ML19260A028

Tech Specs Change Request 36 to Amend DPR-50,App a to Incorporate Fuel Densification Penalty & Eliminate Internal Vent Valve Bypass Flow Penalty Per NRC 760310 Ltr

ML19260A028
Person / Time
Site:	Three Mile Island
Issue date:	07/07/1976
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19260A024	List: ML19260A023 ML19260A028
References
NUDOCS 7910290663
Download: ML19260A028 (11)

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENUSYLVANIA ELECTRIC COMPANY THREE MI'.2 ISLAND NUCLEAR STATION UNIT 1 Operating License ho. DPR-50 Dochet No. 50-289 Technical Snecification Change Beauest No. 36 This Technical Specification Change Request is submitted in rapport of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

METROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President-Generation Sworn and subscribed to me this 7th day of July ,

1976.

Lawrence L. Lawyer Notary Public 1479 017

1. 910 290f63

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Metropolitan Edison Company (Met-Ed)

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

Docket No. 50-289 Operating License No. DPR-50 Technical Suecitication Change Recuest No. 36 The Licensee request 3 that the attached pages replace the corresponding existing Technical Specificatiors pages.

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

a. The Juel 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.

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

Thus, elimination of the internal vent valve flow penalty vill 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 Commission required in its letter of March 10, 1976, "... testing to be 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 proposed 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 flow.

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 has been retracted.

Safety Analysis Justifying Proposed Change Elimination of the vent valve flow penalty has been authorized by the Commission.

Revised densification analysis indicates that the coriect penalties are 5.93% DNBR (versus 1.88% in the Reload Report) and 3.h7% power peaking relative to DNBR (versus 1.06% quoted in the Reload Report).

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1 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 2.13 of the present Technical Specifications (Curve 1).

Curves 2 and 3 represent the corresponding limits for 3 and 2 pump operation, respectively. Each curve is based on the assumption that the reac tor is operating at the maxiwum achievable power level for that pump operating condition. In tbc ariginal cycle 2 submittal (and in the cycle one technical specifications), Curve 1 incorporated the open vent valve penalty, while curves 2 and 3 did not. That is, the four pump limit curve was based upon operation with one vent valve open while the three and two pump limit curves assumed all vent valves remained shut. In revising Figure 2.1-3 to incorporate the corrected DNBR densification penalty, the basis for the four pump limit curve was changed to eliminate the vent valve penalty. The combined effect was to move curve 1 to the right (i.e., at 2185 PSIG, Tout was 621.2F and is nov 62h.hF). The revised curves 2 and 3 incorporated only the increased densification penalty, therefore, the curves moved slightly to t he left (i.e. , for curve 2 at 2185 PSIG, Tout was 625.3F and is now 6?3.2F).

The flux / flow trip setpoint for cycle 2 (1.08) is based on the one pump coastdown analysis. When the revised densification penalty is incorporated and the vent valve penalty is 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 also be shown that a thermal-hydraulic limit of 1.11 on the flux / flow setpoint can be justified by taking credit for 1/2 of the vent valve penalty.

The error found in the TMI-1, Cycle 2 DNER densification penalty calculations resulted from the use of inconsistent heat flux (flux shape) and enthalpy rise in evaluating the DNBR densification penalty. This error only affects the PT envelope and flux / flow ratio.

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

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TABLE OF CONTENTS Section Page 4.6 E4ERGENei POWER SYSTB4 PERIODIC TESTS h-h6 h.7 REACTOR CONTROL ROD SYSTH4 TESTS h-h8 h.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS h-48 4.7.2 CONTROL RCD PROGRAM VERIFICATION h-50 h.8 MAIN STEAM ISOLATION VALVES h-51 h.9 EMERGFNC FEEDWATER PUMPS PERIODIC TESTING 4-52 h.9.1 TEST ~ v -52 h.9.2 ACCEPTANCE CRITERIA h -:, -

k.10 REACTIVITY ANOMALIES h-53 h.11 SITE ENVIRONMDITAL RADIOACTIVITY-SURVEI h-Sh h.12

' CONTROL ROOM FILTERING SYSTEM ~ h-55 4.12.1 OPERATING TESTS 4-55 4.12.2 FILTER TESTS h-55 h.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE h-56 h.lh REACTOR BUILDING PURGE EXFAUST SYSTEM h-57 4.15 MAIN STEAM SYSTD4 INSERVICE INSPECTION h-58 h.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 5 DESIGN FEATURES 5-1 5.1 SITE 5-1 52 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTE4 5-3 5.3 REACTOR 5-h 5 3.1 REACTOR CORE 5-4 5.3.2 REACT 01 COOLANT SYSTEM 5-h 5.h NEW AND SFENT FUEL SOTRAGE FACILITIES 5-6 5.h.1 NEW FUEL STORt.GE 5-6 5.k.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTD4S 5-8 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6.2 6-1 ORGANIZATION 6-2 6.2.1 0FFSITE 6-2 6.2.2 FACILITY STAFF 6-2 6.3 STATION STAFF QUALIFICATIONS 6-3 6.h TRAINING 6-3 6.5 REVIEW & AUDIT 6-3 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) 6-3 6.5.2.A MET-ED CORPORATE TECHNICAL SUPPORT STAFF 6-5 6.5.2.B GENERAL OFFICE REVIEW BOARD (GORB) 6-7 6.6 REPORTABLE OCCURRENCE ACTION 6-10 6.7 OCCURRENCES INVOLVING A SAFETY LIMIT VIOLATION 6-10a 6.8 PROCEDURES 6-11 m

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The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a

-avim"' permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and lov flow rate combinations for the pu=p situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 108 percent and reactor flow rate is 100 percent,

.or flow rate is 92.6 percent and power level is 100 percent.

2. Trip vculd occur when three reactor coolant pumps are operating if power is 80.7 percent and reactor flow rate is Th.7 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 h9.2 percent or flow rate is h5.h percent and the power level is h9 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 penalty in reactor coolant 1 ov through the core was taken for an open core vent valve because of the sore vent valve surveillance program during each refueling outage.

For safety analysis calculstions the maximum calibration and instrumentation errors for the power level vere used.

The power-imbalance boundaries are established in order te prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNER limits. The reactor power i= balance (power in the top half of core minus power in the bottom half of core) reduces the power 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 power level trip and associated reactor pover/ reactor power-imbalance boundaries by 1.08 percent for a one percent flow reduction.

b. Pump monitors The redundant pump monitors prevent the mini:mmt core DNBR from decreasing below 1.3 ty tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

c. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpover 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.

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The low pressure (1800 psig) and variable low pressure (11379 Tout - h91h) trip setpoint shown in Figure 2.3-1 have been established to maintain the

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DNB ratio greater than or equil to 13 for those design accidents that result in a pressure reduction (3, 4).

Due to the calibration and instrumentation errors, the safety analysis used a variable low reactor coolant system pressure trip value of (11 379 Tout - h95h).

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d. Coolant out1ct te=perature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperatures 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 is 619 F. Under the worst case environment, power supply perturbations, and drift, the accuracy of the trip string is+1F. This accuracy was arrived at by su==ing the worst case accuracies of each module. This is a conservative method of error analysis since the nor=al procedure is to use the root mean square method.

Therefore, it is assured that a trip will 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 oppration within the calibrated range. Additional testing has de=onstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated range.

Since it has been established that the channe: vill trip at a value of RC outlet temperature no higher than 620F even in the worst case, and since the chhnnel is fully operational approximately 10% above the calibrated range and exhibits no hystere-is or foldover characteristics, it is concluded that the instrument design is acceptable.

e. Reactor building pressure The high reactor building pressure trip setting limit (k psig) provides positive assurance that a reactor trip will 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 system pressure trip.

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TABLE 2.3-1 ~

• 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 Shutdown Power - 100%) Power - 75%) Operating Power - 49%) Bypass

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

% of rated power

2. Nuclea Power based on 1.08 times flow minus 1.08 times flow minus 108 times flow minue Bypassed flov (2 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 Igonpumpmonitors 91% Bypassed

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max. % of rated power

4. I!igh reactor coolant 2355 b) 2355 2355 1720 y system pressure, psig, m max.

5 Low reactor coolcnt 1800 1800 1800 Bypassed system pressure, psig min.

6. Variablu low reactor (11 379 Tout-h914 ) (1) (11379 Tout-hRh ) (1) ll.379 Tout 4914 ) (1) Bypassed coolant system pressure psig, min. g 7 Reactoc coolant temp. 619 619 619 F., Max. 619

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4 8. High Reactor Building 4 4 4 4 N pressure, psig, max.

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(1) Tout is in degrees Fahrenheit (F)

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CD (2) Reactor coolant system flow, %

N U (3) Administratively controlled reduction set only during reactor shutdown (1 ) Automatically set when other segments of the RPS (as specified) are bypassed 4

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

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k.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE Atelicability 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.

Specification 4.16.1 At intervals not exceeding the refueling interval, each reactor internals vent valve will be tested to verify that no valve is stuck in the open position and that each j valve continues to exhibit freedom of movement. i i

Bases Verifying vent valve freedom of movement insures that coolant flow dues not bypass the core through reactor internals vent valves during operation 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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2600 2400

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E a

5 m

2200 E Acceptable

[ Operation

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[m 2000 /

8 Unacceptable Operation 1600 560 580 600 620 640 6;d Reactor Outlet Temperature, F

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UNIT 1, CYCLE 2 CORE PROTECTION SAFETY LIMIT Figure 2.1-1 1479 025

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2500 g

P = 2355 psig T = 619F

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.a 2300 '

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[ Acceptable

{ Operation U k

[ 2100

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5o g

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8 %g 1900 # '

g 3 Unacceptable Operation; P = 1800 psig

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1700 1500 560 580 600 620 640 000 Reactor Outlet Temperature, F UNIT 1, CYCLE 2 PROTECTION SYSTEM MAXIMUM ALLOWABLE SET POINTS

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Figure 2.3-1 -

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2600 2400

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5 2200 i

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/D 2000 /

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1600 560 580 600 620 640 660 Reactor Outlet. Temperature, F REACTOR COOLANT FLOW CURVE (LB/HR) POWER PU APS OPERATING (TYPE OF LIMIT) i 139.8 x 106 (1005)* 1 1 ?.", Four Pumps (DNBR Limit) 2 104.5 x 106 (74.75) 86.7% Three Pumps (DNBR Limi t) 3 68.8 x 106 (49.25) 5 9 . 1 ", One Pump in Each loop (Quality limi

• 106.5% of Cycle i Design Flow UNIT 1, CYCLE 2 1479 027 CORE PROTECTION SAFETY BASES Figure 2.1-3