IR 05000285/1992025
| ML20126D699 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/18/1992 |
| From: | Chamberlain D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20126D690 | List: |
| References | |
| 50-285-92-25, NUDOCS 9212280113 | |
| Download: ML20126D699 (33) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-285/92-25 Operating License:
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Licensee: Omaha Public Power District (0 PPD)
444 South 16th Street Mall Mail Stop 8E/EP4 Omaha, Nebraska 68102-2247 Facility Name:
Fort Calhoun Station Inspection At: Blair, Nebraska i
Inspection Conducted:
September 28 through October 2 and October 28, 1992 Inspectors:
L. E. Ellershaw, Reactor Inspector, Materials and Quality Programs Section, Division of Reactor Safety M. F. Runyan, Reactor Inspector, Plant Systems Section Division of Reactor Safety i
R. V..Azua, Resident Inspector, Project Section C Division of Reactor Projects I. Barnes, Chief, Materials and Quality Programs Section Division of Reactor Safety Approved:
IOIl#YYL l l8 91 Awight Chanberlain, Deputy Director
' D' ate Division of Reactor Safety 1812pction Summary Areas Insoected:
Routine, announced inspection to determine the effectiveness of the licensee's program for assuring the reliability and operability of safety-related check valves.
Results:
The Fort Calhoun check valve program was found to be sufficient. There o
appeared, however, to be a lack of formal coordination or information sharing between this program activity and other programs which might affect it (paragraph 1.2).
9212280113 921218
{DR.ADOCK 05000285 PDR
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-2-The check valve program is currently based on preventive maintenance o
- activities, which consisted of disassembly and visual inspection.
Acoustic 9 missions monitoring diagnostic testing is planned on being introduced by the end of the year (paragraph 1.2).
Guidelines and positions contained in Generic Letter 89-04, " Guidance on o
Developing Acceptable Inservice Testing Programs," were appropriately addressed. Surveillance tests have been performed in accordance with the guidelines and within the prescribed frequencies (paragraph 1.3).
The licensee had a check valve design review performed in order to o
determine whether valves were of the proper type and size for operating conditions, and were properly oriented and located at a suitable distance from sources of turbulence.
It was noted that several elements of the design review were not in accordance with the guidelines established by the Electric Power Research Institute (paragraph 1.6).
Errors existed between the technical database and isometric drawings and o
it appeared that a comprehensive overview by the licensee of the design-review process had not been performed (paragraph 1.6).
The licensee's review of generic letters, information notices, o
bulletins, and 10 CFR Part 21 reports were satisfactorily documented, and required responses had been made to the NRC (paragraph 1.7),
Check valve performance trending had not been established within the o
check valve program since sufficient. data points did not exist and non-intrusive testing had not been implemented (paragraph 1.8).
Summary of Inspection Findinas Inspector Followup Item 9036-01 was reviewed and remains open pending o
licensee submittal of Revision 2 to Licensee Event Report 91-010-(paragraph 2 1).
Inspection Followup Item 9102-01 was closed (paragraph 2.2).
o Inspection Followup Item 9102-02 was closed (paragraph 2.3).
o Inspection Followup Item 9102-03 was closed (paragraph 2.4).
o Licensee Event Report 91-10 remains open (paragraph 3).
o Attachments Attachment 1 - Persons Contacted and Exit Meeting o
Attachment 2 - Surveillance Test Records Reviewed o
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-3-Attachment 3 - Check Valve f ailure History o
Attachment 4 - Information Presented by The Licensee During the o
Technical Meeting Conducted October 28, 1992
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-4-DETAllS 1 PERFORMANCE OF SAFETY-RELATED CHECK VALVES (2515/110)
The purpose of this inspection was to determine the effectiveness of the licensee's program to provide assurance of the operability and reliability of check valves in safety-related systems.
1.1 Backaround In recent years, numerous deficiencies related to check valves have been identified throughout the nuclear industry.
Information pertaining to these deficiencies has be<n disseminated by the NRC in Information Notices, and by the Institute of Nuclear Power Operations (INP0) in Significant-0perating Experience Reports (S0ERs).
The Fort Calhoun Station (FCS). check valve program was established in response to recommendations contained in INP0's 50ER 86-03, dated October 15, 1986. The licensee awarded a contract to combustion Engineering, Inc. (CE), on August 26, 1988, to determine the scope of check valves to be included in the program and to perform an applications design review of the check valves selected in order to adequately respond to SOER 86-03. The purchase order also requested that Electric Power Research Institute (EPRI) Report NP-5479, " Application Guidelines for Check Valves in.
Nuclear Power Plants," dated January 1988, be used as a guide in the performance of the design review.
CE completed the design applications review and submitted their report to the licensee on March 30, 1989.
In addition, the licensee awarded a contract to Applied Power Associates to perform a review of all FCS check valve maintenance history records in order to provide a report listing all work documents and dates associated with check valve maintenance. This report was issued in July 1988. The licensee established a check valve program basis document titled " Check Valve Reliability Program,"
which was issued as Revision 0 on October 26, 1990. This document provided the history, scope, and technical database established by the CE design applications review and the Applied Power Associates report'.
It also established a check valve program population which currently is comprised of 286 check valves, including all 135 check valves that are in the-In-service Testing Program (IST).
The inspectors reviewed the Piping and Instrumentation Drawings (P& ids) from j
the following safety-related systems in order to establish with reasonable i
certainty that all safety-related check valves had been included in the Check Valve Program:
service water; auxiliary feed water; diesel air start; chemical and volume control, and residual heat removal and low pressuro injection. No exceptions were identified.
1.2 Check Valve Proaram The Check Valve Reliability Program, currently Revision 3, dated September 14, 1992, previded the program elements which are designed to identify existing undetected check valve failures, incipient failures, and to prevent future
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-5-failures through the use of monitoring and/or inspection activities. The two elements that.were implemented included preventive maintenance (disassembly)
and check valve inspections (visual inspection and mechanical exercising).
Procedures for preventive maintenance and inspection had been developed and implemented. A third element, check valve monitoring, had rbt been implemented. The various monitoring techniques (i.e., non-intrusive examination methods) had been evaluated and the licensee stated that they had selected the acoustic emissions monitoring method with an implementation target date of December 31, 1992.
Responsibilities for implementing the various program elements were addressed, as were those of other groups whose actions could impact the check valve program (i.e., design engineering, IST, maintenance, system engineering, etc.).
The inspectors noted, however, that the responsibilities were not integrated and there did not appear to be any required coordination and notification between the check valve program engineer and other groups.
During comparison of IST program and check valve program populations, the inspectors noted that there were five valves in the IST program which were not in the check valve program. Discussions with the IST coordinator and-the check valve program engineer revealed that the five valves (IA-PCV-6680A-1-C,
-6680A-2-C, -6G808-1-C, -66808-2-C, and -6682-C) had been installed May 1, 1990, using modification package MR-FC-87-20.
The operability check was performed and the modification was approved and documented by the System Acceptance Committee in meeting minutes FC-0274-91 on February 14, 1991. At that time, the valves were added to the IST program. The inspectors noted that the required IST surveillance tests were performed on_the valves at the appropriate times.
Since there was no formal mechanism requiring notification, the check valve progrum engineer was unaware of the addition of the five valves, thus, they were not entered into the check valve program.
It was also noted that check valve corrective maintenance activities, when performed, are not formally brought to the attentf on of the check valve program engineer. The engineer must rely upon his review of the Computerized History and Maintenance Planning System to determine what maintenance activities might have had an impact on check valve program valves. His review is documented in a quarterly report to his supervisor.
There did not appear to be any defined methodology for increasing preventive maintenance activities upon identification of valve operating problems.
The engineer explained that engineering judgement would be the basis for increased activities. The inspectors acknowledged the use of engineering judgement; however, it was pointed out that there were neither defined processes nor expectations in.this area.
The inspectors were informed that the lack of formalized coordination between other groups and the check valve program engineer had been self identified; and in addition, according to memorandum PED-SSE-92-0476 dated June 17, 1992, it appeared to be a generic issue with regards to Special Services Engineering Programs. The proposed resolution to this issue stated that changes to configuration control procedures would be drafted and submitted to provide a
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W-6-formal means of notification to the check valve program engineer for the purpose of integrating information. The inspectors were informed that the last of the identified ciianges would become effective by December 31, 1992.
1.3 Surveillance Testina The inspectors selected 15 safety-related check valves (identified in Attachment 2) that were contained in both the IST program and the check valve program and reviewed all applicable records pertaining to surveillance testing performed between January 1, 1990, and the time of this inspection. The inspectors verified that the appropriate inspections and/or surveillance tests were performed in accor(ance with the programs' established frequencies. As a result of the supplementary fashion in which the two programs have been developed (i.e., the check valve program may forgo check valve inspections by taking credit for IST surveillance tests), a review was-performed which verified that the sample valves were either tested and/or inspected and that none of the valves had been inadvertently exempted from both programs. A review of all the applicable surveillance test procedure records associated with the 15 valves was made for technical adequacy and clarity, and no discrepancies were identified.
For full-stroke tests, the licensee met the criteria established in Generic Letter 89-04 (i.e., for those valves not having a visible position indicator, the open position may be verified by passing the maximum required accident condition flow through the valves).
In addition, the inspectors verified that some of the valves that should be back flow tested, such as the chemical and volume control system and volume control tank outlet check valves, did receive those tests. Other valves, such as the cump discharge check valves located in the safety injection and containment spray systems, are not back flow tested due to the availability of a pump
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l discharge isolation valve located downstream from each of the check valves.
The inspectors' review indicated that all of the IST surveillance tests were
performed within the specified frequencies and were in accordance with the IST surveillance test procedures.
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1.4 Maintenance Proaram The inspectors reviewed aspects of the maintenance program for check valves to determine whether processes and programs existed to identify degradation
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The maintenance history for each of the 15 valves was reviewed. As mentioned
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previously, some of the valves had not received preventive maintenance (i.e.,
l disassembled and visually inspected). This was because the licensee, in some instances, took credit for IST surveillance tests, and in other instances, l
performed sample disassemblies. This sampling technique was established and used when the licensee determined that it was not necessary to disassemble and inspect all applicable' valves each refueling cycle.
The technique required that-the valves be grouped by design (manufacturer, size, model number, and materials of construction), application, and that they have the same service conditions including valve orientation.
One valve from a group would receive
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_ group would be schedule' d to receive preventive. maintenance during a' subsequent outage.- Of the valves reviewed, only three had-received preventiveL maintenance since January 1, 1990 (SI-135, SI-208,-and S1-211). The-maintenance records 'showed that all three of the valves were-found to be in.
excellent condition,'with no appreciable wear noted on the-valve internals- (no
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loose or corroded parts).
In addition, the licensee verifled,;while the valves were disassembled, that each was capable of _ full-stroking.
The only maintensnce activities associated with the 15 selected valves, other than visual inspection, were performed on the 3 valves identified above.
These maintenance activities were not the result of worn r ets or of the valves falling to perform their functions.
For. valves-SI-208 and SI-211, the-maintenance activities were directly--related to disassembly of the valves 1for visual inspection.
It was discovered during reassembly that the body pressure sealing area (inspection port) was out of round and required machining-to allow reassembly and prevent leakage. As a result of this. condition and the difficulties involved in correctill and reassembling these two valves, the licensee developed surveillance test procedur_es to perform' full-stroke and partial stroke tcsts during refueling outages and cold shutdown conditions.
The licensee-had also altered their inspection schedule to preclude future-visual inspections of these two valves and to take credit for the IST surveillance tests.
The inspectors requested a historical printout which would show all preventive-maintenance performed on )rogram check valves.
The printout showed_that since-January 1990, _ there have seen a. total of 29 check valves that have:been disassembled and inspected,10 of which were in the IST-program.
1.5 Check Valve Failure Rates Since none of the valves reviewed had a history of failures-since January.1, 1990, the inspectors were unable-to verify the implementation of the licensee's procedures for reporting,- evaluating,t and correcting check. valve problems. The inspectors did, however, review the 1_icensee's procedures which
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included Standing Order Procedure S0-R-4 " Station' Incident Reports,"
Revision 35,- dated October 5,1992,. and Nuclear Operations: Division Quality -
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Procedure N00-QP-31:" Operability and Reportability Determinations,"-
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Revision 3, dated August 28, 1992. As described in Procedure _50-R-4, incidenti reports are written when the as-found data of surveillance tests are outside the acceptance criteria specified in the Technical Specifications or Updated o
Safety Analysis Report-(USAR), or for any significant occurrence in plant operations which-have safety implications.
Procedure N00-QP-31 addressed 'all other conditions that do not rise to the level of an incident report but still require the. determination of operability. These procedures were found to L
provide clear direction as'to when they should be used,~how:theylare to be l
'implemonted, and who is responsible:for their implementation.
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L The overall check valve' failure rate per year'is a statistic maintained by the L
licensee and is part of the check valve program document (Appendix C, " List of L
Known Check Valve Failures").
The check valve program engineer's definition t-L j -.
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-8-Known Check Valve failures"). The check valve program engineer's. definition of failure was the point at which a check valve could no longer perform its intended function.
Valve identifications and the year (s) of failure are documented in Attachment 3 to this report.
1.6 Desian Acolication Review As discussed above, the licensee contracted CE to perform a design application review of FCS's check valves. The scope of this review included check valves installed in systems listed in INP0's SOER 86-03, " Check Valve Failures or Degradation." Additionally, check valves designated as ASME Class 1, 2, or 3; high energy check valves; instrument air accumulator / valve pressure retaining check valves; and check valves identified as being problematic were included-in the scope of the review.
In all, the design applications of 279 check valves were analyzed.
The primary technical basis for the review was EPRI's Report NP-5479, " Application Guidelines for Check Valves in Nuclear Power Plants."
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The inspectors reviewed the final report on the check valve design review, which was transmitted from CE to the licensee on December 20, 1989. The report concluded that "the design review was performed consistent with the requirements of the SOER and confirms that no check valve misapplications were identified that would preclude the performance of their design function." The inspectors contacted CE through the licensee and learned that this statement meant that no gross misapplications were found that would obviously preclude normal function, such as a swing check valve positioned in a vertical line with the disk under the seat. CE identified many misapplications that could cause accelerated wear based on an analysis of flow velocities and the proximity of nearby upstream flow disturbances.
Recommendations made by CE to address these misapplications and the licensee's response to them are discussed later in this section.
The inspectors identified several examples where CE failed to implement the guidance of EPRI NP-5479.
For flow disturbances caused by upstream control valves, EPRI-5479 stated that 10 pipe diameters should be used to define the area of wear-related turbulence.
In its review, CE used 5 pipe diameters for this application. The inspectors questioned CE through the licensee on this non-conservative variance from the EPRI guidance and were informed that the individuals who had performed the review were no longer working for the company. CE was unable to provide a reason for the use of 5 pipe diameters in lieu of 10. Additionally, the inspectors noted that the calculated minimum flow velocities to achieve full disc lift had not been adjusted to account for the presence of nearby disturbances. The EPRI report stated that flow disturbances within 1 pipe diameter.an increase minimum lift velocities by 40 percent. Another factor, which affects the minimum lift velocity,-is the-inlet orientation of the valve, i.e., between horizontal and vertical.
EPRI NP-5479 presents calculations to estimate the effect of an angled orientation on the minimum lift velocity of swing and tilting disk check valves. However, it appeared that CE estimated minimum lift velocities in all cases using calculations based on a horizontally-oriented check valve that was free of
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-9-turbulence. When questioned, CE could not provide a basis for not using the EPRI guidelines for making corrections to minimum flow calculations.
The inspectors concluded that the identified nonconservative deviations from the EPRI guidelines probably did not have a significant impact on the final product of the CE report, which was the assignment of priority levels to check valves foe the purpose of scheduling inspection and testing.
The CE report divided the check valves into three priority levels based on a summation of points assigned for each adverse factor.
The licensee stated that check valves assigned to the lowest priority level (3) were given less attention in the planning of inspection and testing activities.
Some changes to the assigned priority levels would have probably resulted from a more thorough utilization of the EPRI guidelines.
It was speculative whether these changes would have altered the makeup of the licensee's inspection and testing schedule. The accuracy of the database, though, was clearly diminished.
The licensee's failure to perform a meaningful review of the CE report contributed to this situation as well as the fact that CE's services were procure ' under a contract that did not specify 10 CFR Part 50, Appendix B requirements.
The inspectors identified other errors in the CE report regarding the accuracy of recorded technical information.
Through a review of six isometric drawings, the inspectors found examples where the information listed in the technical database was not accurate. Check valves AC-504 and AC-505 were listed as having a horizontal inlet orientation when in fact they were located in a vertical line. This was confirmed during a plant walkdown. These were swing check valve.e, which are not generally placed in vertical lines. The licensee had earlier informed the inspectors that there were no swing check.
valves located in vcrtical lines, unaware of the error-made by CE. The significance of this finding was lessened by the fact that valves AC-504 and AC-505 are located in the condensate system and are, thus, not safety-related.
Check valves FW-161 and FW-162 were listed as having elbows 1.5 pipe diameters upstream,- but the elbows were actually located at 0 pipe diameters using the convention used elsewhere ir the report.
Additionally, the elbow upstream of FW-162 was incorrectly assigned a disturbance orientation of 0 degrees (instead of 90). The discovery of these errors based on such a small review suggested that the data collection process for the design revi9w lacked the normal standards of precision anticipated of a contractor ani reemphasized the significance of the licensee's failure to perform a meaningful review of the CE report, lhe inspectors noted that the CE report lacked a comprehensive depth of analysis in certain respects.
For example, flow disturbances were all treated equally without regard to the type of disturbance' or its varying distance to the check valve.
Thus, an elbow located 0 diameters from a check valve was -
assigned the same importance (in the assignment of points to determine priority levels) as an elbow located at 5 diameters. The relatively minor flow disturbance created by a long elbow was considered equivalent to the vigorous' turbulence caused by a control valve. Also, except for pumps, no consideration was made for the effect of multiple disturbances located within
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x-10-5 or 10 pipe diameters. Only the most immediate upstream disturbance was considered in the review.
The inspectors requested for review several check valve system and minimum velocity calculations.
The licensee stated that none of these calculations had been retained by CE and that they were not available on site. Though not required, retention of the calculations would have provided useful baseline information.
The inspectors reviewed two recent modifications to determine whether the licensee was utilizing EPRI NP-5479 or similar guidance in its design review process for installing new check valves or when modifying existing ones.
Modification HR FC-87-14, Revision 3, installed new check valves CH-469 and SI-343.
Engineering Change Notice (ECN) 92-14, Revision 0, replaced check valves AC-189 and AC-190 with valves of slightly different design.
In neither case was it evident from a review of the modification package that the technical aspects of check valve design were addressed.
Specifically, there was no documented analysis of system versus minimum flow velocities nor discussion of the effects of upstream flow disturbances.
Based on a discussion with the responsible engineer, it appeared that these design elements were considered, but not formally documented. As a related issue, the inspectors were concerned that the licensee's engineers had not received recent training in the area of check valve design. This was reinforced by information provided by the licensee showing that no licensee employee had received formal training on the guidance of EPRI NP-5479 other than a 2-day EPRI workshop in 1987 (which was before the issuance of EPRI NP-5479 in 1988).
In its final report to the licensee, CE provided recommendations to enhance the performance and improve the wear resistance of the 279 analyzed check valves.
These recommendations included reorientation of valves, replacement with a different type of valve, relocation to a less turbulent area in the piping, periodic replacement with the same type of valve, and recommendations to inspect or test the valves on a priority basis. -The licensee had documented a position relative to the CE recommendations in correspondence dated December 17, 1989. Many of the documented positions were interim in nature and contingent on the results of planned inspections or tests. As of the date of this inspection, the licensee had not documented an update to its initial response. At the inspectors request, the licensee provided a verbal status of its current position on the CE recommendations.
In some cases, the licensee implemented or planned.to implement the CE recommendations. However, in the majority of the cases, the licensee stated that inspection or test results showed the check valves to be stable and that there was no plan to implement the recommendations. The inspectors did not have any immediate concerns with respect to the licensee's stated position'on implementing the CE recommendations, but considered it a weakness that disposition of the recommendations was not formally documented.
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-11-1.7 Industry Information The inspectors reviewed the licensee's response to a 10 CFR 50 Part 21 notification, two NRC Bulletins, and six NRC Information Notices pertaining to check valves. The Part 21 notification addressed Velan swing check valves, which had a tendency to jam in the open direction.
The licensee's response was that there were no Veian swing check valves in the plant, an assertion supported by the inspectors' review of the check valve database. A strength was noted in that the response included a discussion of any other types of valves that might have been similarly affected (none were identified).
The inspectors reviewed the licensee's respo'nse to NRC Bulletins 83-03 and 89-02 and NRC Information Notices 86-01, 88-70, 89-62, 90-03, 90-16, and 90-79, all of which pertained to check valves.
None of these generic communications were applicable to the licensee's check valves except Information Notice 88-70. This notice identified a series of deficiencies that had been observed in check valve programs that were reviewed during several NRC inspections.
Among the identified deficiencies were the failure to include check valves in the steam supply to the turbine-driven auxiliary feedwater (AFW) pump in the IST program, AFW check valves not in the IST program, the failure to reverse flow test check valves other than those utilized for containment isolation and reactor coolant pressure boundary, and inadequate leak checks.
As a result of this notice, the licensee added six valves to its check valve program. The inspectors' review of the licensee's check valve program did not reveal any existing deficiencies of the type delineated in Information Notice 88-70.
As in the case of the Part 21 notice discussed above, the licensee's investigation into the information provided by the NRC did not stop at the point of identifying that the subject valve type was not installed in the plant, but included a consideration of whether installed valves had similarities that could make them vulnerable to the same failure mechanisms.
The inspectors concluded that the licensee had appropriately documented their review of each item and had considered generic implications beyond the specific details presented.
1.8 Trendina The check valve program engineer was responsible for generating a summary on a quarterly basis documenting a review of check valve maintenance and testing activities, included was a listing of open and completed check valve maintenance work orders and a historical list of all known check valve failures. The inspectors reviewed several of the quarterly summaries and noted that they appeared to be useful for detecting failure trends.
Since a program for conducting non-intrusive testing had not been implemented, the licensee did not have a trending program for monitoring check valve performance trends. As such, predictive maintenance of check valves was not possible.
The licensee stated that it was considering the implementation of i
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-12-an acoustic monitoring program and that if implemented the information derived from this program would be trended.
2 Followup (92701)
2.1 (00en) Insoector Followuo item 285/9036-01: Auxiliary Steam Pioino in the Diesel Generator Rooms This item identified a concern with high energy auxiliary steam piping in the diesel generator rooms that had not been designed for seismic loadings.
In response to the NRC concern, the licensee performed Engineering Analysis (EA) FC-91-15 to analyze the piping for structural integrity.
The inspectors reviewed EA-FC-91-15, Revision 2.
This analysis concluded that the auxiliary steam piping in the diesel rooms would maintain its pressure boundary integrity and would not become detached from its supports during a design basis seismic event.
The inspectors determined that the engineering analysis provided in EA-FC-91-15, Revision 2, acceptably justified the conclusions of the structural adequacy of the piping. However, additional.information discovered.
by the licensee during this analysis revealed that one of the assumptions made -
in this analysis was not correct. That assuraption was that a high energy line break -(HELB) review had been performed during original plant design and that this review had indicated that the environmental consequences of a critical crack (which is required by Appendix H of the USAR to be assumed for a high energy system such as auxiliary steam) would be acceptable for the as-built
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configuration (i.e., would not result in the loss of both diesel generators).
A subsequent engineering analysis, EA-FC-91-43, Revision 0, concluded that a single critical crack of auxiliary steam in either diesel generator room uld have resulted in the loss of both diesel generators due to steam migration through a normally-open fire damper that would not actuav. in this scenario.
The train separation problem in the diesel rooms was identified by the licensee as a result of finding a similar problem in a different part of the-plant.
Licensee Event Report (LER) 91-10, Revision 0, reported that a critical crack of auxiliary steam piping in Room 07 could result in the' loss of both trains of emergency electrical equipment including both. trains of safety-relattd 480 volt motor control centers, panels to support alternate -
shutdown, and auxiliary feedwater capabilities. The licensee stated that the presence of auxiliary steam piping in Room 57 was unknown until it was identified by a licensee employee who was performing a walkdown in preparation for the NRC electrical distribution system function inspection (EDSFI).
The postulated scenario in Room 57 was identical to the one later discovered in the diesel generator rooms, steam migration through a normally open fire damper that would not actuate-to close at the relatively low temperature of the escaping steam. -At the time this condition was discovered, the 2" auxiliary steam piping supply to the diesel generator room heaters was routed through-Room 57.
The licensee's immediate corrective action for LER 91-10 was to' secure the steam supply to the diesel generator rooms and, thus, remove live steam from Room 57. - Another engineering analysis, EA-FC-91-31, was
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-13-initiated to evaluate the remainder of the auxiliary steam system for similar HELB concerns. A modification (MR-FC-91-17) was initiated to reroute the steam supply to the diesel generator rooms bypassing Room 57.
During the process of performing EA-FC-91-31, the licensee became aware of the diesel generator room concerns and separated that part of the review into
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another engineering analysis, EA-FC-91-43. The conclusion of EA-FC-91-43 was that the environmental conditions resulting from a critical crack in the auxiliary steam piping and the lack of isolation between the rooms could cause the potential for simultaneous loss of both diesel generators and related components.
Assuming a single failure for offsite power, the plant would then be without power. At the time this condition was discovered, the auxiliary steam supply to the diesel generator rooms had already been isolated as part of the immediate corrective actions for the Room 57 problem.
Before the steam supply was reinstated from an alternate piping line, the licensee replaced the fire damper between the diesel generator rooms with a fire door that would remain closed. This modification was performed under MR-FC-91-17. With the fire door in place, a critical crack in the auxiliary steam piping could affect the operability of no more than one diesel generator.
The inspectors reviewed the documents listed above and noted that the modifications necessary to address common mode failures in Room 57 and the diesel generator rooms had been completed. The inspectors, therefore, concluded that the immediate safety concerns had been resolved.
However, two other issues remained in question. One of these was the apparent failure of the initial design process to identify this condition. 'The licensee considered that the problem should have been identified during the original drawing review and physical walkdowns performed for the plant HELB evaluation in 1973.
The other issue involved the licensee's failure to provide a complete report on the environmental separation problem in the diesel generator rooms.
LER 91-10, Revision 0 and Revision 1, discussed the-auxiliary steam supply piping to Room 57 and to the diesel generator rooms, but only described the potential common mode failure in Room 57 and the actions taken to correct that problem. The inspectors notified the licensee that missing from this or any other LER was a description of the diesel generator room problem and the long-term corrective actions taken to resolve it.
The licensee committed to issue a supplement to LER 91-10, to describe the potential common mode failure of both diesel generators and the actions taken to resolve this condition. The reportability issue will be addressed as part of the followup to the issuance of LER 91-10, Revision 2.
The design issues and chronology of events associated with this matter were discussed with the licensee during a management meeting on October 28, 1992.
At that meeting, the licensee answered a series of questions posed by the NRC.
The licensee's handout material for this meeting is enclosed as Attachment 4 to the report.
In consideration of the material presented by the licensee, l
the inspectors concluded that once the Room 57 and diesel generator vulnerabilities were identified the licensee's handling of the design issues j
was appropriate and acceptable.
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-14-The inspector followup item is considered open pending review of the content of the licensee's reporting of the event and will be addressed by HRC followup to LER 91-10, Revision 2.
2.2 (Closed)
Insoection Followuo Item 285/9102-01:
Comoletion of Testina of Safetv-Related Heat Exchanaers Cooled by Raw Water The inspectors verified by review of test procedures that appropriate test program requirements had been established and implemented to verify the heat transfer capability of the component cooling water, letdown cooling, spent fuel pool cooling, and shutdown cooling heat exchangers.
Procedures included in this review were as follows:
(1) SE-PFT-CCW-0004, "SFP Heat Exchanger and-Circulating Pump Performance Test," issued February 22, 1992; (2) SE-PFT-CCW-0002, "AC-4A Shutdown Cooling Heat Exchanger Performance Test,"
issued April 29, 1991; (3) SE-PFT-CCW-0012, "AC-4B Shutdown Cooling Heat Exchanger Performance Test," Issued April 30, 1991; (4) SE-PFT-CCW-0003,
" Letdown Heat Exchanger," issued March 28, 1991; and (5) SE-PFT-CCW-0001,
"AC-1 Heat Exchanger Performance Test," issued April 25, 1992.
No anomalies were noted during this review, with the licensee test data indicting that satisfactory heat transfer capability was being maintained in the safety-related heat exchangers cooled by raw water.
2.3 (Closed) Insoection Followuo item 285/9102-02:
Establishment of a Proaram Basis Document for Inspection of Comoonent Coolina Water and Raw Water System Comoonents The inspectors reviewed the licensee's response letter to the inspection report dated February 13, 1991, and verified that the indicated selections for erosion / corrosion monitoring had been examined. No degradation was noted during these examinations. The inspectors concurred with the licensee's determination that further testing at these locations was not required. The l
inspectors ascertained that additional raw water system locations-had been added to the erosion / corrosion monitoring program as a result of observations of wear near the component cooling water heat exchangers. The inspectors additionally ascertained that the licensee was in the process of acquiring the
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latest EPRI erosion / corrosion program, which should improve the technical t
basis for control of this activity.
2.4 (Closed) Inspection Followuo Item 285/9102-03:
Review of Complete L
As-Built Verification The inspectors reviewed the licensee's actions-that were taken to complete its commitments made with respect to Recommended Action IV of Generic Letter 89-13.
These actions included:
(a) evaluation of tests performed on raw water / component cooling water heat exchangers, to verify capability to remove required post-accident heat loads (documented in OPPD Memoranda PED-FC-92-853 dated September 30, 1991, and PED-SYE-92-0865 dated August 19, 1992); (b) verification that the raw water system had been designed to accommodate a single failure and still meet its design basis accident requirements (documented in OPPD Memorandum PED-FC-92-626 dated April 7, 1992,
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-15-and Stone and Webster Analysis EA-FC-92-027); and (c) reconciliation of system walkdowns and design basis open items with the raw water / component cooling water design basis documents to verify system operability (documented in OPPD Memorandum PED-FC-92-641).
The inspectors concluded from the review that the licensee had satisfactorily addressed the remaining items of Recommended Action IV of Generic Letter 89-13.
3 ONSITE REVIEW OF LICENSEE EVENT REPORTS (92700)
3.1 IDppn) Licensee Event Report 285/91-10: Auxiliary Steam Pioina in Room 57 Outside Desian Basis (HELB)
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The inspectors' review of this issue is described in the discussion of Inspector Fol'owup Item 285/9036-01 in paragraph 2.1 above. The licensee committed to submit a revision to LER 91-10 that provides a description of the related HELB concerns in the diesel generator rooms.
This item will remain open pending review of the revised LE : n.
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ATTACHMENT 1 1 PERSONS CONTACTED OPPD
- C. Boughter, Supervisor, Special Services Engineering
- C. Brunnert, Supervisor, Operations-Quality Assurance
- fS.' Gambhir,-Division Manager, Production Engineering
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- J. Gasper, Manager,: Training
- fW.:Gatas,-Division Manager, Nuclear Operations
- R. Jaworski, Manager, Station Engineering
- fW. Jones, Senior Vice President
- L. Kusek, Manager, Nuclear Safety Review i R. Lewis, Principal Engineer
- D. Lippy, Station Licensing Engineer R. Lippy, Special Services Engineering, IST Program Coordinator
- K. Miller, Supervisor,- Mechanical Engineering
- T. Patterson, Manager, Fort Calhoun Station D. Rollins, Special Services Engineering, Check Valve Program Engineer
- fR. Short, Manager, Nuclear' Licensing
- 'J. Skiles, Manager, Mechanical Engineering
- S. Swearngin, Lead Engineer, Special Services Engineering
- B. Van Sant, Supervisor, Design Engineering-Mechanical NRC
- S.' Collins, Director, Olvision of Reactor Safety.
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- P. Harrell, Section Chief, Project Section C, Division of Reactor Projects..
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f R. Mullikin, Senior Resident Inspector, Fort Calhoun Station
- M. Runyan, Reactor Inspector,' Division of Reactor. Safety i T. Westerman, Section Chief, Plant Systems Section, Division of Reactor Safety In addition to the personnel listed above,- the inspectors contacted other licensee employees during this inspection period.
- Denotes personnel attending the exit meeting onsite on October 2,1992.
- Denotes personnel attending the technical and exit meeting in Region-IV-offices on October 28, 1992.
2 ' EXIT MEETING Exit meetings were conducted on October 2, and~0ctober-28, 1992. -During these-meetings, the inspectors reviewed the scope and findings of_the report.
The licensee did-not identify as proprietary, any information.provided to, or-reviewed by the inspectors.
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ATTACHMENT 2 i
SURVEILLANCE TEST RECORDS REVIEWED
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Yalve No.
Qescription SI-211 Safety injection tank SI-6D discharge check valve
SI-208 Safety injection loop 2A check valve SI-323 Check valve upstream of charging pump tap into high-pressure safety injection header SI-135 Containment spray pump 3A discharge check valve CH-203 Loop 1A charging line check valve SA-288 Diesel generator starting air compressor discharge check valve SI-102 High pressure safety injection pump discharge check valve FO-105 Diesel generator fuel oil check valve IA-HCV-306-C Instrument air check valve -
l IA-HCV-240-C Instrument air check valve SA-282 Diesel generator starting air compressor discharge valve SA-291 Diesel generator starting air compressor discharge check valve-CH-187 Charging pump CH-lC discharge check valve CH-198 Charging pump discharge to RCS check valve CH-189 Charging pump CH-1A-discharge check valve
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ATTACHMENT 3 CHECK VALVE FAILURE HISTORY Yalve Identification
_ Year of Failure FW-164 1978, 1990 AC-189 1983, 1990 AC-190 1984, 1988, 1992 Y-03 1986 AC-104 1988 SA-138 1988
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FW-162 1989 RW-125 1989 CH-187 1990'
CH-188 1990 CH-189 1990 RW-ll5 1990 RW-ll7 1990 RW-121 1990 WD-175 1990 SA-288 1991 IA-LCV-383-1-C 1991 FW-1334 1997 WD-272 1992-WD-273 1992
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ATTACHMENT 4 Information Presented by the Licensee During the Technical Meeting Conducted October 28, 1992-
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AGENDA
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Opening Coments
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Sudesh Gambhir Timeline of Events Jack Skiles Safety Significance Randy Lewis Answers to NRC Questions Jack Skiles Sumary Sudesh Gambhir l
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ANSWERS TO ElGHT NRC QUESTIONS -
October 28, 1992 l
Question 1 What was the Scope of the 1973 NELB evaluation?
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Evaluated Plant design based on A. Giambusso letter dated December 14, 1972.
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OPPD transmitted schedule for completion of analysis via letter dated December 22, 1972.
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Evaluation results lead to initiating modifications to complyLwith A. Giambusso letter. Modifications included installation of spray etc., which was completed by(AS), steel slats on main steam system, shields on Auxiliary Steam September 1973.
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Review to -identify the essential structures and equipment, required for safe and orderly shutdown-which conceivably could be damaged by a rupture of a high energy line outside Containment.
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Developed analytical-program to determine the effect of the postulated high energy line breaks on the-items or areas identified, including auxiliary steam.
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Significant analytical work included:
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Calculation of maximum stresses and identification of-postulated break locations in -the Main Steam and Feedwater piping.
- Calculation of maximum blowdown loads for pipe rupture conditions at locations of maximum stress on critical piping.
Development of conceptual designs to protect the Control
Room,'the electrical penetrations in the Switchgear _ areas l
.below Room 81, and-the Cable Spreading Room.
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Question la:
Should it have found this fssue?
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Based on the scope of the evaluation identified in Question 1, it can be concluded that this issue could have been identified.
However, it is uncertain as to what level of review was specifically performed, orisinally, for systems addressed in l
Section M.4.0 of USAR Appenc ix M.
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lt is known that critical cracks were assumed, jet impingement shields were installed on nearby Cable Trays containing essential cables.
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Design Basis Document Open Item (No. 29) exists regarding insufficient documentatior, to support some of the conclusions of evaluating DB0 Open items for piping. process of progrannatically Appendix M.
OPPD is currently in the '
I The design AE may have $erformed an analysis-and reached a similar
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conclusion to that whit we have reachea in answering Question #B.
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Question 2s How was the problan discovertd in Roan S7 and in the Olesel Generator Roons and discuss any relationship to:
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[0Sf!
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Design Res;nstitution.
c.
MRC Information Notices d.
MRC Inspectton FolIow-Up Iten 9036 01
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Due to our knowledge of design basis reconstitution, IEN 90 53 and l
the follow up item, the OPPD questioning attitude progressed on thic issue as discussed in Question 3.
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The EDSF1 self assessment walkdowns identified the Auxiliary Steam piping in Room 57, which was not in the original Appendix M.
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Question 3s Describe the sequence of events fron inittal discovery to
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the present time.
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December 14, 1972 - Letter From A. Giambusso (AEC) To J. L.- Wilkins
(OPPD) Issued.
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Establuhes criteria for HELB, Temperature > 200*F or j
Pressure > 275 psig.
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Verification that a rupture of a high energy pipe will not j
result in the loss of required redundancy of emergency l
clectrical equipment.
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Loss of ability to cope with accidents due to breaks
... leak too small to cause Reactor accident, but large
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enough to cause an electrical failure.".
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Evaluation to assure that Diesels and Batteries will remain i
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1973 evaluation included Auxiliary Steam (AS) piping.
March 1973 - OPPD letter from J. L. Wilkins to A. Giambutso responds to AEC requirements.
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Identifies essential structures and equipment.
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Identified approach to liELB resolution.
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Identifies modifications required to provide additional
protection.
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Reviews continuing for non-major sources.
May 1973 - Final OPPD Report, " Postulated High Energy Line Rupture Outside Containment" Final Report is issued l
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Scope all systems outside Containment T > 200* F and/or P > 275 psig.
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Excludes systems not_ normally pressurized.
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Circumferential break design temp. > 200* F and P.> 275 psig.
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Critical crack design temp. > 200 * F or. press-) 275_ psig;
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Auxiliary Steam was included in Scope (p. 45).
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-DG's were: identified as essential equipment ^(p. 40).
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Describes Modifications to protect Cable Trays-for Jet
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Impingement.
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August 1989 Plant level Design Basis Document PLDBD HE-11 " Internal Missiles and HELB" is issued.
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0,9en Item 29, Category 3 - Lack of documentation of analysis for HELB affects on non major sources - includes AS.
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August 16,1990 - IEN 90 053 " Potential Failures Of Auxiliary Steam Safety Related Equipment" is issued.perability Of Piping And The Possible Effects On O
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Discusses analysis of environmental effects of AS line ruptures in mild environments at Hillstone 1 and 2 and Haddam Neck.
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Reviewed IEN against USAR App. H, no indication of immediate concern.
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No safety concerns identified.
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Additional evaluation scheduled with Resolution Of Piping Issues At Fort Calhoun Station.
December 11, 1990 - NRC Maintenance Team Follow Up Inspection questioned seismic qualification of AS piping in DG Rnums.
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Seismic question only.
December 12, 1990 - DEN walkdown of AS piping in DG Rooms.
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Supports adequate for a seismic event.
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Concluded that at least one OG should be available in the event of HELB based on:
Availability of Jet Impingement Shield
e Statements in USAR
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Fire dampers observed closed.
February 25, 1991 - NRC Inspection Report 50-285/90-36(HTIFollow-Up).
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Seismic adequacy of piping system.
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Environmental effects.
March 20, 1991 - DEN walkdown of other electrical areas.
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Self assessment in anticipation of EDSF1 inspection.
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Discovers AS line in Room 57.
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March 22, 1991 - Fire Dampers Room 57E, 57, 20 Clo: 1d.
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Precautionary measure.
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Pending AS HELB environmental affects calculation for Room 57.
May 16, 1991 - Auxiliary Steam isolation Valve To Auxiliary Building closed.
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Precautionary measure until EA 90 031 completed.
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Eliminated AS HELB potential in safety related areas of Auxiliary Building.
l May 17, 1991 - HELB Evaluation in Room 57 Completed.
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PED-FC 91-1867 was issued.
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PRC discusses Room 57 HELB, DG HELB.
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PRC concurs Plant is outside its design basis for Room 57 HELB only.
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Call to DEN over DG Rooms - not a concern if the Dampers are closed.
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Event Notification made at 14:20.
" Efforts continuing to research other Plant areas for
HELB information on similar AS piping.
The LER will provide a current status of this effort."
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After the phone call, DEN was notified that Fire Dampers in the DG Rooms were normally lef t open.
July 2,1991 - LER 91-010 " Auxiliary Steam Piping In Room 57 Outside Design Basis" Revision 0 is issued.
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Discusses Room 57.
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Short Term - Closed Auxiliary Building AS isolation valve.
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Long Term - Perform EA-91-031 for Auxiliary Building AS.
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. Perform Mod to AS line in Room 57.
October 2, l'01 - Auxiliary Steam line in Room 81 is cut and capped.
'einanently isolated piping in Room 57.
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E!..ninated source for AS in DG Rooms.
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11, 1991 - EA FC-91-043 'Hiah Energy Line Break Study for October Auxiliary Steam In The Diesel Generator Rooms" is completed.
(formalized preliminary MR FC-91-031)
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Performed to establish basis for MR FC-91-017 " Room 57 HELB'
corrective actions.
Recommends establishing separation between rooms.
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States ' Loss of offsite power is not a credible event for concurrent postulation with AS line break in DG Rooms".
October 15, 1991 - EA FC-91-031 ' Potential-Failures Of Auxiliary Steam Piping And The Possible Effects On The Operability Of Vital Equipment" is issued.
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Covers all AS and Condensate Return (CR) piping.
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Credited Modifications already performed for elimination of i
(ventilation-forDd63(DG-1),14(DG2),and65 concern in Room 57 2).
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Discusses safety related equipment in Corridor 26.
- November 22, 1991 - LER 91-010. Revision 1 is issued.
DiscussesD/GRoomsandCorridor26correctiveactions.
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Discusses MR-FC-91-027 corrective actions.
January 7, 1992 - MR FC 91-017 ' Room 57 HELB, Phase II" is Accepted for Operability.
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Adequate HELB separation between rooms.
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Improved Room ventilation.
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Reduced inlet steam pressure.
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Eliminatod Room 57 piping.
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September 15, 1992 - NRC Review Of Open Item 90-036/01 For Closure ~of Unresolved Open Item U 9225-01, "Re)ortability Of l-
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HELB In DG Rooms" initiated to trac ( NRC Region IV.
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e Review of HELB_ issue in UG Rooms.
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October 11, 1992 - IR 920570;is issued.
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Follow up from NRC concern from September 15,=1992.
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l October *, 1992 - LER 91 010, Revision 2 to be issued.
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Discusses DG and Corridor 26 AS HELB in more detail.
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Discusses DG AS HELB corrective actions.
Expected October 30, 1992.
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Question 4s Why was the problem with the Auxiliary Steam in the Diesel Rooms and the subsequent corrective actions (including the replacement of the fire Daaper with a Fire Door) not reported under 10 CTR 50.727
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llELB concerns in Room 57 had previously been identified and reported.
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When Diesel Generator Room concerns were later confirmed, the potential to affect either Diesel had been eliminated as a result of actions taken to resolve the Room 57 concern.
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The one hour report to the NRC for Room 57 included a closing statement that read " efforts are continuin to research other Plant areas for llELB status of this effort
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Because the same piping served the Diesel Room that also passed through Room 57, and the significance and resolutions were similar, the Diesel Generator issue was treated as one event in further evaluations and corrective actions.
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Revision 1 of LER 91010 included the corrective actions which eliminated those problems.
In retrospect additional details of
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theDGRoomandCorridor26concernsshouldhavebeenaddressed.
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Engineering personnel involved in the HELB issue in Room 57 and the Diesel Generator Room believed that LER 91-010 and its l
subsequent revision encompassed both issues.
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Revision 2 to LER 91 010 is currently being prepared to more clearly state the details relative to the Diesel Generator Rooms and Corridor 26.
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Question 5s Why was the Auxiliary Steam problan in the Diesel Generator Rooms not discovered soon after the NRC identified a concern that a systen failure could affect safety-related equipment in these roons?
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Original verbal NRC concern was conveyed to be a question regarding the seismic adequacy-of AS piping in the DG rooms.
roach to resolving the concern had been discussed with the Our app (As recorded by RTC 90-148.)
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NRC.
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OPPD did perform a limited HELB evaluation and judged there to be adequate separation based on conditions at the time of the walkdown.
(December 12,1990)
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OPPD reviewed the MTI follow-up' transmittal 90-036 with HELB t
Fire Dampers--
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concerns included based on previous walkdowns.The separation was acce
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observed closed.)
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OPPD DEN was not aware that the Dampers were normally open until they were informed by System Engineering on May 17, 1991.
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Question 6s Why did EA-TC-91-015 nake an assumption concerning the validity of previous IIELB reviews when it was recognized that these reviews were conducted a long time ago?
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A limited review of the DG Room configuration, at the time of the walkdown, disclosed that adequate separation of the two DG Rooms existed to prevent a concurrent failure due to a HELB in either room. A single failure of either DG was acceptable since they were not required to mitigate the HELB event (trip.an AS system i.e.
HELB would not result in a Reactor or Turbine
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Jet Impingement Spray Shields are evident confirming that the l
modifications described in USAR Appendix H were performed.
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Question 7s Why was the Design Engineering staff originally unfantilar
with the requirements of & ndix M of the USAR to postulate l
a critical crack in a high energy 1ine?
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DEN is familiar with HELB requirements.
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HELB screening criteria is in gel 3, gel-29, and gel-63.
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Procedures for evaluation of HELB effects is in gel 6.
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Several modifications have been performed for HELB concerns.
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Question 8s Could a stens break of critical crack dimensions in the
'~c Auxiliary Steam line have produced steam in sufficient q
quantities to propagate through the open Fire Daaper in time to disable both Diesel Generators without detection or time
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for Operators to take compensatory actions.
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Loss of both Diesels due to an Auxiliary Steam line break would not have been considered a credible event.
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Loss of Off-site Power, Reactor trip or Turbine trip would not have been caused by the AS line crack in the Diesel Generator Rooms.
- No Reactor trip.
- No Turbine trip.
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Loss of both DG's would require immediate shutdown which could be done safely without Diesel Generators.
Analysis indicates that adequate time exists to prevent damage to both Diesels.
DG 2 exceeds limiting environment in about 5 minutes.
DG-1 still acceptable after 50 minutes.
- Six fire detectors in DG 2 Room Activate on the presence of steam.
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Alarm in Control Room.
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Operator response time to room less than 2 minutes
Operator cetion to discover source and isolate between e
10 and 15 minutes.
Could close RCV-978, MS-183, AS-ll2, all are
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easily accessed.
- Environment in DG-1 Room remains acceptable for DG operation.
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Loss of Off-site Power due to some external event autostarts both Diesels.
- Diesel radiator fans exhaust 100,000 cfm.
- Would evacuate environment before affecting electrical components.
Turbine would trip, thereby eliminating steam supply,
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Omaha Public Power District-3-NRC Inspection Report 50-285/92-25 bec to DMB (IE01) - DRS and DRP bec distrib. by RIV:
J. L. Milhoan Resident inspector DRSS-FIPS Section Chief (DRP/C)
MIS System RIV file DRP RSTS Operator Project Engineer (DRP/C)
Lisa Shea, RM/ALF, MS: MllBB 4503 DRS Section Chief (DRP/TSS)
Senior Resident inspector - Cooper Senior Resident inspector - River Bend L. Ellershaw I. Barnes M. Runyan R. Azua RIV:MQPS:Rl*
RI:PSS*
RI:PSC:DRP*
C:MQPS'
EC /TA-LEllershaw/lb MRunyan RAzua IBarnes GSNbN 11/16/92 11/16/92 11/18/92 11/19/92
__\\M1/92 DD:D %
D;ORP DCherlain Abach
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4/6/92
- Prevfously concurred
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