ML20138A982

Proposed Tech Specs,Changing Main Steam Line High Flow Setpoint for Primary Containment Isolation to 150% of Rated Steam Flow

ML20138A982
Person / Time
Site:	Cooper
Issue date:	03/11/1986
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
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ML20138A974	List: ML20138A969 ML20138A982
References
NUDOCS 8603200230
Download: ML20138A982 (2)

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j COOPER NUCLEAR STATION .

TABLE 3.2.A (Page 1)

PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION Minimum Number Action Required When Instrument of Operable Components Component Operability Instrument I.D. No. Setting Limit. Per Trip System (1) is Not Assured (2)

Main Steam Line High RMP-RM-251, A,B,C,&D 1 3 Times Full Power 2 A or B Rad.

Reactor Low Water Level NBI-LIS-lui. A,B,C,6D >+12.5" Indicated Level 2(4) A or B l

Reactor Low Low Water NBI-LIS-57 A & B #2 >-37" indicated Level 2 A or B Level NBI-LIS-58 A & B #2 Reactor Low Low Low Water NBI-LIS-57 A & B-#1 >-145.5" Indicated Level 2 . A or B Level NBI-LIS-58 A & B #1 I

I Main Steam Line Leak MS-TS-121 A,B,C,6D < 200*F

_ 2(6) B Detection 122, 123, 124, 143, 144, di 145, 146, 147, 148, 149,

? 150

Main Steam Line liigh MS-dPIS-Il6 A,B,C,6D < 150% of Rated Steam 2(3) B l i Flow 117, 118, 119 Flow Main Steam Line Lew MS-PS-134, A,B,C,6D > 825 psig 2(5) B l Pressure

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} liigh Drywell Pressure PC-PS-12, A,B,C,6D < 2 psig 2(4) A or B liigh Reactor Pressure RR-PS-128 A & B < 75 psig 1 D a Main Condenser Low' MS-PS-103, A,B,C,&D > 7" lig (7) 2 A or B Vacuum Reactor Water Cleanup RWCU-dPIS-170 A & B < 200% of System Flow 1 C System liigh Flow 8603200230 860311 i PDR ADOCK 05000298' 4 P PDR I

3.2. BASES

(Cont'd) and the guidelines of 10CFR100 will not be exceeded. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system isolation are initiated in time to meet the above criteria. Reference Paragraph VI.5.3.1 USAR.

The high drywell pressure instrumentation is a diverse signal for mal-tunctions to the ':ater level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 and 6 isolation valves. For the breaks discussed above, this instrumentation-will generally initiate CSCS operation before the low-low-low water level instrumentation; thus the_

results given above are applicable here also. The water level instrumen-tation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel i during a steam line break accident. The primary function of the instru-mentation is to detect a break in the main steam line. For the worst case of accident, main steam line break outside the drywell, a trip setting of 150% of rated steam flow in conjunction with the flow limiters l 1 and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel clad temperatures peak at approximately 1000*F and release of radioactivity to the environs is below 10CFR100 guidelines. Reference Section XIV.6.5 USAR.

Temperature monitoring instrumentation is provided in the main steam tunnel and along the steam line in the turbine building to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause' closure of isolation valves. See Spec. 3.7 for Valve Grcup. The setting is 200*F for the main steam leak detection system.

For large breaks, the high steam flow instrumentation is a backup to the temp. instrumentation.

High radiation monitors in the main steam tunnel have been provided to detect gross fuel failure as in the control rod drop accident. With the established setting of 3 times normal background, and main steam line

, isolation valve closure, fission product release is limited so that 10CFR100 guidelines are not exceeded for this accident. Reference Sec-tion XIV.6.2 US AR.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below Speci-fication 2.1.A.6. The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the turbine bypass valves when not in the RUN Mode is less severe than the loss of feedwater analyzed in Section XIV.5 of the USAR, therefore, closure of the Main Steam Isolation valves for j thermal transient protection when not in RUN mode is not required.

1 The Reactor Water Cleanup System high flow and temperature instrumentation are arranged similar to that for the HPCI. The trip settings are such that core uncovery is prevented and fission product release is within limits.

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MDE-43-0386 March 1986 I DRF B21-00288 SAFETY EVALUATION OF MAIN STEAM LINE HIGH FLOW SETPOINT FOR COOPER NUCLEAR STATION Prepared by:

L. L. Chi, Senior Engineer Application Engineering Services Revie.ted by: _

J. R. Pobre, Principal Engineer Plant Piping Design Reviewed by: f f>6 ,

R. R. Ghosh, Lead System Engineer Reactor Systems Design Reviewed by: a.

p L! Rash, Principal Engineer Licensing Services Approved by: IM C4 L. Sozii, Mdia&dr Application Engineering Services  :

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GENERAL $ ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL ELECTRIC COMPANY

• 175 CURTNER AVENUE

• SAN JOSE, CAUFORNIA 95125

-960-3 F703 30 T "-

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IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please read Carefully The only undertakings of General Electric Company respecting inform-ation in this document are contained in the contract between the customer and General Electric Company, as identified in the purchase i

order for this report and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the customer or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric' Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this docu-ment.

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TABLE OF CONTENTS

f. age i

1. Introduction 1

2. Basis for MSL High Flow Setpoint 2

3. Objective of Safety Evaluation 3

4. Safety Evaluation 4 i

5. Impact on USAR 5

6. References 6 i

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1. INTRODUCTION The purpose of the main steam line (MSL) high flow instrument is to detect large steam line breaks and isolate the reactor thereby limiting the potential for radioactive release outside the containment. The MSL high flow is one of the various methods to isolate the reactor in case of a MSL break outside the containment. The MSL high flow is primarily for detection of large breaks. For small breaks or steam leak, other

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methods of isolation are used. Other methods of isolation for Cooper Nuclear Station (CNS) include temperature sensors and radiation monitors in the main steam tunnel, . low steam line pressure, and low reactor watar level (Level 1). The effectiveness of the other isolation methods depends on the size of the break and, for area dependent monitors, on the location of the monitors.

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2. ~ BASIS FOR MSL HIGH FLOW SETPOINT The MSL flow rate is sensed from a differential pressure instrument in the flow velocity limiter (venturi) in each steam line. The maximum flow loss from a steam line break is limited by the venturi in each steam line. This maximum flow is approximately 200% of nuclear boiler rated (NBR) steam flow. The maximum setpoint for the MSL high flow trip is therefore less than 200% of NBR steam flow. The minimum setpoint

- must be above 100% of NBR steam flow to avoid spurious trips and allow for continuous operation.' A typical value for a BWR/4 is 140% of NBR steam flow. This setpoint would allow on line testing of the mafn steam isolation valves (MSIVs) since closing one valve would result in approximately 133% of NBR steam flow in the remaining three lines.

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3. OBJECTIVE OF SAFETY EVALUATION

, The Technical Specifications for CNS specifies that the setpoint for the differential pressure instrument should be less than that equivalent to 140% of NBR steam flow (Reference 1). However, the differential pressure setpoint was originally established as 118 psid for each of the four steam lines which corresponds to approximately 150% of NBR steam flow. This apparent anomaly is described in more detail in Section 5 of

. this report. Also, this data applies to an ideal steam line configurat-ion where a differential pressure of 48.86 psid is calculated for the 100% NBR steam flow (Reference 2). Due to variation of the steam line configurations, the differential pressures for the four MSLs at CNS have a spread of 48 psid for MSL C to 56 psid for MSL i during normal full power operation. The differential pressure for MSL C is in agreement with the theoretical calculation. This means that the 118 psid setpoint would be considerably less than 150% of NBR steam flow for the other steam lines. The setpoint of 118 psid accounts for such variations of the affects of steam line configurations and is considered as a reason-able setpoint for CNS. Nevertheless, this setpoint is in apparent violation with the plant Technical Specifications. Therefore, a safety evaluation was performed to determine the safety implication of allowing the MSL high flow setpoint to remain at approximately 150% instead of 140% of NBR steam flow.

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4. SAFETY EVALUATION The variation of 'the MSL high flow setpoint from 150 to 140% of the NBR steam flow would not affect the analysis for a guillotine break of the MSL assumed in Section XIV-6 of the Updated Safety Analysis Report (USAR) for CNS (Reference 3). Thus, the safety implication of the setpoint of 150% of rated flow is limited to the difference in ability to detect a break between 150% to 140%. As noted in Reference 3, a setpoint of 140% of rated steam flow would detect steam line breaks

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greater than 0.3 ft 2. The dose release for such a break is only 2.7x10 ~7 of that allowed by 10CFR100. A setpoint of 150% of NBR steam flow would detect steam line breaks greater than 0.38 2ft . Since all breaks below 0.3 ft 2are now being detected by other sensors, high temperature, etc, a high flow setpoint of 150% for CNS would increase i the break detection requirement of the other sensors from 0.3 to 0.38 ft , a mere increase of 0.08 ft . The increase in break size detectable by the high flow instruments does not necessarily affect the ability of the other sensors to detect and isolate the break. The response time for the other sensors would decrease because of the larger break area.

The larger break area would result in larger break flow and depressurize the main steam lines faster. The temperature sensors would also respond faster. If feedwater and other high pressure makeup systems are not available, the low water level isolation setpoint would also be reached earlier with a larger break. A conservative evaluation shows that the difference in total dose release between the two break sizes is approximately 10%. This 10% increase in the extremely small dose calculated for the 140% setpoint will not significantly change the existing margin for the limits allowed by 10CFR100. Therefore, the ability of the plant to detect and isolate a MSL break outside the containment would not be affected by the equivalent setpoint of 150% of NBR steam flow (118 psid) for the MSL high flow. Therefore, the 118 psid setpoint does not present a safety concern and does not have any safety implication.

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5. IMPACT ON USAR The setpoint for the MSL high flow for CNS was set at 150% of NBR steam flow versus 140% because the design of each venturi was based on 105% of i

NBR steam flow. That is, the rated steam flow of the venturi was designed to 105% of NBR and the high flow setpoint became 140% of 105%

of NBR steam flow (1.4 X 1.05 X NBR) which resulted in the 150% setting.

4 The value of 105% of NBR steam flow has been used as a conservative basis for some safety analyses and for consistency with the. rating of l

the turbine which is typically at 105% of NBR.

For the postulated event of MSL break outside the containment, the venturi will limit the break flow through the broken steam line. The maximum flow through the venturi is approximately 200% of rated steam flow. As described above, however, the designed flow is actually 200%

of 105% of NBR steam flow. This raises a concern to the basis of the "200% of rated steam flow" assumed.in the safety analysis in Reference-

3. For each of the four venturis, the 100% of NBR steam flow for CNS is 0 lb/hr or 664 lb/sec, 105% of NBR is 2.51x10 6 2.39x10 lb/hr or 697 6

lb/sec, 200% of NBR is 4.78x10 lb/hr or 1328 lb/sec, and 200% of 105%

6 of NBR is 5.02x10 lb/hr or 1393'lb/sec. The MSL break outside containment analysis in Reference 3 conservatively assumes that the maximum break flow used is 1500 lb/sec which is more than 200% of either the NBR steam flow or the 105% of NBR steam flow. Therefore, the safety analysis for the MSL break outside the containment was performed with a

conservative flow rate which is consistent with the actual size of the

venturi. This demonstrates that sizing the venturi at 105% of NBR steam flow is not an unreviewed safety question.

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6. REFERENCES

1. " Cooper Nuclear Station Lican'se and Technical Specifications",

Docket 50-298, DPR-46.

2. " Flow Element Components".. General Electric Company, Design Specification Data Sheet, 21A1058AT, Rev. O.

3. " Nebraska Public Power District Cooper Nuclear Station Updated Safety Analysis Report", Section XIV-6,' Docket 50-298.

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