ML110040863

Bellefonte Units 3 & 4 Cola (Final Safety Analysis Report), Rev. 3 - FSAR Chapter 05 Reactor Coolant System and Connected Systems - Sections 05.01 - 05.04

ML110040863
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Site:	Bellefonte
Issue date:	12/22/2010
From:	Arent G Tennessee Valley Authority
To:	Document Control Desk, Office of New Reactors
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BELLEFONTE.P02.NP, BELLEFONTE.P02.NP.3, TENNVALLEY, TENNVALLEY.SUBMISSION.6, +reviewedmmc1
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CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS TABLE OF CONTENTS Section Title Page 5.1

SUMMARY

DESCRIPTION................................................................. 5.1-1 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY .... 5.2-1 5.2.1.1 Compliance with 10 CFR 50.55a ............................................. 5.2-1 5.2.3.2.1 Chemistry of Reactor Coolant.................................................. 5.2-1 5.2.4 INSERVICE INSPECTION AND TESTING OF CLASS 1 COMPONENTS ............................................................................. 5.2-2 5.2.4.1 System Boundary Subject to Inspection .................................. 5.2-2 5.2.4.3 Examination Techniques and Procedures ............................... 5.2-4 5.2.4.3.1 Examination Methods .............................................................. 5.2-4 5.2.4.3.2 Qualification of Personnel and Examination Systems for Ultrasonic Examination ............................................................ 5.2-6 5.2.4.4 Inspection Intervals.................................................................. 5.2-6 5.2.4.5 Examination Categories and Requirements ............................ 5.2-6 5.2.4.6 Evaluation of Examination Results .......................................... 5.2-7 5.2.4.8 Relief Requests ....................................................................... 5.2-7 5.2.4.9 Preservice Inspection of Class 1 Components ........................ 5.2-7 5.2.4.10 Program Implementation ......................................................... 5.2-7 5.2.6 COMBINED LICENSE INFORMATION ITEMS............................. 5.2-8 5.2.6.1 ASME Code and Addenda....................................................... 5.2-8 5.2.6.2 Plant-Specific Inspection Program........................................... 5.2-8 5.

2.7 REFERENCES

.............................................................................. 5.2-8 5.3 REACTOR VESSEL ............................................................................ 5.3-1 5.3.2.6 Material Surveillance ............................................................... 5.3-1 5.3.2.6.3 Report of Test Results ............................................................. 5.3-3 5.3.3.2 Operating Procedures.............................................................. 5.3-3 5.3.6 COMBINED LICENSE INFORMATION......................................... 5.3-3 5.3.6.1 Pressure-Temperature Limit Curves........................................ 5.3-3 5.3.6.2 Reactor Vessel Materials Surveillance Program ..................... 5.3-4 5.3.6.4 Reactor Vessel Materials Properties Verification..................... 5.3-4 5.3.6.4.1 Reactor Vessel Materials Properties Verification..................... 5.3-4 5.4 COMPONENT AND SUBSYSTEM DESIGN....................................... 5.4-1 5.4.2.5 Steam Generator Inservice Inspection .................................... 5.4-1 5.4.15 COMBINED LICENSE INFORMATION ITEMS............................. 5.4-1 5.4.16 REFERENCES .............................................................................. 5.4-2 5-i Revision 3

CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIPTION This section of the referenced DCD is incorporated by reference with no departures or supplements.

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5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

5.2.1.1 Compliance with 10 CFR 50.55a Add the following text after the second sentence of the second paragraph of DCD Subsection 5.2.1.1.

COL 5.2-1 If a later Code edition/addenda than the Design Certification Code edition/

addenda is used by the material and/or component supplier, then a code reconciliation to determine acceptability is performed as required by the ASME Code,Section III, NCA-1140. The later Code edition/addenda must be authorized in 10 CFR 50.55a or in a specific authorization as provided in 50.55a(a)(3). Code Cases to be used in design and construction are identified in the DCD; additional Code Cases for design and construction beyond those for the design certification are not required.

Inservice inspection of the reactor coolant pressure boundary is conducted in accordance with the applicable edition and addenda of the ASME Boiler and Pressure Vessel Code Section XI, as described in Subsection 5.2.4. Inservice testing of the reactor coolant pressure boundary components is in accordance with the edition and addenda of the ASME OM Code as discussed in Subsection 3.9.6 for pumps and valves, and as discussed in Subsection 3.9.3.4.4 for dynamic restraints.

5.2.3.2.1 Chemistry of Reactor Coolant Add the following text to the end of DCD Subsection 5.2.3.2.1.

SUP 5.2-1 The water chemistry program is based on industry guidelines as described in EPRI TR-1002884, Pressurized Water Reactor Primary Water Chemistry (Reference 201). The program includes periodic monitoring and control of chemical additives and reactor coolant impurities listed in DCD Table 5.2-2.

Detailed procedures implement the program requirements for sampling and analysis frequencies, and corrective actions for control of reactor water chemistry.

The frequency of sampling water chemistry varies (e.g. continuous, daily, weekly, or as needed) based on plant operating conditions and the EPRI water chemistry guidelines. Whenever corrective actions are taken to address an abnormal chemistry condition, increased sampling is utilized to verify the effectiveness of 5.2-1 Revision 3

these actions. When measured water chemistry parameters are outside the specified range, corrective actions are taken to bring the parameter back within the acceptable range and within the time period specified in the EPRI water chemistry guidelines. Following corrective actions, additional samples are taken and analyzed to verify that the corrective actions were effective in returning the concentrations of contaminants to within the specified range.

Chemistry procedures will provide guidance for the sampling and monitoring of primary coolant properties.

5.2.4 INSERVICE INSPECTION AND TESTING OF CLASS 1 COMPONENTS Add the following after the first paragraph in DCD Subsection 5.2.4:

COL 5.2-2 The initial inservice inspection program incorporates the latest edition and addenda of the ASME Boiler and Pressure Vessel Code approved in 10 CFR 50.55a(b) on the date 12 months before initial fuel load. Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) 12 months before the start of the 120-month inspection interval (or the optional ASME Code cases listed in NRC Regulatory Guide 1.147, that are incorporated by reference in 10 CFR 50.55a(b)), subject to the limitations and modifications listed in 10 CFR 50.55a(b).

5.2.4.1 System Boundary Subject to Inspection Add the following at the end of DCD Subsection 5.2.4.1:

COL 5.2-2 The Class 1 system boundary for both preservice and inservice inspection programs and the system pressure test program includes those items within the Class 1 and Quality Group A (Equipment Class A per DCD Subsection 3.2.2 and DCD Table 3.2-3) boundary. Based on 10 CFR Part 50 and Regulatory Guide 1.26, the Class 1 boundary includes the following:

• Reactor pressure vessel;

• Portions of the Reactor System (RXS);

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• Portions of the Chemical and Volume Control System (CVS);

• Portions of the Incore Instrumentation System (IIS);

• Portions of the Passive Core Cooling System (PXS);

• Portions of the Reactor Coolant System (RCS); and

• Portions of the Normal Residual Heat Removal System (RNS).

Those portions of the above systems within the Class 1 boundary are those items that are part of the reactor coolant pressure boundary as defined in Section 5.2.

Exclusions Portions of the systems within the reactor coolant pressure boundary (RCPB), as defined above, that are excluded from the Class 1 boundary in accordance with 10 CFR Part 50, Section 50.55a, are as follows:

• Those components where, in the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only; or

• Components that are or can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other open). Each open valve is capable of automatic actuation and, assuming the other valve is open, its closure time is such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only.

The description of portions of systems excluded from the RCPB does not address Class 1 components exempt from inservice examinations under ASME Code Section XI rules. The Class 1 components exempt from inservice examinations are defined by ASME Section XI, IWB-1220, except as modified by 10 CFR 50.55a.

The inservice inspection program is augmented for reactor vessel top head inspections by use of the ASME Code Case N-729-1, Alternative Examination Requirements for Pressurized-Water Reactor (PWR) Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, as modified by the conditions specified in 10 CFR 50.55a(g)(6)(ii)(D).

Boric acid corrosion control procedures require inspection of the reactor coolant pressure boundary subject to leakage that can cause boric acid corrosion of the reactor coolant pressure boundary materials. The procedures determine the 5.2-3 Revision 3

principal locations where leaks can cause degradation of the primary pressure boundary by boric acid corrosion. Potential paths of the leaking coolant are established. The boric acid corrosion control procedures also contain methods for conducting examinations and performing engineering evaluations to establish the impact on the reactor coolant pressure boundary when leakage is located.

The boric acid corrosion control procedures consist of:

1. Visual inspections of component surfaces that are potentially exposed to borated water leakage.

2. Discovery of leak path and removal of boric acid residue.

3. Assessment of the corrosion.

4. Follow-up inspection for adequacy of corrective actions, as appropriate.

Add the following text at the end of DCD Subsection 5.2.4.1:

SUP 5.2-2 The inservice inspection program, along with the boric acid corrosion control procedures, provides guidance for inspecting the integrity of bolting and threaded fasteners.

5.2.4.3 Examination Techniques and Procedures Add the following at the end of DCD Subsection 5.2.4.3:

5.2.4.3.1 Examination Methods Ultrasonic Examination of the Reactor Vessel COL 5.2-2 Ultrasonic examination for the RPV is conducted in accordance with the ASME Code,Section XI. The design of the RPV considered the requirements of the ASME Code Section XI with regard to performance of preservice inspection. For the required preservice examinations, the reactor vessel meets the acceptance standards of Section XI, IWB-3510. The RPV shell welds are designed for 100%

accessibility for both preservice and inservice inspection. RPV shell welds may be examined from the inside or outside diameter surfaces (or a combination of those techniques) using automated ultrasonic examination equipment. The RPV nozzle-to-shell welds are 100% accessible for preservice inspection but might have limited areas that may not be accessible from the outer surface for inservice 5.2-4 Revision 3

examination techniques. If accessibility is limited, an inservice inspection program relief request is prepared and submitted for review approval by the NRC.

Inner radius examinations are performed from the outside of the nozzle using several compound angle transducer wedges to obtain complete coverage of the required examination volume. Alternatively, nozzle inner radius examinations may be performed using enhanced visual techniques, as allowed by 10 CFR 50.55a(b)(2)(xxi).

Visual Examination Visual examination methods VT-1, VT-2 and VT-3 are conducted in accordance with ASME Section XI, IWA-2210. In addition, VT-2 examinations meet the requirements of IWA-5240.

Where direct visual VT-1 examinations are conducted without the use of mirrors or with other viewing aids, clearance is provided where feasible for the head and shoulders of a man within a working arm's length of the surface to be examined.

Surface Examination Magnetic particle and liquid penetrant examination techniques are performed in accordance with ASME Section XI, IWA-2221 and IWA-2222, respectively. Direct examination access for magnetic particle (MT) and liquid penetrant (PT) examination is the same as that required for direct visual (VT-1) examination (see Visual Examination), except that additional access is provided as necessary to enable physical contact with the item in order to perform the examination. Remote MT and PT generally are not appropriate as a standard examination process; however, boroscopes and mirrors can be used at close range to improve the angle of vision.

Volumetric Ultrasonic Direct Examination Volumetric ultrasonic direct examination is performed in accordance with ASME Section XI, IWA-2232, which references mandatory Appendix I.

Alternative Examination Techniques As provided by ASME Section XI, IWA-2240, alternative examination methods, a combination of methods, or newly developed techniques may be substituted for the methods specified for a given item in this section, provided that they are demonstrated to be equivalent or superior to the specified method. This provision allows for the use of newly developed examination methods, techniques, etc.,

which may result in improvements in examination reliability and reductions in personnel exposure. In accordance with 10 CFR 50.55a(b)(2)(xix), IWA-2240 as written in the 1997 Addenda of ASME Section XI must be used when applying these provisions.

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5.2.4.3.2 Qualification of Personnel and Examination Systems for Ultrasonic Examination Personnel performing examinations shall be qualified in accordance with ASME Section XI, Appendix VII. Ultrasonic examination systems shall be qualified in accordance with industry accepted programs for implementation of ASME Section XI, Appendix VIII. Qualification to ASME Section XI, Appendix VIII, is in compliance with the provisions of 10 CFR 50.55a.

5.2.4.4 Inspection Intervals Add the following after the second sentence of the first paragraph of DCD Subsection 5.2.4.4:

COL 5.2-2 Because 10 CFR 50.55a(g)(4) requires 120-month inspection intervals, Inspection Program B of IWB-2400 must be chosen. The inspection interval is divided into three periods. Period one comprises the first three years of the interval, period two comprises the next four years of the interval, and period three comprises the remaining three years of the inspection interval. Each period can be extended for up to one year to enable an inspection to coincide with a plant outage. The adjustment of period end dates shall not alter the rules and requirements for scheduling inspection intervals.

5.2.4.5 Examination Categories and Requirements Add the following after the first sentence of DCD Subsection 5.2.4.5:

COL 5.2-2 Class 1 piping supports will be examined in accordance with ASME Section XI, IWF-2500.

Preservice examinations required by design specification and preservice documentation are in accordance with ASME Section III, NB-5280. Components exempt from preservice examination are described in ASME Section III, NB-5283.

Add the following after the last sentence of DCD Subsection 5.2.4.5:

The preservice examination is performed once in accordance with ASME XI, IWB-2200, on all of the items selected for inservice examination, with the exception of the examinations specifically excluded by ASME Section XI from preservice requirements, such as VT-3 examination of valve body and pump 5.2-6 Revision 3

casing internal surfaces (B-L-2 and B-M-2 examination categories, respectively) and the visual VT-2 examinations for category B-P.

5.2.4.6 Evaluation of Examination Results Add the following at the end of DCD Subsection 5.2.4.6:

COL 5.2-2 Components containing flaws or relevant conditions and accepted for continued service in accordance with the requirements of IWB-3132.4 or IWB-3142.4 are subjected to successive period examinations in accordance with the requirements of IWB-2420. Examinations that reveal flaws or relevant conditions exceeding Table IWB-3410-1 acceptance standards are extended to include additional examinations in accordance with the requirements of IWB-2430.

COL 5.2-2 Add Subsections 5.2.4.8, 5.2.4.9, and 5.2.4.10 after the last paragraph of DCD Subsection 5.2.4.7:

5.2.4.8 Relief Requests The specific areas where the applicable ASME Code requirements cannot be met are identified after the initial examinations are performed. Should relief requests be required, they will be developed through the regulatory process and submitted to the NRC for approval in accordance with 10 CFR 50.55a(a)(3) or 50.55a(g)(5).

The relief requests include appropriate justifications and proposed alternative inspection methods.

5.2.4.9 Preservice Inspection of Class 1 Components Preservice examinations required by design specification and preservice documentation are in accordance with ASME Section III, NB-5281. Volumetric and surface examinations are performed as specified in ASME Section III, NB-5282. Components described in ASME Section III, NB-5283 are exempt from preservice examination.

5.2.4.10 Program Implementation The milestones for preservice and inservice inspection program implementation are identified in Table 13.4-201.

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5.2.6 COMBINED LICENSE INFORMATION ITEMS 5.2.6.1 ASME Code and Addenda COL 5.2-1 This COL Item is addressed in Subsection 5.2.1.1.

5.2.6.2 Plant-Specific Inspection Program COL 5.2-2 This COL Item is addressed in Subsections 5.2.4, 5.2.4.1, 5.2.4.3.1, 5.2.4.3.2, 5.2.4.4, 5.2.4.5, 5.2.4.6, 5.2.4.8, 5.2.4.9, and 5.2.4.10.

5.

2.7 REFERENCES

Add the following information at the end of DCD Subsection 5.2.7.

201. EPRI, Pressurized Water Reactor Primary Water Chemistry Guidelines, EPRI TR-1002884, Revision 5, October 2003.

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5.3 REACTOR VESSEL This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

5.3.2.6 Material Surveillance Add the following information between the first and second paragraphs of DCD Subsection 5.3.2.6.

COL 5.3-2 Surveillance test materials are prepared from the actual materials used in fabricating the beltline region of the reactor vessel. Records are maintained of the chemical analyses, fabrication history, mechanical properties and other essential variables pertinent to the fabrication process of the shell forging and weld metal from which the surveillance test materials are prepared. The test materials are processed so that they are representative of the material in the completed reactor vessel.

Three metallurgically different materials prepared from sections of reactor vessel shell forging are used for test specimens. These include base metal, weld metal and heat affected zone (HAZ) material.

Base metal test material is manufactured from a section of ring forging, either the intermediate shell course, the lower shell course, or the transition ring of the reactor pressure vessel. Selection is based on an evaluation of initial toughness (characterized by the reference temperature (RTNDT) and Upper Shelf Energy (USE)), and the predicted effect of chemical composition (nickel and residual copper) and neutron fluence on the toughness (RTNDT shift and decrease in USE) during reactor operation. The ring forging with the highest predicted adjusted RTNDT temperature (initial RTNDT plus RTNDT shift) or that with USE predicted to approach close to the minimum limit of 50 ft-lb at end-of-license (EOL) is selected as the surveillance base metal test material. The means for measuring initial toughness and for predicting irradiation induced toughness changes is consistent with applicable procedures in force at the time the material is being selected. The section of shell forging used for the base metal test block is adjacent to the test material used for fracture toughness tests.

Weld metal and HAZ test material is produced by welding together sections of the forgings from the beltline of the reactor vessel. The HAZ test material is manufactured from a section of the same shell course forging used for base metal test material. The sections of shell course forging used for weld metal and HAZ test material are adjacent to the test material used for fracture toughness tests.

The heat of wire or rod and lot of flux are from the same heat and lot used in making the beltline region welds. Welding parameters duplicate those used for the beltline region welds. The procedures for inspection of the reactor vessel welds are followed for the inspection of the welds in test materials. The surveillance weld 5.3-1 Revision 3

and HAZ material are heat-treated to metallurgical conditions which are representative of the final metallurgical conditions of similar materials in the completed reactor vessel.

Test Specimens are marked to identify the type of materials and the orientation with respect to the test materials. Drawings specify the identification system to be used and include plant identification, type of material, orientation of specimen and sequential number.

Baseline test specimens are provided for establishing the baseline (unirradiated) properties of the reactor vessel materials. The data from tests of these specimens provides the basis for determining the radiation induced property changes of the reactor vessel materials.

Drop weight test specimens of each of base metal, weld metal, and HAZ metal are provided for establishing the nil-ductility transition temperature (NDTT) of the unirradiated surveillance materials. These data form the basis for RTNDT determination from which subsequent radiation induced changes are determined.

Standard Charpy impact test specimens each of base metal (longitudinal (tangential) and transverse (axial)), weld metal, and HAZ material are provided for developing a Charpy impact energy transition curve from fully brittle to fully ductile behavior for defining specific index temperatures for these materials. These data, together with the drop weight NDTT, are used to establish and RTNDT for each material.

Tensile test specimens each of base metal (longitudinal (tangential) and transverse (axial)), weld metal, and HAZ metal are provided to permit a sufficient number of tests for accurately establishing the tensile properties for these materials at a minimum of three test temperatures (e.g., ambient, operating and one intermediate temperature) to define the strength of the material.

The above described test specimens are to be used for determining changes in the strength and toughness of the surveillance materials resulting from neutron irradiation. Sufficient Charpy impact, compact tension and tensile test specimens are provided for establishing the changes in the properties of the surveillance materials over the lifetime of the reactor vessel. The type, quantity, and storage conditions (e.g., surveillance capsules backfilled with inert gas) of test specimens meet or exceed the minimum requirements of ASTM E-185.

Reactor materials do not begin to be affected by neutron fluence until the reactor begins critical operation. Table 13.4-201 provides milestones for reactor vessel material surveillance program implementation.

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Add the following subsection after DCD Subsection 5.3.2.6.2.2.

5.3.2.6.3 Report of Test Results COL 5.3-2 A summary technical report for each capsule withdrawn with the test results is submitted, as specified in 10 CFR 50.4, within one year of the date of capsule withdrawal unless an extension is granted by the Director, Office of Nuclear Reactor Regulation.

The report includes the data required by ASTM E185-82, as specified in paragraph III.B.1 of 10 CFR Part 50, Appendix H, and includes the results of the fracture toughness tests conducted on the beltline materials in the irradiated and unirradiated conditions.

If the test results indicate a change in the Technical Specifications is required, either in the pressure-temperature limits or in the operating procedures required to meet the limits, the expected date for submittal of the revised Technical Specification is provided with the report.

Add the following subsection after DCD Subsection 5.3.3.1.

5.3.3.2 Operating Procedures SUP 5.3-1 Plant operating procedures are developed and maintained to prevent exceeding the pressure-temperature limits identified in reactor coolant system pressure and temperature limits report, as required by Technical Specification 5.6.6, during normal and abnormal operating conditions and system tests.

5.3.6 COMBINED LICENSE INFORMATION 5.3.6.1 Pressure-Temperature Limit Curves Replace the text in DCD Subsection 5.3.6.1 with the following.

COL 5.3-1 The pressure-temperature curves shown in DCD Figures 5.3-2 and 5.3-3 are generic curves for AP1000 reactor vessel design, and they are the limiting curves based on copper and nickel material composition. Plant-specific curves will be developed based on material composition of copper and nickel. Use of plant-specific curves will be addressed during procurement and fabrication of the reactor vessel. As noted in the bases to Technical Specification 3.4.14, use of plant-specific curves requires evaluation of the LTOP system. This includes an evaluation of the setpoint pressure for the RNS relief valve to determine if the 5.3-3 Revision 3

setpoint pressure needs to be changed based on the plant-specific pressure-temperature curves. The development of the plant-specific curves and evaluation of the setpoint pressure are required prior to fuel load.

5.3.6.2 Reactor Vessel Materials Surveillance Program COL 5.3-2 This COL Item is addressed in Subsections 5.3.2.6 and 5.3.2.6.3.

5.3.6.4 Reactor Vessel Materials Properties Verification Replace the text in DCD Subsection 5.3.6.4.1 with the following.

5.3.6.4.1 Reactor Vessel Materials Properties Verification COL 5.3-4 The verification of plant-specific belt line material properties consistent with the requirements in DCD Subsection 5.3.3.1 and DCD Tables 5.3-1 and 5.3-3 will be completed prior to fuel load. The verification will include a pressurized thermal shock evaluation based on as procured reactor vessel material data and the projected neutron fluence for the plant design objective of 60 years. This evaluation report will be submitted for NRC staff review.

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5.4 COMPONENT AND SUBSYSTEM DESIGN This section of the referenced DCD is incorporated by reference with the following departures and/or supplements.

5.4.2.5 Steam Generator Inservice Inspection Add the following information at the end of DCD Subsection 5.4.2.5.

COL 5.4-1 A steam generator tube surveillance program is implemented in accordance with the recommendations and guidance of Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines" (Reference 201). A program for periodic monitoring of degradation of steam generator internals is also implemented in accordance with NEI 97-06. Applicable Electric Power Research Institute (EPRI)

Steam Generator Management Program (SGMP) guidelines are followed as described in the NEI 97-06. The Programs are in compliance with applicable sections of ASME Section XI.

NEI 97-06 and the referenced EPRI SGMP guidelines provide recommendations concerning the inspection of tubes, which cover inspection equipment, baseline inspections, tube selection, sampling and frequency of inspection, methods of recording, required actions based on findings, and tube plugging. The minimum requirements for inservice inspection of steam generators, including plugging criteria, are established in Technical Specification 5.5.4.

The tube surveillance and degradation monitoring programs include provisions to maintain the compatibility of steam generator tubing with primary and secondary coolant to limit the steam generators susceptibility to corrosion. These provisions are in accordance with NEI 97-06.

5.4.15 COMBINED LICENSE INFORMATION ITEMS COL 5.4-1 This COL Item is addressed in Subsection 5.4.2.5.

5.4-1 Revision 3

5.4.16 REFERENCES Insert the following information at the end of DCD Subsection 5.4.16.

201. Nuclear Energy Institute, Steam Generator Program Guidelines, NEI 97-06, Revision 2, May 2005.

5.4-2 Revision 3