ML19221B080
| ML19221B080 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-05.2.2, NUREG-75-87, NUREG-75-87-5.2.2, SRP-05.02.02-01, SRP-5.02.02-1, NUDOCS 7907120371 | |
| Download: ML19221B080 (6) | |
Text
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STANDARD REV EW PLAN OFFICE OF NUCLEAR REACTOR REGULATION OVERPRESSURIZATION PROTECTION SECTION 5.2.2 REVIEW PESPONSIBILITIES Pri. nary - Reactor Systers Branch (RSB)
Secondary - Core Performance Branch (CPB)
Electrical, Instrurentation and Control Systems Branch (EICSB)
Mechanical Engineering Branch (MEB)
I.
(REAS OF REVIEW Ov?rpressure protection for the reactor coolant pressure boundary (RCPB) is provided by u of relief and safety valves. For RCPB overpressure protection, the relief and safety valves opi rate in conjunction with the reactor protection system and the steam generator safety vilves. The reviewer examines the design bases, system and corponent descriptions, and sys*.en
/nalyses and tests described in the applicant's safety analysis report (SAR) in order to evaluate the adequacy o' the overpressure protection which is provided. The areas of review for a boiiing water reactor (CWR) are the reactor coolant system relief and safety valves.
Enr. pressurizeJ water reactor (FWR), the areas of resiew are the pressurizer safety and relief valv%, and the piping from these valves to the quench tank. The review of anticipated transicnts withe ; scram is described in Standard Review Plan (SRP) 15.8.
ihe adequacy of the proposed preoperational and initial startup test programs is exarined as a part of this review. The reviewe' also evaluates the proposed technical specifications to assure that they are adequate in regard to limiting condicions of operation and periodic surveillance testing.
The overpressuro protection components are also reviewed to assure that they have the proper seismic and quality group classification. This aspect of the review is performed as a portior of the effort described in SRP 3.2.1 and SRP 3.2.2.
The MEB, as described in SRP 3.9.3. reviews the design and installation criteria for the overpressure protection components to assure that they are in confor~ance with ASME Boiler and Pressure Vessel Code requirements.
The EICSB, as described in SRP 7.6, evaluates the adequacy of controls and instrumentation of the overpressure protection components with regard to the required features of automatic actuation, remote sensing and indication, renote control, emergency onsite power, and connec-ticns to the reactor protection system.
USNRC STANDARD REVIEW PLAN
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The CPB provides generic evaluation of the mathematical models used to analyze tran-sients that result in an increase in the pressure within the reactor coolant system.
The APCSB reviews the adequacy of the pressure relief and safety valves for the secondary systen of PWR's.
II. ACCEPTANCE CRITERIA The fundamental criterion against which an evaluation of overpressure protection is to be made is General Design Criterion 15 (Ref. 1): "The reactor coolant system and associated auxiliary, control, and protects systems shall t'e designed with sufficient margin to assure that the design conditio: >
',f the reactor coolant pressure boundary are not exceeded during any condition of normal operaticn, including anticipated operational occurrences.'
Further, tha preoperational and initial startup test programs are to neet tho intent of kegulatery Guide 1.68 (Ref. 2).
To tm acceptable, adequate relief and safety val /e capacity nust t'e provided far the primary syste-of PWR's and CWR's.
For FWR's, the secondary systen rust also be provided with relief and safety valves having adequite capacity.
1.
Re_1_ief Valves Fer the design basis nornal operational transients, the relief valve capacity nust be sufficient to limit the pressure so as to prevent u fety valve discharge directly t the coatainment, with the following assumptions:
a.
The reactor is operating at licensed core thernal power level.
All systen and core pa'. meters are at the values within the norml operating b.
ranges which would produce the highest transient pressure, All components, instrumentation, and controls function normally.
c.
2.
Sa fetyv Valves For the nost severe abnormal operational transient, with reactor scram, the safety valve capacity should be sufficient to limit the pressure to less than 110 of the RCPB design pressure, as specified by the ASME Boiler and Pressure Vessel Code (Ref.
3), with suf ficient nargin to account for uncertainties in the design and operation of the plant and assuming:
The reactor is operating at a power qual to the licensed core thermal power a.
level plus an increment suf ficient to account for power reasurement uncertainties.
All systen and core para reters are at the values within the normal operating range, b.
including uncertainties ard technical specifica:icn limits, which would result in the highest transient pressure.
5.2.2-2
C.
The reactor scram is initiated either by the high pressure sicnal or by the second signal from the reactor protection systen, whichever is la'er.
d.
The dischar9e flow is based on the rated capacities spacified in the ASPE Boiler and Pressure Vessel Code for each type of valve.
III.
FEVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to assure that the design criteria and bases and the preliminary design as set forth in the preliminary safety analysis report reet the acceptance criteria given in Section II of this plan.
For operating license (OL) applications, the procedures are used to verify that the initial desior criteria and bases have been appropriately implerented in the final des gn as set i
forth in the final sa fety analysis report and in the report on overpressu e protection. The latter report is requi ed by the ASME Ccde and is to serve as the basis for nany of in-dividual review steps outlined below during the CL review. The CL review also includes the proposed technical specitications, to assure that they are adequate in regard to limiting conditions of operation and periodic surveillance testing.
The follcaing steps are taken by the RSS reviewer in determining that the acceptance criteria of Section II have been et.
These steps should be applied to CP and OL reviews as approrriate. Freviously reviewed designs nay be used as a guide; however, the reviewer rust verify that any changes are justified.
1.
The piping and instru entation diagrams are examined to determine the number, type, and location of the safety and relief valves in both the prim?r; and secondary systems, and of discharge lines, instru entation, and cther co ponents.
2.
All other functions of the ccrponents, instru ents, or controls used for overpress ce protection and the interfaces with ail other systems are identified. The effects of thmse other functions or systems on operation of the overpressure protection systen are determined.
3.
The valve descriptions are examined to deternine type and nanufacturer and to evaluate reliability (e.g., new t - standard design).
4.
The capacities, set points, and setpoint tolerances for all safety and relief valves are identified.
5.
All of the reactor trip signals which occur during overpressure transients, including their setpoints and setpoint tolerances, are identified.
6.
All transients analyzed in Chapter 15 of the SAR that result in an increase in the pressure experienced by the RCFB are examined. The peak predicted pressures are identified and the operating conditions and setpoints used in the analysis are reviewed to assure that they are suitably conservative.
1470 E)f) 5.2.2-3
The information be!cw is provided to the reviewer as guidance and is based on typical previously-reviawed designs.
a.
BWR's - For relief valve sizing, in transients in which a scram is initiated by closure of the main steam isolation valves or fast closure of the turbine stop valves, the highest pressure resWts from instantaneous loss of condenser vacuun or a turbine trip without bypass. For safety valve sizing, in transients in which scram is initiated by high flux or high pressure, the highest pressure results from closure of all nain steam isolation valves. Analysis of previouC/ accept-able designs has shown that the peak pressure is at least 25 psi below the al-lowable, assuming the reactor is operating at 105: of rated power, pressure is 1040 psia, no c ndit is taken for relief valve repacity, one safety valve fails to open, and stram is initiated by high pressure.
b.
PWR's - for relief valve sizing, the valve capacity has been sufficient to accomodate the surge from the design basis step load change. Safety valve sizing is usually based on the maximun surge ratt that results from a turbine trip without bypass. Analysis of previously acceptable designs has shown that the discharge flow from the safety valves in the primary and secondary systems is typically 85! and 100: of their respective rated capacities assu-iing the react )r is initially at 102; of rated. Ser, the uncertainties in power, oressure, and temperature are St. 30 psi, and 4"F, respectively, scran is initiated by low level in the steam generator; and no credit is taken for relief valve operation or Doppler er noderator temperature reactivity feedback.
7.
The applicant's proposed preoperational and initial star; test prograns are revicaed s
i tant of 9egulatory Guide 1.68 (Ref.
ta determine that they are consistent with tha n
2).
At the OL stap, this aspect of the,eview is to assure that sufficient infor-mation is provided by the applicant to identify clearly the test objectives, rethods of testing, and acceptance criteria (Sec par. C.2.b of Regulatory Guide 1.68.)
The reviewer evaluates the proposed test programs to determine if they provide a reasonable assurance that the corponents that provide overpressure protection will perform their safety function. As an alternative to this detailed evaluation, the revieaer may compare the overpressure protection design to that of a previously reviewed plant.
I' che design is essentially identical and if the proposed test programs are essentially the same as perforned previously on other plants, the review-er may conclude that the proposed test programs for overpressu e protection are adequate.
If the proposed design differs significantly from that of previously reviewed designs, the impact of the proposed changes on the preoperational and initial startup testing programs are reviewed at the construction permit stage. This effort should parti ^u-larly evaluate the need far any special design features requireri to perforr. acceptable test programs.
5.2.2-4
8.
The proposed plant technical specifications are reviewed to:
a.
Confirm the suitability of the limiting conditions of operation, including the proposed time limits and reactor operating restrictions for periods when system equipment is inoperable due to repairs and maintenance.
b.
Verify that the frequency and scope of periodic surveillance testing is adequate.
IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains sufficient infornation and his review supports the following kinds of statements and conclusions, which should be included in the staff's safety evaluation report:
1.
BWR's "The pressure relief system prevents overpressurization of the reactor coolant pressure boundary under the most severe transients and limits the reactor pressure during normal operational transients. Overpressure protection will be provided by safety and relief valves located on the four main steam lines between the reactor vessel and the first isolation valve inside the dryxell. The relief and saf,
valves are distributed among the four main steam lines such that a single accident cannot disable the safety, relief, or automatic depressurization functions. The valves discharge through piping to the suppression pool. The valves operate as spring-loaded safety /alves with set pressures that range from to psig. Their total capacity at their set pressure is T of rated steam flow.
"To determine the ability of the pressure relief system to prevent overpressuriza-tion, the applicant analyzed the severe transient of main steam isolation valve
<losure. The analysis was performed assuming that: a) the plant is in operation at design conditions (*J of rated steam flow and a reactor vessel dome pressure of
- psig), and b) the reactor is shut down by a high pressure scram. The calculated neak pressure at the bottom of the vessel is psig, a margin of psi below the code allowable of psig (110% of vessel design pressure). The staff concludes that the design of the pressure relief systems conforms to the Commission's regula-tions and to ap:icable regulatory guides, staff technical positions, and industry standards and is acceptable."
2.
PWR's "The pressure relief system prevents overpressurization of the reactor coolant pressure boundary under the most severe transients and limits the reactor pre.ssure during normal ope ational transients. Overpressure protection for the reactor cool-ant pressure boundary is accomplished by utilizing the safety valves. These valves discharge to the pressurizer quench tank through a comnon header from the pressurizer. The reactor coolcnt system (RCS) safe " valves, in conjunction with the steam generator safety valves, and the reactor protecti vstem, will protect the RCS against overpressure in the event of a complete loss of heat sink.
- Normally, BWR's are analyzed at 105% rated steam flow at a pressure of 1040 psig.
5.2.2-5
!7 057
"The peak RCS pressure following the worst transient is limited to the ASME Code allowable (110% of the design pressure) with no credit taken for operation of RCS relief valves, steam line relief valves, steam dump system, RCS pressurizer level control system, or pressurizer spray. The plant was assumed to be operating et design conditions ( % of rateo power) and the reactor is shut down by a scram. The calculated pressure at the bottom of the vessel is psig, a margin of psi below the code allowable of psig (110$ of vessel design pressure).
The staff concludes that the design of the pressure relief system conforms to the Comission's regulations and to applicable regu?atory guides, staff technical positions, and industry standards, and is acceptable."
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 15, " Reactor Coolant System Design."
2.
Regulatory Guide 1.68, "Preoperational and Initlal Startup Test Programs for Water-Cooled Power Reactors."
3.
ASME Boiler and Pressure Vessel Code,Section III, Article NB-7000, " Protection Ag-inst Overpressure," Anerican Scciety of Mechanical Engineers.
O 5.2.2-6 147 058
pR REGg NU R EG.75/087 f
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STANDARD REVIEW PLAN 4
U.S. NUCLEAR REGULATORY COMMISSION
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 5.2.3 REACTOR COOLANT PRES 5URE BOUNDARY MATERIALS REVIEW RESPONSIBILITIES Primary - Materials Engineering Branch (MTEB)
Secondary - None I.
AREAS OF REVIEW General Design Criteria 1, 13, 14 ar.d 31 require that the reactor coolant pressure bound-ary shall be designed, fabricated, erected and tested so as to have an extremely low probability of a rapidly propagating failure and of a gross rupture.
The Criteria also require that the reactor coolant pressure boundary shall be tested to quality standards commensurate with the safety function to be performed and that instrumentation shall be provided to monitor the variables that can affect the integrity of the reactor coolant pressure boundary.
The following areas, which relate to materials of the reactor coolant pressure boundary (RCPB) other than the reactor pressora vessel, which is covered in Standard Review Plan 5.3.1, " Reactor Vessel Materials", are reviewed.
1.
Material Specifications The specifications for pressure-retaining ferritic materials and austenitic stainless steels, including weld materials, that are used for each component (e.g., vessels, piping, pumps, and valves) of the reactor coolant pressure boundary, are reviewed.
The adequacy and suitability of the ferritic materials, stainless steel, and non-t ferrous metals specified for the above applications are r eviewed.
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2.
Compatibility of Materials with the Reactor Coolant General corrosion and stress corrosion cracking induced by impurities in the reactor l coolant can cause f ailures of the reactor coolant pressure boundary.
The chemistry of the reactor coolant and the additives (such as inhibitors) whose function is to control corrosion are reviewed.
The water chemistry includes the permissible concentrations of chlorides, fluorides, oxygen, hydrogen, and soluble poisons, the methods used to control the concentrations of impurities, and the ph.
USNRC STAND ARD REVIEW PLAN stende,d row w piene ere e,oper.d to, the av dence of t*e o+nc. of Nuci.e, R eto, R.goiet on e+ett roepon.,bie fo, the reve.w of sopocetione io construci end opereto nuc6eer power plente These documente see enede eveden;e to the public es part of the Commiseson e polky to 6riform the nucieer indvetry and the gertere. pub #e of reg.etetory procedures and po4icles Standard rewtow piene see not subetitutee f or requietory guedee or the Consmissaon e regulatione end sofrip4Me Wr4ttt theast le not required The etenderd review plan etctione are heyed to Revienon 2 of the Stenderd Forenet and Content of sateev Anoe eee Reporte v
for Nocient Pwurer Pt*Me Not en sectices of 1954 Stenderd Formet have e corroependmg review plan Puhdehed stendard rewWw piene wiR be rewteed periodicolty. es appropriete to ecceenmodate commente end to reflect new information and enre<5ence C.ed,imprete end ouggest#@ne for Imptoweenent w(H be Conaldered afsd shoqqd be sent to tito u S Nucteer R.gulatory Commiseson Othce of Nuc3ee, Reectet
.t.noc -
lg7 999 7007120394 Rev. 1
The review includes the compatibility of the materials of construction employed in the RCPB with the reactor coolant, contaminants, or radiolytic products to which the systera is exposed. The extent of the corrosion of ferritic low allcy steels and carbon steels in contact with the reactor coolant is reviewed. Similarly, a review is made of possible uses of austenitic stainless steels in the sensitized condition.
The use of austenitic stainless steels in boiling water reactors (BWR's) requires special attention because of the oxygen content of BWR coolant.
3.
Fabrication and Processing of Ferritic Materials a.
The fracture toughness properties of ferritic materials used for pressure-retaining components of the reactor coolant pressure boundary are reviewed.
The fracture toughness tests performed on all ferritic materials used for pressure-retaining RCPB components (i.e., vessels, pumps, valves, and piping) are reviewed.
The test procedures used for Charpy V-notch impact and dropweight testing are reviewed.
Fracture toughness of the material is characterized by its reference tempera-ture, RT This temperature is the higher of the nil-ductility temperature NDT.
(NDT) from the dropweight test or the temperature that is 60 F below the temperature at which Charpy V-notch impact test data are 50 ft-lbs and 35 mils lateral expansion. The limiting RT temperature of the material is reviewed.
NDr b.
The control of welding in ferritic steels is reviewed.
(1) The quality of welds in low alloy steels can be increased significantly by proper controls. In particular, the prcpensity for cold cracks or reheat cracks to form in areas under the bead and in heat-af fected zones (HM) can be minimized by maintaining proper preheat temperatures of the base metal concurrent with controls on other welding variables. The minimum preheat temperature and the maximum interpass temperatures are reviewed.
(2) The quality of electroslag welds in low alloy steel components can be in-creased by maintaining a weld solidification pattern that possesses a strong intergranular bond in the centet of the weld.
The welding vari-ables, which have a significant effect on the weld solidification pattern, must be controlled. The welding variables, solidification pat +. erns, macro-etch tests, and Charpy V notch impact tests of electroslag welds are reviewed.
(3) Experience shows that a welder qualified to weld low-alloy steel or carbon steel components under normal fabricating conditions may not produce ac-ceptable welds if the accessibility to the weld area is restricted.
5.2.3-2 147 060 Rev.,
Limited accessibility can occur when component parts are joined in the final assembly or at the plant site, where other adjacent components or structures prevent the welder from issuming an advantageous position during the welding operation. The adequacy of accessibility dtring the welding of ferritic components is reviewed.
c.
The requirements for nonc'estructive examination of ferritic wrought serless tubular products used for ASME Class 1 components of nuclear power plants are specified in Paragraphs NB-2550 thrcugh NB-2570, ASME Boiler and Pressure Vessel Code (hereafter "the Code"),Section III.
The methods of examination specified for nondestructive examination are reviewed.
4.
Fabrication and Processing of Austenitic Stainless Steel Austenitic stainless steels in a variety of product forms are used for construction of pressure retaining components in the reactor coolant pressure boundary. Unstabi-lized austenitic type stainless steels, which include American Iron and Steel Institute (AISI) Types 304 and 316, are normally used.
Because these compositions are susceptible to stress corrosion cracking when exposed to certain environmental conditions, process controls must be exercised during all stages of component manu-facturing and reactor construction to avoid severe sensitization of the material and to minimize exposure of the stainless steel to contaminants that could lead to stress corrosion cracking.
a.
Sensitization is caused by intergranular precipitation of chromium carbide in dustenitic stainless steels that are exposed to temperatures in the approximate range of 800 F to 1500 F.
Precipitation increases with increasing carbon content and exposure time.
Control of the application and processing of stain-less steel is needed to eliminate the occurrences of stress corrosion cracking in sensitized stainless steel components of nuclear reactors. Test data and service experience demonstrate th3t sensitized stainless stoel is significantly more susceptible to stress corrosion cracking than nonsensitized (solution heat treated) stainless steel.
Special provisions may apply to the use of austenitic stainless steel in bciling I water reactor (BWR) piping because plant operating experience indicates that reactor coolant boundary piping is susceptible to oxygen-assisted stress corro-sion cracking.
b.
The following areas are reviewed: requirements for solution heat treatment of stainless steel; plans to avuid partial or severe sensitization during welding, including information on welding methods, heat input, and interpass temperatures; and a description of the material inspection program that will be used to verify that unstabilized austenitic stainless steels are not susceptible in service to intergranular attack.
147 061 5.2.3-3 Rev. 1
Contamination of austenitic stainless steel with halogens and halogen-bearing compounds (e.g., die lubricants, marking compounds, and masking tape) must be avoided to the maximum degree possible to avoid stress corrosion cracking.
Plans for cleaning and protecting the material against contaminants capable of causing stress corrosion cracking during fabrication, shipment, storage, cen-struction, testing, and operation of components and systems are reviewed. Any pickling used in processing at:stenitic stainless steel components and the restrictions plac ed on pickling sensitized materials are reviewed. The upper limit on the yield strength of austenitic stainless steel materials is reviewed.
c.
Whether sensitized or not, austenitic stainless steel is subject to stress cor-rosion and must be protected from contaminants that can promote cracking.
Thermal insulation is often employed adjacent to, or in direct contact with, stainless steel piping and ccmponents. The contaminants present in the thermal insulation may be leached by spilled or leaking liquids and d(posited on the stainless steel surfaces. The controls on the use of nonmetallic thermal insulation are reviewed.
d.
Austenitic stainless steel is subject to hot cracking (microfissuring) during welding if the weld metal composition or the welding procedure is not preperly controlled. Because cracks formed in this manner are small and difficult to detect by nondestructive testing methods, welding procedures, weld metal compo-sitions, and delta ferrite percentages that minimize the possibility of hot cracking must be specified. As a part of achieving this control, Regulatory Guide 1.31, " Control of Ferrite C.antent in Stainless Steel Weld Metal" contains recommendations for process control through the testing of weld test pads.
The staff recommendations will provide assurance that the ferrite content will be adequate to prevent microfissuring. The adequacy of the proposed welding procedures is reviewed.
The assurance of satisfactory electroslag welds for austenitic stainless steel components can be increased by maintaining a weld solidification pattern with a strong intergranular bond in the center of the weld.
The welding variables that have a significant effect on the weld solidification pattern must be controlled.
A number of electroslag welding prccess variables, such as, slag pool depth, electrode feed rate and oscillation, current, voltage, and slag conductivity, have been shown to influence the weld solidification pattern If the combination.
I of process variables produces a deep pool of molten weld metal, the crystal l
(dendritic) growth direction from the pool sides will join at an obtut,e angle at the center of the weld, and cracks may develop because of the weaker centerline bond between dendrites. A proper combination of process variables promotes a dendritic growth pattern with an acute joining angle, which results in a strong 147 062 Rev. 1
centerline bond.
The weldii,g variables, solidification patterns, and macro etch tests used in the electroslag welding of dustenitic stainless steel are reviewed.
Experience has shown that a welder qualified to weld stainless steel components under normal fabricating conditions may not produce acceptable welds if the ac-cessibility to the weld area is restricted. Limited accessibility can occur when Component parts are joined in the final assembly or at the plant site, where other adjace-t components or structures prevent the welder from assuming an advanta wA position during the welding operation. The adequacy of accessibility of field erected structures, for welding austenitic stainless steel components, is reviewed.
The requirements for nondestructive examination of wrought seamless tubular e.
products used for components of nuclear power plants are specified in Paragraph NB-2550 of the Code, Section 411.
Nondestructive examination techniques applied to tubular products used for components of the RCPB, or other safety-related ASME Class 1 systems that are designed for pressure in excess of 275 psig or temperatures in excess of 200 F, must be capable of detecting unacceptable defects regardless of defect shape, orientation, or location in the product.
The nondestructive examination procedures used for inspection of tubular products are reviewed.
Inservice inspection requirements for the Reactor Coolant Pressure Boundary are described in SRP 5.2.4, " Inservice Inspection and Testing of Reactor Coolant Pressure Boundary."
II.
ACCEPTANCE CRITERIA The acceptance criteria for the areas of review described in Section I of this plan are as follows:
1.
Material Specifications The specifications for permitted materials are those identified in the ASME Code, f
Section III, Appendix I, or described in detail in the ASME Code,Section II, l
Parts A, B, and C.
Regulatory Guide 1.85, " Code Case Acceptability ASME Section III l
Materials," describes the acceptable Code Cases to be used in conjunction with the above specifimations.
Special requirements for EWR piping materials and materials processing are described in Branch Technical Position MIEB 5-7,* " Material Selection and Processing Guidelines for CWR Coolant Pressure Boundary Piping."
i Branch Technical Position MIEB 5-7 is identical to the content of NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping."
147 063 5.2.3-5 Rev. 1
2.
Compatibility of Materials with the Reactor Coolant In boiling water reactors (BWR's), high purity water is maintained. The purity is monitored through continuous on-line reading of the conductivity of the coolar.t and by periodically sampling and chemically analyzing it for pH and chloride content.
An on-line water treatment plant maintains the coolant within Technical Specification l i ri,i t s. In reactor coolants used for BWR's, oxygen seeks a natural level, and no attempt is made to control the amount of oxygen contained in the solution. The acceptance criteria for the chemistry of the BWR coolant are specified in Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors." Acceptable locations for chemistry monitoring sensors are described in Regulatory Guide 1.56.
In reactor coolants used for pressurized water reactors (PWR's), the conductivity measurements tend to show high values due to interference from additions of boric acid.
These additions tend to mask the effect of other impurities. Therefore, sampling and chemical analysis for chlorides, fluorides, and oxygen must be performed on a scheduled basis. Acceptable levels of chloride, fluoride and oxygen for pressurized water reactor coolant purity are stated in Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel."
ferritic low alloy steels and carbon steels, which are used in many principal pressure-retaining components, are clad with a layer of austenitic stainless steel. If cladding is not used, conservative corrosion allowances must be indicated for all exposed surfaces of carbon and low alloy steels, as indicated in the ASME Code,Section III, NB-3120, " Corrosion."
Unstablilized austenitic stainless steel of the AISI Type 3XX series used for com-ponents of the RCPB must conform to requirements of Regulatory Guide No. l.44 and Branch Technical Position MTEB 5-7, includingverificationofnoneensitizationofthel material by an approved test.
3.
Fabrication and Processing of Ferritic Materials The acceptance criteria for fracture toughness are provided by the ASME Code, l
a.
Section III; and 10 CFR Part 50, Appendix G.
The pressure-retaining components of the RCPB that are made of ferritic materials must meet the requirements for fracture toughness during system hydrostatic tests and any condition of normal operation, including anticipated operational occurrences.
With respect to absorbed energy in ft-lbs and iateral expansion as shown by Charpy V-notch (C ) impact tests, all materials must meet the acceptance standards of Article NB-2330 of the Code,Section III, and the requirements of Sections IV.A.2, IV.A.3, and IV.B of Appendix G, 10 CFR Part 50, as follows:
(1) The special acceptance r equirements for f racture toughness of reactor vessels are covered by Standard Review Plan 5.3.1, " Reactor Vessel Mate-ials."
147 064 5.2.3-6 Rev. 1
(2) Materials for piping (i.e., pipes, tubes, and fittings), pumps, and valves, excluding bolting materials, must meet the requirements of the Code,Section III, Paragraph NB-2332, and Appendix G, Paragraph G-3100.
The required C values for piping are specified in Table NB-2332-1 of the Code, Section Ill.
(3) Materials for bolting for which impact tests are required must meet the requirements of the Code,Section III, Paragrapn NB-2333, and Appendix G, Paragraph G-4100.
(4) Calibration of instruments and equipment must meet the requirements of the Code,Section III, Paragraph NB-2360.
b.
The acceptance criteria for control of ferritic steel welding are listed below:
(1) The amount of specified preheat must be in accordance with the requirements of the Code,Section III, Appendix D, Paragraph D-1200, supplemented by Regulatory Guide 1.50, " Control of Preheat Temperature for Welding Low Alloy Steel."
The supplemental acceptance criteria for control of preheat temperature are as follows:
The welding procedure qualification requires that minimum preheat and maximum interpass temperatures be specified and that the welding procedure be qualified at the minimum preheat temperature. For production welds, the preneat temperature should be maintained until a post-weld he + treat-ment has been performed.
The preheat controls described in the Westinghouse Topical Report WCAP-8577 are an acceptable alternate to compliance with those of Regulatory Guide 1.50, " Control of Preheat Temperature for Welding Low Alloy Steel."
Production welding should be monitored to verify that the limits on preheat and interpass temperatures are maintained. In the event that the above criteria are not met, the weld is subject to rejection.
(2) The acceptance criteria for electroslag welds are presented in Regulatory l
Guide 1.34, " Control of Electroslag Weld Properties." These criteria specify acceptable solidification patterns and impa:t test limits (for qualification of welds in Class 1 and Class 2 components) and the criteria for veri'ying conformance during production welding.
(3) Reguiatory Guide 1.71, " Welder Qualification for Areas of Limited Accessibil-ity," provides the following criteria for requalification of welders: the performance qualification should require testing of the welder when 5.2.3-7 c
conditions of accessibility to a production weld are less than 30 to 35 cm (12-14 inches) in any direction from the joint; and requalification is required for different restricted accessibility conditions or when aty of the essential variables listed ir the Code,Section IX, are changed.
Qualification of the Welder or Welding Operators for limited accessibility may be waived provided that 100% Radiographic and/or Ultri;onic examination!
of the completed welded joint is performed. Examination procedures and acceptance standards should meet the requirements of the ASME Section 111 of the Code. Records of the examination reports and radiographs should be retained and made part of the Quality Assurance Documentation for the completed weld.
c.
Acceptance criteria for nondestructive examination of ferritic steel tubular products are given in the ASME Code,Section III, Paragraph NB-2550.
4.
Fabrication and Processing of Austenitic Stainless Steel a.
The acceptance criteria fcr testing, alloy compositions, and heat treatment, to avoid sensitization in austenitic stainless steels, are covered in Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel," and additionai criteria for BWR's are in Branch Technical Position MIEB 5-7.
b.
Controls to avoid stress corrosien cracking in austenitic stainless steels are also covered in Regulatory Guide 1. 44.
This guide provides acceptance criteria on the cleaning and protection of the material against contaminants capable of causing stress corrosion cracking. Acid pickling is to be avoided on fabricated stainless steels. Necessary pickling is to be done only with appropriate controls. Pickling should not be performed upon sensitized stainless steel,s.
The quality of water used for final cleaning or flushing of finished surfaces during installation is in accordance with Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water Cooled Nuclear Power Plants." Vented tanks with deionized or demineralized water are an acceptable source of water for final cleaning or flushing of finished surfaces. The oxygen content of the water need not be controlled.
Laboratory stress corrosion tests and service experience provide the basis for the criterion that cold worked austenitic stainless steels used in the reactor coolant pressure boundary should have an upper limit on the yield strength of 90,000 psi.
The compatibility of austenitic stainless steel materials with thermal insulation c.
is dependent upon the type of insulation. The thermal insulation is acceptable if either reflective metal insulation is employed or a nonmetallic insulation which meets the criteria of Regulatory Guide 1.36, " Nonmetallic Thermal Insulation 5.2.3-o 147 066 Rev. 1
for Austenitic Stainless Steel" is used. The acceptance criteria for nonmetallic insulation fo' stainless steel are based on the levels of leachable contaminants in the material and are preseM ed in position C.2.b and Figure 1 of the guide.
d.
The acceptance criteria for desta ferrite in austenitic stainless steel welds are given in Regulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal."
These acceptance criteria cover (1) verification of delta ferrite content of filler metals, (2) ferrite measurement, (3) instrumentation, (4) acceptability of test results, and (5) documentation of weld pad verification test.
The acceptance criteria for electroslag welds in austenitic stainless steel are given in Regulatory Guide 1.34, " Control of Electroslag Weld Properties." These criteria specify acceptable solidification patterns for qualification of austenitic stainless steel welds and the basis for verifying confornance during production welding.
Regulatory Guide 1.71, " Welder Qualification for Areas of Limited Accessibility,"
provides the following criteria for requalification of welders:
(1) The performance qualification should require testing of the welder when conditions of accessibility to a production weld are less than 30 to 35 cm l (12-14 inches) in any direction from the joint.
9 (2) Requalification is required for different restricted accessibility conditions or when other essential variables listed in the Code,Section IX, are changed. An alternate acceptance criterion is as stated in II.3.b.
l e.
The acceptance criteria for nondestructive examination of austenitic stainless steel tubular products are stated in the ASME Code,Section III, Paragraph i
NB-2550.
III. REVIEW PROCEDURES The reviewer will select and emphasize material from the procedures described below, as may be apprcpriate for a particular case.
For each area of review described in Section I of this plan, the following review procedures are followed.
1.
Material Specifications The material specifications for each major pressure-retainlag component or part used in the RCPB are compared with the acceptable specifications listed in the Code, Sections II and III, as stated in the acceptance criteria. Exceptions to the material specifications of the Code are clearly identified, and the basis evaluated. The reviewer judges the significance of the exceptions and, taking into account precedents set in earlier cases, determines the acceptability of the proposed exceptions.
In those instances where the Miterials Engineer ~ng Branch takes exceptioc to the use of 5.2.3-9 Rev. I
a specific material or questions certain aspects of a specification, the applicant is advised which material is not acceptable, and for what reason.
2.
Compatibility of Materials with the Reactor Coolant The reviewer verifies that the following information is provided at each respective stage of the review process:
a.
At the construction permit stage of review:
(1) A list of the materials of construction of the components of the reactor coolant pressure boundary that are exposed to the reactor coolant, in-cluding a description of material compatibility with the coolant, con-taminants, and radiolytic products to which the materials may be exposed in service.
(2) A li-of the materials of construction of the RCPB, and a description of material compatibility with external insulation and with the environment in the event of reactor coolant leakage.
(3) The limitations imposed on concentrations of chloride and fluoride ions and oxygen in the reactcr coolant, and the extent of monitoring such limitations.
(4) The fabrication and cleaning controls imposed on stainless steel components to minimize contamination with chloride and fluoride ions.
(5) The controls and limits that are specified for leacha.le impurities in thermal insulation, as identified in Section II.4.c.
(6) For BWR's, performance monitoring recommended in Regulatory Guide No.
- 1. 56, l
" Maintenance of Water Purity ia Boiling Water Reactors."
b.
At the operating license stage of the review process:
(1) The items listed under 2.a above, to prov- > assurance that any changes are noted that may have occurred during the period between the sut,mittal of SAR's.
(2) A list of the instrumentation and equipment that will monitor and control the purity of the reactor coolant, including water purity indicators and alarms provided in the control room.
3.
Fabrication and Processing of Ferritic Materials a.
The information submitted by the applicant relative to tests for fracture toughness is reviewed fsr conformance with the acceptance criteria stated In Section II.3.a.
These tests include Charpy V-notch impact and dropw2;ght tests. A description of the tests is reviewed, and the locations of the test specimens and their oricolation are verified. Information regarding calibration of g
instruments and equipment is reviewed for conformance with the acceptance criteria stated in Section II.3.' (4).
5.2.3-10 Rev. 1
In the event that none of the fracture toughness Msts has been performed, the preliminary safety analysis report (PSAR) must contain a statement of the appli-cant's intention to perform this work in accordance with the Code, iection III, Paragraph NB-2300 and Appendix G; and the requirements of 10 CFR 50, Appendix G.
The final safety analysis report (FSAR) is reviewed to assure that all the impact tests required by NB-2340 have been performed.
f b.
The control of welding in fer-itic steels is reviewed as described below:
(1) The information submitted by the applicant regarding the control of preheat temperatures for welding low alloy steel is reviewed for conformance with the acceptance criteria stated in Section II.3.b.(1).
(2) The electroslag welJ information submitted by the applicant is reviewed for conformance to the acceptance criteria discussed in Section II.3.b.(2).
The information in the SAR is reviewed to verify that macroetch tests have been made (to assure that an acceptable weld solidification pattern is obtained) and that impact tests specified in Regulatory Guide 1.34 meet the acceptance criteria discussed oreviously in Section II.3.b.(2).
(3) The ASME Code,Section III, requires adherence to the requirements of Section IX, " Welding Qualifications." One of the requirements is welder qualification for production welds. However, there is a need for supplement-ing this section of the Code because the assurance of providing satisfactory welds in locations of restricted dire' nhysical and visual accessibility can be increased significantly by qualifying the welder under conditions simulating the space limitations under which the actual welds will be made.
Regulatory Guide 1.71, " Welder Qualification for Limited Accessibility,"
provides the necessary supplement to the Code,Section IX, in this respect.
The information submitted by the applicant is reviewed for conformance with acceptance criteria discussed in Section II.3.b.(3).
The ASME Code,Section III, NB-2550 specifies the ultrasonic method for examina-c.
tion of ferritic steel tubular products.
4.
Fabrication and Processing 'f Austenitic Stainless Steels a.
The information submitted by the applicant in the following areas is reviewed for conformance with the acceptance criteria stated in Section II.4.a regarding:
(1) The desirable stage in the sequence of processing for solution heat treat-ment, the rates of cooling, and the quenching media.
(2) Controls to prevent sensitization during welding, as described in Regulatory Guide 1.44.
5.2.3-11 Rev. I
(3) Controls to verify non-sensitization, as described in Regulatory Guide 1. 44. l (4) For BWR's, additional processing controls, as described in Branch Technical Position MfEB S-7.
In the event that information in the above areas is not supplied, sufficient justification for the deviation must be presented.
b.
The information submitted by the applicant is reviewed for conformance with the acceptar.ce criteria discussed in Section II.4.b as follows:
Verification is sought that process controls are exercised during all stages of component manufacture and reactor construction to minimize the exposure of aus-tenitic stainless steels to contaminants that could lead to stress corrosion cracking.
Information is also checked to assure that precautions have been taken to require removal of all cleaning solutions, processing compounds, degreasing agents, and any other foreign material from the surfaces of the component at any stage of processing prior to any elevated temperature treatrrent and prior to hydrotests.
The reviewer verifies that a statement is contained in the SAR that pickling of sensitized austenitic stainless is avoided and that the quality of water used l
for final cleaning or flusning of finished surf aces t uring installation is in accordance with acceptance criteria discussed in Section II.4.b.
Because excessive cold work in austenitic stainless steel can render this material susceptible to stress corrosion cracking, control n ust be exerted by the appli-cant, by placing an upper limit on the yield strength, in accordance with the acceptance criteria discussed in Sectinn II.4.b.
Verification is obtained that the applicant has such a control measure.
The information submitted by the applicant is reviewed to determine the type of c.
insulation used and to determine its compatibility with the austenitic stainless steel used in construction of the component.
There are no compatibility concerns with the use of reflective metal insulation; the chief compatibility concern is with the use of nonmetallic insulation. A review is performed to assure that any such material specified by the applicant is in conformance with the acceptance criteria stated in Section II.4.c.
Verifi-cation is obtained that the material has been chemically analyzed by methods equivalent to those prescribed in Regulatory Guide 1.36 and that evidence is ob-tained that the levels of leachable contaminants are such that stress corrosion of stainless steel will not result from use of the insulation.
d.
The information submitted by the applicant regarding control of delta ferrite in austenitic stainless steel welds is reviewed to determine its conformance with S.2.3-12 Rev. 1
the acceptance criteria stated in Section II.4.d.
The information submitted must state that appropriate filler metal acceptance tests have been conducted and that a certified materials test eeport has been received. The information should state, also, the applicant's program for compliance with the staff positions in Regulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal."
The information submitted by the applicant regarding control of electroslag weld properties for austentic stainless steel materials i s rev iewed f or conformance with the acceptance criteria discussed in Section II.4.d.
Tha
..iew of information on the control of electroslag weld properties in aus-tenitic stainless steels is essentially the same as that discussed previously for ferritic steels.
However, because electroslag welded austenitic stainless stetis have very high impact resistance and because the Code,Section III, is not concerned with impact testing of these welds, the checks are: (1) a macroetch test is used to provide assurance that the solidification pattern is in ac-cordance with the requirement of the acceptance criteria shown in Section II.4.d, and (2) wrought stainless steel parts are solution heat treated after welding.
The review procedure for information submitted on welder qualification for limited accessibility areas, applicable to austenitic stainless steels, is the sdme as that for ferritic steels, which has been discussed previously under Section III.3.b.(3).
The procedures fcr review of nondestructive examination of tubular products e.
fabricated from austenitic stainless steel are the same as those discussed for similar ferritic products in Section III.3.c of this plan, and the acceptance criteria are as shown in Section II.4.e.
5.
General If the information contained in the safet, analysis reports or the plant Technical Specifications does not comply with the a,propriate acceptance criteria, or if the information provided is inadequate to establish such compliance, a request for additional information is prepared and transmitted. Such requests identify not only the necessary additional information but also the changes needed in the SAR or the Technical Specifications. Subsequent amendments received in response to these re-quests are reviewed for compliance with the applicable acceptance criteria.
IV.
EVALUAi!ON FINDINGS The reviewer verifies that sufficient and adequate information has been provided to satisfy the requirements of the review plan and that his evaluation supports conclusions of the following type, to be included in the staff's safety evaluation report:
147 071 5.2.3-13 Rev. 1
"The materials used for construction of components of the reactor coolant pressure t,oundary (RCPB) have been identified by specification and found to be in conform-ance with the requiremerts of Section III of the ASME Code.
"The materials of construction of the RCIB exposed to the reactor coolant have been identified and all of the materials are compatible with the expected environment, as proven by extensive testing and satisfactory performance. General corrosion of all materials, except unclad carbon and low alloy steel, is negligible. For these materials, conservative corrosion allowances have been provided for all exposed surfaces in accordance with the requirements of the Code,Section III.
"The materials of construction for the RCPB are compatible with the thermal insula-tion used in these areas and are in conformance with the recommendations of Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steels."
"The controls imposed on reactor coolant chemistry are in conformance with the recom-mendations of Regulatory Guide 1.44, " Control of Sensitized Stainless Steel," and Regulatory Guide 1.56, " Maintenance of Weter Purity in BWR's," and provide reasonable assurance that the RCPB components will be adequately protected during operation from conditions that could critically lead to stress corrosion of the materials and loss of structural integrity of a component. The instrumentation and sampling provisions for monitoring reactor coola7t water chemistry provide adequate measurement capability for detecting significant changes on a timely basis. Compliance with the recommenda-tions of these Regulatory Guides constitutes an acceptable basis for satisfying the applicable requirements of General Design Criteria 14 and 31.
"The fracture toughness tests required by the ASME Code, augmented by Appendix G, 10 CFR 50, provide reasonable assurance that adequate safety margins against nonductile behavior or rapidly propagating fracture can be established for all pressure retaining components of the reactor coolant pressure boundary.
"The use of Appendix G of the ASME Code,Section III, and the results of fracture toughness tests performed in accordance with the Code and NRC Regulations in establish-ing safe operating procedures, provides adequate safety margins during operating, testing, maintenance, and postulated accident conditions. Compliance with these Code provisions and NRC Regulations canstitutes an acceptable basis for satisfying the l
requirements of General Design Criterion 31.
"The controls imposed on welding preheat temperatures are in conformance with the recommendations of Regulatory Guide 1.50, " Control of Preheat Temperature for Welding low Alloy Steels." These controls provide reasonable assurance that cracking of componunts made from low alloy steels will not occur during fabrication and minimize the possibility of subsequent cracking due to residual stresses being retained in the weldment.
5.2.3-14 10j 0/2 Rev. 1
"The controls imposed on electroslag welding of ferritic steels are in accordance with the recommendations of Regulatory Guide 1.34, " Control of Electroslag Weld Properties," and provide assurance that welds fabricated by the process will have high integrity and will have a sufficient degree of toughness to furnish adequate safety margins during operating, testing, maintenance, and postulated accident conditions.
"The controls imposed upon components constructed of austenitic stainless steel used in the reactor coolant pressure boundary conform to the recommendations of Regulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal," and Regulatory Guide 1.34, " Control of Electroslag Weld Properties." Material selection, fabrication practices, examination procedures, and protection procedures performed in accordance with these recommendations provide reasonable assurance that the austenitic stainless steel in the reactor coolant pressure boundary will be in a metallurgical condition which precludes susceptibility to stress corrosion cracking during service. Conformance with these Regulatory Guides constitutes an acceptable basis for meeting in part the requirements of General Design Criteria 1 and 14."
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Plants."
2.
10 CFR Part 50, Appendix G, " Fracture Toughness Requirements."
3.
ASME Boiler and Pressure Vessel Code,Section II, Parts A, B, and C,Section III, and Section IX, American Society of Mechanical Engineers.
4.
ASTM, A-262, Practice E, " Copper-Copper Sulfate-Sulfuric Acid Test for Detecting Susceptibility to Intergranular Attack in Stainless Steels," Annual Book of ASTM Standards, American Society of Testing and Materials.
5.
ASTM E 23, " Notched Bar Impact Testing of Metallic Materials," Annual Book of ASTM l
Standards, American Society of Testing and Materials.
6.
ASTM E-208, " Standard Method for Conducting Dropweight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels,' Annual Book of ASTM Standards, American Society for Testing and Materials.
7.
Regulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal."
8.
Regulatory Guide 1.34, " Control of Electroslag Weld Properties."
9.
Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steel."
10.
Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of F luid Systems and Associated Compnnents of Water Cooled Nuclear Power Plants."
147 073 5.2.3-15 Rev. 1
11.
Regulatory Guide 1.43, " Control of Stainless Steel Weld Cladding of Low-Alloy Steel."
12.
Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel."
13.
Regulatory Guide 1.50, " Control of Preheat Temperature for Welding of Low-Alloy Steel."
14.
Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors."
15.
Regulatory Guide 1. 71, " Welder Qualification for Areas of Limited Accessibility."
16.
Regulatory Guide 1.85, " Code Case Acceptability ASME Section III Materials."
17.
Branch Technical Position MTEB 5-7 " Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," appended.
18.
WCAP-8577, "The Application of Preheat Temperatures Af ter-Welding Pressure Vessel Steels" Westinghouse Electric Corporation (Sept. 1975, Approved by Letter J.
F.
Stolz to C.
Eicheldinger, June 18, 1976).
9 9
147 074 5.2.3-16 Rev. 1
BRANCH TECHNICAL POSITION MIEB 5-7*
MATERIAL SELECTION AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE BOUNDARY PIPING I.
INTRODUCTION Small, hairline cracks in austenitic stainless steel piping in boiling water reactor (BWR) facilities were observed as early as 1965.
In each case, it was believed that the situation had been corrected or substantially reduced by better control of welding, contaminants and/or design modifications. In September, 1974, when the first of a series of cracks in the piping of the more modern BWRs was found at Dresden Unit No.
2.,
the then Atomic Energy Commission (AEC) initiated an intensive investigation to evaluate the cause, extent, and safety implications of the observed cracking. In January 1975, a special Pipe Cracking Study Group was formed to coordinate and accelerate the staff's ccqtinuing investigations of the occurrences of pipe cracking. This group included representatives of the Nuclear Regulatory Commission (NRC) and their consultants. In October 1975, the Study Group issued a report, NUREG-75/067 " Technical Report, Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants."
During the same general time span, the General Electric Company (GE) conducted an independent evaluation of the cracking occurrences and submitted their findings and recommendations to the NRC.
This paper sets forth the NRC technical position based on the information available at this time.
Plant operating history indicates that Types 304 and 316 austenitic stainless steel,
.ing in the reactor coolant pressure boundary of boiling water reactors are susceptible to stress corrosion cracking. Studies have shown that such cracking is caused by a combination of the presence of significant amounts of oxygen in the coolant, high stresses, and some sensitiza-tion of metal adjacent to welds.
Such cracks have occurred in the heat affected zones adjacent to welds but are not expected to occur outside these areas, provided that the pipe material is properly annealed.
Pipe runs containing stagnant or low velocity fluids have been observed to be more susceptible to stress corrosion cracking than pipes containing a continuously flowing fluid during plant operation. Historically, these cracks have been identified either by volumetric examination, by leak detection systems, or by visual inspection. Because of the inherent high material toughness of austenitic stainless steel piping, stress corrosion cracking is unlikely to cause a rapidly propagating failure resulting in a loss of coolant accident.
Although the probability is extremely low that these stress corrosion cracks will propagate far enough to create a significant safety hazard to the pubilc, the presence of such cracks is undesirable. Steps should therefore be taken to minimize stress corrosion cracking in BWR piping systems to eliminate this condition and to improve overall plant reliability.
- This Branch Technical Position is identical, except for typographical corrections, to the staff position given in NUREG-0313.
(July 1977) 1 [r /7
() 7 I)
U/i I
5.2.3-17 Rev. I
It is the purpose of this position to set forth acceptable methods to reduce the stress corrosion cracking susceptibility of BWR piping and thereby also provide an increased level of reactor coolant pressure boundary integrity. Recognizing that the most straight-forward and desirable approach or methods may not be practicable, or even possible, for all plants, the bases for varying degrees of conformance to our guidelines are provided.
Augmented inservice inspection and leak detection requirements are established for plants that have not fully implemented the provisions contained in Part II of this document.
II.
SUMMARY
OF ACCEPTABLE METHODS TO MINIMIZE CRACK SUSCEPTIBILITY The material selection and processing guidelines listed below identify altern6tive acceptable methods to minimize susceptibility to stress corrosion in BWR pressure boundary piping.
It is expected that adoption of these practices will result in a high degree of protection against stress corrosion cracking.
1.
Corrosion Resistant Materials All pipe and fitting material including weld metal should be of a type and grade that has been shown to be highly resistant to oxygen-assisted stress corrosion in the as-installed condition. Unstabilized wrought austenitic stainless steel with >0.035%
carbon does not meet this requirement unless all such piping including welds is in the solution annealed condition. The acceptability of alternative materials, processes, or other methods to provide an adequate degree of corrosion resistance will be made on a case-by-case basis.
2.
Corrosion Resistant "Saf e Ends" All unstabilized wrought austenitic stainless steel piping with carbon contents
>0.035% should be in the solution annealed condition. If welds joining these materials are not solution annealed, they should be made between cast (or weld overlaid) austenitic stainless steel surfaces (5% minimum ferrite) or other materials having high resistance to oxygen-assisted stress corrosion. The joint design must be such that any unstabilized wrought austenitic stainless steel containing >0.035% carbon, which may become sensitized as a result of the welding process, is not exposed to the reactor coolant.
3.
Other proposed methods to provide protection against stress corrosion cracking will be reviewed on a case-by-case basis.
Regulatory Guide 1.44 " Control of the Use of Sensitized Stainless Steel," dated May 1973 will be revited to provide additional guidance on acceptable practices.
III. INSERVICE INSPECTION AND LEAK DETECTION REQUIREMENTS FOR BWRs WITH VARYING CONFORMANCE TO MATERIAL SELECTION AND PROCESSING GUIDELINES 1.
For plants where all ASME Code Class I reactor coolant prassure boundary piping subject to inservice inspections under Section XI meets the guidelines stated in Part II, no augmented inservice inspection or leak detection requirements are necessary.
5.2.3-18 14 7
().,6
/
Rev. 1
2.
Piping in all other plants is subject to additional inservice inspection and leak detection requirements, as described below.
The degree of inspection of such piping depends on whether the specific piping runs are conforming or non-conforming, and on whether the specific piping runs are classified as " Serv'ce Sensitive." " Service Sensitive" lines are defined as those that have experienced cracking in service, or that are considered to be particularly susceptible to cracking because of high stress, or because they contain relatively stagnant, intermittent, or low flow coolant.
Examples of piping runs considered to be service sensitive include, (but are not limited to): core spray lines, recirculating by pass lines (or " stub tubes" on plants that have removed the by pass lines), CRD hydraulic return lines, isolation condenser lines, and shut down heat exchanger lines.
A.
For non-conforming lines that are not service sensitive:
(1) Inservice inspection fo the non-conforming lines should be conducted in accordance with the schedule specified in ASMF Code,Section XI - Subsection IWB, as required by the applicable examination Categories B-F and B-J, with the exception that the required examination should be completed in no more than 80 months (two thirds of the time prescribed in the schedule in the ASME Boiler and Pressure Vessel Code Section XI).
If examinations conducted during the first 80 month period reveal no incidence of stress corrosion cracking, the examination schedule thereafter can revert to the schedule prescribed in Section XI cf the ASME Boiler and Pressure Vessel Code.
The piping areas subject to examination, the method of examination, the allowable indication standards and examination procedures should comply with the requirements of the Edition and Addenda of the ASME Code,Section XI identified as applicable by 10 CFR Part 50, Section 50.55a, Paragraph (g), " Code and Standards."
(2) Tha reactor coolant leakage detection system should be operated under the following Technical Specfication requirements in order to enhance the discovery of unidentified leakage that may include through-wall cracks in austenitic stainless steel piping:
a.
The source of reactor coolant leakage should be identifiable to the extent practical, using leakage detection and collection systems that meet the position described in Section C, Regulatory Position of Regula-tory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," or an acceptable equivalent system.
b.
Plant shutdown should be initiated for inspection and corrective action when the leakage nystem indicates, within a period of four hours or less, an increase in the rate of unidentif':d leakage in excess of two gallons per minute, or when the total inidentified leakage attains a rate of five gallons per minute, whichever occurs first
}7 {}((
5.2.3-19 Rev. 1
c.
Unidentifed leakage should include all leakage other than:
1.
Leakage into closed systems, such as pump seal or valve packing leakage that is ca.otured, metered, and conducted to a sump or collecting tank, 2.
Leakage into the containment atmosphere from sources that are specifically located and known either not to interfere with the operation of the unidentified leakage detection systein, or not to be f rom a through-wall crack in the piping within the reactor coolant pressure boundary.
B.
For non-conforming lines that are service sensitive:
(1) The leakage detection requirements described in III.A above, should be implemented.
(2) The welds and adjoining areas of bypass piping of the discharge valves in the main recirculation loops, and of the austenitic stainless steel reactor core spray piping up to and including the second isolation valve should be examined at each reactor refueling outage or at other scheduled or un-scheduled plant shutdowns. Successive examinations need not be closer than six months, if shutdowns occur more frequently than six months. This require-ment applies to all bypass lines whether the 4-inch valve is kept open or closed during operation.
In the event these examinations find the piping free of unacceptable indica-tions for three successive inspecticns, the examination may be extended to each 36 month period (plus or minus by as much as 12 months) coinciding with a refueling outage. In these cases, the successive examination may be limited to cne bypass pipe run, and one reactor core spray piping run.
(3) The welds and adjoining areas of other service sensitive piping should be examined on a sampling basis.
For example, if a system consists of several branch runs with essentially symmetric piping configurations that perform similar system functions, an acceptable inspection program should include at least one, but not less than 25%, of the similar branch runs.
The frequency of such examinations should be as dascribed in 2 above.
If unacce5 table flaw indications are detected in any branch run, the remaining oranch runs among the group should be examined.
In the event the examinations reveal no unacceptable indications within three successive inspections, the examination schedule may revert to the ASME Boiler and Pressure Vessel Code,Section XI, " Inservice Inspection of Nuclear Power Plant Components" with the exception that the required examination should be completed during each 80 month period (two-thirds the time prescribed in the schedule in the ASME Code Section XI).
5.2.3-20 f f i O t'B-7 Rev. 1
(4) The method of examination, the allowable indication standards and examina-tion procedures should comply with the requirments of the Edition and Addenda of the ASME Code,Section XI identified as applicable by 10 CFR Part 50, Section 50.55a, Paragraph (g), " Codes and Standards."
IV.
IMPLEMENTATION OF MATERIAL SELECTION AND PROCESSING GUIDELINES 1.
For plants that apply for a construction permit after the issue date of this document *,
als ASME Code Clacs I reactor coolant pressure boundary lines should conf orm to the guidelines stated in Part II.**
2.
For plants under review, but for which a construction permit has not yet been issued, all service sensitive lines should conform to the gJidelines stated in Part II.
Other ASME Code Class I reactor coolant pressure boundary lines should conform to Part II to the extent practicable.
3.
For plants that have been issued a construction permit, ASME Code Class I reactor coolant pressure boundary lines should conform to the guidelines stated in Part II to the extent practicable.
4.
For plants that have been issued an operating license, service sensitive lines should be modified to conform to the guidelines stated in Part II, to the extent practicable.
Lines in which cracking is experienced should be replaced with piping that conforms to the guidelines stated in Part II.
V.
GENERAL RECOMMENDATIONS The measures outlined in Part II of this document provide for positive actions that are consistent with the current technology. The implementation of these actions should markedly reduce the susceptibility to stress corrosion cracking in BWRs.
It is recognized that additional techniques are available to limit the corrosion potential of BWR coolant pressure boundary materials and improve the overall system integrity.
These include plant design and operational considerations to reduce system exposure to potentially agressive environment, improve material fabrication and welding techniques and provisions for volumetric inspection capability in the design of weld joints.
Specifically, consideration should be given to:
1.
Minimizing the total extent of the coolant pressure boundary with special emphasis on stagnant or low flow lines.
2.
Reducing the oxygen content of the primary coolant.
- Dated July 19//
- After revision, Regulatory Guide 1.44 may be used as guidance for acceptable materials, process or other methods.
5.2.3-21 1
I Rev. I