ML20055E589

Forwards Request for Addl Info Re BWR Vessel Internals License Renewal Industry Rept

ML20055E589
Person / Time
Issue date:	07/06/1990
From:	Jocelyn Craig Office of Nuclear Reactor Regulation
To:	Griffing E NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT &
References
TAC-76161, NUDOCS 9007120134
Download: ML20055E589 (12)

Text

{{#Wiki_filter:' G-July 6, 1990 i Mr. Edward P. Griffing lianager, Technical Division Nucletr Management and Resources Council 1770 Eye Street, N.W. Suite 300 Washington, D.C. 20006-2496

Dear tir.,

Criffing:

SUBJECT:

EWR VESSEL INTERNALS - INDUSTRY REPORT (TAC No. 76161) By letter dated February 23, 1990, ilVMARC forwarded a topical report entitled, " Boiling Water Reactor Vessel Internals License Renewal Industry Report," for NRC review and approval. The staff has performed a review of the subject report and has determined that substantial information needs to be provided in order for the staff to complete its review. In general, our concerns focus on the lack of detail contained in the report and the weaknesses in the techacal bases provided to justify '.:onclusions made in the report. Specifically, more detail is needed to justify the determinations that a particular component is or is not safety significant, that. a particular mode of degradation is or is not significant for a particular component, and to support the proposed methods for dealing with degradation in safety significant components. Our specific request for additional information is enclosed. We will be happy to meet with you to discuss these issues. We agree with NUMARC that having approved industry technical reports are important to defining the scope of issues, review and corrective measures in tiie technical area for each lead plant. Sinccrely, /s/ P. T. Kuo for John W. Craig, Director License Renewal Project tirectorate Division of Reactor Projects - III, 14 IV, V and Special Projects 90 Office of Nuclear Reactor Regulation

Enclosure:

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. (,f - .ii UNITED STATES + g#.:, /,;g / NUCLEAR REGULATORY COMMISSION W A$HINGTON, D. C. 20555 ,r \\...o f' July 6,1990 Mr. Edward P. Griffing. lianager, Technicti Division I;uclecr ManageLent and Resources Council 1776 Lye Street, I;.W. Suite 300 Washington, D.C. E00064490 Letr fir. Griffing:

SUBJECT:

BWR YESSEL lliTERt:ALS - IliDUSTRY REPORT (TAC lio. 76101) l E.y letter dated Februcry 23, 1990, I;UMARC forwtrded a topict1 report entitled, " Boiling Kctcr Reactor Vessel Internals License Renewal Inoustry Report," for i;CC review and approvol. The staff has perfori.ied a review of 16 subject report and hcs determined that substtr.tial inforn.ation needs to be provided in order for the staff to con.plete its review. In general, our concerns focus cn tFe lack of detcil contained in the report and the weaknesses in the technical bases provided to justify conclusions made in the report. Specifically, more detail is needed to justify the determinations that a particular. component is or is not sLfety significant, that a particular noode of degradction is cr is not significant for a particular component, and to support the proposed methods for dealing with degradation in sLfety significant ccmponents. Our specific request for additions 1 inforn.ation is enclosed. We will be happy to meet with you to discuss these issues. We agree with 14UMARC that having approved industry technical reports are important to defining the scope of issues, review and corrective measures in the technical-erea for ecch lead plant. Sincerely, l) [ M.W / Ot,,,7. b. Craig, Director License Renewal Project irectorate Division of Reactor Proj cts - Ill, IV, Y and Special Projects Office of fluclear Reactor Regulation

Enclosure:

As stated l L l ~

~ i. = o-t - REQUEST FOR ALDITIONAL lt:f0RMAT10h C0liCERtilt<G o BWR VESSEL lhTERilALS LICE!!SE REl;EWAL REPORT r 1: 1. General Coments 1. Chapter 3 of the report is full of statements such as "these types of' events are not anticipated," " repairs could be implemented before plant-safet) is affected," and " analyses have shown" which are used to justify .the classification of a particular component as not being safety significant. Although the assessment of components may be correct, little or no docunentation is provided. Please provide tore detail and supporting evidence to justify your conclusions. E. Chapters 5 and 6 have sin.ilar prot <1 ems as Chapter 3. More specifically, [ Sections 5.3.2, 5.3.3, 5.3.4, 5.3.5, 5.3.6,.and 5.3.8 refer to evaluations b, for which no details and no references are given. There is heavy reliance on the detection-of degradation by inservice inspections (whether volumetric or visual), however, the report does not discuss or reference the specific nature of the inspections required or the adequacy of inservice inspection to detect specific degradation mechanisms. In Chapter 6, various " aging degradetion management" strategies are listed. Again, the details of these strategies and how they would be applico or ~ =- appropriate references are completely lacking. Often the choice of an appropriate strategy for a particular component is dismissed as a " plant unique decision. L 3-tow temperature s'asitization (LTS) of 304 types of stainless steel is not discussed in this report. LTS could occur either in the base I L material or in heat affected zones. LTS could render the components susceptible to IGSCC. Provide thc justification for not considering this i aging effect in this report. 4. The report recognizes that environmental effects are not explicitly covered in the ASi1E Code and that time to crack initiation is dependent on the loading spectrum. However, when the report addresses fatigue l'ife evaluation, it reverts to using the Cs de procedures. Research results -have shown that the boiling water reactor (BWR) environment significantly decreases fatigue ' life and in some cases the margin iri the Code curve.is-A completely elin.inated. Further, simple linear addition of the effects of individual cyc~les is used for estimating fatigue usage without regard to . loading spectrum. The loading spectrum can have an even more significant effect on time to crack initiation under the environmental conditions of a BWR. The report should specify exactly how environmental condition effects on fatigue crack initiation and growth will be considered to demonstrate an acceptable fatigue life for the renewal period for internals base material end weldments. i -[. k k jp, . ". -{"i j ; ' ' ' ' fvi 'i-~'

-c. L. In EPLI I;P-51Ellt, "EWL Pilot Plant Life Extension St".,y at the Monticello Plent: Phase 1" and LPRI I;P-5830M, "BUE Pilot FD..t Life Exter.sion Study at the Monticello Plant: Interir.. Pha se 2". The various degradation prctlens that CWF vessel internals have experienced are discussed. More iroportantly, the reports discuss the need for such things as rebuilding 10 percent of the CRDs each outage, nothods to UT shroud-to-shroud support cy linder welds and core spray inlet tee attachnrots, methods to UT top guide -in central core region for IASCC, underweter 151 cr. core plate for IASCC, replacement of jet pump assemblies following 30 years of plant operatior, methods to perform UT of jet pump riser elbow to thermal siteve weld. The NUMARC report appehrs to ignore the published it. formation because the replacerient of vessel internals, their removal for inspection and 151 inspection procedures is not discussed in depth. Instead, the industry report enphasizes justification for not having to do a detailed examination and replacement of the vessel internals. Under the PLEX follow-or, activities in EPRI lip 5181M, reactor pressure vessel it,ternals (RPyl) reconmendation #4 stetes that at least one jet pump inlet-roixer subassembly should be disassembled, dye peretrant tested, and examined for erosion effects following 30 and 40 years of plant operation. Similarly RPV1 reconnendation 47 identifies the need to develop UT inspection techniques as soon as practical for the jet pump riser elbow to thermal sleeve weld region and to develop pot.ntial rtpair techniques, Many of the RPV1 recor:r.endations are not mentioned in the industry report. Please describe how these recommendations are addiessec in the report or provide the technical basis for why these recommendations need not be included in this report. 6. In Sectiori 1.1, seven components were categorized as "not safety sigr'ificarit" such that further evaluation of these components was not required. The analysis to denionstrate that the f ailure.cf such equipment would not compromise the performance of safety related equipment during extended life operation is generally inadequate. For exauplc, the report asserted (Section 3.3.1) that a steam dryer f ailare would r ot have adverse saiety consequences because in the past there have been no loose parts resulting from cracking (even though cracling occurred). The report did not address what the consequence to safety would be if dryer f ailure actually occurred. The same comments apply to all of the other seven components, including the shroud head, neutron source holder, etc. 7. A resiew of vendor / supplier /manufactu"er's recommendations useful to undetstanding und managing aging (i.e., GE SILs) was lackin; and should be provided, along with an analysis of their relevance to extended life operation. 8. The vessel internals interfaces were not well established. The welds that attach the various internals to the reactor vessel are an example of an interf ace that may need to be inspected.

L.. T.. 9.- _ Age related degradation is a common mode failure mechanism. Thus, where raultiple cotr.porients are involved, the management of aging for extended life operation should consider the increased likelihood that multialt.- (simultereous) failures will occur. This should be addressed in t,e case of e.g., control blades, shroud head bolts, jet pump sensing lines,_etc. 10.- The report did not indicate the need for benchmark or base-line levels of parameter (s) or indicator (s) for recordkeeping and trending. Th'e report should discuss the need for trending of degradation data taken periodically in order to evaluate extended life operability. 11, Where criteria for carrying out corrective maintenance are cited.it is vague as to how they are to be determined, and how accurate they will be in deterraining whether corrective maintenance is needed.

10. The report indicates that in some cases more data'will be needed to determine the state of degradation. The report should also indicate, where existing inspection or monitoring methods do not exist or are inadequate and what specific itcensees must do to improve the quality of the data or inspection methods.

13. An in-depth review of applicable NRC Bulletins, NUREGs, and GE Service Information Letters dealing with reactor internals (Table 3-1) would be useful. This shculd include a summary of their contents and applicability, and a description of the problems addressed, their gencric applicability, the proposed or required fixes, and, most importantly, the implications with regard to extended life.

14. For_ utilities referencing the ITR, the selection of a strategy to manage aging should not be icft to the individual utilities. Specific guidelines for the' utilities should be provided wherever.possible.

For example, Section 1.5 entitled "Agirg Manageraent Programs" identifies issues for which aging degradation cannot be shown to be within established bounding limits and suggests that plant specific approaches are needed to monitor such aging. However, no plant specific approach is identified.

15. The references presented were not readily ava'lable to the reviewers, and as such, the adequacy of the references in term; of completeness and correctness could not be established. The references should be made available for the NRC staff to review.

16. - The subject eport did not discuss the need to monitor or manage the ' potential degradation of the reactor vessel interior attachment welds. The attachment welds may be safety significant even though the associated internal components are not safety significant because cracks initiated from those welds have the potential of propagating into the reactor vessel material. Therefore, it is necessary to inspect those welds to ensure the integrity of the reactor vessel. Recent IGSCC piping inspection performed in several domer. tic and foreign BWR plants reported that cracks originated from the Inconel 182 butters were found to propagate into the nozzle materials. In a routine inservice inspection of a reactor vessel head-of a domestic BWR plant, cracks

r. -4 origir ated in the cledding were found to propagate a short distance ti.Lo the rcactor vessel materici. In view cf the reported field experiences, we believe that augmented inspection should be perfornied to confirm the integrity of-those interior attachment welds, We.have found attachment welds in the fo110 win 5 reactor vessel internal s conponents: Steam dryer holddown brackets and support brackets . Steam separator guide rod brackets Shroud support to bottom head Jet jump upper and lower braces Sample holder brackets (6) Feedwcter sparger brackets -(7 ) Core spray header brackets, sparger piping and nozzle mounting brackets and end brackets. The above list racy not be complete because it will vary dt, < ding on the design of a specific plant. Therefore, the licensees should identify all internal attachment welds in their reactor vessels and evaluate their susceptibility to IGSCC. An aging management program for those attachment welds susceptible to IGSCC should be proposed for MRC review and approval.

17. The' process used to evaluate the components'(Figure 1-1) assumes that.if aging (of a specific component is adequately managed during its designi.e., 4 life

)_ operation. This assumption is questionable. The issue is of concern c.g., to the IEEE and ASME Codes and standards organizations. The report should discuss this issue, particularly with reference to current activities and documents of the ASME B&PV and 0&M Code-committees. In Figure 1-1, there should be a box indicated where the current aging management plans will be evaluated as to adequacy for license renewal. -18. A review of the current ASME activities with regard to the relevance of Section XI to extended life operation should be included and any implications discussed. -II. Specific Coments 1. In Section 1.E "No Significant Degradation - Generic" certai,n processes / mechanisms in this group are stated to not cause significant degradation of any BWF, vessel internal component. This is an area where it raust be readily demonstrated that significant data exists for the full range of conditions that are expected during the original licensing period. E. In Section 1.1 and Section 3.1 several componerts (steam dryer, steam separator, shroud head, feedwater sparger, steam dryer support ring, neutron source holders and jet pump sensing line) are identified as not safety significant. This is based on the consideration that they do not-perform an essential safety function. The subject report did not discuss whether the failure of such components would affect the safety function of other safety significant components and inappropriately concluded without substantiation that no loose parts would be generated in the future or e

4, during extended life. Furthermore, the report should also address the safety impact of loose parts on plant operation as a result'of complete failure of such components. 3. In Section 5.3.11, the CRD housing is cons to have satisfactory plant programs in_ nanaging the aging degracom..i by fatigue and stress corrosion. cracking. This determination is based mainly on.the considerati_on that the ASME Section XI inservice inspection programs would detect any significant degradation. However, it is also stated in Section 5.3.11 that exen.ptions to these requirements are permitted under-IWB-1E20 of Section_XI. The subject report did not discuss the details of the exemptions nor explain why the existing inservice inspection programs and assessment procedurcs are considered adequate. 4. The subject report identified nine components (core spray sparger, access hole cover, SRM/1RM dry tubes, control blades, shroud head bolts, jet pump, CRD housings and orificed fuel support) as having satisfactory aging management programs. However, the information provided to justify this conclusion is very limit (d with no plant-specific details. Therefore, we request that each licensee should submit their, programs for NRC review to ensure that their inspection programs are adequate and effective and that their safety evaluation, flaw evaluation procedures, and the mitigction, repair and refurbishment programs are acceptable to NRC-staff. 5. The subject report indicates that inspection, monitoring, water chemistry controls and refurbishment are_ effective aging management programs for stress corrosion cracking of core spray line internal piping, top guide, shroud and core plates. These are independently acceptable methods, inspection must be included in all programs to determine whether monitoring, water chemistry controls. or refurbishment are effective. We request that inspection be performed for all_IGSCC susceptible portions of those components such as creviced and sensitized areas and areas with .high tensile residual stresses. For areas not accessible for such inspection, alternative examinations should be proposed and approved by the staff as part 4 this report review. 6. In Section 5.1.2, it is indicated that the slip dissolution model can be used to predict crack growth. We cannot make a meaningful evaluation regarding the acceptability of this model because detailed information is not provided..Therefore, the use of this model to calculate crack growth is not-currently approved and would require the submittal of additional information for NRC staff review and approval. 7. Section 3.0 of this report identifies those BWR reactor internal components which.are safety-related. Section 4.1.2 states that design calculations, startup test measurements and field experience confirm that fatigue is not significant for those safety-related components listed in Table 4-1, and therefore, the report concludes that fatigue is not'a significant aging degradation mechanism for these components and that these components need not be evaluated further for purposes of license renewal. The staff cannot reach this same conclusion without more detailed information relative to the bases for the report conclusions.

7jp ~. ir + y As an example, if.a licensee applying for license renewal is allowed to reference this Industry Report without a plant-specific review by. the staff, then this report should contain a sumr.ary of the results of:the ~ four steps requested Lelow (or equivalent evaluations) for all of the a safety-related reactor internal components identified in Section 3.0. A.- Using data 'from sources such as prototype plant preoperational= flow induced vibration tests of reactor internals (Regulatory Guide -1.20), applicable model tests and actual plant operating histories, - estiriate an enveloping nunter of stress alternating cycles which have been or will be irnposed or, these components up to the time of license renewal application.. This estimate should include. actual; plant transients and an acccunting for partial transients, if applicable. B. Using (1) the data from 1'above, (2) enveloping alternating stress intensity values associated with applicable cycles for each component, and (3) methodology based on ASME Section.Ill, KB'3222.4 or HG 3222.4, calculate the enveloping cumulative usage factors for each con.ponent up to the time of license renewal application. These analyses should be based on the current Design Fatigue Curves in ASME Section Ill, Appendix 1. C. Using a conservative extrapolation of the data from 1 above, and the methodology from 2 above, calculate the enveloping curnulative usage factors for the projected 60 year lifetime of the plant. To justify the conclusions discussed in Section 4.1.2,-the enveloping cumulative usage factors ?or the components for which fatigue is not significant should all be below the ASME Section 111 allowable of 1.0. However, if the cumulative usage factor for any component is 0.7 or higher, an aging management program should be proposed to monitor any potential age-related degradation due to fatigue. D. This report should require that each license for license renewal verify that the cumulative usage factors for its reactor internals are enveloped by those froni above. Using the above example or comparable procedures, revis.e Section 4.1.2 to provide a more detailed basis for the conclusions reached therein. 8. Section 3.1.14 classifies the jet pump sensing line as a non-nuclear safety component. However, in Section 5.3.8, the basis for not recommending an aging management program to detect possible fatigue degradation in the. jet pump is that the jet pump sensing lines would be able to detect such degradation. Since the sensing line is to be used in this manner, then the following recuirem nts should be included in the report: A. Provide a requiremeret in t'is ren rt that a licensee applying for license renewal must confom N,ne of the three methods of jet pump operability surveillance which are recommended in NUREG/CR 3052, " Closeout of IE Bulletin 80-07: BWR Jet Pump Assembly failure," dated Hovember 1984. These methods are: i

n .3 7-7 c a.- IE Etlletin 80-07, Item B.2.b. b. Technical Specification, Jet Pumps Surveillance P,equirement, Type 3 (Types 1 and 2 are unacceptable for.this purpose). General Electric Service Information Letter (330, " Jet Punp c. Beara Cracks," dated June 9,1980. B. Although the sensins line itself is a non-nuclear safety component, .it is being relied upon to detect possible degradation of a .sefety-related component.- Provide the. justification for using a - non-sLfety significant component to provide adequate protection of a safety-significant component; or this report should include a requireraent to periodically monitor all jet pump sensing lines to assure that they are operable and are providing relitble z_ information. 9. In Section 1.3 cf the report, the. category of "No Si5nificant Degradation - (pecific" is introduced. Components in thic category are stated to not exserience any significant degradation due to neutron irradiation enurittleraent,, thermal aging entrittlement, and fatigue, because of material selection, distance from the core, loading magnitudes, repiccement at regular intervals, or geometric location. More detailed informatiun on the basis f_or establishing that no plant unique features are.present that would invalidate the recomniendations of this section are needed. 10. In discussing the Orificed Fuel Support on pages 1-7 and 1-8, the report = indicates that current inspection procedures per Section XI of the ASME Code.are considered to be sufficient to detect any significant degradation = ar.d that' the effects of aging embrittlement can be bounded within acceptable limits. Other than making these unsubstantiateo statenients, no. evaluations m or analyses were presented. Provide the evaluations or technical bases to support the stated conclusions.

11. On page 1-9, the report states that current inspection procedures may not detect stress corrosion ciacking (SCC) in the core shroud, top guide and core plate.

However, the aging degradation management discussion in Section 6 relies heavily on detection by NDE, but gives no discussion of the effectiveness of the procedures. Please explain this apparent discrepancy in stated aging management methods. 11. In Section 3.1, the report provides a descriptive overview of the EWR pressure vessel intetnal components. The chapter addresses, component by component, their f unction and design and states whether the component is safety significant. The' criteria for being excluded as not being safety significant is not stated. Because of the possibility that a non-safety component, when failing, could effect safety related components, the "h criteria for exclusion should be made more explicit in the text of the report. =_

13. On page 3-13, the report note: that GE has recommended that the neutron source holders be removed from the rer: tor after the first fuel cycle, and that in'many instances they have been.

It then concludes that these components s e eim- . - -um--

a o, T pose nc potential for-loose parts. The report should specify aging management-practices for those instances where the utilities have chosen not to follow the GE reconiniendation to remove the source holders. y

14. On page 3-17, the-report states that the feedwater spargers are not considered safety significant. The High Prtssure Coolv' Injection Systen.

(HPCI) is safety related but the report claims that even ..ae feedwater sparger contained significant tracks, the flow rate f rom the HPCI would rcmain the same. { However, if the feedwater spurger contained significant cracks and the HPCI system were activated,_ the possibility of the sparger breaking up and generating loose parts was not addressed nor the effect of these loose t parts on' safety during a situation when the HPCI system is needed. E

15. Section 3.2. of the report is quite general in nature. While reference is L

made to regulations, codes, and standards; a statenent of the operating conditions, loads, cycles, environment, and other such factors that must be known to Gesign these components is not presented nor is reference made to where such information may be found. The information on inspection art testing _.is also quite general in nature. Insufficient detail is presented to allow a knowledgeable person to-review and waluate the adequacy of-the inspection and. testing and that is performed. A description of the contents, and a review of its relevance to extended life operation, should be made of all of. the docuruents cited in Section 3.2.3.

16. On page 3-34, for each component for which degradation is described, the report should make a statement as to its generality among BWRs, and how extended life operation will affect the frequency or extent of the degradation or failure.

On page 3-35, the report states that crack growth beyond the cold-work -17. layer would be limited due to the resistant (annealed) stainless steel. There is significant laboratory data that shows above a threshold K' crack stowth rates in annealed material can be as high as in sensitNE8 materials. 'In any situation where initiation is possible by any means '(cold work, thermal fatigue), the stress levels must be evaluated to assure that the resulting stress intensity levels are below the threshold levels. On ptw ' 4 of the report, the quoted references deal primarily with 18. irradu&n performance under fast reactor conditions. Provide the raticWe why this data base can be used in BWR conditions. Also the statt :a.s that there is no embrittlement of stainless steel is true only in the sense that unlike ferritic steels, the austenitic steels do not undergo cleavage. However, the reductions in infracture toughness are large and certainly could impact the analysis of cracked components. The foramaterialwithfluenceof1.5x10')bebasedonanactualmeasure 4 quoted values of K on page 5 appear to n/cm' (see Ref. 3 in EPRI I Np-050). There is no reason to believe that this is a lower bound value for all materials and higher fluence levels.

19. On page 4-5, the discussion on thermal aging of cast stainless steel I

components does not address the potentici effect of irradiation on the rete of aging degradation of material toughness. It is likely that neutron

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t irradiation will accelerate the embrittlenient of the duplex material, although this has not been investigated for temperaturas comparable to a BWR operation.

This concern must be addressed for evaluation of the lons-term therral aging performance of the orificed fuel support.

20. On ptge 4-6, the report needs to address what methodology will be used to estiniate the level of toughness loss as a function of material chen,istry and microstructure and time at teniperature.

Further, an evaluation is needed of what constitutes significant toughness loss and what will be cone to nianage the problem. Further, the report states that recent microstructural studies of aged duplex stainless steels revealed the-forn.ation of three unidentified precipitates. However, the information reported is not recent; over the past several years these phases have been identified and analytical models have been developed for predicting toughness of cast stainless steel as a function of n.aterial chtmistry, ndcrostructure and time at temperature. The report also states that vEry little fracture toughness data are available for unaged or aged cast stainless steel. This also is old information and in fact, considerabit data now exists for the aged and unaged cast stainless steel.

21. On page 4-9, the use of the word " sensitizes" nicy cause some confusion.

Perhaps it would be better to say that nuclear exposure can also lead to SCC.

21. On page 4-10 of the report, the stated range of temperatures (850-1000 F)-

that can cause sensitization in austenitic stainless steels is much too A teniperature range of 900 to 1500*F should be used. narrow. On page 441 of the report, the data base which led to a thgshold fluence E3. 2 of 5 x 10' n/cm' should be examined. On page 4-10, 1 x 10 n/cm is suggested as a potential threshold level. Please provide-a detailed discusson on the uncertainty associated with an application of the threshold fluence. 24. Chapter 4 provides an overview of observed or anticipated mechanisms for degradation. The presented information lacks detail in some cases, and provides an inadequate amount of reference material. The effect of each mechanism on each safety significant component is addressed in Table 4.2. This table should be expanded to cover ali vessel internals and all mechanisms.

25. On page 5-2, the report states that the inservice inspection (ISI) requirements for BWR internals are covered in Table IWB-2500-1 of the ASME Code and also that-the Code is developing standards for more internals inspections. The visual inspection requirements in the code for safety related internals are not considered adequate. The report does not cover the adequacy of the currently required inspections nor does it address how the adequacy of the standards under development will be judged and what adequate inspections will be conducted during extended life.

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e ~ 20.;.On page 5-5 of the report, cyclic crack growth is discussed in generel terms. What specific fatigue-crach growth curves have been or are proposed for use?

27. On page 5-9, the report concludas that linear' elastic fracture machanics (LEFM) breaks down in the analysis-of the growth rates of very; short cracks.

presumably this implies that for analysis,-some minimum length. must be assumed, but as in the entire discussion of fatigue analysis no specific values are given. Provide the specific valles to be used. 20. The report states that cracking of the source range lionitor/ intermediate range monitor (SRM/lRM) dry tubes has not been observad in. the pressure boundary portion of the tubes. However, it did not address why such cracking in the press' ire boundary portion of the tubes is not expected in the future and what the consequences / actions would be.

29. On-pages 5-12, it is stated that " Failure of the control blades could impair the ability to shut down the reactor. However, the failures observed will not compromise their shutdown capabilities." No justifica-tion was given for assuming thht failures that may be observed in the future will similarly not impair the ability to shut down the reactor.

Provide the basis for the stated conclusions. 30. Sections 6.2 through 6.5 addresses stress corrosion cracking.of the susceptible individual components. The recommended approaches in each case include periodic volumetric inspection, SCC monitoring, water cheniistry control, and refurbishment, if necessary. The report should also provide guidance on the use of trending or aging n.anagement programs. 31.- Techniques for controlling the electrochemical potential (ECP) in the recirculation piping have been demonstrated. Controlling the ECP in-the core region is more difficult. What data are available to show that this is a practical solution for the cracking of internal components? 32. Volunietric' inspection for IGSCC even in recirculation piping is~ difficult to do reliably. How were (are) the inspection techniques for the top guide and other internal components to be qualified?

33. On pages 6-5, Refurbishment" is mentioned several times as an

" acceptable" innhod of aging management. However, it is not clear for reasons of access, exposure, and material condition that refurbishment of. components really is always a realistic alternative. For example, repairs of irradiated stainless steel components at the Savannah River involving welding have proven to be extremely difficult because of helium generation in the material during irradiation. ........ _ _. _, _, _ _ _ _}}