c. Babcock & Wilcox calculations confirmed that elimination of the internal vent valve bypass flow penalty, as authorized by Nuclear Regulatory Commission letter of March 10, 1976, would more than compensate for this error.

Thus, elimination of the internal vent valve flow penalty will allow continued use of present pressure temperature curves until revised curves included are authorized.

As a prerequisite for eliminating the vent valve flow penalty, the Ccemission required in its letter of March 10, 1976, "... testing to te conducted each refueling outage to confirm that no vent valve is stuck in an open position and that each valve continues to exhibit complete freedom of movement." This surveillance requirement was performed during the last refueling outage. This prpposed change incorporates this surveillance requirement into technical specifications, as well as revised figures 2.1-1, 2.1-3, and 2.3-1 which include credit for elimination of vent valve bypass flev.

Note: The proposed technical specification h.16 included in Technical Specification Change Request No. 13 (still under review) is no longer needed, due to equipment modification. Therefore, Technical Specification Change Request No. 13 vill be retracted.

Safety Analysic Justifyinc Prorosed Change ) II}

Elimination of the vent valve flow penalty has been authorized by the Commission.

Revised densification analysis indicates that the correct penalties are 5 93% DNER (versus 1.885 in the Reload Report) and 3.h75 pcver peaking relative to DNER (versus 1.C65 quoted in the Reload Report).

The variable low pressure trip setpoint for cycle 2 operation is based on the four pump open vent valve pressure-temperature limit curve presented in figure 8.3 of the Relcad Report. The corresponding limit curve, based 7 911110 / J

on cloced vent valves and incorporating the reviced dencification penalty, vill be approximately 3F lecc restrictive (max. allowable To will be 3F higher at a given prescure).

The flux / flow trip setpoint for cycle 2 (1.08) is baced on the one pump coastdown analysis. '4 hen the revised dencification penalty is incorporated and the vent valve penalty ic eliminated, the thermal-hydraulic limiting flux / flow setpoint is greater than 1.12 (this limit must be at least 1.11 to justify the tech spec setpoint of 1.08). It can alco be chown that a thermal-hydraulic limit of 1.11 on the flux / flow cetpoint can be justified by taking credit for 1/2 of the vent valve penalty.

Baced upon the above, it is determined that this change does not conctitute a threat to the health and safety of the public, nor does it involve an unreviewed safety question.

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The power level trip set point produced by the power-to-flc ratio provides both high power level and lov flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The pcuer level trip set point produced by the pcVer to flow ratio provides overpower DI!B protection for all =cdes of purp operaticn. For every flow rate there is a naximum permissible power level, and for every pcVer level there is a minimum permissible lov flow rate. Typical power level and lov flow rate combinations for the purp situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pu=ps are cperating if power is 1C8 percent and reactor flow rate is 100 percent, or flow rate is 92.6 percent and power level in 100 percent.

2. Trip would occur when three reacter coolant pumps are operating if power is 80.7 percent and reactor flow rate is Th.T percent or flow rate is 69.2 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.9 percent and reactor flow rate is L9.2 percent or flow rate ic h5.h percent and the power level is 49 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

No pensity in reactor coolant flow through the core was taken for en open core vent valve because of the core vent valve surveillance progre.m during each refueling outage.

Fcr safety analysis calculations the maximum calibration and instrunentation errors for the power level vere used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from teing exceeded. These thernal limits are either power peaking kW/ft limits or DIGR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the pcVer level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the pcver level trip and associated reactor pover/ reactor pover-imbalance boundaries by 1.08 percent for a one percent flow reduction.

b. Pump monitors The redundant pu=p monitors prevent the mini =tm core D:ER frca decreasing belov 1.3 by tripping the reactor due to the loss of reactor coolant purp(s). The pump monitors also restrict the power level for the number of pumps in operation.

c. Reactor coolant system pressure 9701 L . ,

1F7

.J During a startup accident from 1cv power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpcuer trip set point. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.

2-6

8 The 1cv pressure (1800 psig) and variable low pressure (ll.379 Tout - h91h) trip setpoint shown in Figure 2.3-1 have been established to =aintain the D:iB ratio greater than or equil to 13 for those design accidents that result

{- in a presaure reduction (3, h).

Due to the calibration and instru=entation errors, the safety analysis used a variable lcv reactor coolant systen pressure trip value of (11379 Tout - h954 ).

l

d. Coolant outlet te=perature The high reactor coolant outlet te=perature trip setting li=1t (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant te=peratures in the operating range.

The calibrated range of the temperature channels of the RPS is 520 to 620 F. The trip setpoint of the channel _a 619 F. Under the vorst case environment, power supply perturbations, and drift, the accuracy of the trip string ic+1F. This accuracy was arrived at by summing the worst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square =ethod.

Therefore, it is assured that a trip vill occur at a value no higher than 620F even under vorst case conditions. The safety analysis used a high temperature trip set point of 620F.

The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, linearity,

( repeatability, etc. This does not imply that the equipment is restricted to cppration within the calibrated range. Additicnal testing has demonstrated that in fact, the temperature channel is fully operaticnal approximately 107, above the calibrated range.

Since it has been established that the channel vill trip at a value of RC outlet tc=perature no higher than 620F even in the vorst case, and since the channel is fully operational approxicately 10f, above the eglibrated range and exhibits no hystere-is or foldover characteristics, it is concluded that the instrument design is acceptable.

c. Reactor building pressure The high reactor building pressure trip setting limit (h psAg) p provides positive assurance that a reactor trip vill occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant syste= pressure trip.

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.m ,.* m m TABLE 2 3-1 -

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REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS

, Four Reactor Coolant Three Reactor Coolant One Reactor Coolant Pumps Operating Pumps Cperating Pump Operating in (Nominal Operating .(Nominal Operating Each Loop (Nominal Shutdovn Power - 100%) Power - 75%) Operating Power - h9%) Bypass

1. Nuclear power, Max. 105.5 105 5 105 5 5 0 (3)

% of rated power

2. Nucleap Power based on 1.08 times flow minus 1.08 times flow minus 108 times flow minue Bypassed flov (21 and imbalance reduction due to reduction due to reduction due to max. of rated power imbalance (s) imbalance (s) imbalance (s) 3 lear power based NA NA 91% B3 passed Ivgonpumpmonitors, max. % of rated power

4. High reactor coolant 2355 2355 2355 1720 Ch) y system pressure, psig, M) max.

5 Low reactor coolant 1800 1800 1800 Bypassed system pressure, psig ,

min.

6. Variable low reactor (11 379 Tout-h914) (1) (11.379 Tout-h914) (1) 11.379 Tout h91h ) (1) Bypassed coolant system pressure psig, min.

g

[j 7 Reactor coolant temp. 619 619 619 619 7g F., Max, r>

8. High Reactor Building h h h 4

- -- pressure, psig, max.

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U1 (1) Tout is in degrees Fahrenheit (F)

(2) Recctor coolant system flow, %

(3) Administrative 1y controlled reduction set only during reactor shutdown (1 ) Automatically set when other segments of the RPS (as specified) are bypassed 1

(5) The pump monitors also produce a trip on: (a) loss of two reactor coolant pu=ps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operatien.

h.16 REACTOR INTEPNALS VENT VALVES SURVEIL:1.UCE Anolicability Applies to Reactor Internals Vent Valves.

Objective To verify that no reactor internals vent valve is stuck in the open position and that each valve continues to exhibit freedom of movement.

Scecification h.16.1 At intervals not exceeding the refueling interval, each reactor internals vent valve vill be tested to verify that no valve is stuck in the open position and that each valve continues to exhibit freedom of movement.

Bases Verifying vent valve freedom of movement insures that coolant flow does not bypass the core through reactor internals vent valves during operaticn and therefore insures the conservatism of Core Protection Safety limits as delineated in figures 2.1-1 and 2.1-3, and the flux / flow trip setpoint.

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