ML20046A369
| ML20046A369 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/22/1993 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | Milhoan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20046A364 | List: |
| References | |
| NSD930779, NUDOCS 9307280008 | |
| Download: ML20046A369 (6) | |
Text
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GENERAL OFFICE
u NSD930779 June 22, 1993 Mr. James L. Milhoan Regional Administrator U.S. Nuclear Regulatory Commission Region I'J 611 Ryan Plaza Drive. Suite 1000 Arlington. I 76011
Subject:
Status of 1ssues Related to Unit Startup
- ooper Nuclear Station NRC Docket No.30-298. License No. OPR-e6
Reference:
S. R. Horn to J. *. Milhoan (NSD930690), dated June 7,1993, Status of Issues Related to Unit Startup
Dear Mr. Milhoan:
The referenced letter provided the status of ten issues that had been identified by the NRC during the 1993 refueling outage at Cooper Nuclear Station iCNS) as requiring resolution.
This letter provides the current status of cur efforts with regard :o resolving these ten issues and documents completion of those actions reouired to support plant startup.
Surveillance Testint of the secendarv Containment A Seconcarv :ontain=ent Action Plan was developec to identify, review, and i=plement certain programmatie :hanges necessarv to =aintain preservation of the Secondarv Centainment capability durinz the operating cycle. All items of this action plan nave recently been completen in support of plant startup from the 1993 refueling outage.
The District vill also revise ene CNS Technical Specifications to incor: orate new requirements towarc providing a better assurance tha: the secondary c:ntainment integrity is =aintained througnout the plant opera:Ing cycle.
These Technical Specification changes will be suomitted for NF.C review in August 1993.
Diesel Generator Fuel Oil Psrt :ulate Levels In the reference. we stated : hat CNS Technical 5:ecification and procedure changes were under review to address Diesel Generator Fuel Oil Particulate Levels and sampling of storage and day tanks for water. Se revisions to the acplicable CNS procedure will be SORC approveo on June 21. 1993, and will establish administrative limits and controls to require action to ensure the Fuel Oil Particulate Level remains below 10mg/1. and that the appropriate water sampling is performed.
9307280008 930721 PDR ADDCK 05000298 G
PDR;
i NSD930779 Page 2 of 6 A change to the CNS Technical Specifications that will incorporate new limiting conditions for operatior and surveillance requirements for Diesel Generator Fuel Oil that is consistent with the new BWR standard Technical Specifications has been approved by the oversight committees and will be submitted for NRC review l
by June 23, 1993. These proposed Technical Specification changes will include new requirements for an acceptable Fuel Oil Particulate Level (10mg/1) and will also require the sampling for and the removal of detected water in the fuel oil storage tanks and day tanks.
In addition, modifications have been performed on the Diesel Generator Fuel Oil day tanks to allow sampling and removal of any water from the very bottom of the tank in order to gain the full benefit of sampling for and removal of water.
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All actions and programmatic changes required to assure diesel generator j
operability in support of plant startup will be completed by June 23. 1993. The District continues to work with the industry fuel expert discussed in the reference, to further aasure continued maintenance of the overall quality of the Diesel Generator Fuel 011.
~'.is effort is expected to be completed following plant startup.
H,/03 Analyzers The primary problems with the H /02 Analyzer system have been associated with the i
2 condensation and collection of water in the sample lines and with sample pump reliability.
During the 1993 refueling outage the pump internals have been modified to enhance their reliability and the previously installed particulate filters on the suction of these pumps have been removed.
Extensive post-modification testing has been conducted on the Hz/03 Analyzer system to verify that the previous problems with water condensation in the instrument sample lines have been resolved. This has included strenuous testing and where necessary correction of the slope and configuration of the instrument lines to ensure no low points exist that would allow water pockets to form. Heat tracing on system piping and tubing has also been upgraded and satisfactorily tested. To further ensure the effectivene.es of this system upgrade. an enhanced instrument surveillance program will be implemented during the initial stages of power operation.
~'his program will verify the previous water problems have been satisfactorily resolved.
Satisfactory testing to full accident pressure of the H /02 Analyzer system 2
tubing within the instrument :abinets has been completed. All portions of the H /02 Analyzer piping that would be subjected to primary containment accident 2
pressure have now been tested with acceptable results achieved. The additional leakage measured for the piping within the H /02 instrument cabinets has been 2
factored into our previous :ntegrated Leak Rate Test results and the total leakage rate remains well below allowable criteria.
A review of other primary containment cabinet sampling system configurations is underway to verify that no similar situations of incomplete testing exists.
Concerns noted during this review will be corrected prior to plant startup from the 1993 refueling outage.
In addition, a detailed review of all containment penetrations and their associated Appendix J testing requirements will be performed during the next operational cycle and will be completed prior to startup from the next scheduled refueling outage.
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NSD930779 Paga 3 of 6 Local Leak Rate Testine Criteria for One Inch and Less Containment Penetrations As stated in the reference, Appendix J Type B and Type C testing is being conducted for containment penetrations to the CNS licensed primary containment configuration and there is no generic exemption for Appendix J testing of Primary Containment Penetrations one inch and less in diameter.
Service Water System Integrity As stated in the reference, various aspects of Service Water System integrity were reviewed by Messrs. J. Medoff and F. Grubelich of the NRC-NRR during the week of May 10, 1993, We understand that the only remaining open issue 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 safety significant issues considering the number of inspections being performed, the 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 has been identified as an apparent violation in NRC :nspection Report 93-17.
We therefore will address this item in future correspondence, but, unless otherwise notified, it is not being considered an issue requiring resolution prior to plant startup.
Pressure Isolation Valve and Local Leak Rate Testint Pressure Isolation Valve (PIV) Leak Rate Testing has been conducted on the appropriate valves in the Residual Hea; Removal, High Pressure Coolant Injection, and Reactor Core Isolation Cooling Systems. Testing has been performed per ~ the guidance of ASME Section XI, IMV-3420 and all final measured leakage rates are within the specified acceptance criteria.
Final tasting has yet to be done on valve RHR-M025A which is presently undergoing repair vork. Satisfactory PIV Leak Rate Testing of RHR-M025A will be completed prior :: plant startup from the 1993 refueling outage. Due to the existing plant configuration and time remaining to startup, testing of the Core Spray injection check valves will take place during the Primary System Leakage Test (vessel hydro) currently scheduled for June 29, 1993. As committed to in the reference. plant mocifications vill be performed during the next scheduled refueling outage to allow P:V testing of the Core Spray inboard motor operated injection valves.
Based on the acceptable PIV Leak Rate Test results to date, which include those for the repaired valves RHR-M017 and RHR-M018.
ne tested valves have been verified to be capable of performing their pressure isolation function for the upcoming operating cycle. Acceptable testing of the Core Spray injection check valves, prior to plant startup, will verify the ability of one valve in each Core Spray injection path to be capable of pressure isolation and when combined with the Appendix J testing of the inboard motor operated injection valve. provides assurance that the integrity of the high/ low pressure interface exists in these lines for the upcoming operating cycle.
Concerning 10CFR50 Appendix J testing of certain primary containment isolation valves in the reverse (opposite to accident) direction, the District submitted a request for schedular exemption to Appendix J Testing for ten valves on June 7, 1993. NRC approval of this exemption request. is required prior to allowing plant startup to commence. The exemption request, when granted, will allow plant
NSD930779 Page 4 of 6 operation for one operating cycle, During this period, evaluations will be performed to demonstrate the adequacy of testing the valves
'.a the reverse l
direction, or modifications will be implemented to allow leakage testing to be performed in the accident direction. As stated in the reference, 29 other valves at CNS continue to be tested in the reverse direction because analysis has determined that such testing will produce equivalent or more conservative results j
than testing in the accident direction.
j Dutstanding Condition Reauirine Evaluation Prior to Startup The establishment and duties of a temporary Corrective Action Program Overview Group (CAPOG) as well as other efforts to address this issue were discussed in the reference. Through interviews with appropriate station personnel and review of certain past documentation related to our Corrective Action Program, various issues or questions have been identified for review and resolution. A number of f
l items were determined to require resolution to assure component or system operability prior to certain key outage milestones (e.g.,
Vessel Pressure Testing, establishment of Primary Containment Integrity, Plant Startup). These issues are tracked to assure the status of their completion prior to the i
performance of an outage milestone. 3e status of this effort was discussed with Mr. Jim Gagliardo (NRC-Region IV) the week of June 14, 1993, during his visit to CNS, and he was given a copy of the issues identified to date and their status.
The identification, evaluation, and resolution of the issues stemming from the above efforts, as well as the improved performance by District personnel in identifying and documenting deficiencies and concerns provide assurance that this issue is being adequately addressed.
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?J'V-RM-4 Renlacement i
A design change to install automatic primary containment isolation valves on both j
the sample and return lines for the primary containment atmosphere monitor (RMV-RM-4) has been completed and post-modification testing to verify the isolation function performance is in progress.
Se testing of the isolation function will be satisfactorily comp'eted prior to startup from the 1993 refueling outage. As stated in the reference. the old containment atmospheric monitor will be operated in parallel with the new monitor for a period of time af ter plant startup, in order to verify the reliability of the new monitor during actual operating conditions.
3e old monitor will continue to be the primary monitor until the reliability of the new monitor is verified.
All previous concerns associated with the old monitor have been addressed to ensure its operability and reliability curing plant operation.
Safety Relief Valve Drift A General Electric evaluation has been perfor=ed using approved analysis methodologies discussed in NEDO-24011 P-A-US, to calculate the peak vessel pressure that would result for certain two stage Target Rock Safety Relief Valve (SRV) setpoint drift assumed to occur during the upcoming Cycle 16 operation.
SRV setpoint drift data from the previous five operating cycles have been reviewed to ensure that the evaluated SRV setpoint drift bounds the setpoint drift, which has been experienced to date.
,, =....
NSD930779 Page 5 of 6 i
The evaluation concluded that sufficient margin existed to the Reactor Coolan:
System Integrity Safety Limit even with postulated amounts of SRV sotpoint drift j
higher than have been previously experienced. This evaluation will be factored i
into a component operability evaluation that wtil be performed to ascertain the SRV's operability for the upcoming Cycle lo operation.
This cperability evaluation will be completed and reviewed by the Station Operations Review Committee prior co startup frcm the 1993 refueling outage.
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As stated in the reference, the District plans to install f n some of the valves I
l during the next scheduled refueling outage, a new design of pilot disk assembly.
The upgraded pilot disk assembly was recently developed by General Electric and 4
the Boiling Water Reactors Owners' Group (BWROG) to address the SRV setpoint
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drift issue. The District will follow the BWROG reccamendations regarding the i
number of valves that should receive the upgraded pilot disk for the initial cycle.
The District will also continue to follow industry progress for other possible fixes for serpoint drift outside of those being developed by the BWROG.
l Based on our evaluation of past SRV setpoint drift data at CNS and the evaluation j
performed by General Electric, SRV setpoint drif t does not present an unreviewed safety question at CNS.
l Testinz of Manual Valves Since the submittal of the reference, the District has evaluated NRC inspection i
findings at the Monticello plant related to the testing of manual vcives. Based on this evaluation no additional valves were identified that need to be incorporate-nto the CNS In-Service Testing (IST) program.
However, another third party u iew specific to manual valves will be performed during the next a
operating cycle.
As discussed in the reference, we agreed that periodic exercising of manucl l
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 eterently in the j
PM program, or are not otherwise exercised on a once per cycle basis, wil) be j
added.
3is effort will be completed by the end of 1993.
j As discussed above, with the exceptions noted, all actions committed in the reference to be performed prior to plant startup from the 1993 refueling outage j
have been completed.
The remaining committed actions yet to be completed will be performed and your staff vill be formally advised of their completion prior te plant startup. Should you have any additional questions or require additional information, please contact ce at your earliest convenience, Sin ' rely, h
mm G /11 Horn Nuc ear Power Group Manager
/ dis
NSD930779 Page 6 of 6 cc:
U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC NRC Resident Inspector Cooper Nuclear Station