Regulatory Guide 1.162

(Draft Was Issued as DG-1027) Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels

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Issue date:	02/29/1996
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RG-1.162
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RCso U.S. NUCLEAR REGULATORY COMMISSION February 1996 REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.162 (Draft was Issued as DG-1027)

FORMAT AND CONTENT OF REPORT

FOR THERMAL ANNEAUNG OF REACTOR PRESSURE VESSELS

A. INTRODUCTION

pleteness of the information provided, would assist the NRC staff in locating specific information, and would The thermal annealing rule, § 50.66, "Require aid in shortening the time needed for the review ments for Thermal Annealing of the Reactor Pressure process.

Vessel," of 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facility," provides that: This regulatory guide also describes alternative methods that are acceptable to the NRC for determin For those light water nuclear power reactors ing the recovery of fracture toughness after the thermal where neutron radiation has reduced the frac annealing and for estimating the degree of post ture toughness of the reactor vessel materials, annealing reembrittlement expected during subse a thermal annealing may be applied to the quent plant operations; 10 CFR 50.66 requires these to reactor vessel to recover the fracture tough be reported.

ness of the material. The use of a thermal an This regulatory guide contains guidance on man nealing treatment is subject to the require datory information collections that are contained as re ments in this section. A report describing the quirements in 10 CFR Part 50 and that are subject to licensee's plan for conducting the thermal an the Paperwork Reduction Act of 1980 (44 U.S.C.

nealing must be submitted in accordance with 3501 et seq.). These requirements were approved by

§ 50.4 at least three years prior to the date at the Office of Management and Budget, approval num which the limiting fracture toughness criteria ber 3150-0011.

in § 50.61 or Appendix G to Part 50 would be exceeded. The public reporting burden for this collection of information Is estimated to be an average of 6000

This regulatory guide describes a format and con hours per respondent, including the time for reviewing tent acceptable to the NRC staff for the Thermal An instructions, searching existing data sources, gathering nealing Report to be submitted to the NRC for describ and maintaining the data needed, and completing and ing the licensee's plan for thermal annealing a reactor reviewing the collection of information. Send com vessel. This guide also describes the Thermal Anneal ments regarding this burden estimate or any other ing Results Report that is required by 10 CFR 50.66 to aspect of this collection of information, including sug be submitted after the thermal annealing. Use of this gestions for reducing the burden, to the Information format by the applicant would help ensure the corn- and Records Management Branch (T-6 F 33), U.S.

USNRC REGULATORY GUIDES Written comments may be submitted to the Rules Review and Directives Branch, DFIPS, ADM, U.S. Nuclear Regulatory Commission, Washing Regulatory Guides are issued to describe and make avallable to the public ton, DC 20658-0001.

such Information as methods acceptable to the NRC staff for Implement ing specific parts of the Commission's regulations, techniques used by The guides are issued in the following ten broad divisions:

the staff In evaluating specific problems or postulated accidents, and 1. Power Reactors 6. Products data needed by the NRC staff In its review of applications for permits and 2. Research and Test Reactors 7. Transportation licenses. Regulatory guides are not substitutes for regulations. and com 3. Fuels and Materials Facilities 8. Occupational Health plianoe with them Is not required. Methods and solutions different from 4. Environmental and Siting 9. Antitrust and Financial Review those set out In the guides will be acceptable If they provide a basis for the 5. Materials and Plant Protection 10. General findings requisite to the Issuance or continuance of a permit or license by the Commission. Single copies of regulatory guides may be obtained free of charge by writ the Office of Administration, Attention: Distribution and Services tion, U.S. Nuclear Regulatory Commission. Washington, OC

This guide was Issued after consideration of comments received from the 20655-0001. or by fax at (301)415-2260.

public. Comments and suggestions for Improvements In these guides are Issued guides may also be purchased from the National Technical Infor encouraged at all times, and guides will be revised, as appropriate, to mation Service on a standing order basis. Details on this service may be acconmmodate comments and to reflect new Information or experience. obtained by writing NTIS, 6285 Port Royal Road, Springfield, VA 22161.

Nuclear Regulatory Commission, Washington, DC annealing" at a temperature of 650°F using the reactor

20555-000 1; and to the Desk Officer, Office of Infor coolant pumps as the heat source. In addition, at least mation and Regulatory Affairs, NEOB-10202 12 Russian-designed VVER-440 PWRs, which operate

(3150-0011), Office of Management and Budget, at conditions similar to U.S. PWRs, have been an Washington, DC 20503. nealed in Russia and Eastern Europe at temperatures of approximately 850*F, using dry air and radiant Is.

B. DISCUSSION

heaters as the heat source. Details of the thermal an nealing of the Novovoronezh Unit 3 have been re BACKGROUND ported (Ref. 3) by a U.S. delegation that witnessed the operation.

Criterion 31 of Appendix A, "General Design Cri teria for Nuclear Power Plants," to 10 CFR Part 50 CURRENT STATE OF KNOWLEDGE ON

requires that: THERMAL ANNEALING

The reactor shall be designed with sufficient A significant amount of information has been re margin to assure that when stressed under op ported in the literature on thermal annealing and on erating, maintenance, testing, and postulated the effects of thermal annealing variables (e.g., tem accident conditions, (1) the boundary behave perature, time, materials chemistry, fluence levels), on in a nonbrittle manner and (2) the probability the recovery of toughness properties. Server (Ref. 4)

of rapidly propagating fracture is minimized. summarized the state of knowledge, as of 1985, for A major concern in this regard is that the material in-place thermal annealing of commercial reactor pres properties of reactor vessels degrade progressively sure vessels. He reviewed data on annealing recovery when exposed to neutron radiation during service, re and reirradiation effects for high-copper welds and sulting in a loss in fracture toughness and ductility. To concluded that significant recovery occurs for anneal maintain adequate toughness and preclude nonductile ing at 850*F for both the transition temperature shift failure in the vessel, a number of mitigating actions are (ARTNDT) and reduction in Charpy upper-shelf ener taken during the operating life of a reactor: periodic gy. He also reviewed engineering studies and con changes are made in the pressure-temperature (P-T) cluded that annealing of U.S. reactors at 850*F is fea limits required to preclude nonductile fracture of the sible using existing commercial heat treating methods, materials during startup and cooldown, limitations are but that plant-specific engineering problems would placed on the reduction of Charpy upper-shelf energy need to be resolved. Server (Ref. 4) also performed a to maintain an adequate margin of safety against duc thermal and structural analysis for a typical PWR vessel K1 tile fracture, and additional restrictions are placed on annealed at 850 0 F, which predicted that vessel dimen toughness properties by screening criteria imposed to sional stability would be maintained and that post avoid vessel failure from pressurized thermal shock. anneal residual stresses would not be significant. How ever, Server's results indicated that excessive bending If neutron radiation embritdement becomes so se of the attached piping from differential thermal expan vere that the required margins cannot be maintained, sion of the vessel could be a problem that required

10 CFR 50.61 and Appendix G to 10 CFR Part 50 careful temperature control.

permit the application of a thermal annealing treat ment to recover the toughness properties of the vessel Mager and others (Refs. 5 and 6) reported on re materials, which would avoid premature retirement of search to determine the extent of fracture toughness the reactor pressure vessel. Thermal annealing, the recovery as a function of annealing time and tempera heating of the reactor vessel beltline to a temperature ture for materials that are sensitive to neutron well above the operating temperature of the reactor for embrittlement. They concluded that excellent recov an extended period of time sufficient to remove the ery of all properties could be achieved by annealing at microstructural changes caused by radiation, is the 850*F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, and that the reembrittlement af only known method for restoring toughness properties ter annealing would follow the same trend as the pre to materials degraded by neutron radiation. The re annealing embrittlement rate. These reports also de quirements for conducting thermal annealing and re scribe a thermal annealing procedure developed for starting the plant after annealing are set forth in field application.

10 CFR 50.66. Additional research that includes data on Although thermal annealing has not yet been irradiation-anneal-reirradiation property trends for applied to a U.S. commercial power reactor, it has reactor pressure vessel welds has been reported by been successfully applied to other reactors. Two reac Hawthorne and Hiser (Ref. 7).

tor vessels that have been successfully annealed are the More recently*, Eason et al. performed analyses of Army's SM-1A in 1967 (Ref. 1), and the BR-3 in existing data on annealing of irradiated pressure vessel Mol, Belgium, in 1984 (Ref. 2). Both of these reactors steels using both mechanistic and statistical consider operated at temperatures low enough to permit "wet ations. Eason et al. also developed improved correla-

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tion models for estimating Charpy upper-shelf energy The thermal annealing rule, 10 CFR 50.66, per and transition temperature after radiation and anneal mits the thermal annealing of reactor vessels to restore ing. This work is reported in NUREG/CR-6327 (Ref. fracture toughness of the reactor vessel material that

8), and it provides the basis for equations for estimat was reduced by neutron radiation. Section 50.66(a)

>* ing recovery of fracture toughness following annealing requires that, prior to initiation of thermal annealing, a (Section 3.1.3 of this regulatory guide). Thermal Annealing Report be submitted to describe the licensee's plan for conducting the thermal anneal.

General guidance for inservice annealing may be The report must be submitted at least 3 years prior to found in ASTM Standard E 509-86 (Ref. 9). ASTM the date at which the limiting fracture toughness crite Standard E 509-86 contains general procedures for ria in 10 CFR 50.61 or Appendix G to Part 50 would conducting an in-service thermal anneal of a reactor be exceeded. This 3-year period is specified to provide vessel and for demonstrating the effectiveness and de the NRC staff with sufficient time to review the thermal gree of recovery. ASTM Standard E 509-86 also pro annealing program. Within 3 years of submittal of a vides direction for a post-anneal vessel radiation sur licensee's Thermal Annealing Report and at least 30

veillance program. days prior to the start of the thermal annealing,

10 CFR 50.66(a) requires the NRC staff to review the Thermal Annealing Report and place the results of its CURRENT REGULATORY REQUIREMENTS evaluation in its Public Document Room. In order to FOR FRACTURE TOUGHNESS OF VESSELS provide for public participation in the regulatory proc ess, 10 CFR 50.66(0 requires that the NRC hold a Fracture toughness requirements for light-water public meeting a minimum of 30 days before the li cooled reactor pressure vessels are addressed in sever censee starts to thermal anneal the reactor vessel. The al regulations. Appendix G, "Fracture Toughness Re licensee may. begin thermal annealing after the NRC

quirements," to 10 CFR Part 50 provides the fracture has placed the results of its evaluation of the Thermal toughness requirements for vessels during normal op Annealing Report in the Public Document Room and eration and anticipated accident conditions. Appendix after the public meeting is held.

G also permits the use of thermal annealing to restore fracture toughness degraded by neutron radiation The Thermal Annealing Report is required by when embrittlement degrades mechanical properties to 10 CFR 50.66(b) to include (1) a Thermal Annealing such an extent that adequate margins of safety cannot Operating Plan, (2) a Requalification Inspection and be demonstrated during operation. Appendix H, Test Program, (3) a Fracture Toughness Recovery and

"Reactor Vessel Material Surveillance Program Re "Reembrittlement Trend Assurance Program, and quirements," to 10 CFR Part 50 requires surveillance (4) the Identification of Unreviewed Safety Questions programs to monitor irradiation embrittlement of reac and Technical Specification Changes.

tor vessel beltline materials. The pressurized thermal The rule also provides three methods for deter shock (PTS) rule, 10 CFR 50.61, establishes screening mining the percent recovery. When surveillance speci-.

criteria for embrittlement beyond which the plant may mens are available from a credible surveillance pro not operate without further justification.

gram, the percent recovery for the reactor vessel is The application of these regulations in the late required to be determined by using a test program that

1980s and early 1990s demonstrated a need for clarifi applies the actual annealing conditions to the irra cation and improved guidance. The NRC review of the diated surveillance specimens. If surveillance speci reactor pressure vessel integrity of the Yankee Nuclear mens are not available, the applicant may elect to de Power Station highlighted the need for such changes termine the percent recovery by testing materials and resulted in a detailed plan described in removed from the reactor vessel beltline region. The SECY-91-333 (Ref. 10) and SECY-92-283 third method permits uses of a generic computational (Ref. 11). method if adequate justification is provided. Use of the procedure described in this regulatory guide in Sec To implement this plan, a proposed rule to amend tion 3.1.3, "Computational Methods," is considered the regulations was issued on October 4, 1994 (59 FR appropriate justification for this application.

50513). This rule was issued in final form on Decem ber 19, 1995 (60 FR 65456). The rule includes a new Upon completion of the thermal annealing and the

§ 50.66, "Requirements for Thermal Annealing of the associated tests and analyses, the applicant must con firm in writing to the Director of the Office of Nuclear Reactor Pressure Vessel," which sets forth the NRC's requirements for annealing of reactor pressure vessels. Reactor Regulation (NRR) that the thermal anneal was Also, changes were made to both Appendix G and the performed in accordance with the Thermal Annealing PTS rule to reference the thermal annealing require Operating Plan and the Requalification Inspection and ments in § 50.66, as an option for reducing embrittle Test Program. Within 15 days of the licensee's written ment when the toughness requirements of those rules confirmation that the thermal annealing was com cannot otherwise be met. pleted in accordance with the Thermal Annealing

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Plan, and prior to restart, the NRC will (1) briefly doc quirements for partial annealing are analogous to situa ument whether the thermal annealing was performed tions in which the thermal annealing was completed;

in compliance with the licensee's Thermal Annealing that is, when the partial annealing was otherwise per Operating Plan and the Requalification Inspection and formed in compliance with the Thermal Annealing Op Test Program, placing the documentation in the NRC erating Plan and relevant portions of the Requalifica Public Document Room, and (2) hold a public meeting tion Inspection and Test Program, the licensee submits K

to permit the licensee to explain the results of the reac written confirmation of such compliance and may re tor vessel annealing to the NRC and the public, allow start following, among other things, holding a public the NRC to discuss its inspection of the reactor vessel meeting on the annealing. By contrast, if the partial annealing, and provide an opportunity for the public to annealing was not performed in accordance with the comment to the NRC on the thermal annealing. The Thermal Annealing Operating Plan and relevant licensee may restart its reactor after the meeting has portions of the Requalification Inspection and Test been completed, unless the NRC orders otherwise. Program, the licensee is required by 10 CFR

Within 45 days of the licensee's written confirmation 50.66(c) (3) (iii) to submit a summary of lack of com that the thermal annealing was completed in accor pliance, submit a justification for subsequent opera dance with the Thermal Annealing Operating Plan and tions, identify any changes to the facility that are attrib the Requalification Inspection and Test Program, the utable to noncompliances that constitute unreviewed NRC staff will complete full documentation of the safety questions, and identify changes to the technical NRC's inspection of the licensee's annealing process specifications that are required for operation as a result and place the documentation in the NRC's Public Doc of the noncompliances with the Thermal Annealing ument Room. Operating Plan and relevant portions of the Requalifi cation Inspection and Test Program. If unreviewed If the thermal annealing was completed but not safety questions or changes to technical specifications performed in accordance with the Thermal Annealing are identified as necessary for resumed operation, the Operating Plan and the Requalification Inspection and licensee may restart only after the Director of NRR au Test Program, including the bounding conditions of thorizes restart and the public meeting on the thermal the temperature and times, according to 10 CFR annealing is held.

50.66(c) (2) the licensee must submit a summary of the lack of compliance and a justification for subsequent According to 10 CFR 50.66(d), every licensee operations. The licensee must also identify any who either completes a thermal annealing or termi changes to the facility that are attributable to the non nates an annealing but elects to take full or partial compliances that constitute unreviewed safety ques tions and any changes to the technical specifications credit for the annealing must provide a Thermal An nealing Results Report detailing (1) the time and tem K

that are required for operation as a result of the non perature profiles of the actual thermal anneal, (2) the compliances. This identification does not relieve the post-anneal RTNDT and Charpy upper-shelf energy licensee from complying with applicable requirements values of the reactor material to be used in subsequent of the Commission's regulations and the operating li operations, (3) the projected postanneal reembrittle cense; and if these requirements cannot be met as a ment trends for both RTNDT and Charpy upper-shelf result of the annealing operation, the licensee must ob energy, and (4) the projected values of RTpTS and tain the appropriate exemption per 10 CFR 50.12. If Charpy upper-shelf energy at the end of the proposed unreviewed safety questions or changes to technical period of operation addressed in the application. The specifications are not identified as necessary for re report must be submitted within three months of com sumed operation, the licensee may restart after the pleting the thermal anneal, unless an extension is au NRC staff places a summary of its inspection of the thorized by the Director of NRR.

thermal annealing in the NRC Public Document Room C. FORMAT AND CONTENT FOR THE

and the NRC holds a public meeting on the thermal annealing. On the other hand, if unreviewed safety THERMAL ANNEALING REPORT

questions or changes to technical specifications are The format described here is acceptable to the identified as necessary for resumed operation, the li NRC staff for the Thermal Annealing Report that is to censee may restart only after the Director of the Office be submitted to the Director of NRR for annealing a of Nuclear Reactor Regulation authorizes restart, the reactor vessel to restore fracture toughness of the reac summary of the NRC staff inspection is placed in the tor vessel material. This format addresses the contents NRC Public Document Room, and a public meeting on of the Thermal Annealing Operating Plan, the Requal the thermal annealing is held. ification Inspection and Test Program, the Fracture Toughness Recovery and Reembrittlement Trend As The thermal annealing rule also sets forth in surance Program, and the Identification of Unre

10 CFR 50.66(c) (3) (i) the requirements that a licens viewed Safety Questions and Technical Specifications ee must follow if the thermal annealing was terminated Changes. It also describes acceptance criteria that the prior to completion. In general, the process and re- staff will use in evaluating the applicant's proposed

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programs for determining fracture toughness recovery RTNDT and Charpy upper-shelf energy values, deter and establishing reembrittlement rates. This regulatory mined by either analysis or testing.

guide applies to thermal annealing procedures that use heaters (electric or gas) for heating the reactor vessel, 1.2 Description of the Reactor Pressure Vessel

) the "dry" anneal method. Use of the "wet" anneal method, which applies heat generated by the pump to This section of the report should provide a de tailed description of the reactor pressure vessel and heat the reactor coolant, will be reviewed separately, identify those parts of the vessel to be annealed. It on a case-by-case basis. should also include all vessel data used for determining the Thermal Annealing Operating Plan, the proposed

1. THERMAL ANNEALING OPERATING inspections and tests, and the programs for recovery PLAN and reembrittlement.

The Thermal Annealing Operating Plan should in Information to be reported on each heat of materi clude sufficient information to permit an independent al in the reactor vessel beltline region should include evaluation of all the elements that went into its devel material compositions, including all elements relevant opment. The following sections provide guidance on to irradiation behavior, mechanical properties, fabri the format and content acceptable to the NRC staff for cation techniques, nondestructive test results, and the operating plan, the information that should be in neutron fluence exposures. The initial RTNDT as spec cluded in the plan, the minimum level of detail for this ified in Branch Technical Position MTEB-5-2 in information, and the necessary supporting data. NUREG-0800 (Ref. 12) and NB-2300 of the ASME

In all cases, the information described in this guide Boiler and Pressure Vessel Code (Ref. 13), along with may be referenced if it has been submitted previously the initial Charpy upper-shelf energy as defined in in another document, including any updates to pre ASTM Standard E 185 (Ref. 14), should be reported for each heat of material. Material heats of base metal vious submittals.

and weld metal that will be used for measuring percent recovery and for subsequent surveillance purposes, if

1.1 General Considerations any, should be identified.

This first section should present introductory and All reactor vessel dimensions should be reported, general information. It should identify the reactor and give the reasons that thermal annealing is being pro including diameter, wall thickness, cladding thickness, nozzle dimensions, flange dimensions, and transition j posed, including any regulatory requirement being challenged by the loss in fracture toughness. The pro section dimensions. The dimensions of the gaps be jected percent recovery from annealing and the pro tween the vessel and other potentially affected compo nents such as adjacent concrete structures, internal jected rate of reembrittlement in subsequent reactor permanent structures, and insulation should also be re operations should be identified, as well as the expected remaining operating life after annealing. The projected ported. Attachments to the reactor that could be af fected by the annealing operation and the expected annealing response and reirradiation response should be determined using the provisions of this guide in Sec effects should be identified and described. Examples of such effects are:

tion 3, "Fracture Toughness Recovery and Reem brittlement Assurance Program." In using these provi "* Changes in properties of the vessel insulation, sions, the projected recovery should be determined "* Effects of thermal growth of the reactor on sliding using the annealing time and temperature proposed in support structures, the application.

"* Overheating of instrumentation and attachments.

The operating history of the reactor prior to an nealing should be described in this section, including 1.3 Equipment, Components, and Structures the power-time-temperature history during power op Affected by Thermal Annealing erations to permit evaluation of temperature and flu This section of the report should provide a de ence conditions for the reactor vessel; these data may scription of all equipment, structures, and components be either actual recorded data or data deduced from that could be affected by the annealing operation, other plant information that is identified as such. The either thermally or mechanically, and the expected specific reactor vessel beitline temperatures during the effects to the level necessary to assess the effects of reactor operation should be reported. annealing on the equipment, structures, and compo This section should describe the results of the on nents. Examples of these effects include degradation going surveillance program, including the number of of the biological shield because of loss in strength or specimens, initial values for the reference temperature reduction in neutron and gamma absorption capacity

• > (RTNDT) and Charpy upper-shelf energy, and all data and the effects of vessel growth and distortions on at on shifts of RTNDT and decrease of Charpy upper tached piping. All significant thermal and mechanical shelf energy. It should also provide the pre-annealing loadings projected for each item should be identified,

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as well as actions proposed to avoid damage from these result in unacceptable degradation of the fatigue life of loadings. these components.

The biological shield should be described, includ The parameters to be evaluated in the thermal and ing its dimensions, materials, irradiation exposures, stress analyses should include the annealing tempera any unique features, and all cooling provisions to be ture, hold time at the annealing temperature, heating and cooling rates, the effect of insulation around the K

used for controlling temperatures. If the biological shield is a tank, any provision for circulating the tank vessel Including the bottom head, the active heating coolant should be described. If the biological shield is a length of the heating device, the physical constraints concrete structure, the properties of the concrete on the vessel, structural characteristics of attached pip should be reported as well as the properties of other ing assemblies, and any other restraints.

concrete structures exposed to higher than normal The thermal analysis should establish the tempera temperatures. The existing design temperature limita ture profiles for the inside and outside surfaces of the tions for the concrete should be described. If the de vessel wall during heatup, start and end of steady-state sign temperature limitations are to be exceeded during conditions, and cooldown conditions. The effects of the thermal annealing operation, an acceptable maxi localized high temperatures should be evaluated for mum temperature for the concrete should be estab degradation of the concrete adjacent to the vessel, for lished as addressed in Paragraph 1.4. changes, if any, in thermal and mechanical properties The piping attached to the vessel should be de of the reactor insulation and for detrimental effects, if scribed. This description should include material any, on containment and the biological shield. Maxi types, dimensions, and restraints such as supports and mum concrete temperatures should be based on exist snubbers. The design requirements with respect to ing design limits or provisions of Section III, Division temperature and bending stress or strain limitations 2, of the ASME Boiler and Pressure Vessel Code should be identified for the piping. Further, any indi (Ref. 15). If the design temperature limits or the cations of potential flaws found during inspections of ASME Boiler and Pressure Vessel Code limits for the the piping should be described. adjacent concrete structure are projected to be ex ceeded during the annealing operation, an acceptable Any other equipment or instrumentation that maximum temperature for the concrete must be estab could be affected by the thermal annealing should be fished using appropriate test data. Test data on proper described. A description of the overall containment as ties should address irradiated concrete of appropriate it relates to core removal and storage, as well as the type and exposures to time-temperature conditions annealing of the vessel, should be included. Any spe that bound the expected conditions of the concrete cial requirements should be described in detail. For during annealing.

example, storage of core internals may require a coffer dam approach to isolate the coolant from the heating The structural analysis should evaluate, for the equipment in the drained vessel, in which case the complete annealing cycle, residual deformations, re modifications, the equipment, and the method for en sidual stresses, elastic-plastic-creep effects (Ref. 16),

suring the integrity of the isolation seals should be distortions, bending, piping displacements, effects of thermal gradients (axial, azimuthal, and through detailed.

wall), and restraints on the vessel, including nozzles and flange and attached piping. Any potential interfer

1.4 Thermal and Stress Analyses ence with other equipment, components, or supports This section of the report should provide an evalu should be evaluated. This section should specify the ation of the effects of mechanical and thermal stresses limiting parameters established by these analyses, in and temperatures on the vessel, containment, biologi cluding maximum temperature, maximum stress, and cal shield, attached piping and appurtenances, and ad limiting heatup and cooldown rates.

jacent equipment, components, and structures that demonstrates that the annealing operation will not be 1.5 Thermal Annealing Operating Conditions detrimental to reactor operation. This evaluation This section of the report should describe the pro should include detailed thermal and structural analyses posed thermal annealing operating conditions, includ that establish appropriate time and temperature pro ing bounding conditions of temperature and time, and files, including the heatup and cooldown rates of the heatup and cooldown schedules. The annealing annealing operation, so that dimensional stability of parameters should be selected to provide sufficient re the system will be maintained. The analyses should covery of fracture toughness to satisfy the require demonstrate that localized temperatures, thermal ments of 10 CFR 50.60 and 10 CFR 50.61, or any stress gradients, and subsequent residual stresses will other objective identified in the application, for the not result in unacceptable dimensional changes or dis proposed post-anneal period of operation. The anneal tortions in the vessel, attached piping, and appurte ing parameters should be compatible with design stress K

nances and that the thermal annealing cycle will not limits of the reactor and any other component or

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structure expected to experience significant tempera cent equipment, components, and structures. Instru ture or stress gradients during the annealing operation. mentation should also be installed to determine the Limitations such as the physical constraints resulting stress profiles in these items, including the effects of from attached piping, supports, snubbers, and other thermal gradients in the axial, azimuthal, and through components and the thermal and mechanical stresses thickness directions during all transient and steady generated in the vessel and piping during the annealing state aspects of the annealing operation. The accuracy operation should also be considered. and reliability of the measurements should be demon strated. The stresses and strains caused by temperature This section should identify the proposed anneal gradients may be established by analysis in combina ing temperature, time at temperature, heatup rate, tion with on-line measurements of temperature or cooldown rate, and the limitations and permitted vari displacements.

ations in these conditions. The limitation on tempera ture variations should include axial, azimuthal, and The annealing procedure should detail the opera through-wall gradients and the permissible tempera tional steps to be taken during the annealing operation ture profiles in the vessel during heatup, cooldown, and should include all quality assurance measures and steady-state heating. The bases used to establish needed to ensure an effective annealing operation.

these annealing parameters should be described. The annealing procedure should identify the controls that will be in place and how these will be applied and The time and temperature parameters identified maintained throughout the annealing operation. The in the Thermal Annealing Operating Plan should be annealing procedure should describe how the heat based on the thermal and stress analyses described in treatment equipment will be installed and removed Section 1.4 and should represent the bounding times, from the vessel; what procedures will be instituted to temperatures, and heatup and cooldown schedules for control radioactive contamination before, during, and the thermal annealing operation that should not be vio after the annealing operation; and how the vessel will lated during the annealing operation. If these bound be drained and dried prior to annealing. The proce ing conditions for times and temperatures are violated dures should detail the precautions to be taken to pre during the thermal annealing operation, the analysis clude cooling water leakage into the vessel during the will no longer be valid and the annealing operation is annealing operation; such leakage could result in a considered not in accordance with the Thermal An steam explosion or a thermal shock to the vessel.

nealing Operating Plan. In that case, the licensee should follow Section 5.2 of this regulatory guide. 1.7 Proposed Annealing Equipment This section of the report should provide a de

1.6 Description of Annealing Method, scription of the equipment to be used for the in-service Instrumentation, and Procedures annealing. It should describe the heating apparatus This section of the report should describe in detail and the general plant layout to support the annealing the method selected for annealing the vessel as well as operation; the controls and instrumentation, including the proposed instrumentation and procedures to be redundant controls; and equipment for measuring and applied during the annealing operation. The annealing recording the temperatures and temperature profiles.

operation should not degrade the reactor or other This section should describe how the equipment will equipment, components, and structures to such an ex operate, as well as what provisions will be made to pro tent that their ability to perform intended safety func tect personnel from radiation exposure and to protect tions can no longer be maintained. The annealing op instruments and equipment from temperature effects eration must be compatible with the original design during the annealing operation.

limits of the reactor system or the incompatibility The heating apparatus should be designed, pro should be described and justified. In such cases, addi vided with instrumentation, and controlled so that the tional design analysis may be required.

entire section of the vessel to be annealed is effectively The annealing method should be determined held at a uniform temperature, within the bounds es based on constraints from reactor design and accessi tablished by the Thermal Annealing Operating Plan, bility to the reactor vessel to allow insertion of equip throughout the annealing period. Redundancy in heat ment and instrumentation. Selection of the method ing devices, controls, and instrumentation should be also should be based on the expected structural effects discussed in the operating plan. The temperature con on the primary system that result from temperature trol system should be able to control temperatures suf gradients in the pressure vessel. ficiently to avert adverse effects from thermal gradients during heatup, annealing, and cooldown operations.

The reactor vessel and adjacent equipment, com ponents, and structures should have instrumentation 1.8 ALARA Considerations that permits on-line measurement of temperatures at

"locationsthat are needed to assess the entire tempera This section of the report should describe the steps ture profile of the reactor pressure vessel and the adja- to be taken to minimize occupational exposure, in

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accordance with the "as low as is reasonably achiev plan (see Section 1.5 of this guide) are not exceeded.

able" (ALARA) principle and the provisions of 10 Temperature measurements should be made at suffi CFR 20.1206. Special training of the personnel who cient locations to establish temperature profiles for will actually perform the annealing operations should both the inside and outside surfaces of the reactor ves be described. Equipment and procedures for monitor sel. These measurements should be made for the en ing and control of airborne radioactive particles during tire length of the vessel along axial directions where it K

the operation should be identified. This section should is physically possible, at a minimum of two different specifically address precautions to be taken to avoid azimuthal locations (which should include the top and excessive exposure from radiation streaming when the bottom positions of the heating zone), at locations reactor internals are being removed and stored, when within the heating zone where there may be cold spots the reactor coolant is removed from the reactor, and (e.g., at joints between the heaters), at locations on when the heating equipment is being moved into and each nozzle, and on any other component that is ex out of the reactor vessel. It should also describe steps pected to be significantly affected by the annealing taken to minimize occupational exposure from radio treatment. Measurements should be made with suffi active waste processing, radioactive materials decon cient frequency to identify any temperature excursions tamination, and radioactive waste shipment. that could lead to violating the established temperature limits. When appropriate, the measurement devices

1.9 Summary of the Thermal Annealing should be in physical contact with the component Operating Plan when the temperature is being measured. The meas The Thermal Annealing Report should contain a urement records should be retained for possible review summary of the Thermal Annealing Operating Plan and inspection by the NRC, in accordance with the re that includes the highlights of each section, the key pa quirements of 10 CFR 50.71(c), until the facility li rameters of annealing, and the major conclusions of cense is terminated. The temperature measurements the plan. The projected percent recovery and the pro should be monitored and compared to pre-established jected reembrittlement rate should be identified, as tolerance bounds during heatup, steady-state opera well as the projected end-of-license values of RTNDT tion, and cooldown to ensure that temperature and and Charpy upper-shelf energy after annealing. stress limits have not been exceeded.

Stress limitations should be monitored by a proce

2. REQUALIFICATION INSPECTION AND dure established by the licensee that uses strain gauges TEST PROGRAM or alternative methods, for example, deflection meas The inspection and test program to requalify the urements or temperature measurements. Experimen K

annealed reactor vessel should include the detailed tal evidence of the validity of alternative methods monitoring, inspections, and tests proposed to demon should be provided. Measurements should be made to strate that the limitations in the operating plan on tem establish stress levels at the vessel locations of highest peratures, heat treatment times, temperature profiles, stress, on the vessel nozzles, on the flange, and on and stresses have not been exceeded. The detailed high-stress piping locations.

monitoring, inspections, and tests should also establish the thermal annealing time and temperature to be used This section should describe the measurement in quantifying the fracture toughness recovery. The type, the number of measurements to be made for program should also demonstrate that the annealing each component, measurement sensitivity, measure operation has not degraded the reactor vessel, at ment frequency, and recording method.

tached piping or appurtenances, or the adjacent con crete structures to a degree that could affect the safe 2.2 Inspection Program operation of the reactor after annealing.

This section should describe the inspection pro The program should identify the limiting parame gram proposed to affirm that the annealing operation ters established for the thermal annealing operation has not damaged the reactor vessel or related equip conditions, identify the physical measurements and ment, components, or structures. The inspection pro tests to be made to ensure that these conditions are not gram, as a minimum, should include a pre- and post exceeded, describe the instrumentation to be used for anneal visual examination of critical regions of the making these measurements and tests, and state the vessel, piping, and any other equipment, component, quality assurance provisions to be applied.

or structure that might be affected by the annealing operation. The inspection program should also de

2.1 Monitoring the Annealing Process scribe a nondestructive examination program for the This section should identify the measurements, reactor vessel beltline region that will ensure that the with their locations, that will be used to monitor the vessel will continue to perform its safety function after annealing process and to make certain that the pro the annealing operation. The description of the inspec posed annealing conditions evaluated in the operating tion program should include acceptance criteria, type

1.162-8

and number of examinations, qualification require 3.1.1 Vessel Surveillance Program Method ments, and reporting requirements. If the plant's surveillance program has resulted in

"credible" data (as defined in the PTS rule, 10 CFR

) 2.3 Testing Program 50.6 1), and broken specimens from that program have been retained (as recommended in NRC Information This section should describe the testing program Notice No. 90-52, "Retention of Broken Charpy that will be performed to demonstrate the effectiveness Specimens," Reference 17), the thermal annealing of the annealing operation and to assure that the reac rule (10 CFR 50.66) requires that broken specimens tor vessel, attached piping and appurtenances, and ad from surveillance specimens and any remaining untest jacent concrete will continue to perform their intended ed surveillance specimens be used to evaluate anneal safety function following the annealing operation. This ing recovery on a material-specific basis. The broken program is expected to be unique to each plant and specimens should be reconstituted (see Section 3.1.4 should be established by the applicant. The testing pro of this guide) to form new, full-size specimens with the gram may, for example, test the effects on the vessel of insert material being the only material from the original annealing, the integrity of the concrete, the operability surveillance specimen. These reconstituted specimens, of instrumentation, and the post-anneal functioning of and any untested specimens from the original speci affected components, equipment, and structures. men complement, should be annealed at time and temperature conditions that are equal to or are bounded by the actual vessel annealing conditions.

3. FRACTURE TOUGHNESS RECOVERY AND

REEMBRITrLEMENT ASSURANCE This method may be applied to broken specimens PROGRAM from a single capsule or multiple capsules from the sur The fracture toughness recovery and reembrittle veillance program. Specimens from at least two cap ment assurance program should describe the methods sules should be used, with the fluences of the two cap to be used for quantifying the percent recovery, the sules spanning the peak fluence of the reactor pressure reembrittlement trend and for establishing the post vessel beltline; this ensures that an interpolation of the anneal RTNDT and Charpy upper-shelf energy values. annealing recovery is possible. If broken specimens These tasks are important for evaluation of the safety from only a single capsule are used, the specific surveil margins of the reactor pressure vessel in subsequent lance capsule chosen should be the one for which the operating periods. The methods outlined below pro fluence most closely matches the peak fluence of the vide experimental and computational means for quan reactor pressure vessel beltline.

tifying both the recovery of fracture toughness follow As an alternative, materials test reactor (MTR) ir ing the thermal anneal and the reembrittlement rate radiations of the vessel-specific limiting material may with subsequent plant operation. be used as a method for determining percent recovery on a material-specific basis (see Section 3.3 of this

3.1 Fracture Toughness Recovery Program guide).

This section of the assurance program should de Methods for testing the specimens and using the scribe the method planned to determine the percent resultant data are discussed in Sections 3.1.5 and recovery, including any computations or tests. The 3.1.6 of this guide.

methods discussed below provide experimental and The assurance program should describe the plans computational means for determining the percent re for using the surveillance results, including a descrip covery of ARTNDT, Rt, and the percent recovery of tion of the procedure for generating post-anneal prop Charpy upper-shelf energy, RUSE. erties (e.g., reconstituted specimens or whole pre viously untested specimens) and the method for using As provided in the thermal annealing rule (10

the surveillance measurements of percent recovery to CFR 50.66), one of three methods may be used to evaluate the recovery in fracture toughness following evaluate the percent recovery for the vessel material.

the thermal annealing. One method requires the use of 3.1.2 Irradiated Vessel Material Method surveillance specimens from "credible" surveillance programs (as defined in the PTS rule, 10 CFR 50.61) An alternative method for determining the per to develop material-specific data, if such specimens are cent recovery uses the results of a verification test pro available. The most accurate, but difficult, second gram employing materials removed from the beltline method uses material removed from the reactor pres region of the reactor vessel. For this method, the sam sure vessel beltline to develop plant-specific data. The ples removed from the vessel are used to evaluate the third method uses generic computations to estimate as-irradiated or pre-anneal condition of the material

> the recovery. These three methods are described be and the post-anneal condition of the material. The low. Values of percent recovery (RUSE and RI) may post-anneal condition is evaluated from specimens that not exceed 100 percent. have been annealed at the time and temperature

1.162-9

conditions equal to or bounded by the reactor vessel pliance with the applicable stress limits of Section III of annealing conditions. the ASME Code must include consideration of the ef The number of samples to be removed from the fects of fatigue and corrosion on the exposed base metal following removal of the samples, and the analy vessel depends on many factors, including the size of ses should consider any thermal and mechanical ef the samples, the reason for the annealing (determining fects on the surface and near-surface material remain both RTNDT and Charpy upper-shelf energy requires ing after removal of the sample.

more tests than determining only one of these quanti ties), the testing plans (specimen size and type), and The assurance program should describe the meth the acceptability of removing the samples from the ves od proposed to characterize the depression remaining sel. The samples removed from the vessel beltdine can after sample removal to ensure that the condition of be used to fabricate full-size Charpy specimens, inserts the remaining material is bounded by the assumptions for reconstitution into full-size Charpy specimens, or in the analysis and is acceptable. Any proposed re sub-size Charpy specimens (see subsections 3.1.4.3 pairs, including weld repair, should be described in the and 3.1.6.4). Other test methods for irradiated vessel Thermal Annealing Operating Plan.

material may be used if appropriate justification is 3.1.2.2 Testing of Material Removed from the provided. Vessel Beltline. One impediment to the quantitative This method is plant-specific and, as described be use of data from testing of material removed from the low, several criteria must be satisfied to demonstrate vessel beltine in determining the percent recovery of the acceptability of this method for a specific applica ARTNDT and Charpy upper-shelf energy is that this tion or plant. material represents surface or near-surface properties of the material. In contrast, ASTM Standard E 185 The assurance program should provide a complete (Ref. 14) requires the use of material from the 1/4T

description of the plans for using samples removed location of plate and forging products, and more than from the vessel beltline, including the method for re 0.5 in. from the surface of weld metals, to determine moving the samples; the number, size, and location of ARTNDT and Charpy upper-shelf energy.

the samples; analyses to demonstrate acceptability of To permit a quantitative use of material removed the sample removal; the experimental plans for using from the vessel beltiine to determine the percent re the samples (size and number of specimens, test plans covery of ARTNDT and Charpy upper-shelf energy, and procedures, etc.); and the method for determin samples removed from the vessel beltline should be ing the percent of recovery of the vessel material from used to evaluate both the pre-anneal and the post the results of the tests from the samples.

anneal properties of the near-surface materia

l. Speci

3.1.2.1 Acceptability of Removing Material mens used to evaluate the post-anneal properties from the Vessel. The acceptability of the method used should be annealed at time and temperature condi for removing samples from the vessel beltline is based tions that equal or are bounded by the actual vessel on local and global considerations, both of which annealing conditions. The resulting percent recovery should be addressed in the assurance program. The of transition temperature at the 30 ft-lb level may be global considerations concern the impact on overall used to determine the percent recovery for ARTNDT,

vessel integrity of the depression, hole, or surface dis R1. The resulting percent recovery of the Charpy continuity remaining after removal of the sample, and upper-shelf energy may be used to determine the Char they are addressed through analysis. The local consid py upper-shelf energy, RUSE.

erations concern thermal and mechanical effects and Samples removed from the vessel beltiine material surface quality effects on the surrounding material re can be used to develop test specimens in several man maining after removal of the samples. ners. The samples can be used to fabricate fullsize The sample removal process should be described Charpy specimens, inserts for reconstitution into full in the assurance program, including a description of size Charpy specimens, sub-size Charpy specimens or the measures to ensure the identification and docu other test specimens of the approved plan. Preparation mentation of the orientation of the sample relative to and use of these various specimen types is described in the vessel. Section 3.14.

The removal of samples from the vessel beltline 3.1.3 Computational Method will result in a depression or other surface discontinuity The computational method uses generic equations in the vessel wall. As required in the thermal annealing (Equations I and 2) to determine the percent recovery rule (10 CFR 50.66), it must be demonstrated that the of Charpy upper-shelf energy (USE) and ARTNDT re resulting depressions satisfy the stress limits of the spectively. Alternative computational methods may be applicable portions of the ASME Code Section III, re used if appropriate justification is provided. When de gardless of the applicable section of the Code for the termining the projected percent recovery for the an vessel design. The analyses used to demonstrate com- nealing plan, the proposed lower-bound annealing

1.162-10

time and temperature are used in Equations 1 and 2. 3.1.4 Specimen Handling and Preparation Procedures However, when computing the post-anneal percent recovery, the actual annealing time and the lower 3.1.4.1 Specimen Handling Procedures. For re bound of the range of actual annealing temperatures constituting surveillance specimens or removing sam determined from the instrumentation (see Section 2.1 ples from the vessel beltine, the assurance program of this guide) should be used. should describe the methods to be used for marking and handling the test materials to ensure that the

9 RUSE = {[1-0.586 exp(-ta/15. )] x [0.570AUSEi + orientation of the material relative to the vessel is un

4

(0.120Ta-10 ) Cu+0.0389Ta-17.6]} x ambiguous and traceable.

(100/AUSEji (Equation 1) 3.1.4.2 Specimen Orientation. The assurance program should address the orientation of the speci where mens to be tested. For testing materials from a "cred ible" surveillance program for the evaluation of per RUSE = percent recovery of USE from cent recovery of ARTNDT, it is preferable to test the annealing, post-anneal specimens in the same orientation as the AUSEi = (mean USE unirradiated - mean original surveillance tests. In contrast, for evaluation of USE after irradiation), Charpy upper-shelf energy, it may be preferable to use ta = time at annealing temperature in specimens oriented in the transverse direction, the hours, T-L orientation, according to ASTM Standard E 399 Ta = annealing temperature in °F, (Ref. 18).

Cu = copper content of material in weight-percent. 3.1.4.3 Reconstitution of Charpy Specimens.

Reconstitution of Charpy specimens is used to provide R== [0.5 + 0.5 tanh((aTa - a2)/95.7}1]* 100 new full-size Charpy specimens from the broken pieces (Equation 2) of previously tested specimens and to conserve materi al. Several methods are available for welding end-tabs where onto the test material section, with the goal of each method to provide a structurally sound and testable Rt = percent recovery of transition temperature from annealing, specimen (i.e., the specimen does not fracture at the reconstitution welds) without overheating (possibly in Sal = 1 + 0.0151 ln(ta)

0.424Cu( 3 .2 8 - 0.00306Ta) ducing annealing) the test section.

a2 = 0.584(Ti + 637), for Ta k 800OF The assurance program should describe the proce or a2 = 0.584T1 - 15.51n(O) + 833 for dures to be used for the reconstitution process. The Ta < 750 0F procedures and criteria of ASTM Standard E 1253-88 where (Ref. 19), "Standard Guide for Reconstitution of Irra diated Charpy Specimens," are sufficient to demon flux rate, n/(cm2-s), strate an adequate reconstitution method. Other pro TC temperature of irradiation. posed methods should have appropriate justification.

Cu. copper content of material in weight percent with maximum 3.1.5 Specimen Testing value of 0.3%.

3.1.5.1 Test Procedures. The assurance pro The current Rt equation is not accurate between gram should describe the procedures used for testing annealing temperatures of 750 and 800 0 F. Until a the Charpy specimens, either full-size or sub-size. For complete equation is developed an extension of the ef testing full-size Charpy specimens, either reconstituted fect of the flux term (a2) is assumed to a temperature specimens or specimens fabricated from samples re of 775*F. Between 775 and 800°F, a linear interpola moved from the vessel beltline, testing should be per tion between Equation 2 evaluated at 775 °F with the formed using equipment and testing procedures similar flux term and Equation 2 evaluated at 8000 without to those used to develop surveillance data, as outlined the flux term should be made. in ASTM Standards E 185-82 (Ref. 14) and E 23-88 Since plant operational characteristics do not re (Ref. 20).

sult in a unique value of irradiation temperature Testing of sub-size Charpy specimens should fol throughout a plant's lifetime, the method used for low the general procedures and methods in ASTM

evaluating Ti should be described in the assurance Standard E 23-88 (Ref. 20) for testing of full-size program. Charpy specimens. For testing sub-size Charpy speci Equations 1 and 2 are documented in Reference 8 mens, a description of the general test procedures and and represent mean values of the percent recovery of the testing equipment should be provided in the assur RTNDT and USE. ance program. In addition, a method for applying the

1.162-11

results from the tests of the sub-size specimens should reference temperature (RTNDT) and Charpy upper be described. shelf energy are evaluated using Equations 3 and 4:

3. 1.S.2 Test Plan. The assurance program should RTNDT(A) = RTNDT + ARTNDT x describe the test plan, including the number of speci

(100 - W) / 100 (Equation 3)

mens to be tested and the method for selecting test temperatures. This description should include the K

steps taken to ensure that a reasonable measure of CvUSE(A) = CvUSE(u) [I - D x recovery is achieved by the testing; the description (100 - RUSE)/100001 (Equation 4)

should also include the proposed method for handling where uncertainty in the test results.

RTNDT(A) = reference temperature, RTNDT,

Testing to evaluate percent recovery of ARTNDT of the material in the post should result in an unambiguous evaluation of the tem anneal condition in OF,

perature at which the average Charpy energy versus RTNDT(U) = reference temperature, RTNDT,

temperature curve achieves an energy level of 30 ft-lb. of the material in the It is preferable that the testing cover a broad range of preservice or unirradiated results based on the shear percentage (from near 0 condition in °F,

percent to greater than 95 percent) to permit a more ARTNDT = mean value of the transition complete assessment of the Charpy data trends for the temperature shift, or change material and to preclude false trends from data cluster in RTNDT, from irradiation ing around the 30 ft-lb level. (before annealing) in °F,

Testing to evaluate Charpy upper-shelf energy re Rt = percent recovery of ARTNDT

covery should provide an unambiguous definition of from annealing, the Charpy upper-shelf energy for the material. All CvUSE(A) = Charpy upper-shelf energy of tests used in evaluating the Charpy upper-shelf energy the material in the post should result in 100 percent shear. anneal condition in ft-lb, CvUSE(u) = Charpy upper-shelf energy of the material in the preservice

3.1.6 Quantification of Post-Anneal Initial or unirradiated condition in Properties ft-lb, Quantification of the post-anneal initial proper ties, RTNDT and Charpy upper-shelf energy, is depen D = percent decrease in Charpy upper-shelf energy from K

dent on the method used to determine the percent re irradiation (before annealing),

covery by annealing. and RUSE = percent recovery of Charpy

3.1.6.1 Vessel Surveillance Program Method.

upper-shelf energy from The assurance program should describe the surveil annealing.

lance results to be used in evaluating percent recovery, along with the proposed method to relate the observed The values of RTNDT(A) and CvUSE(A), calcu percent recovery from the surveillance results to the lated using Equations 3 and 4, respectively, should be percent recovery of the vessel material. used as the values of reference temperature (RTNDT)

and Charpy upper-shelf energy, respectively, at the ini One method for relating the observed recovery to tiation of continued plant operation.

the vessel recovery is to compare the measured recov

3.1.6.2 Irradiated Vessel Material Method.

ery from the test (or tests) of surveillance specimens to Since the Charpy data evaluated by this method repre the percent recovery from Equations 1 and 2 for the sent the surface or near-surface properties of the plate, surveillance capsule. The average ratio (or the ratio weld, or forging, the measured values cannot be used from a single capsule) between the measured recovery directly to represent the plate or forging 1/4T or weld from the surveillance capsules and that evaluated from bulk properties as called for by Regulatory Guide 1.99, Equations 1 and 2 for the surveillance capsule provides Revision 2 (Ref. 21). However, evaluations of the pre a material-specific adjustment for the generic equa anneal and the post-anneal properties of the sample tion. In this method, the vessel percent recovery would removed from the vessel beltline will provide sufficient be calculated by multiplying the percent recovery de data to evaluate the expected recovery at the plate or termined from Equations 1 and 2 for the vessel by the forging II4T level or the weld bulk properties.

average ratio adjustment. Values of Rt and RUSE de termined from surveillance data may not exceed 100. The assurance program should describe the proce dures used to evaluate percent recovery for the vessel Once suitable values of Rt and RUSE have been materials from the measurements resulting from this determined from the surveillance data, the post-anneal method.

1.162-12

If the samples removed from the vessel beltline are the 1/4T location of plate and from the peak flux location for the material, one pro forging, or the weld bulk cedure to evaluate the post-anneal properties from the properties in ft-lb, measured percent recovery uses the following D = percent decrease in Charpy equations: upper-shelf energy from irradi ation embrittlement, for the

1/4T location of plate and RTNDT(A) = RTNDT(U) + ARTNDT x [1 - (iTs 1 (Equation 5) forging, or the weld bulk TTSA) (RS / ATs)]

properties,

= from the measured surface CvUSE(A) = CvUSE u (1 - D / 100) data, the Charpy upper-shelf (CvSA /CVS) (Equation 6) energy for the pre-anneal condition in ft-lb, where CVSA = from the measured surface data, the Charpy upper-shelf RTNDT(A) = reference temperature, energy for the post-anneal RTNDT, of the material in the condition in ft-lb.

post-anneal condition in *F,

RTNDT(U) = reference temperature, The values of RTNDT(A) and CvUSE(A) calcu RTNDT, of the material in the lated using Equations 5 and 6 should be used as the preservice or unirradiated values of reference temperature (RTNDT) and Charpy condition in *F, uppershelf energy, respectively, at the initiation of ARTNDT = mean value of the transition continued plant operation.

temperature shift, or change in RTNDT, caused by If the samples removed from the vessel beitline do irradiation (before annealing) not come from the vessel peak flux location, the vessel in -F, embrittlement and the annealing recovery will not be

"iTs1 = from the measured surface as great as that for the vessel peak flux location, and data, the transition tempera Equations 5 and 6 should underestimate the percent ture at the 30 ft-lb energy recovery of the vessel. In such cases, the assurance level for the pre-anneal program should describe the procedures used to evalu condition in *F, ate percent recovery for the vessel materials from the

"7SA = from the measured surface measurements resulting from this method.

data, the transition tempera 3.1.6.3 Computational Method. For the compu ture at the 30 ft-lb energy tational method, the postanneal initial RTNDT and level for the post-anneal Charpy upper-shelf energy values for each beltiine ma condition in *F, terial should be evaluated using Equations 3 and 4, us RS= the ratio of Rt from Equation ing the values of Rt and RUSE from Equations 2 and 1.

2 for the 1/4T location to Rt from Equation 2 for the The values of RTNDT(A) and CvUSE(A), calcu surface of the plate, forging or lated using Equations 3 and 4, respectively, should be weld, used as the values of reference temperature (RTNDT)

ATs = the mean value of the transi and Charpy upper-shelf energy, respectively, at plant tion temperature shift (in *F) restart.

at the surface of the plate, forging or weld, determined 3.1.6.4 Sub-size Charpy Specimens. Sub-size using the surface fluence and Charpy specimens can provide very useful information the same calculational method concerning the toughness of the material and the re used to evaluate ARTNDT for covery of irradiation embrittlement by annealing. Data the 1/4T location, from sub-size Charpy specimens is not uniquely corre CvUSE(A) = Charpy upper-shelf energy of lated to data from full-size Charpy specimens in an the material in the post-anneal absolute sense, but quantities such as irradiation condition, for the 1/4T embrittlement shift and annealing recovery can be eva location of plate and forging, luated from sub-size specimen data. Although some or the weld bulk properties in work has been under way in the United States In this ft-lb, area using several sub-size Charpy specimen designs, CvUSE(u) = Charpy upper-shelf energy of there is no consensus in the U.S. technical community the material in the preservice or the nuclear industry as to a preferred specimen de or unirradiated condition, for sign or an appropriate correlation method.

1.162-13

The assurance program should describe the over where all test plan for the use of sub-size Charpy specimens, RTNDT = reference temperature, including test specimen design and test procedures. In RTNDT, of the material in the addition, the assurance program should describe the irradiated condition in OF,

method proposed for quantitative use of the sub-size RTNDT(U) = reference temperature, Charpy specimen data in evaluating percent recovery, RTNDT, of the material in the including any experimental demonstration validating preservice or unirradiated the method. condition in OF,

ARTNDT = mean value of the transition

3.2 Reembrittlement Trend Assurance Program temperature shift, or change As specified in the thermal annealing rule (10

in RTNDT, from irradiation in CFR 50.66(b) (3) (ii) (B)), the reembrittlement trend OF, and M = margin term in OF to account of both RTNDT and Charpy upper-shelf energy is to be estimated to establish the projected embrittlement at for uncertainties in the values the end of the proposed period of plant operation, and of RTNDT(U), nickel and is to be monitored during post-anneal reactor opera copper content, fluence and the calculational procedures, tions to confirm these estimates, using a surveillance program that conforms to the intent of Appendix H of as determined from Equation 8.

10 CFR Part 50. An appropriate method for estimating the reembrittlement trend is using a "lateral shift" (Equation 8)

M - 2 Ja1 2 + a,2 method. For the "lateral shift" method, the reem brittlement trend is the same as the embrittlement where trend used for the pre-anneal operating period, regard less of whether the embrittlement was determined al = standard deviation of RTNDT(U) in OF,

using the procedures of the PTS Rule (10 CFR 50.61(c)) for embrittlement of RTNDT, or the proce =

=a standard deviation of ARTNDT in OF.

dures of Revision 2 of Regulatory Guide 1.99 (Ref. 21) From Revision 2 of Regulatory Guide 1.99 (Ref.

for embrittlement of Charpy upper-shelf energy, or 21), VA is 17°F for base metals and aA is 28*F for whether embrittlement was determined from "cred weld metals.

ible" surveillance data.

Further, ARTNDT is given by:

The assurance program should describe the pro gram to estimate reembrittlement trends prior to the ARTNDT = (CF) (f)(0.28 - 0.10 log f)

development of "credible" data from the reembrittle (Equation 9)

ment surveillance program and should describe the surveillance program to be used for post-anneal plant where operation.

CF chemistry factor (in OF) based on the nickel and copper content of the

3.2.1 Lateral Shift Method material, or based upon results from

3.2.1.1 Description of the Lateral Shift Meth the surveillance program if the od. As illustrated in Figure 1 for both RTNDT and program is "credible" according to the Charpy upper-shelf energy, the lateral shift method re criteria in the PTS rule (10 CFR

sults in a shift of the initial irradiation embrittlement 50.61), and curve along the fluence axis, using the post-anneal f = the best-estimate neutron fluence (in properties (ARTNDT and Charpy upper-shelf energy) units of 1019 n/cm 2 , E > 1 MeV), at as the basis point. This method has been found to be the clad-base metal interface on the conservative in bounding subsequent embrittlement. inside surface of the vessel.

3.2.1.2 Reembrittlement of RTNDT. The reem For reembrittlement, the lateral shift is accom brittlement of RTNDT is established using the same plished by determining the "transition recovery flu embrittlement trend as the pre-anneal operating peri ence," ft, by solving for the fluence value that satisfies od, with a lateral shift. Equation 10:

For the pre-anneal operating period, RTNDT is RTNDT(A) - RTNDT(U) = (CF) (fl) (0.28 - 0.10 log ft)

given by: (Equation 10)

RTNDT = RTNDT(U) + ARTNDT + M where RTNDT(A) is the reference temperature, K

(Equation 7) RTNDT, of the material in the post-anneal condition.

1.162-14

LATERAL SHIFT METHOD

Fluence (n/cm2 , E > 1 MeV)

LATERAL SHIFT METHOD

wI

Ruence (nIcr2, E > I MoV)

Figure 1

1.162-15

For reembrittlement, the reference temperature rately model the various Charpy uppershelf energy de (RTNDT) is evaluated by crease curves. Using the equations in Reference 21:

RTNDT = RTNDT(U) + ARTNDT + M CvUSE = CvUSE(u) x [I - D/100]

(Equation 11) (Equation 14) K

for base metals:

where D = (100 Cu + 9) (00.2368 ARTNDT = the mean value of the shift in for weld metals:

reference temperature caused D = (100 Cu + 14) (f0.2368 by irradiation (as given below by Equation 12) in OF, and the upper bound:

M = margin term in OF to account D = 42.39 (f)0.1502 for uncertainties in the values of RTNDT(U), nickel and where copper content, fluence, and CvUSE = Charpy upper-shelf energy of the calculational procedures, the material in the irradiated as given by Equation 13. condition (before annealing)

in ft-lb, ARTNDT = (CF) (f + ft)[0.28 - 0.10 log (f + ft)] CvUSE(U) = Charpy upper-shelf energy of (Equation 12) the material in the preservice or unirradiated condition in where ft-lb, D = percent decrease in Charpy CF = the same chemistry factor (in OF) upper-shelf energy from used for the pre-anneal operating irradiation (before annealing),

period, based on the nickel and Cu = copper content (weight-percent)

copper content of the material or for the subject material, and the results of the "credible" f = the best-estimate total neutron surveillance program, fluence (in units of 1019 n/cm 2 ,

f f the increment of best-estimate E > I MeV), at the clad-base neutron fluence (in units of 1019 metal interface on the inside n/cm 2 , E > 1 MeV), at the clad-base surface of the vessel.

metal interface on the inside surface The value of D is the lesser of that from the appro of the vessel, accumulated during priate equation for the material type and that from the plant operation subsequent to the upper bound equation. For "credible" surveillance annealing operation, and data, guidance is given in Revision 2 of Regulatory ft= the "transition recovery fluence,"

Guide 1.99 (Ref. 21) for determining percent decrease evaluated from Equation 10.

in Charpy upper-shelf energy based on the surveillance results.

M - 2 la_ + ao2 (Equation 13) For reembrittlement, the lateral shift is accom plished by determining the "shelf recovery fluence,"

fs, from:

where

01 = standard deviation of for base metals:

RTNDT(U) in OF,

= standard deviation of ARTNDT in OF,

1 - (CvUSE(A)/CvUSE(u)) 14.223

3.2.1.3 Reembrittlement of the Charpy Upper Shelf Energy. The reembrittlement of the Charpy upper-shelf energy is evaluated using the same for weld metals:

[

embrittlement trend as the pre-anneal operating peri

]

od, with a lateral shift. For the pre-anneal operating 4.223 period, Revision 2 of Regulatory Guide 1.99 (Ref. 21) I1- (CvUJSE(A)/CvUSE(u))

K`1 gives upper-shelf energy decrease in a graphical form 100 Cu +14 only. Merkle (Ref. 22) developed equations that accu-

1.162-16

the upper bound:

f'

(CvUSE(A)/CvUSE(u))

1 - "42.39 I6.658 3.3 Use of Materials Test Reactor (MTR)

Irradiations If archival pieces of the limiting vessel material are available, materials test reactor (MTR) irradiations may be used to evaluate the recovery and reembrittle ment trends. Plans for using archival material should (Equation 15) be described in the assurance program, including the traceability of the material to the vessel, the proposed For both weld metal and base metal, the correct experimental matrix, and the method proposed for us value of fs is the larger of the values from the appropri ing the results of the MTR irradiations.

ate equation for the material type and the upper bound equation. 4. IDENTIFICATION OF UNREVIEWED

SAFETY QUESTIONS AND TECHNICAL

Reembrittlement of Charpy upper-shelf energy is SPECIFICATION CHANGES

evaluated from: Any changes to the facility that are described in the updated final safety analysis report as unreviewed CvUSE = CvUSE~u) x [1 - D/100] safety questions and any changes to the technical spec (Equation 16) ifications that are necessary to either conduct the ther mal annealing or operate the nuclear power reactor for base metals: 0 236 following the annealing should be identified in this sec D = (100 Cu + 9) (f + fs) - s tion. The section should demonstrate that the Com for weld metals: mission's requirements are complied with and that D = (100 Cu + 14) (f + fs) 0 "2 36 8 there is reasonable assurance of adequate protection to the public health and safety following the changes.

the upper bound: S. COMPLETION OR TERMINATION OF

D = 42.39 (f + fs)°-15°2 THERMAL ANNEAL

where 5.1 Annealing Completed, In Compliance CvUSE(u) = Charpy upper-shelf energy of If the thermal annealing was completed in accor the material in the preservice dance with the Thermal Annealing Operating Plan and

2 or unirradiated condition in the Requalification Inspection and Test Program, the ft-lb, licensee must so confirm in writing to the Director, Of Cu = the copper content (weight fice of Nuclear Reactor Regulation (NRR). Within 15 percent) for the subject days of the licensee's written confirmation that the material, thermal annealing was completed in accordance with f = the increment of best-estimate the Thermal Annealing Plan, and prior to restart, the total neutron fluence (in units NRC will (1) briefly document whether the thermal an of 1019 n/cm 2 , E > 1 MeV), nealing was performed in compliance with the licens at the clad-base metal inter ee's Thermal Annealing Operating Plan and the Re face on the inside surface qualification Inspection and Test Program, with the of the vessel, accumulated documentation to be placed in the NRC Public Docu during subsequent plant ment Room, and (2) hold a public meeting on the an operation after the annealing nealing. The purposes for the public meeting are to operation, and (1) permit the licensee to explain the results of the fs= the "shelf recovery fluence," reactor vessel annealing to the NRC and the public, evaluated from Equation 15. (2) allow the NRC to discuss its inspection of the reac tor vessel annealing, and (3) provide an opportunity For "credible" surveillance data, the values of "9" for the public to comment to the NRC on the thermal and "14" in Equations 15 and 16 are replaced by val annealing. The licensee may restart its reactor after the ues that are based on the surveillance results. meeting has been completed, unless the NRC orders otherwise. Within 45 days of the licensee's written con

3.2.2 Surveillance Method firmation that the thermal annealing was completed in For the surveillance method, the reembrittlement accordance with the Thermal Annealing Operating trend is determined from the surveillance results of a Plan and the Requalification Inspection and Test Pro program that conforms to the intent of Appendix H of gram, the NRC staff will complete full documentation

10 CFR Part 50, once the surveillance program and of the NRC's inspection of the licensee's annealing results from the program have met the credibility re process and place the documentation in the NRC Pub quirements in the PTS rule (10 CFR 50.61). lic Document Room.

1.162-17

5.2 Annealing Completed, Not in Compliance 5.3.2 Licensee Elects To Take Credit for any Recovery If the thermal annealing was completed but not If the partial annealing was otherwise performed in performed in accordance with the Thermal Annealing accordance with the Thermal Annealing Operating Operating Plan and the Requalification Inspection and Test Program, including the bounding conditions of the temperature and times, the licensee must submit a Plan and relevant portions of the Requalification In spection and Test Program and the licensee elects to K

take full or partial credit for the partial annealing, the summary of lack of compliance and a justification for licensee must confirm in writing to the Director, NRR,

subsequent operations. The licensee must also identify that the partial annealing was otherwise performed in any changes to the facility that are attributable to the compliance with the Thermal Annealing Operating noncompliances that constitute unreviewed safety Plan and relevant portions of the Requlification In questions and any changes to the technical specifica spection and Test Program. The licensee may restart tions that are required for operation as a result of the its reactor after the NRC places a summary of its in noncompliances. This identification does not relieve spection of the thermal annealing in the NRC Public the licensee from complying with applicable require Document Room and the NRC holds a public meeting ments of the Commission's regulations and the operat on the thermal annealing.

ing license; and if these requirements cannot be met as a result of the annealing operation, the licensee must 5.3.3 Termination, Not in Compliance obtain the appropriate exemption per 10 CFR 50.12. If the partial annealing was not performed in ac If unreviewed safety questions or changes to technical cordance with the Thermal Annealing Operating Plan specifications are not identified as necessary for re and relevant portions of the Requalification Inspection sumed operation, the licensee may restart after the and Test Program, the licensee is to submit a summary NRC staff places a summary of its inspection of the of lack of compliance with the Thermal Annealing Op thermal annealing in the NRC Public Document Room erating Plan and the Requalification Inspection and and the NRC holds a public meeting on the thermal Test Program and a justification for subsequent opera annealing. On the other hand, if unreviewed safety tion, to the Director of NRR. Any changes to the facil questions or changes to technical specifications are ity as described in the updated final safety analysis re identified as necessary for resumed operation, the li port that are attributable to the noncompliances and censee may restart only after the Director of NRR au constitute unreviewed safety questions, and any thorizes restart, the summary of the NRC staff inspec changes to the technical specifications that are re tion is placed in the NRC Public Document Room, and quired as a result of the noncompliances, must also be a public meeting is held on the thermal annealing. identified.

If no unreviewed safety questions or changes to

5.3 Termination Prior to Completion of Anneal technical specifications are identified, the licensee may restart its reactor after the NRC places a summary of its If the thermal annealing was terminated prior to inspection of the thermal annealing in the Public Doc completion, the licensee should immediately notify the ument Room, and the NRC holds a public meeting on NRC of the premature termination of the thermal the thermal annealing.

anneal. If any unreviewed safety questions or changes to technical specifications are identified, the licensee may

5.3.1 Licensee Elects Not To Take Credit restart its reactor only after approval is obtained from for any Recovery the Director, Office of Nuclear Reactor Regulation, the summary of the NRC staff inspection is placed in If the partial annealing was otherwise performed in the public document room, and a public meeting on accordance with the Thermal Annealing Operating the thermal annealing is held.

Plan and relevant portions of the Requalification In spection and Test Program, and the licensee does not 6. THERMAL ANNEALING RESULTS

elect to take credit for any recovery, the licensee need REPORT

not submit the Thermal Annealing Results Report de Every licensee who either completes a thermal an scribed in Section 6 of this regulatory guide but instead nealing or terminates an annealing but elects to take must confirm in writing to the Director, NRR, that the full or partial credit for the annealing must provide a partial annealing was otherwise performed in accor report that includes the results of the annealing opera dance with the Thermal Annealing Operating Plan and tion and verifies compliance with the approved plan.

relevant portions of the Requalification Inspection and This report is to be submitted within three months of Test Program. The licensee may restart its reactor after completing the thermal anneal, unless an extension is the NRC places a summary of its inspection of the ther authorized by the Director, Office of Nuclear Reactor mal annealing in the Public Document Room and the Regulation. This report should provide the following K

NRC holds a public meeting on the thermal annealing. information:

1.162-18

(1) The time and temperature profiles of the ac 6.3 Determination of Percent Recovery tual thermal annealing, The method for determining the percent recovery of RTNDT and Charpy upper-shelf energy should be

(2) The post-anneal RTNDT and Charpy upper described. The actual time-temperature parameters of shelf energy values of the reactor vessel mate the vessel annealing operation should be used and re rials for use in subsequent reactor operation, ported. If the percent recovery is determined from testing credible surveillance specimens or from testing materials removed from the beltline region of the reac

(3) The projected post-anneal reembrittlement tor vessel, and the testing was subsequent to the an trends for both RTNDT and Charpy upper nealing operation or was not reported in the Thermal shelf energy, and Annealing Operating Plan, the results of these tests should be reported. In this case, the report should include the initial unirradiated properties, the as

(4) The projected values of RTpTS and Charpy irradiated properties just prior to annealing, and the upper-shelf energy at the end of the proposed period of operation addressed in the properties of the test specimens in the irradiated and application. annealed conditions. The report should provide sup porting evidence for the licensee's report that the an nealing conditions for the test specimens were equal to

6.1 Description of the Overall Process or bounded by the annealing conditions of the reactor vessel. If the percent recovery is determined by calcu A detailed summary of the annealing operation, lation, the evaluation of percent recovery should use including an actual time-temperature history, should the actual vessel lower-bound annealing time and tem be provided. The summary should identify the location perature. In all cases, the report should include the and method of attachment for the specific post-anneal RTNDT and Charpy upper-shelf energy instrumentation, including all temperature measure values.

ment devices applied to the vessel and other structures and components. A history of the times and tempera tures for each temperature measurement device 6.4 Determination of Reembrittlement Trend should be included, showing the actual temperatures The program for determining the reembrittlement for the beginning of annealing, the heat-up period, the trend based on both the reference temperature and the

, steady-state conditions, and the cool-down period. Charpy upper-shelf energy should be described, in Sufficient detail should be included to permit determi cluding the results of any analyses or tests that establish nation of the heat-up and cool-down rates and the vari these reembrittlement trends. The program should ations in temperature measurements during the entire specifically identify the post-anneal "starting" values of cycle. A summary of key measurements should be pro reference temperature and Charpy upper-shelf energy, vided that shows that the proposed annealing condi as well as the projected embrittlement path of these tions, specifically the time and temperature parameters values with increasing neutron fluence, including the and stress allowables established in the application, basis for this projection. To the degree that this in were not exceeded. Additionally, a summary of the formation is the same as the information in the Ther worker exposures incurred during the annealing proc mal Annealing Operation Plan, the plan may be ess should be included. referenced.

6.2 Evaluation of Requalificatlon Inspections 6.5 Changes to Surveillance Program and Tests Changes to the surveillance program previously The results and evaluations of any inspections and described in the Thermal Annealing Report (10 CFR

tests used to requalify the annealed reactor vessel, the 50.66(b) (3) (ii) (B)) that are the result of the annealing attached piping or appurtenances, and the adjacent operation should be described in detail. If no changes concrete structures should be reported. The Thermal are necessary, the Thermal Annealing Results Report Annealing Results Report should include the results of should so state.

all inspections and tests to demonstrate that the an nealing operation has not caused degradation of the 6.6 Allowable Operating Period reactor vessel, the insulation, the attached piping or appurtenances, containment, and the adjacent con Based on the degree of recovery and the projected crete to a degree that could affect the safe operation of reembrittlement trend, an analysis should be provided the reactor. The report should describe the evaluation to demonstrate the period of operation for which the and disposition of any indications detected during the requirements of 10 CFR 50.60 and 10 CFR 50.61 will post-anneal inspections. be satisfied.

1.162-19

7. PUBLIC INFORMATION AND 7.2 Completion or Termination of Thermal PARTICIPATION Annealing Within 15 days after the NRC's receipt of the

7.1 Thermal Annealing Report licensee's written submittal on the completion or termination of thermal annealing as described in Sec Upon receipt of a Thermal Annealing Report, and tion 5 of this regulatory guide, the NRC staff will place K

a minimum of 30 days before the licensee starts ther in the NRC Public Document Room a summary of its mal annealing, the NRC will: inspection of the licensee's thermal annealing, and the Commission will hold a public meeting on the anneal

(1) Notify and solicit comments from local and State governments in the vicinity of the site where the ing. The purposes of this public meeting are (1) for the licensee to explain to the NRC and the public the re thermal annealing will take place and any Indian Nation or other indigenous people that have treaty or sults of the reactor pressure vessel annealing, (2) for statutory rights that could be affected by the thermal the NRC to discuss its inspection of the reactor vessel annealing, and (3) for the NRC to receive public com annealing, ments on the annealing.

(2) Publish a notice of a public meeting in the Fed eral Register and in a forum, such as local newspapers, 7.3 NRC Inspection Report which is readily accessible to individuals in the vicinity Within 45 days of NRC's receipt of a licensee's of the site, to solicit comments from the public, and submittals described in Section 5 of this regulatory guide, the NRC staff will fully document its inspection

(3) Hold a public meeting on the licensee's Ther of the licensee's annealing process and place this docu mal Annealing Report. mentation in the NRC Public Document Room.

1-.

1.162-20

REFERENCES

Vessel Steels," NUREG/CR-6327 (MCS

1. U. Potapovs, J. R. Hawthorne, and C. Z. Serpan, Jr., "Notch Ductility Properties of SM-1A Reac 950302), May 1995.3 tor Pressure Vessel Following the In-Place American Society for Testing and Materials,

9.

Annealing Operation," Nuclear Applications, "Recommended Guide for In-Service Annealing Vol. 5, No. 6, pp. 389-409, 1968.

of Water-Cooled Nuclear Reactor Vessels,"

2. A. Fabry et al., "Annealing of the BR-3 Reactor ASTM E 509-86, Philadelphia, 1986.

Pressure Vessel," in Proceedings of the Twelfth 10. NRC, "Additional Requirements for Yankee Water Reactor Safety Research Information Rowe Pressure Vessel Issues," SECY-91-333, Meeting, NUREG/CP-0058, Vol. 4, pp.

October 22, 1991.3

144-175, NRC, January 1985.1

11. NRC, "Action Plans To Implement the Lessons

3. N. M. Cole and T. Friderichs, "Report on An Learned from the Yankee Rowe Reactor Vessel nealing of the Novovoronezh Unit 3 Reactor Ves Embrittlement Issue," SECY-92-283, Au sel in the USSR," NUREG/CR-5760 (MPR gust 14, 1992.3 Associates, Inc., MPR-1230), NRC, July 1991.

12. NRC, "Standard Review Plan for the Review of

4. W. L. Server, "In-Place Thermal Annealing of Safety Analysis Reports for Nuclear Power Nuclear Reactor Pressure Vessels," NUREG/ Plants," NUREG-0800, June 1987.1 CR-4212 (EG&G Idaho, Inc., EGG-MS-6708),

NRC, April 1985. 13. American Society of Mechanical Engineers,

"Rules for Construction of Nuclear Power

5. T. R. Mager, "Feasibility of and Methodology for Plants,"Section III, Division 1, Subsection NB,

Thermal Annealing an Embrittled Reactor Ves of ASME Boiler and Pressure Vessel Code, New sel," EPRI NP-2712, Vol. 2, Electric Power York, 1993.

Research Institute, Palo Alto, CA, November

1982.2 14. American Society for Testing and Materials,

"Standard Practice for Conducting Surveillance

6. T. R. Mager et al., "Thermal Annealing of an Tests for Light-Water Cooled Nuclear Power Embrittled Reactor Vessel, Feasibility and Meth Reactors," ASTM E 185-82, Philadelphia, odology," EPRI NP-6113-SD, Electric Power 1982.

Research Institute, Palo Alto, CA, January

1989.2 15. American Society of Mechanical Engineers,

"Nuclear Power Plant Components,"Section III,

7. J. R. Hawthorne and A. L. Hiser, "Investigations Division 2, of the ASME Boiler and Pressure of Irradiation-Anneal-Reirradiation (IAR) Prop Vessel Code, New York, 1993.

erties Trends of RPV Welds-Phase 2, Final Re port," NUREG/CR-5492 (Materials Engineering 16. G. B. Reddy and D. J. Ayres, "High Associates, Inc., MEA-2088), NRC, January Temperature Elastic-Plastic and Creep Properties

1990.1 for SA533 Grade B, Class 1 and SA508 Materi als," EPRI Report NP-2763, December 1982.2

8. E. D. Eason et al., "Models for Embrittlement Recovery Due to Annealing of Reactor Pressure 17. NRC Information Notice No. 90-52, "Retention of Broken Charpy Specimens," August 14,

1990.3

18. American Society for Testing and Materials,

"Standard Test Method for Plane-Strain Frac ture Toughness of Metallic Materials," ASTM E

399-83, Philadelphia, 1983.

1Copies are available for inspection or copying for a fee from the 19. American Society for Testing and Materials, NRC Public Document Room at 2120 L Street NW., Washing ton, DC; the PDR's mailing address Is Mail Stop LL-6, Wash "Standard Guide for Reconstitution of Irradiated ington, DC 20555-0001; telephone (202)634-3273; fax Charpy Specimens," ASTM E 1253-88, Phila

(202)634-3343. Copies may be purchased at current rates delphia, 1988.

from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 (telephone (202) 512-1800) or from the National Technical Information Service by writing 3 Coples are available for Inspection or copying for a fee from the NTIS at 5285 Port Royal Road, Springfield, VA 22161. NRC Public Document Room at 2120 L Street NW., Washing

> 2 Copies may be purchased from EPRI's Research Reports Cen ton, DC. the PDR's mailing address is Mail Stop LL-6. Wash Ington, DC 20555-0001; telephone (202)634-3273; fax ter, P.O. Box 50490, Palo Alto, CA 94303 (telephone (202)634-3343.

(415)965-4081).

1.162-21

20. American Society for Testing and Materials, Materials," Regulatory Guide 1.99, Revision 2,

"Standard Test Methods for Notched Bar Impact May 1988.4 Testing of Metallic Materials," ASTM E 23-88, Philadelphia, 1988. 22. R. D. Cheverton et al., "Review of Reactor Pres sure Vessel Evaluation Report for Yankee Rowe

21. U.S. Nuclear Regulatory Commission, Nuclear Power Station (YAEC No. 1735),"

NUREG/CR-5799 (ORNL/TM-11982), NRC,

K

"Radiation Embrittlement of Reactor Vessel March 1992.1

4 Single copies of regulatory guides may be obtained free of charge by writing the Office of Administration, Attn: Distribution and Services Section, USNRC, Washington, DC 20555-0001, or by fax at (301)415-2260. Copies are also available for inspection or copying for a fee from the NRC Public Document Room at 2120 L Street NW., Washington, DC; the PDR's mailing address is Mail K

Stop LL-6, Washington, DC 20555-0001; telephone (202)634-3273; fax (202)634-3343.

1.162-22

REGULATORY ANALYSIS

A separate regulatory analysis was not prepared the guide. A copy of the regulatory analysis is available for inspection and copying for a fee at the NRC Public for this regulatory guide. The regulatory analysis prepared for 10 CFR 50.66, "Requirements for Ther Document Room, 2120 L Street NW, Washing mal Annealing of the Reactor Pressure Vessel," pro ton, DC; the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555-0001; phone vides the regulatory basis for this guide and examines

(202)634-3273; fax (202)634-3343.

the costs and benefits of the rule as implemented by

1.162-23

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