ML040750334

Draft Regulatory Guide 1.139, Guidance for Residual Heat Removal

ML040750334
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Issue date:	05/31/1978
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RG-1.139
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U.S. NUCLEAR REGULATORY COMMISSION May 1978 REGULATORY GUIDE OFFICE OF STANDARDS DEVELOPMENT

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i-r REGULATORY GUIDE GUIDANCE FOR RESIOUA USNRC REGULATORY GUIDES Regulatory Guides re Issued to describe and make available to the Public methods acteatable to the NRC staff of Imolementing apecific Parts of the Commiwon's reouiationsto delineate ttchniesue used by the etaff in teluating seecific problems or poetuiatd accidents, or to provide guidance to applicants. Regulatory Guidet are not stbetitutes for regulations, and compliance with them is not requir d.

Methods and solutions different from those wee out in the guides voll be accept abe It they provide a bai, for the findings requiaite to the Isiance or continuance of a permit or license by the Commision.

Comments and aiessions for improsements in thes guides we encouraged It 11 times, and uiden will be revised. en eapropriate. to accommodate comments and to reflect new Information or experience.

However, comments on this guiderlf received within about two months after Its ieuance. will be particularly useful in evaluating the nwed for an earty reviaion.

Comments should be sent to the Secretary of the Commi4nonUS.Nuctavo Aegu.

tatory Commisson, Washington. D.C.

20555. Attention. Docketiung and Service Brandc.

The guide rt re inued in the following ten broad divrsions

1. Power Reactors

6. Productt

2. Reseatch and TentReactors

7. Transportation

3. FuoisandMatelralsFacilites S. OccuPational Health

4. Environmentalnd Siting

9. Antirust Review S. Materials and Plant Protection

10. Genral Requets tfor single conis of issued guides lthich may be regroducedl or tor place-ment on en automatic distribution lot for single conis of future uides in soecific disisions should be mader in writing to the US. Nucleer Regulatory Commisson.

Washington, D.C.

20555. Attention:

Drectvor Divison of Document Control.

A. INTRODUCTION General Design Criterion (GDC) 19, "Control Room," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utilization Facilities,"

requires that it be possible to take actions from the control room to maintain the power plant in a safe condition during normal operation or in the case of an accident. GDC 34, "Residual Heat Removal," requires that a system to remove residual heat be provided. GDC 34 defines the system's safety function as the transfer of fission product decay heat and other residual heat from the reactor core after the reactor is shut down so that acceptable design limits of the fuel and the reactor coolant pressure boundary are not exceeded.

Furthermore, GOC 34 requires that the system safety function can be accomplished assuming the availability of only onsite or offsite power, coincident with a single failure.

This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to the removal of decay heat and sensible heat after a reactor shutdown.

B. DISCUSSION The safe shutdown of a nuclear power plant following an accident not related to a loss-of-coolant accident CLOCA) has been typically interpreted as a hot standby.

Consequently, considerable emphasis has been placed on the hot-standby condition of a power plant in case of an accident or abnormal occurrence. A similar degree of emphasis has been placed on long-term cooling, which is typically achieved by the residual heat removal (RHR) system.

The RHR system starts to operate when the reactor coolant pressure and temperature are sub-stantially lower than their hot-standby condition values. It is the intent of this guide to place the same degree of emphasis on the entire range of reactor coolant temperatures and pressures, including the range between hot standby and RHR operation conditions.

The importance of reliable systems that remove decay heat from the reactor coolant system (RCS) while the latter is at or near normal operating temperatures is indicated by the results of WASH-1400, "Reactor Safety Study" (RSS).

The capability of a typical PWR plant and a typical BWR plant to remove decay heat following a plant trip was evaluated in the RSS on a probabilistic basis. The evaluation included both those events in which the reactor protection system (RPS) failed (Anticipated Transients Without Scram) and events in which the RPS func-tioned as designed.

For these types of events, it was considered acceptable to maintain the reactor at or near normal operating temperature and pressure for a long time.

However, in the event of a plant trip even with a successful operation of the RPS, systems or equipment failures that led to the inability to remove decay heat resulted in a higher proba-bility of a core melt than that predicted for a large LOCA for both PWRs and BWRs. Conse-quently, a significant safety benefit will be gained by upgrading those systems and equipment needed to maintain the RCS at the hot-standby condition for extended periods or those needed to cool and depressurize the RCS so that the RHR system can be operated.

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Furthermore, even though it may generally be considered safe to maintain a reactor in a hot-standby condition for a long time, experience shows that thei-e have been events that required eventual cooldown and long-term cooling until the RCS was cold enough to perform inspection and repairs. It is therefore obvious that the ability to transfer heat from the reactor to the environment after a shutdown is an important safety function for both PWRs and BWRs.

Consequently, it is essential that a power plant have the capability to go from hot-standby to cold-shutdown conditions (when this is determined to be the safest course of action) under any accident conditions.

These accident conditions can conceivably include a safe shutdown earthquake (SSE) and an extended loss of offsite power that may have resulted from that SSE.

In that case, all components and equipment that are not seismic Category I and all systems or parts of systems that depend solely on offsite power sources for their operation would be assumed inoperable.

Under these circumstances, a plant safe shutdown (including cooldown) within a reasonable time requires systems designed to safety grade standards and operable from the control room.

However, limited operator actions outside the control room may be permitted if suitably justified.

Four processes are necessary to achieve a cold shutdown in a power plant: (1) the inser-tion of the control rods, with or without boration to the cold shutdown concentration, (2) heat rejection to the surroundings, (3) depressurization, and (4) long-term cooling.

These processes are discussed below.

1. Boration of the RCS to the required cold shutdown concentration provides an addi-tional reactivity control measure to ensure that the reactor will not become critical during and after the RCS cooling.

a.

For pressurized water reactors (PWRs), the boration of the RCS is used in addi-tion to the insertion of the control rods.

Boration is achieved by the chemical and volume control system (CVCS).

It is important that this safety function can be achieved in all accident conditions, including an SSE and an extended loss of offsite power. In case of a loss of offsite power, the only means of mixing the injected boron solution with the reactor coolant is natural convection circulation.

b. For boiling water reactors (BWRs), the boration of the RCS is achieved by the standby liquid control system (SLCS).

However, that system is activated only if the control rods fail to shut down the reactor.

2.

Heat rejection to the surroundings is the only way to avoid a core melt under normal or accident shutdown conditions.

a.

For PWRs, heat rejection is achieved by the main steam system and either the normal or the auxiliary feedwater system.

In case of an SSE, only seismic Category I components and equipment are assumed operable. During a loss of offsite power, the auxiliary feedwater system provides cooling water to the steam generators.

Without offsite power, 1.139-2

reactor cooling depends solely on natural convection circulation induced by the cooling effect of heat transfer to the steam generators.

Natural circulation may lead to slow and unequal cooling.

Slow cooling will result in longer cooldown times, which, in turn, require a larger clean feedwater inventory. Unequal cooling may lead to hot spots and high vessel stresses. It is essential that adequate cooling be provided under these circumstances to keep the integrity of the reactor coolant pressure boundary and maintain the reactor core in a coolable form.

b.

For BWRs, heat rejection is achieved by the main steam system and either (1) the normal feedwater system in conjunction with the main condenser or (2) the reactor core isolation cooling (RCIC) system in conjunction with the condensate storage tank, the residual neat removal heat exchangers in the steam condensing mode, and the pressure suppression pool.

Further heat rejection is achieved by the RHR system after the RCS has been sufficiently depressurized.

3.

For all current designs of PWRs and BWRs, depressurization of the RCS is a prereq-uisite to the operation of the RHR system in the long-term cooling mode; therefore, it is important that systems or components required to depressurize the RCS be designed to withstand severe postulated accident conditions and be able to perform their intended functions.

a.

For PWRs, depressurization of the RCS can be achieved by the pressurizer in conjunction with one or more of these components:

(1) the main pressurizer spray, (2) the auxiliary pressurizer spray, or (3) the pressurizer relief valves.

b.

For BWRs, depressurization of the RCS is achieved by dumping steam to either the main condenser or the pressure suppression pool.

Steam condensation during the RHR system operation in the steam condensing mode will help depressurize the RCS.

4.

For both PWRs and BWRs, following the reactor shutdown and both the initial and the intermediate cooldown periods, long-term cooling is necessary to prevent heat accumulation in the RCS. This function is usually accomplished by the RHR system.

In all current plant designs, the RHR system has a lower design pressure than the RCS.

In most of these designs it is located largely outside the containment. However, in some plant designs, the RHR system is located inside the containment.

C. REGULATORY POSITION

1. FUNCTIONAL The systems necessary to take the reactor from normal operating conditions to cold shutdown, including the RHR system, should satisfy the functional guidance presented below.

a.

The design should be such that the reactor can be taken from normal operating condi-tions to cold shutdown using only safety-grade systems that satisfy General Design Criteria I through S.

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b.

These systems should have suitable redundancy in components and features and suitable interconnection, leak detection and containment, and isolation capabilities to ensure that, for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available), the system safety function can be accomplished assuming a single failure.

In demonstrating that the system can perform its function assuming a single failure, limited operator action outside the control room would be acceptable if suitably justified.

c.

The systems should be capable of bringing the reactor to a cold-shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following shutdown with only offsite power or onsite power available, assuming the most limiting single failure.

2.

RHR SYSTEM ISOLATION

a.

Isolation of the suction side of the RHR system should be provided by at least two power-operated valves in series, with valve positions indicated in the control room. Alarms in the control room should be provided to alert the operator if either valve is open when the RCS pressure exceeds the RHR system design pressure.

The valves should have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure. Failure of a power supply should not cause any valve to change posi-tion.

Independent diverse protective measures should be provided to close any open valve in the event of an increase in the RCS pressure above the RHR system design pressure.

b.

One of the following should be provided on the discharge side of the RHR system to isolate it from the RCS:

(1) The valves, position indicators, alarms, and interlocks described in item a.

(2) One or more check valves in series with a normally closed power-operated valve with its position indicated in the control room.

If the RHR system discharge line is used for an ECCS function, the power-operated valve should be opened upon receipt of a safety-injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.

(3) Three check valves in series, or (4) Two check valves in series, provided there are design provisions to permit periodic testing of the check valves for leak tightness and the testing is performed at least annually.

3.

RHR SYSTEM PRESSURE RELIEF

a.

To protect the RHR system against accidental overpressurization when it is in opera-tion (not isolated from the RCS), pressure relief in the RHR system should be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most 1.139-4

limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS should be considered when selecting the pressure relieving capacity of the RHR system.

For example, during shutdown cooling in a PWR with no steam bubble in the pressuri-zer, inadvertent operation of an additional charging pump should be considered in selecting the design bases.

Fluid discharged through the RHR system pressure relief valves should be collected and contained so that a relief valve that is stuck in the open position will not:

a.

Result in flooding of any safety-related equipment.

b.

Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.

c.

Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside the containment.

If interlocks are provided to automatically close the isolation valves when the RCS pres-sure exceeds the RHR design pressure, adequate relief capacity should be provided during the time that the valves are closing.

4.

RHR SYSTEM PUMP PROTECTION The design and operating procedures of the RHR system should include provisions to prevent damage to the RHR system pumps due to overheating, cavitation, or loss of adequate pump suction head.

S.

RHR SYSTEM TESTING For the RHR system, the isolation valve operability and interlock circuits should be designed to permit on-line testing when operating in the RHR mode.

System testing should meet the requirements of IEEE Standard 338 and the recommendations of Regulatory Guide 1.22.

The preoperational and initial startup test program should be in conformance with Regula-tory Guide 1.68.

The programs for pressurized water reactors should include tests with sup-porting analysis to confirm (a) that adequate mixing of borated water added prior to or during cooldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve such mixing and (b) that the cooldown under natural circulation conditions can be achieved within the limits specified in the emergency operating procedures.

Comparison with the performance of previously tested plants of similar design may be sub-stituted for these tests.

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6.

AUXILIARY FEEDWATER SUPPLY The seismic Category I water supply for the auxiliary feedwater system for a PWR should have sufficient inventory to permit operation at hot-standby conditions for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followed by cooldown to the conditions permitting operation of the RHR system. The inventory needed for cooldown should be based on the longest cooldown time needed with either only onsite or only offsite power available with an assumed single failure.

7.

OPERATIONAL PROCEDURES The operational procedures for bringing the plant from normal operating power to cold shutdown should be in conformance with Regulatory Guide 1.33.

For pressurized water reactors, the operational procedures should include specific procedures and information required for cooldown under natural circulation conditions.

D. IMPLEMENTATION The purpose of this section is to provide information to applicants regarding the NRC staff's plans for implementing this regulatory guide.

Except in those cases in which the applicant proposes an acceptable alternative method for complying with the specified portions of the Commission's regulations, the method described herein will be used in the evaluation of submittals in connection with applications for construction permits for all plants (standard and custom), manufacturing licenses, and preliminary design approvals docketed on or after January 1, 1978.

All applications docketed before January 1, 1978, will be reviewed against this guide on a case-by-case basis.

V C. S. GOVERNMENT PRtNTINGC OFFlCE:

1979-72O-357/12.

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