ML19031A649

10/26/1977 Letter Submit Report Reactor Coolant System Overpressurization Overpressure Protection.

ML19031A649
Person / Time
Site:	Salem
Issue date:	10/26/1977
From:	Librizzi F Public Service Electric & Gas Co
To:	Lear G Office of Nuclear Reactor Regulation
References
Download: ML19031A649 (22)

Text

PS~G 9 Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000

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October 26, 1977 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. c. 20555 Attention: Mr. George Lear, Chief Operating Reactors Branch 3 Division of Operating Reactors Gentlemen:

OVERPRESSURE PROTECTION NO. 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272 By our letter of March 25, 1977, Public Service committed to the installation of an interim system to prevent reactor pressure overpressurization by December 31, 1977 and to have a final system installed by the end of the first refueling outage. Enclosed is a report of the design which will meet both of these commitments and which we plan to have operational by December 31, 1977.

Should you have any questions concerning this report, we will be pleased to discuss them with you.

Very truly yours, F. P. Librizzi General Manager -

Electric Production Encl.

P77 98 39 The Energy People 95-0942

J r.*'

REACTOR COOLANT SYSTEM OVERPRESSURIZATION OVERPRESSURE PROTECTION NO. 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272

TABLE OF CONTENTS Section Title I. Pressure Transient Analysis II. System Description and Design Criteria III. Design Evaluation IV. Administrative Controls

v. Conclusion Appendix A Westinghouse Report-Pressure Mitigating Systems Transient Analysis Report LIST OF FIGURES Figure No. Title Page 1 Reactor Coolant System 13 Pressure-Temperature versus Cooldown Rates 2 Reactor Coolant System 14 Pressure-Temperature* versus Heatup Rate Schematic and Logic Diagrams 3 CD-249108 15 4 CD-249109-Rev. 1 16 5 CD-249110 17 6 CD-249111-Rev. 1 18 7 CD-249112 19 LIST OF TABLES Table No. Title 1 Comparison of Salem and Westinghouse 12 Reference Parameters I. Pressure Transient Analysis Examination of past and postulated Reactor Coolant System overpressurization events of recent concern has resulted in the identification of two basic types of overpressurization mechanisms. They are categorized as mass input and heat input. For the mass input case the most severe transient has been identified as the inadvertent starting of a Safety Injection Pump, due to its high flow capacity. For the heat input case, the most severe transient is the inadvertent starting of a Reactor Coolant Pump with the presence of a significant temperature differential between the Reactor Coolant System and the Steam Generator in the affected loop.

Background and justification for the above determination are described further in Appendix A of this report.

A. Mass Input Case The Reactor Coolant System pressure transient to be expected in the Salem Plant for the identified worst case mass input incident was calculated in accordance with the algorithm described in Appendix A. Using the assumptions, inherent in that algorithm, the peak mass input induced pressure was calculated at 446 psig. This peak pressure transient is a function of mass input rate, valve setpoint, valve opening time, RCS volume and the assumption that only one relief valve is available. The peak pressure transient when compared to the 13EFPY (Effective Full Power Years)

Pressure Temperature limit curves (Figures 1 and 2),

is within limits down to a reactor coolant temperature of 1800F during heatup and 1500F during cooldown.

Below these temperatures the peak pressure exceeds the Technical Specification curves by 20 and 46 psig, respectively.

There are several compelling reasons to believe th~t the actual pressure transients which would be experienced in the worst case would not exceed the pressure limits of 10CFR50 Appendix G even though slight infringement on the Technical Specification curves is indicated.

These reasons are as follows:

(1) The calculation of the pressure transient was based on the use of a single pressurizer relief valve.

The Salem Pressurizer Overpressure Protection System (POPS), provides for actuation of a second pressurizer relief valve and the use of the RHR suction line relief r (1) (Cont'd) valve (which has a comparable capacity and the same setpoint) .

The effect on overpressurization of the second pressurizer relief valve was not calculated for the Salem Plant due to the unavailability of an appropriate algorithm in Appendix A. It should be noted, however, that sensitivity studies conducted by Westinghouse for a 6,000 cubic foot plant showed that the addition of a second pressurizer relief valve reduced the over-pressurization (and, hence, the peak pressure) by 30 psi.

(2) The pressure transient calculated was based on pressurizer relief valve opening time.

Sensitivity studies performed by Westinghouse indicated that the pressure transient is extremely sensitive to relief valve opening times, i.e.

the quicker acting the valve, the more effective it is in mitigating the pressure transient. In comparison, the RHR relief valve opens within a fraction of a second whereas the pressurizer relief valves open in two seconds.

Thus, given the expected availability of the RHR suction relief valve, a significantly lower peak is indicated.

(3) The reference parameter algorithm (Appendix A) was developed using conservative bounding input parameters and parameter sensitivity studies which provided a linearized method for calculating a plant specific analysis. As a result of the assumptions made and the linearization method used for the bounding algorithm, pressure transient calculations yield greater magnitudes than would be realized under actual process conditions, and are therefore conservative.

(4) The Technical Specification Curves include significant conservatisms in both temperature and pressure to account for instrument error.

B. Heat Input Case The calculations for the heat input case were performed in accordance with the supplement to the Westinghouse Report (Appendix A). The peak pressure transient was calculated by interpolating combinations of the following variables; relief valve setpoint, relief valve opening time, reactor coolant system volume and steam generator heat transfer area. In all cases, the calculated peak pressure was less than the worst case previously described under Mass Input Case.

II. SYSTEM DESCRIPTION AND DESIGN CRITERIA The POPS is a two-train overpressurization mitigating system which uses separate and independent pressure transmitters to open the two pressurizer power operated relief valves (lPRl and 1PR2) in the event Reactor Coolant System Pressure exceeds the preset value of 375 psi. This automatic action takes place provided the system has been armed by placing two key-locked pushbuttons in the "ON" position as shown in Figures 3 thru 7. The system will be armed whenever the reactor coolant system is below 3120F.

Each relief valve is actuated by its own logic relay which is energized by a bistable device. The bistable

• is energized when reactor coolant pressure exceeds the setpoint. Existing installed pressure sensors are used to develop the signal for valve actuation. These are the same sensors which provide automatic closure of the RHR suction paths at 600 psi.

Operation of the POPS is governed by two adminis-tratively controlled, keylocked pushbuttons which perform three functions. When the Reactor Coolant System temperature is less than 312oF, the system is armed by depressing the "ON" pushbutton for each POPS train.

An actuation signal to open the motor operated valves upstream of the relief valves is initiated as well as an alarm to indicate that the POPS is armed if temperature should subsequently increase above 3120F. In this mode of operation, the relief valve will be opened automatically if Reactor Coolant System pressure exceeds 375 psi.

When Reactor Coolant System temperature increases above 3120F, the "OFF" pushbutton for each POPS train is depressed. This action removes the opening permissive signal to the relief valve, removes the opening signal from the associated motor operated valve, and provides an alarm to indicate that the system is d.isarmed if temperature is subsequently decreased below 3120~. Upon actuation, the valves will open and will reset when system pressure decreases below 375 psi.

Reactor Coolant System pressure and temperature instrumentation are provided which enable the operator to monitor the above parameters. An alarm is provided on the main control console to inform the operator of a POPS system initiation. Valve position indicating lights inform the operator that the valves have opened. In addition, a computer generated alarm informs the operator of an impending pressure excursion beyond the Technical Specification limits.

Testing provisions in the POPS circuitry allow for test opening of the relief valves prior to arming of the system below 3120F. The "TEST" pushbutton, when depressed, will operate the relief valve provided that the associated upstr~am motor operated valve is closed. Other portions of the POPS can be tested in a manner similar to other plant protection systems.

This design meets the four design criteria established in the NRC's November 17, 1976 letter.

(1) "Credit for Operator Action" - No credit can be taken for operator action until 10 minutes after the operator is aware that a pressure transient is in progress.

The POPS requires no operator action other than to arm the system prior to operation with reactor coolant system temperature less than 3120F. All protective action is performed automatically.

(2) "Single Failure Criteria" - The pressure protection system should be designed to protect the vessel given a single failure that initiates the pressure transient. In this area redundant or diverse pressure protection sistems would be considered as meeting the single failure criteria.

The POPS incorporates redundancy and separation of pressure transmitters, logic, and valves in a channelized system. Single failures within the POPS will not defeat the safety function. Single failures which are capable of initiating a pressure transient cannot cause failures in the POPS which would render it unable to provide protection.

( 3) "Testability" - The equipment design should include.

some provision for testing on a schedule consistent with the frequency that the system is used for pressure protection.

The POPS design provides for testing of the analog circuitry any time the RHR suction valves from the Reactor Coolant System are closed. The relief valves lPRl and 1PR2 can be tested prior to entry into a water-solid condition by use of the POPS functional test pushbutton. The POPS is designed to function during low temperature, low pressure operating conditions and therefore, periodic testing of the system during power operation is not planned.

(4) "Seismic Design and IEEE 279 Criteria" - Ideally, the pressure protection system should meet both seismic category 1 and IEEE 279 criteria. The basic objective, however, is that the system should not be vulnerable to an event which both causes a pressure transient and causes a failure of equipment needed to terminate the transient.

The POPS design meets seismic Category I criteria for all equipment required to open the relief valves.

The instrumentation and actuating circuitry meet the applicable requirements of IEEE 279 - 1971, provided the system is armed by the operator.

III. DESIGN EVALUATION The POPS is designed as a "protection grade" system in accordance with the applicable portions of IEEE 279 - 1971.

The use of proven devices provides assurance that the system is compatible with other protection system equip-ment. The use of administrative controls to arm the POPS is considered acceptable due to the infrequency of low pressure, low temperature operation.

The effects of various failures have been considered in the POPS design. These failures considered include "loss of control air" and "loss of station power." Failures within the POPS cannot cause a loss of protective function, due to the two-train design, and failures capable of causing an overpressurization event cannot cause failures within the POPS or prevent operation of the system.

The "loss of air" situation is provided for by provision of an air accumulator for each relief valve. The accumulators are sized to provide control air for up to 100 cycles of valve opening and closing. The accumulators are designed to seismic Category I requirements. The air accumulators are provided with an alarm which will alert the operator to a low air pressure condition. The accumulator design thus precludes a total loss of control air to the relief valves.

A "loss of station power" will not affect the POPS since the protection logic power is provided by inverters and control power for the relief valves originates at the batteries.

In the event that one relief valve were to open on a false signal or transmitter failure at a time when protection is not required, a depressurization of the Reactor Coolant System would occur. Any such depressurization would be less severe than those analyzed in FSAR Section 14.1.2. The discharge through the relief valve can be terminated by operator action, thus minimizing the effects of the transient.

IV. ADMINISTRATIVE CONTROLS The following components and conditions have Administrative Controls placed on them to prevent any overpressurizing of the reactor coolant system.

(1) A reactor coolant pump is not to be started unless there is a bubble in the pressurizer, except when filling and venting the Reactor Coolant System.

It can be jogged for 1 to 1 1/2 minutes while venting the system (OI's II-1.3.1 and II-1.3.4).

(2) To prevent the accidental starting of a safety injection or charging pump when RCS temperature is below 3500F, the power supplies to both safety injection pumps and to one centrifugal charging pump are deenergized and racked out.

Additionally, when the RCS temperature is reduced below 2000F, the remaining centrifugal charging pump's electrical supply will be deenergized and racked out. (OI-I-3.6)

V. CONCLUSION In summary, we believe that the material presented in this report meets all of the NRC staff's concerns related to overpressurization incidents. This system described herein provides satisfactory overpressure protection during low pressure, low temperature operation for both water solid and non-solid modes of operation, up to 3120F.

RWS:jk P77 98 39/52 TABLE I COMPARISON OF SALEM AND WESTINGHOUSE REFERENCE PARAMETERS Westinghouse Actual "reference" Salem plant plant A. GENERAL

1. RCS volume, ft3 13,000 12,800

2. Pressurizer relief valve

a. opening time, sec. 3 2

b. rating @ 1 psig. 50 gpm @ 600F 55 gpm @ 600F

c. size, inches 2 2

a. plug characteristic linear linear

3. RHR suction relief valve

a. opening time, sec. not considered << 1

b. rating @ setpoint of 375 psig + 10%

accumulation not considered 900 gpm @ 600F

c. size, inches not considered 3

d. type not considered spring relief valve B. MASS INPUT

1. SI pump input @

375 psig 129.3 lb/sec 10 8. 4 lb/sec Westinghouse Westinghouse Lower Limit Salem Upper Limit

c. HEAT INPUT

1. Relief Valve Setpoint, 350 375 400 psig ( s)

2. Relief Valve Opening 1.5 2.0 3.0 Time, sec (AT)

3. Reactor Coolant System Volume ft3 (VRCS) 6,000 12,800 13,000

4. Steam Generator Heat Transfer area, ft2 51,500 58,000 r

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APPENDIX A WESTINGHOUSE REPORT - PRESSURE MITIGATING SYSTEMS TRANSIENT ANALYSIS REPORT