IR 05000282/1997015
| ML20217C489 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/19/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20217C451 | List: |
| References | |
| 50-282-97-15, 50-306-97-15, 72-0010-97-15, 72-10-97-15, NUDOCS 9710010417 | |
| Download: ML20217C489 (22) | |
Text
. .
l U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket Nos: 50 202,50 306,72 10 License Nos: DPR 42, DPR 60, SNM 2506 Report No: 50 282/97015(DRP);-50 306/97015(DRP);
7210/97015(DRP)
Licensee: Northern States Power Company Facility: Prairie Island Nuclear Generating Plant Location: 1717 Wakonade Drive East Welch, MN b5089 Dates: June 26 August 5,1997 Inspectors: S, Ray, Senior Residont inspector R. Bywater, Resident inspector M. Farber, Reactor inspector
_ _
P. Krohn, Resident inspector Approved by: J. W. McCormick-Barger, Chief ~
Reactor Projects Branch 7 pOl 9710010417 970919 PDR ADOCK 05000282 G PDR
.
- - -m
_
, .
j i
l EXECUTIVE SUMMARY-
<
Prairie Island Nuclear Generating Plant, Units 1 & 2
, NRC Inspection Report 50 282/97015(DRP); 50 306/97015(DRP); 7210/97015(DRP)
,. This inspection was performed by the resident inspectors and included aspects of licensee 1 operations, maintenance, engineering, and plant support. The inspection also included
! followup inspection by a regional specialist of findings from the maintenance rule baseline i inspection.
,
Ooerations i
! e The conduct of operations was generally good during this inspection period j (Section 01,1),
j e The licensee identified and adequately addressed a fuel handling error in the spent j fuel pool Verification requirements added to procedures as corrective actions from previous fuel handling errors were not rigorously followed by operators (Section 01.2).
e The inspectors identified several deficiencies in attachments to the Emergency
Operating Procedures when compared to the licensee's writers guide (Section 01.3).
l Maintenance
,
!
- The conduct of maintenance and surveillance activities was generally good during j this inspection period (Section M1.1).
! e The inspectors Ncntified a procedure adherence violation when operators did not
- declare the Unit 2 control rod deviation monitor inoperable during a surveillance test (Section M1.1).
i
- * During an inservice testing program audit, the licensee's quality services staff
_
identified that the licensee's staf f did not evaluate the condition of a residual heat
- removal pump when vibration levels during a surveillance test were recorded at the
alert value. The procedure did not contain the appropriate acceptance criteria (Section M7.1),
o A followup inspection of the NRC's previously performed maintenance rule baseline inspection was conducted. All of the findings from this previous inspection were satisfactorily addressed and no new weaknesses were identified (Section M8),
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--_
.
Ennincorina e The inspectors identified concerns with habitability of the control room following actuation of a c: on dioxide fire suppression system. Also, the inspectors
!
identified that the presence of carbon dioxide in the control room, including the l impact of a rupture of the carbon dioxide storage tank, was not included within the scope of the licensee's evaluation of control roorn habitability in response to Three Mile Island Action Item Ill.D.3.4 and that other commitments in response to tnis Action item regarding preparations for an accidental release of toxic gas may have been inappropriately deleted (Section E2.2).
Plant Suonort e No discrepancies were noted during routino observations (Section IV).
_
.. - .
Reoort Details Summarv of Plant Status Unit 1 operated at or near full power for the entire inspection period. Unit 2 operated at or I near full power for the entire inspection period except for a brief power reduction to about 40 percent power on July 26 27 to conduct turbine valve testing and condenser cleaning, l
'
l. OperatioDE
)
i 01 Conduct of Operations
{
01.1 General Comments Insoection Scope (71707)
Using lnspection Procedure 71707, the inspectors conducted frequent reviews of
_
plant operations. These reviews included observations of control room evolutions, shif t turnovers, and logkeeping, as well as evaluation of operability decision Section 13, " Plant Operations," of the Updated Safety Analysis Report was i- reviewed as part of the inspection, Observations and Findinas The inspectors observed proper control room marining, close attention to control panels, generally good use of communication protocols and procedures, and detailed shift briefs in which all members of the crew contributed. No significant problems with normal plant operations were noted, Conclusions All normal operations of the plant during the inspection period were conducted wel .2 Fuel Handlina Error in Soent Fuel Pool Insoection Scooe (92901)
On July 7; 1997, the inspectors were informed by licensee management that operators had installed a spent fuel assembly into the wrong location in the spent fuel pool while conducting preparatory moves to support Westinghouse _ fuel inspections. The inspectors reviewed the circumstances surrounding and corrective actions for the even b, Observations and Findinas On July 7, an operator was conducting a series of spent fuel movements in the spent fuel pool to position assemblies for an upcoming specialinspection projec _ . _ _ . . _ . _ _ _ . __ _ _ _ _ _ _ _ _ _ __ . _ _ _ . . _ _
. -
,
!
i The operator was performing fuel moves under the direction of a shHt supervisor and was under occasional observation by a quality services inspector and a nuclear
, enginee Af ter numerous moves were successfully perforrned, the supervisor misread the fuel transfer log and told the operator to place tho assembly in the wrong locatio .
The operator did not verify that the correct location had been ordored by 1 independently reviewing the lo Section Work Instruction SWI O 41, " Duties and Responsibilities of Fuel Handling -
Personnel," Revision 2, Step 6.2.C, required that prior to inserting an element in a location, the spent fuel pool operator observe the fuel transfer log. The operator had been independently reviewing the log for each of the previous fuel moves, but discontinued the practice as the day went on. The supervisor, quality services inspector, and nuclear engineer all had opportunities to correct the failure of the operator to independently verify the log pnor to the event, but all failed to take actio The licensee's corrective actions for the event included the following:
e temporarily stopping all fuel moves; '
e counseling of the individuals involved; e issuance of Nonconformance Report 2010799; >
e issuance by the plant manager of a Site Notice describing the event and l emphasizing management expectations for fuel movements; and e requiring a nuclear engineer to provide a third independent concurrence for l
all fuel move L The plant manager authorized fuel movements to restart on July 9 and the inspectors observed that the corrective actions were effectively implemented for j the remaining fuel moves.
! The event, caused by a failure to follow a requirement in SWI O 41, was of low
'
safety significance and was identified and corrected by the licensee. This non-repetitive, licensee identified and corrected violation is being treated as a Non Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement Polic (50-282/97015 01(DRP): 50 306/97015-01(DRP))
, c. Conclusions
!
Licensee management treated the event seriously with concerns expressed to the inspectors by both the Vice President, Nuclear Generation and President, Northern States Power Generation. The licensee attributed the event to complacency which had built in af ter a long series of fuel movements that day. Prncedures for moving fuel, which had been recently upgraded because of other fuel handing errors in the i past, were adequate to prevent an error of this nature if they had been properly
! followed. Y l
. , .-. - - _ _ - . - _ _ - , .- . . - - - - - - . _ . . - . . ~ . . - - -
.
.
01.3 Unit 2 Power Reduction for Turbine Valve Testina and Condenser Cleanina insoection Scoos (71707)
The inspectors observed portions of the Unit 2 power reduction on July 26 27,1997, to perform turbine valve testing and condenser cleaning, Observations and Findinas The inspectors observed the shift manager and Unit 2 shift supervisor conduct a pre evolution briefing for a planned power reduction to approximately 40 percent power. The operators performed the power reduction in accordance with i procedure 2C1.4, " Unit 2 Power Operation," Revision 14. The briefing was l adequate and the operators performed the power reduction evolution wel l During the power reduction, alarms weis received when three control rod position indications deviated from bank demand position. In response, the operators performed Surveillance Procedure SP 2319, " Rod Position Verification," Revision 5, to verify that none of the control rods were actually misaligned. The inspectors'
observations and concerns with this activity are discussed in Section M Conclusions Operators performed wellin conducting the power reduction; however, problems with the performance of the rod position verification test were identifie Operations Procedures and Documentation 03.1 Emeraency Ooeratina Procedure (EOP) Attachments Insoection Scoce (92901)
The inspectors reviewed the attachments to the licensee's EOPs and noted various
' discrepancies when evaluated against Procedure H14.7, " Emergency Operating Procedure Writers Guide," Revision 1, The attachments reviewed were:
- Attachment A " Natural Circulation Conditions" e Attachment B, " Main Steamline Isolation" e Attachment C, " Plant System Alignments" e Attachment D, " Post LOCA [ Loss of Coolant Accident] Alignment of 12(22)
RHR 1 Residual Heat Removal] for Shutdown Cooling" e Attachment E, "SG (Steam Generator] Wide Range Level Adverse Conditions"
- Attachment F, " Reactor Vessel Vent Time Calculation" e Attachment G, " Unit 1(2) Containment isolation Valve Locations" e Attachment I, " Containment Closure Procedure" e Attachment J, " Isolate Unit 1(2) Moisture Separator Reheaters"
- Attachment K, " Unit 1(2) Alignment for Switchover to Recirculation"
.
.
e Table EO 1, " Auto Actions Guide" l e Figure ES03A 1, " Condensate Storage Requirements"
!
e Figure ES03A 2, " Condensate Storage Requirements" l e Figure ES03A 3, " Pressure Temperature Limits for Natural Circulation l
Cooldown With CRDM (Control Rod Drive Mechanism) Fans"
- Figure ES03B 1, " Condensate Storage Requirements" e Figure ES03B 2, " Condensate Storage Requirements" e Figure ES03B 3, " Pressure / Temperature Limits for Natural Circulation Cooldown Without CRDM Fans" e Figure ESO41. " Pressure / Temperature Limits for Natural Circulation Cooldown With Steam Void in Vessel" e Figure ECA11 1, " Minimum injection Flow Rate Versus Time After Trip" e Figure ECA31 1, " Containment Sump Level vs. RWST [ Refueling Water Storage Tank) Level" e Figure FRP1 1, " Post Soak Cooldown Limit" e Figure FRP21, " Technical Specification P/T [ Pressure /Temperaturel Limits Cooldown Curve" e Figure FRP2-2, "Cooldown Limit" e Figure FR 31, " Technical Specification P/T Limits Cooldown Curve" e Figure Fr,g3 2, " Hydrogen Flow Rate vs. RCS [ Reactor Coolant System]
Pres",ure" b. Observations and Findinos Procedure H14.7, Section 4.4.B.3. " Attachments," required that the EOP Introduction Page include "A listing of any attachments included in the procedure IEOP)," and stated that the attachments are intended to provide supplementalinformation to the user during the use of that specific procedure and may be in the form of figures, tables, charts, curves, etc."
The inspectors noted the following discrepancies:
e None of the tables or figures in the EOPs were listed in the introduction Page of any of the EOPs to which the tables and figures were attache However, Attachments A through K were listed in the introduction Page, o Several of the attachments did not provide " supplemental information" but in fact contained procedure steps required to be performed to mitigate the emergencies. The procedures were either to be performed by out-plant operators or a combination of control room and out-plant operator Attachments B, C, D, I, J, and K were actually procedure e The attachments that were actually procedures, as listed above, generally were not written in the same format, content, and style required by H1 for the EOPs they were attached to, e- Attachment K, Step 3, had multiple unrelated actions within the same step contrary to H14.7, Section 4.7.B.1, which stated, "Each step should deal
. .
i with only one idea." Step 3 contained about 11 separate actions for the auxiliary building operator to take in several ditferent locations and involved several different systems for several different purposes. There were no placekeeping alds to help the operator verify that no actions were missed, o The substeps in Attachment K, Step 3, contained no indication of whether they needed to be performed in order, contrary to H14.7, Section 4.7.B.10, which stated, " Alphanumeric designation of steps indicates that they should be performed in the order shown, unless otherwise stated. Listing or designating with a bullet (e) indicates that there is no preference to the order of performance." The Step 3 substeps had neither alphanumeric designations nor bullets, o Attachment E was actually a figure b:+ was not numbered similar to the other figures in the EOP e The same attachments, figures, and tables were often attached to several ;
different EOPs. However they had the revision number of the EOP they were attached to. Thus the same attachment, figure, or table might have several differont revision numbers and there was no way to verify that they '
were actually identical. The most significant example was Attachment E which was atta thed to 54 different EOPs and had 13 different revision number Conclusions Following the writers guide for EOP development is not a regulatory requirement but is an expectation of plant management. Failure to follow the writers guide indicates that a weak review of the procedures was accomplished in the EOP development and approval process. The inspectors were especially concerned with Attachment K which relates to manual actions to place the unit in containment sump recirculation. This is one of the most safety significant and time critical actions for reducing core damage frequency according to the licensee's Individual Plant Examination Report submitted in response to NRC Generic Letter 88 20
" Individual Plant Examination for Severe Accident Vulnerabilities -
08 Miscellaneous Operations issues (92700)
08.1 (Closed) Licensee Event Reoort (LER) 282/97009: " Unavoidable Momentary Non-compliances with Technical Specification Requirements Which Do Not Provide a Time Interval to Establish Different Plant Equipment' Configurations Upon Moving into Different Plant Conditions." This LER reports a condition, first identified by the inspectors, which was a result of inadequately _ worded Technical Specifications, whereby the licensee could not literally comply with the Technical Specification This condition had no safety significance and the licensee intended to correct the Technical Specification in a future amendment reques .
.
08.2 Withdrawal of Aonlication for Offsite indeoendent Soent Fuel Storaae installation
.
On July 22,1997, the licensee issued a letter to the NRC withdrawing its i
!
August 7,1996, application for a license to construct and operate an Independent Spent Fuel Storage Installation (ISFSI) at an away from reactor site located in Florence Township, Goodhue County, under Docket No. 7218. The action was taken after the Minnesota Supreme Court declined to hear an appeal of a previous l ruling by tho Minnesota Environmental Quality Board to deny a Certificate of Site i
Comparability for the facilit . Maintenance i
M1 Conduct of Maintenance M1.1 General Comments Insoection Scone (61726,62707. 92902)
The inspectors observed all or portions of the following maintenance and surveillance activities, included in the inspection was a review of the surveillance procedures (SPs) or work orders (WOs) listed as well as the appropriate Updated Safety Analysis Report (USAR) sections regarding the activities. The inspectors verified that the surveillance procedures observed met the requirements of the Technical Specifications except where note e SP 1115 Spent Fuel Pool Special Ventilation System Test (Revision 15)
e SP 1128 Monthly Backwash of Emergency Bay intake Pipe (Revision 0)
e SP 1249 Caustic Addition Standpipe Level Functional Test (Revision 7)
e SP 2088 Safety injection Pumps Test (Revision 38)
e SP 2095 Bus 26 Load Sequencer Test (Revision 10)
e SP 2102 22 Turbine Driven Auxiliary Feedwater Pump Test (Revision 51)
o SP 2319 Rod Position Verification (Revision 5)
e WO 9706943 Spent Fuel Moves e WO 9707591 Investigate D6 Fuel Oil Storage Tank Leak Alarm o WO 9707654 Remove Instrument Tubes from Fuel Assemblies for Analysis e WO 9707785 Setpoint Change on 11 Caustic Standpipe Low Level Alarm o WO 9708069 investigate 121 Air Compressor Remote Alarm
- Observations and Findinas For all of the work observed, procedures were properly used and followed, except as noted belo . . . ..
. .. ..
. _ _ --
. .
.
.
.. . ..
.
.
e Work Order 9707785 and SP 1249 were completed to correct an operator-identified design problem. The caustic standpipe low level alarm was set such that it would not alarm untillevel was slightly below the sightglas Thus, when the alarm actuated, operators were unable to verify the actual level or determine the rate of level decrease locally. The WO raised the low level alarm setpoint to be within the sightglass range. The change was in the conservativo directio * Work Order 9706943 and WO 9707654 were part of a project in cooperation with Westinghouso to conduct detailed examinations of certain spent fuel elements to gather data to help determine the cause of control rods failing to fully insert at some plants as discussed in NRC Bulletin 96 01,
" Control Rod Insertion Problems." A fuel handling error which occurred during the project is discussed in Section 01.2 of this repor * Regarding SP 2319, " Rod Position Verification," Revision 5, on July 26, 1997, the inspectors identified problems with the performance of the procedur The procedure was implemented during a Unit 2 power reduction when control rod misalignment alarms were received for dif ferences between individual rod position indicators (IRPIs) and the rod group demand positions for three control rods in the D control bank. The purpose of the procedure was to determine if the misalignment was actual or merely the result of a problem with the IRPIs. Rod deviation alarms were a common occurrence during power changes due, in part, to the non linear response of the IRPI magnetic coil stacks as a function of temperature. This had been identified earlier as an operator workaroun As discussed in NRC Inspection Report 50-282/97011(DRP);
50 306/97011(DRP), Nonconformance Report (NCR) 2010776 was issued to address the condition where the surveillance test determined that the control rod was not actually misaligned but that the IRPl was indicating inaccurately. In that case, the IRPI could be considered operable if it correctly indicated rod motion, but the rod deviation monitor must be considered inoperable because it would not be capable of detecting certain subsequent actual rod position deviations for the affected control rod. On July 25,1997, the licensee issued Temporary Memorandum (TM)
No. TMA 1997 0122, to procedure SP 2319, Revision 5, which revised a flow chart attached to SP 2319 to address operability of the rod deviation monito When the first rod deviation alarm was received on July 26 at 10:13 pm, the shif t supervisor requested that a reactor operator perform SP 2319. There was no pre-job briefing prior to implementation of the procedure and the reactor operator did not identify to the shift supor.;sor that a new TM had been written for this procedure. The TM stated, "Use new flow chart to determine IRPI operability as a substitute for existing flow chart. New flow
.
.
chart is a result of NCR 20' )776." Additionally, the procedure stated, "If this verification proves thi . the rods are not misaligned, then the rod shall not be considered misa' 'ed per TS (Technical Specification) 3.10.E.2 and the IRPI operability r' ., determined per the attached flow chart."
The inspectors observe. that the reactor operator completed procedure steps in SP 2319 that c ifirmed that none of the three rods were actually misaligned. However, the reactor operator did not refer to the flow chart to evaluate operability of the IRPIs or the rod deviation monitor. After the reactor operator completed the procedure, the shif t supervisor and shif t manager reviewed it. After completion of their reviews, the procedure was logged as satisf actorily completed with no rod misalignment at 10:50 p At that time, the rod deviation alarm had not cleare The inspectors asked the shift supervisor why the rod deviation monitor was not declared inoperable per the requirements of the SP 2319 flow char After further review and discussions with the shift manager and nuclear engineer, the shift supervisor declared the rod deviation monitor inoperable and ordered that the requirements of TS 3,10.1 be implemented for logging IRPl Criterion V to Appendix B of 10 CFR Part 50 requires that procedures affecting quality be of a type appropriate to the circumstances and be accomplished in accordance with these procedures. On July 26,1997, the rod deviation monitor was not declared inoperable as required by SP 231 This is a violation. (50 306/97015-02(DRP))
During an interview with the shift supervisor following this event, he stated that he had considered operability of the rod deviation monitor during performance of SP 2319 and he believed that it was operating properly and that he did not need to take the action identified in the procedure. The shift supervisor had not beel aware, however, of the recently identified concerns with IRPl and rod deviation monitor operability in the Nonconformance Report which were the basis for the Temporary Memorandum. Also, the shift supervisor stated that operators were not trained on the use of the flowchart and that its use was not identified as a procedure step in the body of the procedur c. Conclusions The majority of inspector-observed maintenance and surveillance activities were conducted well with good communications Job planning, work practices, and coordination between departments. However, problems were identified during the performance of SP 2319. The procedure was inadequately reviewed by operators prior to its use. Also, IRPI and rod deviation monitor operability had not been properly assessed by licensee staff after rod alignment was verified because instructions contained in the SP 2319 flowchart were not implemente _ _ _ _ _ _. _ _ _ . _ _ . _ _ _ _ _ .__ _ _ _ _ . _ _ _ _ _ _
. .
!
!
- M7 Quality Assurance in Maintenance Activities i
M7.1 Failure to Evaluate RHP Pumo Hioh Vibration Level i insoection Scone (92901)
!
During June 1997, while Unit 1 was in cold shutdown, a licensee quality services
'
audit of the ASME (American Society of Mechanical Engineers)Section XI Inservice Testing Program identified some inconsistencies regarding acceptance criteria in Revision 7 of Surveillance Procedure SP 10928, " Safety injection Check Valve test (Head Off) Part b: RWST [ Refueling Water Storage Tankl to RHR 1 Residual Heat l
.
Rernovall Flow Path Verification." The inspectors reviewed the circumstances 4 surrounding this finding, i
j- Observations and Findinas The purpose of SP 1092B was to obtain new ASME Section XI baseline flow and
vibration levels for the No.11 and No.12 RHR pumps. The procedure stated that
- acceptance criteria did not apply and that the test results would be evaluated by the system engineer. However, even for baselining purposes, the vibration acceptance criteria should have been specified in the surveillance procedure (per ASME requirements). During performance of the January 13,1996, test, a vibration levelin the alert range for the No.12 RHR pump was recorded (the procedure did not specify what the alert range was) and was not subsequently evaluated in the required 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> interval per Section XI requirement The licensee believed that the pump never experienced high vibration levels. Test data taken before and after the January 13,1996, test showed normal vibration levels. The licensee believed that the vibration instrument was positioned to read the alert setpoint instead of the actual vibration level during performance of the test. This resulted in the alert setpoint being recorded instead of the actual vibrations of the operating pump. The licensee concluded that the No.12 RHR pump vibration levels remained normal throughout the period in question and that the pump was therefore operabl The licensee's corrective actions for the event included
e engineering management emphasizing adherence to the system turnover process described in administrative work instructions to system engineers; e ASME Inservice Testing requirements training for mechanical system engineers; and
e requesting operations training staff to review with operators the proper method of taking vibration data from panel Technical Specification 4.2 requires that pumps and valves be tested in accordance with the requirements of ASME Section XI (1989 edition)/ ANSI (American National Standards Institute) Operations and Maintenance Standard, Part 6 (1988 edition).
The requirement to analyze the high vibrations levels cecorded during the
- _ _ _ - - _ _-
surveillance test within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> was not met. This was a violation of Technical Specification 4.2. The violation was not considered safety significant because the pump was subsequently determined to be operable. This non repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Polic (50 282/97015 03(DRP)) Conclusions The finding by the licensee's quality services organization regarding the failure of the licensee to include acceptance criteria in SP 1092 B was considered an excellent finding. Management treated the event seriously and documented the event in LER 50 282/97010. " Failure to Evaluate the Condition of a Residual Heat Removal Pump When the Vibration Level during a Surveillance Was Recorded at the Alert Value." The LER is further discussed in Section M8.7 of this report. The inspectors agreed with the licensee's conclusion that No.12 RHR pump vibration levels remained norme! throughout the period in question and that the pump was operabl M8 Miscellaneous Malntenance issues (92700,92902)
Sections M8.1 through M8.2 pertain to a maintenance rule followup inspection conducted by a region based inspector on July 22 through 24,1997. The purpose of the inspection was to review the status of findings reported in the maintenance rule baseline inspection (Inspec, tion Report 50 282/96012(DRS);
50-306/96012(DRS)). The following licensee documents were reviewed as part of the inspection:
- H24, Revision 2, " Maintenance Rule Program"
- PM 358610, Revision 0, " Periodic Structures Monitoring"
Maintenance Rule Function /MPFF (Maintenance Preventable Functional Failure) System Basis Document System AC through ZZ
- Maintenance Rule System Specific Basis Document
- Maintenance Rule System Basis Document Volume 1C, " Structures Monitoring Program"
- Table of the Maintenance Rule Scope Determination and Performance Criteria
- 1906 Annual e ,sipment Performance Report on Systems, Structures, and Components
- Expert Committee meeting minutes dated October 10,1996
- Expert Committee meeting minutes dated February 5,1997
- Expert Committee meeting minutes dated May 30,1997
- Expert Committee meeting minutes dated June 13,1997
- Expert Committee meeting minutes dated July 8,1997
- Expert Committee meeting minutes dated July 18,1997
l I
. -
M8.1 (Closed) Violation (VIO) 50 282/96012 01(DRS): 50 306/96012 01(DRS)):
Scoping Omissions. The violation involved the exclusion of the following structures, systems, or components (SSCs) from the scope of the Maintenance Rule (MR) program:
e communications e electrical cable trays e circulating water traveling screens e circulating water bay Examination of expert commi ttee meeting minutes, system basis documents, and a .
Table of the Maintenance Rule Scope Determination and Performance Criteria revealed that the four SSCs were placed in the scope of the MR at the conclusion of the October baseline inspection. Further examination of these documents revealed that there were no SSCs that were improperly excluded from the scope of the MR progra M8.2 -LClosed) Insoection Followuo item (IFI) 50 282/96012 02(DRS): 50 30619&D_t2-l 02(DRS): Completion of Required Reliability / Unavailability Balance. This item was opened to ensure a review of the reliability / unavailability balance required by Section (a)(3) of the MR. This balance was to be completed during the 1996 annual evaluation, which at the time of the baseline inspection had not been completed. A review of the 1996 Annual Equipment Performance Report on Systems, Structures, and Components revealed that the balance had been appropriately evaluated and that resolution of unavailability issues with some SSCs was appropriate. The review also revealed that the report was comprehensive, consistent with the guidance in Nuclear Management Resource Council 93 01, and in compliance with Section (a)(3) of the M M8.3 (Closed) VIO 50 282/96012-03a(DRS): 50-306/96012 03a(DRS): Lack of Unavailability Criteria. This issue involved a failure to establish specific unavailability criteria for four risk significant SSCs. Examination of expert committee meeting minutes, system basis documents, and a Table of the Maintenance Rule Scope Determination and performance Criteria revealed that the four systems identified during the baseline inspection and listed below now had appropriate unavailability criteria:
o safeguards buses 15/16 room coolers - < 168 brs^/ ear / train e safeguards chilled water < 168 hrs / year / train e reactor protection - < 144 hrs / year / train e nuclear instruments - < 144 hrs / year / train M8.4 (Closed) VIO 50-282/96012 03b(DRS): 50 306/96012-03b(DRS): Inappropriate Reliability Criteria. Expert committee meeting minutes, a Summary of Maintenance Rule Performance Criteria PRA (Probabilistic Risk Assessment) Sensitivity Analysis, and a Table of the Maintenance Rule Scope Determination and Performance Criteria were reviewed. The licensee's immediate corrective action was to establish general reliability criteria consisting of the more conservative of:
.. ._ _
_
.. _ _ _ _ _ _ . _ _ _ . . _______m _ _ _ . _ . . _ . ._ _ _ ... _ . -
,
<
l
.
- < 2 MPFFs/2 years, or I
- Demands or Run time < PRA assumptions for the specific failure mode of
- I concer !
The second consideration brought reliability criteria directly in line with the J
assumptions in the PRA. This resulted in changes to the reliability of many of the SSCs;in many cases a reliability criterion of 1 MPFF was established. Sensitivity
1 studies were conducted through June 1997 which further refined reliability criteria for several SSCs. The final sensitivity study which combined both reliability and
'
.
unavailability criteria resulted in a core damage frequency of 7.8E 5/yr. This was considered acceptable.
Goal. This violation involved the establishment of a performance goal for the i i 480 Volts Alternating Current (VAC) electrical distribution system which was tied
,
to the industry failure rate for 480VAC circuit breakers. Expert committee meeting
- minutes and the Table of the Maintenance Rule Scope Determination and ,
Performance Criteria were reviewed. The inspectors confirmed that the goal had been changed to no failure of risk significant 480VAC breakers due to auxiliary l contact lubrication problems during the 19961997 time period. This goal was considered acceptable, and at the time of this inspection, the goal was being met, g M8.6 (Closed) Unresolved item (URI) 50 282/96012-05(DRS): URI 50-306/96012-
'
05(DRS): Structure Monitoring Program. This item involved the lack of specific i
d guidance for implementation of the structure monitoririg program. Subsequent to the intpection, the licensee received an NRC letter dated October 1,1996, which authorized use of Nuclear Energy Institute (NEI) 96-03, " Industry Guideline for Monitoring the Condition of Structures at Nuclear Power Plants." The licensee
'
implemented this guidance through the addition of System Basis Document -
, Volume 1C, " Structures Monitoring Program," and procedure PM 3586 10,
,
" Periodic Structures inspection,".to be used to evaluate structures between the 5-l year civil engineering walkdowns. Based on their review of these two documents, 3 the inspectors concluded that NEl 96 03 guidance was appropriately implemented.
, Volume 1C of the system basis document provided guidance on the following i areas:
i e identification of structures
'
e staff responsibilities
- e examination objectives
- e methods of inspection
'
e follow-up actions
.* cause determinations
- e results evaluations e corrective actions
- e trending and industry data
,
, 15
- .- - , __, ---
. , - . . _ . _ -
. - _. . - - - - . - - . - - .- - , . -
. .
Procedure PM 358610 provided a checklist, based on NEl 96-03 guidance, which defined the scope of items for evaluation while performing the associated inspection, a list of common structural deficiencies, a tabulation of plant structures and deficient conditions, acceptance criteria, and required actions for deficient conditions. These were considered acceptabl M8.7 (Ocen) LER 50 282/97010: " Failure to Evaluate the Condition of a Residual Heat Removal Pump When the Vibration Level during a Surveillance Was Recorded at the Alert Value." The issues documented in this LER are discussed in Section M7.1 of this report. The LER is considered open pending completion of the corrective actions listed in the LE lil. Enaineerina E2 Engineering Support of Facilities and Equipment E Review of USAR Commitments (37551,92903)
While performing the inspections discussed in this report, the inspectors revi6wed the applicable portions of the USAR that related to the areas inspected and used the USAR as an engineering / technical support basis document. The inspectors compared plant practices, procedures, and/or paramoters to the USAR descriptions as discussed in each section. The inspectors verified that the USAR wording was consistent with the observed plant practices, procedures, and parameter E2.2 Control Room Habitability issues insnection Scoce (37551,929031 Three Mile Island (TMI) Action item lli.D.3.4, " Control Room Habitability," as promulgated in NUREG-0737, " Clarification of TMl Action Plan Requirements,"
required licensees to assure that control room operatois will be adequately protected against the effects of accidentai release of toxic and radioactive gase The inspectors reviewed various issues regarding habitability of the control room, including ventilation system response in the event of actuation of the carbon dioxide (Cardox) fire suppression system in the relay room, evaluation of toxic gas hazards, and operator preparedness in the event of a potentially incapacitating toxic gas release. The USAR, ventilation system descriptions and drawings, and various licensing submittals and responses were included in this revie Observations and Findinas The control room and relay room ventilation system provide a small amount of air ventilation to the relay and computer room, located directly beneath the control room. The relay and computer room return fan returns air directly to the control room environment and is designed to shut off and the isolation dampers are
. . . . . . .. .
. .
.
.
... .
- - - . _ . _ _ - - - _ _ - - . - _ _ - _-
. .
designed to close in the event of a Cardox system actuation or upon start of a control room ventilation cleanup fan. The inspectors asked the licensee if the shutoff logic for this fan was ever teste During its evaluation, the licensee identified that the Caidex system actuation interlock for the isolation dampers in the ducts between the control room and relay room had not been previously tested. The licensee then conducted a test and determined that the actuation heaters designed to melt the fusible links had not been properly installed and that the dampers would not have closed had there been a spurious actuation of the Cardox system. Of note, an actual fire in the relay room
,
may have caused the fusible links to melt and close the dampers. However, some amount of carbon dioxide would have entered the control room during a fire or if a spurious actuation of the system had occurred. The licensee initiated an evaluatiun
.
to determine the effect on control room habitability. This was considered an
Unresolved item, pending the lice 1see's completion of its evaluation and
subsequent NRC review. (URI 5t' 282/97015 04(DRP): 50 306/97015 04(DRP))
The inspectors reviewed the licensing correspondence regarding TMI Action item lli.D.3.4. The licensee had performed studies of onsite and offsite releases of hazardous substances and the impact on control room habitability. The inspectors identified that the licensee's December 13,1991, submittal to the NRC in support of removing the chlorine detection system from the Technical Specifications included within the scope of review, all onsite chemicals on its hazardous substance inventory that exceeded reportable limits of the Superfund Amendments and Reauthorization Act. Due to an oversight, carbon dioxide was not on this list and it was not included until February 1997. The licensee did not have a program in place to re evaluate any additionalimpact on control room habitability if hazardous substances in excess of Superfund Amendments and Resuthorization Act limits were stored onsite. A Nonconformance Report was initiated to document this and evaluate the impact of a Cardox tank rupture on the control room. This will be reviewed as part of the above Unresolved ite The inspectors reviewed the licensee's preparedness for a toxic gas releas Although the NRC had approved the elimination of toxic gas monitors on March 4, 1985 (for ammonia, hydrogen chloride, and formaldehyde) and on September 29, 1992 (for chlorine), the NRC stated in its March 4,1985, letter to the licencae tha "we require that you institute the training program as you proposed to assure toxic gas detection by the operators which includes donning self-contained breathing apparatus during such emergencies and to manually isolate the control room in the event of noticeable toxic chemical odors " The inspectors determined that the licensee was no longer meeting this commitment, and by the end of the inspection period, the licensee had not provided documentation that retracted this commitmen There are normally seven licensed watchstanding operators in the control room (five reactor operators and two senior reactor operators). In addition, a third senior reactor operator, the shift manager, would normally bo in the control roorn to perform shift technical advisor duties following an accident. However, there were
.
!
only four sets of self contained breathing apparatus (SCBA) staged in thr control room. Additionally, operators who were required to wear corrective lenses as a condition of their license did not have SCBA compatible corrective lenses staged in the control room. At the conclusion of the inspection period, after the finding was identified by the inspectors, the licensee was in the process of procuring SCBA-compatible corrective lenses for operators in need of the As an immediate corrective action in response to the toxic gas control room l habitability concern, the licensee closed all outside air dampers to the control room ventilation system and initiated Nonconformance Report 2010820 to evaluate the issue for reportability and long term corrective actions. The inspectors will further ioview this item for possible enforcement action as part of the above Unresolved Iter Conclusicq The inspectorn identihed several concerns with the licensee's assurance that the I
- control room would remain habitable or that operators would be protected and able to continue performing their duties in the event of a Cardox system actuation or toxic gas releste. The issues remained unresolved pending additional evaluation by the licensee and NRC review of licensee adherence to TMI Action item Ill.D. requirement IV. Plant Suncort R1 Radiological Protection and Chemistry Controls (71750)
During normal resident inspection activities, routine observations were conducted in the areas of radiological protection and chemistry controls using Inspection Procedure 7175 No discrepancies were noted. -
P1 Conduct of Emergency Preparedness Activities (71750)
During normal resiacnt inspection activities, routine observations were conducted in the area of emergency preparodness using inspection Procedure 71750. No discrepancies were note S1 Conduct of Security and Safeguards Activities (71750)
During normal resident inspection activities, routine observations were conducted in the areas of security and safeguards activities using inspection Procedure 71750. No discrepancies were note ._. _ . _ . _ . . . . . _ . _ _ _ . _ . _ _ - _ . . ~ . . - _ . . _ . _ _ . _ . . . ._ ._. _ _.m _.._.. . _ . . . _ .
. .
,. :
-
\
0 I Mfnaaemat Meetinas
!
'
X1 Exit Meeting Summery e
The inspectors presented the inspection results to members of licensee management at the
- conclusion of the inspection on August 5,1997. The licensee acknowledged the findings
! presented.
1.-
'
- The inspectors asked the licensee whether any materials examined during the inspection l- should be considered proprietary. No proprietary information was identified, i
i
+
4
I-i i
i
j <
9
i
i
,
- - , - . - , . ~ . . - . . ~ . . . . . - . - - - . . - - . - - - - - - - . - - - - - -
-- . -_
-.-
,
"
_
PARTIAL LIST OF PERSONS CONTACTED -
- Licensee
--J. Sorensen, Plant Manager .
.
- K."Albrecht, General Superintendent Engineering,' Electrical /l&C T, Amundson, General Superintendent Ei.gineering, Mechanical
. J. Goldsmith, General Superintendent Engineering, Generation Services -
J. Hill, Manager, Quality Services-G. Lenertzi General Superintendent, Plant Maintenance
'J. Maki, Outage Manager D. Schuelke,' General Superintendent, Radiation Protection and Chemistry T. Silverberg, General Superintendent, Plant Operations M. Sleigh, Superintendent, Security
_
INSPECTION PROCEDURES USED IP 37551: Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Oper'ations-IP 71750: - Plant Support Activities IP 92700: Onsite Follow up of Written Reports of Nonroutine Events at Power Reactor -
Facilities IP 92901: Followup - Operations IP 92902: Followup'- Maintenance
'IP 92903: Followup - Engineering ITEMS OPENED, CLOSED, AND DISCUSSED Onened 50 282(306)/97015-01 NCV Fuel Handling Error in Spent _ Fuel Pool 50-306/97015-02 VIO Failure to Follow Procedures to Declare Control Rod Deviation Monitor inoperable During Rod Position Verification 50 282/97015-03- NCV RHR Pump Vibration Not Evaluated per ASME Section
'
' XI Requirements 50 282(306)/97015-04 URI . Control Room Habitability - Evaluation and Preparedness for Toxic Gas Release 50-282/97010 LER ' Failure to Evaluate the Condition of a Residual Heat Removal Pump When the Vibration Level during a Surveillance Was Recorded at the Alert Value -
__ _
-
i
.
t
':
Closed
--
50 282/97009 LER Unavoidable Momentary Non-compliances with Technical Specification Requirements Which Do Not Provide a Time interval to Establish Different Plant Equipment Configurations Upon Moving into different Plant Conditions 50 282(306)/96012 01 VIO Maintenance Rule Scoping omissions 50 282(306)/96012-02 IFl Completion of Required Reliability / Unavailability Balance 50-282(306)/96012 03a VIO Lack of Unavailability Criteria
- 50 282(306)/96012-03b VIO Inappropriate Reliability Criteria 50-282(306)/96012-04 VIO Inappropriate (a)(1) Goal
_
50 282(306)/96012 05 URI Structure Monitoring Program l
UST OF ACRONYMS USED ANSI ' American National Standards Institute ASME American Society of Mechanical Engineers CFR Code of Federal Regulations CRDM Control Rod Drive Mechanism-DRP- Division of Reactor Projects DRS Division of Reactor Safety EOP Emergency Operating Procedure i
'
IFl inspection Followup Item IP inspection Procedure IRPI Individual Rod Position Indicator LER Licensee Event Report LOCA Loss of Coolant Accident MPFF Maintenance Preventable Functional Failure MR Maintenance Rule NCR Nonconformance Report
- NEl _ Nuclear Energy Institute -
NRC Nuclear Regulatory Commission PDR Public Document Room PM Preventive Maintenance PRA Probabilistic Risk Assessment P/T Pressure /Temperaturo RCS Reactor Coolant System RHR Residual Heat Removal RWST Refueling Water Storage Tank SCBA Self-Contained Breathing Apparatus SG- Steam Generator SP Surveillance Procedure SSC' Structure, System, or Component SWI Section Work instruction-TM Temporary Memorandum TMI Three Mile Island
_ _ _ _ _ _ _ _ - _ _
. o TS Technical Specifiertion URI Unresolved item-USAR Updated Safety Analysis Report VAC Volts Alternating Current VIO Violation WO Work Order
. .. ..
.
..
_