IR 05000454/1989009
| ML20244C755 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 04/12/1989 |
| From: | Bongiovanni A, Hasse R, Lougheed V, Phillips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20244C749 | List: |
| References | |
| 50-454-89-09, 50-454-89-9, 50-455-89-06, 50-455-89-6, NUDOCS 8904200399 | |
| Download: ML20244C755 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
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Reports No. 50-454/89009;50-455/89006(DRS)
Docket No. 50-454; 50-455 Licenses No. NPF-37; NPF-66 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Nuclear Power Station, Units 1 and 2 Inspection At: Byron, IL 61010-9750 Inspection Conducted:
February 27 through March 16, 1989 e//x/M Inspectors: R ss Team Leader Date hovnni Yh/H A.
Date V. hAtM k
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P. Lougheed V//>//f Dat'e i
M. f, pu k i
F. Phillips, Chief 9//>///
Approved By.
Operational Programs Section Date
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i Inspection Summary Inspection on February 27 through March 16, 1989 (Reports No. 50-454/89009(DRS);
l No. 50-455/89006(DRS))
Areas Inspected:
Routine, announced inspection of design changes and modifications
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and the dedication of commercial grade equipment for safety-related applications.
This inspection was conducted in accordance with Inspection Modules 37700, 38703,
and 37828.
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Results:
Three violations were identified during this inspection: A violation I
of 10 CFR Part 50, Section 50.59(b)(1), failure to document the basis for l
concluding no unreviewed safety question existed, with multiple examples; a violation of 10 CFR Part 50, Appendix B, Criterion III, failure to provide
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adequate design control, with three examples; a violation of 10 CFR Part 50,
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Appendix B, Criterion XI, Test Control.
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The. licensee displayed ~a weakness in the area of 10 CFR 50.59 evaluations.
This appeared to be due, in part, to a misconception that the regulation does not apply to non-safety-related equipment and systems.. Weaknesses in-
'l the' design control'and post modification testing areas.were also identified.
This appeared to be caused in part by a weakness in the communication between
'the large number of organizations involved in the design change process.
Three Open Items were identified relating to dedication of commercial grade equipment (Paragraph 4.c), post-modification testing (Paragraph 3.b.(2)),
and failure.to include newly installed GE HPAISI relays.in the preventive maintenance program (Paragraph 3.b.(1)).
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There was evidence of strength in the area of self-assessment. Recent QA audits-had identified concerns similar to those identified by the inspectors;
. however, these were recent audits and corrective actions had not been completed.
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DETAILS
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Persons Contacted Commonwealth Edison Co. (CECO)
R. Pleniewicz,, Station Manager
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R. Ward, Technical Superintendent l
R.-- Flahine, Technical Staff Supervisor l
R.Steder,ModificationCoordinator(Byron)
j W. Pirnat, Regulatory Assurance-J. Phelan, Lead Electrical Engineer, Pressurized Water Reactor Engineering (PWRE)
G. Stauffer, Regulatory Assurance D. Winchester, QA Superintendent i
L. Stern, PWRE Engineer l
G. Contrady, PWRE Engineer E. Zittle, Regulatory Assurance N. Hallis, Modification Coordinator, PWRE W. Dijstelbergan, Field Engineering Supervisor, PWRE T. Kroll, Engineer, Project and Construction Services (PACS)
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W. Dean, Nuclear Safety E. Wobler, QA/QC Coordinator, PACS
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- C. Moerke, Project Engineer, Byron /Braidwood, PWRE U.S. Nuclear Regulatory Commission N. Gilles, Resident Inspector j
Other personnel were contacted as a matter of routine during the l
inspection.
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- With this exception, all personnel listed above attended the exit I
interview held on March 16, 1989.
2.
Status of Previous Inspection ~ Findings a.
(Closed) Open Item (50-454/87021-01; 50-455/87020-01):
No mechanism existed to assure all requirements contained in a cancelled procedure were covered by other procedures. The licensee had revised the Permanent Procedure Request Form (BAP 1310-T3) to require an assessment to determine if all requirements are covered l
by other procedures prior to requesting the deletion of a
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procedure. The inspectors were satisfied that this provided
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adequate control for deleting procedures.
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(Closed) Open Item (50-454/88022-01; 50-455/88019-01):
Scope of
post-modification ter'ing specified by PWRE in Modification Approval letters.
Tl open item is considered closed. The concern i
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will be combined with other testing concerns identified during this inspection and tracked as open item (454/89009-02; 50-455/89006-03) (see Paragraph 3.b.(2)).
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3.
Design Change and Modifications j
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The inspectors reviewed eight permanent plant modifications and a
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I selection of temporary alterations to determine if they had been conducted in accordance with regulatory and programmatic requirements.
a.
Description of Modifications Reviewed (1) Modification M6-2-88-004, " Revise Steam Dump Wiring": This modification moved the arming signal for the steam dumps from the auxiliary reactor trip breakers (which are normally de-energized with power removed) to the breaker and cell contacts.
Completion of the modification allows the steam dumps to be armed on reactor trip.
(2) Modification M6-2-85-128, " Revise AFW Test Valve Logic to Open on an Undervoltage and SI Signal":
This modification added cables and wiring so that valve AF-004A would open on an undervoltage and an SI signal. The valve is a test valve located directly downstream of the motor driven auxiliary feedwater pump. It is normally open, but is closed during the monthly surveillance testing of the pump. Prior to the modification,.if the pump was being tested and auxiliary feedwater was required, the valve would have to be opened at a local test panel.
(3) Modification M6-2-86-180, " Reactor Vessel Level Monitoring System and Refueling Cavity Level Measurement": This modification installed the Reactor Vessel Level Monitoring System. This system allows direct indication of the reactor vessel and refueling cavity level during refueling operations to eliminate problems that have occurred with the use of Tygon tube as a means of determining vessel level.
(4) Modification M6-2-87-166, " Replacement of the Diesel Generator Agastat Relay": This modification involved the replacement of the GPOR model Agastat relay with a more reliable GE HPAISI relay for the diesel generator (DG) governor circuit relay, 4EX3. The replacement resulted from relay failures at the Braidwood and Byron stations.
(5) Modification M6-2-84-118, "MSIV Nitrogen Charging Header":
This modification involved the addition of a nitrogen charging / pressure testing manifold and the associated nitrogen supply piping for each MSIV accumulator.
Each manifold consisted of two pressure gauges, nitrogen charging taps, i
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. accumulator supply piping with flexible hose connections, and
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pipe supports. Portions of the modification were considered.
l safety-related and seismically qualified.
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- (6) ~ Modification M6-2-87-132, " Install Time Delay Relay In Auxiliary Feedwater Pump Trip Logic": This modification installed a 2.5 second time delay in the low RWST suction pressure trip logic for the motor driven auxiliary.feedwater pump (AFWP) and a 4.0-second delay in the essential service water (ESW) ' suction l
. valve arming logic for the same condition. The intent of the
- modification was to prevent AFWP Trip and. arming of the ESW valves due to a low suction pressure transient during AFWP start.
(7) Modification M6-2-88-32, " Modify Trim on Auxiliary Feedwater Flow Control Valves": This modification replaced the trim in the AFW flow control valves to provide more stable flow control at low flow rates.
(8) Modification M6-2-88-87, " Remove Reactor Coolant Pump (RCP)
Trip Signal From Reactor Coolant Loop Stop Valve Limit Switches": This modification removed the RCP trip signal from the reactor coolant loop stop valve limit switches. The trip signal was considered to serve no useful purpose since the stop valves were locked open and de-energized prior to starting a RCP
'and a spurious RCP trip could result from a limit switch failure.
(9) Temporary Alterations: 'The inspectors reviewed approximately 25 Temporary Alternations. Temporary Alterations were defined as temporary changes to the plant involving no change in the function of the system or component involved.
b.
Results of Inspection The typical design change process involved many organizations.
The design is prepared by a consultant Architect Engineering Firm (A/E). The design is then transmitted to the licensee's. corporate engineeringgroup(PWRE). This organization reviews the design using a series of check lists to assure all relevant issues have been addressed. They then prepare a Modification Approval Letter transmitting the design details to the station. This letter contains, in part, the design details, post modification testing acceptance criteria, safety classification, and QA requirements.
The station then details the installation activity including the preparation of post modification testing procedures and the preparation of the installation work package (NWR). This effort is coordinated by the Station Technical Support staff. The work package is then turned over to Project and Construction Services (PACS) which is a corporate support group. The actual installation is then performed by a contractor to PACS. Specific modifications may deviate from this process which would be documented in a project plan.
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A The -inspectors identified 'several concerns which are discussed below:-
(1) Design Control
'The Modification Approval Letter for modification M6-2-88-004 identified the modification as safety-related and specifically required inspection of all terminations. While the terminals involved ~with the steam dumps were classified as non-safety-related, the modification required revision of wiring within the reactor protection switchgear which was safety-related.
The NWR prepared for the installation was also classified as safety-related. Contrary to this classification and the requirements of the Modification Approval Letter, the modification was installed as non-safety.related with no QC inspection performed. The rationale provided to the inspectors by the installing organizations was that the terminals involved and:the steam dump system were non-safety related and therefore the modification could be installed as non-safety-relateo. No effort was made to obtain the concurrence of the design ared engineering (This concurrence was obtained after the inspectors organizations prior to initiating the installation activity.
identified the discrepancy.) This failure to adhere.to the design requirements specified by the engineering organization is considered a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control (455/88006-01A).
Modification M6-2-84-118 involved the addition of.a nitrogen charging / pressure testing manifold to each MSIV accumulator..
The manifold was installed as safety-related in the safety
. valve room.
The design criteria, as stated in the modification
' approval letter and design drawings, specified that the safety-related non-ASME (Class "H") piping was separated from the non-safety-related class
"D" piping with a safety-related, seismically qualified globe valve. The inspector reviewed the purchase orders and receipt inspection documentation for these valves (2MS210 A-H thru 2MS212 A-H).-
The valves were purchased as commercial grade and dedicated for safety-related application.
Calculation BY-CQD-220559-00, performed by Sargent and Lundy,
seismically qualified the newly installed valves for Unit 1 and I
was referenced for the identical Unit 2 modification. The inspector noted that valves 1MS210 E, F, G, H and IMS211 E, F, G and H (also 2MS210 E, F, G, H and 2MS211 E, F, G, and H) were not included in the evaluation. Discussions with the licensee showed that these valves were inadvertently excluded from the evaluation. The licensee later provided the inspector with a new calculation which seismically qualified the valves for both Units.
Failure to seismically qualify these sixteen valves as required by design specifications is another example of a violation of 10 CFR 50, Appendix B, Criterion III (454/89009-01; 455/89006-01B(DRS)).
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The. replacement of the GPOR model Agastat relay in modification M6-2-87-166 was a result of several relay failures due to high'
contact resistance experienced at the Braidwood and Byron stations during DG surveillance. The existing' relays degraded due to a buildup of an oxide film on the contacts causing high resistance and failure. The engineering package included correspondence between Cooper-Bessemer Reciprocating (C&B) and Sargent and Lundy (S&L) regarding the replacement of the.
existing relay with a GE HFA 100 series model.
In a letter from C&B to S&L dated March 11, 1988, it was stated that GE personnel confirmed that the GE HFA relay should not be used in-low voltage circuits such as in the DG panels and suggested the l
use of mercury-wetted relays. The low voltage, low current circuit would not provide enough energy to burn through.the oxide film and would cause a buildup of an oxide layer on the contacts. The wiping action of the contact would aid in controlling the buildup; however, it would not eliminate the film. S&L responded via a letter dated March 17, 1988 and.
indicated that periodic cleaning and maintenance of the new relays would be required to compensate for film buildup.
Discussions with the licensee showed that the site personnel were unaware of the above correspondence and did not require the relays to be added to their preventative maintenance program.
The inspectors were concerned that although these relays were technically acceptable to use as a replacement, the licensee was still susceptible to the original root cause of the prior relay failures, namely, failure of these relays due to high contact resistance. The inspectors were also concerned that the Vendor Manual for the new relay could not be-located at the site. This failure to implement all design requirements (i.e.,
periodic contact cleaning) is another example of a violation of 10 CFR 50, Appendix B, Criterion III (455/89006-01C).
The inspectors had a concern with the manner in which the new supports to the Unit 1 accumulator fill lines were added (i.e.
by Temporary Alteration 88-1-61).
Review of procedures QAP-3-51, " Modifications" and BAP 330-2, " Temporary Alterations" did not indicate that a change of this nature (which involved a functional change) could be handled as a temporary alteration.
Review of the change has indicated that prior engineering evaluation (via a letter from the A/E) was obtained and that the work had been done in accordance with the applicable codes, and that a formal modification was in process to remove the temporary alteration. Discussions with the licensee indicated that they were aware that the normal design control processes had been bypassed, and that this should not have been done.
The root cause was the lack of a program for processing minor modifications. The generic problem had been identified in a QA audit. The failure to adhere to the design control process is a violation of 10 CFR 50, Appendix B, Criterion V.
Since the
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licensee identified.the problem and corrective action was in progress, ~no Notice of violation will be issued in accordance -
with.10 CFR Part 2, Section V.A.
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(2)- Post Modification-Testing The-inspectors identified several concerns with testing -
performed after modification-installation.
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- The' intent of modification M6-2-87-132 was to install a time delay in the motor driven AFWP logic to prevent pump trip
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and ESW suction valve arming due to a low suction pressure transient frequently encountered during pump start.
The logic diagram included in the engineering package was not consistent with the design intent in that it required the presence of both a " pump running" signal and a lcw suction pressure signal to. initiate the time delay relay -(TDR). With this logic, the time delay would occur regardless of when the-
low suction pressure signal initiated. The wiring schematic reflected the correct logic in that the TDR initiated on pump-start.- The acceptance criteria for the post modification test given in the Modification Approval Letter also reflected the correct logic; however, the test procedure called.for inserting the low suction pressure signal prior to pump start. -Thus,.the test as performed could not distinguish between the two logics and did not confirm that the correct logic had been installed.
The inspector had a concern with the discrepancy between the logic diagram and wiring schematic. The significance would-depend to some extent on the use of the logic diagram in.the design process -(i.e., the wiring schematic developed from the logic diagram or the logic diagram being'an independent depiction of the logic). The inspector had-greater concern
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with the failure of the post modification test to adequately confirm that the acceptance criteria had been met. There is no requirement that the post modification test be reviewed by i
the design organization. The discrepancy in this case could
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indicate the need for such a review or the need for the design organization to provide more detail. an the test logic.
The failure to confirm that the acceptance criteria had been met is a violation of 10 CFR 50, Appendix B, Criterion XI (455/89006-02).
The concerns descrioed below will be combined with open item
(454/88022-01;455/88019-01) addressing similar concerns and
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tracked as a single open item (454/89009-02; 455/89006-03).
For modification M6-2-83-004, the engineering organization l
specified only that a modification test be done with no consideration of operability testing. Onsite review also did not require any operability testing. The l
installation was performed as non-QA with no verification
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of the installation, although it was in a safety-related cabinet. Since no operability test was specified, the i
inspector was concerned that a reactor protection trip i
function may have inadvertently been affected with no means to detect it. When questioned, the licensee was
able to show that the normal reactor protection trip surveillance were performed after the modification had been ccmpleted. However, the general concern (whether i
adequate post-modification operability testing is being specified) remains.
During review of the engineering evaluation (via checklists)
for modification M6-2-85-128, it w'is noticed that certain tests were marked as being required. Particularly these j
were testing for the presence of sneak circuits and testing j
for valve failure position on loss of power. These tests l
were not included in the Modification Approval Letter, and were not performed. Since the site personnel do not see the engineering checklists, it is necessary for any required testing to appear in the Modification Approval Letter.
In a discussion with the licensee, it was stated that the sneak circuit testing was considered to be included, i
although not specifically addressed, and that testing on
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loss of power was not necessary, since this function had not been altered by the modification. The inspector's concern is that there is a " disconnect" between the engineering checklist requirements and the actual testing.
(3) Procedures The inspectors identified two concerns related to procedures.
The inspectors reviewed the procedures affected by modification M6-2-84-118 and noted one discrepancy. Procedure B0P-NT-M2, Revision 4 entitled, " Nitrogen System Valve Lineup," had an incorrect operating position and physical location for valve 2NT081. The licensee agreed to revise the procedure. The inspector performed a system walkdown and verified that the valve was placed in the correct (open) position.
Procedure 2BVS 7.1.5-2, Revision 2. " Main Steam Isolation Valves Partial Stroke Test" was conducted in conjunction with the modification testing on February 26, 1989. During this test, twenty-three typographical ermrs were identified which included references to the wrong +tt and incorrect equipment.
The inspector was concerned tha' A, Unsite Review (OSR)
function approved the procedure revision on February 21, 1989 and did not identify any of these errors. This is considered a weakness in quality verification by OSR. A temporary revision to the procedure correcting the errors was issued prior to
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conducting the test.
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(4)
10 CFD. 50.59 Evaluations The inspectors identified a general concern with respect to the documentation supporting 10 CFR 50.59 evaluations. While the inspectors identified no specific safety concerns, the failure to document the evaluation basis and/or adequately address the 10 CFR 50.59 criteria increases the probability that an unreviewed safety question would not be identified.
Specific deficiencies are discussed below:
The 50.59 evaluation for modification M6-2-88-004 did not provide justification to support the conclusion that no unreviewed safety question existed. The justification used
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"non-safety-related" rather than evaluating the modification i
with respect to the 50.59 criteria. The completed engineering checklists used by the licensee to direct attention to possible concerns did not provide backup support for the 50.59 evaluation. The failure to adequately document the bases as
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to why a modification was not an unreviewed safety question
is a violation of the requirements of 10 CFR 50.59(b)(1)
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(455/89006-04A).
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Additional axamples of the use of "non-safety-related" as a justification as to why a change did not constitute an unreviewed safety question were found in Temporary Alterations 88-0-1-009, 88-0-019, and 88-0-035. Additionally, Temporary Alteration 88-1-051 addressed the criteria by reference to a letter by the A/E. However, the letter did not address the criteria of 10 CFR 50.59 in any manner. These are
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additional examples of a violation of 10 CFR 50.59(b)(1)
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The inspectors had a concern with the documentation of the safety evaluation for Temporary Alteration 88-1-39 in that j
it did not adequately address the design purpose of the i
setpoint on a vacuum breaker when the breaker cover was
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removed (thus bypassing it). Further discussions with the
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licensee and review of supporting documentation showed that the concern was considered and resolved, but was not documented i
on the safety evaluation. The inspector considers this
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i specific concern resolved, although noting that inadequate or poorly documented safety evaluations were fairly common with the temporary alterations.
4.
Dedication of Commercial Grade Equipment for Safety Related Applications j
The inspectors reviewed a random sample of the QC receipt inspections and procurement packages to assess the licensee's commercial grade
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procurement program. The inspectors also held discussions with the i
station QC supervisor and the Technical Staff procurement engineer.
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Results of this inspection were as follows:
a.
The licensee procures the majority of their items as " Safety-Related, Part 21 applicable." According to both the QC supervisor and the
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procurement engineer, obtaining items under Part 21 has not been I
a problem and is their preferred procurement method.
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b.
A large portion of the commercial grade procurement are from the diesel generator manufacturer, Cooper Bessemer. This is due to Cooper having dropped their Part 21 program.
For items ordered from Cooper, a certificate of conformance (stating that the item is equal to or better than the original part) is required.
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Station Procedure BVP-100-2, " Classification / Procurement of Parts /
Components" contained no program for the dedication process with respect to identification and verification of critical characteristics.
The present system consists of an informal process of jotting instructions for QC receiving on the back of the procurement form.
This is not consistent with the requirements of the recently revised Quality Assurance Procedure QAP 4-51, Section 3.4 of the corporate QA manual which requires the identification of critical characteristics and the means for verifying them.
Resolution of this concern will be tracked as an open item (454/89009-04; 455/89006-05).
5.
Quality Verification Effectiveness The inspections reviewed two QA audits concerning the design change and modification program to determine if this licensee oversight group had identified concerns similar to those identified by the inspectors. These audits had identified concerns in the areas of design centrol and post modification testing.
These were two areas in which the inspectors had identified concerns (although not identical concerns). The audit findings were being addressed. This indicates that the licensee audit activities are effective in identifying problems. Since these were recent findings, effectiveness in obtaining resolution of the issues
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could not be assessed. The inspectors did have a concern with the j
failure of the onsite review function to identify the numerous errors in i
a post modification test procedure (see Paragraph 3.b.(2)). This could j
be an isolated case and precipitated by the fact that the procedure was j
a " markup" of the procedure used to test a similar modification on the
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other unit.
6.
Open Items f
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
the inspection are discussed in Paragraphs 3.b.(1), 3.b.(2), and 4.c.
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Exit Interview l
The inspectors met with licensee representatives (denoted in Paragraph 1)
at the conclusion of the inspection on March 16, 1989, and summarized the purpose, scope and findings of the inspection. The licensee stated
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that the inspectors had no access to proprietary information.
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