ML19253A269
| ML19253A269 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/30/1972 |
| From: | OAK RIDGE NATIONAL LABORATORY |
| To: | |
| References | |
| ORNL-TM-3782, NUDOCS 7908210197 | |
| Download: ML19253A269 (96) | |
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s ~ 1 s. 's ORNL TM-3782 a, e 4 Contract No. W-7405-eng-26 General Engineering Division THE SELECTION AND PROCUREMENT OF PRESSURE RELIEF VALVES FOR LIGHT-WATER-COOLED NUCLEAR REACTOR SYSTEF3 f prepared by Oak Ridge National Laboratory and MPR Associates, Incorporated (under Subcontract No. 3040 with Union Carbide Corporation, Nuclear Division) l JUNE 1972 t OAK RIDGE NATIONAL IABORATORY Oak Ridge, Tennessee 37830 operated by UNION CARBIDE CORPORATION Nuclear Division 2 I for the U.S. ATCMIC ENERGY COMMISSION 804 152 es lii i FOREWORD r c. This document was originally prepared by MPR Associates, Incorporated, under Subcontrac't Number 3040 with Union Carbide Corpora-tion, Nuclear Division, as an activity of the RDT Standards Program at Oak Ridge Nctional Laboratory. The information presented herein was com-piled by W. R. Schmidt aad L. M. Corocoran of MPR Associates under the direction of W. A. Bush of Oak Ridge National Laboratory. This (.ocument is intended to provide background information and guidance in the selec-tion and procurement of spring-loaded safety valves, pilot-operated pres-sure relief valves, and power-actuated pressure relief valves to be used in light-water-cooled nuclear powe plants under the purview of the United States Atomic Energy Cocnission Division of Reactor Develotant and Technology. The cooperation and essistance provided in compiling the information i presented herein by representatives of Combustion Engineering, Nuclear Power Division; Crosby Valve and Gage Co=pany; Dresser Industries; General Electric Company, Atomic Power Equipment Division; and the Target Rock Corporation are hereby acknowledged. e 804.153 c. V i CONTENTS Abstract 1 a 1. INTRODUCTION 1 2. CAUSE;; JD CONTROL OF OVERPRESSURE IN WATER-COOLED NUCLEAR REACTOR POWER 3YSTEMS 3 2.1 Overpressure Causative Conditions 3 2.1.1 Pressurized-Water Reactor Systems 3 2.1.2 Boiling-Water Reactor Systems 6 2.2 Overpressure Control 8 3. SPECIAL REQUIREMENTS FOR OVERPRESSURE PROTECTION AND PRESSURE RELIEF VALVE SYSTEMS IN NUCLEAR POWER PLANTS 11 3.1 Effect of Overpressure ca Raactor Power 11 3.1.1 Pressurized-9ater Reactor Systems 11 3.1.2 Boiling-Water Reactor Systems 12 i. 3.2 Loss of Coolant and Nuclear Safety 14 3.3 Isolation of Pressure Relief Valves 15 l 3.4 Code Requirements for Overpressure Protection 17 3.4.1 Nuclear Power System Requireme sts 17 3.4.2 Pressure Relief Valve Requirements 18 3.4.3 Rated Capacity and Certification Requirements 20 4.
- 'RESSURE RELIEF VALVE SYSTEMS USED IN WATER-COOLED REACTOR PMNTS 22 4.1 Pressure Relief Valve Systems in Pressurized-Water Reactor Plants 22 4.1.1 Systems in Existing Plants 26 4.t.2 Features of Systems Considered for Future Plants..
28 (a) Pilot-Operated Pressure Ralf ef Valves 29 (b) Power-Actuated Pressure Relief Valves 29 (c) Anti-Simmer Devices 29 42 Prassure Enlief Valve Systems in Boiling-Water Reactor Plants 30 4.2.1 Pre-Section-III Overpressure Protection Systems... 30 4.2.2 S e c tion-III Ove rp re s su re Pro te c tion Sys tems....... 33 804 154 i vi 5. DESIGN FEATURES AND OPERATING CHARACTERISTICS OF THREE BASIC TYPES OF NUCLEAR PRESSURE RELIEF VALVES 37 5.1 Spring-Ioaded Safety Valves 37 5.1.1 Mechanical Design 37 5.1.2 Operation..........'.............................. 40 5.1.3 Auxillary Features 44 (a) Hand Lifting I4 vers 44 (b) Anti-Simer Devices......'................... 45 (c) Remote (Pneumatic) Operators 45 5.2 Pilot-Operated Pressure Relief Valves 45 5.2.1 Three-Stage Pilot-Operated Pressure Relief Valve.. 46 5.2.2 Two-Stage Pilot-Operated Pressure Relief valve 52 5.3 Power-Actuated Pressure Relief Valves 54 6. GUIDELINES IN THE SELECTION OF SPECIFICATION REQUIREMENTS FOR PRESSURE RELIEF VALVES 57 6.1 Design Requirements 57 f 6.1.1 valve Type 57 (a) Spring-Loaded Safety Valves 57 (b) Pilot-Operated Pressure Relief Valves 58 (c) Power-Actuated Pressure Relief Valves 59 6.1.2 Port Arrangement and Size 60 l 61 l 6.1.3 Orientation 6.1.4 Design Conditions 62 6.1.5 Reactor Coolant Chemistry 64 l 6.1.6 Auxiliary Devices 64 (a) Remote Operators 64 (b) Anti-Simmer Devices 66 6.1.7 Requirements for Auxiliary Power Systems 66 i (a) Electrical Systems 67 (b) Pneumatic Systems 67 6.1.3 Seismic Requirements 67 6.1.9 External Loadings 70 (a) Piping Expansion 70 (b) Flow Reaction Forces 71 (c) Seismic Acceleration 71 hk =- vil 6.2 Performance Raquirements 72 6.2.1 Capacity 72 a 6.2.2 S e t Pre s'su re S e le c tion,........................... 72 (a) Code Raquirements 73 (b) Syntam Transients and Valve Imakaga 73 (c) System Perturbations Resulting From Valve Operation 73 6.2.3 Blowdown 74 (a) System Limitations 74 (b) Valve Design 74 6.2.4 Back Pressure ~75 (a) Static Back Pressure 75 (b) Dynamic Back Pressure 75 6.3 Material Raquirements 75 6.4 Quality Assurance Requirements 78 6.4.1 Examination and Test Witness Points 79 6.4.2 Performance Tests 79 7. MAIKTENANCE AND TESTING OF PRESSURE RELIEF VALVES 81 7.1 Receipt Inspection and Testing 81 7.2 Pre-Operational Examination and Tests 82 7.3 Periodic Testing and Maintenance 83 7.3.1 Examination 83 7.3.2 Refurbishing 84 7.3.3 Tes ting.......................................... 84 (a) Iow-Flow Test With Steam 84 (b) Low-Flow Test With Air or Nitrogen 84 (c) Hydraulic Test 85 (d) Tests With Special Fixtures 85 7.4 Non-Routine Maintenance 85 GLOSSARY 89 4 804 156 ix 3 LIST OF FIGURES Figure Page Number , ' Title Number 2.1 Schematic Diagram of a Typical Pressurized-Water 4 Reactor Nuclear Energy System 2.2 Schematic Diagram of a Typical Boiling-Water Reactor 7 Nuclear Energy System 4.1 Typical Pressure Ralief Valve System for a Pressurized-23 Water Reactor 4.2 Typical Water Seal Arrangements for Pressure Relief 25 Valve s 4.3 Total Capacity of Code-Raquired Pressure Relief Valve 28 Systems as a Function of Electrical Power Output for Various Pressurized-Water Reactor Plants 4.4 Typical Pressure Relief Valve System for a Boiling-31 Water Reactor Plant 5.1 Exposed-Spring Type of Spring-Loaded Safety Valve 38 5.2 Enclosed-Spring Type of Spring-Loaded Safety Valve 39 I 5.3 Operatienal Diagram of Spring-loaded Safety Valve 41 5.4 Lift as a Function of Pressure for Spring-Loaded 43 Safety Valves 5.5 Three-Stage Pilot-Operated Pressure Relief Valve 47 5.6 Pre-Load Region of Pilot valve 46 57 Operational Diagram of Three-Stage Pilot-Operated 48 Pressure Relief Valve in Closed Position 5.8 Operational Diagram of Three-Stage Pilot-Operated 49 Pressure Relief Valve in Open Position 5.9 Valve Seating Force and Bellows Expansion as a Func-51 tion of System Pressure 5.10 Two-Stage Pilot-Operated Pressure Relief Valve 53 I 5.11 Solenoid-Operated Power-Actuated Pressure Relief 55 valve 6.1 Typical Seismic desponse Spectra for a Nuclear Power 69 Plant 804 1SI7 e e ee u e= +- - = wem eme oe e-e ,e e - e +*e-we ewMN*** xi LIST OF TABLES Table Page Number Title Number 4.1 Characteristics of a Typical Pre-Section-III over-32 pressure Protection System for a Boiling-Water Reactor Plant (Oyster Creek No. 1) 4.2 Characteristics of Overpressure Protection Systems 34 for Three Section-III Boiling-Water Reactor Plants 6.1 Design Data for Typical Rasctor Coolant System 63 Transients 6.2 Typical Reactor Coolant Chemical Conditions in Cur-65 rent Pressurized-Water and Boiling-Water Reactor Plants 6.3 Materials Used in Nuclear Pressure Relief Valve Parts 77 v. 9 9 1 e I '4 804 158 I 1 THE SEIICTION AND PROCUREMENT OF PRESSURE RELIEF VALVES FOR LIGHI-WATER-COOLED NUCLEAR REACTOR SYSTEMS
- Abstract Protection againat. overpressure cust be provided in nuclear reactor systems. Overpressure causal conditions and controls and special requirements for pressure relief valves used in pressurized-water and boiling-water reac-tor systems are discussed to provide background informa-tion. The pressure relief systems currentiy used in both pressurized-and boiling-water reactor plants as well as those considered for future plants are discussed. The three basic types of pressure relief valves used for over-pressure protection in nuclear power tilants are described; guidelines for their selection with respect to design, performance, and materials are presented; and their rec-ormended maintenance and testing procedures are discussed.
1. INTRODUCTION This docu=ent was prepared to provide background information and guidelines for the selection and procurement of pressure relief valves to be used in central station light-water-cooled nuclear reactor power plants. The information and guidelines presented herein are based on discussions with manufacturers of pressure relief valves used in nuclear service, designers of reactor plants involved in the selection and pro-curement of pressure relief valves, and others active in the design, fab-rication, testing, and operation of central station nuclear plants. l The information presented herein is primarily directed to the plant owner, operator, or purchaser of pressure relief valves who is not closely familiar with the selection, specification, design, or use of pressure i relief valves for nuclear serv 1ce. This document is not intended to serve as a design manual for pressure relief valves or as a specification for overpressure protection system requirements and the selection of pressure relief valves. Specific requirements for the selection and pro-curement of pressure relief valves =ust necessarily be based on the type, 804 159 .-. -- -.~. - - - = - [ 2 power rating, design for the overpressure protection system, and other safety devices and controls for a particular plant. However, the general information discussed in this document should provide background informa-j tion useful to the designer responsible for the procurement of pressure relief valves and assist the plant ownar or operator in evaluating the I adaquacy of his overpressure protection system. The intended primary objectives of this document are i 1. to inform the plant owner or purchaser of relief valves of the various applications of these valves and their combinations in existing pressurized-water and boiling-water nuclear reactor plants; 2. to describe the need for overpressure protection, the reactor plant conditions which may lead to overpressure, and the special safety I considerations involved in the provision of overpressure protection I in nuclear plants; 3. to acquaint the plant owner or purchaser of relief valves with the various types of pressure relief valves used in nuclear plents and to describe their operating principles; and 4. to provide background information and guidelines useful in the pro-curement of pressure relief valves for nuclear service. i I I d 804 160 3 s 2. CAUSES AND CONTROL OF OVERPRESSURE IN WATER-COOLED NUCLEAR REACTOR POWER SYSTEMS In general, overpressure in water-cooled reactor plants occurs as a result of a mismatch of the plant electrical load and the reactor power. Power mismatches in which the power demand is less than the power gener-l ated by the reactor generally result in increases in the temperature of the reactor coolant, expansion of the reactor coolant, and increases in i I the pressure in the reactor coolant system. Such power mismatches must [ be accommodated by the plant design because the time constants inherent to reactor systems and controls make it impossible for the reactor to respond instantly to changes in power demand. 2.1 Overpressure Causative Conditions t Power mismatches which result in pressure increases in the reactor coolant system can occur as a result of either a reduction in load or an f increase in reactor power. The actual system conditions which cause the changes in demand or reactor power differ somewhat in pressurized-water reactor systems and boiling-water reactor systems and are therefore dis-cussed separately. 2.1.1 Pressurized-Water Reactor Systems I A typical pressurized-water reactor system is illustrated schemati-cally in Fig. 2.1. The reactor coolant system consists of the reactor l vessel, which contains the reactor core, and two to four reactor coolant { loops, each of which contains a steam generator and one or more reactor coolant pumps. Heat is removed fram the reactor core by the reactor coolant in the primary system and is transferred to the secondary system feedwater in the steam generators. (The resultant steam is used to drive the steam plant turbine generators.) The pressurizer, which contains a saturated steam volume in the upper half and reactor coolant in the lower Ob 9 tie 4 0 p _ _ _ _ _ _ _ _ _ _ _ __ t E,,s c[os, t 0,.u.! a,1xvf al,_ _ _, __ _, _ _ _ _,,_ _,,,,_ _ IU stalus.tetets U I
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In the pressurized-water reactor system achamatically depicted in Fig. 2.1, power mismatches and the resultant system overpressure can occur as a result of (1) a reduction in the plant load or (2) an increase in reactor power. 1. A reduction in the plant load can occur as a result of nomal power changes in the system or as a result of an abnormal loss of power, such as the opening of the main electrical circuit breakers, automatic or inadvertent closing of the main turbine root valves (turbine trip), or closure of the main steam isolation valves between the steam generators and the turbines. In addition, in typical central station pressurized-water reactor plants, the volume of the steam space in the Yressurizer is insufficient to accomodate the total volumetric expansion of reactor coolant that occurs with large electrical load reductions. As a result, reactor coolant is bled from the system during such load changes. Mal-function of these bleed or letdown systems can also lead to overpressure conditions in the reactor coolant system. 2-An increase in reactor power and the possible overpressure of j the.eactor coolant system may be caused in several ways in a pressurized-water reactor plant. These include inadvertent withdrawal of control rods, changes in the level of neutron poisons in the reactor core, and a reduction in the inlet temperature of the reactor coolant and moderator. Inadvertent control rod withdrawal is postulated to occur as a result of I malfunction of the reactor control system or operator error. The level of neutron poison in the reactor core may be reduced in plants which e= ploy soluble neutron poison, such as boric acid, by the accidental injection of water that is low in the required concentration of neutron poison (boric acid). Inadvertent reduction in the inlet temperature of the reactor coolant can occur in plants -hat contain isolation valves on 804 163 i l l 6 I each reactor coolant loop by startup of an idla and partially cooled-down loop. The effect of the reduced temperature of the reactor. coolant is to i= prove the effectiveness of the water as a neutron moderator, thereby l increasing the population of thermal neutrons which cause fission reactions. Of these postulated normal and abnormal conditions, the most signif-icant from the standpoint of systets overpressure in most plants are the conditions which result in complete loss of electrical load. In some pressurized-water reactor plants, the inadvertent rod withdrawal accident i g may be the most significsne. i 2.1.2 Boiling-Water Reactor Systems A typical boiling-water reactor system is illustrated schematically in Fig. 2.2. This system differs from the pressurized-water reactor sys-tem in that the steam required to drive the turbine generators is gener-ated directly by the reactor core in the reactor vessel. 'Ihe re fore. there are no steam generators, and since the reactor coolant system con-tains significant amounts of steam as well as water, a pressurizer is not required. As in pressurized-water reactor systers, overpressure in -boiling-water reactor systems resulting from power mismatches can occur as a result of (1) a reductio'n in the plant load or (2) an increase in reactor power. 1. A reduction in the plant load can occur as a result of normal -system power changes or as a result of an abnormal loss of power, such as opening of the main electrical circuit breakers, automatic or inadver-closing of the main turbine root valves (turbine trip), or closure tent of the main steam isolation valves between the reactor and the turbines. 2. An increase in reactor power and the possible overpressure of the reactor coolant system can be caused in several ways in a boiling-water reactor plant. These include inadvertent withdrawal of control rods, changes in the level of neutron poisons in the reactor core, and a reduction in the inlet temperacure of the reactor coolaat and n:oderator. In addition, because of the direct cycle of the boiling-water reactor hk -_5-.- s e s 1 8f54-3 IIAtill CCXIlitIII: c--------------------------i ,,,,u.uilu sin :: Tim I l Ptttt-ACTB11tB Pt(Itttt t!LitT l .. r 'i illil 88 Fitti IFill!!I tit 33581 l g g till!F i! Lit
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system, a loss of feedwater heaters can result in the injection of j
relatively cool water into the core, thereby adding reactivity. The loss of feedwater heaters must therefore be considered in an evaluation of the conditions which result in overpresspre in a' boiling-water reactor system.
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As La pressurized-water reactor systama, the condition in a boiling-water reactor system that is the most severe with respect to overpre s are
. is the complete loss of electrical'lcad.
2.2 Overpressurs Control The overpressure which results from the normal and abnormal load changes discussed in the preceding subsection must be considered in the
.atructural design of the reactor coolant and associated systems, and it
.must be ILmited to acceptable values for all conceivable conditions. The
- basis for evaluating the effect of overpressure on the structural design of the reactor coolant system isSection III, Naclear Power Plant Cempo-nents, of the ASME Boiler and Pressure Vessel Code. This code provides for aufficient msrgin in the component designs to accommodate an over-i pressure of up to 10%'of the design pressure of the system. Ove rpres sure
- in the reactor coolant system =ust therefore be limited to values less than 107. of the system design pressure. This may be accomplished in sev-eral ways, and they include (1) a rapid reducti,n of reactor power in Tesponse to the reduction in power demand, (2) maintaining or restoring 1 power demand by diverting the main stream flow past the turbine generator to a suitable steam dump, (3) actuation of the pressurizer spray in
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pressurized-water reactor systems, and (4) the discharge of reactor cool-meat in the foes of steam or water from the reactor coolant system. In I
- Fractice, all of these means are utilized to control overpressure.
l The primary means of rapidly reducing reactor power in response to j
loss of load is a reactor scram:
the rapid insertion of control rods and j
shutdown of the reactor.
In pressurized-water and boiling-water reactor plants, resccor scram is normally initiated automatically upon turbine trip, high pressure in the reactor coolant system, or high temperature in i
the reactor outlet. Reactor scram can also be initiated manuallya Rapid
g bine restoration of power demand is possible by osans of f art acting tur bypass valves which, upon receipt of a signal indicating that a turbine h tur-trip has occurred, open and divert the main stream flow around t e bine and into the condenser. The,j ressurizer spray in pressurized-water reactor systems is normally detuated by a pressure switch at a preset sys-h driving tem pressure, or the overpressure itself may be uti.1.ized as t e head to effect pressurizer spray.
The final means of overpressure control in both pressurized-water from the and boiling-water reactor plants is by the discharge of fluid lf-system through pressure relief valves, some of which must be se d
actuated at a preset overpressure and some of which may be actuate These pressure remotely by ar, electrical signal at a preset pressure.
in relief valves are normally installed on the top of the pressurizar iling-pressurized-water reactor systems and on the me.in steam lines in bo water reactor systems.
All of the aforementioned overpressure p;otection systems and devices h degree of are installed in most large nur lear plants to provide a hig ble values.
assurance that system overpre,sure will be limited to accepta essen-In addition, the design for these systems is such that they art For example, in the design of pressure relief systems tially redundant.
tly in pressurized-water and boiling-water reactor plants. it is presen f the safety accepted practice to determine the total required capacity o (turbine trip) valves on the bat.is of a complete loss-of-load transient and the assumptions that l does the reacter scram resulting from the turbine trip scram signa high-1 not occur (scram is assumed to occur only_as a result of the pressure or high-temperature scram signal),
h the turbine bypass valves which divert the steam directly to t e 2
condenser malfunction and do not open, t
does not the pressurizer spray in pressurized-water reactor sys ems 3.
function, and hich no credit is taken for the capacity of pressure relief valves w 4.
do not meee the rrquirements of the ASME Code.
In actuality, if any two, and in some cases only one, of these high probability safety features operate as designed (and there is a very
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10 lief valves at least two or more vould operate), the pressure re ditions of the pastulated that vould not be required to open even under the con Thus, the basis for the design of pres-co=plete loss-of-load transient.
tral sta-sure relief systems for reactor coolant systams in present cen This conservative design phi-tion nuclear plants is very conservative.
However, it losophy is considered to be justified for nuclear plants.
is important that all of the reactor safety and control systems which into censidera-provide overpressure protection be undarstood and taken t s E addi-tion when evaluating nuclear everp ' ssure protection sys em.
lf and its response tion, the nuclear characteristics of the reactor itseThese considerations are dis I
to overpressure shculd be considered.
in more detail in Fection 3 of this document.
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11 SPECIAL REQUIREMENTS FOR OVERPRESSURE PROTECTION AND PRESSURE RELIEF VALVE SYSTEMS IN NUCLEAR POWER 3
When evaluating overpre.ssure pr'otection and pressure relief valve i
systems for nuclear power plants, special consideration should be g ven d
t r and to the effects of overpressure on reactor power in pressurize -wa e boiling-water reactor systems and to the loss of coolant and nuclear Attention should also be directed to the isolation of pressure safety.
relief valms and to the code requirements for overpressure protection.
Effect of Overpressure on Reactor Power _
31 As previously discussed, mismatches in reactor power and electrical loads can result in increases in the tac:perature and pressure of the The effects of these temperature and pressure reactor coolant system.
d changes on react ~c power differ considerably in pressurized-water an boiling-water reactor plants and have a significant influence on the cotal* capacity of the pressure relief valves required to limit overpres-The effects ef temperature and overpressure sure to acceptable values.
l on reactor power and the required relief valve capacity for typica h
pressurized-water and boiling-water reactor rystems are discussed in t e following paragraphs.
Pressurized-Water Reactor Systems _
i 3 1.1 f
In contrast to fossil-fueled be U.rs, pressurized-water reactors are f
n T%f b to Ocause of the nuclear character-essentially self-regulating.
A loss of load in the sec-istics of the reactor core. rr
.qerntor.
ondary (steam) system of a pt:3surizu; water reactor results in an An increase in increase in the temperature and pra.ssure of ehe coolant.
veness as a neutron the te=perature of the coolant reduces its effe(
. - reacciens take moderator and reduces the rate at which nuclear
'or power place in the reactor core, thereby automatically redu,.
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12 in response to a load reduction. In contrast, an increase in the pressure 4
of the coolant increases the effectiveness of the moderator and, together with other nuclear reactivity effects, tends to increase reactor power.
However, the dominant effect is the < temperature effect on reactivity, and in the pressurized-water reactor, the net effect of the increase in the temperature and pressure of the reactor coolant is to reduce reactor As a result, even if no action were taken in the event of a com-power.
plate loss-of-load transient, the pressure in the system would peak and l
gradually decrease to a stable equilibrium value.
The magnitude of the peak overpressure in the absence of reactor i
scram, pressurizar spray, or pressure relief valve operation is dependent upon the particular pressurized-water reactor plant itself.
j However, for typical pressurized-water reactor plants with two to four coolant loops and vertical U-tube steam generators, the maximum calculated system pres-sure when no pressure relief valves are provided is usually only 50 to 100 psi greater than the acceptable overpressure which occurs when pres-sure relief valves are provided and do operate. As a result of these con-siderations, the total pressure relieving capacity required for a reactor
. coolant system in a pressurized-water reactor is relatively small and typically amounts to 0 5 to 17. of the total rated steam flow of the plant.
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3.1.2 Boiling-Water Reactor Svstems I
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The effect of overpressure and incressed reactor coolant temperature on a boiling-water reactor is somewhat different than that on a pressurized-water reactor.
Considerable boiling occurs in the core of a boiling-water reactor.
An increase in the system pressure resulting from e loss-of-load transient tends to collapse the steam bubbles or voids,
%ereby increasing the ratio of water to steam in the core.
Since water is a significantly better neutron moderator than steam, the effect of the initial pressure increase is to increase core reactivity.
The increase in moderator temperature that occurs with the pressure increase tends to have a negative influence on reactivity in the same manner as in a
. pressurized-water reactor. However, the effect of the pressure on the 804 170
13 voids in the core is predominant. Bus, the initial response of the boiling-water reactor to an increase in pressure and temperature will be to incr2ase reactor power. This effect will eventually be reversed since, as the temperature of tho' coplant increases, a substantial amount of boil-ing will occur and result in the generation of a large percentage of voids in the core. The eventual increase in the void fraction, together with the moderator temperature e'ffects on reactivity, will reduce the core power to an equilibrium value.
However, to avoid the condition in which a large portion of the core is blanketed by steam and to minimize the resulting risk of core overheat-ing, it is necessary in a boiling-water reactor to carefully control pres-sure increases such as those which accompany a large load reduction.
This task is made difficult by the fact that the initial overpressure tends to increase core power. In addition, in the direct-cycle boiling-water reactor system, the transport time of the reactor coolant from the reactor to the turbine is significantly less than the total transport time from reactor to turbine in the pressurized-water reactor system, where a loss-of-load transient must first be reflected in the steam gen-I erator and then through the primary system to the reactor befers the effect of the power mismatch is significant. As a result, the peak over-pressure resulting from a loss of load will occur much more rapidly in a boiling-water reactor system than in a pressurized-water reactor system.
It is therefore important that corrective action in the form of reactor 1.L scram, actuation of turbine bypass valves, and/or operation of pressure relief valves occur rapidly (typically within a few hundred milliseconds of the loss-of-load incident). his requires fast-response pressure relief valves and a total pressure relief valve dischargepacf ty in the order ofgtotal rated, a g. This total relieving capicity required for the boiling-water reactor plant is significantly greater than that required for a pressurized-water reactor plant of similar size.
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l 14 3.2 1.oss of Coolant and Nuclear Safety One of the most severe accidents in a water-cooled nuclear reactor plant that can be postulated is the loss of reactor coolgnt resulting from an uncontrolled depressurization or " blowdown" of the reactor cool-ant system caused by a piping system rupture or component malfunction.
In the absence of any corrective action, such an incident would result in rapid depressurization of the reactor coolant system with concurrent loss of coolant from the reactor. This loss of coolant in an operating plant could result in rapid overheating and possible meltdown of the reactor core with a possible release of large amounts of radioactive fission products. Accordingly, a significant amount of design, stress analysis, and quality control effort is expended in the design of nuclear plants to assure that a rupture of the primary system will not occur for all conceivable normal and accident conditions. In addition, redundant emer-gency systems are provided to minin.ize the consequences of an uncontrolled blowdown of the reactor coolant system by injection of additional cooling water directly into the reactor vessel and core. Thus, a prima saferf consideration in a nuclear plant is the prevention of uncontrolled loss of reactor coolant during-operation of the reactor.
For these reasons, the overpressure protection systems themselves
=ust be designed so that they do not cause or contribute to uncontrolled blowdown of the reactor coolant system. Specifically, it is of utmost l
impo lear plan [that de pressure relief valves not only open properly if called upon to do so, but that, once opened, they close and reseat properly at the required resent pressure. It is this require-ment for pressure relief valvu systems in nuclear plants that differs substantially from that for conventional fossil-fueled plants and requires l
a different and somewhat more restrictive design philosophy.
Inadvertent opening of a pressure relief valve, failure of a pres-f I
sure relief valve to re-close, qt at the pre 2-**
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.une= a t % 2 m u u sr A co r=9tureof e s which is non-isolable from the reactor vessel and core. Unlike the
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15 design of pressure relief valves for conventional plants, egal attention must be given to the pressure integrity and reliability in_re-closing _.ni pressure relief valves for nuclear systems as is heircapabig ity to open properly. JBecause of the nu=ber of redundant systems that provide overpressure protectida in nuclear plants and because of the fact that there is no way to provide redundancy to protect against a failed-open pressure relief valve, it may be argued that the pressure integrity If i
and capability of a pressure relief valve to close and remain closed is of more importance in a nuclear plant than its ability to open. This is especially so if the design of the pressure relief system and other i
nuclear safety sy.tems can tolerate failure of any one pressure relief valve to open.
As a result of these considerations, the design, material, manufac-turing, tet ting, and quality assurance requirements for pressure relief valves used in nuclear systems must assure that all functional and pres-sure retaining parts are designed and manufactured to a quality level con-sistent with that for other critical c.omponents in the reactor coolant system that are directly connected to the reactor vessel. The manufac-turers of pressure relief valves for nuclear service are required to pro-vide more sophisticated design analyses, more extensive nondestructive examination and testing, and more restrictive quality assurance require-l ments than are normally required for pressure relief valves manufactured for conventional service. This is especially true for the pressure-g retaining parts of spring-loaded safety valves, such as the main spring, j
the external spring retention structure, and bolting, as well as other l
i parts not actually in contact with the reactor coolant but whose failure l
could lead to uncontrolled loss of reactor coolant through the valve.
l 3.3 Isolation of Pressure Relief Valves The i=portance of designing and manufacturing pressure relief valves so that they will not cause or contribute to an uncontrolled blowdown of the reactor coolant system was stressed in the foregoing discussion. One means of attaining additional protection against this eventuality is to 804 173
I u
16 install isolation valves between the reactor and the pressure relief I
valves. These isolation valves would normally be open under all oper-l ating conditions. However, if a pressure relief valve were to inadver-i tently open or otherwise fail and not reseat, its isolation valve could be remotely or manually closed to ILnit the loss of reactor coolant to an acceptable amount.
However, the installation of isolation (stop) valves between the reactor coolant system and the pressure relief valves is prohibited by Section III of the ASMZ Boiler and Pressure Vessel Code unless, as stated it. Article NB-7153 of Section III, the isolation valves are " constructed and installed with positive controls and interlocks so that the relieving capacity requirements... are met under all conditions of operation of both the system and the stop valves". This provision can be met if the isolation valve actuating device or controls are normally locked open or otherwise positively controlled and if valve position detectors and inter-locks are provided to preclude plant operation in the event that any of the isolation valves are closed.
Thus, it is permissible to provide isolation cannh411ty_for pressure
~
s
+
s relief valves against loys ant resulting frgg, malfunc31r,a-af,~these vah However, in large nuclear power plants
~ ~. -
which have a large number of pressure relief valves installed inside the reactor containment structure where they are inaccessible during reactor 1
operation, the remote valve operators, interlocks, and controls required I
to permit installation of isolation valves for each pressure relief valve I
awould add a significant amount of complexity to the reactor safety sys-Therefore, isolation valves are not normally used in the large tems.
commercial nuclear plants built in the United States.
An alternative method of obtaining isolation capability for pressure trelief valves without using the controls.nd electrical interlocks pre-viously discussed is to provide essentially two independent pressure relief valve systems, each of which would provide the required relieving capacity, and to use isolation valves in each system that are mechanically interlocked so that it would not be physically possible to isolate both pressure relief valve systems at the s a=e
'.1=e. An example of such a system is one in which a three-way isolation valve is installed upstream 804 174
s 17
.of two pressure relief valves, one of which is redundant. The design for the three-way valve must be such that only one of the pressure relief valves can be isolated at any one time; that is, the three-way valve must have no center-off position. A system similar to this is installed in the nuclear-powered merchant ship NS Savannah. Obviously, such a system is practical in only those plants which require a relatively small number of pressure relief valves. This type of system has not been used in large central station nuclear generating plants.
3.4 Code Recuirements for Overpressure Protection The reactor vessel and all components in the reactor coolant system are designed, manufactured, and tested in accordance with the rules of Section III, Nuclear Power Plant Components, of the ASME Boiler and Pres-sure Vessel Code. Requirements which must be met by component manufac-turers are stipulated in this code. It also requireg that the plant owner provide pressure relieving devices to protect the components in i
service from steady-state and transient pressure conditions in excess of i
the pressure for which they are intended. The requirements for overpres-5 sure protection are specified in Article NB-7000, Protection Against Overpressure, of Section III. These requirements are discussed briefly in the following paragraphs.
3.4.1 Nuclear Power System Requirements All nuclear power systems designed in accordance with the rules of Section III of the ASME Boiler and Pressure Vessel Code are required to have installed a total rated pressure relieving capacity sufficient to prevent the pressure within the system from rising more than 107. above the design pressure of the system, taking into account any pressure drops occurring at full flow, any back-pressure effects, and the operation of other protective systems which complement the pressure relief valves.
The set pressure of at least one of the installed pressure relief valves 804 175
18 must be not greater than the design pressure of the system. The pressure relief valves must be installed so that the total relieving capacity is never less than that required to meet the rules of Article NB-7000 of S',ction III of the AS*E Boiler and, Pressure Vessel Code. If stop valves are installed in the pressure relief lines, positive controls and inter-locks are required to prevent reduction of the available relieving capac-ity below thae required.
The provisions of the design for the.anclear power system that are intended to meet the requirements for overpressure protection set fo.th
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in Article NB-7000 of Sectica III are required to be the subject of a report on everpressure protection prepared by or on behalf of the owner of the plant.
'l'ne overpressure protection system and how it complies with the requirements of Article NB-7000 of Section III must be described in this report. In general, the areas to be covered in this report are 1.
a description of the overpressure protection systems, ircluding the i
type and number of pressure relief valves, their total rated capacity, the valve set pressures and other valve design parameters such as blowdown, accumulation, etc.;
.2.
the basis for the design of the overpressure protection systems, including a description and summary of results of analyses which justify the adequacy of the pressure relieving capacity provided and the basis for selection of the system transient condition which led
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to the design overpressure; 3.
a description of other safety systems which supplement the pressure relief valve systems in the provision of overpressure protection and I
the specific assumptions made relative to the operation of these sys-tems during the design overpressure conditions; and i
4.
capacity ce'rtifications for the pressure relief valve designs used.
3.4.2 Pressure Relief Valve Requirements Article NB-7000 of Section III also places requirements -on the types, marking, and certification of pressure relief valves i.talled in nuclear e tergy systems for overpressure protection. The three types of pressure 804. 176
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19 relief valve most commonly installed in nuclear power systems are (1) the spring-loaded safe.ty valve, (2) the pilot-operated pressure relief valve, and (3) the power-actuated pressure relief valve.
1.
The spring-loaded.salety. valve is essentially an angle valve in which a plug is held in plac5 against the system pressure by a spring.
The valve sit pressure is determined by the spring preload; that is, the pressure load at set pressure.
2.
The pilot-operated pressure relief valve is essentially an angle valve or a globe valve in which a plug seals against system pressure and is held in place by system pressure (a reverse-seated disk).
The set pressure is determined by a pilot valve which senses the system pressure, and when opened by the system pressure, the pilot valve hydraulically actuates the main valve using system pressure.
3.
The power-actuated pressure relief valve is essentially the same as a pilot-operated pressure relief valve except that the set pressure is determined by an external control aircuit which actuates the pilot valve, which La turn actuates the main valve hydraulically using the system pressure.
P,rior to issuance of the Summer 1969 Addenda to Section III of the ASHI Boiler and Pressure Vessel Code, only self-actuated pressure relief valves were considered acceptable for use as part of the Code-required overpressure protection.
However, the use of self-actuated valves, such
- as spring-loaded safety valves (including those equipped with anti-siccer devices) and pilot-operated pressure relief valves, as well as power-actuated pressure relief valves equivalent to spring-loaded safety valves is now permitted by Section III, subject to the following requirements.
1.
Spring-loaded safety, valves must be opened by direct action of the fluid pressure in the protected system against the spring.
Balanced spring-loaded val.as must be equipped with a means to verify the,integ-rity of the back-pressure balancing device. Valves equipped with anti-sibmer devices that raise the seat load during normal operation are sub-ject to the following additional requirements.
The main valve must open automatically at the set pressure a.
by self-actuation even if any component cf the anti-st= mar devi~ e should c
fail.
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The main valve must revert to its normal set pressure on a
any failure of the energy or signal sources.
The auxiliary load of the device must not raise the set pres-c.
sure of the valve by more than 10L of the system design pressure.
d.
The auxiliary load must be automatically removed at the set pressure of the valve.
2.
The pilot control devices of pilot-operated pressure relief valves must be actuated directly by the fluid pressure of the protected system. Where operation of the pilot control device depends on the integrity of a pressure sensing element, such as a bellows, means cust be provided for detecting a failure of the sensing element. In addition, the main unloading valve must open wid2 a full opening pop action in direct response to the operation of the pilot valve. The main unloading valve end the pilot control device, treated as a combination, must meet all the requirements applied to the pressure relief valve.
3.
Power-actuated pressure relief valves which depend upon an external energy source for their operation must be capacity certified in the same manner as self-actuated pressure relief valves. To be able to take credit for the relicving capacity of these valves to the extent per-mitted by Section III, the control system and the external energy sources for valve operation must have redundancy and independence at least equal to that required for other control and safety circuits associated with the nuclear energy system. Further, the main unloading valve and its auxiliary devices, treated as a combination, must meet all Code require-ments applicable to pressure relief valves.
l
_3.4.3 Rated Capacity and Certification Requirements Each pressure relief valve design uJed must be flow tested and its capacity must be certified. However, the percentage of the certified capacity of each valve type that may be credited toward the total rated Telieving capacity of the system differs slightly among the individual types of pressure relief valves.
B04 170 7-
21 The raced capacity of a given valve or valve design is determined and certified in accordance with Article NB-7800 of Section III of the ASME Boiler and Pressure Vessel Code. This article requires that capac-ity certification tests be ' performed, using saturated steam, at a test facility which has been approved by the ASME Boiler and Pressure Vessel Comittee and in the presence of an abserver a sthorised by this comittee.
The article contains detailed. procedural steps that must be complied with to obtain official capacity certification. However, when applied to typ-ical pressure relief valve designs used in large central station nuclear planta, these test procedures mst be modified to reflect the fact that there are no ASME-approved test facilities with sufficient capacity to test a large nuclear pressure relief velve. Modifications to these requirements that are normally made to pemit capacity certification of large pressure relief valvas include (1) testing at substantially reduced flows and pressures with extrapolation of t.he test data to higher pres-sures and (2) testing of reduced-scale models of valves that duplicate the internal flow passage configuration and lift o'f the full-scale valves.
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4.
PRESSURE RELIEF VALVE SYSTEMS USED IN WATER-COOLED REACTOR PIANTS Numerous safety systems and controls are iucorperated in the light-water-cooled reactor plants built in the United States to effectively control or ILait overpressure excursions resulting from reactor power and load mismatches. The system of interest in this discussion is the pres-sure relief valve system which consists of one or more of the three basic types of pressure relief valve described briefly in Subsection 3.4.2.
These pressure relief valves may be self-actuated by the pressure in the i
reactor coolant system, or they may be remocely actuated by an externally supplied signal. The main design features and operating characteristics of these three types of valves are described in detail in Section 5 of this document. The nature of the various pressure relief valve systems employing these three cypes of valves that have been or are being used in commercial pressurized-water and boiling-water reactor plants are f
described in the following subsections.
4.1 Pressure Relief Valve Systems in Pressurized-Water Reactor Plants 1
A typical pressure relief valve system for a pressurized-water reactor is illustrated in Fig. 4.1.
As indicated in Fig. 4.1, the pres-i sure relief valves of a pressurized-water reactor system are mounted on i
.short headers located on top of the pressurizar. This is permissible in i
plants in which it is not physically possible to isolate the pressurizer from t'= reactor coolant system which is protected by the pressure relief In plants where isolation valves are used in the surge line to i-.ves.
the pressurizer, it is necessary to provide separate overpressure protec-tion for the reactor coolant system. However, pressurizer isolation valves are not normally provided in large central station plants, and it is current practice in the design of pressurized-water reactor plants to install all of the required pressure relief valves on the pressurizer.
Rith this type of installation, the effect of the pressure drop in the 804 IB0
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24 surge line to the pressurizer during discharge of the pressure relief 4
valves cust be considered when selecting the set pressures and required i
relieving capacities for the, pressure relief valves.
As illustrated in Fig. 4.1, the outlets of the pressure relief valves are typically piped to a cocmon header from which a single dis-charge line is directed to the quench tank.
This closed discharge system is used to prevent unnecessary contamination of the reactor containment in the event of valve operation or leakage. The relief valve. discharge is admitted to the quench tank, which is partially filled with water.c ambient temperature, through submerged spray nozzles to quench the steam and minimize pressure buildup in the tank during discharge of the relief valves.
The quench tank and connected piping may be pressurized to a pressure slightly in excess of atmospheric pressure with an inert gas, such as nitrogen, for corrosion control.
The quench tank and connected piping are protected from overpressure by a small pressure relief valve and/or rupture disk.
This valve and/or rupture disk is vented directly to the reactor containment.
However, a
.more detailed discussion of the overpressure protection system for the t;uench tank is not within the scope of this document.
The pressure drep in the discharge line or lines during valve oper-
.ation will result in a dynamic back pressure on the pressure relief
- valves.
Unless these back pressures are accounted for in the valve desiga, static back pressure will alter the pressure at which the valve opens and the dynamic back pressure will affect the operational charac-teristics such as capacity, blowdown, and accumulation.
Because of these back-pressure effects, the types of pressure relief valves used in j
. pressurized-water reactor systems are those that are not significantly
.affected by static or dynamic back pressure.
These types include back-gressure-balanced spring-1saded safety valves, power-actuaten pressure relief valves, and pilot-operated pressure celief valves.
In all of these types, the valve bonnets, main spring and spindle, and other actu-
-ating devices are suitably enclosed or sealed to prevent leakage to the atmosphr e.
A variation in the pressure relief valve system described in the pre-cuding paragraphs that is being used in some pressurized-water reactor B04
25 plants involves the use of a water seal upstream of the pressure relief valves. Radiolytic decomposition of water in the core produces free oxy-gen in the reactor coolant system. In pressurized-waterweactor systems, free hydrogen (approximately 15 to 20 ce/kg of water) is introduced into s
the system to reduce the oxygen concencration.
It is widely accepted that this free hydrogen, which tends to accumulata in the steam space of the pressurizar, leaks throug' h the spring-loaded safety valves, deterio-rates the valve seating surfaces, and results in steam leakage. Although the deteriorative effect of hydrogen on valve seating surfaces has not been demonstrated, the inability of spring. loaded safety valves to hold hydrogen and the frequent steam leakage from spring-loaded safety valves at pressures above approximately 857. of the valve set pressure have been demons trated. Therefore, to avoid possible valve leakage difficulties, water seals are being provided in a number of pressurized-water reactor sys tems.
As illustrated in Fig. 4.2, a water seal is a valve inlet piping arrangement which permits water to collect on the upstream side of the rtttsutt title 7 faLVE ITte-4 r
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26 valve seat. The water may be condensed from the steam by cooling the pipe or it may be introduced by a water feed line. Some pressurized-water reactor suppliers and utilities believe that these water secls sig-cificantly reduce both hydrogen and steam leakage.
4.1.1 Systems in Existing Plants Pressure relief valve overpressure protection systems currently in use in typical pressurized-water reactor plants consist of spring-loaded safety valves and power-actuated pressure relief valves mounted on the
-top of the pressurizer steam space. In typical pressurized-water reactor plants with an electrical output of 800 to 1000 MW, two or three spring-loaded safety valves with capacities in the range of 250,000 to 600,000 lb/hr are used together with one or two power-actuated pressure relief valves with a total capacity of less than half the capacity of the installed safety valves.
The total relieving capacity required by Section III of the ASME Boiler and Pressure Vessel Code is provided by the spring-loaded safety valves because the capacity of the power-actuated pressure relief valves could not be included in the capacity required by the Code prior to issu-rance of the 1969 Su=mer Addenda to Section III. Ihe purpose of the power-actuated pressure relief valves, which are normally set at a pres-
.sure somewhat below the set pressures of the safety valves, is to limit the system pressure to a value which will minimize si2::mering and leakage of the safety valves during normal operation and which will prevent open-
.ing of the safety valves for all but the most severe plant transien ts.
Typical set pressures for spring-loaded safety valves and power-actuated pressure relief valves currently used in central station pressurized-
-uater reactor plants are tabulated below. The pressure at which opera-tion of the pressurizer spray is normally initiated is also givea.
Sys tem Design pressure 2485 psig Normal operating pressure range 2050 to 2250 psig Set pressure range Spring-loaded safety valve 2485 to 2565 psig Power-actuated pressure relief valve 2335.to 2400 psig Pressurizer spray actuation range 2200 to 2300 psig
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27 The total prescure relief valve capacity required to satisfy the rules of Section III for a given pressurized-water reactor plant is nor-e mally determined by means of complex digital computer analyses which account for the nuclear characteristics of the reactor core, the dynamics s
of the reactor coolant system (temperature, flow rates, pressure drops,
. ate.), the conditions in the staara generators and secondary (steam) sys-tem, and the operational characteristics of the safety valves (set pres-sures of individual valves, accumulation, and blowdown). The various system conditions which lead to overpressure (discussed in Section 2) are analyzed for several assumed safety valve capacities. Based on the results of these detailed analyses, the total safety valve capacity is selected to limit the maximum pressure in any part of the system to no more than 1107. of its minimum design pressure. The nu=ber and capacity of the individual safety valves are then selected on the basis of commer-cially available valve sizes and economic considerations.
As discussed in Subsection 2.2, the total pressure relief valve
,~
capacity necessary to meet the requirements of Section III of the ASMZ Boiler and Pressure Vessel Code is in most cases based on a postulated complete loss-of-load transient and the pessimistic assumptions that reactor scram resulting from turbine trip, actuation of turbine bypass valve, initiari m of pressurizer. spray, and operation of the power-actuated relief valves do not occur. When sizing power-actuated relief valves, it is normally assumed that these other systems do operate as designed. Thus, the capacity of the power-actuated valves is set at a value which will limit the pressure of the system to values below the set pressures of the safety valves required by Section III.
The total capacity of Code-required pressure relief valve systems as a function of the electrical power output of a number of representative pressurized-water reactor plants is illustrated in Fig. 4.3.
A clear relationship between the total relief valve capacity and the electrical pdwer output for each of the pressurized-water reactor plants is shown.
However, this illustration is presented only for informational purposes and should not be used as a basis for sizing the pressure relief valve capacity of a particular pressurized-water reactor plant.
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Total Capacity of Code-Required Pressure Relief Valve i
v ISystects as a Function of Electrical Power Output for Various Pressurized-
-Mater Reactor Plants.
4.1.2 Features of Systems Considered for Future Plants The pressure relief valve systems that have been used in existing central station pressurized-water reactor plants to date are essentially the same. Alternative systems or features that are under consideration
29 for future plants or are presently offered by suppliers include pilot operated pressure relief valves, power-actuated pressure relief valves,
't and anti-si==er devices.
(a) Pilot-operated Preseure Relief Valves. The use of pilot-s operatcd pressure relief valves is permitted by Section III of the ASME Boiler and Pressure Vessel Code, and this type of valve has the potential for providing improved leak tightness. Since pilot-actuated pr Jsure relief valves can be both self actuated and remotely actuated, they.can be used to perform the functions of both the spring-loaded safety valves and the redundant power-actuated pressure relief valves presently spec-ified for pressure relief valve systems in pressuri=ed-water reactor sys-An optional overpressure protection system which employs pilot-tems.
operated pressure relief valves has been offered by one supplier of pressurized-water reactor systems.
(b) Power-Actuated Pressure Relief Valves.Section III of the ASME Boiler and Pressure Vessel Code was modified in the summer of 1969 to permit the use of power-actuated pressure relief valves for overpressure protection of nuclear energy systems if (1) redundancy of the control circuits equal to that of other reactor safety systems is provided and (2) credit is taken for only one-half of the relieving capacity of the l
installed power-actuated valves. It is therefore possible to reduce the number of spring-loaded safety valves in a plant by meeting the Section-III requirements for installed power-actuated pressure relief valves.
Such an overpressure protection system would permit some reduction in the total capacity provided by spring-loaded safety valves. However, such a change in the present pressurized-water reactor system may not be eco-nemically attractive accause of the cost involved in the provision of redundant control circuits and valves.
(c) Anti-Simmer Devices. Anti-simer devices are now being offered by manufacturers of pressure relief valves as a means of reducing or eliminating leakage and si=mering of spring-loaded safety valves, and the use of such devices is now permitted by Section III.
It is therefore possible that these auxiliary devices will be used in future pressure relief valve systems which employ spring-loaded safety valves.
804 187
30 4.2 Pressure Relief Valve Svstems in Boiling-Water Reactor Plants The pressure relief valve systems used in boiling-water reactor plants prior to and af ter issuance of Section III of the ASME Boiler and Pressure Vessel Code are discussed in the following subsections.
t 4.2.1 Pre-Section-III Overpressure Protection Sys tems i
The pressure relief valve systems used in boiling-wa*.er reactor plants designed prior to the introitetion and use of the first edition of Section I!I cf the ASNE Boiler and Pressure Vessel Code censisted of a large nu=ber of spring-loaded safety valves to provide the total pres-sure relieving capacity required for r.he system and a smaller number of power-actuated pressure relief valvr.s to supplement the spring-loaded i
I safety valves. The pressure relie.f valves in a boiling-water, reactor system, which does not use a prassurizer as do pressurized-water reactor l
systems, are installed on the main steam lines inside the containment vessel immediately adjacent to and non-isolable from the reactor vessel, as illustrated in Fig. 4.4.
The total capacity provided by the spring-loaded safety valves in a pre-Section-III boiling-water reactor plant is approximately 100 to JS07. of the total rated steam flew of the plant, and it is obtained through the use of a suitable number of the largest commercially avail-able spring-loaded safe ty valves (typically, fif teen to twenty 600,000-lb/hr valves). Because of the large pressure relief capacity required for the boiling-water reactor system and the small probability for oper-ation of the spring-loaded safety valves (estimated by designers to be ene operation per 40-year plant life), these safety valves discharge
.directly to the containment vessei (referred to as the dry well). Since 1
the spring-loaded safety valves are not connected to a closed discharge system, static and dynamic back-pressure effects are insignificant and l
these valves are not provided with back-pressure balancing devices. Nor is there a need for a bellows or other seal to limit or contain leakage of steam past the valve spindle during valve discharge or seat leakage
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Typical Pressure Relief Valve System for a Boiling-Water Reactor Plant.
because the valves discharge directly to the dry well. Therefore, the exposed-spring unbalanced type of spring-loaded safety valve, described in Section 5 of this document, is normally used in boiling-water reactor systems.
The power-actu ated pressure relief valves used in the pressure relief valve systems of boiling-water reactor plants designed prior to the issu-ance of Section III serve to minimize pressure surges that might otherwise actuate the spring-loaded safety valves or cause sirmering of the safety I
valves. These valves are electrically actuated in response to a signal from a remote pressure sensing element, and their total capacity is typ-ically about 307. of the total steam flow of the plant. However, as is the case for overpressure protection systems in pressurized-water reactor planta, no credit is taken for the capacity of the power-actuated pressure relief valves used in boiling-water reactor plants when evaluating the hk
32 capacity requirements of Section III; that is, it is conservatively assumed in the analysis of the design loss-of-load transient that these valves do not operate.
Operation of the power-actuatsd pressure relief valves is relatively frequent when compared with operation of the spring-loaded safety valves, and the discharge from these valves is piped from their location on the main steam headers in the dry well to a vapor suppression pool within the The vapor suppression pool is partially filled with containment vr lume.
Iha dis-water and is t innected to the dry well by large pipes or ducts.
charge lines from the power-actuated relief valves pass through these ducts and discharge alow the water level in the suppression pool where the discharged steam is condensed. Back-pressure effects resulting from the pressure drop in the discharge lines and the head of water over tha outlet of the discharge piping c:ust be considered for the power-actuatei f
relief valves.
The significant characteristics of a typical overpressure protection system for a boi?.ing-water reactor plant (Oyster Creek No.1) designed prior to the issuance of Section III of the ASME Boiler and Pressure Ves-sel Code are given in Table 4.1.
Table 4.1.
Characteristics of a Typical Pre-Sec'. ion-III Overpressure Protection System for a Boiling-Water Reactor Plant (Oyster Creek No. 1)
System 1250 psig Design pressure, 1000 psig Normal operating pressure, Spring-loaded safety valve 1210 to 1240 psig Set pressure, 16 Nurbe r,
634,000 lb/hr Individual capacity, Power-actuated pressure relief valve 1025 psig Set pressure, 4
- Number, 600,000 lb/hr Individual capacity,
- * ~
I 33 Section-III Overpressure Protection Systems 4.2.2 In boiling-water reactor systems designed in accordance with the d
the rules of Section III of the ASME Boiler and Pressure Vessel Co e, l been function of power-actuated pressure relief valves has, in genera,
i l ded safety replaced by pilot-operated pressure relief valves or spr ng-oa As in the pre-Section-III systems, valves equipped with remote operators.
ontainment on the d
h the pressure relief valves are installed insi e t e c ble from the reac-main steam lines imediately adjacent to and non-isola tor vessel.
The total installed pressure relieving capacity in boiling-water Section III is reactor systems designed in accordance with the rules of the issuance of significantly less than that in systems designed prior to l
The primary reason for this reduction is the assumption Section III.
t ms that made by designers of Section-III boiling-water reactor sys e nying the reactor scram occurs as a result of the pressure surge accompa Ihis assumption, which vas not made in design loss-of-load transient.
t ms, is consid-the design of pre-Section-III overpressure protection sys e f ty systems, ered by designers to be warranted because there are many sa e s
including reactor scram, in nuclear plants that contribute to overpre -
fact is recognized sure protection of the reactor coolant system and this in Section III of the ASME Boiler and Pressure Vessel Code.
i l
The pertinent characteristics of the overpressure protection systems t
l j
for three Section-III boiling-water reactor plants are given in Tab e i
The considerable increase in the total pressure relieving capac ty l a is related 4.2 of the Brown's Ferry plant over that of the Millstone p ar d
line. This
)
to the increased core power density of the 1967 pro uctlts in an increase in the system-power-to-system-volume ratio resu h
(A increased sensitivity of the system pressure to power mismatc es.
sim-reduction in rjstem volume for a given system power would produce a ilar effect.)
i The considerable increase in the total pressure relieving c.spac ty l
is related to of the Zi=er plant over that of the Brown's Ferry p ant rators inste ad of the use of spring-loaded safety valves with remote ope
e Table 4.2.
Characteristics of Overpressure Protection Systems for Three Section-III Boiling-Water Reactor Plants 1965 Product Line 1967 Product Line 1959 Product Line Plant H111 stone Brown's Ferry Zimmer Power output, HW(e)
~650
~1100
~1100 Design pressure, psig 1250 1250 Operating pressure, psig 1035 1015 Required relief capacity,
~30
~63
~63 i
% of steam flow Source of relief capacity 1hree 800,000-lb/hr Eleven 800,000-lb/hr Eleven 900,000-lb /hr pilot-operated relief pilot-operated relief spring-loaded safety valves set at ~ 1100 pst valves set at ~ 1100 psi valves with remote oper-ators set at ~ 1210 to y
1250 psi Required safety capacity,
~10
~13
~40
% of steam flow not including relief capacity l
Source of safety capacity Two 650,000-lb/hr Two 900,000-lb/hr or Five 900,000-lb/hr spring-loaded safety three 600,000-lb/hr spring-loaded safety valves set at ~1230 psi spring-loaded safety valves set at ~1210 valves set at ~1230 psi to 1250 pai Remarks liigher power and density Approximately same power than and approximately density and system vol-same system as Hillstone uma as Brown's Ferry i
i i
I cya CD
.p=
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35 pilot-actuated pressure relief valves. Although these spring-loaded safety valves are operated remotely at a pressure of 1100 psi, for the
.se of satisfying the requirements of Section III, they are not assumed to open until actuated by system pressure at their set pressures of 1210 to 1250 psi. The nature of a boiling-water reactor system is such that an overpressure transient tends to be initially self-perpetuating.
Therefore, staps to halt an overpressure transient become less effective with the passage of time after initiation of the transient. For this reason, the opening of 11 pressure relief valves at pressures of 1210 to 1250 psi is less effective than the opening of 11 pressure relief valves of slightly less capacity at a pressure of approximately 1100 psi. Addi-tional spring-loaded safety valves are therefore required.
The spring-loaded safety valves used in current boiling-water reac-tor systems designed in accordance with the rules of Section III are exposed-spring unbalanced safety valves, and they discharge directly to the dry well as in pre-Section-III pressure relief valve systems. The pilot-operated valves and the remotely operable spring-loaded safety valves serve a dual function. First, they provide cost of the total plant pressure relief capacity required by Section III. Second, they minimize pressure surges associated with normal plant operation in the same manner as the power-actuated pressure relief valves used in pre-Section-III boiling-water reactor plants and in pressurized-water reac-tor plants. In effect, the pilot operated valus and the remotely oper-able spring-loaded safety valves in Section-III boilirg-water reactor systems replace the power-actuated pressure relief valves previously used and eliminate a like capacity of conventional spring-loaded safety valves. As for the power-actuated relief valves used in pre 3ection-III boiling-water reactor plar*?.s, the discharge from the dual-function valves l
is piped to the vapor suppression pool. Accordingly, back-pressure f
effects c:ust be considered in the design and procurement of these valves.
An additional feangre of the pressure relief system described in this subsection not directly related to system overpressure protection is its use in some boiling-water reactor planen a
.n engineered safe-guard system.
It is accepted practice in the design of reactor systems to provide auxilitty cooling and water injection or spray systems to cool I
\\9 a
e 36 the reactor core in the unlikely event of a piping system rupture that is followed by the loss of reactor coolant. For certain sizes of pipe rupture in some boiling-water reactor plants, the rate of fluid loss would exceed the normal makeup capability of the system but would not be rapid enough to cause the reduction of system pressure necessary to per-l mit the low-pressure emergency core spray systems to function when re quired. In such a situation, the core could become uncovered and over-heat. To prevent core overheating in the event of a rupture in the reac-tor coolant syster, the coolant system is automatically vented to the containment by means of the pressure relief valves. This is accomplished
.I in the previously described Section-III overpressure protection system i
by pilot-operated pressure relief valves or spring-loaded safety valves equipped with auxiliary air operators which allow them to be remotely actuated at any system pressure (above a few hundred psi). This function,
{
which is referred to as the automatic blowdown function, does not affect the capability of the valves to operate by self-actuation as pressure relief. valves if required for plant overpressure protection.
8 i
i
37 i
1 5.
DESIGN FEATURES AND OPERATING CHARACTERISTICS OF THREE BASIC TYPES OF NUCLEAR PRESSURE RELIEF VALVES There are three basic types of pressure relief valves that are used for overpressure protection in nuclear plants. As outlined briefly in Subsection 3.4.2, these are (1).the spring-loaded safety valve, (2) the pilot-operated pressure relief valve, and (3) the power-actuated pressure relief valve. The main design features and operating characteristics of.
these three types of valves are discussed in this section.
5.1 Spring-Ioaded Safety Valves The spring-loaded safety valves used in nuclear power reactor sys-tems are similar to the safety valves tradicionally used in steam boiler and petroleum industry applications. Two design variations of the spring-loaded safety valve that are used in nuclear energy systems are the exposed-spring type and the enclosed-spring (sealed bonnet) type. The mechanical design and basic operation of these two types of spring-loaded valves are discussed in the following parag aphs.
5.1.1 Mechanical Design i
The exposed-spring type of spring-loaded safety valve is illustrated in Fig. 5.1, and the enclosed-spring type is illustrated in Fig. 5.2.
As is apparent from these illustrations, the mechanical designs of the two valve types are somewhat different. The design of the exposed-spring i
safe ty valve, illustrated in Fig. 5.1, is such that some leakage between the disk holder and the disk guide to the ambient can occur during valve discharge. Valves of this type are used in boiling-water reactor plant applications in which the safety valves discharge directly to the reactor containment volume. In this application, slight leakage past the spindle to a bient is insignificant when compared with the normal discharge of fluid to the containment.
804 195
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The design of the enclosed-spring sealed bonnet safety valve, shown in Fig. 5.2, differs from that of the exposed-spring salve in that the
=ain valve spring is totally enclosed by a sealed bonnet and a back
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Enclosed-Spring Type of Spring-Loaded Safety Valve.
pressure balancing bellows is included to eliminate the effect of b mk pressure on the set pressure of the safety valve. Valves of this type are used in pressurized-water reactor plant applications in which the D
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N 40 discharge from the safety valver is piped to a quench tank for contain=ent of the discharged steam to minimi=e contamination. Consistent with this objective, a bonnet is provided to enclose the safety valve spring so that any leakage past the spindle during
- valve discharge will be contained.
In the pressure relief system of a pressurized-water reactor plant in which the safety valves are piped to a closed discharge system, these valves may be subjected to either dynamic or static back-pressure changes.
A discharging safety valve maybe subjected to dynamic back pressure caused by the pressure drop in the downstream piping. This back pressure occurs only during operation of the safety valves, and it is a function of the i
valve flow rate, flow resistance in the discharge system, andvalve design.
i Dynamic back pressure primarily affects the performance characteristics of the valve such as its capacity, accumulation, and blowdown.
Static back pressure may occur as a result of overpressure of static gas in the l
quench tank and discharge system or from the discharge of other safety 1
valves on a coz:: mon header and discharge line to the sa=a tank.
S ta tic back pressure affects the operational characteristics of the valve and reduces valve lif t pressure by the amount of change in back pressure.
Accordingly, the enclosed-spring safety valves used in pressurized-water reactor systems are provided with a back-pressure balancing bellows to prevent the downstream pressure from acting over the top of the disk
,and altering the valve set pressure.
In addition to this balancing bel-lows, a back-pressure balancing piston is normally provided to eliminate the effects of back pressure on.the valve set pressure in the event of bellows failure.
Section III of the ASME Boiler and Pressure Vessel Code requires that balanced safety valves be provided with this supplementary back-pressure balancing device, and it also requires that this type of valve le equipped with a means, such as a pressure switch, to verify the integ-rity of the bellows and to warn the plant operator of any bellows failure.
5.1.2 operatton The basic operation of both the exposed-spring and the enclosed-spring (sealed bonnet) types of safety valve is identical.
This cpera-tien is described by using the si=plified diagram shown in Fig. 5.3.
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Operational Diagram of Spring-Loaded Safety Valve.
Prior to valve lif t, reactor coolant in the form of steam is contained in the inlet nozzle (L) of the valve. This steam exerts an upward force on the vs1ve disk (B), which is held closed by the spring (I) pre-load act-ing through the valve spindle (K).
The spring pre-load is transmitted to the spindle through the lower spring retaining washer (J), and it is main-tained by the upper spring retaining washer (G) and the set pressure adjust =ent screw (F). The set pressure adjustment screw is anchored to the inlet piping through the spring retention structure (H) and the valve D
D) vol BD
I 42 body (E). Although only the inlet nozzle and the disk are actually in contact with the high-pressure inlet fluid, the inlet nozzle, disk, spin-die, main spring, spring retaining washers, set pressure adjustment screw, spring retention structure, valve body, inlet flange bolting, and the bolting between the body and the spring retention structure all maintain the inlet pressure boundary and are therefore pressure retaining parts.
As the system pressure increases, the valve seating force decreases until it becomes zero at the set pressure of the valve. This is expressed
=sthematically as S = F - pA,
97here S = seating force, I = spring pre-load, p = system pressure, and A = effective pressure area at the seating surface.
Therefore, at the valve set pressure, F = pA.
Small quantities of steam then begin to escape into the huddling chamber (C), shown in Fig. 5.3.
~
The huddling chamber is the open annular region between the nozzle (L) and the nozzle ring (D) at.the bottem and the disk (B) at the top.
Flow frem I
the huddling chamber is restricted by the adjusting rings in a manner that raises the dynamic pressure in this region above the back pressure
.in the valve.
The additional upward load *against the valve disk resulting from the tuddling chacber pressure acting against the disk upsets the force balance' between the spring pre-load and the system pressure and causes the valve l
co " pop" open to a new equilibrium position. From observation of typical mlve openings, the initial equilibrium position of the valve disk is at l
-ac ou t 707, of full lif t.
The additional force necessary to obtain the l
- full-flow lif t is produced by the flow reaction forces caused by the steam i= pinging on the disk and being diverted outward and downward. These dynamic flow and pressure forces cause the valve to attain full lif t and to reach rated flow within a pressure 37. above the valve set pressure. A typical plot of valve lift as a function of pressure for spring-loaded safety valves is illustrated in Fig. 5.4.
804 200
l i
44 i
2.
Accumulation. When the valve opens, the adjusting rings divert the discharge in such a manner that the flow reaction forces acting on the valve disk open the valve "to. full lif t.
Manipulation of the adjust-ing rings will increase or decraase lhe flow reaction forces acting on
.j the valve disk, thereby controlling the accumulation of the valve.
t l
(According to valve manufacturers, the upper adjusting ring on existing I
valvas, such as those illustrated in Figs. 5.1 and 5.2, has the greater ef fect on valve accumulation.)
3.
Blowdown. As in the case of valve accumulation, blowdown' of the valve is a function of the flow reaction forces acting on the valve disk, and manipulation cif the adjusting rings will increase or decrease the blowdown.
(According to valve manufacturers, the upper adjusting ring on existing valves also has the greater affect on valve blowdown.)
Movement of either of the adjusting rings affects simmer, accuc:ula-tion, and blowdown. Therefore it is not possible to independently adjust for any one desired characteristic. In practice the proper settings of the adjusting rings are a co= promise that best meets the objectives of the desired performance, and they are arrived at through a trial and error process by a person experienced in the design and operation of safe ty valves.
5.1.3 Auxiliarv Features l
l The auxiliary features that can be used with spring-loaded safety valves include 1 end l'.f ting levers, anti-si=cer devices, and remote (pneumatic) oper stors.
(a) Hand Lifting lavers. Spring-loaded safety valves often have ihand lif tiy levers which permit manual actuation of the valves at system pressures above about 757. of the valve set pressure. This manual over-ride feature is a carry-over from boiler safety valves. Hand lifting levers are required by some local jurisdictions but are not required by Section III of the ASME Boiler and Pressure Vessel Code for nclear service.
804 201
45 Another auxiliary feature of safety (b) Anti-Siemer Devices.
Ves-valves now recognized by Section III of the ASME Boiler and PressureAs sel Code that is available as an option is the anti-sicx:ter device.
discussed in Subsection 51.2 ' manipulation of the valve adjusting rings However, simmer or leakage may be un-will aid in controlling sicmar.
avoidable in systems which frequently operate at pressures near the set Manufacturers of spring-loaded safety pressure of the safety valve.
Such a device con-valves have therefore developed anti-simmer devices.
Pres-sists of a pneumatic chamber mounted on top of the safety valve.
surization of the chamber loads a plunger which rests on the valve spin-When die and results in a supplemental seating force on the valve disk.
the system pressure reaches the set pressure of the valve, a pressure l
sensing switch deactivates the anti-stomer device, allowing the valve to i
" pop" open by self-actuaLAon in the normal manner.
Spring-loaded safety valves are (c) Remote (Pneumatic) Operators.
Thesa operators available with remotely actuated pneumatic operators.
are mounted on the valve body, and when pressurized by remote actuation i
of a solenoid valve, they operate the valve by moving a hand lif t ng Remote operators are lever or similar lever arm to lif t the valve disk.
of fered for those applications in which the capability to actuate safety valves by a remote signal is required.
Pilot-Operated Pressure Relief Valves 52 Pilot-operated pressure relief valves were developed specifically I
for use in nuclear energy systems to obtain improved leak tightness over Unlike spring-loaded safety that provided by spring-loaded safety valves.
valves, pilot-operated pressure relief valves are available in a variety Nevertheless, all pilot-operated of arrangements and configurations.
pressure relief valves presently available can be characterized as hydrau-lic pilot-operated reverse-seated globe valves which open in direct response to the system pressure by means of a self-energized pilot valve.
This basic design minimizes leakage of the main valve at pressures up to 804 202
46 Two types of pilot-operated pressure relief valves the set pressure.
that have extensive operating experience in nuclear energy systems are
~
the three-stage type and the two-stage type..
5.2.1
'Ihree-Stage Pilot-Operated Pressure Ralief Valve A three-stage pilot-operated pressure relief valve which u.es a pilot valve bellows as a pressure sensing device is illustrated in Fig.
This particular valve has a rated capacity in excess of 800,000 5.5.
lb/hr at pressures above 1100 psi, and it has been used in boiling-water reactor systems.
A schematic diagram of the pilot valve prior to pressurization is '
illustrated in Fig, 5.6.
The pre-load clearance between the pre-load ETR4-13
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Pre-Load Region of Pilot Valve.
As the valve is spacer of the pilot valve and the valve disk is closed.
pressurized, the areas in the three-stage pilot-operated pressure relief These areas valve that contain high-pressure fluid become pressurized.
are shaded in the operational diagram of the three-stage valve -in the closed position illustrated in Fig. 5 7.
804 203
47
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sure Relie f Vahe in Closed Position.
' ich reference to Fig. 5 7, presaurization of the pilot valve causes (J). When the the bellows (I) to expand and reduce the abut =ent r; h bellows pressure in the pilot valve reaches the valve set pressure, t e has expanded until the abut =ent gap has been closed and the pilot disk (C) has been pulled open.
As illustrated in Fig. 5 8, opening of the pilot valve disk (C) pres-l i ton (Q). Le resulting
_suri::es the region above the second-stage va ve p s 1
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pressure differential across the second-stage valve piston causes the second-stage valve disk (S) to open and vent the region above the main Since the diaceter cf the main valve piston exceeds valve piston (M).
the diameter of the seat for the main valve disk (0), the pressure dif-ferential across the main valve piston causes the main valve to open and relieve fluid from the system.
In addition to the self-actuated operating mode just described, the valve illustrated in Fig. 5 5 (page 47) is also remotely operable by actu-Remo te ation of the air operator denoted by the letter "E" in Fig. 5.7 PO4 206
.ee
~
50 actuation of a solenoid valve ptessurizes the air operator, which pushes the second st;ge valve open to vent the region above the main piston and causes the valve to open as during self-actuation except that the pilot valve remains closed during this Inode of operation. By remote actuation of the air operator, the valve may be caused to open at any pressure between the set pressure and a few psi; that is, that pressure necessary to produce a net load on the pisten in excess of the relatively small pre-load on the main valve spring (F). Failure of the air operator cannot prevent the self actuation mode of valve operation.
l Unlike a spring-loaded safety valve, the valve design illustrated in Fig. 5.5 (page 47) does not provide for any adjustability to control blow-The a:cumulation is essentially zero down, accu =ulation, or si::x::ar.
i because of the insediate full-open re sponse of the main valve to opening of the pilot valve. The reverse seating of the pilot, second-stage, and main valves resulta in virtually no seat leakage or siz:x:ering because of The relationship between system pressure and the high seating forces.
seating force for reverse-seated main and pilot valves and for spring-From the curves shown in Fig.
loaded valves is illustrated in Fig. 5.9 5 9, it is clear that the seat leakaga characteristics exhibited by a r
raverse-seated pressure relief valve vill be considerably better than those exhibited by the spring-loaded safety valve.
The blevdown of the valve illustrated in Fig. 5 5 (page 47) is equal f
to the change in system pressure from the closing of the abutment gap to I
t P in Fig. 5.9).
This pressure l
the opening of the pilot valve (Pa 3
i difference (the blevdown) :::ultiplied by the effective pressure area of f
the bellows is equal to the load necessary to unseat the pilot valve disk (the ef fective seating area of the pilot valve disk multiplied by l
Therefore, the the pressure drop across the disk of the pilot valve).
'blevdown of the valve illustrated in Fig. 5.5 is dependent on the set pressure, the back pressure, and the ratio of the disk seating area of Stated the pilot valve to the effective pressure area of the bellows.
mathematically, the pilot disk seating force = (?g - F )A B
D' 804 207
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Valve Seating Force and Bellows Expansion as a Function of System Pressure.
where P3 = set pressure, PB = back pressure, and AD = effective seating area of pilot disk.
The load necessary to pull the pilot disk open 1D = 6P A = (P -Pb' gB g
B where APg = increase in system pressure between closing of the abut =ent gap and opening of the pilot valve and A = effective pressure area of the bellows.
3 "he re fore,
blevdown = AP = (P
-P g
3 B
'Itus, it is apparent that the three-stage pilot-operated pressure relief valve shown in Fig. 5.5 (page 47) has no blowdown adjustment capa-bility after fabrication. It is also apparent that this valve is sensitive hk,
52 to back pressure.
However, this back-pressure sensitivity is negligible when co= pared with that of spring-loaded safety valves without back pres-sure balancing devices, and valves such as the valve shown in Fig. 5.5 are installed with closed dischar,ge systems without need for back pres-sure corrections.
The set pressure of the valve illustrated in Fig. 5.5 is adjustable by maans of the set pressure adjustment screw, which pre-loads the bel-lows.
An increase in spring pre-load increases the pressure load neces-sary to expand the bellows enough to lift the pilot disk. Therefore, an increase in spring pre-load increases the set pressure, and a decrease in spring pre-load decreases the set pressure.
- 5. 2. 2 Two-Stage Pilot-Ocerated Pressure Relief Valve A two-stage self-energized pilot-operated pressure relief valve that use_s a pilb: valve bellows as a pressure sensing device is shown in Fig.
5.10.
Valves such as this are typically used when capacity requirenants do not greatly exceed 100,000 lb/hr.
(Three-stage valves are used for i
capacities above 100,000 lb/hr to maintain short response times.)
Operation of the pilot valvss in both the two-stage valve shown in Fig. 5.10 and the three-stage valve shown in Fig. 5.5 is identical.
The bellows (I) of the two-stage valve shown in Fig. 5.10 is expanded by the increasing system pressure until the abutnent gap (J) is closed and the pilot seat (L) ir, opened. Opening of the pilot valve pressurizes the region above the second-stage valve piston (M), and the resulting pressure dif-ferential across the piston causes the valve to open. Unlike the three-stage valve, in which the second stage valve vents the main piston cham-
.ber, the second stage of the two-stage valve is the main valve and its
. actuation relieves stees from the system.
In addition to the self-actuated operating mode just described, the two-stage valve cau also be remotely operated by an air operator.
The
. air operator, not shcwn in Fig. 5.10, is =ounted on the pilot valve bon-(A) and is also attached to the free end of the pilot valve bellows.
net Actuation of a solenoid valve pressurizes the air operator and produces 804. ?.09
53
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Two-Stage Pilot-Operated Pressure Relief Valve.
an axial load on the bellows of the pilot valve, thereby extending tLa bellows and causing valve actuation in the sa=a manner as in self-actuation.
Like the three-staga valve design illustrated in Fig. 5.5 (page 47),
the two-stage valve design illustrated in Fig. 5.10 does not provide for adjustability of si=er, accu =ulation, or blowdown.
As for the three-stage valve, the accu =ulation and si:=er of the two-stage valve are essen-tially =ero, while che approximate blowdown is designed into the valve onm rn 0
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54 by selection of the proper ratio of pilot disk seating area to effective pressure area of the bellows.
The three-stage valve design, shown in Fig. 5.5 (page 47) employs a her=etically sealed bellows assembly which necessitates set pressure adjust = ant by pre-loading the bellows with the set pressure adjustment spring. However, the two-stage valve design shown in Fig. 5.10 incorpo-rates a pilot valve stes (B) which is threaded into the free end of the bellows through packing. By screwing this stem in or out, the abutment gap is increased or decreased, thereby increasing or decreasing the set pressure.
5.3 Power-Actuated Pressure Relief Valves l
All power-actuated pressure relief valves in use in nuclear energy syste=s are characterized by a pilot-actuated reverse-seated main (globe) valve which is hydraulically operated by a pilot valve in response to a signal from an external control circuit. Although power-actuated pres-
-sure relief valves are used extensively in nuclear energy systems, only a limited number of ' valve designs have been used. The operation of these
= valves is similar to that of the pilot-operated pressure relief valves 2except for the me thod of pilot valve actuation. The pilot valve of a power-actuated pressure relief valve is opened by a solenoid or air oper-l 4 tor powered by an external energy source (electrical or pneumatic) in
- response to a remotely generated control signal. A typical solenoid-operated power-actuated pressure relief valve is shown in Fig. 5.11.
With reference to Fig. 5.11, steam at system pressure is contained in the valve inlet upstream of the main valve disk (C), in the cavity tehind the main valve disk (B), its discharge line (D), and in the cavity upstream of the pilot valve disk (E). The steam has access to these lat-car regions through the blend hole in the main valve pressure chamber (A).
Epon receipt of an electrical signal _fren either a pressure sensing switch er the reactor plant control system, the valve solenoid is energized, thereby stroking the plunger (G) and opening the pilot valve by =eans of the pilot valve actuation lever (F). Opening of the pilot valve (E) vents 804 2II v_. = _,_,
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56 the main valve pressure cha=ber (B). The resulting pressure differential across the main valve disk (c) causes the valve to open and discharge at full rated flow. Operationally, the power-actuated pressure relief valve and the three-stage pilot-ope, rated pressure relief valve differ ent.* in their means of actuation (solenoid or air operator and self-energized pilot valve, respectively).
k l
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i 57 6.
GUIDELINES IN TRE SELECTION OF SPECIFICATION REQUIREMENTS FOR PRESSURE RELIEF VALVES The discussion presente'd in this section is intended to provide background information and guidelinas relativo to the key design, perform-ance, material, and quality assurance requirements for pressure relief valves. This information is presented to assist in the selection of those requirements which must be specified for the procurement of spring-loaded safety valves, pilot-operated pressure relief valves, and power-actuated pressure relief valves.
6.1 Design Recuirecents Tha design requirements to be specified must account for the partic-ular type of valve desired, its port sizes and arrangement, orientation, and design conditions; the chemical conditions of the reactor coolant, the auxiliary devices to be used on the valve and the requirements for the auxiliary power systems, the seismic requirements for the pressure relief valves in the system, and the external loadings which may be imposed on the valve.
6.1.1 valve Type Considerations in the selection of requirements for spring-loaded safety valves, pilot-operated pressure relief valves, and power-actuated pressure relief valves are discussed in the following paragraphs.
(a) Spring-Loaded Safe ty valves. ine type of pressure relief valve most commonly used in commercial central station nuclear p. ants is the spring-loaded safety valve. This valve type, which is similar to the conventional safety valve used in fossil-fueled boiler applications, has the most safety valve service experience and its use requires minimal' departure from conventional safety valve technology. However, problems in obtaining good valve leak tightness, especially af ter several opening
.804 214
a 58 and re-closing cycles, and the potential for valve si r.ar at pressures near the set pressure of the valve have led the designers of reactor plants to supplement spring-loaded safety valves with water seals and I
power-ac.uated pressure relief valves set at pressures below the set pres-sure of the safety valves. The addititual costs and complexities associ-ated with these supplementary features should be considered in the selec-tion of spring-loaded safety valves for nuclear service.
Of the two types of spring-loaded safety valves used in nuclear plants (discussed in Subsection 5.1), the exposed spring valve is normally specified for applications in which leakage of reactor coolant past the valve spindle during discharge is inconsequential. This is the case in boiling-water reactor applications in which the safety valves discharge directly into the contain=ent volr.:e. The enclosed-spring (sealed-bonnet)
. safety valve should be specified for applications in which the valve dis-reharge is piped to a closed discharge system. The enclosed-spring safe ty valve eliminates leakage past the valve spindle during discharge,.
4-also incorporates back-pressure balancing devices which minimize the effect of back pressure (resulting from the closed discharge system) on the set oressure and operation of the valve.
(b e Pilot-Operated Pressure Relief Valves. The pilot-operated pres-l
.sure relief valve was developed for nuclear service in the late 1950's, l
.and its use has been limited to, nuclear power applications. As a result,
.the service experience available for this type of valve is considerably less than that for conventional spring-loaded safety valves. Care should oe taken in the selection of pilot-operated pressure relief valves to assure that the manufacturer has valve service and/or test experience for conditions comparable to those required for the intended application. If
. sui. cable for the application, pilot-operated pressure relief valves offer the advantages of a good history of leak tightness even af ter numerous operating cycles and they can serve the functions of power-actuated relief valves as well as Code-required safety valves. The operating features typical of power-actuated valves, such as fast response and abi.11ty to operate the valve remotely through the use of an external power source, are additional advantages of the pilot-operated pressure relief valve.
804 26
4 59 A disadvantage in the use of pilot-operated pressure relief valves that cuse be considered in the design of the pressure relief system is I
that in accordance with the current edition of Section III of the ASME Boiler and Pressure kassel. Code, credit may be taken for only 757. of the installed capacity of these valves. Ihus, although these valves eliminate the need for power-actuated relief valves, approximately 337. excess capac-ity must be provided in pilot-operated relief valves.
(However, where the design of the pilot-operated pressure relief valves can be codified to qualify them as valves equivalent to spring-loaded safety valves in accordance with the rules of Article NB-7000 of Section III, credit may be taken for the full 1007. of the installed valve capacity.) The eco-nomic advantages or disadvantages of these features will depend upon the specific application involved and the cost of the various types of valvas at the time they are procured.
(c) Power-Actuated Pressure Relief Valves. Power-actuated pressure relief valves have been used in central station nuclear plants to supple-ment the safety valves required by Section III of the ASIE Boiler and Pressure Vessel Code and for functions not related to overpressure pro-tection. Until the summer of 1969, reactor designers were not permitted
- by Section III to take credit for the capacity of power-actuated valves in meeting the Code requirements for overpressure protection. However, it is permissible to take credit for up to 507. of the installed capaci:.y of power-actuated pressure relief valves in meeting the Section III requirements for overpressure protection for plants designed af ter mid-1969 if the valve controls and power supply have redundancy and reliabil-ity equivalent to that provided in other critical reactor safety systems.
This provision of the current edition of Section III permits a reduction in the total required capacity of convent.ional spring-loaded safety valves by 507. of the installed capacity of power-actuated relief valves.
Accordingly, in new overpressure protection systems in chich power-actuated pressure relief valves are used to supplement conventional spring-loaded safety valves, consideration should be given the economic advantage of taking credit for the capacity of the power-actuated relief valves as compared with the additional costs involved in provision of the 1
804 N
i 3
i 60 required reliability and redundancy in the valve control systeu and power supply. As in the case of pilot-operated pressure relief valves, the post economically attractive system will depend on the specific application and the relative costs of the various types of valves at the time procured.
If it is technically and econo'mically advantageous to use power-actuated relief valves as a part of the Code-required pressure relieving capacity, the capacity of these valves unst be certified in the ascie man-ner as is that of conventional spring-loaded safety valves. If the power-actuated ralves are to be used as Coda-required safety valves, cap'acity certification should be specified in the ordering data.
6.1.2 Port Arrangement and Size All spring-loaded safety valves have an angle body configuration; that is, the inlet axis forms a right angle with the outlet axis.
Pilot-operated and power-actuated pressure relief valves are available in a number of configurations, depending upon the specific valve type and man-u f acture r.
These configurations include the angle body pattern, the globe valve body pattern with port axes in line, and the globe valve body pat-tern with port ar.as parallel but offset. All pressure relief valves used in con =ercial nuclear reactor plants are provided with bolted connections
-to per=it periodic removal for testing, examination, refurbishing, and re adjus ting.
Determination of the proper inlet port size for the intended applica-tion is usually made by the reactor designer. As a mini =um, the following f actors must be considered in the selection of the size and schedule of the valv; inlet port and piping.
1.
The size of the inlet pipe must be such that for the rated capac-12y, the fluid velocity in the piping is acceptable (typically, a maximum of 40 to 50 ft/sec).
2.
The size of the inlet pipe c:ust be such that when the total length of inlet piping from the protected vessel to the safety valve is considered, the maxi =um pressure drop at full rated flow capacity is less than the expected minimum blowdown of the valve.
' Ibis is necessary to prevent instability or chatter of the valve.
OA 217
61 3.
The schedule of the inlet piping and fisages must be such that the inlet piping sad flange connection are structurally adequate for the maximum loadings expected to.. result from inlet pressure, piping forces and moments, valve opening, and flow reaction forces during valve s
discharge.
The size of the valve outlet port is usually selected by the valve manufacturer on the basis of availability.of valve designs and proper operation of the valve. However, in cases where the valve outlet is piped to e closed discharge systcm, the reactor designer should make sure that the size of the outlet is large enough to kriep the downstream pres-sure drop during valve discharge within the back-pressure limits specified for the valve.
6.1.3 Orientatien According to the manufacturers of safety valves, it is important
~
that spring-loaded safety valves be installed with,the axis of the main spring, disk, spindle, and inlet nozzle in a vertical position. This orientation assures that loadings on critical operating parts of the valve.are as symmetrical as possible. The spring-loaded safety valves used in commercial nuclear plants are therefore installed in the vertical, I
upright position. Although it is not known that these valves would -oper-i ate unsatisfactorily in a non-vertical orientation, the presently avail-able experience dictates that all spring-loaded safety valves be installed wsth the axis of the main spindle and inlet piping in the vertical position.
Valve manufacturers have indicated that, in general, the installed orientation of pilot-operated and power-actuated pressure relief valves is not critical, and these types of valves have been used in various 5
i orientations in nuclear plants. In the absence of any specific reconnen-dations from the valve manufacturer concerning preferred valve orienta-tion, it is considered advisable to orient pilot-operated and power-actuated relief valves so that the pilot or other actuating device and I
the associated flow passages are self-draining to the =axi=um extent possible.
1 804 M8 L
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62 6.1.4 Design Conditions The design conditions specified for pressure relief valves are typically identical to the design ponditions specified for the reactor vessel in boiling-water reactor plants and for the pressurizar in pressurized-water r,eactor plants. Typical values of design and operating temperatures and pressures for pressure relief valves in boiling-water and pressurized-water reacnr plants are as fo'llows.
.p Boiling-Water.
. Pressurized-Water
..c..,..
Reactor Plants Reactor Plants
. m t.Temperatuye
,..: r.d e a.' -
~
Design 575'F-Y h::.i.m:670 +to 700*FM'c -7, e..-
~
Operating 540 to 550*F 650*F Pressure Design 1250 psia 2500 psia Operating 1000 to 1050 psia 2050 to 2250 psia In addition to the basic design temperature and pressure', the oper-ating f.ransient conditions expected for the reactor coolant system must be specified since these conditions may affect the performance of the 1
pressure relief valves and these conditions are required for structural analyses of the valves. The operating r:ransients specified for typical pressurized-water and boiling-water reactor coolant systems and typical design numbers of cycles for each are given in Table 6.1.
For the valve manufacturer to design a pressure relief valve that will be exposed to reactor coolant system transients such as those given in Table 6.1, the purchaser mst specify the system pressure and te=per-ature response to the transients. However, the system response to tran-sients is generally contained only in proprietary specifications for reactor coolant systems since this response mst be determined by cooplex system analyses. When procuring a pressure relief valve, the valve pur-chaser r.hould, where possible, attach the appropriate pressure and tem-perature response curves contained in the specification for the reactor coolant system.
In addition to the operating transients of the reactor coolant sys-tem, the number of operating cycles anticipated for the pressure relief valve must be specified. The design nu=ber of valve operating cycles is hk e
63 Table 6.1.
Design Data for Typical Itaactor coolant System rranstants Number of Deeurrences for Combustion Babcocis General I
Westinghouse Engineering
& Wilcox glectric Tranatent Coedition 3600-Hid(t) PVR 2400-Hid(t) PVR 2500-NW(t) PVR BVR Beacup and cooldova 70'F-580*F.70*F ae 200 500 480 120 100*F/hr
$40*F. 580 *F-540 *F 1MO s
Ramp loading and usloedios at 5% full power /mia 29,000 13,000 (0 to 1001 (15 eo 100%
full power) full power) at 10% full power /mia 15,000 12,000 (10 to 1001 (15 to 100%
full power) full power) step loading and unloading 2000 2000 2000 at 10% full power Large step unloading 200 40 160 40 (50% full power)
(loss of 100%
(100 to 5%
(turbine trip) turbine load) full poo r)
Raactor trip 400 400 400 200 Eydrostatic test 5
10 10 3
(3110 pet)
(3110 pet)
(3123 psi)
(1560 psi) laak test at 2500 psi 200 320 180 i
(1250 psi) l loss of coolant flow at 400 40 100% full power loss of secondary systen 5
pre s sure 10e Steady-acate fluctuations
=
+6*F and +100 poi in 1 min Daily reduction to 751 powr.
10,000 Weekly reduction to 5C% power 2,000 Control rod worth test 50,000 loss of feedwater heaters 80 loss of feedwater flow 80 Improper start of shutdown 10 recirculation loop quite low for the pressure relief systems currently in use.
For spring-loaded safety valves, which are normally prevented from operating by power-actuated pressure relief valves, the expected number of operating cycles over the life of the plant is conservatively less than 50.
The usage of pilot-operated and power-actuated pressure relief valves will depend to a great extent on the set pressures of the valves; the nature of other pressure control systems, such as the pressurirer spray; and the operating history of the plant. In current pressurized-w. ter and boiling-water reactor plants, it is not expected that the pilot-operated or
. L : n.: :
64 power-actuated pressure relief valves will be operated c: ore than 1000 cycles over the 40-year plant life. Thus, in the absence of specific infor=ation on the particular application, several hundred cycles for spring-loaded safety valves and se'veral thousand cycles for pilot-e operated and power-actuated pressure relief valves would be concidered sufficiently conservative design numbers for valve operating cycles.
6.1.5 Reactor Coolant Chemiserv
.... c.. n.. :s
- u. -.
Of considerable significance in the design and selection of materials for pressure relief valves are the chemical conditions of the reactor coolant. It is most important that definitive reactor coolant chemical conditions representative of the coolant chemistry expected in the plant be specified in the procurement of presaura relief valves. 'the most sig-nificant difference between the water chemistry conditions of current pressurized _ water reactor plants and those of boiling-water reactor
~
plants is that boric acid (a neutron poison) is used in the reactor cool-ant of most cocnercial pressurized-water reactor plants as one means of controlling the reactivity of the reactor system. Typical chemical con-dicions of the reactor coolant in current pr.ssurized-water and boiling-water reactor plants are, given in Table 6.2.
6.1.6 Auxiliary Devices As discussed in Subsection 5.1.3, manufacturers offer optional auxil-iary devices for use on spring-loaded safety valves, and such devices are also available for pilot-operated pressure relief valves. Factors to be considered in the selection of remote operators and anti-simmer devices
'or pressure relief valves are discussed in the following paragraphs.
i (a) Remote Operators. Remotely actuated electro-pneumatic operators can be supplied on spring-loaded safety valves or pilot-operated pressure relief valves when required. These operators permit actuation of a valve at pressures below the valve set pressure, and they do not prevent the self-actuation mode of valve operation. Therefore, the addition of hk
.em
e 65 Table 6.2.
Typical Reactor Coolant Chemical Conditions in Current Pressurized-water and Boiling-Water Reactor Plants Chemical condition PWR Plants BWR P!ents Conductivity,
< 1.0 tt
< 1 to 10 micromhos/cm pH at 25'C 4 to 10.5 5 5 to 8.5 Max 0,,
ppm 0.25 0.43 Max halogens, ppm
< 0.25
< 0.1 to 1.0 H, em /kg of H,0 15 to 35 a
Max total suspended 2.0 0.5
- solids,. ppm --
pil control agent 0.2 to 25 (LiOH), ppm Boric acid, ppe 3000 to 15,000
<1 Silica, ppm
<1 remote operators to spring-loaded safety valves or pilot-operated pressure, a
relief valves should be considered for reactor system applications in which the capability to remotely actuate valves to discharge reactor cool-ant from the system over a wide range of operating prassures is desirable.
However, spring-loaded safety valves should be equipped with remote oper-ators only af ter careful consideration of the possible deterioration of seating surf aces caused by frequent operation of the valve. Since fre-quent operation of spring-loaded safety valves has traditionally been minimized to avoid valve leakage problems, the use of remote operators a these valves should be undertaken only in those applications where the expected frequency of remote operation and the specific. operating condi-tions during remote operstion are within the range of conditions over which the valve manuf acturer can demonstrate satisfactory valve performance.
When requiring re=ote operatort, for pressure relief valves, care
=us t be taken to assure thst the ambient condition extremes, valve d;.ty cycles, and the full range of system pressure over which the remote oper-ator is required to function are specified. Specification of ambient con-dition extrames is particularly important in appliestions where the remote hk,
66 ditions in operator is required to function under possible accident con h as 75 psi at which the reactor containment may be pressurized to as hig These conditions reduce the effective pressure tec:peratures up to 300*F.
differential availabla for operation of pneumatic operators and may gov-ern the type of solenoid used for electric operators.
Specification of anticip'ated duty cycle requirements is particularly important for solenoid-actuated remote operators bnsuae the acceptability i tion of the solenoid is dependant on the duty cycle (suration of energ za Specifica-and duration of de-energization) and the ambient tec:perature.
range of operating pressuras over, which the _ pressure tion of the full relief valve cust operate in respense to the remote actuator is necessary In addition, pilot-operated pressure for sizing of the remote operator.
relief valves normally depend upon the pressure in the system for opera-i only tien of the main valve and are therefore capable of remote actuat on l less at 2ystem pressures above a certain pressure, which is typical y D.
dian 100 psi.
~
Most manufacturers of spring-loaded safety (b) Anti-Sicner Devices _.
These electro-valves now offer anti-simmer devices as optional equipment.
pneumatic devices are attached to the top of the spring retention struc-force to ture of the valve, and they provide a supplemental disk seating Although the use of such devices is minimi=e seat leakage and simnering.
d now permitted by Section III of the ASME Boiler and Pressure Vessel Co e i
t under certain conditicas, their effectiveness and reliability have not ye If anti-siemer devices are consid-been demonstrated in nuclear service.
ered necessary by a particular valve manufacturer or by the reactor designer, the valve manufacturer should be required to provide test data the.
and/or results of satisfactory service experience that demonstrate In addition, the ambient conditions to which suitability of the device.
these devices will be exposed cust be accurately specified.
d Reautrements for Auxiliary Power Svstees_
6.1.7 1
elief valves and pressure relief valves Power-sceusted pressure iliary power vich remote operators or anti-siemer devices require aux
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Therefore, the characteristics of the plant sources for their operation.
i of electrical and pneumatic systems that are available for operat on d
remotely actuated valves ::ust be specified when such valves are procure.
The characteristics of the power systems are necessarily dependent upon f
tion relative to the design of a particular nuclear plant, but the in oma i
agraphs is electrical and pneumatic systems discussed in the follow ng par l
generally applicable to all current centesi atation nuclear p ants.
Current central station nuclear plaats (a) tiectrical Systems.
or 120/208-volt altar-typically will have available 120-volt, 208-volt, d a 440-125/250-volt direct current scurce; an nating current soutcas; a However, the volt, 480-volt, or 600-volt alternating current source. is not s.1 ways alternating current source in the 400 to 600 volt range lief valves available within the containmont volume (where the pressure re Virtually all remotely actuated pressure relief valves are located).
h installed in nuclear energy systems are controlled and posered by t e f
Thus, the direct 125/250-volt direct-current electrical power sevree.
f remote current source should be used for the control and operecion o operator and anti-sics::er devices when available.
Pneumatic valve operators are designed for (b) Pneu:natic Systems _.
ble in standard 80 _o 150 psi instrument air systems normally availa In the absence of other requirements, the pressurs nuclear power plants.
i res-relief valve manufacturer should be given the normal and maximum a r p design sures available in the 80 to 150 psi instrument air system as a for pneumatic valve operators.
requirement 6.1.8 S eismic _.Recuireewn ts_
The loadings that could result forn seismic disturbances cust be This is effected by using the spec-considered by the valve manuf acturer.
in ified arcimum anticipated inertial lod factors imposed on ths valves These inertial load factors (G-fr:cors) hori: int.a1 and vertical loadings.
Neing analyzed to are ewiciplied by the weight of the valve or valve e -
.gn purposes.
obtain an equivalent static load that may be ased o The inertial load f actors which cust be specified represent the respense 214
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of the valve, the response of the piping on which the valve is mounted, and the response of tha valve and piping supports to the ground motion
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This ground motion and the characteristics during a seismic disturbance.
of the valve piping and supports (tpa degrees of freedom, spring con-to plant. There-stants, and natural frequencies) will differ from plan fore, the seismic. load factors for a given pressure relief valve applica-tion es2st be detarained for each plant.
The determination of suitable seismia load factors for a.given appli-The plant cation is made by using seismic response spectra for the plant.
response spectra for the design earthquake is usually providad in the form
..t..
af a plot showing the variation in the maxirmin response of a single-degree-of-freedom oscillator as a function of the natural frequency of vibrat.cn when subjected to the base acceleration represented by the This accelerogram is a plot of accelerogram of the design earthquake.
ground acceleration versus time for an actual earthquake that is consid-ered representative of what might be expected in the geographical area of the particuale plant. A typical nuclear power plant response spectra plot is illustrated in Fig. 6.1.
From the spectra in this plot, it is possible to determine the maximum response of a single-degree-of-freedom system and, by supefposition, the maxi =um response of a multi-degree-of-freedom system.
As an exemple of the use of the response spectra, assume that a valve l
under consideration is located on piping which has three degrees of free-dem in the horizontal and vertical directions. Further, assu=e that the natural frequencies of vibration of the piping are 0.08, 0.2, and 1.0 cps The dac: ping factor typically used 'or welded struc-in each direction.
tures such as piping is 0.57..
From Fig. 6.1, the corresponding maxi =um hori=ontal accelerations in one direction are as follms.
Natural f.
Frequency G-Factor (cps)
(7.G) 0.08 0.77 0.20 4.0 1.00 24.0 SeeOWeeV MeM-M
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I 70 A conservative total maxi =um horizontal acceleration in one direction can be obtained by direct addition, yielding a value of approximately 297.G or 0.29G.
This is the accelegation in one horizontal directice. The total maximus acceleration in the other horizontal direction is also 0.29G. Dis, by vector a-tion, results in a total -m
'-- horizontal acceleration of [(0.29G)2 + (0.29G)2p/2 = 0.41G. Where vertical response spectra are not provided, the vertical response for each vibrational mode is typically taken to be two-thirds of the value of the corresponding horizontal acceleration. Thus, the inertial load factors for this exam-ple would be 0.41G in the horizontal direction and 0.27G in the vertical direction.
6.1.9 External Loadings The external loads that may be imposed on the inlet and outlet noz-zies of a valve must be considered by the valve manufacturer. These loads consist of forces and moments resulting from thermal expansion of i
the attached piping, flow reaction forcan resulting frem valve discharge, and seismic loads. 'It is most important i: hat the valve manufacturer be required to design and analyze his valves for these external loads not only for structural reasons but also because these loads may affect the operational characteristics and seat tightness of the v.'1ves.
Hence, it
.is important that the valve purchaser specify to the valve manufacturer both the piping expansion loads and the seismic loads. If the specified external loads are considered excessive by the valve manufacturer, it may be necessary to re-evaluate the design of the inlet and discharge piping systems and supports for the pressure relief valves to minimize l
the external loads resulting from piping expansion and contraction.
(a) Piping Expansion. Pressure relief valves whose inlets and I
outlets are both rigidly connected to the plant piping systems are sub-jected to piping reactions resulting from the thermal expansion or con-traction of the piping during plant and valve operation. Typical of this condition are the pressure relief valves in a pressurized-water reactor plant. A comon complaint of valve manufacturers is that system designers
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design a system in the cold condition with low external loads on the
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valves and fail to design for the hot pipe condition in which themal expansion causes significant toading on the valve. It is therefore impor-tant that when specifying pi;fing reactions, the valve purchaser consider the vorst esse of piping expansion or contraction. tiowever, for operabil-ity and leak tightness, pressure relief valves are strain dependent rather than stres L pendent. Therefore, the applicable maximum piping reactions are subject to negotiation with the valve manufacturer.
(b) Flow Reaction Forces. When discharging, pressure relief valves produce loads on the connecting piping as a result of the momenti.2 and pressure changes in the fluid as it passes through the valve and connect-ing piping. These loads are normally calculated by the valve manufac-turer, using elementary themodynamic and flow relations in conjunction with the inlet conditions and the orifice area or total flow rate. In addition to these steady-state flow forces, pressure relief valves are subjected to transitory reaction forces during valve opening and closing that are of ten considerably in excess of the steady-state reaction forces'.
'ihese transitory forces are caused by the shock of stopping the dish, spindle, and other parts that are set in motion by valve opening or clos-ing. The transitory reaction force is dependent upon the individual valve design and the arrangement of the connecting piping, and it must usually be calculated or determined experimentally by the manufacturer.
(c) Seismic Acceleration. During an earthquake, pressure relief valves are subjected to inertial forces associated with the accelerations they experience as a result of the excitation of the connecting piping.
The manner in which such accelerations are determined is discussed briefly l
in Subsection 6.1.8.
These accelerations or inertial load factors are
't multiplied by the weight of the valve and the results are applied as equivalent static loads at the center of gravity of the valve. The seis-mic loads so applied are reacted as forces and moments on the inlet and outlet r.)z=les of the valve and cust therefore be considered in the anal-ysis of.. external loadings.
72
_6. 2 Performance Requirements The factors to be considered in selecting the performance requirements to be specified include the tot.1 capacity required for the pressure relief valves in a nuclear plant, selection of the valve set f
pressures, the blowdown range of the pressure relief valves, and the back
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pressures affecting the valves.
6.2.1 Caoacity
-- ed: 1--evi+ c - - -, :.e : 3 - :1 +.:: ::. d 5 i:"; -
i The total relief valve capacity required for a nuclear plant is that relieving capacity necessary to prevent the pressure in the protected sys-t tem from axceeding 110% of the system design pressure during any normal or abnormal operating condition. Suppliers of nuclear energy systems emphasize that there is no single or short-cut method of approximating the pressure relief valve capacity required for a given plant. Typical i
pressure relief valve capacity requirements for pressurized-water and i
5
. boiling-water reactor plants are discussed in Section 4 of this document.
I The capacities of individual pressure relief valves are usually J
l determined on the basis of the availability of existing valves and eco-nomics.
Designs for pressurized-water reactor systems generally call for two or three pressu;
- relief valves, depending on the acaber, availabil-ity, and cost of the various valve sizes necessary to develop the total I
reliaving capacity required. The largest pressure relief valves avail-l able for nuclear energy systems (capacities of 600,000 to 800,000 lb/hr)
I are currently used in boiling-water reactor systems.
I 6.2.2 Set Pressure Selection Selection of valve set pressures is governed by consideration of the rules of Section III of the ASME Boiler and Pressure Vessel Code, system transients and valve leakage, and system perturbations resulting from valve operation.
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-a 73 (a) Code Recuirements.Section III of the ASME Boiler and Pressure Vessel Code specifies an upper limit on the set pressures of the required pressure relief valves. This maxi =um. set pressure is the lower of the limits set by the fe ' lowing tslo Section-III requirements.
1.
The set pressures of the required pressure relief valves shall be such as to prevent the
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pressure in the protected system from exceeding 110% of the system design pressure.
2.
At least one of the required pressure relief valves shall be set to open at a pressure no greater than the system design pressure.
(b) System Transients and Valve Leakage. Pressure transients in the protected system set a lower limit on the set pressures of the required pressure relief valves. In order to prevent or limit seat and disk deterioration caused by valve sinner, the set pressures of the valves must be significantly in excess of the peak pressures of normal system transients to prevent unnecessary actuation of the valves and valve simmer. Manufacturers of spring-loaded safety valves recommend that the pressure dead band between the valve set pressure and the peak system pressure during normal transients be at least 10% of the valve set pressure. However, reactor designers usually provide a greater dead band than this. Manufacturers of pilot-operated pressure relief v.alves nor-mally recommend that the pressure differential between the normal peak pressure and the valve set pressure be a minimum of about 50 psi; that is, about 2.5% of the set pressure in pressurized-water reactor systems and 5% of the set pressure in boiling-water reactor systems.
(c) Svstem Perturbations Resulting From Valve Operation. Within the limits set by the previously discussed considerations, valve set pres-sures are normally distributed to avoid simultaneous opening of the installed pressure relief valves. This serves to reduce system perturba-I tions caused by the sudden loss of fluid from the system, and it also
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minimizes the reaction force which may be applied to the system at any one time as a result of valve openings. Typical pressure relief valve e
set pressures for current pressurized-water and boiling-water reactor systems are discussed in Section 4 of this document.
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_,m 74 6.2.3 Blowdown Typical blowdowns for pressure relief valves in acclear energy systems are 4 or 57..
The source of these values and the typically spec-ified accumulation of 37. for pressure relief valves in nuclear energy systems is the Section-III requirement that the valves be rated at a pres-sure not exceeding 1037. of the valve set pressure with a blowdown not ea eeding 57. of the set pressure. The available blowdown range for pres-sure relief valves is limited by system limitations and valve design.
(a) Svatem Limitations.l System considerations place upper and lower limits on the permissible valve blowdown. The blowdown must be low enough to rinimize loss of reactor coolant and the resnicant reduction of system pressure. At the same time, the blowdown must be high enough to keep the pressure drops in the upstream piping during full flow from exceeding the blowdown of the valve.
If the upstream pressure drop exceeds the valve blowdown, the valve vill start to close immediately upon the initiation of flow caused by the pessure drop and it will reopen af ter the flow has been reduced and the inlet pressure has again increased.
Such instability or l' chatter" results in rapid deterioration of the seat-ing surfaces of the valve, and more importantly it significantly reduces the effective es pacity of the valve.
~
(b) Valve Design. The blowdown and accumulation of spring-loaded safety valves are both determined by the arrangement of adjusting rings which control the flow reaction forces acting on the valve disk.
There-fore, tLe blevdown and accu =ulation of this type of valve are partially interdependent.
In general, as the flow forces acting on the valve disk increase, the blowdown increases and the accumulation decreases; and as the flow forces decrease, the blowdown decreases and the accumulation
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increases.
I The blowdown of pilot-operated pressure relief valves is determined i
by the design of the pilot valve and is not adjustable.
It is therefore important that the required blowdown be clearly specified at the time of valve purchase to allow the valve to be designed for the required e
blowdown.
804 2M
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75 The opening and closing of power-actuated pressure relief valves are I
l both remotely controlled by the control system (manual or automatic).
Thus, the blowdown of power-actuated pressure relief valves is determined s
solely by the control system..
6.2.4 Back Pressure The back pressure in a pressure relief valve can affect the perform-ance of the valve in a number of ways, and it should therefore be speci-fied so that the valve can be properly designed and certified for the intended service condi'tions. The back pressures that must be specified are the static and dynamic back pressures.
(a) Static Back Pressure. Static back pressure is the pressure in the outlet of the valve when it is closed. The static back pressure in pressure relief valves which discharge directly to the reactor contain-ment is equal to the ambient pressure in the reactor containment. The static back pressure in pressure relief valves which discharge to a closed discharge system is equal to the pressure in that closed; system.
In determining. the static back pressure of a closed discharge system, the static overpressure in the system and the pressure increase resulting from the operation of other pressure relief valves that discharge into i
the same closed system must be considered. The range of static back pressures must be spect.fied to the valve manufacturer.
(b) Dynamic Back Pressure. The dynanic back pressure is the pres-sure in the outlet of the valve when it is discharging. The maximum dynamic back pressure consists of the static back pressure plus the maxi-l mum pressure drop in the discharge piping system resulting from the maxi-mum rated flow of the valve.
i 6.3 Material Recuirements 1
Acceptable materials of construction for the pressure-retaining parts of valves are prescribed by Section III of the ASME Boiler and,
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76 Pressure Vessel Code. The selection of materials for non-pressure-
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retaining valve parts and for valve parts in contact with the reactor coolant during static (valve closed) operation and during valve discharge is generally based on the recocuandations of valve manufacturers and reactor designers and on service experience in central station nuclear i
plants. Materials that have typically been uaed for pressure relief valve parts are given in Table 6.3.
Of particular significance in the selection of materials for a spec-ific relief valve application is the chemistry of the reactor coolant.
As discussed in Subsection 6.1.5, a main difference between the chemistry of the reactor coolant in current pressurized-water reactor plants and that in boiling-water reactor plants is that boric acid is added to the reactor coolant of pressurized-water reactor plants. Test data and serv-ice experience indicate that corrosion resistant materials, such as stain-less steels, Inconal, and Stallites, c:ust be used for all valve parts that will be in contact vita borated reactor coolant. Carbon and low-
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alloy steels are attacked rapidly by borated coolant, and their use should i
be avoided in these applications. Recent service experienco in central station pressurized-water reactor plants has indicated that even carbon steel valve parts outside -the coolant boundary may be subject to severe corrosion attack by borated water leaking through the valves or gasketed joints.
Accordingly, in applications involving borated reactor coolant, l
cor-rosion resistant materials should be used for all pa-ts of pressure relief valves normally in contact with reactor coolant and for all parts, such as the valve body and downstream outlet, which may contact borated water during valve discharge or in the event of valve leakage. For internal I
working parts or seating surfaces, it.is prudent that the manufacturer select and recorcend materials which he considers satisfactory for the j
intended application.
For borated reactor coolant applications, the valve manufacturer should be requested to provide test data and/or rest i
of satisfactory service experience that demonstrate that the materials are suitable for use in the specified reactor coolant chemistry.
An area of concern in the selection and treat =ent of stainless stet for nuclear service is the susceptibility to stress corrosion cracking c grg 2D i
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77 Table 6.3.
Materials Used in Nuclear Pressure Relic.f Valve Parts Valve Part Material I
Inlet nozzle forging Type 304 or 316 austenitic stainless steel Body and other pressure-Carbon steel castings and, inborated systems, retaining parts in con-type 304 or 316 austenitic stainless steel tact with reactor cool-forging or casting (carbon steel bodies in ant borated systems have experienced extensive corrosion due to leakage and valve operation and are not normally used)
Spindle or stem Austenitic or Martensitic stainless steel Disk Forged type 304 or 316 austenitic stainless stui or Inconal Pressure sensigg bellows Nonferrous corrosion resistant material Back-pressure balancing Austenitic stainless steel bellows Bonnet, yoke, and other Carbon steel bars, castings, and forgings pressure-retain.__ parts not contacting coolant Main springs of spring-Alloy sirring steel loaded safety valves Bolring Alloy steel Seating surfaces Corrosion resistant materials whose combina-tions include austenitic stainless steel on austenitic stainless steel (type 304,316, and 1990DL), Stellite on Stellite for spring-loaded safety valves, Stellite on Stellite and Stellite on Martensitic (400 series) stainless steels for power-actuated relief valves, and Stellite on Stellite for pilot-operated relief valves. Relatively soft stainless steels are used in spring-loaded valves to provide improved sealing, while hard-faced seats are used in pilot-operated valves, which have high seating forces, to provide wear resistance for frequent operation Internal wear surfaces Stellite and Colmonoy hard facing Internal springs, pins, Corrosion resistant materials such as austenitic washers, piston rings, stainless steel, Inconel, cast Stellite, Haynes liners, etc.
25, and type 17-4 pH stainless steel (tempered at 1100*F or above)
Seals Stainless steel self-energized 0-rings and spiral-wound stainless steel & asbestos (Flexicallic) gasiets hk k
m 78 typ.
3CW. and 316 austenitic stainless steels that have been heated in the tewersture range of 800 to 1600*F for any significant period of
- time, testing of these austenitic stainle.ss steels in this temperature rense results in precipitation of carbidas at the grain boundaries of the esterial; a condition referred to as sensitization. This sensitization etsatf tcastly reduces the resistance of these materials to stress cor:o-stos cracking because of contaminants such as chloridas and fluoride /..
This is particularly true for rolled or vraaght stainless steel forms.
1t 3 tt is difficult to assure that no such contaminants will M present entsag the seaufacture, tasting, and operation of pressure rel!caf valves, the see of sensitized austenitic stainless steels should bc :.soided where peestbte.
Possible alternatives include the use of materials wt.ich are not as e.aceptible to sensitization as types 304 and 316 stainless steel and
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e t tmamat t oo of the he at trest:nents which cause sensitization. Alternate g,
metertels that have improved resistance to sensitization include Inconel, i
- C stabitsed stainlass steels such as type 347, and low-carbon stainless 3
etsels (carboo content less than 0.037.) such as types 304L and 316L. The p
heat treataset operationa cournonly employed in valve manufacture that may
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tot 4Larestonal ecxitrol purposes and the preheating of valve parts to the idQ;., essen tiaation temperature range for the purpose of applying weld-deposited hard f actag such as Ste111ta. Where such heat treatments are required, pg % feasthtlity of limiting the maximum temperature to below 800*F should M.,~%e comeL4ered.
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w37 6.4 Quality Assurance Recuirements A
Pattars rem 1Lar to the procurement of pressure relief valves that
.h 14 h canaldered to the selection of quality assurance requirecents 5;.
- **lectice of test witness points and performance tests.
see)eets are briefly discussed herein.
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i 79 6.4.1 Examination and Test Witness Points 4
During the course of manufacture, examination, and testing of pressure relief valves, the va'lve purchaser or his representative chould monitor and audit the performance of the valve manufacturer in meeting all apacified requirements to assure compliance with them. In addition, key examinations and tests on one or mora of the valves ordered should ha witnessed by the valve purchaser or his representative. This is accomplished by specifying the examination and test witness points during valve production at the time of or shortly af ter placement of the order.
Witness points should normally include important tests that are per-formed onIv ence for a given valve (hydrostatic and valve performance tests). 0-her witness points should be selected so that at least one i
nondestructive examination of each type required may be witnessed during l
1 the course of valve manufacture. Examples of witness points which are considered important are 1.
a typical liquid penetrant examination of a pressure-retaining part, hard facing, and/or bellows; 2.
a typical ultrasonic examination of pressure-retainiag forgings or bolting; 3.
review of selected casting and welding radiographs; 4.
hydrostatic tests; i
5.
p eformance tests; 6.
final cleanliness inspection; and 7.
final preparation for shipment.
6.4.2 Performance Tests i
All pressure relief valves should be subjected to production perform-ance tests. In addition, special lead unit tests should be perfor=ed to qualify a new or unproven design or design feature. The need for lead unit qualification tests of a given valve design should be based on dis-cussions with the prospective valve supplier and, where possible, with previous users of the valve design.
In general, lead unit qualification 804. 236 1
ink -
80 testing is considered warranted if (1) the valve design represents a new and previously unused concept or if (2) the design conditions and service requirements differ substantially,from those for which valid test data and/or service experience are available. For example, lead unit tests would be warranted for valve design concepts.tred by c valve supplier for the first time or for cases in which a valve 'nat has operated satisf ac-torily with steam is required to operated with water.
Cases in which special lead unit qualifiestion tests may not be required include reasonable extrapolaticus of a proven valve design to different sizes or pressure ratings, the use of an established functional ec bination of preren corponent parts in a different mechanical arrange-or the use of a valve design for operating conditions whf ch,
- ment, although new, are bracketed by satisfactory service experience at some-what different temperatures and pressures.
i i
804 237
81 l
MAINTENANCE AND TESTING OF PRESSURE RELIEF VALVES 1
7 The functions required for proper maintenance and testing of pressure i t inspection relief valves prior to and af ter installation include rece p i dic testing and testing, pre-operational examination and testing, per oi tenance.
and maintenance af ter installation, and some non-routine ma n Receipt Inspection and Testing 7,'
Generally, an order for pressure relief valves requires production bility and the testing of all pressure relief valves to assure opera hipment, and proper set pressure adjustment, final inspection prior to s i
f and dam-suitable preservation and packaging to prevent contaminat on o As a result of these provisions age to the valves during normal shipment.
d adjust-and the impracticality of performing meaningful valve tests an i
f reactor cool-ments in the field, pre-installation performance test ng o d is not normally ant pressure relief valves is not considered verranted an i
However, receipt inspection of the valves at the plant con-performed.
Receipt inspection should include struction site should be performed.
f ship-visual examination of the valves and packaging for evidence o d anf 1.
ping damage, proper sealing of the inlet and outlet ports, an i
indication of tampering; ing verification of proper valve size and identification by compar i s; 2
shipping papers with data on valve name place and physical dimens on d seal for examination of valve set pressure adjustment lock vire an 3
evidence of any readjustment or tampering; and l
for visual inspection of accessible internal passages of va ves 4.
cleanliness and freedom from foreign materials.
d FollowinF receipt inspeccion, the valve covers and packaging shoul d and be carefully restored, and the valvas should be stored under locke d protected from controlled storage conditions, preferably under cover an The importance of controlled or bonded storage of the the environment.
the construction site cannot be overemphasized pressure relitf valves at
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i since it must be assured that the valves are in no way readjusted, i
damaged, or tampered with af ter completion of the manufacturer's final s
If during receipt inspection, site storage, or instal-perfomance tests.
lation of the valves in the system", it appears that tampering has or I
could have occurred, the valves should be returned to the manufacturer for ratesting and re-verification of the set pressures and operability.
Pre-Operational Examination and Tests _
7.2 Af ter installation of the valves in the system and prior to initial Sub-operation of the system, precautions similar to those describe,d in h i d section 7.1 should be taken to minimite the risk of damage or unaut or ze readjustment of the valves while they are installed in the plant during Prior to system operation, the valves should be 2
the construction period.
These examinations should subjected to pre-operational examinations.
include visual examination for evidence of damage, deterioration or tampering, 1.
and cleanliness,;
leak-tightness test of the valve to flanged piping joints (nomally 2.
perfomed with the valve gagged during tha hydrostatic and tightness 1
tests of the system);
installation check where applicable to verify proper installation of 3
d electrical wiring, pneumatic connections, and pilot sensing lines; an final examination to assure that the gagging device and any tec:porary 4.
i l
testing devices have been removed and all valve adjust =ents are su t-ably locked prior to operation of the system.
If the foregoing guidelines for receipt inspection, storage, and pre-i l
operational examination are met satisfactorily, pre-operational funct ona testing of the valves is not considered necessary.
804 239 g
7 83 7.3 Periodic Testing and Maintenance e
Suppliers of nuclear energy systems and manufacturers of pressure relief valves recomend periodic testing, examination, and readjustment of pressure relief valves. Pressure relief. valves installed in pressur -
ized water reactor systems are usually removed from the system, retur-bished, and ratested annually, while those installed in boiling-water reactor systems are usually refurbished and ratested at two -aar inter-vals. These periodic tests are normally performed in onjunction with the plant refueling outages, which are scheduled a. approximate yearly intervals in Isrge central station nuclear plants.
The functions to be performed during periodic examination, refurbish-ing, and testing are within the capability of the maintenance personnel of most plants if the necessary facilities for set pressure testing of the j
valves are available. However, unless the plant operator is experienced in the maintenance testing of pressure relief valves and their adjustment and unless he is equipped with all of the required special tooling and test facilities, the disassembly, refurbishing, adjustment, and production testing of these valves should be performed by the valve manufacturer.
When this is not possible and the valve refurbishing, readjustment, and production testing operations must be performed at the plant, considera-tion should be given to the possibility of performing the work under t e
supervision of the service representative of the valve manufacturer.
m i
l 7.3.1 Examination Periodic examination involves removal of the valve from the system and a careful visual examination of the accessible surfaces. Mirrors, boroscopes, or other visual aids should be used to the maximum extent possible to permit examination of the internal parts and surfaces of the
'i valve. Care should be taken during this examination not to disturb the valve adjusting rings or other adjustments. If this examination reveals no abnormal conditions and the service history of the valve has been satisfactory, the valve should be bench tested and returned to service.
804 240
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7.3.2 Re furbishing If the periodic examination of the valve discussed in Subsection
+
7.3.1 or the operating history of the valve indicate any need for mainta-nance or abnormal condition, the valve ~ should be.disasse= bled, inspected, refurbished as required, cleaned, and reassembled. Under these condi-tions, refurbishing nomally should includa reconditioning of seats, disks, and other critical wear surfaces and replacement of all normally replaceable gaskets and seals ~.
7.3.3 Testing Periodic bench testing should be performed to verify the operability, set pressure, and seat tightness of the valves. The test methods that can be used include a low-flow test with saturated steam at operating temperature and pressure conditions, low-flow test with nitrogen or air at room temperature, hydraulic test with hot water or water at room tem-2 perature, and tests with special fixtures.
(a) tow-Flow Test With Steam. In the low-flow test made with sat-
.urated steam at the operating temperature and pressure conditions, the valve should be allowed to reach equilibrium conditions and it should then be popped to verify the accuracy and repeatability of the set pres-t sure. If readjustment is necessary, the valve should be ratested af ter each adjustment until the desired set pressures are obtained. Set pres-sures measured during this test should not vary by more than + 17. from the desired set pressure.
(b) Low-Flow Test With Air or Nitroge3 The low-flow test made with nitrogen or air at room temperature is conducted in the sa=e manner as the previeusly described test made with steam. However, the difference between the test temperature of the fluid and the lated temperature of I
the steam on the lif t pressure of the valve cust be accounced for. This is usually done by using a calibration curve, supplied by the valve man-
.ufacturer, that relates measured lif t pressure for saturated steam at rated conditions to the =easured lif t pressure for nitrogen or air at s
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room temperature for the same valve setting. The requirement that this calibration curve be provided by the valve manufacturer should be included l
in the ordering data of the procurement document if such testing is anticipated.
(c) Hydraulie Test. Hydraulic testing is normally applicable only to pilot-operated pressure relief valves, and it is not usually recom-mended for spring-loaded safety valves because their operating character-istics with water are normally significantly different from u.
with steam. When specifically reconcended by the valve manufacturer, set pres-sure verification tests may be perfore:ed by using water at room tempera-ture or at the design temperatrze. In this case, the use of calibrat1.on curves which relate measured set pressures under design conditiona to the measured set pressures under test cour'U.lons is required.
(d) Tests With Special Fixtures. Some manufacturers of pressure relief valves offer special testing fixtures to facilitate periodic sot pressure verification testing. One such device normally used on ' spring '
loaded safety valves consists of a hydraulic cylinder which is attached I
to the top of the main velve spindle. This device applies an upward force on the sytndle until the valve is opened. The hydraulic cylinder and associated controls and instrumentation are calibrated so 'that the pressure of the hydraulic cylinder at which the spring pre-load of the valm is overcome can be related directly to the set pressure of the valve. In any test setup that involves the use of special test fixtures l
to simplify bench testing or to permit set pressure testing of the valve when it is in the system line, care should be taken to assure that the special test fixtures and the test procedures have been qualified and that the necessary calibration data were obtained under conditions which duplicate those expected in the plant.
I 7.4 Non-Routine Maintenance Non-routine valve maintenance and repair necessitated by valve leak-age or malfunction should preferably be performed by the valve manufacturer.
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c 86 However, when it is necessary or desirable that such verk be performed by the plant operator, the maintenance procedures specified in the valve technical manual should be strictly < followed. Valve mainter.ance, adjust-cent, and testing not covered in the techuical manual that must be per-formed by the plant operator should not be performed without a written procedure agreed to by the valve manufacturer. In addition, these oper-ations should preferably ba performed under the supervision of a repre-sentative of the valve manufacturer.
The maintenance operations nomally included in technical manuals for pressure relief valves are valve disassembly and reassecbly; seat and disk reconditioning (lapping); adjustments for set pressure, blevdown, accu =ulation, and sinner; set pressure testing; and installation requirements.
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GLOSSARY b
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89 GLOSSARY accumulation: the increase in system pressure above the lif t pressure required to obtain the specified design capacity, expressed in percent of the set pressure.
adiusting rings: rings threaded to the nozzle and disk guida of spring-loaded safety valves used to alter and control valve accu =ulation, blowdown, and sinx::er.
anti-sicmer device: an auxiliary device for spring-loaded safety valves that provides a supplemental force on the valve spindle to increase the valve seating force. The purpose of the device is to mintsi=e seat leakage and valve sL=ser at system pressures near the set pres-sure of the valve. At the valve set pressure, the device is de-energized to permit opening of the valve at the desired pressure.
back pressure:
the pressure in the outlet of the valve (also see static back pressure and dynamic back pressure).
back-oressure balancing devices: devices (bellows or pistons) which min-imi=e the effect of back pressure on the valve lift pressure.
I back-pressure sensitivity: the average effect of back pressure on lif t pressure expressed as the psi change in lif t pressure per 100-psi change in back pressure.
blowdown:
the difference between the lif t pressure and the pressure at which ths valve reseats, expressed in percent of the set pressure.
body:
that portion of the valve that connec6s the inlet and outlet flanges. The valve body directs the outlet flow, contains the valve internals, and contains or supports the pilot valve if there is one.
bonnet: that part of a spring-loaded safety valve that encloses the spring and maintains the spring pre-load. It may be subjected to internal pressure in the event of a failure of the back-pressure balancing bellows.
The bonnet is also that ~ part of a pilot-operated pressure relief valve that encloses the pilot valve to prevent leakage to the ambient in the event of failure of the pressure boundary of the pilot valv:.
caoacity:
the certified fluid discharge rate frem the main valve, expressed in Ib/hr of steam.
certification: a written and signed statement describing an action which has been performed.
critical surfaces: those surfaces which if damaged chemically (stain, corrosion products, etc.) or physically (abrasiora, nicks, etc.)
would render the component unsuitable for the it tended service or degrade the performance of the valve in service.
disk: the movable member of the valve seating surfaces.
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90 disk quide: a hollow cylindri.al sleeve that is concentric to and in contact with the outer surface of the disk (or disk holder, if any) in a spring-loaded safet? valve. This device provides guidance for disk movement during valve operation and also s
.apports one or more of the adjusting rings.
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dynamic back pressure: the pressure acting in the piping at the outlet port of the main valve during discharge of the valve.
gagginR device:
a special device which prevents valve lift and discharge when the inlat pressure exceeds the lif t pressure of the valve.
The gagging device is installed and used only during hydraulic testing of the system.
hand lifting device: a device for direct manual actuation of the valve at pressures in excess of 75% of the valve set pressure. Such devices are common on spring-loaded safety valves.
huddling chamber:
term applied to the annular region between the nozzle, lower (or nozzle) adjusting ring, disk, and disk guide in spring-loaded safety valves. Rapid buildup of pressure in this region at the onset of valve opening causes the valve to " pop" open.
inlet nozzle:
a cylindrical or conical component of a spring-loaded safety valve that fits into the inlet port of the valve and contains the inlet fluid and pressure.
internal parts:
those parts which are required for functional, support, or other reasons but whose failure would not result in leakage of reactor coolant to the surrounding atmosphere. Internal chambers and internal adjusting rings are. internal parts of a valve.
lift pressure:
the actual' pressure at which initial opening of the valve occurs in response (directly or indirectly) to an increase in _ystee pressure.
main spring:
the spring which provides the load necessary to close the main valve disk.
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main soring retentf m s tructure : the yoke or bonnet of a spring-loaded safety valve that retains the spring, spindle, and disk and trans-mits the main spring load to the valve body.
main valve:
the valve through which the rated capacity flows. The main valve of pilot-operated and power-actuated pressure relief valves is actuated by a pilot valve or an electrical or electro-pneumatic operator. Equivalent terms include main unloading valve and main relief valve.
coerational cvele:
a full opening of the valve followed by a full clos-ing, including normal operation of all principal functional parts.
,j ihis meaning is of ten implied by use of the terrs "a cycle" or
" valve cycle".
oilot-ooerated pressure relief valve :
i a pressure relief valve which typically consists of a main valve that is hydraulically actuated by the pressurizing fluid upon opening of a pilot valve. The main valve provides the desired relieving capacity. The pilot valve, 804 246-N e
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91 which may consist of one or mere stages, is sized for proper hydraulic operacion of the main valve, and it normally provides negligible pressure relieving capacity as compared with that of the main valve.
pilot valve: an independent valve who9e operation servss to hydraulically actuate the main valve.
pilot valve bellows: a relatively rigid machined pressure-retaining part of the pilot valve with a relatively low axial spring constant that acts as the pressure sensing element in pilot-operated pressure relief valves.
pilot valve containment: a pressure retaining enclosure required by Section III of the ASME Poiler and Pressure Vessel Code for the pilot valves of pilot-operated pressure relief valves to prevent leakage to the ambient in the event of pilot valve failure. This f'
containment is often referred to as a bonnet.
power-actuated pressure relief valve: a pressure relief valve that typ-ically consists of a main valve and an electrically actuated valve operator which controls both the opening and closing of the main valve.
pressure relief valve: an assembly consisting of pressure retaining and operational parts with the function of relieving fluid to prevent excessive pressure in a protected system.
pressure retaining parts: Those valve parts that fonn the inlet pressure boundary or are directly or indirectly stressed by inlet pressure.
Pressure Jutaining parts include those parts whose failure could result in inadvertent opening of the valve and/or loss of reactor coolant from the protected system. Examples of pressure retaining parts are the valve body, bolting stressed by inlet pressure or the main spring, the disk, inlet nozzle, main spring, spindle, and the spring retention structure of spring-loaded safety valves; the body of the pilot valve and the bellows and enclosure of pilot-operated and power-actuated pressure relief valves; and all other connecting parts and closures stressed by internal pressure. Equivalent terms include pressure containing parts and pressure boundary parts.
reactor coolant:
the fluid which cools the nuclear power reactor and serves to transfer the heat produced to a desired location. Reactor j
coolant in the form of steam is the pressurizing fluid in the system for which the valves discussed in this docu=ent provide overpressure protection.
remote actuator: an electrically controlled pneumatic or electrical device which controls both the opening and closing of the main valve, generally by means of a pilot. valve which hydraulically operates the r.
main valve.
safety valve: see spring-loaded safe ty valve.
seat: stationary member of the valve seating surfaces.
seating force: compressive load on the valve seating surfaces.
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92 seating surfaces: mating surfaces of the disk (movable me=ber) and the (stationary member), the separation of which is termed valve seat opening and results in valve discharge.
secondarv orifice: term referring to the restricted flow area (resulting frem the arrangement of the valve adjusting rings)at the outside edge (radially away frem the seating surf ace.5) of the huddling cha=ber.
set pressure: the inlet pressure at which initial opening of the valve is intended to occur.
set oressure adiustment range:
the range of pressures over which the valve set pressure is adjustable and within which the valve perform-ance meets the specified requirenants.
set oressure adiustment screw or nut: the screw or nut which is used to compress the main spring of a spring-loaded safety valve and the set pressure adjustment spring of a pilot-operated pressure relief valve, thereby adjusting the set pressure of the valve.
set oressure adiustment sering: a spring used to pre-load the bellows of the pilot valve and adjust the set pressure in pilot-operated pres-sure relief valves.
set oressure tolerance: the allowable deviation between the lif t pres-sure and set pressure, expcessed in percent of the set pr2ssure (the r
allowable deviation in performance reproducibility).
si==er:
term applied to extensive seat leakage at inlet pressures
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approaching the set pressure of the valve. Simmar is characteristic of spring-loaded, safety valves in the absence of anti-st=mer devices.
soindle:
ene stem or shaf t of a spring-loaded safety valve that transmits the main spring load to the disk.
soring: see main spring.
soring-loade. s afe ty valve : a pressure relief valve that typically con-sists ou a disk held closed by a spring whose pre-load directly j
counters the load of the system pressure on the valve disk. The valve opens when the system pressure load on the disk exceeds the spring pre-load.
soring retention structure: the yoke or bonnet of a spring-loaded safety valve that maintains the spring pre-load.
static back oressure:
the pressure acting downstream of a closed valve.
tencerature sensitivitv: the average effect of the equilibrium tempera-ture of the valve on lif t pressure, expressed in a psi change in lift pressure per 100*F change in the temperature of the reactor coolant.
tria: internal parts of the valve such as the seat, disk, and wear and hard-f acing materials.
ycke: the part of a spring-loaded safety valve that maintains the spring pre-load and is attached to the body of the valve by a flange or lugs.
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