ML20213D777

RCS Leakage Detection Assessment for Elimination of RCS Main Loop Pipe Break Protective Devices

ML20213D777
Person / Time
Site:	North Anna
Issue date:	10/31/1986
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML19292G230	List: ML20213D730 ML20213D760 ML20213D777 ML20213D792 ML20213E055
References
NUDOCS 8611120229
Download: ML20213D777 (16)

Text

ATTACHMENT 4 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNITS 1 AND 2 REACTOR COOLANT SYSTEM LEAKAGE DETECTION ASSESSMENT FOR ELIMINATION OF REACTOR COOLANT SYSTEM MAIN LOOP PIPE BREAK PROTECTIVE DEVICES OCTOBER 1986 8611120229 861106 DR ADOCK 0500 8 60-KKD-4699B

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- INTRODUCTION 2 I I'. REGULATORY ~ POSITION 3 r . III.

SUMMARY

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- IV. REFERENCES. '13 L

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- I. INTRODUCTION

- This report is submitted in support of Virginia Electric and Power Company's request in compliance ~ to General Design Criteria 4 (GDC-4) for an amendment to license NPF-4 and NPF-7 for North Anna Units 1 and 2, to exclude protection'against.the dynamic effects of postulated main reactor coolant pipe rupture from the design basis of primary system components / supports:and piping. The technical basis for this

-licensing amendment is based upon the fracture mechanics analyses referred to as " leak-before-break."

NRC Generic Letter 84-04 provided the NRC safety evaluation of the submittal of Westinghouse's owners group on Unresolved Safety Issue A-2 and concluded that an acceptable technical basis exists so that the blowdown loads from double-ended pipe breaks in the reactor coolant system (RCS) primary loop need not be considered in the design basis, provided that certain conditions can be met. One condition is that an acceptable capability exists to detect RCS leakage. North Anna was not included in the plants reviewed under A-2. owners group.

But individual plants were encouraged to submit similar plant -

specific fracture mechanics studies. General Design Criteria-4 was revised. effective May 12, 1986 to utilize leak before break technology to exclude dynamic effects of postulated break of main reactor coolant-loop. Use of leak before break technology requires an assessment of leakage detection system. The guidance for evaluation of leakage detection system is to be taken from NUREG 1061, Volume 3, which emphasizes an assessment with respect to Reg. Guide 1.45.

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An' assessment of the North Anna Units 1 and 2 reactor coolant leakage detection system is the subject of this report. It compares the North Anna leakage detection systems with the nine Regulatory Guide 1.45 design guidelines and considers the exceptions given in NRC Generic Letter 84-04. This evaluation demonstrated that these leakage detection systems are capable, with margin, of detecting leakage from the postulated through-wall flaws.

II. REGULATORY POSITION NRC Generic Letter 84-04 and NUREG 1061 Vol. 3 states that, " Leakage detection systems should be demonstrated to be sufficient to provide adequate margin to detect the leakage from the postulated through-wall flaw utilizing the guidance of Regulatory Guide 1.45," Reactor Coolant Pressure Boundary Leakage Detection Systems, "with the exception that the seismic qualification of the airborne particulate radiation moni-tor is not necessary. At least one leakage detection system with a sensitivity capable of detecting 1 gallon per minute (gpm) in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> be operable."

Regulatory Guide 1.45 states that "The source of reactor coolant leak-age should be identified to the extent practical. Reactor coolant pressure boundary leakage detection and collection systems should be selected and designed to include the following:"

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Design Guideline 1:

" Leakage to the primary reactor containment from identified sources should be collected or otherwise isolated so that:

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a. the flow rates are monitored separately from unidentified leakage, and

b. -the total flow rate can be established and monitored."

Assessment ,

Identified sources of primary system leakage are collected in the reactor containment in the primary drain transfer tank- (PDTT) and pressurizer relief tank (PRT). The~PDTT receives leak-off from various primary valve packings, #2 and #3 RCP seal leak-off, and leakage from the reactor vessel _ flange leakage detection system.

The PRT receives leakage from the pressurizer power operated relief valves (PORV's). Flow rates from identified leakage sources,are determined by the PRT and PDTT level changes over time as a part of procedures PT 52.2 and PT 52.2A (RCS Leak Rate Determination). Additionally, these tanks have High level alarms to enhance the ability to detect a sudden increase in identified leakage.

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Design Guideline 2

" Leakage to the primary reactor containment from unidentified sources should be collected and the flow rate monitored with an accuracy of one gallon per minute (gpm) or better."

Assessment:

Leakage to the primary containment from unidentified sources is collected in a common sump within the containment. Though not directly monitored, flow rate is calculated to within the required accuracy by observing level change over a fixed time interval. Again, sump high level alarms detect a sudden change in unidentified leakage. Normally, at North Anna Power Station, unidentified leakage is calculated by subtracting identified leakages (PDTT and PRT) from total leakage make up to the volume control tank (VCT).

Design Guideline 3:

"At least three separate detection methods should be employed and two of these methods should be (1) sump level and flow monitoring, and (2) airborne particulate radioactivity monitoring. The third method may be selected from the following:

a. monitoring of condensate flow rate from air coolers, and 60-KKD-4699B 5

b. monitoring of airborne activity.

Humidity, temperature, or pressure monitoring of the containment atmosphere should be considered as alarms or indirect indication of leakage to the containment."

Assessment Leakage from the RCS to the containment is indicated in the con-trol room by one or more of the following methods:

1. The containment air particulate radioactivity monitoring system is indicated, recorded, and alarmed in the control room.

2 .- The containment gas monitor is indicated, recorded, and alarmed in the control room.

3. Abnormal makeup water requirements to maintain normal level in the pressurizer are.an indication of a leak. The demineralized water and concentrated boric acid makeup flow

. rates are both recorded and alarmed in the main control room.

4. The instrumentation for containment pressure, partial pres-sure, temperature, dew point and sump level all indicate in the control room. This instrumentation provides an indirect indication of leakage into the containment.

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5. Reactor vessel leak-off through the reactor vessel head flange is collected and will cause a high temperature in the drain line, which actuates an alarm in the control room.

Design Guideline 4:

" Provisions should be made to monitor systems connected to the Reactor Coolant. Pressure Boundaries (RCPB) for signs of intersystem leakage.

Methods should include radioactivity monitoring and indicators to show abnormal water levels or flow in the unaffected area."

Assessment The condenser air ejector monitor, the component cooling water monitor, and the steam generator blowdown monitors indicate primary system leakage into interconnected systems. Leakage into the low head safety injection system (Event V) is precluded by redundant check valves which are subject to routine surveillance.

Leakage into this system would ultimately-be detected by the refueling water storage tank (RWST) level detection system.

Process and effluent radiation monitoring system and particulate activity monitoring system would detect leakage into condenser, component cooling and blowdown.

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Design Guideline 5:

"The sensitivity and response time of each leakage detection system in regulatory position 3 above employed _for unidentified leakage should be adequate to detect a' leakage rate, or -its equivalent, of I gpm in -

less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />."

Assessment UFSAR Section 5.2.4.1 discusses leakage detection and' sensitivity and to the extent to which R.G. 1.45 is being. met. Generic Let-ter 84-04 allows an exception to Regulatory Guide 1.45 by requir-ing that "at least one leakage detection system with a sensitivity capable of detection 1 gpm in 4. hours must be opera-ble." The sensitivity and response time of the methods described for detection of RCS leakage are dependent on RCS activity, pre-vious leakage rate which can increase steady state count rates on the process and effluent radiation monitoring system and particu-late radiation monitoring system, frequency of surveillance, etc.

Different-detection methods would be effective over a broad range of operating conditions. Given the depth and redundancy of leak-age detection methods described above, we are confident a 1-gpm leakage rate would be detected within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during steady-state operation.

Additionally, with indication of primary coolant leakage, using the methods previously listed, North Anna Power Station Abnormal 60-KKD-4699B 8

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Procedure AP-16L provides guidance' for ' response to _ increased leakage for determining the cause of:the-leakage and the leak rate.- Periodic Test Procedures PT-52.2 and PT-52.2A are used to-determine the leak rate. The' indicating parameters are continually monitored as a function of control room operation and data recorded in the control room at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

. Design Guideline 6:

"The. leakage detection systems should be capable of performing their functions following seismic events that do not require plant shutdown.

The airborne particulate radioactivity: monitoring system should remain functional when s'ubjected to the SSE."

Assessment Generic Letter 84-04 allows an exception to Regulatory Guide 1.45 by stating that "the seismic qualification of the airborne par-ticulate radiation monitor is not necessary."

The containment pressure instrumentation at North Anna Units 1 and_2 is seismically qualified and satisfies the requirements of-~

Regulatory Guide 1.97 for post accident monitoring. The contain-ment sump narrow range level instrumentation will be seismically qualified prior to restart of each unit with the reactor coolant system supports redesigned.

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Design Guideline 7:

" Indicators and alarms for each leakage detection system should be provided in the main control room. Procedures for converting various indications to a common leakage equivalent should be available to the operators. The calibration of the indicators should account for needed independent variables."

Assessment Indicators and alarms are prov'ided in the control room for all five leakage detection systems. Operator curves are used to con-vert leakage to _ vessels to a flow rate value. Periodic Test Procedures PT-52.2 and PT 52.2A are used to quantify identified and unidentified RCS leakage.

Design Guideline 8:

"The leakage detection systems should be equipped with provisions to readily permit testing for operability and calibration during plant operation."

Assessment The instrumentation used for leakage detection at North Anna Units 1 and 2 can be calibrated or tested for operability at any time.

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Design Guideline 9:

"The technical specifications should include the limiting conditions for_ identified and unidentified leakage and address the availability of various types of instruments to assure-adequate coverage at all times."

Assessment Technical Specification 3/4.4.6 gives limiting conditions for

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both identified and unidentified leakage. Also, it requires that "at least three means shall be available-to detect reactor cool-ant system leakage. One of those means must depend on radioac-tivity monitoring leakage detection."

III.

SUMMARY

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The following summarizes each of the nine regulatory position design guidelines cited in Section II:

1. North Anna Units 1 and 2 satisfy the regulatory position of detecting identified and unidentified leakage.

2. Periodic Test Procedure PT-52.2 and PT-52.2A, " Reactor Coolant System Leak Rate," are capable of measuring flow rates with an accuracy better than 1 gpm.

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3. - North Anna has five methods for ' leakage detecting:

a. Containment gaseous radioactivity monitor,

b. Containment air particulate radioactivity monitor,

c. Containment structure leakage monitoring system.

d. Containment sump monitoring,

e. Reactor coolant system makeup rate

4. North Anna uses the following monitors to detect primary system leakage into interconnected systems:

- Condenser air ejector monitor

- Component cooling water monitor

- Steam generator blowdown monitor

5. North Anna is capable of detecting a leakage rate of 1 gpm in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

6. NRC Generic Letter 84-04 does not require the airborne particulate monitors to be seismically qualified. The containment pressure and sump narrow range level instrumentation are seismically qualified.

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7. . Indicators for all' the leakage detection systems are provided in i the control-room. The containment temperature and dew point indi-cators are displayed and alarmed via the plant process computer.

All others have alarms. Various parameters used to determine leakage rate are converted using operator curves and leakage (total, identified, and unidentified).-

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8. The instrumentation for leakage detection can be calibrated or tested for operability at any time.

9. North Anna Technical Specification 3/4.4.6 gives limiting condi-tions for both identified and unidentified leakage.

CONCLUSIONS This evaluation shows that North Anna Units 1 and 2 have systems which are capable, with margin, of detecting the leakage from the postulated through-wall flaws.

IV.. REFERENCES

1. Regulatory Guide 1.45, " Reactor Coolant Dressure Boundary Leakage Detection Systems."

2. NRC Generic Letter 84-04, " Safety Evaluation of Westinghouse Topf-cal Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1984.

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~3. NUREG-1061] Volume 3, Report of the U.S. Nuclear Regulatory

' Commission Piping Review Committee; Evaluation of Pipe Breaks, _

November, 1984;

4. ' North Anna UFSAR Section 5.2.4 " Reactor Coolant Pressure Boundary Leakage Detection Systems."

5. orth Anna Technical Specification.3/4.4.6, " Reactor Coolant Sys-tem Leakage."

6. North Anna Power Station Periodic Test Procedure PT-37 " Radiation Monitoring Equipment Check."

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7. North Anna Power Station Periodic Test Procedures' PT-52.2 and PT -

52.2A, " Reactor Coolant System Leak Rate."'

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8. North Anna Power Station Abnormal Procedure AP-5.2 " Radiation Mon-itoring System."

9. North Anna Power Station Abnormal Procedure AP-16 " Excessive Pri-

' mary Plant Leakage."

10.~ North Anna Power Station Abnormal Procedure AP-18" Increasing. Con-tainment Pressure."

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11. Alarm Response Procedure AR-9, Annunciator J-7, Containment Partial Pressure +0.25 PSI.

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12. Alarm Response Procedure AR-9, Annunciator J-14, Containment Partial Pressure +0.1 PSI.

13. Alarm Response Procedure AR-9, Annunciator J-41, Reactor Con-I tainment Sump High level,0* 8.6 in.

14. Alarm Response Procedure AR-9, Annunciator J-59, Incore Instrument Room Sump High_ level 18" inch.

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