ML18065A621
| ML18065A621 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/03/1996 |
| From: | Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.147, RTR-REGGD-1.147 NUDOCS 9604120071 | |
| Download: ML18065A621 (12) | |
Text
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consumers Power -
POW ERi Nii MICHlliAN'S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 April 3, 1996 U S Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT REQUEST FOR APPROVAL TO USE CODE CASE N-557, "IN-PLACE DRY ANNEALING OF A PWR NUCLEAR REACTOR VESSEL (SECTION XI, DIVISION 1 )"
In accordance with 1 O CFR 50.55a (a)(3), NRC approval is requested for use of ASME Code Case N-557, "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1 )." This approval is requested until Code Case N-557 is included in Regulatory Guide 1.147, lnservice Inspection Code Case Acceptability -ASME Section XI Division 1. Implementation of this code case, when approved by the NRC staff, is expected during the 1998 Refueling Outage.
ASME Code Case N-557 permits the in-place thermal anneal under the rules of the ASME B&PV Code Section XI, and provides the requirements for the anneal of a nuclear pressure vessel in a commercial nuclear power plant.
Pursuant to 1 O CFR 50.55a (a)(3), the justification in Attachment 1 shows that implementation of Code Case N-557 will result in maintaining the structural integrity of the reactor coolant pressure boundary, and will provide an acceptable level of quality and safety. Attachment 2 provides a pre-publication copy of ASME Code Case N-557.
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I I
9604120071 960403' 'I j
PDR ADOCK 05000255 :.
G PDR
- I A CMS' ENERGY COMPANY
SUMMARY
OF COMMITMENTS This letter contains no new commitments and no revisions to existing commitments.
Richard W Smedley Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades
- Attachments
ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 REQUEST FOR USAGE APPROVAL OF CODE CASE N-557 "IN-PLACE DRY ANNEALING OF A PWR NUCLEAR REACTOR VESSEL (SECTION XI, DIVISION 1 )"
JUSTIFICATION FOR USE OF ASME CODE CASE N-557 5 Pages
Code:
Subject:
JUSTIFICATION FOR USE OF ASME CODE CASE N-557 Class 1 "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1 )"
Component: Reactor Vessel and Associated Primary Coolant System Piping.
Code Reguirement The original design, fabrication, and construction of the Palisades reactor vessel was in accordance with the ASME Boiler and Pressure Vessel Code, Section Ill, 1965 edition, including all addenda through Winter 1965 (ASME B&PV Code, Section Ill, 1965, W65A). Replacement parts, components and repairs satisfy the* requirements of the original construction code and the requirements specified in ASME B&PV Code,Section XI "Rules tor lnservice Inspection of Nuclear Power Plant Components".
While annealing a reactor vessel is similar to a repair activity, the current definition of a repair in Section XI does not include annealing. As a result, editions of Section XI through 1995 do not address annealing. This Code Case establishes the requirements which permit an in-place thermal anneal under the rules of the ASME B&PV Code Section XI applicable to operating PWR nuclear power plants.
Basis for Justification:
Pursuant to 1 O CFR 50.55a(a)(3) and footnote 6 to this section, NRC authorization is requested to utilize Code Case N-557 "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1 )." The basis for this request is that the requirements of the code case provide an acceptable level of quality and safety to ensure the post annealing structural integrity of the reactor coolant system pressure boundary, per 10 CFR 50.55a(a)(3)(i).
In an operating nuclear power plant, the high energy neutrons emitted by the nuclear reaction gradually decrease the material toughness and ductility of the beltline region material of the ferritic steel reactor vessel. The rate of damage is* a function of the radiation dose and of certain trace elements in the vessel welds and base metal.
Federal regulations require a minimum level of material toughness which is monitored by a surveillance program (10 CFR Part 50, Appendix H). These material property degradations can be reversed by performing a thermal anneal.
1
Thermal annealing is the process of heating the reactor vessel beltline (the region enveloping the active fuel length) to a temperature well above the operating temperature of the reactor for an extended period of time. This process will remove the microstructural changes caused by radiation and restore the vessel material's toughness. It is the only known method for restoring toughness properties to materials degraded by neutron radiation.
The Russians have successfully annealed 12 WER-440 PWRs at temperatures of approximately 850°F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. Dry thermal annealing of the Palisades reactor vessel will be the first application of this process to a commercial reactor in the United States. The annealing time and temperature to be used at Palisades will be very similar to those successfully used by the Russians. A significant difference between the Russian anneal and the Palisades anneal is the zone being heated. The Palisades reactor vessel will be heated along its entire active fuel length while the Russiar;i reactor was only heated over approximately a 5 foot s~ction in the center of the active fuel length. The major impact of this difference is that the magnitude of the stresses in the nozzle region of the reactor will be greater at Palisades since the resultant thermal gradients in this region will be higher; while the stresses in the vessel shell are expected to be much lower at Palisades due to a more uniform temperature distribution in the shell region of the vessel. A detailed transient thermal stress analysis will be performed for the Palisades reactor to insure all stresses are within the guidelines
- established by the Code Case.
Editions of Section XI *through 1995 do not address annealing. Annealing a reactor vessel to restore material properties reduced by irradiation is similar to a repair activity, such as a post weld heat treatment. However, the current definition of repair in Section XI does not include annealing. This Code Case provides the rules and requirements to perform an in-place dry anneal of a nuclear reactor vessel in a commercial nuclear power plant and permits an in-place thermal anneal under the rules of the ASME B&PV Code Section XI. The focus of this Code Case is the structural integrity of the reactor coolant pressure boundary. Restoration of material properties is addressed in NUREG/CR 6327 "Models for Embrittlement Recovery Due to Annealing of Reactor Pressure Vessel Steels" and RG 1.162 "Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels.. "
- 2
An in-place anneal is analogous to the local post weld heat treatment (PWHT) the vessel has received as a construction activity and a similar philosophy should control.
The similarities and differences between an anneal and a PWHT are discussed below.
Similarities:
(1)
The anneal is a non-operational event involving only a thermal treatment of an open top reactor vessel and, as such, there are essentially no primary loadings.
The integrity of the component is not affected by the stresses and strains during the anneal any more than it is during a fabrication PWHT.
(2)
The stresses during the anneal, just as with a PWHT, have no fatigue implications because this is a one time construction type of activity and there are no significant primary loadings.
(3)
Residual stresses due to differential thermal expansion between base metal and clad will be present after annealing.
Differences:
(1)
The metal temperatures in the annealing zone during the anneal are substantially lower than those for a PWHT while the hold time at temperature is substantially longer." Despite these differences in temperature and *hold time, annealing is analogous to a low temperature PWHT in which time and temperature limits during anneal are set such that no significant time dependent deformations (creep) will occur.
(2)
Insulation design and heat input distribution are used during a fabrication PWHT to control the axial thermal gradients at the edges of the heated band, as required by NB-4624.3. For an in-place anneal, the operational reactor vessel insulation will be utilized and the axial temperature gradients will be controlled by an axially varying heat input at the edges of the annealing zone.
(3)
For most local PWHT's during fabrication, the component is an ihdependent entity and effects on other items/structures are not a factor. For an in-place anneal of a reactor vessel, effects on surrounding structures and attached components must be considered as well as reassembly fits and tolerances.
Previous studies of in-place annealing have shown that the highest stresses as a result of the anneal occur at the primary piping - nozzle interface. The Code Case uses conservative limits based on Section Ill philosophy for these stresses to assure that damage does not occur. Preanneal and postanneal fits and tolerances will be measured to insure that no unacceptable deformations take place during the annealing process.
3
Based on these similarities and resolution of differences discussed above, the in-place anneal of a reactor vessel is considered a local heat treatment process. Therefore, the IWA-4000 requirements for a repair activity are appropriate to provide rules consistent with Section XI goals and principles. Application of these requirements was considered in Code Case N-557. These requirements are similar to presently authorized repairs and will provide an equivalent level of quality and safety.
Palisades has established administrative limits and annealing conditions based on the results of thermal and stress analyses conducted at annealing conditions. Adherence to the limits and conditions will insure that the limiting parameters addressed in the Code Case.are not exceeded. The details of these limits and conditions are presented in the preliminary Palisades Thermal Annealing Report (TAR) which has been submitted to the NRC for approval.
The final Palisades TAR will be submitted after completion of the Marble Hill Annealing Demonstration Project. The TAR discusses compliance to all qualitative requirements of the Code Case in the context required by RG 1.162. The following discussion concentrates on the quantitative requirements of the Code Case which are not specifically specified in RG 1.162.
The limiting parameters addressed in Code Case N-557 take into consideration the material, temperature, stress, and strain requirements needed to assure an effective and acceptable anneal of the Palisades reactor vessel. These limiting parameters are classified into three categories: time at temperature, temperature, and stress.
Time at Temperature Limits The limiting parameters associated with time at temperature are to satisfy the material property recovery requirements, and to limit the potential for creep and other forms of elevated temperature metallurgical degradation.
Code Case N-557 indicates that for metal temperatures exceeding 900°F but less than 940°F and with less than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> of hold time vessel steels will not undergo significant time dependent behavior, such as creep. Similarly for temperatures greater than 850°F but less than 900°F, a hold time less than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> will provide the same assurance. The Palisades reactor vessel anneal will be conducted at a temperature range of 800-900°F for a 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> hold time. These temperatures and times are well within the limits addressed in the code case.
4
Temperature Limits The limiting parameters associated with temperature are to ensure that degradation of specific equipment, components, or structures does not occur.
Code Case N-557 specifies a maximum temperature of 940°F to ensure that reactor vessel material would not be subject to significant time dependent material degradation due to annealing temperatures. The maximum allowed annealing temperature at Palisades will not exceed 940°F.
Stress Limits The limiting parameters associated with induced stresses caused by the annealing process are intended to avoid harmful permanent set and the potential for ductile flaw growth.
ASME Code Case N-557 requires that mechanical and thermal loadings on the reactor coolant pressure boundary and its supports be evaluated and suggests criteria and a methodology to perform the evaluation. This guidance specifies the loadings which require evaluation and the associated stress categories and limits of stress intensity.
Allowable stress intensity values are provided for SA-302 Grade B(Nickel Modified) plate with 0.4 to 0. 7 % *Ni which includes the Palisades reactor vessel plate material
- and SA508 Grade 2 Class 1 forgings which is the Palisades reactor vessel nozzle material.
The Code Case primary plus secondary stress limit of 3 Sm establishes the maximum allowed stress in pressure boundary components. This stress limit will prevent component damage resulting from the annealing operation. Palisades will fully comply with the guidance provided in the Co~e Case.
Conclusions ASME B&PV Code Section XI does not presently address annealing of a reactor pressure vessel.
Code Case N-557 provides the requirements to perform an in-place dry anneal of a reactor pressure vessel under the rules of the ASME B&PV Code Section XI applicable to operating nuclear power plants. These rules and requirements will be rigidly implemented as outlined in the Palisades Thermal Annealing Report. Therefore, application of the Code Case will insure that the anneal of the Palisades reactor vessel will be conducted in a manner which will provide an acceptable level of quality and
- safety to insure the post annealing structural integrity of the reactor coolant system pressur~ boundary.
5
. ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 CODE CASE N-557 "IN-PLACE DRY ANNEALING OF A PWR NUCLEAR REACTOR VESSEL (SECTION XI, DIVISION 1 )"
3 Pages
Case* N-557 (ISi #95-43, N/D 95-14, BC 95-454)
IN-PLACE DRY ANNEALING OF A PWR NUCLEAR REACTOR VESSEL (Section XI, Division 1)
Inquiry: What requirements may be used for an in-place dry anneal1 of a PWR reactor vessel?
Reply: It is the opinion of the Committee that an in-place dry anneal of an PWR reactor vessel may be conducted using the following requirements.
1.0 The applicable requirements of the 1995 Edition, IWA-4130 through 4170 and IWA-4910 for a repair shall be met.
- 2. O The vessel temperature shall not exceed 940 F. The vessel temperature shall not exceed 900 F for more than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, nor 850 F for more than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />.
3.0 The region of the vessel to be annealed, the annealing temperature, the hold time at temperature, and the heating and cooling rates shali be defined by the Owner in the Repair/Replacement Plan. The region to be annealed shall be a circumferential band.
4.0 The temperature of the reactor vessel extending upward and downward from the edges of the heated region shall be gradually diminished to avoid harmful thermal gradients.
5.0 Temperature shall be monitored by thermocouples (or equivalent) in direct contact with the vessel wall. Thermocouples (or equivalent) shall be placed to adequately control and monitor temperatures, rates and gradients as specified in the Repair/Replacement Plan.
- Time-temperature recordings of the annealing treatment shall be made available for review by the Inspector.
6.0 The Owner shall determine and document what mechanical and thermal loadings on the reactor coolant pressure boundary components and their supports are to be evaluated.
The purpose of the evaluation is to preclude harmful permanent set and the potential for ductile flaw growth. One way of accomplishing this is by use of the criteria and methodology in 6.1 and 6.2, which is only guidance for the evaluation. However, compliance with this guidance will eliminate the need for any further evaluation. For the upper ranges of time and temperature in 2.0, Code Case N-499-1 provides further guidance.
[1] A heat treatment conducted to improve beltline material toughness reduced by neutron irradiation Page 1 of 3
e Case 'N-557 {ISi #95-43, N/D 95-14, BC 95-454)
IN-PLACE\\DRY ANNEALING OF A PWR NUCLEAR REACTOR VESSEL {Section XI, Division 1) 6.1 The reactor vessel loadings that require evaluation shall be evaluated using the stress categories and limits of stress intensities in Fig. 1. For components other than the reactor vessel, and for supports, evaluation of these loadings shall be performed in accordance with the Construction Code.
6.2 Table 1 provides allowable stress intensity values for reactor vessels constructed of any combination of the following materials:
SA-533 Type B Class 1 plate [formerly SA-533 Grade B Class 1 plate or SA-302 Grade B (Nickel Modified) plate with nickel content in the.range 0.40 to 0.70 weight percent]
SA-508 Grade 2 Class 1 forgings (formerly SA-508 Class 2 forgings)
SA-508 Grade 3 Class 1 forgings (formerly SA-508 Class 3 forgings) 7.0 The Owner shall specify, in the Repair/Replacement Plan, any deformation limits to be satisfied.
8.0 Post-anneal volumetric examination, consisting of all of the examinations required during the current inspection interval; shall be performed on that portion of the vessel where the temperature exceeded 700 F. The acceptance criteria, type and number of examinations,
- qualification requirements, and reporting requirements shall be those of the Owner's
- lnservice Inspection Program.
9.0 Use of this Case sh~ll be documented on an NIS-2 form.
Code Case applicability:
197 4 Edition with the Summer 1975 Addenda through the 1995 Edition with the 1995 Addenda.
Table 1 Design Stress Intensity Values, Sm, for Use in Annealing Activity Evaluation.
Temp., F 70 700 750 800 850 900 940 Sm, ksi 26.7 26.7 26.7 26.7 25.5 24.0 22.5 Page 2 of 3
.... ~
~----------------------------*
Stress Categoiy Plirrary secondary
- ------- --*-- **-*---- ------------*----* - --*--------- ---*----*------*-------~------
General Merrbrone
,Local Membrane Bending Expansion Membrane plus Bending Peale and Fatigue Description Symbol (Nole (2))
Combination ol stress cOfl'lXX)ents and allowable Units ol stress Intensities Legend Average prlrrary stress oaoss solld section. Exel~
enects*or dlscon-tlnuttles and con-centrations. Pro-duced bV mechanical loads.
Pm Average stress across any solid section.
Considers enects or dlsconllrnitles bUt not concentrations.
PrOducedb'{
mechanical loads.
Component or prtmary stress proportional lo distance lrom centroid ot solid secllon. Excludes enects ot disconti-nuities and concen-trations. PrOduced bV mechanical loads.
(Note(l))
Stresses which resun from the constraint of tree end displace-ment. Considers enecls ot dlscontl-nunles bUI not local stress concentration (not oppllcoble to vessels).
Pe
~
Sell-equlllblollng stress necessary to sollsty continuity ot structure. Occurs at structural dlsconll-nuttles. Con be caused bV mechanical loads or dlnerenllcil thermal expansion. Excludes local stress concentrations.
Q Evaluation not required F
0 AlloWoble value PL +Pb + Pe+....... --8
[t"'1* [6JI.,___;
I Calculated vo1ue
~--~
NOTES:
Loadings present during anneal activity PL+ Pb (l J Bending component ot primary stress for piping Sholl be the stress proportional to the distance from centroid or pipe cross section.
(2) lhe S',fTlbolS Pm. I\\. Pa. Pe. Q. and F dO not repiesent si'igle quantttles. bUI sets ol six quantnies representing the six stress components Ol; 01, or, tr, t/, and tr.
(3) The value ol Sm shalt be determined from Tobie l using the concurrent temperature at the point being evaluated.
(4) The structural strength of the clodding may be neglected.
(5) When other Items ore attached to the reactor vessel In the high temperature region (T> 700 F). these Items shall be evaluated for their enect on the pressure boundary.
(6) The stresses In colegOI'{ Q ore thOse ports ot the total stress that ore produced bV thermal gradients. structural dlsconllmJtles. etc.. and they do not Include prllT'Ory stresses that may also exist al the some point. However. It should be noted that a detailed stress analysis frequently gives the comblnoilon or primary and secondary stre~s directly, aqd, when appropriate. the calculated voiue represents the total of Pm + Pb + Q, and not Q alone. *
( 7) The special stress llmlls or NB-322 7.5, Nozzle Piping Transition, tor stresses resulting rrom the omeol activity. shalt be met.
FIG. I STRl:SS CAlEGOR.IE.'1 AND LIMITS OF S'IRLSS lNTENSllY FOR ANNEALING E.Vl\\LLll\\TION
- ~...... _..
C<Xlv CO$(; N*XXX (ISi #<)!).4:l. N/D ')b. i.~J I'< 1Ut: 3 or 3