IR 05000387/1990024
| ML17157A569 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/12/1991 |
| From: | Eapen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157A568 | List: |
| References | |
| 50-387-90-24, 50-388-90-24, NUDOCS 9102210015 | |
| Download: ML17157A569 (16) | |
Text
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U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.
50-387/90-24 50-388/90-24 Docket Nos.
50-387 50-388 License Nos.
NPF-14 NPF-22 Licensee:
Penns lvania Power 8 Li ht Com an 2 North Ninth Street Allentown Penns lvania 18101 Facility Name:
Sus uehanna Steam-Electric Station Inspection At:
Cor orate Office, Allentown Penns lvania Sus uehanna Steam Electric Station Unit 152, Berwick, Penns lvania Inspection Conducted:
November 5 to November
1990 I
Inspectors:
D.
T'. Moy. Reactor Engineer L. J. Prividy, Senior Reactor Engineer R. Mathew, Reactor Engineer G.
Rangarao, Reactor Engineer Approved By date Dr.
.
K. Eapen, Chi.f, Special Test Programs Section, Engineering Branch, DRS IIns ection Summar:
Routine Announced Ins ection on November 5-16 1990 Re ort Nos.
50-387/90-24 50-388/90-24 A~Id:
Th f f
hf control of design, design changes, modification and temporary modifications.
Also included in the scope of this inspection were organization, staffing, communication between departments, quality assurance, training, management support and review of licensee's actions on previous NRC concerns.
Results:
No violations or deviations were identified.
One item remained unresolved regarding internal adjustment of molded case circuit breakers.
9102210015 910212 PDR ADOCK 05000387 PDR
DETAILS 1.0 Persons Contacted Penns lvania Power 5 Li ht Com an J.
A. Blakeslee, Asst.
Superintendent of Plant-SSES R.
M. Bogar, Electrical Maintenance Supervisor E.
S. Bragger, Sr. Project Engineer-Nuclear
- F. G. Butler, Manager, Nuclear Design C. T. Coddington, Sr. Project Engineer-Licensing N. Covington, Maintenance Engineer
"T. C. Dalpiaz, Asst.
Superintendent-Outages E.
W. Figara, Supv. of Maintenance G.
T. Glaser, 18C Production Supervisor M.
M. Golden, Plant Engineering Supv.
T. A. Gorman, Supv.
Engineering-Nuclear'.
Klinger, Installation Engineering Group Supervisor G. J.
Kuczynski, Tech.-Engineering Supv.
C.
A. Myers, Manager, Nuclear Project
"J.
E. O'ullivan, Senior Project Engineer L.
D. Oneil, Supv.
Engineering-Nuclear
"D.
F. Roth, Service Compliance Engineer K.
M. Roush, Supv.-Nuclear Instruction M. Scheibner, Engineer H.
G. Stanley, Superintendent of Plant-SSES
- R.
R. Wehry, Compliance Engineer
- C.
R. Whirl, Assistant Manager-NQA W.
W. Williams, Project Licensing Specialist U.
S Nuclear Re viator Commission
"G.
S. Barber, Sr.
Resident Inspector J.
R. Stair, Resident Inspector
- Denotes those present at.the exit meeting on November 16, 1990.
2.0 Objectives of the Ins ection The objective of the inspection was to assess the adequacy of the licensee's program for engineering support to plant operations.
This support includes that for design changes and other engineering work requests by the plan.0 En ineerin Or anization and Staffin The inspector reviewed the licensee's procedure
"Conduct of Technical Support,"
AD-QA-410, Revision 4, dated June 30, 1989 and other documenta-tion.
He also. held discussions with engineering management pers'onnel to determine the organization and adequacy o'f the staffing level of the engineering technical department at Susquehanna site.
The inspectors also reviewed training, qualification and experience levels of assigned personnel in various positions, and interviewed the staff engineers and managers to assess their knowledge and familiarity with their assignments.
Find i n ing department, and that findings were closed out in a timely manner gA audits are performed by PP&L bn a yearly basis.
The audits identified several findings and observations in the modification proces During this review, the inspectors noted that there has been a significant improvement i'n resolving the issues in a timely manner.
The audit findings and observations for the above audits were addressed by the licensee satisfactorily.
The inspectors determined that gA involvement in monitoring engineering effectiveness is adequate.
No unacceptable conditions were identified.
8.0 Technical Trainin The licensee has established an extensive fifteen ( 15) weeks technical training program in a period of five (5) years for both station technical support and other operations and maintenance staffs.
This training combines both classroom instruction and self study programs with direct working experience in the fields of BMR technology, system engineering, simulator training, quality assurance and safety determination.
All the operation and maintenance engineers participate in this program in addition to their SRO and maintenance training.
As per the licensee, this program is mandatory for all the site technical personnel.
The licensee also provides the basic level training in areas of administrative and gA/gC procedures, station qualified reviewer work and GET/plant access training.
The licensee indicated that their training facility has sixteen ( 16)
classrooms, the ( 1) simulator and five (5) laboratories which are well equipped for electrical, control instrumentation, mechanical, chemical and radiological fields'he licensee also has a total of sixty-eight (68) technical instructors to provide instructions both in classroom and laboratories, out of. which twenty-five (25) are hired contractors.
The
'raining chief and his staff report directly to the Nuclear Vice President.
The licensee's training program is described in Procedures NTP-(A-63. 1 and 63.2.
The licensee stated that the training program and these procedures were developed based upon INPO documents85-001, 88-002,88-012, 88-015 and 88-0022 and ANSI Standards.
At present, 90% of the staff at the site'ave completed their training except the newly hired who are currently undergoing the program.
The total curriculum provides the technical know
"how and on-the-job training for various engineering disciplines.
9.0 The licensee indicated that all the engineering staffs are budgeted to attend seminars and symposiums in their fields and seventy-five (75) of the staff represent IEEE, ANS and other technical committees,.
The licensee participates in the EPRI, NUMARC and BWR Owner's group and encourages the technical staff to take active roles to meet the station up-to-date requirements.
This training program appears to be adequate based on discussions with engineering personnel and the depth of the knowledge of engineering personnel observed during this inspection.
~En ineerin Work Re nests EWRs The inspector discussed the EWR backlog with the corporate engineer responsible for tracking all EWRs.
The EWR backlog as. of mid.-November had 2,651 open items.
The licensee indicated that the EWR backlog had been trending upward and that management was aware of the increasing trend.
The inspector reviewed the EWR backlog contained in a partial computer printout at the site from the compliance group.
There were approximately 100 EWRs that were older than two years.
The inspector randomly selected
.
14 of these EWRs.
It was noted that half of the EWRs addressed maintenance related issues (e.g.,
spare parts).
The inspector discussed this observation with the maintenance manager who indicated that engineering generally responded to maintenance needs in a timely manner.
10.0 Followu Ins ection Re ardin Adjustment of Molded Case Circuit Breakers MCCBs at Sus uehanna During NRC inspection 90-02, an allegation regarding the licensee's practice of permitting maintenance personnel to.disassemble and make internal adjustments to MCCB trip units in accordance with maintenance procedure MT-GE-008 was substantiated.
Concerns were also raised by the alleger regarding the licensee's practice of disassembly and adjustment of MCCB trip units without specific vendor published instructions or approval and industry guidance.
During this inspection, the inspectors noted that the licensee's procedure MT-GE-008 does not provide detailed information regarding the operations involved in disassembly/reassembly of the breakers.
The licensee stated that they did not contact. the vendor to verify whether their practice of MCCB adjustments could lead to degradation of the breaker.
The licensee stated that this activity'is performed in a controlled environment by
qualified personnel.
It should be noted that NRC generic letter 83-28 informed all licensees that an effective vendor interface program should be implemented for all safety related equipment.
The inspectors noted that no specific vendor published instructions or industry standards are available for these type of MCCB trip adjustments except for general maintenance and testing guidance.
MCCB manufacturers do 'not publish procedures for user's use.
~ The inspectors noted that even though the field test equipment and existing test methods were adequate to establish the basic function of breakers, it did not establish the original qualification tests performed by the'manu-facturer.
Verification tests to determine the functional capability of a breaker through out its range of operation requires, strict'adherence to manufacturer's detailed test procedures, shop drawings and actual service conditions.
Furthermore, the licensee did not perform an evalua'tion to determine the impact on the MCCB qualification due to the disassembly/reas-sembly operation and the application of lock-tite and use of a heat gun for this trip adjustment.
The inspectors inquired about the licensee's
CFR 50.59 review that
=
justified the reuse of a circuit breaker which has been unsealed and adjusted without the manufacturer's approval, and industry guidelines, and which has lost its UL approval due to disassembly of the MCCB.
During this review, the licensee stated that they are in the process of performing a
10CFR 50.59 evaluation to determine whether any unreviewed safety, question exists regarding this activity.
So far, the licensee identified ten breakers which had been adjusted.
The inspectors reviewed test data for two of the breakers (1B21613 and
.
1B23753) which were documented in Work Authorizations S22518 and S22520.
These breakers were Westinghouse HFB type magnetic adjustable type breakers which have an external adjustment dial.
The information regarding other breakers was not available during this inspection.
The licensee stated that they had adjusted internal trip settings for these breakers, since the as-found trip settings of each pole was different and the required set point would not be achieved with the external adjustment dial.
The licensee did not discuss these adjustments with the breaker vendor.
The licensee made the decision to correct the settings by opening the sealed molded circuit breakers and performing internal trip adjustments using their maintenance procedure MT-GE-008.
The as-left set points for these breakers meet the requirements mentioned in procedure MT-GE-008 and manufacturer's time/current curve.
The inspector noted that the licensee is not testing these breakers in the minimum and maximum position of magnetic set point dial and returning the trip setting to the as-found position prior to installation.
The licensee's practice of testing these breakers is not consistent with industry standards NEMA AB-1 and UL489..
The inspectors observed that these breakers were never tested after their original functional test dated July 1982 due to the lack of a formal maintenance testing progra Subsequent to the inspection, the safety evaluation. report for MCCB internal trip adjustment using Procedure MT-GE-008 was provided by the licensee.
This was to determine whether the changes, as a result of the modification, satisfied the requirements delineat'ed in 10CFR 50.59..
The licensee's safety evaluation describes the process by which intern'al magnetic trip adjustments are done to achieve the set point specified by engineering design documents.
The licensee's report stated that the internal adjustment process has no affect on equipment qualification and the process does not involve the installation of new components or materials, nor does it involve the removal of any existing component or material from the breakers The report further states that the breaker is restored to its original configuration and verified through the functional test to assure its proper breaker operation.
However, the inspectors noted that the licensee has not conducted any qualification testing or evaluations to determine if the breakers satisfy the original breaker qualification specifications.
The licensee's safety evaluation concluded that the MCCB trip adjustment and tests did not involve an unreviewed safety question.
This is an unresolved item pending 1) licensee's completion of a
CFR 50.59 evaluation; 2) discussion of the MCCB adjustments with the vendor to identity potential problem associated with this practice; 3) establish that the breaker satisfies all original functional criteria and; 4)
re-evaluation of their trip adjustment and breaker maintenance test procedure to confirm the as.-left MCCBs will function as intended with periodic testing (50-387/90-24-01; 50-388/90-24-01).
11.0 Licensee's Actions on Previous NRC Concerns.
11. 1 0 en Unresolved Item 50-387/89-29-01 and'0-388/88-32-01 This item pertains to the 'lack of adequate calculations to address the available voltage of the equipment under reduced voltage condi-tions and also 'during design bases events.'his issue was originally identified during NRC inspection 90-200.
During this inspection, the inspectors noted that the licensee had submitted a technical specifi-cation amendment request No.
89 to address the undervoltage issue.
This request is being reviewed by the NRC Office of Nuclear Reactor Regulation.
The licensee stated that they are in the process of initiating modifications to the 120 VAC system to eliminate the voltage concern.
This item remains open pending completion of the modification by the licensee.
11.2 0 en Unresolved Item" 50-387/88-21-01 and 50-388/88-24-01 During the NRC Appendix R inspection, the team observed that most of the discrepancies identified in the common power source concern analysis report SEA-EE-40 were addressed except for few load center breakers.
An additional concern was raised by the team regarding the
, lack of scheduled maintenance on circuit breakers at the 480V load center level'and belo.3 During this inspection, the inspectors noted that the licensee had e'valuated and corrected the discrepancies identified in report SEA-EE-040.
Appropriate set point documents were revised and the field breaker/relay set points wepe verified and corrected.
The licensee stated that during the review of this issue, they decided to increase the set point for breaker 2B24022 to improve the existing coordination.
This work is scheduled to be completed during the next Unit 2 refueling outage.
The inspectors did not find any unacceptable conditions and had no further questions regarding this issue.
The licensee stated that a formal breaker maintenance program to address all of load center, motor control center and distribution panel breakers (480Vac and below) is being developed to ensure proper breaker operation.
This item remains open pending the licerisee's implementation of a formal breaker maintenance program.
Closed Unresolved Item Ho. 50-387/88-18-01 dnd 50-338/88-21-01 On August 4, 1988, the 1'censee submitted to the NRC pursuant,to 10CFR 50.73(a)(2)(iv),
the Licensee Event Report 88-012-00.
This report documented the isolation of Reactor Wat'er Clean Up (RWCU)
system, an Engineered Safety Feature ( ESF)
system, whenever a
large Unit
pump motor was started.
The licensee evaluated the cause of the RWCU actuations and determined that a problem existed with the steam leak detection Riley Temperature modules, which were installed during the Unit 2 refueling outage.
These new Riley modules (model no.
164 e 5687 P103, 104)
had inherent problems involving their actuation during the voltage dips associated with the starting of a large motor, whereas the original old modules (model no.
163 c
1940 P001,2,3,4)
were not susceptible to these voltage dips.
The licensee decided to use the old Riley module after evaluating a
Based upon the evaluation, the licensee replaced all of the new modules with the old Riley modules for the RWCU system of Unit 1 and 2.
The new Riley modules are still used for other systems in Unit 2.
The licensee stated that the existing procedures would not allow them to start more than one large motor at a time as precautionary means.
Furthermore, other systems have more redundant channels and wider set point margin.
Hence, the licensee concluded there was a low potential to ESF actuation as a result of voltage dips in other systems that
.
use the new Riley module The licensee conducts monthly'surveillance and quarterly calibration testing programs on the old model Riley-modules.
If relay chatter or failure occurs during test of any module, it will be replaced by an old model Riley. module.
This item is closed.
11.4 0 en Unresolved Item No. 50-387/90-06-01 and 50-388/90-06-01 Control Room RG 1.97 Instruments Not S ecificall Identified During RG 1.97 inspection, during March 12-16, 1990, the NRC inspector-observed that the Category 1 instruments of both Unit 1 and Unit 2 had not been specifically identified as recommended by paragraph 1.46 of RG 1.97 regarding equipment identification.
NUREG 0737, Supplement
and NRC Generic Letter 86-33 were issued to the licensee on December 17, 1982, requesting the licensee to document any deviations from RG 1.97.
However, this deviation was not documented in the licensee's submittal to the NRC, dated May 31, 1984.
The inspectors reviewed the Susquehanna FSAR, paragraph 3. 13. 1, Compliance with RG 1.97 for the discussed deviations.
The FSAR indicates that only the instrumentation for accident monitoring not.
specifically identified on the control panels had not been evaluated against RG 1.97, Revision 1.
The licensee's position on RG 1.97, Revision 2 was transmitted to the NRC in letter PLA-965, dated November 13, 1981.
The inspectors'eview determined that the instrument identification deviation was not di,scussed in this letter.
No justification was provided in the FSAR as to why the specific identification was not required.
The NRR review of the Detailed Control Room Design Review (DCRDR) was issued oq May 30, 1990 to the licensee.
This review indicates that the licensee will incorporate the status of accident monitoring instrumentation labeling through the normal control room design change process and the staff will evaluate the licensee's actions related to RG 1.97 gu'idance on accident instrumentation labelling in a separate action.
This item remains open pending the licensee's evaluation and completion of RG 1;97 instrumentation labelling.
12.0 Plant Walkdown The inspector made a tour of the plant including control building, turbine building to observe the work in progress, housekeeping and cleanliness..
No unacceptable conditions were noted during this walkdow.0 Mana ement Meetin s
At the conclusion of the site and corporate inspection, both mini exit meeting (November 9, 1990)
and final exit meeting (November 16, 1990)
were conducted with the licensee's representatives to discuss the results and conclusions of the inspection.
At no time during the engineering inspection was written material provided to the licensee by the inspectors.
The licensee did not indicated that proprietary information was involved within the scope of this iri'spection.