ML20198P958

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Forwards Operability Determination for ISI Components That Were Not Fully Examined in Second ten-year ISI Interval as Requested by Ts,For NRC Info
ML20198P958
Person / Time
Site: Monticello, Prairie Island  Xcel Energy icon.png
Issue date: 01/13/1998
From: Richard Anderson
NORTHERN STATES POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9801220300
Download: ML20198P958 (37)
v • d • e


Text

Northern States Power Company 414 Nicohet Mall Y M iU$7) m Y January 13,1998 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20S55 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR 22 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 50-282 License No. DPR-42 50 306 DPR-60 information Regarding Operability of the Systems with Components That Have Had Limited Inservice Inspection Examinations During implementation of the ASME Section XIinservice Inspection Program

References:

1. Monticello LER 97004," Failure to Submit Relief Requests for Limited inservice Inspection Examinations"
2. Prairie Island LER 19702, " Failure to Submit Relief Requests for l Nited inservice inspection Examinations"
3. Prairie Island Letter, " Reply to Notice of Violation (Inspection Repon 97003),

Failure to Submit Rela Requests for Limited inservice Inspection Examinations" A meeting was held October 30,1997 in NRR offices at White Flint in order to provide further information to the Staff regarding the Monticello and Pral;ie Island plants'

" Failure to Submit Relief Requests for Limited Inservice Inspection Examinations," as previously discusse.1in References 1,2, and 3. At that meeting NSP representatives presented information contained in the plants'" Operability Determination For ISI Components That Were Not Fully Examined in 2nd Ten Year ISI interval as Required by Techrilcal Specifications."

Attached to this letter is a copy of that operability determination for the information of 7

the NRC Staff. Note that the operability determination addresses both Monticello and ,

F"loM P4jp L.'lllll!!lllify)l.ljillll

l USNRC NORTHERN STATES POWER COMPANY January 13,1996  !

Page 2 i i

Prairie Island because both plants were affected by the same policy which was determined to be in error regarding the necessity for submitting Relief Requests.

In this submittal we have made no new Nuclear Regulatory Commission commitments, the corrective actions to prevent further violations were previously committed to in the LERs, References 1 and 2.

Please contact Jack Leveille (612-3881121 Ext. 4662)if you have any questions related to this letter, ud Dr'7v loger O Anderson Director Nuclear Energy Engineering c: Regional Administrator -- Region Ill, NRC Prairie Island Senior Resident inspector, NRC Monticello Senior Resident inspector NRC Prairie Island NRR Project Manager, NRC Monticello NRR Project Manager, NRC Kevin Connaughton, NRR Project Manager, NRC J E Silberg

Attachment:

Operability Determination for ISI Components That Were Not Fully Examined in 2nd Ten Year ISI Interval as Required by Technical Specifications U M TPM4.R C i s

Operability Determination 1 For ISI Components That Were Not Fully Examined in Second Ten Year ISI Interval as Required by Technical Specifications

1. Determine What Equipment is Degraded or Potentially Nonconforming.

Class 1,2 and 3 components with ISI inspection limitations.

A summary quantification of the compononts affected listed by ASME Item number and examination method are shown Figures 1,2,3 and 10. In all cases the examinations have been performed to the extent practical.

II. Determine the Safety Function (s) Performed By the Equipment Inservice inspection examinations are performed on pressure-containing components and their supports to assure the integrity of the pressure boundary and the protection of the health and safety of the public.

ill. Determine ;he Circumstances of the Potential Nonconformance The following 10 CFR 50.55a paragraphs apply to the inservice inspection of components in accordance with the ASME Section XI code:

50.55a(g)(1): For a boiling or pressurized water cooled nuclear power facility whose construction permit was issued prior to January 1,1971, components (including supports) must meet the requirements of paragraphs (g) (4) and (5) of this section to the extent practical.

50.55a(g)(4): Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code CIsss 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions of the ASME Boller and Pressure Vessel Code ... to the extent practical within the limitations of desip, geometry and materials of construction of the componente.

50.55a(g)(5)(!v): Where an examination requirement by the code or addenda is determined to be impractical by the licensee and is not included in the revised inservice inspection program as permitted by paragraph (g)(4) of this section, the basis for this determination must be demonstrated to the satisfaction of the Commission ...

NSP's interpretation and implementation of these paragraphs has been questioneo, in the past, NSP has considered these requirements to mean that if design, geometry, or material of construction prevent full code compliance, (g)(1) end (g)(4) allowed examination to the extent practical and (g)(5)(iv) did not apply. Such components were examined to the extent practical and it was 1

__ _ m _ _ _ _ _ _ _ ._ _ _ _ ._ _ _ _ _ _ _

f i

believed that 10 CFR 50.55a requirements were met. Such examinations were  !

classified as " limited" and clearly identified on the examination report and in the  ;

summary report sent to the NRC after each refueling outage.  !

t Both Prairie Island and Monticello implemented these regulations the same. At ,

Prairie Island the NRC stated that (g)(5)(iv) applied to these limited exams. This  !

concern was passed on to Monticello regarding cornpliance with Technical  !

Specification 4.15 since reilef requests were not submitted for Monticello ,

components with design and geometry limitations. ,

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. IV. Determine the Requirement or Commitment Established for the Equipment, and  :

'; Why the Requirement or Equipment May Not be Met.

Section 4.15.A.1 of the Monticello Technical Specifications and Section 4.2.A.1 i of the Prairie Island Technical Specifications require Inservice Inspection to be  !

E completed in accordance with "Section XI of the ASME Boller and Pressure  ;

Vessel Code and applicable Agenda at, required by 10 CFR 50, Section  :

50.55[a)(g), except where relief has been granted by the Commission pursuant  !

to 10 CFR 50, [Section) 50.55[a)(g)(6)(1)." The Section XI Code contains  ;

requirements to examine essentially 100% of the weld length. Due to design and geometry limitations,100% coverage could not be achleved on some welds examined during the Second Ten Year Interval. i Since Monticello and Prairie Island's construction permits were dated prior to January 1,1971, no relief requests have been submitted on these limited i examinations. This was based on the mir. interpretation of the words "to the  :

extent practical"in 10 CFR 50, St etion 50.55a(g) and the words in Section 50.55.a(g)(4), quoted below.

(4) . . . to the extent practical within the limitations of design, geometry and

-materials of construction of the components.

It was interpreted that these sections meant that interferences inherent ir, th e c design constituted impracticality and were, therefore, exempted. This was the approach taken for the ISI Program since Section XI was implemented at Monticello and Pralrie Island.

- 10 CFR 50.55a(g)(5)(iv) requires that relief requests be submitted within 12 months of the end of the 10 Year Interval to which they apply. The non-compliance is for the Second Ten Year Interval examinations. The current Third Ten Year Interval programs are still in compliance since several years remain to ,

submit applicable relief requests under (g)(5)(iv). ,

The Technical Specification surveillances were completed to the extent practical for the Second Ten Year Interval; Relief requests for limited examinations ,

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performed during the Third Ten Year Interval will be submitted to the NRC Staff  ;

for review and approval. l, V. Determine by What Means and When the Potentially Nonconforming Equipment Was First Discovered.

Following discussions with members of the NRC Staff in February of 1997, it was -  !

determined that the Code of Federal Regulations had been interpreted i differently by the NRC. j VI. Determine the Basis for Declaring the Affected System Operable .

A. Pressure Boundary Integrity l The operability determination of a safety system is tsessed in several key areas. The key areas are as follows: does the system meet its design basis and system requirements; does the system meet its functional  :

requirement in accordance with the Technical Specific > tion; and is the system integrity maintained. The only area that is affected by the nonconforming condition is the assurance of the system's ability to i maintain integrity. The other key areas for operability assessment are not affected by this nonconforming condition.

B. Code Required Inspections are Based on a Sampling Design The intent of the ISI program is to ensure the integrity of Class 1,2, and 3 piping components and their supports after they are placed into service.

Service related degradation in the systems within the scope of ISI normally manifests itself in fatigue cracking or stress corrosion cracking of  !

welded joints. These types of cracking are normally detected on the inside surface of a pipe or vessel with a volumetric method such as ultrasonic testing (UT) or on the outside surface with a surface method such penetrant testing (PT) or magnetic particle testing (MT).

Other service related degradation from operation might be over-stress (such as may come from a water hammer incident) or general corrosion  !

such as from boric acid leaking. These types of service related degradation are typically detected with visual examinations.  ;

Some examinations in the Second Ten Year Interval were limited, however, the ISI program still provides a high confidence level in the integrity of the systems. Only a sampling of the weld population is required to be inspected by the Section XI code. In many cases, the ISI -

program sampling requirements exceed the minimum code requirements, see Figures 4,5 and 6.

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A sampling program allows confidence in the integrity of the pressure coundary to be established without examining all of the welds in the system. Besides only being required to examine a percentage of the welds, Code Case N-460 (endorsed by the NRC) allows for only a portion of each inspected weld to be examined (i.e., greater than 90% meets the

" essentially 100%" requirement). For example, Figure 5 (for Prairie Island) shows that only 25% of the piping welds need to be examined for the reactor coolant piping. One category of the reactor cooling piping welds for Prairie Island Unit 1 was checked: B9.11 (circumferential welds > 4" NPS). The requirement is 25% of the welds need to be inspected and 42% of the welds are in the UT program schedule. If the examinations averaged only 70% coverage of the weld volumes with the UT exams, the program would exceed the 22.5% minimum weld volume required for the whole piping system (assuming that only 25% of the wolds would have been examined with only 90% coverage for each weld).

Another Prairie Island example is the requirement for weld examinations of the i, team generator tubesheet-to head welds. The requirement is that one steam generator per unit be excmined. All of these welds are examined at Prairie Island. An example for Monticello is the 89.11 welds.

Again the requirement is 25% of the wolds and there are 30.5% of the welds in the program schadule; if the weld coverage only averaged 75%

por each examined weld, the 22.5% per total piping system weld coverage would be exceeded. From a sampling perspective of establis',ing confidence in the integrity of the system, it is concluded that the sampling percentage critoria for total weld coverage is met, even though the coverage criteria par each examined weld is not met, in addition, the NRC acknowledges that an acceptable level of confidence is established by a sampling method; (1) the regulations approve the sampling schedules of the Code, (2) Reg Guide 1147 endorses Code Case N-460, which allows 90% coverage of a weld to meet the

" essentially 100%" requirement, and (3) the regulations allow for relief to be granted where full compliance with the inspection requirements are not practical.

All limited examinations perfermed to date were retrieved and reviewed for this evaluation. The asults of this review for each unit are shown on Figures 7,8 and 9 along with a summary of the flaws identified by the inspection program. NSP maintains a high confidence level that the limited examinations performed providtrj a valid representative sampling for the inservice inspection program.

The Code requirements and the NRC regulations provide a logical justification that the means of building assurance is not based on 100%

inspection coverage. For these reasons, the reduced crea or volume 4

coverages of the examinations for some components has not diminished the confidence level for the systems in the ASME Section XI ISI programs for Prairie Island and Monticello.

C. Service History With the exception of the recirculation loop at Monticello and the steam generators at Prairie Island, very few significant indications have been found with ISI over 20 years of operating Monticello and Prairie Island.

Because so few indications have been found during examinations, there is high confidence that there are no indications in the un-inspectable areas of components with limitations.

The recirculation piping at Monticello was replaced in 1984 with Type 316NG stainless steel, The new piping material had a maximum of 0.02%

car _ bon and all welding was tightly controlled to minimize the susceptibility to stress corrosion cracking. All shop welds were solution annealed after fabrication and fields welds were given Induction Heating Stress improvement (lHSI) prior to placing the unit back in service. This modification significantly reduced the susceptibility of the recirculation piping to stress corrosion cracking. Monticello uses hydrogen water chemistry to further protect austenitic stainless steel in the reactor coolant piping from stress corrosion cracking.

Steam generators at Prairie Island represent a case where it was determined that to rely on sampling would not be prudent. Because of the susceptibility to degradation of steam generator tubes,100% tube length eddy current exams (tube length exams with bobbin coil and rotating pancake coil technology of regions highly susceptible to stress corrosion -

cracking) have been performed each refueling outage for the last 15 years.

D. Additional Test / Inspection Methods Hydrostatic tests are performed during regular inspection intervals to ensure the piping system is capable of maintaining pressure integrity.

System integrity is monitored continuously during normal operation by many direct and indirect methods, e.g., containment radiation monitoring, containment air monitoring, containment leakage detection and monitoring, containment temperature monitoring, etc.

E. Materials Used Materials used in the pressure retaining components and supports at Monticello and Prairie Island have a long history of reliable service in 5

these applications With a few exceptions related to design or the fabrication practice, the incidence of service-induced cracking of these materials in systems within the ISI program is very low.

F. Monitor Industry Experience Industry events are monitored for applicability to Monticello and Prairie Island, INPO Nuclear Nehvork and NRC Information Notices are evaluated for applicability, NSP participates in both BWROG and WOG materials subcommittees, in addition. NSP is also a member of the

' Welding Research Council and the Pressure Vessel Research Council.

Any material problems at other plants are evaluated and plant inspections are planned as necessary.

G. Programs to Monitor the Material Condition of Plant Components The ISI Program at Monticello and Prairie Island have monitored the integrity of the plants' piping over the last 20 plus years. The program has been evaluated by the NRC as well as NSP Quality Services audits and found to be an effective program.

The program has identified a small number of service-related degradations which have been aggressively addressed. Prairie Island has an extensive and proactive eddy current inspection program to monitor the steam generator tubes. Aggressive action was taken by Monticello to replace the recirculation piping that was found to be degraded.

A pro-active pipe wall thinning (erosion / corrosion) program was initiated in the early 1980's and has been effective in identifying degradation and preventing service-related failures.

H. ASME Class 3 Components Class 3 components that are required to be examined per ASME code are  ;

listed in Table IWD 2500-1. These requirements have been modified by Code Case N-491, Both Monticello and Prairie Island units are applying Code case N-491 in their ISI programs. The examination requirements for Class 3 components are 10% of Class 3 piping supports. Figure 10 contains the number of Class 3 piping supports, the number of examinations completed in the Third Ten Year Interval and the number of

-inspections scheduled for completion in the Third Ten Year Interval.

Figure 10 shows that the number of examinations for the Class 3 components far exceed the required number of examinations, it is believed that the number and extent of exam limitations for Class 3 6

4 components is very small, if any exist. Since the number of inspected components greatly exceeds the number required, there is a high confidence that examination of Class 3 components will meet code requirements.

l. Components with Limited Exams in the Second Ten Year Interval But Not To Be Examined in the Third Ten Year Interval There are three categories of components with limited exams in the Second Ten Year interval:

e those repeated in the Third Ten Year Interval which have already r been re examined and for which relief requests will be submitted shortly e those to be repeated in the Third Ten Year Interval which have not yet been re-examined and for which relief requests will be submitted following the re-examinations e those which will not be re-examined in the Third Ten Year Interval Eecause there will be no relief requests submitted for the third category, it has been deemed prudent to provide a more detailed look at the safety significant components included in this group.

To determine the safety significance of items with limited examinations in the Second Ten Year Interval, but not scheduled for examination in the Third, a two part review was performed. The first part was to identify the components with limited examinations that were not included in the Third Ten Year Interval but were in the Second Ten Year Interval. Once identified, these components were compared to the maintenance rule risk significant items and then quantified, thus narrowing the scope of items to evaluate. Secondly, the assessment of this set of items with Prairie Island and Monticello focused on service degradation mechanisms, other potential type degradation mechanisms, service degradation safety significance, and industry experience.

All the items addressed are not being examined in the Third Ten Year Interval inservice inspection programs due to changes in ASME Section XI code requirements. The First Ten Year Interval inservice inspection programs for Monticello and Prairie Island were developed from the 74S75 Code which required that 25% of the Class I welds be examined each interval, thereby examining 100% of the Class I welds by the end of life,40 years. In 1978 the Code was revised to target detection of inservice degradation. The 25% sample of the Class I welds to be examined were concentrated on terminal ends, dissimilar metri welds and welds with high stress levels. These welds were to be re-Inspected over the life of the plant to detect inservice degradation.  !

7 e

The result of the code change was that Monticello and Prairie Island scheduled some welds for examination in Second Ten Year Intervals that were not examined in the First Ten Year Intervals. The Third Ten Year Interval scheduled all dissimilar and terminal end welds and the remainder of the required 25% sample was randomly selected from other welds previously examined in either the First or Second, or both intervals.

Thereby inservice degradation could be monitored over a 10 or 20 year period.

1. Prairie Island in the Prairie Island Units 1 and 2 inservice inspection program, the components with limited examinations that were included in the Second Ten Year Interval, but not the Third, are Class 11 components exempted per IWC-1221(e), those components that are statically pressurized. These include portions of boric acid tank and piping, accumulator tank and discharge piping, and containment sump pump piping. The Class ll Category C-F weld components that are included in the program, but not inspected, due to base metal thickness requirements, do not have associated hangers inspected; these components are in the RHR discharge pipe and SI suction pipe.

The maintenance rule risk significant components include the following pipes (for rupture scenarios}: the RHR suction, the accumulator return, the safety injection to reactor vessel, the main steam, and the reactor coolant. Of the components exempted in the Third Ten Year Interval, only the accumulator discharge piping has any nsk significant welds or supports. For Prairie Island Unit 1 the risk significant items with limited examinations are 13 circumferential welds. For Prairie Island Unit 2 the risk significant items with limited examinations are 3 circumferential welds. Stress corrosion cracking is the principal service degradation mechanism for this system. However, these components are unlikely to develop stress corrosion cracking due to the materials of construction, weld joint configuration and service conditions.

The following is the service degradation history of the Prairie Island units excluding the steam generator.

The boric acid lines experienced stress corrosion cracking in weld areas resulting in minor leakage. These lines were replaced in 1985 with very low carbon stainless steel, to avoid sensitization of weld heat affected zones. Additionally, the welds were either post weld heat treated, or a water backed welding technique was used 8

to control residual stresses from welding. To ensure that the boric acid remains in solution, a rocirculation system was added, and the type of heat tracing was changed to eliminate the potential for overheating. The material and operational changes are adequate to prevent reoccurrence, therefore additional examination is not necessary.

In 1992 the Feedwater nozzle extensions were replaced. They were replaced due to flow accelerated corrosion concerns, which turned out to be unfounded. A thermal monitoring system was also installed at this time to monitor potential thermal fatigue in response to NRC Bulletin 88-08. These nozzles remain in the Section XI program.

During a plant walkdown a leak was discovered originating from one of the accumulator tank level transmitter nozzles.

Westinghouse performed a metallurgical analysis of the leaking nozzle ard concluded that the nozzle leak was due to stress corrosion cracking caused by improper installation. Misalignment during welding provided an area of high stress concentration, thereey providing a preforential site for stress corrosion cracking.

This Class ll nozi le is outside the scope of Section XI. Ultrasonic testing of all accumulator tank instrument nozzles was added to the inservice inspection program in 1988 and continues today.

2. Monticello in the Monticello inservice inspection program, the risk significant components with limited examinations that were included in the Second Ten Year Interval, but not the Third, are Class ll components that are beyond the last shutoff valve in an open ended system and exempted per IWC-1222(d), or are less thar' 3/8 inch nominal wall thickness category C F components exempted per Table IWC-2500-1,1986 Edition. Some components in the containment spray and core spray systems were examined during the Second Ten Year Interval but are not scheduled for examination in the Third Ten Year Interval due to these exemptions, respectively. These components are made of carbon steel, material highly resistant to stress corrosion cracking.

Tha following is the service degradation history of Monticello excluding the reactor core internals.

In 1982 stress corrosion cracking was discovered on the recirculation system piping. Modifications included recirculation 9

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pipe replacement with 316NG materials, low heat input welding and induction heat stress improvement after welding. Similar f replacements have been made to the core spray piping system.

Plant operation includes hydrogen water chemistry wFeh greatly reduces the potential for stress corrosion cracking. All of these '

welds remain in the Section XI program; augmented examination continues under guldance from NUREG 0313 Revision 2.

3. General ,

The Prairie Island and Monticello plants have experienced e ome -

pipe degradation. These experiences are consistent with industry experiences, as noted in a white paper on ASME Code Case N-560 and Welding Research Council Bulletin No. 382.

The ASME Code Case N 560 white paper notes that the majority of flaws found in Category B J welds have been caused by factors outside the scopc of current selection criteria (Stress Corrosion Cracking and Thermal Stratification). A survey of the industry shows that, (for the respondents this includes 37332 Category B .!

welds) 156 welds had service induced flaws of which; 151 were attributed to stress corrosion cracking. One third of the flaws were found with ASME XI ultrasonic examination; the other two thirds of i the flaws were found by visual inspection or leakage.

Weiding Research Council Bulletin No. 382 lists the potential ,

degradation mechanisms as general corrosion, stress corrosion cracking, erosion corrosion, cavitation, mechanical fatigue, thermal fatiguo, thermal stratificatinn and water hammer events. It also shows that there have been no severe failures in Class I piping; cracking / leakage is the principal failure mode. The non-Class 1 >

piping principal failure mechanism is erosion / corrosion. Stress corrosion cracking has not caused a single serious accident nor is likely to. -

Stress corrosion cracking has been the principal degradation mechanism at Prairie Island and Monticello. The leakage resulting from through wall stress corrosion cracking can be detected by ,

system leak testing, system walkdowns or leak detection systems without severe consequences.

Thermal fatigue and stratification has not besn a problem at either Monticello or Prairie Island. Erosion / corrosion is monitored at Monticello and Prairie Island with a specific wall thinning inspection program. Mechanical fatigue and water hammer events are not 10

s. +m.. - - .e.. , ,.nm, a , ,

practical to manage by perledic inservice examinations. ASME Code Case N 560, Appendix 1, paragraph 2.3(g) states, "High cycle i mechanical vibratory fatigue is not addressed as a damage  ;

mechanism by this Code Case. The nature of this mechanism is  !

such that generally almost the entire fatigue life of the component is expended during the crack initiation phase. Once a crack, initiates, failure quick.y follows. Consequently, this mechanism does not lend itself to management by periodic inservice examinations (i.e. volumetric, % we, etc.). Frequent system walkdowns, leakage monitoring wystems, and current ASME -

Section XI system leak test requirements are practical measures to ,

address this issue." Volumetric examination does not provide the means to monitor piping degradation induced by fatigue. Routine system walkdowns, leakage monitoring systems, and current ASMc Seebn XI system leak test requirements do provide the meer 'o monitor piping degradation induced by fatigue, in cenclusion, plant experiences have shown that there are no unusual degradatien mechanisms and Prairie Island s and Monticello's experiences are similar to the industry's e, oerience. Ctress corrosion cracking has been observed at both sites; industry degradation mechanisms have been stress corrosion cracking and thermal fatigue.

Stress corrosion cracking manifests itself in small leaks, consequently visual examination through frequent system walkdowns, leakage monitoring and system leak testing is the most practical, economic means to monitor this degradation mechanism. In the twenty plus years of operation at each of the plants, any design or fabrication defects or problems should have been revested, and problems found have been properly addressed. Monticello has replaced susceptible pipe with less susceptible materials and used improved welding methods. 'iduction heating stress improvement and hydrogen water chemistry have been utilized to improve resistance to stress corrosion cracking All inservice h'.pections at Monticello and Prairie Island have been done to the greatest extent possible. Limitations are due to geometry, matorials and design. The only degradation mechanism has been stress corrrsion cracking. Industry experience has shown that it is not a concern in primary coolant systems at PWRs. BWRs manage this degradation ,

mechanism by following the guidelines in NUREG 0313, Revision 2.

Monticello has replaced susceptible pipe with less susceptible materials and by using improved welding methods, in addition, lHSI and hydrogen water chemistry have been utilized to improve resistance to stress corrosion cracking.

11

Tha large extent of weld volume examination achieved in the safety >

significant systems at Prairie Island and Monticello ensures safety of those systams.- Stross corrosion crack 5g is unlikely in these systems, its - .

occurrence is not usually isolated, and it does not lead to catastrophic y failure. _ Therefore, limited examir ations are not a significant safety concern. ,

a References White Paper: ASME Code Case N-560, Evaluation e' Inservice inspection Requiremer,ts for Class 1, Cetsgory B-J h issure. Retaining Welds in Piping, Report No. 92-01-01 Revision 1, Juiy 1995.

ASME Nde Case N-560, A3ernate Examination Requirements for Class I, Category B-J Piping WeldsSection XI, Division 1, dated August 9, 1996. ,

Prairie Island Safety Evaluation No. 232: Accumulator Tank Nozzle Leakage and Repair.

J. Summary The Inservice Inspection Program is the surveillance that assures continued integrity of reactor coolant pressure boundary components, their supports and other systems importent to plant safety. Non-destructive examinations are performed to detect service related degradation. Since 100% weld length (volume) can not be achieved due to design and geometry limitations on some components, reasonable assurance that pressure integrity is maintained for the affected systems  ;

is based on the full examination of similar components (i.e. representative sampling), exanination to the extent practical of the component not fully examined, the use of materials with low susceptibility to crack generation, and the system pressure tests.

l At 5r.s%d UT examination techniques are being used to the extent pactic$ to obtain more repeatable and reliable examinations.

For these reasons, the components in the Second Ten Year Inter tal with limited examinations are operable. Components in the Third Ten Year Interval are also operable since relief requests will be submitted in accordance with the requirements of 10CFR50.55a.

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Vil. Corrective Actions

A.1 Actions Planned ,

The actions planned are stated in'the Licensee Event Reports: 97004 ,

(Monticello) and 19702 (Prairie Island).. l

-B; Justification for Action Schedule -

This schedule was detennined to provide NSP steff time'during outages -

to obtain physical data needed to accurady v.tegorire the inspection j limitations for limited examinations already completed during the Third -

Ten Year interval. Since component oport.2bility has been assured and

. j the ext minations with limitations were co:npleted to the exteat possible, this schedule provides some flexibility to address ALARA and resource concerns.  ;

' Vill. . Summary r

< ' Pressure containing components have been examined for over 20 years at all 'i three units. Trending has revealed no major service related degradation, except

for the recirculation piping at Monticello and the Steam Generators at Prairie Island. These issues have been addressed.- The evidence supports that the  !

~

l Integrity _of the reactor coolant pressure boundary and other systems important ,

to safety are intact. Continued Inservice Inspection during the Third Ten Year i Interval will provide additional assurance of system integrity.

  • The components partially examined during the Second Ten Year Interval are operable based en the above discutsions.

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i 1 i i i 2 I i I 3 i I I 4 4 i i 5 s of Lamd.ed Laame m Une Number of Components Number of Limited 2nd Ten Year interval & Part Number of Llrr Scheduled for Exam 6 nation Examinatione Scheduled of the 3rd bu;HQT Examinations in th<

Number of Components in During the 3rd Ten Year DurNO thr,3rd Tsn Year Scheduled During the 3rd YwarIntervalNQI P the 3rd Ten Year interva! Interval inwrval Ten Year Interval 3rd Ten Year Ini m l g a y a p op '?p op p .c sp op p gA E E uJ 1 E XA 1 ]E Wa . E 1 WA j

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B10.10 B12.20 1 1 812 50 44 18 2 B13.10 10 10 013 20 1 1 B13 30 1 1 014.10 96 6 4 1 B15.10 815.11 2 2 C1.10 4 2 2 C1.20 4 2 2 1 C2.31 4 4 C3 40 C5.10 31 31 5 5 1 1 C5 51 713 713 80 80 31 5 5 1 2 C5 81 2 2 C7.10 12 12 C7.20 12 12 F.A. CL i 180 163 1

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  • EST1 MATED" Number of Limited httfagg,
rt of the *g1IlMAILQ" Numt>er of Limited YkunL
  • ESTIMATED" Number of Umited Volumetric Examinations from Co5smns 3,4 and 5 Examinations from Columns 3,4 and 5 Examinations from Columns 3. 4 and 8 nrvil s a a s a a e a , a i -

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Figure 4 Scheduled Examinations Exceed Requirements for Class 1 and 2 ASME Code Components for Monticello ISI Third Ten Year ISI Third Ten Year interval items Interval items Monticello - 86 Section XI Required to be - Scheduled to be ASME Code Category examined Examined Examination Category B-A, Pressure B1.11, B1.12, B1.11, B1.12, Retaining Welds in Reactor Vessel B1.21, B1.22, B1.21, B1.22, B1.30, B1.40 B1.30, B1.40 Examination Category B-D, Full B3.90, B3.100 B3.90, B3.100 Penetration Welds Of Nozzles in Vessels -

Inspection Program B Examination Category B-F, Pressure 85.10, B5.20 B5.10, B5.20 Retaining Dissimilar Metal Welds Examination Category B-G-1, Pressure 86.10, B6.20, B6.10, B6.20, Retaining Bolting, Greater Than 2 in. In 86.40, B6.50, B6.40, B6.50, Diameter _

B6.180, B6.200 B6.180, B6.200 Examination Category B-G-2, Pressure 87.10, B7.50, B7.10, 87.50, Retaining Bolting,2 in. And Less In 87.70, B7.80 B7.70, B7.80 Diameter Examination Category B-J, Pressure 89.10, B9.11, B9.10, B9.11, Retaining Welds in Piping B9.21, B9.31, B9.21, B9.31, B9.32, B9.40. Note 89.32, B9.40.

1d: required 25% Scheduled 30.6%

(surface and for surface and 32%

volumetric) of for volumetric components be examinations examined Examination Category B-L-1, Pressure B12.20, 812.50 B12.20, B12.50 Retaining Welds in Pump Casings; B-M-1, Pressure Retaining Welds in Valve Bodies; B-L-2, Pump Casings; B-M-2, Valve Bodies

Figura 4 Scheduled Examinations Exceed Requirements for Class 1 and 2 ASME Code Components for Monticello ISI Third Ten Year ISI Third Ten Year interval items Interval items Monticello - 86 Section XI Reauired to be Scheduled to be ASME Code Category examined examined Examination Category B-N 1, Interior Of B13.10, B13.20, B13.10, B13.20, Reactor Vessel. B-N-2, Integrally Welded B13.30 B13.30 Core Support Structures And Interior Attachments To Reactor Vessels. B-N 3, Removable Core Support Structures Examination Category B-0, Pressure B14.10 814.10 Retaining Welds in Control Rod Housings Exarnination Category B P, All Pressure B15.10, B15.11, B15.10, 815.11.

Retaining Components Leakage Test, Done per code Hydrostatic Test except where relief has been granted Examination Category C-A, Pressure C1.10, C1.20 C1.10, C1.20 Retaining Welds in Pressure Vessels Examination Category C-8, Pressure C2.31 C2.31 Retaining Nozzle Welds in Vessels Examination Category C-F-1, Pressure C5.10. Note 2: The C5.10. Scheduled Retaining Welds in Austenitic Stainless welds selected for 16.1% for surface Steel Or High Alloy Piping examination shall and 16.1% for include 7.5%, . volumetric examinations Examination CEegory C-F-2, Pressure C5.51, C5.81. Note C5.51, C5.81.

Retaining Welds In Carbon or Low Alloy 2 The welds Scheduled 11.5%

Steel Piping selected for for surface and examination shall 11.2% for include 7.5%, volumetric examinations Examination Category C-H, All Pressure C7.10, C7.20 C7.10, C7.20 Retaining Components-i

Figure 4 Scheduled Examinations Exceed Requirements for Class 1 and 2 ASME Code Components for Monticello ISl Third Ten Year ISI Third Ten Year Interval items Interval items Monticello - 86 Section XI ASME Code Category $x*a$*eI !xa'nfn"e7 Examination Category F-A, Supports F1.10. Extent of F1.10. Scheduled Examination - 25% 90.6% Visual Examination. Note:

The percentage will be reduced when code case N-491 is applied Examination Category F-A, Supports F1.20. Extent of F1.20. Scheduled Examination - 15% 92.5% Visual Examination. Note:

The percentage will be reduced when code case N-491 is applied l

l l

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1 l Figure 5 Scheduled Examinations Exceed Requirements for Prairie Island Unit 1 IS: Third Ten Year ISI Third Ten Year Interval items Interval items PI Unit 1 - 89 Section XI Reauired to be Scheduled to be ASME Code Category Examined Examined i Examination Category B-A, Pressure B1.11, B1.21, B1.11, B1.21, Retaining Welds in Reactor Vessel B1.30, B1.40 B1.30, B1.40 Examination Category B-B, Pressure B2.11, B2.12, B2.11, B2.12, Retaining Welds in Vessels Other Than B2.40. Note 1: The B2.40. For item Reactor Vessels examination may be B2.40, both S/Gs limited to one Tubesheet-to Head vessel among the Weld Scheduled group of vessels performing a similar function Examination Category B-D, Full 83.90, B3.100, B3.90, B3.100, Penetration Welds Of Nozzles in Vessels - B3.120, B3.140 B3.120, B3.140 Inspection Program B Examination Category B-E, Pressure B4.11, B4.12, B4.11, B4.12, Retaining Partial Penetration Weids in 8 4.13, B4.20 B4.13, B4.20 Vessels Examination Category B-F, Pressure 85.10, B5.40 B5.10, B5.40 Retaining Dissimilar Metal Welds Examination Category B-G-1, Pressure 86.10, B6.30, B6.10, B6.30, Retaining Bolting, Greater Than 2 in. In 86.40, B6.50, B6.40, B6.50, Diameter B6.180, B6.190 B6.180, B6.190 Examination Category B-G-2, Pressure 87.10, B7.20, B7.10, B7.20, Retaining Bolting,2 in. And Less in- B7.30, B7.50, B7.30, B7.50, Diameter 87.60, B7.70 B7.60, B7.70 Examination Category B-J, Pressure 89.10, B9.11, B9.10, B9.11, Retaining Welds in Piping B9.21, B9.31, B9.21, B9.31, B9.32, B9.40. Note 89.32, B9.40.

1d: required 25% Scheduled 33% for (surface and surface and 42% for volumetric) of volumetric components be examinations examined

Figuro 5 Scheduled Examinations Exceed Requirements for Prairie Island Unit i ISI Third Ten Year ISI Third Ten Year Interval items Interval items PI Unit 1 - 89 Section XI Required to be Scheduled to be ASME Code Category examined examined Examination Category B-L-1, Pressure B12.10, B12.20, B12.10, B12.20, Retaining Welds in Pump Casings: B M-1, B12.50 B12.50 Pressure Retaining Welds In Valve Bodies; B-L-2, Pump Casings; B-M-2, Valve Bodies Examination Category B-N-1, Interior Of B13.10, B13.50, B13.10, B13.50, Reactor Vessel. B-N-2, Integrally Welded B13.60, B13.70 B13.60, B13.70 Core Support Structures And Interior Attachments To Reactor Vessels. B N-3, Removable Core Support Structures Examination Category B-O, Pressure B14.10 B14.10 Retaining Welds in Control Rod Housings Examination Category B P, Pressure B15.10, B15.11, B15.10, B15.11, Retaining Components Leakage Test, Done per code Hydrostatic Test except where relief has been granted Examination Category C-A, Pressure C1.10, C1.20, C1.10, C1.20, Retaining Welds in Pressure Vessels C1.30 C1.30 Examination Category C-8, Pressure C2.11, C2.21, C2.11, C2.21, Retaining Nozzle Welds in Vessels C2.22 C2.22 Examination Category C-F-1, Pressure C5.10, C5.11, C5.10, C5.11, Retaining Welds in Austenitic Stainless C5.21, C5.30, C5.21, C5.30, Steel Or High Alloy Piping C5.41. Note 2: The C5.41. Scheduled welds selected for 8.1% for surface examination shall and 7.9% for include 7.5%, . volumetric examinations Examination Category C-F-2, Pressure C5.50, C5.50 & C5.50, C5.50 &

Retaining Welds in Carbon or Low Alloy HELB, C5.51, C5.51 HELB, C5.51, C5.51 Steel Piping & HELB, C5.80. & HELB, C5.80.

Note 2: The welds Scheduled 21.4%

selected for for surface and examination shall 19.5% for include 7.5%, volumetric examinations

l Figura 5 Scheduled Examinations Exceed Requirements for Prairie Island Unit 1 ISI Third Ten Year ISI Third Ten Year Interval items Interval items El Unit 1 - 89 Section XI Reautred to be Scheduled to be ASME Code Category examined examined Examination Category C-G, Pressure C610 C6.10 Retaining Welds in Pumps And Valves Examination Category C-H, All Pressure C7.10, C7.20 C7.10, C7.20 Retaining Components Examination Category F-A, Supports F1.10. Extent of F1.10. Scheduled Examination - 25% 72.3% Visual Examination. Note:

The percentage will be reduced when code case N-491 is applied Examination Category F-A, Supports F1.20. Extent of F1.20. Scheduled Examination - 15% 54.8% Visual Examination. Note:

The percentage will be reduced when code case N-491 is applied

Figuro 6 Scheduled Examinations Exceed Requirements for Prairie island Unit 2 ISI Third Ten Year ISI Third Ten Year Interval items Interval items Pl Unit 2 - 89 Section XI Roauired to be Scheduled to be ASME Code Category Examined Examined Examination Category B-A, Pressure B1.11, B1.21, B1.11, B1.21, Retaining Welds in Reactor Vessel B1.30, B1.40 B1.30, B1.40 Examination Category B-8, Pressure B2.11, B2.12, B2.11, B2.12, Retaining Welds in Vessels Other Than B2.40, B2.51. Note B2.40, B2.51. For Reactor Vessels 1: The examination item B2.40, both may be limited to S/Gs Tubesheet-to one vessel among Head Weld the group of vessels Scheduled performing a similar function Examination Category B-D, Full B3.90, B3.100, B3.90, B3.100, Penetration Welds Of Nozzles in Vessels - 83.120, B3.140 83.120, B3.140 Inspection Program B Examination Category B-E, Pressure - B4.11, B4.12, B4.11, B4.12, Retaining Partial Penetration Welds in B4.13, B4.20 B4.13, B4.20 Vesseis Examination Category B-F, Pressure B5.10, B5.40 B5.10, B5.40 Retaining Dissimilar Metal Welds Examination Category B-G-1, Pressure B6.10, B6.30, B6.10, B6.30, Retaining Botting, Greater Than 2 in. In B6.40, B6.50, B6.40, B6.50, Diameter B6.180, B6.190 B6.180, B6.190 Examination Category B-G-2, Pressure B7.10, B7.20, B7.10, B7.20, Retaining Bolting,2 in. And Less in 87.30, B7.50, B7.30, B7.50, Diameter B7.60, B7.70 B7.60, B7.70 Examination Category B-J, Pressure B9.10, B9.11, B9.10, 89.11, Retaining Welds in Piping B9.21, B9.31, B9.21, B9.31, B9.32, B9.40. Note B9.32, B9.40.

1d required 25% Scheduled 27.7%

(surface and for surface and volumetric) of 31.3% for components be volumetric examined examinations

Figuro 6 Scheduled Examinations Exceed Requirements for Prairie Island Unit 2 ISI Third Ten Year ISI Third Ten Year interval items interval items PI Unit 2 - 89 Section XI Required to be Scheduled to be ASME Code Category examir,ed Examined Examination Category B-L-1, Pressure B12.10, B12.20 B12.10, B12.20 Retaining Welds In Pump Casings; B-M-1, Pressure Retaining Welds in Valve Bodies; B-L-2, Pump Casings; B-M-2, Valve Bodies Examination Category B-N-1, Interior Of B13.10, B13.50, B13.10, B13.50, Reactor Vessel. B-N-2, Integrally Welded B13.60, B13.70 B13.60, B13.70 Core Support Structures And interior Attachments To Reactor Vessels. B N-3, Removable Core Support Structures Examination Category B-0, Pressure B14.10 B14.10

__ Retaining Welds in Control Rod Housings Examination Category B-P, Pressure B15.10, B15.11, B15.10, B15.11, Retaining Components Leakage Test, Done per code Hydrostatic Test except where relief has been granted Examination Category C-A, Pressure C1.10, C1.20, C1.10, C1.20, Retaining Welds in Pressure Vessels C1.30 C1.30 Examination Category C-B, Pressure C2.11, C2.21, C2.11, C2.21, Retaining Nozzle Welds in Vessels C2.22 C2.22 Examination Category C-F-1, Pressure C5.10, C5.11, C5.10, C5.11, Retaining Welds in Austenitic Stainless C5.21, C5.30, C5.21, C5.30, Steel Or High Alloy Piping C5.41. Note 2: The C5.41. Scheduled welds selected for 9.2% for surface examination shall and 9.1% for include 7.5%, volumetric examinations Exainination Category C-F-2, Pressure C5.50, C5.50 & C5.50, C5.50 &

Retaining Welds in Carbon or Low Alloy HELB, C5.51, HELB, C5.51, Steel Piping C5.80, C5.81. Note C5.80, C5.81, 2: The welds Scheduled 22% for selected for surface and 23.2%

examination shall for volumetric include 7.5%, examinations

Figura 6 Scheduled Examinations Exceed Requirements for Prairie Island Unit 2 ISI Third Ten Year ISI Third Ten Year interval items Interval items PI Unit 2 - 89 Section XI Reautred to be Scheduled to be ASME Code Category examined Examined Examination Category C-G, Pressure C6.10 C6.10 Retaining Welds in Pumps And Valves Examination Category C-H, All Pressure C7.10, C7.20 C7.10, C7.20 Retaining Components Examination Category F-A, Supports F1.10. Extent of F1.10. Scheduled Examination - 25% 30.3% Visual Examination Examination Category F-A, Supports F1.20. Extent of F1.20. Scheduled Examination - 15% 45.6% Visual Examination

Figure 7 Service History for Monticello .

Page 1 of 2 At Monticello, in the ISl program there are:

. 464 components that require visual examinations (VT) .,

. 349 components that require volumetric examination (UT)

. ~ 366 components that require surface examinations (MT & PT)

Of those, ISI ultrasonic examinations have found no flaws that indicate service degradation of the pressure boundary. All of the flaws found by ISI to date in the pressure boundary are believed to be fabrication related.

I Examples of the flaws reported in ISI NDE reports are:

e linear volumetric indication in the reactor vessel head to flange weld

. pitting found on the reactor head washers e items that do not strictly match drawing configuration e missing or loose bolts e items with insufficient thread engagement This experience would lead to the conclusion that service related fatigue cracking or stress corrosion cracking is either not an active mechanism of service degradation or degradation is currently below the limits of detectability. These ISI results are not surprising compared with the rest of the BWR fleet and Monticello's operating experience.

To summarize, the examinations in Monticello's ISI program that are limited due to geometry, materials or design are es follows (See Figure 1, Columns 3,4, & 5):

. 4 visual examinations e 30 surface examinations

. 407 volumetric examinations For visual examinations (See Figure 1, Column 6):

. 3 examinations are limited but more '.han 90 % of the surface has been examined

. . 1 examination is lim!ted but between 25% and 50% of the surface has been examined For volumetric examinations (See Figure 1, Column 7):

~

. 140 are limited but more than 90% of the volume has been completely examined ,

. 142 are limited but between 75% and 90% of the volume has been completely- ,

examined _ _

. -- 87 are limited but between 50% and 75% of the volume has been completely l examined l

Figure 7 Service History for Monticello Page 2 of 2 e 14 is limited but between 25% and 50% of the volume has been completely examined

. 24 are limited and less than 25% of the volume has been completely examined For surface examinations (See Figure 1, Column 8):

. 6 are limited but more than 90% of the surface has been examined

. 7 are limitad but between 75% and 90% of the surface has been examined

. 13 are limited but between 50% and 75% of the surface has been examined

. 1 is limited but between 25% and 50 of the surface has been examined

. 3 are limited and less than 25% of the surface has been examined With volumetric examinations most of the limitations are due to the inability to interrogate the volume required with all of the required scans. Usually it is not possible to perform one of the axial scans due to interference with the component. However, axial scans from the other direction have been done in most cases so the only potential undetectable flaws are those oriented in such a way that they are only detectable with .

the scan that was not performed.

Figure 8 Service History for Prairie Island Unit 1  :

Page 1 of 2 At Prairie Island 1, in the ISI program there are:

. 430 components that require visual examinations (VT)

. 200 components that require volumetric examination (UT)

. 323 components that require surface examinations (MT & PT)

Of those, ISI ultrasonic examinations have found no flaws that indicate service degradation of the pressure boundary. All of the flaws found by ISI to date in the pressure boundary are believed to be fabrication related.

Examples of the flaws reported in ISI NDE reports are:

. linear volumetric indications in the steam generator shell welds, transition cone and tube sheet to shell welds

. linear volumetric indication in the reactor head to flange weld

. linear surface indications in stainless steel piping

, e linear MT indication in a reactor coolant pump flywheel (this was a Tech Spec required examination)

. items that do not strictly match drawing configuration

. missing or loose bolts

. items with insufficient thread engagement

. beni or damaged hangers This experience would lead to the conclusion that service related fatigue cracking or stress corrosion cracking is either not an active mechanism of service degradation or degradation is currently below the limits of detectability. These ISI results are not surprising compared with the rest of the PWR fleet and Prairie Island's operating experience.

To summarize, the examinations in Prairie Island Unit 1 ISI program that are limited due to geometry, materials or design are as follows (See Figure 2, Columns 3,4, & 5):

. 46 visual examinations

. 74 surface examinations e 339 volumetric examination For visual examinations (See Figure 2, Column 6):

. 20 are limited but more than 90% of the surface has been examined

. 6 are limited but between 75% and 90% of the surface has been examined

. .1 are limited but between 50% and 75% of the surface has been examineo

. 19 are limited and less than 25% of the surface has been examined

Figure 8 -

Service History for Prairie Islanet Unit 1 Page 2 of 2 For volumetric examinations (See Figure 2, Column 7):

. - 172 are limited but more than 90% of the volume has been completely examined

. .105 are limited but between 75% and 90% of the volume has been completely examined ,

e 46 are limited but between 50% and 75% of the volume has been completely examined e 16 are limited but between 25% and 50% of the volumo has been completely examined For surface examinations (See Figure 2, Column 8):

e 29 are limited but more than 90% of the surface has been examinea a 24 are limited but between 75% and 90% of the surface has been examined

. 5 are limited but between 50% and 75% of the surface has been examined

. 8 are limited but between 25% and 50 of the surface has been examined '

. 8 are limited and less than 25% of the surface has been examined With volumetric examinations most of the limitations are due to the inability to interrogate the volume required with all of the required scans. Usually it is not possible to perform one of the axial scans due to interference with the component. However, axial scans from the other direction have been done in most cases so the only potential undetectable flaws are those oriented in such a way that they are only detectable with the scan that was not performed.

d

Figure 9 Service History for Prairie Island Unit 2 Page 1 of 2 At Prairie Island 2, in the ISI program there are:

. - 281 components that require visual examinations (VT)

  • 206 components that require volumetric examination (UT) e 307 components that require surface examinations (MT & PT)

Of those, ISI ultrasonic examinations have found no flaws that indicate service degradation of the prescure boundary. All of the flaws found by ISI to date in the pressure boundarv are believed to be fabrication related.

Examples of the flaws reported in ISI NDE reports are:

  • linear volumetric indicatione in the steam generator shell welds, transition cone and tube sheet to shell welds e linear volumetric indication in the reactor head to flange weld

. linear surface indicatioris in stain!ess steel piping e linear MT indication in a reactor coolant pump flywheel (this was a Tech Spec required examination)

. items that do not strictly match drawing configuration

. boric acid residue on nuts e missing or loose bolts e items with insufficient thread engagement

  • bent or damaged hangers This experience would lead to the conclusion that service related fatigue cracking or stress corrosion cracking is either not an active mechanism of service degradation or degradation is currently below the limits of detectability. These ISI results are not surprising compared with the rest of the PWR fleet and Prairie Island's operating experience.-

' To summarize, the examinations in Prairie Island Unit 2 ISI program that are limited

'due to geometry, materials or design are as follows (See Figure 3, Coluinns 3,4, & 5):

. 38 visual examinations

e. 69 surface examinations

. 364 volumetric examination 1 For. visual examinations (See Figure 3, Column 6):  !

4 are limited but more than 90% of the surface has been examined

.- 7 are limited but between 75% and 90% of the surface has been examined

  • '3 are limited but between 50% and 75% of the surface has been examined j l

Figure 9 Service History for Prairie Island Unit 2 Page 2 of 2 e 2 are limited but between 25% and 50 of the surface has been examined

+ 22 are limited and less than 25% of the surface has been examined For volumetric examinations (See Figure 3, Column 7):

. 77 are limited but more than 90% of the volume has been completLy examined

. 184 are limited but between 75% and 90% of the volume has been completely examined e 84 are limited but between 50% and 75% of the volume has been completely examined

. 12 are limited but between 25% and 50% of the volume has been completely examined

. 7 are limited and less than 25% of the volume has been completely examined For surface examinations (See Figure 3, Column 8):

. 11 are limited but more than 90% of the surface has been examined

. 31 are limited but between 75% and 90% of the surface has been examined

. 12 are limited but between 50% and 75% of the surface has been examined

. 6 are limited but between 25% and 50 of the surface has been examined a 9 are !!mited and less than 25% of the surface has been examined With volumotric examinations most of the limitations are due to the inability to interrogate the volume required with all of the required scans. Usually it is not possible to perform one of the axial scans due to interference with the component. However, axial scans from the other direction have been done in most cases so the only potential undstoctable flaws are those oriented in such a way that they are only detectable with tha scan that was not performed.

(

p

1 I

Figure 10 Class 3 ASME Code Components Examinations for Monticello and Prairie Island ASME Class 3 Examinations still Class 3 Piping Examinations Piping Supports Completed scheduled in Third Supports in Third Ton Year Ten Year Interval Intervcl _

MONTICELLO 126 38 66 PRAIRIE ISLAND 442 101 23

_ UNIT 1 PRAIRIE ILLAND 298 18 23 UNIT 2 l

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