ML20236L242
| ML20236L242 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/06/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20236L244 | List: |
| References | |
| 50-282-98-07, 50-282-98-7, 50-306-98-07, 50-306-98-7, NUDOCS 9807100282 | |
| Download: ML20236L242 (16) | |
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Northern states Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. E.
Welch, MN 55089 July 6,1998 10 CFR Part 2 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING STATION Docket Nos. 50-282 License Nos. DPR-42 50-3CG DPR-60 Response to Notice of Violation (Inspection Report 98007)
Your letter of June 5,1998, which transmitted Inspection Report No. 98007, required responses to Notice of Violations. Our response to the notice is contained in the attachment to this letter.
Your letter of June 5,1998 also expressed concern regarding the quality of engineering and engineering support at Prairie Island. As we discussed at the SALP Public Meeting held at Prairie Island on May 12,1998, NSP has several initiatives underway in the engineering functional area. We discussed improvements in procedure adequacy, engineering reviews, USAR and Improved Standard Technical Specifications, corrective action program, and the quality of NRC submittals. We also discussed engineering staffing levels to support these initiatives without losing focus on maintaining the material condition of the plant. As recognized by the SALP Board, the engineering staff at Prairie Island provides
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strong support to operations, maintenance, and other organizations and continues to //
identify and address technical issues. We entinue to strive to not have a mixed I perception of engineering by the NRC staff. NSP has aiways been and remains committed to a strong engineering organization that provides high quality support to the operation of Prairie Island.
The results of the revised safety evaluation for temporary modification 95T047 to the In l Cooling Water Strainer Backwash Valves was requested in your June 5,1998, letter, ,
The safety evaluation was revised on June 22,1998, but this revision has since I been retracted. The results of a new revision to the safety evaluation will be l transmitted separately at a later time.
9807100282 900706 PDR ADOCK 05000282 G PDR I
E USNRC NORTHERN STATES POWER COMPANY July 6,1998 Pcgs 2 in this response we have made three new NRC commitmer'~ as indicated with bold italics in Attachment 1. If you have any questions concernii J this response, please j contact John Stanton at 612-388-1121 x4083.
- b i Joel P. Sorensen l Plant Manager l Prairie Island Nuclear Generating Plant ,
l l c: (see next page)
]
! c: Regional Administrator-Ill, NRC '
l NRR Project Manager, NRC 1
Senior Resident inspector, NRC i l State of Minnesota, Attn: Kris Sanda
! J. E Silberg
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l l Attachment 1 Response to Notice of Violation (Inspection Report 98007) I Attachment 2 PINGP Calculation No. ENG-ME-370.
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l Attachment 1 Response to Notice of Violation (Inspection Report 98007) l 1
, Violation 1:
Criterion XVI, " Corrective Action," of Appendix B of 10 CFR Part 50 requires in part, that measures be established to assure that conditions adverse to quality are promptly j identified and corrected.
! Contrary to the above, on March 23 and 24,1998, the licensee identified and took l corrective actions for concerns with the potential effects from a feedwater line break on the ability of the main steam isolation valves to close. However, the licensee failed to identify or take corrective actions for the potential effect of flooding on a main steam isolation junction box in the room adjacent to the Unit 2 loop A main steam isolation valves, a condition adverse to quality, until the concern was identified by the NRC on April 22,1998.
This is a Severity Level IV. violation (Supplement 1).
Reason for Violation 1:
L The specific MSIV flooding issue identified on April 22,1998 was originally undocumented and assessed via engineering judgement. Following the NRC identifying their concerns on the specific flooding issue, NSP determined the use of l engineering judgement alone was inadequate. The issue was then formally documented and immediate corrective actions were taken. Subsequent analysis has borne _out the engineering judgement (i.e., the credible postulated design basis breaks or cracks in the area of the room adjacent to the Unit 2 Loop A MSIVs cannot cause submergence of the MSIVjunction box.)
Violation 1 - Corrective Actions Taken and Results Achieved:
The doors to the affected room (room adjacent to the Unit 2 Loop A MSIVs) have been locked open. Upon further review, NSP has determined that these doors serve no particular function (e.g., fire door, special vent zone boundary, etc.). Thus, NSP plans to permanently remove the doors.
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Attachment 1 Pags 2 Based on additional review of the design basis feedwater line break locations NSP plans to retract Supplement 1 of LER 1-98-05. The Prairie Island Nuclear Generating Plant (PINGP) Updated Safety Analysis Report (USAR) Appendix l identifies the criteria to determine design basis break locations in the high energy line systems. The criteria
, specify that two design break locations be chosen between the pipe terminal ends at j the highest stressed points. These locations for Unit 2 main feedwater are found in calculation ENG-ME-319 (as summarized in preliminary calculation ENG-ME-357). The high stress locations in the 16 feedwater line are between the main feedwater regulating valve and its inlet isolation valve, and in the turbine building near the I feedwater heater. Circumferential and longitudinal breaks at these locations do not l support jet flow over the open grating into the doorway and enclosed room containing Unit 2 Train A MSIV junction box.
USAR Appendix I also specifies each branch run be evaluated for design basis line breaks. The design basis line break locations for the 4" feedwater regulating valve bypass line can flow into the doorway and towards the enclosed room containing the Unit 2 Train A MSIV junction box. Calculation ENG-ME-370 (see attached) was prepared to determine the potential flow rate out of this design basis break, and into the room with the junction box. This flow rate was determined to be 3548 gpm. Calculation ENG-ME-364 determined that the holes in the floor of the enclosed room containing the Unit 2 Train A MSIV junction box would allow over 4196 gpm to flow out of the room.
Therefore, with the doors on the enclosed room closed, no equipment would be L affected by a design basis line break.
Smaller design basis cracks were also evaluated. USAR Appendix I defines the design .
basis crack as a single open crack of a size of one-half the pipe diameter in length, and l one-half the pipe wall thickness in width. Design basis cracks result in less flow area I and less flow than the design basis break in the 4"line.
l Thus, NSP has concluded that the flooding issue described and evaluated above did not represent a condition adverse to quality. NSP plans to retract LER 1-98-05 Supplement 1, and update the Prairie Island Nonconformance Report (NCR) 19980570 with the results of these analyses.
Violation 1 - Corrective Steps To Avoid Further Violations:
This violation identified an instance where engineering personnel performing a self-assecsment did not completely document and thoroughly evaluate the full extent of all potentially postulated issues. Based on the experience gained in assessing the issues surrounding this violation PINGP personnel have a clearer understanding of the need to document engineering judgement and when the use of engineering judgement alone may not be sufficient to completely resolve a self-identified issue. PINGP Management will more clearly communicate to engineers when engineering judgement should be documented and when more thorough follow-up evaluations are expected.
Attachment 1 j l Page 3 i 1
k Violation 1 - Date When Full Compliance Will Be Achieved:
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Compliance has been re-established.
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Attachment 1 Pagn 4 o Violation 2:
Criterion V,." Instructions, Procedures, and Drawings," of Appendix B of 10 CFR Part 50
. requires in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.
Contrary to the above, on April 30,1998, Surveillance Procedure SP 1306,"D2 Diesel Generator Functional Test," Revision 3, a procedure for a Technical Specification required test affecting quality, was not appropriate to the circumstances because it did not contain adequate isolation instructions to prevent an inadvertent start attempt of the diesel engine.
l This is a Severity Level IV violation (Supplement 1).
Reason for Violation 2:
. : The surveillance procedure, SP-1306' Rev 3, did not contain any step to verify the l presence of several" electrical"isolations that are required for the execution of SP-1306, and which are established by the preventative maintenance procedure, PM 3001-4-D22 ,
Prior to the execution of SP-1306 on April 30,1998 the procedure was reviewed and a change was made to the Prerequisites and Initial Condition section. Because it is necessary to lift the " mechanical" isolations established in PM 3001-2-D2 to prepare the diesel-generator for the performance of SP-1306, step 6.1 was changed to read "D2 Diesel Generator is logged out per PM 3001-2-D2". The step had previously read "D2 Diesel Generator is tagged out per PM 3001-2-D2." The absence of any step referring to the " electrical" isolations established in PM 3001-4-D2 was not identified during the
. review of this update. ( The prerequisite step in equivalent surveillance, SP-1150', for the other Unit 1 diesel generator, D1, does reference the proper procedure, PM 3001-4-D15 ) .
Violation 2 - Corrective Actions Taken and Results Achieved:
Procedures SP-1306, PM 3001-4-D2, SP-1150 and PM 3001-4-D1 are being revised
' to provide better cmss-referencing between the preventive maintenance procedure and its respective surveillance procedure. These revisions will be
'i SP 1306, "D2 Diesel Generator Test' 2 PM 3001-4-D2, . "D2 Diesel Generator inspection - Electricar 3 PM 3001-2-D2, *D2 Diesel Generator 18 Month Inspection" 4 SP 1150,' 'D1 Diesel Generator Functional Test" 5 PM 3001-4-D1,
- Diesel Generator inspection - Electrical"
r AttachmInt 1 Paga5 completed prior to their next required performance ( Unit 1 diesel generator 18 month outage).
Procedures SP-1306, PM 3001-4-D2, SP-1150 and PM 3001-4-D1 have been quarantined to prevent their use until the required revisions are completed.
Similar procedures for the DS and D6 diesel generators on Unit 2 have been reviewed.
The surveillance procedures for D5 and D6 contain a prerequisite that the diesel generator be isolated per the electrical preventive maintenance procedure.
Violation 2 - Corrective Steps To Avoid Further Violations:
This is an isolated error involving an inadequate procedural reference related to the division of work between two procedures. This was not a situation involving an
, inadequate review of diesel generator logic circuits as proposed in section M3.1.b of the l Prairie Island Inspection Report 98007 and the associated Notice of Violation.
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) Because this is an isolated procedure inadequacy with no generic implications to other l
procedures, no further corrective action is deemed necessary.
, An Error Reduction Task Force investigation, ERTF Report 98-06, of this event is in l progress. When this investigation is complete its recommendations will be reviewed by l management.
Violation 2 - Date When Full Compliance Will Be Achieved:
Full compliance will be achieved by the completion of the next Unit 1 diesel generator l 18 month outage.
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Attichment 1 '
Pagn 6 Violation 3: .
Criterion V, " Instructions, Procedures; and Drawings," of Appendix B of 10 CFR Part 50 requires in part, that activities affecting quality be prescribed by documented
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instructions, procedures, or drawings, of a type appropriate to the circumstances.
i Contrary to the above, on February 9 and 10,1998, Work Orders 9800533 and 9800534 for the installation of auxiliary feedwater system flow element orifice plates, an i ' activity affecting quality, were not appropriate to the circumstances because the work orders did not contain adequate instructions to ensure the proper orientation of the
. orifice plates. - As a result, the orifice plates were installed backwards.
This is a Severity Level IV violation (Supplement 1).
Reason for Violation 3:
A review of work orders 9800533 and 9800534 determined that, while proper orifice plate orientation was indicated on a drawing in each of these work orders, this information was not sufficient to ensure that the orifice plates were properly installed.
Given the importance of the auxiliary feedwater (AF) system it is recognized that the guidance provided was not appropriate to the circumstances.
Violation 3 - Corrective Actions Taken and Results Achieved:
Work orders were issued to reverse these orifice plates during the next Unit 2 Refueling Outage (November,1998).
L ' Prairie Island Nuclear Generating Plant (PINGP) Non-Conformance Report (NCR) 1 - 19980890 was issued to evaluate the consequences of operating Unit 2 with the AFW flow orifices in this degraded condition. This evaluation has been completed with a disposition of "use-as-is". With the orifices installed backwards, indicated flow is estimated to be approximately 20% less th'an actual. This inaccuracy was reviewed versus normal operating procedures and emergency operating procedures (EOPs).
Under normal operating procedures, AF system flow need not be quantified. Under EOPs certain AF system flow setpoints are applied. NCR 19980890 evaluated these setpoints and determined all EOP-related AF system flow setpoints would be satisfied or the error introduced would be insignificant.
NSP evaluated the most limiting case of AF system flow with respect to EOP setpoints and determined there was some margin above the EOP setpoint with inaccuracy in indicated flow. Nevertheless, NSP did evaluate what would happen if inaccurate AF system flow indication did cause an EOP setpoint to be reached (subsequently causing
I Attichment i Paga 7 an early transition into a Functional Restoration (FR) procedure). The effects of an early transition to the FR procedure for loss of heat sink - namely, securing reactor coolant pumps (RCPs) was determined to be an analyzed, safe action. It was further determined that AF system flow would exceed the setpoint once the steam generators were vented and the operators would transition out of the FR procedure.
Consideration was 'given to re-calibrating instruments to account for the reversed orifices. This action was determined to be unnecessary. Also, consideration was given to issuing a temporary memo against the FR procedure to adjust its setpoint to account for the reversed orifices. This action was also determined to be unnecessary, i Violation 3 - Corrective Steps To Avoid Further Violations:
NSP is undertaking many efforts to improve the adequacy of procedures at Prairie Island. One initiative instituted subsequent to this event is the use of a checklist by mechanical maintenance to ensure work orders written contain adequate guidance prior to implementation.
The PINGP Error Reduction Task Force has also initiated an investigation of this particular event. This investigation is being tracked and will be documented per Error Reduction Task 19980935.
Violation 3 - Date When Full Compliance Will Be Achieved:
With respect to the improperly installed flow orifices, compilance will be restored upon completion of the Work Orders to correctly install these orifices. This
- should be completed during the next Unit 2 Refueling Outage (the planned start l
i date of which is Novemberof1998).
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Attachment 1 Page 8 l Violation 4: j j
Criterion XVI, " Corrective Action," of Appendix B of 10 CFR Part 50 requires in part, that {
[ ' measures be established to assure that conditions adverse to quality are promptly Identified and corrected. ]
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i Contrary to the above, inadequate acceptance criteria, conditions adverse to quality, were identified in May 1987 in Surveillance Procedures SP 2101, "21 Motor-Driven Auxiliary Feedwater Pump Once every Refueling Shutdown," Revision 26, and l l SP 2103, "22 Motor-Driven Auxiliary Feedwater Pump Once every Refueling 1 l Shutdown," Revision 29, and were still not corrected when the procedures were used l
! on March 4 and 5,1998.
This is a Severity. Level IV violation (Supplement 1).
l Background Discussion for Violation 4:
Summary ,
l During the course of the August 8,1997 Predecisional Enforcement Conference held to l discuss the apparent violations at Prairie Island identified in NRC Inspection Report 50-
! ' 282(306)/97008(DRS), representatives of Northern States Power Company - Prairie Island stated that all monthly and outage test procedures for all AFW pumps had been l updated with appropriate acceptance criteria. On April 26,1998 it was determined that i 22 TD AFW pump had been run on March 4,1998, using a surveillance procedure in which the acceptance criteria had not been updated. Further investigation revealed f that 21 MD AFW pump had been run on March 5,1998, also using a procedure that had not been updated.
Discussion in May of 1997, in response to an inquiry made by the NRC during it's System Operational Performance inspection, Prairie Island staff concluded that the acceptance criteria contained in its AFW pump testing surveillance were inadequate. It was determined that the acceptance criteria, which had been established using the ASME
- Section XI requirement of +/- 10% from the established pump operating curve, were not sufficient to ensure that these pumps could meet their design flow requirements.
Engineering calculation ENG-ME-321 was performed to determine the required acceptance criteria. It was approved for use on' June 10,1997.
In conjunction with the development of ENG-ME-321, procedure change requests were issued for each of the eight affected surveillance procedures (one monthly surveillance and one refueling surveillance for each of the four AFW pumps). These procedure L._ __ _ __...______._.__.___.____________________________a
Attachm:nt 1 Pagn 9 change requests were issued on June 8,1997 with one exception. A change request had been issued May 5,1997 to change the acceptance criteria in the monthly surveillance for 21 MD AFW pump. This change request was amended per ENG-ME-321.
In addition to the above procedure change requests, on June 8,1997 temporary memos were issued to incorporate the new acceptance criteria into any surveillance procedure that might be performed prior to the completion of the procedure change process. These temporary memos were issued for each of the monthly surveillance and for the Unit i refueling surveillance. However, based upon the fact that the next scheduled refueling for Unit 2 was not scheduled until November of 1998 (16 months later) it was assumed that there was sufficient time to ensure that the change process for the unit 2 refueling surveillance would be ' completed prior to their next use. l On August 8, at the Predecisional Enforcement Conference the statement was made ;
that all monthly and outage test procedures for all AFW pumps had been updated with appropriate acceptance criteria. This statement was based on the first page of a condition report which indicated that temporary memos had been used to update all such test procedures. At that time the conference participants did not realize that a subsequent electronic screen stated that six of the eight affected surveillance were updated with temporary memos and the remaining two surveillance had procedure change requ"sts submitted. None of the personnel assigned to update these surveillance were participants in the predecisional enforcement conference. A discussion did occur between these personnel and conference participants regarding whether a procedure that was amended via a temporary memo could be categorized as updated. This discussion set a trap which resulted in a misunderstanding between the two groups. When the conference participant reviewed the first page of the condition report indicating that temporary memos had been completed, no further verifications were made, thinking that temporary memos had been written for all of the procedures because of the previous conversation regarding the status of a procedure amended via a temporary memo.
On November 17,1997 the decision was made to include the two Unit 2 refueling surveillance in the Prairie Island surveillance procedure upgrade program. As such l- they were to receive additional changes and be submitted to an extensive review process. At that time the acceptance criteria changes were still outstanding so they were incorporated into the more extensive procedure upgrade. This resulted in l
additional delay of the implementation of the acceptance criteria changes.
On January 24,1998 Prairie Island's Unit 2 was forced to shut down to repair a leak in a Part Length Control Rod Drive Mechanism. The scope of this outage expanded to include the removal of all four of Unit 2's Part Length CRDMs. Since Unit 2 was in cold shutdown, testing was scheduled for all ASME Section XI compononts required to be tested at a cold shutdown frequency. This requirement applied to the AFW check valves but not to the pumps themselves. Because the pump and check valve testing
l Attachment 1 l Pags10 l
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l for a given AFW pump and its associated valves is accomplished under one refueling i surveillance (two surveillance per unit), the refueling surveillance for both Unit 2 AFW l pumps were performed to meet Section XI requirements. At this point the pump I acceptance criteria changes for these procedures had not been incorporated into the standing surveillance. As a result, while testing of the pumps themselves was not required, the performance of the pumps was evaluated based upon incorrect acceptance criteria.
This error was identified on April 26,1998 while the procedure change request l facilitating the acceptance criteria and procedure improvement changes to 22 TD AFW I
pump's refueling surveillance was being reviewed for final approval. On April 27,1998 it was determined that 21 MD AFW pump's refueling surveillance had also been performed while Unit 2 was shut down for removal of the Part Length CRDMs and that the incorrect acceptance cr?eria had been used. In both cases this was the first and only use of these surveillance procedures, with the unacceptable acceptance criteria, since the determination of the correct acceptance criteria.
Reason for Violation 4:
As discussed above, the primary reason for the violation was the comfortable feeling that the expected time span before the next scheduled use was sufficient for the procedure changes to be implemented in a timely manner; the process did not ensure that the procedures could not be used until those changes were completed.
Violation 4 - Corrective Actions Taken and Results Achieved:
The two Unit 2 refueling surveillance procedures were quarantined April 29,1998 - to remain in quarantine until the associated procedure changes have been implemented.
The procedure changes have subsequently been incorporated and the revised procedures have been issued; the procedures are still quarantined, but for a different purpose.
A review of all applicable pump performance data was performed and resulted in the determination that both Unit 2 AFW pumps remained operable.
Violation 4 - Corrective Steps To Avoid Further Violations:
The procedure change process has been modified by way of a revision to the form used to implement the process. This revised form requires the individual making the initial technical review to determine whether the change is needed prior to the next use l of the procedure, if this is the case, the procedure change form and process now requires that the procedure immediately be quarantined or have a temporary change made to the procedure.
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Plant management will provide expectations to the plant etaff on the closure of corrective action items to a change process (procedure change request, modification, l etc.). Note that this is an existing commitment, made in correspondence to the NRC, I dated May 22,1998.
Violation 4 - Date When Full Compliance Will Be Achieved:
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Attichment 1 Pags 12 Violation 5:
Criterion XI, " Test Control," of Appendix B of 10 CFR Part 50 requires, in part, that a test program be established to assure,that all testing required to demonstrate that
. structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The I test program shall include, as appropriate, proof tests prior to installation, preoperational tests, and operational tests .during nuclear power plant operation.
Contrary to the above,
- a. on October 3,1995, preoperational testing of a safety-related backup air supply for the air-operated valves in the cooling water system strainers, installed per Temporary Modification 95P047, did not include testing of the check valves which would be required to isolate the backup system from the nonsafety-related instrument air system during a loss of instrument air pressure, and
'b as of May 11,1998, the test program for the safety-related backup air supply did not include operational tests, and operational tests had not been conducted -
since the preoperational testing.
l This is a Severity Level IV violation (Supplement 1).
Reason for Violation 5:
In 1995 the Service Water Operational Performance Inspection self assessment identified an operability concern regarding the performance of the cooling water system
- with a failure of instrument air. A loss of air pressure to the cooling water strainer backwash valves would result in all four valves moving to their full open positions. This would divert some cooling water away from the supported heat loads. The operability i
. concern was whether under this condition sufficient flow would be supplied to each required heat load during a design basis accident. The status of the then available analytical model for the cooling water system did not provide a conclusive answer, so conservative compensatory action was taken. In September 1995 temporary modification 95T047 was initiated to provide a backup source of air pressure to the
-cooling' water backwash strainer valves in the event of a loss of instrument air pressure.
The backup air supply was designed, procured and installed as safety related temporary modification'. The SWAGELOK Check valves received a seat leak test and i
6Although it was envisioned at the time that the temporary modifiuntion would be replaced with a permanent modification this has not been done.
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Attachment 1 Pags 13 a hydrostatic pressure test at the manufacturer. Further preoperational testing of the check valves was not determined to be necessary. The temporary modification was intended as a short-term compensatory measure and because it was expected to be replaced with a permanent modification, periodic testing of the check valves was not identified as needed.
NSP agrees that these valves should be tested periodically, and it has been identified that every three years would be an appropriate interval. Since these valves were installed on September 15,1995, periodic testing of the check valves is not yet due.
Violation 5 - Corrective Actions Taken and Results Achieved:
The survelliance procedure, SP 1151 " Cooling Water System Test", will be revised to include testing of the check valves in the backup air supply system to the cooling water backwash strainer valves. This revised surveillance will be performed to test the air check valves by September 15,1998.
Violation 5 - Corrective Steps To Avoid Further Violations:
While this violation appears to be an isolated event, currently open temporary modifications will be reviewed to confirm this judgement. If evidence is uncovered that demonstrates that this is not an isolated event, the temporary modification process will be evaluated to identify those actions needed to cddress this issue.
Violation 5 - Date When Full Compliance Will Be Achieved:
Full compliance will be achieved by September 15,1998.
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Attachment 2 PINGP Calculation No. ENG-ME-370 l
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