ML20210L971
| ML20210L971 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/15/1997 |
| From: | Hill W NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-263-97-07, 50-263-97-7, NUDOCS 9708210389 | |
| Download: ML20210L971 (13) | |
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Northem States Power Company Monticeho Nuclear oenerating Plant 2007 West Hwy 75 Monticello. Minnesota 55362-9637 August 15,1997 10 CFR Part 2 US Nuclear Rogulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR 22 Roply to Notico of Violation Contained in M, C Insocction Report No. 50-263/97007 R
Pursuant to the provls:ons of 10 CFR Pat 12, Section 2.201, our replies to the notico of violation and unresolved itom contained in your letter of July 16,1997, are provided in Attachments A and B, respectively, i
Attachment A, Repbe to Notico of Violation, contains the following new NRC commitment:
The Monticollo Maintenance Rulo Program Document EWI 05.02.01 shall bo revised to describo the now methodology used to develop the reliability critoria by October 15,1997.
Attachment D, Hoply to Unresolved item, contains the following two new NRC commitments that will bo incorporated into our ongoing Maintenance Rulo Program:
The Structures System Basis Document will be revised to clarity what constitutos the initial baseline inspection and what deficiencies should be loggod in the civil engineer's notebook.
Surveillanco 1385, Periodic StructuralInspection, will be revised to provido additional guidance for the civil engincor as to when a deficiency is acceptable, reforonco existing structural monitoring programs, clarify what comprised the initial baseline inspection, reviso Tablo 1 to divido largo structures into smaller sections, and to develop forms and guidance to improve the plant's permanont record by providing a separato inspection review shoot for each structure, a separato structural deficiency form, and a deficiency documentatinn process.
Please contact Marcus Voth, Program Manager Licensing (612 2715116),if you require further information.
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William J H
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Plant Mana or Monticello Nuclear Generating Plant PDR ADOCK 05000263 ff,fhfffffffhf;lf 9708210309 970015 O
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REPLY TO NOTICE OF VIOLATION Attachm:nt A Page 2 August 15,1997 1
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Regional Administrator Ill, NRC NRR Project Manager, NRC Sr Resident inspector, NRC State of Minnesota Attn: Kris Sanda J Silberg Attachments A Reply to Notice of Violation B. Reply to Unresolved item i
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e REPLY TO NOTICE OF VIOLATION Attachment A Page 3 August 15,1997 Violation:
10 CFR 50.65(a)(1) requiros, in part, the holders of an operating licenso shall monitor the performance or 04dition of structures, systems or components (SSCs), against licensoo-ostablishod goals, in a mannor sufficient to provide reasonablo assurance that such SSCs as definodin 10 CFR 50.65(b), are capablo of fulfilling theirintanded functions. Such goals shall bo established commensurate with safety. When the performance or condition of a SSC does not moot established goals, appropdato corrective action shall be taken.
10 CFR 50.65(a)(2) requltos, in part, that monitoring as specified in 10 CFR 50.65 paragraph (a)(1)is not required whero it has boon demonstrated that the performance or condition of a SSC is being offectively controlled through the performance of appropriato preventivo maintenanco, such that the SSC romains capable of performing its Intended function.
Contrary to 10 CFR 50.65(a)(2), as of July 10,1996, the time that the licensoo elected to not monitor the performanco or condition of certain SSCs against licensoo-established goals pursuant to the requiromonts of Section (a)(1), the licensoo had not demonstrated that the performance or condition of cortain SSCs within tho scope of to CFR 50.65 were being c!fectively controlled through the performanco of appropriato proventive maintenanco, as evidenced by the following oxamplos, each of which constituto a separato violation:
Violation Examolo A l
i A.
The liconsoo had not demonstrated that the performanco of the primary containment isolation system, the dicsol fuel oil standby pump, the reactor building componetut cooling water standbypump, andprimaiy radiation monitors were being offectively
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controlled through the performanco of appropriate proventivo maintenance in accordance with the requirements of 10 CFR 50.65(a)(2). Specifically, the licensee had failed to establish adequato measures to ovaluato tho ellectiveness of the performance of appropriato proventivo maintenance on those SSCs prior to placing theso SSCs under Section (a)(2). The licensco had performed proventivo maintenance on the SSCs; however, the appropriatonoss had not been adequately evaluated for SSC reliability or unavailability, whkh was necessary to demonstrate the effectiveness of preventive maintenanco. As a result, it was not demonstrated that the systems were controlled such that they remained capable of perforrning their intended function. Thereforo, the licensoe's technicalbasis forplacing those SSCs under the requirements of 10 CFR 50.65(a)(2) was inadequato and these standby SSCs should have boon monitored in accordance with 10 CFR 50.65(a)(1)(26397007-01(DRS)).
This is a Severity LevelIV violation (Supplomont I).
NSP Response to Violation NSP acknowledges the abovo Notice of Violation.
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REPLY TO NOTICE OF VIOLATION Attachment A Page 4 August 15,1997 Although these systems were not being monitored at the appropriate levelin accordance with the Maintenance Rule, they were being monitored by other plant programs and -
processes (such as the Work Order Process, the Recurring Problems Program, the Repeat Work / Rework Investigation Process, and the Condition Report Process) outside of the Maintenance Rule that would have identified ineffective maintenance.
Reason For Violation The systems identified in this violation were being monitored with plant level performance critoria instead of specific system or train level criteria such as unavailability or reliability because they had been classified as "normally operating." The standby functions of components in those systems that are not in continuous operation were not addressed. Since the performance monitoring for these systems was not at i
the appropriate level, NSP did not have a technical basis for classifying these systems as 10 CFR 50.65(a)(2) and, by default, should have clast.ified them as 10 CFR 50.65(a)(1).
For systems such as Diesel Oil Stcrage (DOL) and Reactor Building Closed Cooling Water (RBC) that are normally operating and have a redundant pump in standby, the Expert Panel did not require the development of specific performance critoria for the standby pump in accordance with the guidance of NUMARC 93-01. The Panel focused on the fact that the function was normally operating and that a loss of function would be detected by plant level performance criteria. The standby pump was inadvertently overlooked. Therefore, the Panel wrongly concluded that monitoring at the plant level wts sufficient.
The Expert Panel initially conaldered the Primary Containment Isolation System (PCIS) and the Process Radiation Monitoring System (PRM) to be normally operating systems since these systems are normally energized and could be considered to be " monitoring" their input signals. Therefore, these systems were being monitored with plant level performance criteria in accordance with the guidance of NUMARC 93 01 for normally operating, non-risk significant systems. However, Revision 2 of Regulatory Guide 1.160, " Monitoring the E4ectiveness at Nuclear Power Plants," issued in March of 1997, provides more specific information on the definition of " standby." This expanded definition requires the PCIS and PRM systems to be considered Mandby" systems and not normally operating systems.
Corrective Action Taken and Gesults Achieved Specific performance criteria were developed for the stantLy structures, systems, or components (SSCs) identified above. The Expert Panel reviewed and approved these criteria on August 12,1997.
All other normally operating systems in the Maintenance Rule were reviewed to ensure that they are tru!y normally operating considering the guidance in Reg Guide 1.160, Revision 2, concerning the definition of " standby." The standby pump for the Stator Cooling System and the standby Emergency Seal Oil Pump (ESOP) for the Hydrogen Seal Oil System were identified as additional standby SSCs requiring specific l
e REPLY TO NOTICE OF VIOLATION Attachment A Page 5 August 15,1997 performance criteria. Specific criteria were developed for these components. The Expert Panel reviewed and approved these criteria on August 12,1997. No additional SSCs were identified during this review.
Historical performance data from July 10,1993 to the present was reviewed and compared to the performance criteria for preliminary (a)(1)/(a)(2) determinations. The preliminary determination shows that the (a)(1)/(a)(2) status of the affected systems remains unchanged. This means that none of the affected systems are exceeding their performance criteria. The Expert Panel will finalize these determinations at the meetlog to review the 3rd Quarter Maintenance Rule Periodic Assessment Report.
All of the above activities will be documented in the 3rd Quarter Maintenance Rule Periodic Assessment Report.
The Maintenance Rule Expert Panel reviewed the causes of this violation on June 26, 1997 as training to prevent recurrence.
Corrective Action to be Taken to Avoid Further Violations No further actions are required.
_Date When Full Comollance Will Be Achieved Full compliance has been achieved.
4 REPLY TO NOTICE OF VIOLATION Attachment A Page 6 August 15,1997 fjolation hamDIe B B.
Tho licensee had not demonstrated that the performance of the residual heat removal pumps, the residualheat removalservice waterpumps, the standby aervice water pumps, the safoty rollet valves, and the reactor core isolation cooling system were boing offact!vely controlled through the performanco of appropriate preventive maintonance in accordance with the requirements of to CFR 50.65(a)(2). Specifically, the licensee had failed to establish adequate measures to evaluate the effectiveness of the performance.
of appropriate proventivo maintenance on thoso high satoty (risk) significant SSCs prior to placing those SSCs under Section (a)(2). The measures were inadequate because there was not a technicaljustification to allow these SSCs to have reliability and availability measures that were inconsistent with the assumptions in the licensee's probabilistic risk assessment (PRA). The measures established for those SSCs would allow failuro rates that were between one and two orders of magnitude greater than the reliability and availability assumptions in the licensee's PRA. Thus, multiple failures or high unavailability would nood to occur before the licensee considers moving these SSCs to the (a)(1) category. As a result, it was not demonstrated that the systems were controlled such that they remained capable of performing their intended function.
Therefore, the licensee's technicalbasis forplacing these SSCs under the requirements i
of 10 CFR 50.65(a)(2) was inahquate and these high safety (risk) SSCs should have been monitored in accordance s 'th 10 CFR 50.65(a)(1) (263/97007-02(DRS)).
This is a Severity LovelIV violation (Supplement I).
NSP Response to V!olation NSP acknowledges the above Notice of Violation.
Reason For Violation The PRA sensitivity study which combined the effects of both the availability and reliability criteria did not adequately show that performance criteria were consistent with PRA assumptions.
Initial unavailability criteria were developed by calculating the delta probability change when a set change in baseline Core Damage Frequency (CDF) is divided by the Birnbaum value. The NEl/EPRI PSA Application Guide was used to select this sct change in baseline CDF of approximately 27% These initial values were then compared against similar BWR 3/4 Individual performance criteria to establish a draft set of performance criteria. A PRA sensitivity study was then performed on the draft performance criteria by requantifying the PRA with all SSC unavailabilities set at the perfermance criteria value to determine the overall effect and also the major core damage contributors. After several minor adjustments in the draft performance criteria, as a result of the sensitivity study, the criteria were reviewed by the Expert Panel for consistency between systems and appropriateness of the absolute values. The performance criteria were finally reviewed by the associated System Engineer and compared against Monticello historical data from January 1992 through June 1996. A
REPLY TO NOTICE OF VIOLATION Attachment A Page 7 August 15,1997 final PRA sensitivity study was then completed with the finalized unavailability performance criteria.
For the setting of the initial reliabilit'/ performance criteria for Maintenance Rule systems requiring this measure, the Expert Panel used the Emergency Diesel Generator (EDG)
System as a benchmark. This system already had a reliability program in place due to the requirements of Reg Guide 1.155, Reg Guide 1.9, and NUMARC 87 00 " Station Blackout". The Monticello EDGs have a targeted reliability of 95% NUMARC 87-00 specifies acceptable reliability values for a 95% reliable Diesel Generator System as follows:
3 failures out of last 20 tries, 5 failures out of last 50 tries, and 8 failures out of last 100 tries.
l These values and the relative importance of the Diesel Generator were compared to l
other Monticello systems to establish an initial reliability maintenance preventable l
functional failure (MPFF) value, initial values were then reviewed by the Expert Panel for consistency betwoon systems and appropriateness of absolute values. The performance criteria were reviewed by the associated System Engineer and compared against Monticello historical data from January 1992 through June 1996. Finally, MPFF values were compared against system unavailability values for overall consistency.
Finally, a sensitivity study was performed using both the availability and reliability criteria to evaluate the combined effect on CDF. NSP believed that this study provided the
" link" needed to show that the performance criteria were consistent with the assumptions used in the PRA. However, because the methodology used in the study did not use the reliability performance criteria values in all cases, NSP was not able to adequately show that the performance criteria were consistent with the PRA assumptions.
Corrective Action Taken and Results Achieved New reliability criteria for the risk-significant systems modelod in the PRA were derived based on a methodology discussed with the NRC PRA inspectors during the course of their Maintenance Rule inspection. This methodology used the Poisson Distribution for normally operating systems and the Binomial Distribution for standby systems to calculate the probability of incurring 0,1,2,3 or more failures over a two year period.
This method uses the actual PRA assumptions to calculate the reliability criteria. If the probability of a given number of failures over the two-year time period was greater than or equal to five percent, this number of failures was the maximum acceptable number of failures over the two-year time period. Reliability criteria for most systems were able to be developed using only this methodology. However, for systems where the results of these calculations did not provide predictive criteria (such as zero allowed failures) or were not realistic, a sensitivity study was performed with these systems using more appropriate reliability criteria. This sensitivity study also included the cifects of all the
REPLY TO NOTICE OF VIOLATION Attachment A Page 8 August 15,1997 availability criteria set to their maximum values for risk significant systems modeled in the PRA. - The result was a CDF of less that 1E-4 which is generally recognized as the acceptance limit. The Expert Panel reviewed anu approved the new performance criteria on August 12,1997.
Historical performance data from July 10,1993 to the present was reviewed and compared to the performance ciiteria for preliminary (a)(1)/(a)(2) determinations. The preliminary determination shows that the (a)(1)/(a)(2) status of the offected systems remains unchanged. This means that none of the affected systems are exceMog their performance criteria, The Expert Panel will finalizo these determinations at the meeting to review the 3rd Quarter Maintenance Rule Periodic Assessment Report.
All of the above activities will be documented in the 3rd Quarter Maintenance Rule Periodic Assessment Report.
The Maintenance Rule Expert Panel reviewed the causes of this violation on June 26, 1997 as training to prevent recurrence.
Corrective Action to be Taken to Avoid Further Violations l
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The Monticello Maintenance Rule Program Document EWI 05.02.01, Sections 6.3.2 and l
6.3.3 shall be revised to describe the new methodology used to develop the reliability
. criteria by October 15,1997.
Date When Full Comollance Will Be Achieved Full compliance has been achieved.
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REPLY TO UNRESOLVED ITEM Attichm:nt B Page 1 August 15,1997 Unresolveditem Tanks, supports, buildings, and onclosures were adequately scoped as structures under the MR. The guidanco and documentation of the structure monitoring program were weak and it was unclearif the Intent of the MR and the NUMARC guidance was being met with the program currently in place.
Monticello Response NSP believes that the existing program meets the intent of the Maintenance Rule and the NUMARC/NEl guidance since the inspection procedure and system basis document were developed directly from the Rule and the NUMARC/NEl guidance documents available in June of 1996. The initial baseline inspection was performed in June of 1996 for the in-scope structures. NSP agrees that the guidance and documentation for the Maintenance Rule Structural Monitoring Program should be enhanced.
Maintenance Rule Structural Monitorina Procram, Enhancements NSP believes that we were in full compliance at the time of the NRC's inspection.
However, further actions have and will be taken to enhance the program and address the inspectors' concerns. These actions are outlined below.
Monticello Surveillance Procedure 1385 was revised from revision 0 to revision 1 to include more specific items to be inspected, more specific guidance as to what to inspect for, and the NRC comments on NEl 96-03, Revision D, as documented in the letter datt,d October 1,1996 from Thomas T. Martin (NRC) to Thomas E. Tipton (NEI).
This surveillance was approved on August 12,1997.
The Structures System Basis Document will be revised to eliminate the discrepancy over what comprised the initial baseline inspection. Specific guidance will also be added to address what type of deficiencies should be kept in the civil engineer's deficiency notebook.
Surveillance 1385, Periodic Structural Inspection, will be revised to include criteria, where possible, to assist the civil engineer in determining if a deficiency is acceptable.
References to existing structural monitoring programs will be incorporated. The definition of what comprised the initial baseline inspection will be clarified. Table 1 of this surveillance will be revised to divide large structures, such as the Reactor and Turbine Buildings, into sections or zones to aid in the performance and review of the inspection. Each structure will also have its own inspection review sheet instead of all being located on the same sheet. Also, a separate structural deficiency form shall be added with guidance as to when it should be used. This will provide a specific process for documenting deficiencies and ensure that they become part of the plant's permanent records.
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hEPLY TO UNRESOLVED ITEM Att chm:nt B Prge 2 August 15,1997 Response to inspection Report items Section M1.6.b.4 of NRC Inspection Report 50-263/97007 provided comments concerning the Monticello Maintenance Rule Structural Monitoring Program. In response, NSP provides the following additional information concerning these inspection comments. NSP feels that this additionalinformation further supports our position that the program meets the intent of the Rule and the NUMARC/NEl guidance.
The inspection report states, The completed baseline inspections consisred of a list of ten structures that were considered as either acceptable, acceptable with deficiencies, or unacceptable. The licensee identified the condition of all structures within scope of the MR as acceptable; no deficiencies with structures were documented in the MR program. Discussions with the resident inspector staff revealed several deficiencies where operability evaluations were in place or structural repairs had been completed subsequent to the baseline inspection.
A review of all work orders and condition reports for the Structures System was conducted for the time period between the completion of the structural baseline inspection (June 1996) and the NRC Maintenance Inspection (June 1997). No structural deficiencies were identified that should have been dccumented in the Maintenance Rule Structural Monitoring Program. Upon further discussion with the NRC Senior Resident inspector on this issue, it was determined that two deficiencies were identified to the NRC Maintenance Rule inspectors. The first was the possible inability of the Reactor Building steel superstructure above the refuel floor to withstand tornado loading conditions. The second was a damaged secondary containment penetration seal between the Reactor Water Cleanup (RWCU) Heat Exchanger Room and the pipe chase to the Cleanup Phcse Separator Tank Room. Each will be explained in detail below.
Analyses performed to resolve a design basis document (DBD) follow on item (FOI) showed that the Reactor Building steel superstructure above the refuel floor may not be able to withstand the tornado winds described in the licensing basis of the plant. This was not due to degradation, but due to an original design oversight. As a result, NSP decided to upgrade the superstructure via a modification. This modification was performed after the hitial structural baseline inspection. However, from a Maintenance Rule standpoint, this is clearly an original design deficiency. No preventive maintenance activities would have prevented this outcome, and the baseline walkdown would not be expected to identify a design deficiency of this nature. Since the purpose of the Maintenance Rule is to monitor the effectiveness of maintenance, this deficiency is outside the scope of the Maintenance Rule and need not be documented as a deficiency in the Maintenance Rule Structural Monitoring Program. The upgraded superstructure will be inspected for degradation periodically as part of the normal surveillance activities of the Maintenance Rule Structural Monitoring Program just as the original superstructure was inspected.
The damaged penetration seal was discovered on April 19,1996 and a condition report was initiated to document this discovery. Since the primary function of this penetration
REPLY TO UNRESOLVED ITEM Attichm:nt B Page 3 August 15,1997 was to provide a secondary containment barrier and the damage was not of a structural nature, the damaged penetration seal would have bee 1 counted as a maintenance preventable functional failure (MPFF) against the SecorMary Containment System if it had caused secondary containment to be inoperable. Ho,'ever, the seal did not cause a loss of secondary containment. The leakage from this penetration was within the allowed leakage for secondary containment and was included in the leakage measured by Monticello Surveillance Test 0151 1, Secondary Containment Capability Test.
Therefore, this event was not documented in the Maintenance Rule Structural Monitoring Program or in the Maintenance Rule Access Database under the Secondary Containment System.
In summary, neither of these deficiencies were of a type requiring documentation in the Maintenance Rule Structural Monitoring Program.
The inspection report states, The structural engineer stated that a notebook was kept ofidentified deficiencies; however, the inspectors were unclear how this was incorporated into the MR program.
The Structures System Basis Document, Revision 1, states that the civil engineer responsible for the periodic structuralinspection maintains a working copy of noted l
deficiencies incit ding photos, notes, LERs, Condition Reports, etc. for reference in future structural inspections. Although future inspections will review all areas of in-scope structures, the notebook information is intended to ensure that future inspections specifically include the areas where deficiencies have already been identified. Specific guidance on what items were to be included in the deficiency notebook was left to the discretion of the civil engineer.
The inspection report states, Credit was taken for structuralinspections performedin 1986 as part of the baseline inspections. This was consideredinappropriate since these inspections were performed over 10 years ago and the program required inspecting the condition of major structures on a 5-year frequency.
An evaluation of critical concrete structures was performed in 1986 as part of Monticello's Pilot Plant License Renewal efforts in cooperation with the Electric Power Research Institute (EPRI) and the U. S. Department of Energy (reference EPRI TR-103842s " Class l Structures License Renewal Industry Report" and NSPE-12-4312
" PLEX Topical Report for Critical Concrete Structures". These evaluations were considered inputs to the structural initial baseline evaluation since they were used to provide insights as to where structural concrete problems may exist. A structural walkdown by a registered civil engineer was completed in June 1996 of all structures included within the scope of the Maintenance Rule at that time. This walkdown, using the inputs from the 1986 evaluations, fulfilled the " initial baseline" inspection requirement for the in-scope structures as discussed in NEl 96 03.
' REPLY TO UNRESOLVED ITEM Att:chmInt B Page 4 August 15,1997 The Monticello Maintenance Rule Program Document, EWi 05.02.01, Revision 1, states in Section 6.4 that the initial baseline is comprised of both the 1986 evaluations and the walkdcwn completed in June of 1996. The Structures System Basis Document, Revision 1, incorrectly states that the 1986 reviews themselves will be used as the initial baseline inspection although both the 1986 evaluations and the 1996 walkdown are discussed in detail. NSP acknowledges that these documents do not provide a clear understanding of what comprised the initial baseline inspection. These documents will be revised to clarify this issue as stated earlier in this response.
The inspection report states, Regulatory Guide 1.160, Revision 2, stated that a structure would be considered (a)(1)if either (1) degradation was to the extent that the structure may not meet its design basis or (2) the structure had degraded to the extent, that, if the degradation were allowed to continue uncorrected until the next normally scheduled assessment, the structure may not meet its design basis.
The program did not appear to require placing a structure into (a)(1) based on case (2). The licensee stated that the program would be revised to address this issue.
The performance criteria section of the Structures System Basis Document, Revision 1, defines the classification " acceptable with deficiencies" as structures which are capable of performing their structural functione...but are degraded or have deficiencies which could deteriorate to an unacceptable condition if not analyzed or corrected prior to the next scheduled examination. The basis document goes on to require that any structures identified as being " acceptable with deficiencies" be reviewed by the Maintenance Rule Expert Panel for final (a)(1)/(a)(2) determination and appropriate goal setting action plans. The Monticello Maintenance Rule Program Document, EWl-05.02.01, Revision 1, Section 6.4, states, "A structure should be considered for (a)(1) status if it is listed as " unacceptable" or if it is exhibiting a trend that will lead to
" unacceptable" status." These examples indicate that the program did account for placing a structure into (a)(1) based on case (2) above.
The inspection report states, The structure monitoring procedure did not provide guidance to ensure all areas of a structure were inspected. For example, with regard to the turbine building and the reactor building, it was unclear how each room or zone within these buildings were inspected. The procedure also did not give any guidance as to what should be inspected (e.g., component supports, block walls, roofs, structural steel) or criteria defining what was acceptable for these inspections.
Monticello Surveillance Procedure 1385, Periodic Structural Inspection, Revision 0, listed the structures to be inspected in Table 1, inspection Results. By listing the structure in this table, it was understood that the entire structure would be inspected by the civil engineer performing the inspection. The surveillance also specifically listed concrete, masonry walls, structural steel, roof systems, siding, windows / doors, earthen structures / dams, settlement, and structural / seismic gaps as items to be inspected.
Defining what was an acceptable deficiency or defect was left to the expertise of the
' REPLY TO UNRESOLVED ITEM Attachmtnt B
.__ Page 5 August 15,1997 civil engineer._ For this specific reason, a registered professional engineer in the state of Minnesota was a mandatory qualificatior, for the individual performing the inspection.
Surveillance Procedure 1385 will be revised as discussed earlier in this response.
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