ML20046A371
| ML20046A371 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/07/1993 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | Milhoan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20046A364 | List: |
| References | |
| NSD930690, NUDOCS 9307280011 | |
| Download: ML20046A371 (5) | |
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GENERAL OFFICE P.O. BOX 499. COLUMBUS. NEBRASKA 88602 0499 I
TELEPHONE i402) s64 8561 Nebraska Public Power Deistr. t rAx m ss>sssi ic NSD930690
(% T June 7, 1993
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l 3-t Mr James L. Milhoan Reh;ional Administrator l
U.S. Nuclear Regulatory Commission Region IV l
611 Ryan Plaza Drive. Suite 1000 Arlington, Texas 76011 l
Subject:
Status of Issues Related to Unit Startup Cooper Nuclear Station NRC Docket No. 50-298, License No. DPR-46 l
I
Reference:
G.
R.
Horn to J.
L. Milhoan (NSD930497), dated Ap'ril 29, 1992, Sta'tus of Secondary Containment and Diesel Genera".or Fuel Oil Issues
Dear Mr. Milhoan:
The referenced letter provided the status of two issues that had previously been verbally identified by the NRC as requiring resolution prior to fuel reload at Cooper Nuclear Station.
Five other issues, liscea in the reference, had also been previously verbally identified by the NRC that require resolution prior to plant startup.
In addition, based upon telephone conversations between Mr. Bill Beach, NRC Region IV, and Mr. John Meacham. NPPD, on May 18,1993, three j
more issues were identified that also require resolution prior to unit startup.
1 This letter provides the current sta;.as of our efforts with regard to resolving these ten issues.
As of the date of this letter, fuel has been reloaded and reactor vessel reassembly is in progress.
j Surveillance Testi r of the fecendarv Contain ent As stated in the reference, the District continues to review certain programmatic changes designed to enb-
-he preservation of the secondary containment capability during the ari.
cycle.
These changes include upgrading the i
Joors and other penetration seals and loop existing periodic mainte enre seals, the frequency fot l'
they are performed, and the surveillance requirements. This effort is des.g sted as the Secondary Containment Action Plan and work is continuing towards i t.s completion.
31s plan was discussed with Mr. Elmo Collins (NRC-Region IV) during an inspection the week of May 10, 1993.
The plan is approximately 80% complete, with the remaining actions being primarily associated with preventive maintenance changes. The fuel reload issues have been resolved.
The remaining actions will be completed. prior to plant startup from the 1993 Refueling Outage.
Diesel Generator Fuel Oil Particulate Levels Diesel Generator fuel oil particulate levels, as of the date of this letter, remain below the 10 mg/l limit. Technical Specification and procedure changes are currently under review to specifically address this issue and other issues such as water sampling of the storage tanks. As noted in the referenced letter, we have retained the services of an industry fuel oil expert and are currently working with him to review and enhance our maintenance program to further assure continued Diesel Generator fuel oil quality.
3e fuel reload issues have been resolved.
9307280011 930721 9
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( J- [ 'JU PDR ADOCK 05000298 G
PDRg
James L. Milhoan Page 2 of 5 NSD930690 H2/02 Ans.ly:ers The primary problems identified with the H2/02 Analyzer System have been associated with the condensation and collection of water in the sample lines and with sample pump reliability.
During this outage, the pump internals were modified to enhance the performance of the pump with regard to its ability to pump a moisture laden medium.
The sample tubing has also been inspected and where necessary is being resloped to facilitate improved water drainage.
The previously installed particulate filters on the suction of the pumps were determined to be creating moisture problems (by acting as condensation chambers and thereby exacerbating the water problem) and were removed. Extensive pre and post-modif~ication testing will be conducted to verify that the previous water problems have been adequately resolved or additional measures will be taken.
The operability of the system during the past operating cycle will also be evaluated, following startup from the 1993 Refueling Outage. You will be advised of the results of :nis evaluation.
While the aforementioned modification efforts were underway, it was discovered that inadequate pressure testing of certain portions of the piping system within the analyzer cabinet had taken place. Certain components within the cabinet had not been subjected to full primary containment accident pressure to verify their integrity as part of the post modification testing following installation of the upgraded system.
3erefore, satisfactory testing of the remaining portions of the system within the cabinets will be completed prior to plant startup and any leakage found will be documented in our ILRT program. Other primary containment sampling system configurations will also be reviewed to ensure no other similar situations exist.
Local Leak Pate Testine Criteria for One Inch and Less Containment Penetrations As discussed with Mr. Phil Harrell, NRC-Region IV, during an inspection the week of May 10, 1933, there is no generic exemption for Appendix J testing of primary containment penetrations one inch and less in diameter. Appendix J Type B and Type C testing is conducted for contain=ent penetrations to the Cooper Nuclear Station licensed primary containment configuration.
Service Water System Interr W During NRC inspection activities the week of May 10, 1993, various arpects of Service Water Svs:em intecrity were reviewed by Messrs.
J.
Medoff and F. Grubelich of :he NRC-NRR.~ We understand that the only remaining open issue
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related to the Service Water System concerns its classification.
On May 14, 1993, Mr. Medoff stated at the exit meeting that the Service Water System did not appear to present any safet issues considering the number of inspections being performed, :y significant ne operating characteristics of the system, and the upgrades that have been completed.
However. the issue of compliance with 10CFR50.55a.g.(1) regarding the classification of the Service Water System was identified as an '.'nresolved Item pending further NRR review.
We therefore understand that this issue will be pursued further, but, unless otherwise notified, it is not being considered an issue requiring resolution prior to plant startup.
Pressure Isolation Valve 6 Local Leak Rate Testine A Pressure Isolatien Valve (PIV) testing program is currently bei..g implemented to further ensure nuclear safety at CNS. As you are aware, there is no current licensing commitment at this ' time, however", the recent leakage :esting of RHR-MO-17 and RHR MO-18 valves indicated that it is prudent. This testing will also provide trending to better predict required maintenance. ASME Section XI, leak rate testin is being used as guidance in determining test IWV-3420 A major portion o[e,this testing has been completed and no leakage rates methods.
in excess of the acceptance criteria have been encountered to date.
The only
Jamns L. Milhoan Page 3 of 5 NSD930690 issue of note in this testing is that the PIVs for the core Spray System do not currently have test connections to allow for Plv testing. The injection check valves will, nowever, be tested during the Primary System Leakage Test (Class 1 system pressure test) prior to plant startup.
A modification to install test connections for the Core Spray motor operated inboard in1ection valves will be performed prior to startup from the next scheduled refueling outage, that will allow testing of the required motor operated valves on the pump discharge piping.
The need to modify and resubmit the District's previous response to Generic Letter 87-06, to reflect this PIV testing program is currently under evaluation.
With regard to Local Leak Rate Testing in the reverse (non-accident) direction, it was determined during the 1993 Refueling Outage that such testing for certain large valves does not always yield conservative results. A total of 47 valves were identified which were being tested in the reverse direction.
Of these, eight already have test connectio.ts to allow testing in the accident direction and will be so tested prf r to startup from the 1993 Refueling Outage. Of the remaining 39 valves, 29 have been determined to still produce equivalent or more conservative results when tested in the reverse versus accident direction. The remaining ten valves are being evaluated for schedular exemptions from 10CFR50 Appendix J requirements. Any required exemption requests will be submitted to the NRC for approval prior to plant startup from the 1993 Refueling Outage.
Outstandinz Conditions Fecuirine Evaluation Prior to Startup This issue is being addressed, in part, through the recent establishment of a temporary Corrective Action Program Overview Group (CAPOG).
The Group is comprised of three senior management level personnel and reports directly to the Site Manager. The group was assigned the specific task of reviewing certain past documentation related to the CNS Corrective Action Program (majority of the documentation review covers the past two years) and interviewing appropriate station personnel to determine whether any known conditions exist that require corrective action.
While the review effort is still in progress, issues have been, and are being, identified and resolved.
A portion of this effort was discussed with Mr. Elmo Collins during the NRC inspection activities of the week of May 10, 1993. Any safety significant issues discovered through this review will be brought to your attention following specified reporting procedures, and where necessary, resolved prior to plant startup.
We believe that the efforts of the CAP 0G, and the existing programmatic controls for identifying and correcting problems. as well as the enhanced personnel mindset for identifying and documenting perceived plant concerns, are adequately responding to the concerns expressed by the NRC.
RMV-RM-4 Replacement RMV-RM-4 is the radiation monitoring instru=ent portion of the Drywell Air Sampling System. ~his instrument was scheduled for replacement during the 1993 Refueling Outage due to past reliability problems with the existing instrument and the continuing maintenance required. During procurement activities for the new instrument, a replacement unit was not ' ocated which was qualified to containment design basis pressure requirements. As a result, the most suitable available instrument for this application was procured, and the ensuing design activities were centered around its installation. An incorrect determination was made while addressing primary containment intetrity that the new monitor could be satisfactorily installed without automatic cout'ainment isolation valves and still meet all d'esign requirements.
This shortcoming was identified by both Nuclear Power Group management and Quality Assurance separately, and subsequently by,your Mr. Claude Johnson, as a concern during the present outage.
A design enange has been generated and approved to install automatic primary containment isolation valves on both the sample and return lines for RMV-RM-4 prior to startup from the 1993 Refueling Outage. This effort is in progress. Additionally, since the new instrument has not been proven in service during plant operation. the old instrument will be operated in parallel for a period of time af ter startup, to further assure
James L. Milhocn Pags 4 of 5 NSD930690 availability of the measurement functions.
Previous concerns with the old monitor are being addressed to ensure its reliability until the new monitor is determined to provide acceptable and reliable service during plant operation.
A technical specification change to reflect the addition of the isolation valves will be submitted by August 1993.
Safety Relief Valve Drift For the past several years, Boiling Water Reactors (BURS) with two stage Target Rock Safety / Relief Valves (SRVs) have experienced setpoint drift in excess of the ASME allowable 1% tolerance.
The General Electric Company (GE) has performed comprehensive studies to investigate the root cause of the phenomenon and provide recommendations to prevent or mitigate the problem. GE has recently developed, for field testing, a stellite-platinum pilot disk to alleviate the observed disk-to-seat bonding problem.
We have been closely following developments on this generic industry issue and fully intend to utilize the materials that are being developed to address it.
CNS was one of the first BWR plants to evaluate the safety consequences of SRV setpoint drift. The first CNS SRV setpoint drift evaluation was performed by GE for cycle 11 to address the pertinent safety issues.
The resulting analysis showed that CNS could safely operate with Upper Level (UL) setpoint values of 1210 psig for the SRVs and 1277 psig for the Code Safet.y Valves (SVs).
This analysis assumed that all SRVs and SVs simultaneously drift (to 1210 psig and 1277 psig, respectively).
Note that these values exceed a 3% tolerance.
A Technical Specification change adopting this 3% tolerance as well as the UL values was submitted on January 26, 1990. We are still working towards approval of this Technical Specification change adopting the 3% tolerance value.
Subsequent valve testing showed setpoint drifts in excess of 1% for cycles 13, 14 and 15.
As a result, CE performed cycle specific evaluations and concluded that CNS can operate safely with the observed SRV setpoint drift.
A primary concern of SRV drift is its effect on vessel overpressurization events.
The limiting overpressurization event for CNS is the Main Steam Isolation Valve closure with Flux scram (MSIVF). This event is analyzed each cycle at the UL setpoints previously mentioned.
The cycle 14 MSIVF analyses performed as a result of SRV setpoint drift resulted in only a 1 psi increase in the peak vessel pressure with approximately 65 psi margin remaining available. A cycle 15 SRV setpoint drift analysis has also been perfor=ed where a similar drift was observed. It was concluded for the cycle 15 analysis, that based on the margin available from the cycle 14 analysis, there is significant margin to the ASME overpressure limit. Therefore, the limiting vessel overpressurization event for the upcoming cycle 16 will not present a safety concern for any plausible SRV drifts based on past experience.
The "as found" data for cycle 11 through cycle 15 was recently reviewed for any noticeable failure trends, but no apparent consistency has been identified among the valves.
Based on the observed failures, however, we initiated action last operating cycle to install a stellite-platinum pilot disk in one of the SRVs during the 1993 Outage. Due to the developmental nature of the new design, the stellite-platinum pilot disks were not available in ti=e to be installed during the present refueling outage. The District still plans to install this new disk design in a portion of the Target Rock SRVs during the next refueling outage to see if this approach will resolve this concern. If the new design proves to be effective, the stellite-platinum pilot disks will be installed in all of the Target Rock SRVs at CNS.
In conclusion, settoint drift of two stage Target Rock SRVs has been experienced by the majority o'f BWR plants in the United States.
This potential safety concern has been reviewed by the NRC, the BWR Owners Group, the General Electric Company, and the Nebraska Public Power District.
While no definite cycle dependency trends have been found, the CNS cycle specific analyses performed to
Jamas L. Milhozn Page 5 of 5 NSD930690 date have demonstrated that CNS will not violate any safety limits during the next operating cycle and the observed SRV setpoint drift does not present an unreviewed safety question.
Testinc of Manual Valves Nebraska Public Power District currently believes we are in compliance with the Technical Specifications and ASME Section XI with respect to exercising manual valves. A third party review of the CNS IST program was conducted in 1989 and no manual valves were identified for inclusion in the program at that time. We are, however, aware of related NRC inspection findings at the Monticello plant in November 1992.
We are currently evaluating the applicability of these findings to CNS. In the event that applicable manual valves are identified, they will be added to the IST Program.
In addition, another third party review specific to manual valves will be performed during the next operating cycle.
On a related, but separate, matter discussed with Mr. Phil Harrell during NRC inspection activities of the week of May 10, 1993, we agreed that periodic exercising of manual valves identified in the Emergency Operating Procedures Support Procedures (ESPs) is a good practice.
Numerous valves are currently exercised once per cycle under the CNS Preventive Maintenance (PM) Program. The scope of valves in the ESP program will be reviewed and any valves identified which are not currently in the PM program, or are not otherwise exercised on a once per cycle basis, will be added. This effort will be completed by the end of 1993.
The status of issues provided herein is intended to keep you abreast of our efforts in resolving the ten identified issues. We continue to work toward fully resolving the appropriate issues prior to unit startup, which is currently scheduled for June 22, 1993. A final letter will be transmitted prior to that time informing you of that fact. In the meantime, should you have any additional questions or require additional information, please contact me at your earliest convenience.
Sin ere i
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Horn Nu ear Power Group Manager
/dls cc:
U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC NRC Resident Inspector Cooper Nuclear Station
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