IR 05000220/1992003
| ML17056B735 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/09/1992 |
| From: | Anderson C, Lohmeier A, Roy Mathew NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17056B734 | List: |
| References | |
| 50-220-92-03, 50-220-92-3, 50-410-92-03, 50-410-92-3, NUDOCS 9203270066 | |
| Download: ML17056B735 (46) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
2 j N
. ~222 2-
~lf 2-Docket Nos.
~2~2
+-410 License Nos. QPR~Q N~PF-Licensee:
Ni MhwkP wr tion Facility Name:
Nine Mile P int Nu I r tion ni
2
'Inspection At:
cri aan c
Nw Y rk Inspection Conducted:
anu 27-F
1
Inspectors:
R. K. Mathew, Reactor Engineer, Electrical Section, EB, DRS date A. Lohmeier, Sr. Reactor Engineer, Materials Secti n, EB, DRS 3/alt date Approved by:
C. J.
nderson, Chief, Electrical Section, Engineering Branch, DRS y gv date A~N: D lg,d lg
g d
Nig ','l '
'g
modifications, onsite and corporate engineering interface/communication, engineering organization and its capability, quality assurance, training, management support and staffing.
and licensee actions to'esolve identified weaknesses in the previous engineering and technical support portion of the SALP.
~Re pelt:
Modifications and design changes reviewed were of good quality and technically accurate.
Increased engineering management involvement, project team oversight including oversight of consultants to resolve issues were observed to be good during this period.
A failure of a surveillance test of diesel generator at Unit 2 was identified as the result of a lack of timely engineering review of an NRC Information Notice.
The licensee event reports and 9203270066 920316 PDR ADOCK 05000220
deficiency evaluation reports reviewed had the proper reportability, operability, root cause analysis and corrective actions.
The quality assurance audit and program were found to be adequate.
The progress in the design basis reconstitution program at Unit 1 and configuration management control program at Unit 2 indicates management's continued support-and effort to establish a better configuration control and design basis documentation for the plant. A technical training and qualification program for engineers has been established and is being implemented during the SALP cycle.
Good planning and engineering support exist at Nine Mile for emergency conditions.
A backlog of several engineering action items remained high during this SALP period including temporary modifications which calls for continued management attention.
Niagara Mohawk is implementing a right sizing process depending on performance and other factors.
The backlog of engineering activities and timeliness of delivering engineering products have to be taken into account during the implementation of this process.
Good communication/interface exists at Nine Mile between corporate and plant staff.
Increased management involvement, controls, and initiatives to assure quality of engineering products were observed.
Capability to resolve technical issues and deliver quality engineering support continued to be goo of the In
'
The purpose of the inspection was to determine the effectiveness of the licensee's engineering organization to provide engineering and technical support,and guidance to the Nine Mile Point Units 1 and 2 to assure safety.
The scope included:
the review of design, design changes and modifications in accordance with plant procedures, the requirements and commitments specified in the facilities Technical Specifications (TS), NRC rules and regulations, safety analysis report and quality assurance (QA)
program; the licensee's organization structure and adequacy. of staffing in the engineering area; communication/interface between corporate and site organizations; management support; workload and backlog; plant outage activities; assessment of weaknesses identified in the previous engineering and technical, support portion of the systematic assessment of licensee performance (SALP) report; QA audits; capability of engineering staff to resolve technical issues; and technical training.
1.1 A mini i
nrolsf rDei n h n e dM ifi ti n The inspectors reviewed administrative procedures and engineering procedures to determine whether the engineering activities are specified and controlled by approved procedures.
The procedures reviewed included plant modifications, design change initiation, design input, design verification, safety evaluations, design document changes, configuration management,
-station operations review committee (SORC) reviews, and the modification/simple design change program.
The review indicated that the licensee's procedures provided adequate administrative guidelines and controls to ensure that design, design changes
'nd modifications performed do not involve an unreviewed safety question.
.Appropriate requirements and guidelines are provided for the 10CFR 50.59 screening review and safety evaluations, design input, design calculations and design verifications.
The inspectors noted that the licensee was implementing a procedure update program to be consistent with the Nuclear Division Directives.
The nuclear engineering procedures update was completed January 1992.
The implementation of the new procedures is expected to be completed within 30 days to permit the engineering staff to obtain training on these new procedures, In many areas, the new procedures enhanced work processes and made minor changes to responsibilities/assignments.
Twenty-,
three new 'procedures or major revisions were issued during this procedures update program.
Some of the procedures include NEP-CON-303, "Q-list Changes/Revs and Component Q List Data Base Changes," NEP-DES-103,
"NMP1 System Design Basis Documents," NEP-DES-37, "Vendor Technical Manual - Review and Processing",
NEP-PRO-300, "Nuclear Engineering Procedures Program", and NEP-PTM-302, "Plant Modifications".
I
The inspectors concluded that the licensee has established adequate measures to ensure that plant design changes, modifications and engineering activities are prescribed and controlled by approp'riate procedures.
Y~h"i T
The engineering staff technical training program had been previously identified as a weakness by the NRC. In response, the licensee implemented an initial critical training program to address identified weaknesses until broad-based training is established.
The critical training program was completed in 1990.
A review was performed to determine the progress made by the licensee during this inspection period.
The licensee established a broad-based technical training program for the corporate engineering staff in January 1992.
The program covers approximately 74 technical areas.
The areas include nuclear reactor systems, safety evaluation, process controls, root cause analysis, codes, standards, QA fundamentals and other specialized training.
The implementation of the training started on January 6, 1992.
The existing training is expected to be completed within 3 years.
A continuing training program for corporate engineers is being determined by the engineering training advisory committee.
However, the. plant staff was already receiving the required training.
An engineering training matrix was developed, based on the required, needed and wanted training input from supervisors.
The responsibility for scheduling and completing required training rests with the functional group supervisor.
Furthermore, the licensee has established an engineering qualification program based on. their job assignments.
The plan is generated by engineering supervision and validated by job incumbents detailing tasks, courses and procedures of which the engineer must demonstrate knowledge and ability.
The existing qualification program will enable the managers to assign tasks to the proper engineering staff.
Based on discussions with corporate and site personnel, the inspectors found them to be technically competent and very familiar with the areas of their responsibility.
The training and qualification program developed to address training needs of the engineering/technical staff seems to be comprehensive and designed to enhance the knowledge and skills of nuclear engineering personne ani i n The inspectors reviewed the licensee's organization and staffing levels to ascertain the level to which they contribute to the technical support of the plant.
The licensee's engineering/technical support to assure safe plant operation is provided by the plant site engineering group and system engineers at the plant site and corporate engineers at Salina Meadows.
The Vice President of Nuclear Engineering now has 5 groups under his direction.
These
.
include:
Unit 1 Engineering, Unit 2 Engineering, Technology Services, Performance Services and Independent Safety Engineering Group gSEG).
The engineering and technical support for each Unit is provided by the respective Units 1 and 2 Engineering managers.
The site engineering group, reports to the respective Unit Engineering managers.
Design authority is given to the corporate and site engineering group.
They are responsible for the development of modifications, simple design changes and resolution of technical issues.
The system engineering group, headed by a manager, provides technical support for each Unit and is responsible for operational problems and reports to the Plant Manager.
The total nuclear engineering staff budgeted for Units 1 and 2 during the 1990-1991 SALP period was 123 and 138, respectively.
The Units 1 and 2 engineering staffing at the end of 1991 was 112 and 114, respectively.
The licensee is implementing a right-sizing plan (based on bench marking, performance, and business plan goals) to reduce the engineering staff.
The proposed staffing at the end of 1995 willbe 92 for Unit 1 and 102 for Unit 2.
The unitized engineering organization directed by respective managers established at Niagara Mohawk provides adequate support for the station.
The site engineering group with design authority stationed at each plant provides adequate representation of the corporate engineering group to expedite the engineering and technical support for the stations.
En ineerin B ckl n
Pri ritization The review of the engineering performance monitoring program for the month of December 1991, indicated that there was a backlog for both Units in several items such as simple design changes, walk-ins that needed engineering dispositions, temporary modifications, change papers and work requests.
Unit 2 engineering department has 295 vendor manuals that needed to be updated.
The licensee's schedule to complete this backlog is September 199 A review of the Unit 2 engineering drawings update project indicated that a large backlog exists (approximately 10,800 drawings).
The licensee stated that the drawings update project started in 1991 and the expected completion for the drawings update is 1995.
The inspectors reviewed critical drawings that are needed for the operating staff in the control room.
No major concerns were identified with the exception that some of the drawings had invalid modification references.
This was brought to the licensee's attention, and the licensee reviewed all the drawings in the control room to assure that the modifications referenced in these drawings were, in fact, to be incorporated in these drawings.
The licensee's review of all the drawings indicated that 67 drawings were found marked with references to the "Mod Pending/As-built."
The licensee stated that the modifications referenced in these drawings were cancelled/deleted and no redlining of drawings were needed as a result of this discrepancy.
The licensee immediately deleted the modification reference numbers from these drawings, and also replaced them with new copies.
A review of the number of open modifications for both Units indicated that the modification backlog has been reduced substantially, from 500 to 96.
This backlog reduction meets the engineering branch's goal of 100.
The inspectors noted that the licensee has established a business plan for the nuclear division and at the branch level.
Goals are set and performance is monitored on a monthly basis to assure that the goals are accomplished.
The licensee management is closely monitoring and taking initiatives to reduce the existing engineering backlog.
To enhance plant safety and provide better direct plant support, Niagara Mohawk has established the Integrated Priority System /PS) in accordance with procedures NDP-34.
The IPS applies to planned work in the nuclear division and support organization. 't is assigned with six levels of priority.
Allsafety significant projects are priority 1, and other work which affects safety systems are priority 2.
The effectiveness of the system is evidenced by the fact that all priority projects are on schedule and are reviewed on a periodic basis.
For priorities 3 through 6, work is assigned a merit score indicative of the value of its outcome.
The merit score within a priority level is a secondary measure used for planning and completing work in the nuclear division.
The system for assigning priorities to plant modifications appeared to have the proper safety perspectiv A review of Unit'2 second refueling outage planning activity and engineering participation in the outage activities determined that corporate engineers and site engineers effectively support these activities.
The inspectors reviewed the projects proposed for the forthcoming outages.
The work items were identified as pre-outage or outage activities and the review indicated that modifications required for the outage were design completed and issued to the plant for implementation.
The engineering support personnel required to cover the plant outage activities were identified and were already participating in the pre-outage work activities.
The site engineering group was actively involved in maintenance work activities to eliminate engineering holds on corrective actions for work requests to support the outage.
mm ni
'
In rf An effective interface between the station and engineering personnel exists at Nine Mile Point 1 and 2.
This was evidenced by the staffing of site and system engineering groups for each plant to support the engineering/technical needs of the plant.
Effective interface between operations, maintenance and engineering was established by the participation of representatives from site engineering, corporate and system engineers at the daily meetings.
In order to improve communications between the engineering staff at Salina Meadows and other organizations at the plant site, the licensee had established a daily plant status meeting, a plant manager's meeting, and an afternoon meeting to discuss action items and subsequent telephone calls needed to resolve issues.
The active participation of management representatives from different organizations at these meetings complimented the effective communication at Nine Mile.
Furthermore, the nuclear generation/engineering interface meetings are held quarterly, and plant manager and engineering manager meetings are held biweekly to resolve and prioritize engineering work activities.
R vi w fPr vi ALP A review was performed to determine the licensee's actions to address the areas identified by the NRC as needing improvement for the engineering and technical support section of the last SALP. The areas needing improvement were 1) timeliness and quality of engineering work, 2) increased management oversight of engineering work, 3) timely implementation of technical training program, and 4) thorough review of technical issues to assure qualit The licensee initiative to improve timeliness and quality of engineering was evidenced by the following actions:
a.
Performance measurement system was established in 1991 to monitor engineering performance, meeting schedules, monitoring backlog and amount of rework.
b.
Accountability meetings to emphasize lessons learned.
C.
Increased engineering involvement in planning and pre-planning of outage activities.
,. d.
Implemented simple design change process to expedite engineering support for the'station and the design authority delegated to site engineering group, e.
Independent SALP assessment process to monitor quality of engineering'ork.
Interface meetings such as focus meeting, engineering/plant manager's meetings and daily morning meetings to discuss the engineering activities.
g Top ten list to identify and focus important issues for timely resolution.
2.
Increased management oversight of engineering work was established by the following:
a.
Monthly review of top 10 list to focus issues at each Unit.
b.
Safety review and audit board (SRAB) engineering subcommittee was formed to monitor the engineering activities and submit report to the SRAB.
C.
Reorganization of engineering organization during the last SALP period for unitizing the engineering activities.
Clear lines of responsibilities are defined to effectively utilize personnel to support the Units.
d.
Implementing business plan, goals, performance monitoring system and performance measuring feedback system to identify and focus on issue e.
Deficiency/Event Reporting (DER) process to identify concerns and take appropriate corrective actions.
3.
A technical training and qualification program was established and being implemented during this SALP cycle.
This is discussed in Section 1.2.
4.
In order to improve the quality of review of technical issues, the following items were established:
a, Plant critical issues that are identified are reviewed through the top ten list. Task managers are assigned for implementation and followup of the issues.
b.
Root cause analysis performed by quality assurance operating event assessment group and thorough root cause investigation training for those involved in the effort.
C.
Review of plant modifications by technical review committee, design verification by independent groups and SORC review.
Accountability meetings/lessons learned meetings were held to emphasize the need for SORC review, of modifications to eliminate the past problems in SORC review.
d.
Modifications/complex issues performed through project/task management team and team concept.
e, Senior engineering review team to review design basis.
reconstitution effort at Unit 1.
1.7 T
hnical u
rt to Res Iv Plant I sues The inspectors reviewed the licensee's engineering organization's involvement and capabilities to resolve issues at the plant. A review of sample Licensee Event Reports and Deficiency Event Reports indicated that root cause investigations are thorough and technically accurate.
Reportability and operability reviews were properly performed and coordinated with appropriate groups.
Engineering was actively participating, and effectively contributing to the analyses, evaluations, and proper resolution of events and problems.
Capability to resolve technical issues and to deliver quality engineering products continued to be good.
For example, 1) Research and development effort to investigate the performance of recirculation pump seal failure and replacing it with new seals at Unit 1; 2) Good technical evaluation and corrective action for recirculation pump sample line failure at Unit 2; 3) Sound
technical evaluation to support operation of Unit 1 above 80% power with one high pressure feedwater heater out of service; and 4) Good technical evaluation and resolution of the crack problem for the high pressure core spray safe'end at Unit 2.
Increased management involvement, controls and initiatives to assure quality of engineering products were observed by the following examples:
Implementation of business plan and performance monitoring to assure quality focus on issues and management oversight of engineering work activities.
2.
Top 10 technical issues to prioritize and focus on issues.
3.
4.
5.
SRAB subcommittee and ISEG activities to oversee engineering work activities and performance.
DER process, root cause investigation process, internal SALP type assessment programs to assure quality.
Project team effort and assigning task managers to resolve issues through team approach.
6.
Reduction of litannunciators at Unit 2 to enhance operator performance and replacement of motor generator sets with static units to improve equipment performance.
7.
Continued management support to complete design basis reconstitution effort at Unit 1 and configuration management programs at Unit 2.
Review fT hni 1 I e
ter D in Line -
nit 2 The licensee, in a program to mitigate pipe weld failures at the junction of the heater drain line to the condenser wall nozzles, discussed with the inspector, a program to determine the root cause of failure of the pipe welds and to provide for correct positioning of piping restraints to mitigate effects causing pipe weld failure.
The measurement program provides for the installation of instrumentation such as thermocouples and strain gages which willindicate the nature of the forces giving rise to weld failure. The program is an observational one and includes piping analyses.
No permanent modifications have been installe.
Nzzl f En K -2-ni 2 A circumferential inner surface crack'was discovered in the core spray nozzle to safe end weld which was 41% depth/wall thickness ratio and 11.3% length/circumferential ratio.
The nozzle diameter is 10-inches and the weld wall thickness is 8.5 inches.
This type of cracking has occurred in nozzle safe end welds of many operating reactors.
There have been several approaches to releiring these welds considered by the licensee including weld repair and replacement, induction heating stress improvement gHSI), weld overlay of the crack region, and mechanical stress improvement (MSIP).
The licensee selected the MSIP for this repair.
The process provides for a compressive residual stress field in the crack tip region which precludes crack extension.
Because of the large depth of crack, the licensee cooperated with NRR in evaluation of the crack, its anticipated progression calculated using fracture mechanics analysis, inspection at mid-cycle using a rotating ultrasonic test set up, and establishment of a 50% limiton crack depth.
The crack willbe inspected at the next shutdown to observe the crack progression and determine whether an alternate approach must be used.
This modification demonstrates the licensee awareness of state-of-the-art technology in materials failure prevention.
The licensee has followed the recommendations of those knowledgeable in stress improvement techniques and provided for an appropriate solution to the problem.
1.8 T m m
ifi
'
-M The inspectors reviewed the licensee's temporary modification program to assure that temporary installations are performed and controlled by approved procedures.
The temporary modification is performed in accordance with procedure AP 6.1.. The temporary modifications are maintained by the system engineering group and engineering disposition/evaluation is performed by the site engineering group.
A sample review of temporary modifications for each Unit indicated that the licensee is performing adequate 10CFR 50.59 evaluation and technical review. A review of the temporary modification log kept in the control rooms for Units 1 and 2 identified that the number of open
temporary modifications have increased from-78 to 92 for Unit 1 and 90 to 110 for Unit 2 since last year.
Even though the open T-Mods were periodically reviewed, several T-Mods remained. open for 3-4 years.
The'nspectors concluded that increased licensee management attention is needed to reduce the backlog as well as to eliminate long standing T-Mods at both plants.
mr n
R n
i
'
The inspectors reviewed the engineering staff's plan for supporting the plant during an emergency condition.
The review indicated that a 1992 engineering on-call list and engineering emergency duties phone list were established by the licensee's Emergency Planning Coordinator.
An on call Coordinator was assigned for each week and a primary contact and support staff for each
.
department were established.
The coordinators and primary discipline contacts were provided with pagers required to be on at all times.
Also, they were required to comply with the fitness for duty program.
Based on the site area emergency event in 1991, management had issued a memorandum to all the responsible emergency duty assignees providing the direction and instructions to follow in order to assure appropriate engineering response.
The inspectors concluded that the licensee engineering has a good plan to respond for emergency duties.
DeinB i
R nstituti n BR nfi uration Mana emen Pr m
This area was reviewed to determine the licensees management's initiative in maintaining good design basis and configuration control for the safe operation of the plant.
This program at Unit 1 was initiated in 1990 as a prototype project.
Last year (1991) was the first year of full production with approximately 15 personnel to support the effort and is expected to be completed in 5-6 years.
The program is integrated under four groups consisting of 1) design basis reconstitution/configuration management upgrade, 2) evaluation, monitoring, and maintenance of plant conditions, 3) plant reevaluations and upgrades, and 4) engineering resources and training methodologies.
Thirty-seven systems are identified in this program.
To date, six systems are completed, and two are partially complete.
The DBR effort consists of conducting extensive as-built verification, developing design criteria documents and verifying assumptions of calculations and completeness of data (verification and validation process).
A review of the completed system. design basis document for service water showed that it clearly describes the system requirements, system overview, system design description, system operating
.
limits, test and surveillance requirements, maintenance considerations and regulatory conditions.
The licensee has a design basis discrepancy resolution group consisting of a team of highly qualified engineering and operations
.
personnel to resolve the identified discrepancies and open issues.
A review of the Unit 2 configuration management (CM) program indicated that seven members were working on this project.
The project was started in 1986 and is expected to be completed in 1995.
The configuration management requirements are delineated in the nuclear division directive, NDD-CON. The licensee developed an automated computer database configuration management system that integrates the master equipment list, control documents system, modification tracking system, material management system, cable raceway system and plant change request system.
This computer capability enhances the information available to the engineering staff for performing modification and design changes as well as updating various document information and improve maintenance planning.
Design basis documents exist for Unit 2, but, presently the licensee is upgrading and implementing system design basis documents for the plant.
Twelve system design basis documents are scheduled for completion during 1992.
To date, one system design basis document has been completed.
The inspectors concluded that Niagara Mohawk has taken a significant initiative to recover/reconstitute the design basis of NMP-1 and improve the configuration process at NMP-2. Significant resources have been budgeted and plans are in place.
2.0 D
i n han es and M ificati n Pr ram Im lementation The inspectors reviewed selected design changes and modifications for Units 1 and 2 to ascertain that they are performed in conformance with the requirements of the Technical Specifications (TS), 10CFR, the Safety Analysis Report, the licensee's Quality Assurance Program and in accordance with licensee's procedures.
Also, technical quality of modifications, thoroughness of design analysis, design input, technical review and safety evaluations, management involvement and review and resolution of problems from a safety standpoint were evaluated.
The following modifications and simple design changes were reviewed to determine the abov ~
.
2.1 n
rR
'Pr ram-ni 2-Ph III M N
N2Y MX Operating BWR plants have experienced significant snubber related problems which resulted in increased costs to the licensees in extended outage, testing, inspections and modifications.
Through utilization of relaxed pipe rupture criteria of NRC in Generic Letter 87-11 and ASME Code Case N411, snubber reduction programs have shown economic and safety benefits for several licensees.
Niagara Mohawk Power Corporation (NMPC) Unit 2 has evaluated the potential savings through a snubber reduction program which results in an estimated benefit/cost ratio of 2.39.
NMPC Unit 2 has completed 2 phases of a 3 phase program to reduce the number of snubbers.
Phase I of the program consisted of project initiation, removal of category II snubbers from the Technical Specifications (per procedure NEL-054) in that the failure would not affect a safety related system, and an engineering analysis of 250 snubbers which resulted in replacing 129 snubbers with rigid struts.
Phase IIof the program provided stress analyses on representative subsystems to establish a baseline for program costs.
During this phase, 218 snubbers were removed and replaced with rigid struts.
Phase IIIof the program willconsist of analysis of the balance of piping subsystems.
During the second outage, 312 snubbers willbe evaluated and during the third and fourth refueling outage, 900 snubbers willbe reviewed for removal.
The inspector reviewed the project reports, engineering design standards, modification work request, and safety analysis reviews.
These were found to be consistent with regulatory requirements and standards.
The snubber reduction program of NMPC is well managed, has sound economic and technical bases, and a well controlled implementation plan.
Benefits of this program are reduced maintenance and inspection costs, increased plant safety through a lower probability of unanalyzed conditions, and reduced number of outage extensions through a lower population of snubbers.
It is estimated that the snubber population willbe reduced to approximately 400 from 1236 technical specification snubbers.
2.2 R trBil in l
ed lin B L H
Exch er
-
ni
During the extended outage in 1989, while in-service testing of pumps, flow from 2 pumps was put through a single RBCLC heat exchanger.
An engineering investigation revealed tube failure had occurred in the heat exchanger.
The heat exchanger design flow was 4250 gpm (shell side) and the two pumps could provide for up to 9000 gpm normal flow. Hydraulic tests
indicated that flow induced vibration would occur at flows above 3500 gpm in a single heat exchanger.
Analytic review indicated that gap velocities between adjacent tubes were sufficiently high as to cause. flow induced vibration at rated flow.
The licensee repaired the failed heat exchanger by removing tubes in the region of the inlet and outlet shell side nozzles, such that flow redistribution would occur in these regions, to reduce the gap velocities.
Furthermore, the flow through each heat exchange would be limited to 3000 gpm, to generally reduce tube gap velocities below that which caused flow induced vibration.
The heat exchangers are under consideration for future modification; either replacement of existing heat exchangers with larger flow capacity, or unloading the present system by providing other cooling means for a portion of the load through the addition of another heat exchanger.
The inspector reviewed analytic documents relating to thermal/hydraulic analysis and benchmarking, and the certification of compliance to NRC standards (10CFR 50.59).. The consultants selected by the licensee for the analytic and experimental investigations of stress, thermal hydraulic, and vibrational behavior were appropriate for the specialized nature of the investigations.
Overall, management of the repair and modification program was effective.
It is noted that, during the course of procurement of the heat exchangers, design reviews by the licensee did not reveal an inadequacy in shell side flow limitation to preclude tube flow induced vibration. While the issue of flow induced vibration is a complex one, the utilization of appropriate expertise would have uncovered a potential problem.
Hence, design reviews should be staffed by those reasonably expert in the equipment to be purchased.
A walkdown of the heat exchangers at the site revealed leakage to be emanating from flanges at the butterfly valve. Allleakage had been marked with contamination alert tags and collection containers placed under the leaks.
One leak was noted to drip directly on a heat exchanger shell without collection.
This was called to the attention of the licensee and appropriate action was taken to tighten the flange bolting such that the leaks were eliminated.
The two operating heat exchangers revealed no sounds of internal noise from vibrating tubes.
In general, the engineering and program management of this repair technical investigation program was effectively controlled and a solution to a difficult
'problem was obtained which provided an understanding of the equipment technology which would provide for more reliable operation in the futur.3 R
rB il in P l r ni 2 ER2-2-
-
I Recurring problems with the reactor building polar crane over past years included failure of the hold down studs on the crane rail, a crack on the truck capsule cap of the B grids outer idler wheels,'a crack in the rail bottom flange, and bearing failure.
The polar crane is used in the refueling process and it lifts internal elements of the reactor.
Examination of the crane rails by the licensee indicated that at several locations of rail splicing, the gap exceeded the allowable maximum specified gap.
Consequently, the travel of the wheels over the larger gaps causes an irregular traveling motion (bumping) which could be troublesome under heavier loading.
The licensee has discussed the problem with the crane manufacturer and equipment specialists and, as a result of their discussions, is acting to effect a repair program.
The excessive clearance between rail sections willbe reduced by welding spacer pieces between the rail sections.
The cracks in the crane elements and the rail willbe repaired and all bearings are to be replaced.
The bearings are permanently lubricated, but the licensee is considering the possibility of providing'for lubrication fittings. Of major significance in the licensee's action plan is=to limitthe crane load to-25 tons.
This program is being handled as a maintenance repair.
The inspector reviewed licensee documents relating to the repairs, including the engineering design change, the operation and maintenance manual for the crane, crane support detail drawings, polar crane splice calculations, and root cause evaluations.
The licensee has acted appropriately in providing for the specialized expertise and involvement of NMPC engineering in this issue.
2.4 R
c r W ter leanu tern Mod N M2Y88MX058 Feedw er tratificati n -
ni 2 In response to the potential for thermal stratification in feedwater systems identified by USNRC Information Notice 84-87, NMPC monitored the NMP-2 feedwater system for thermal stratification during startup and initial operating periods of the plant.
Monitoring revealed stratification temperature distributions which could significantly reduce the useful life of piping in the break exclusion area thermal tee region.
Three modes of operation were-particularly significant in resulting stress developed which could affect pipe cyclic life: startup, reactor scram, and shutdow As a result of the monitoring observations, operating procedures were changed such as the reject WCS flow to the condenser where reactor water exceeds 200'F and reactor power is less than 20%.
This operation procedure, however, gives rise to chemistry problems in the condenser.
An analysis by a consultant indicated that directing all flow to one active feedwater line would reduce the temperature differentials and shifting flow to either of two feedwater lines would double the predicted pipe life. It was decided, therefore, that two 8-inch motor operated globe valves would be added to direct the flow to either or both feedwater system lines.
Subsequent to the addition of the two valves, the licensee willmonitor the
'resulting thermal gradients in the system such that a pre-determined delta T in
~
the system willnot be exceeded.
The inspector reviewed the licensee safety evaluation report description of the modification to be made together with the certification of compliance to NRC standards (10CFR 50.59) and found the documentation consistent with regulatory requirements.
The actions of the licensee displayed appropriate technological bases for the actions taken, consultants utilized and test plans to supplement the installation of the modifications to the WCS.
2.5 FeedwaterFl w
n 1V lv -
ni 1M N
1-
-
Feedwater Flow Control Valves (FCVs) 13A and 13B have experienced recurring problems with stem/plug failures, Furthermore, the FCVs lead stellite.hardfaced surfaces provided for a source of Cobalt 60 at NMP1.
Review of the problem by a consulting engineering firm resulted in recommendations to replace the two globe valves, 13A and 13B, with a single Cobalt-free valve of different design.
The valve stem was determined to have failed by flow induced vibration in fatigue.
The replacement valve design was selected from a group of three design options.
The preferred design option is a state-of-the-art control valve capable of maintaining low local velocities within the valve to minimize wear, reduce leakage, and give tighter sealing.
Reduced erosion willminimize deposits to the reactor.
The inspector reviewed the project report describing the technical bases for selecting the replacement system, drawings showing the difference between the original valve design and the replacement valve.
These valves are not safety related.
The licensee through his consultant demonstrated sound engineering approach to the resolution of the recurring valve stem failure problem and provided the additional benefit of reducing the source of more Cobalt 60 in the RC.6
I mate R In
'
-
ni 1 (Mod¹ 80-72)
~.v This modification was initiated by the licensee to meet the requirements of 10CFR 50.62 for the installation of a diversified alternate injection system.
Allother modifications required to meet the rule were completed prior to this modification. The ARI system is redundant to the backup scram valves and provides a path for reactor shutdown that is diverse and independent of the RPS.
The ARI logic sensors are arranged in a redundant two-out-of-two logic channel with either of two logic channels being able to initiate ARI. The two 3-way solenoid valves are arranged in parallel for air inlet and in series for air exhaust so that both valves must be actuated to cause reactor scram.
The power supply for these valves is powered by proper fuses and the logics are completely isolated from the RPS circuits.
The inspector performed a walkdown to verify the installation of some of the components for this modification. The as-built configuration matches the design configuration and no unacceptable conditions were noted.
The procurement aspect for this modification was reviewed.
The review indicated that it met the QA guidance given in NRC Generic Letter 85-06.
The SORC review, safety evaluation, post modification testing, design verification and design input were thorough and in accordance with procedures.
rner'R m
vel Alarm -
nit 1 (Mod¹88-075)
This modification package was developed to address an observation made by the NRC during the detailed control room design review.
Presently, there are sump level alarms for each sump that initiate an annunciator alarm in the control room.
As part of the validation of emergency operating procedures (EOPs), the review team determined that during certain accident scenarios, operators may have to initiate various protective actions depending upon the water levels in the corner rooms.
This modification is adding level instrumentation that willeliminate the requirements for operating personnel to enter the corner rooms to assess the water level during periods of potentially high radiation or temperature as directed by the EOPs.
This modification is scheduled to be completed in the 1992 refueling outage and willprovide the alarm signals to the plant computer and reflash capability to the existing main control room annunciator window. This modification affects only the non safety related components and circuits.
No interfaces exist with the safety related systems.
The modification, even though it is classified non safety, enhances the operating staff's ability to execute the EOPs effectively from a human factors engineering standpoint and improves the plant safety.
The conceptual engineering packages, design input and preliminary safety evaluations reviewed were thorough and the project team was very knowledgeable about the requirements of the modification.8 R
l nin
'
P w r
-
n'
(Mod¹PN2Y89MX-042)
This modification was initiated to address the overloading of uninterruptible power supplies (UPS), 2VBB-UPS-1C and 1D.
On April 13, 1989, Nine Mile
'nit 2 experienced a plant scram due to a turbine trip caused by a loose wire connection in the main generator potential transformer cubicle.
During this event, UPS-1D tripped due to an overload condition.
This was described in LER 89-014.
This resulted in a loss of half the Gaitronics communication system in the plant.
Subsequently, the licensee initiated a modification to add numerous communications equipment in the plant.
During the development of this modification, the licensee identified the UPS-IC to be overloaded.
A temporary modification to reduce load was initiated until this modification was completed.
In order to eliminate the past maintenance outages on these UPSs, this modification willreplace the existing 2 UPSs with new state-of-the-art models and 30% reduction on the existing loads.
The load reduction is done by lowering the voltage requirements of light fixtures'and changing the power supply to some of the essential lighting fed from UPSs to normal power supply.
These are scheduled to be completed by this refueling outage.
The replacement of the UPSs are scheduled for completion after the 1992 outage and the design is only partially complete.
The safety ev'aluation, project report and design verification were found to be comprehensive.
This modification had adequate SORC review and management review and effective communication was observed during the discussion of this modification with system engineers and corporate engineers.
2.9 Re lac men f r ITE Molded e Br ker
-
nit 2 (Mod¹ N2-88-193)
This modification was written to replace obsolete and unavailable molded case circuit breakers installed in panel boards and motor control centers that have no spares or insufficient spares available.
This willreplace approximately 150 breakers of twenty-four different types.
This project willpreclude future system outages, limiting condition of operation, and plant outages due to unavailable spare parts and breaker failures.
There are thirty-six modification packages developed to complete this modification.
Sixte'en packages are scheduled for completion this outage.
A review of the modification-determined that breakers were qualified and tested appropriately, procurement documentation met the 10CFR Part 21 and Appendix B requirements,,
replacement breakers have better characteristics than those installed before and met all the design requirements.
Safety evaluation, design verification and SORC review were performed in accordance with procedures and were, found to be thorough.
Good interface between corporate and site (systems and maintenance)
engineers was observe The following simple design changes were also reviewed.
SC2-0069-91 - Unit 2 - This change revised the cooling tower screen level alarm setpoint.
2.
SC2-0045-92 - Unit 2 - This change was written to replace Division I and II diesel generator governor control relays due to oxidized contacts.
An occurrence report %0-171 was written and a special report to the NRC was made on December 19, 1990, to report a diesel generator valid failure during surveillance test due to oxidized contacts of low voltage and low current agastat relays used in governor controls.
The licensee immediately replaced the relays with new ones and retested the diesel successfully.
The inspector noted that the relay contact oxidation problem in the feedwater control system was known to the licensee in 1988 (through General Electric Service letter) and appropriate corrective action was taken during 1989-1990.
Also, in 1988, an Information Notice 88-98 was issued by the NRC to address the oxidation problems in low voltage applications.
The licensee failed-to review the applicability of the Information Notice gN) to other systems.
The inspector noted that ifthe licensee had taken timely corrective action, then the diesel failure would have been eliminated.
The licensee stated that presently, any generic information such as an information notice is reviewed through their DER process to receive adequate review.
The licensee is replacing the suspect relays with non-oxidizing type relays during the second refueling outage.
The licensee is performing further review to determine ifany other relays are affected by this.
The design change package review showed that appropriate technical review, design verification and'design review were performed in accordance with procedure.
3.
SDC-SC1-0137-91 - Unit 1 - Setpoint change to eliminate spurious alarms for containment oxygen analyzers.
4.
SDC-0278-91 - Unit 1 - Setpoint changes for battery rooms and EDG rooms thermostats.
The above simple design change review indicated that the design changes were performed in accordance with procedures and technical review, 10CFR 50.59 review and design verifications were found to be adequat In summary,- modifications and design changes reviewed were of good quality and technically accurate.
Engineering management involvement, project team, oversight, including oversight, of consultants to resolve issues weie observed to-be good.
Engineers and project team members were very knowledgeable of their modification and design changes.
10CFR 50.59 screening process, safety evaluation, design input, SORC and technical review, post modification testing and adherence to the procedures were found to be thorough and in accordance with procedures and applicable regulatory requirements.
Participation of.
system engineers and site engineers were noted during this process.
3.0 ualit A
urance Au it / urveillances - Units 1 and 2 NMPC has an active auditing and surveillance program to evaluate engineering performance.
Audits over the past year have included "Nuclear Engineering and Licensing Activities (91006-RG/IN) and "Maintenance, Modification and Testing (92001-RG/IN).
The audits include simple design changes, assignment of responsibilities, engineering training, processing NRC information, preparation for EDSFI and modifications/simple design changes.
Surveillances evaluating engineering performance included drawing and design transmittals, MOV program, design basis reconstitution, engineering support, torque switch settings, NMPC and vendor calculations, and vendor manuals.
Results of the audits and surveillances provided for a self-assessment of engineering performance including identification of strengths and opportunities for improvement.
The inspector reviewed a list of the scope, strengths and opportunities for improvement provided by the Quality Assurance Division and found it to be adequate in addressing essential engineering functions and quality of work products.
4.0 Exit Interview At the conclusion of the inspection on February 7, 1992, the inspector met with licensee representatives denoted in Attachment 1.
The inspector summarized the scope and results of the inspection at that tim ATI'A HMENT 1 Nia ara Mohawk P wer o
rati n R. Abott, Manager, Unit'2 Engineering M. Annett, Design Engineer, Unit 1 J. Bebko, General Supervisor M. Bullis, Training B. Connolly, Supervisor, NMP1 QE Surveillance J. Conway, Manager, Technical Support, Unit 2 B: Crandell, Systems Engineer A. Curran, Site Licensing K. Dahlberg, Unit 1 Plant Manager J. Dillon, Supervisor, NMP2 QE Surveillance B. Drews, Manager, Technical Support, Unit 1 P. Finnerty, Supervisor, Engineering - Cost/Schedule P. Francisco, DBR Program J. Gawler; Design Engineer, Unit 1 D. Goodney, Electrical Design, Unit 1 G. Inch, Engineer, Unit 1 M. Jones, Supervisor, Unit 2, Plant Evaluation E. Klein, Project Manager L, Klosowski, Supervisor, Unit 1 Designs J..Kroehler, Manager, QA Engineering E. Lapidus, Unit 2, Designs P. Mangano, Supervisor, Site Engineering, Unit 2 M. McCormick, Plant Manager, Unit 2 T. McMahon, Electrical Design, Unit 1 W. Nowicki, Sr. Engineer, Unit 1 G. Pace, Supervisor, Procurement D. Pike, Supervisor, Unit 1, Plant Evaluation A. Pinter, Site Licensing Group A. Raju, Electrical Designs, Unit 2 M. Ritzwell, Project Engineer, Unit 2 D. Sandwick, Supervisor, NMP1, Project Management J. Spadafore, Program-Director, ISEG J ~ Sullivan, Supervisor, Unit 2, Project Management K. Sweet, Manager, Unit 1,-Maintenance T. Syrell, ISEG C. Terry, Vice President, Nuclear Engineering J. Vierling, Project Engineer B. Walker, Supervisor, Site Engineering, Unit 1 K. Ward, General Supervisor, Design, Unit 2 D. Weaver, Supervisor, Procurement Engineers
Attachment
R. Wyatt, Electrical Design W. Yaeger, Manager, Unit 1, Engineering U.S. Nuclear Re ul mmi i n W. Schmidt, Sr. Resident Inspector Denotes those attending 'the exit meeting conducted on February 7, 1992.