IR 05000315/1986005
| ML17326B216 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/28/1986 |
| From: | Guthrie S, Hard J, Hasse R, Hawkins F, Mccormickbarge, Walker H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17326B214 | List: |
| References | |
| 50-315-86-05, 50-315-86-5, 50-316-86-05, 50-316-86-5, NUDOCS 8604020468 | |
| Download: ML17326B216 (29) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/86005(DRS);
50-316/86005(DRS)
Docket Nos. 50-315; 50-316 Licensee:
American Electric Power Service Corporation Indiana and Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 Licenses No. DPR-58; DPR-74 Facility Name:
D.
C.
Cook Nuclear Plant, Units 1 and
Inspection At:
D.
C.
Cook Site, Bridgman, MI Inspection Conducted:
February 3 through February 14, 1986 (onsite)
February 18 through March 3, 1986 (in-office)
Inspectors:
R.
asse Date H. Walker Date S. Guthrie Da e
Z7 8(.
Da e M. McCormi k-Barger
D e
Approved By:
F. Hawkins, Chief equality Assurance Programs Section Ins ection Summar te Ins ection on Februar 3 throu h March 3, 1986 Re orts No. 50-315/86005 DRS
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The inspection was conducted in accordance with NRC Inspection Procedure No. 37702.
86OeOaO<W euonym PDR ADOCK 05000815
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Results:
Three violations with five examples were identified - Criterion III
~far ure to adequately control design changes)
Paragraphs 2.b.(2)
and 2.b.(4);
Criterion XVI (failure to promptly identify conditions adverse to quality and implement corrective actions)
Paragraphs 2.b.(8)(a)
and 2.b.(8)(b);
and
CFR 50.59 (failure to perform safety evaluation)
Paragraph 2.b.(8)(c).
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Persons Contacted DETAILS Indiana and Michi an Electric Com an ( I&M and American Electric Power Services Cor oration EPSC
- W. Smith, Jr., Plant Manager
- A. Blind, Assistant Plant Manager
- T. Beilman, Planning Superintendent
- J. Riechling, QC
- R. Hennen, Technical Engineer
- D. McAlhany, QA/AEPSC
- M. Horvath, QA/AEPSC
- D. Yount, Document Control
- R. Allen, Planning
- B. Bradley, Technical Engineering
- T. Adams, I8M Construction
- H. Wood, Planning
- J. Hoss, Planning
- R. Russell, Planning
- L. Gibson, Supervisor, Technical Engineering
- E. Abshagen, Planning
- B. Svensson, Assistant Plant Manager, Operations
- J. McElligot, QA/AEPSC
- P. Barrett, Safety and Licensing/AEPSC USNRC 2.
- C. Hehl, Chief, Reactor Projects Section 2A
- B. Jorgensen, Senior Resident Inspector
- C. Wolfsen, Resident Inspector Other personnel were contacted as a matter of routine during the inspection.
- Denotes those attending the exit interview on February 14, 1986.
De~si n
C~han es and Modifications The purpose of this inspection was to provide an in-depth review of the implementation of the licensee's design change and modification program.
The inspection covered permanent modifications, temporary modifications, and set point changes.
a.
Ins ection Methodolo The inspectors reviewed design change packages for completeness, adequacy, and adherence to applicable procedures.
The design change coordinators (DCC) were interviewed as appropriate.
The inspectors also performed a walkdown of installed modifications, where possible, to ensure conformance of the installation to the desig Hl
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b.
Ins ection Results (2)
Procedure PMI-5040 Revision 7, "Oesi
~n Chan e Contr~ol Pro ram" The inspector noted that Paragraph 4.3. 1 of Revision 7 to PNI-5040 allowed both steps and sections of the procedure to be waived by the responsible department superintendent.
No limita-tions were placed on the steps that could be waived and no guidance was provided for conditions that might require these waivers.
Under the present wording, regulatory requirements, including control functions could be waived.
The inspector found no instances where this waiver had been used.
Licensee personnel stated that this paragraph would be changed to ensure that adequate controls are maintained and that regulatory requirements are met.
This item is unresolved pending review of both the revised procedure and any waivers issued per this paragraph prior to its revision (315/86005-01; 316/86005-01).
In general, Revision 7 represented a considerable improvement over Revision
from the standpoint of administrative and technical completeness.
Re uest for Chan
~e RFC Packa e DC-12-2665, "Install CVCS Cross-tie Pi in "
(3)
The inspector noted that a pipe support (Dwg. No. 1-ASI-L-923)
had been re-designed by red-lining the drawing.
Although two individuals had initialed the red-lined changes, the inspector was informed that the change was performed jointly by these individuals.
There was no indication that either an independent design review or verification had been performed on the pipe support re-design.
The RFC records also indicated that two pipe lugs of a size different from the original design had been installed adjacent to the support without a design change, red-lining of the drawing or independent design review; however, the lug installation was later replaced due to surface cracks in the pipe.
The later re-installation of the lugs was performed with a properly red-lined drawing, but also without an independent design review.
These failures to provide independent design verification and document a design change are in violation of
CFR 50, Appendix B, Criterion III (315/86005-02A; 316/86005-02A).
A Audit A-85-04 During the review of audit records for gA audit gA-85-04, the inspector noted that a checklist item dealing with the review for operability of modified systems and the updating of control room drawings had been omitted from the audit.
The inspector was informed that this was due to insufficient time.
Because of the importance of this item, the inspector was concerned about the omission.
Before the end of the inspection, the inspector was provided a draft copy of a letter indicating that a surveillance
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had been scheduled to cover the omitted checklist items.
This letter also indicated that future checklist items not audited would be reviewed for significance by either the lead auditor or the Plant gA Supervisor.
The inspector has no further concerns in this area.
SSC CC-12-2656.
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The licensee replaced certain carbon steel components in the steam generator blowdown system with stainless steel components.
These components were the wear plates inside both the normal and startup flash tanks, plus sections of inlet piping for the normal flash tanks.
During performance of the modification work, crafts-men discovered that the wear plate in the Unit 2 normal flash tank had already been changed.
This earlier change was not reflected in the as-built plant records.
Also, at the time of the inspection, the inspector noted that even though the design change implemented by DC-12-2656 had been completed in tray 1984 and June 1985 for Units 2 and 1 respectively, the tank drawings had not been updated to show the revised material of the wear plates.
These failures to maintain up-to-date drawings is considered to be in violation of 10 CFR 50, Appendix B, Criterion I II (315/86005-02B; 316/86005-02B).
RFC DC-12-2778,
"RVLIS U rade" (6)
The upgrade referred to in this change involved the removal of unnecessary bypass valves in the RVLIS DP transmitter racks and of certain wires in the RVLIS circuits which activate alarm annunciators.
Plant Control and Instrumentation (C8 I) personnel told the inspector that when the technicians examined the RVLIS cabinets in preparation for removing the annunciator wiring, they discovered that the wiring appeared to have never been present.
In addition, some mislabeling of sensing line RTD's became apparent.
Because of these discrepancies, C&I personnel are conducting a significant program of preparing accurate as-built one-line wiring diagrams for the installed RVLIS systems.
The inspector also noted that power to RVLIS was being provided temporarily from a lighting panel.
This is being done until an appropriate time arises for switching RVLIS to the inverters (see Paragraph 2.b.(7).
At the time of this inspection, the RVLIS for neither unit could be considered operable.
The inspector considered this to be acceptable after discussions with NRR representatives.
RFC DC-12-2713
"Coolin Provisions for GRID Cabinets" Critical Reactor Instrument Distribution (GRID) systems are comprised by four inverters in each unit.
High temperatures in these CRIDs, especially during warm summer months, have caused inverter failures in the past.
These failures in turn have
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resulted in unplanned reactor shut downs.
This change provided supplemental cooling air to each of the eight CRIDs, plus other ventilation modifications to reduce the ambient room air temperature.
The inspector had no concerns relative to this modification.
RFC DC-01-2766,
"Re lace CRID Inverters" The 5 kva CRID inverters are being replaced with 7 1/2 kva units.
This is necessary because of additional loads being placed on the regulated AC system which is powered by the CRID system.
The Unit 1 inverters have been replaced and the Unit 2 inverters are to be replaced during the next outage.
During review of the design change package, the inspector noted that a minor change to a footnote in the Technical Specifications was required because of the modified design.
The inspector discussed the matter with the NRR Licensing Project Manager (LPM) for D. C. Cook, since
CFR 50.59 requires prior Comoission approval if a facility change involves a change in the Technical Specifications.
The LPM stated that he had been working closely with the licensee on the inverter upgrade and that he had been reviewing their plans.
On this basis, the inspector has no further questions on this matter.
RFC DC-12-2760
"Re lace Ventilation Damper Operators in Ductwork
R These operators are located outside the duct in the vertical segment of the ventilation shaft.
The RFC was written on an expedited basis to allow for procurement of parts with long lead times, as documented by a Temporary Waiver Letter (TWL)
dated August 10, 1984.
The design change represents the licensee's long term corrective action for a series of Diesel Generator (DG) inverter failures in March 1983, which rendered the DG inoperable (Technical Specifica-tion 3.8. 1. 1.6.)
The licensee reported in the Licensee Event Report (LER) 315/83-023 (CR 1-3-83-262) that the inverter apparently failed because of high DG room temperatures (up to 108 degrees F) caused by ventilation damper failure.
Dampers had been normally closed to prevent freezing air from entering the DG room and opened as needed.
LER investigation revealed that the dampers were wired shut and the motor operators electrically disconnected.
In response to an NRC position, as documented in Condition Report (CR) No. 1-03-83-263, that a
DG must be declared inoperable if supply ventilation dampers are incapable of opening, the dampers were wired open.
The inspector determined that neither the Design Change Coordinator (DCC) nor operating shift personnel were either aware of the status of the damper or able to identify where the degraded status of the system was documented.
The licensee later produced
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evidence that procedures for testing and operation of the DG had been modified by way of a temporary change to reflect the degraded condition.
Maintenance department personnel informed the inspector that, while the dampers are currently wired opened, a
piece of herculite material was placed over the supply ventilation damper downstream of the fan to prevent freezing air from entering the room.
The herculite was equipped with a rope to be used by the operator for removal when the DG's were operating.
Operators interviewed on shift by the inspector were not aware of the need to remove the covering.
This created the potential for loss of the DG on high temperature in the DG room during future starts.
A subsequent check by the inspector showed that operators on shift were made aware of the condition during the inspection.
The inspector also identified the following additional concerns related to this modification:
(a )
Problems with Unit
DG room ventilation and the determination that the damper motors were undersized and seriously degraded are documented in correspondence dating back to April 1982.
The inspector expressed his concern over the timely processing of CR's related to this problem after noting that CR No. 1-03-83-263, written in March 1983, was not closed until August 1985 when RFC-12-2760 was issued to replace the damper motors.
However, the modification is still not completed.
The inspector noted that closure of the CR by assignment of an RFC number, which may then be delayed indefinitely, does not provide adequate control over the resolution of documented deficiencies.
In this instance, lack of action on both resolutions of CR-1-3-83-263 and performance of RFC-12-2760 has left the licensee vulnerable to further DG unavailability for approximately three years.
This failure to promptly take action to correct the adverse damper motor deficiencies, is in violation of 10 CFR 50, Appendix B, Criterion XVI (315/86005-03A).
On March 3, 1986, subsequent to the onsite inspection, the licensee informed the inspector that a completion date of June 1, 1986, had been established for damper motor replacement.
(b)
The licensee's failure to identify the root cause of the inverter failures appears to have led in part to inappro-priate long term corrective action.
Following the immediate response of operators to restore breaker alignments and verify operability of the other DG, corrective action as reported on LER 83-023 (March 1983), consisted of opening ventilation supply dampers to reduce DG room temperatures.
Corrective action for the related CR was not documented until January 11, 1985, when RFC-12-2760 was written to replace the degraded damper motors.
This failure to identify the degraded damper motors as the root cause of the inverter failures and take appropriate corrective action is in violation of 10 CFR 50, Appendix B, Criterion XVI (315/86005-03B).
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(c)
(d)
(e)
The inspector expressed concern that the use of herculite as a substitute for permanently installed dampers, which are motor operated and automatically controlled as a function of DG operations, constitutes a temporary modification.
However, the licensee had performed no safety evaluation for this modification as required by 10 CFR 50.59.
Considering the ventilation systems demonstrated ability to cause a loss of the OG, an unreviewed safety question may well exist which did not receive the safety review required by 10 CFR 50.59.
Failure to recognize changes to the OG room ventilation supply as a temporary modification and perform a safety evaluation is in violation of 10 CFR 50.59 (315/86005-04).
The inspector reviewed the recent revision to PYiI-2140,
"Temporary Modifications" to determine if changes to handling of temporary modifications would have prevented the unreviewed modification from being installed.
The provisions of the current revision clearly require performance of a
CFR 50.59 safety evaluation prior to performing any temporary modification to operable systems or equipment.
The evaluation requires Plant Nuclear Safety Review Committee (PNSRC) concurrence.
The inspector expressed concern that the timeliness of corrective action to resolve a problem involving a piece of equipment essential to plant safety was excessively long.
The original incident occurred in t1arch 1983 and a temporary waiver letter initiating the corrective action was issued in August 1984.
Purchase orders were issued at that time, and parts were received in June 1985.
The RFC is not outage related, and it has not been performed because the DCC was directed to address modifications having a higher priority, particularly those satisfying commitments to NRC.
The inspector discussed with licensee personnel the current methods of establishing design modification priorities.
Currently, top priority is given to those design changes which satisfy commitments to the NRC.
The system does not presently assess the relative safety significance of other modifications which may have greater impact on operating safety, but which are not related to an NRC commitment.
The inspector stressed the need for an integrated schedule with sufficient flexibilityto permit prioritization of design changes on the basis of the modification's contribution to safety.
As part of the inspector's review of this RFC, an analysis of the licensee's receipt inspection procedures and practices was conducted.
The inspector noted that upon receipt, material and equipment is only inspected for visual damage.
The receipt inspectors, who are part of the stores
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department, have the option of ordering tests from Quality Control inspectors if in their judgement a test is warranted.
Certification of materials and testing traditionally performed as a
QA function is performed and the appropriate documentation supplied by the vendor.
The inspector reviewed the QA Department's procedures to verify the vendor's qualifications to provide tests performed in lieu of a receipt inspection program.
While the licensee's certification procedure appears adequate to determine if a vendor's quality assurance program provides an acceptable level of control over product manufacture, it does not address the qualifications of the vendor to perform final product testing not,associated with the manufacturing process.
Although licensee procurement specifications clearly list tests and certifications to be provided by the vendor as part of the procurement process, there is nothing to indicate whether the vendor is qualified to perform those tests on behalf of the licensee.
The QA process which evaluates vendors for inclusion on the Quality Suppliers List (QSL) does not presently review the procurement specification to determine vendor capability to perform the tests specified.
This item is considered open (315/86005-05).
The inspector reviewed the licensee's procedure for handling CR's (Plant Manager Instruction (PMI)-7030, Revisions
and 7 ).
Revision 7 had recently been issued.
Based on the review, the inspector expressed concern that the procedures did not provide adequate instructions to ensure the identification of an adverse condition's root cause.
Nore specifically, while the licensee makes a thorough effort to document the particulars of a condition the licensee's effort to determine the cause is usually incomplete.
While both Revisions 6 and 7 of PNI-7030 have checklists of cause codes, such as personnel error, and component failure, the CR's are deficient in root cause determination, except in the case of personnel error.
Licensee personnel who were interviewed were of the opinion that the incident description sufficiently identified the cause; they made no distinction between the process or condition that brought the event to the licensee's attention and the root cause that created that process or condition.
The CR's related to this event described the loss of DG availability on March 10, 1983, due to inverter failure and reported to management that the fuses blew when the room's temperature became excessively high after the dampers were wired shut.
In actuality, the root cause was unreliable damper operators (which resulted in the dampers being wired
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shut).
Identification of the true root cause at the time of the incident would have likely caused management to promptly pursue long term corrective action in the form of a design change.
Instead, a period of 18 months elapsed between the event and the publication of the TWL to initiate the design change.
While the inspector found evidence that plant management was. involved in the resolution of the issue, it is clear that management analysis and action could have been expedited by a more definitive analysis process.
The inspector also noted in his review of PMI-7030 that the gA Supervisor was to periodically issue status lists of CR's to inform the Plant Manager of repetitive occurrences and provide comparisons to previous years.
Responsibility for the report subsequently shifted to gC and Shift Technical Advisor (STA) groups, but no group ever actually produced the required periodic reports.
The inspector's review determined that there is a plant wide project to develop a
computer based trending and analysis program to provide trending of CR's, LERs, and approximately 70 other documents and processes.
The details of the program, which include periodic reporting, specific sorting for particular managers, narrative assessment of deviating trends, and recommended corrective actions are to be presented to the NRC Region III management in the next several weeks.
(9)
RFC DC-01-2768,
"Re lace Natural Rubber Seals In The Unit 1 Containment ir Lock DC-01-2768 was issued to replace natural rubber seals in the Unit 1 containment air lock with new seals of ethylene propylene polydienemers (EPDM).
Seal replacement was necessary due to end of service life on the airlock equipment hatch, airlock flange and airlock door.
The same modification had been previously completed on Unit 2.
The material change was required by the unavailability of the natural rubber seals.
Unit 2 experience indicated that the new seal material worked well on hatches which are regularly cycled; however, the licensee observed that once compressed and released, the material does not recompress well.
The Unit 2 equipment hatch and flange seals had been leak tested successfully while the airlock door, which is operated more frequently, had experienced several tests where zero leakage was not achieved.
Consequently, completion of the Unit 1 airlock door was deferred pending analysis of the material by the vendor and licensee.
The inspector made the following observations:
(a)
During review of the airlock doorseal issue, the licensee had determined in April 1984, that the natural rubber door seals installed in Unit I during plant construction did not meet design parameters for containment vessel personnel airlocks as required by ASME Boiler and Pressure Vessel Code Section III.
The ASME Code requirements are addressed in AEPSC Civil Engineering Specification No. DCC-CE-125-(CN,
"Design Specification for Containment Vessel Personnel Locks Unit I and II," dated August 23, 1968.
The design requirements for the natural rubber seal material did not appear to envelope the temperatures to which the material would be exposed during a Main Steam Line Break (MSLB) accident inside containment.
Thus, the licensee had operated the unit for nearly two years with the knowledge that containment integrity could be in jeopardy during a
MSLB.
Following this inspection, licensee personnel provided new data to Region III that satisfactorily addressed the issue of environmental suitability of the rubber seal material.
Based on a recently revised analysis performed after this inspection, information presented by the licensee indicates that the rubber material will actually withstand the conditions imposed by the MSLB.
The inspector is concerned regarding the adequacy of management systems that failed to recognize and address this condition which had the potential to degrade containment integrity.
Specifically, A period of approximately seven years preceded initial plant startup in 1975 during which the licensee would reasonably be expected to recognize the inconsistency between the seal design specification and the actual environmental conditions during an MSLB.
A period of nearly two years elapsed since the licensee first determined that the rubber material might not be environmentally suitable, during which time no action was taken to either establish suitability or replace the material.
Knowing that the rubber seal material installed in Unit I was potentially environmentally unqualified (until the recent reanalysis),
and beyond the manufacturer's stated service life, the licensee has been slow in completing work in Unit 1.
The inspec-tor's review of leak test results for the Unit 2 airlock following EPDM seal installation showed that the difficulties encountered with Unit 2 sealing, the reason for delaying Unit I seal replacement, were corrected within six months of being placed in service.
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The licensee has established a target date of late summer, 1986, for the Unit 1 door seal replacement.
Pending completion of the replacement and action to correct management system deficiencies which permitted these conditions to occur and persist, this matter is unresolved (315/86005-06).
(b)
The inspector reviewed the job orders and procedures used to install the new seals in the Unit 1 airlock flange, equipment hatch flange, and airlock door windows.
The licensee referenced Procedure No.
12MPH5021.001.032,
"Containment Airlock and Equipment Hatch Removal and Replacement,"
on the job order.
It did not address the door window seals.
As a result the licensee's contractor refused to perform the window seal work without a procedure and the work was deferred to a later outage.
Because interfacing work was in progress under different job order numbers, the seal replacement procedure's signoffs and significant steps frequently made reference to other job orders rather than documenting the step with a signature.
In addition, the job order under which work was actually performed did not reference the RFC as required by PMI-2290, creating the possibility that those performing the work were unaware of the scope of the design change and testing requirements.
The inspector found no evidence that the work was deficient in any respect, but expressed the concern that the number of interfaces between documents created a potential for some design change requirements to be omitted during the installation process.
(10)
RFC CC-)2-2905, "I
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C ~F1 The inspector found all activities associated with this modification to be acceptable.
(ll)
FC C-!2-1915,
"F 0 I R~I~Pi 0 5 on Blowdown Pi in The inspector found all activities associated with this modification to be acceptable.
(12) Set Point Chan e Control The inspectors reviewed set point changes to determine if they had been implemented in accordance with design change require-ments.
The set point changes were 02-PM-536,
"Change Rod Bank D Hi-alarm"; 01-PM-528,
"Change Pressurizer PORV Lift and Reseat Setpoints";
and RFC-12-2577,
"Reset DG Pre-position Agastat Timers."
The inspector determined that the three changes were implemented properly and that the safety evaluations were well documented and complete.
The inspector had one concern involving the apparent lack of a formal methodology for including such
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factors as set-point drift, instrument inaccuracies, and calculational uncertainties in set-point margins as recommended by Regulatory Guide (RG)
1. 105, "Instrument Setpoints."
The licensee is not committed to this RG.
This is considered an open item requiring further NRC review (315/86005-07; 316/86005-04).
The inspectors also reviewed LER 85-069 (Unit one) involving an erroneous setpoint to determine if appropriate corrective action had been taken.
During that review, the inspector identified one concern.
Specifically, the licensee determined during startup testing of Unit I for Cycle 9 that the trip setpoint for Intermediate Range Channel N-35 was above the Technical Specifi-cation limit.
The LER stated that the cause was lower core neutron leakage than anticipated.
The trip setpoints are established by multiplying previous cycle detector current by the ratio of upcoming cycle previous cycle assembly powers in the area facing the Intermediate Range Nonitor (IRH) detectors.
By the time of this inspection, the licensee had concluded that changes in core leakage could not account for the amount of error in the trip set point and decided to use a safety factor based on previous startup data to account for instrument drift in future cycles.
Properly, the drift over the entire previous cycle should be used in making this determination.
The nuclear engineering group had arranged for this data to be trended for the current and future cycles by adding it to the computer trend block.
However, future trend data would be of little help in determining the extent to which instrument drift contributed to the root cause of the erroneous Cycle 9 setpoint.
The set point drift data for previous cycles is available from log sheets completed each shift by the operations group; however, the nuclear engineering group was unaware of this.
At the end of this inspection, the licensee had not determined the root cause of the erroneous setpoint.
While they are taking actions to improve the methods for determining setpoint margins in the trip setpoint, it is not certain that these actions will address the root cause for this specific case.
This is considered an unresolved item pending determination of the root cause and completion of appropriate corrective actions (315/86005-08).
General Comments and Recommendations (2)
Safety evaluations reviewed were well documented and clearly identify those issues addressed.
The licensee should take immediate action to ensure that the root causes of all conditions adverse to quality are promptly determined and that the corrective actions are implemented with deliberate speed.
Also, the licensee must ensure that interim actions receive the same safety evaluations as the final corrective action.
(see Paragraphs Z.b(8), 2.b.(9)(a),
and 2.c.(7)).
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(3)
The licensee must exercise more care to ensure that all documentation is complete, especially with respect to signatures and dates.
(4) It is recommended that the licensee investigate the issue described in Paragraph 2.b.( 12), involving the lack of knowledge by one group of data available from another group.
This may be indicative of a more general problem in intergroup communications.
(5)
The role of the Design Change Coordinators (DCC) appears to have dramatically improved on previous design control procedures which had authority and responsibility fragmented throughout many areas of the organization.
However, the current work load of the DCC, with each of the six coordinators having an average of thirty active design changes in progress at any one time, appears excessive.
While the inspector found no evidence that the quality of the modifications have been sacrificed, the quantity of safety-related and safety interface design changes in backlog implies that there is a significant body of work with the potential to impact on plant safety.
The inspector understands that the licensee is considering expansion of the Planning Department Staff and Facilities.
This should be aggressively pursued.
(6)
Several DCC's interviewed by the inspector pointed out that significant improvement could be made in the processing of design changes if the site design group were expanded.
In addition, those DCC's interviewed indicated that the geographic distance between the Lead Engineer at AEPSC and the site impeded the progress of modifications by complicating communication and limiting the Lead Engineer's involvement.
(7)
The licensee's method for assigning modification priorities should be revised to ensure that the most safety significant modifications are accomplished first (see Paragraph 2.b.(8)(d)).
3.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations.
Unresolved items disclosed during this inspection are presented in Paragraphs 2.b. (1), 2.b. (9) (a),
and 2.b. (12).
4.
~0 Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both.
Open items disclosed during this inspection are presented in Paragraphs 2.b.(8)(e)
and 2.b.(12).
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5.
Exi.t Intervi ew The inspectors met with licensee representatives (denoted in Paragraph 1)
at the end of the onsite inspection on February 14, 1986, and summarized the purpose, scope, and findings of the inspection.
During the interview, the licensee indicated that the inspectors had no access to proprietary information.
Further discussions between licensee personnel and Region III management and inspection staff occurred through March 3, 1986.
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