IR 05000237/1992010
| ML17177A477 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/08/1992 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17177A476 | List: |
| References | |
| 50-237-92-10, 50-249-92-10, NUDOCS 9206170108 | |
| Download: ML17177A477 (35) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION Ill Report No ~237/92010(0RP); 50-249/92010(0RP)
. Docket No "237; 50-249 license Nos. DPR-19; DPR-25 licensee:
Commonwealth Edison Company Opus *West 111 *
1400 Opus Place Downers Grove, IL 60515 Fac11 ity Name:. Dresden Nuclear Station, Units 2 and 3 Inspection At: *Dresden Site, Morris9 Illinois Inspection Conducted: April 11 through May 28, 1992 Inspectors:
w~ Rogers M. Peck R. Landsman D. Uao
- A. Markley M. Miller K. Shembarger
~
R. Zuffa, Illinois DeP, rtment of Nuclear Safety (/( C t,/B /9. Approved By:
Date Inspection Summarv
- Inspection from April 11 through May 28. 1992 <Reports No. 50-237/92010(PRP>:
- 50-249/92010(PRP>>.
.
Areas Inspected:
A. routine, unannounced safety inspection was conducted by the resident inspectors and an Illinois Department of Nuclear Safety inspector. The inspection included followup on previously identified items and licensee event reports; review of operational safety, plant startup from refuel, monthlY maintenance activities, monthly surveillance activities;
events followup; design changes review;. safety assessment and quality verification;.temporary instruction followup; Unit 2 turbine building concrete cracking; and meetings and other activities; Resylts; Of the areas inspected, one non-cited violation was identified in
_paragraph Two open items were identified in paragraph Four unresolved
.items were identified in paragraphs 4, 7, 8, and 10. One SIMS item, TI
- 2515/112 was closed in paragraph 1 Plant--Qperations *control room operators were professional except in one-*
instance. Operators were generally knowledgeable of annunciator alarm Good
- control of activities during startups were noted. Reactor operator log.
9206170108 920608
.PDR ADOCK 05000237 G
quality was good but, shift engineer log quality was weake Some Technical Specification knowledge deficiencies were noted. Operator identification and resolution of housekeeping deficiencies were not readily evident. Weaknesses were observed in operator knowledge of administrative controls. -
Maintenance/Syrye1Jlance Maintenance activities in the field appeared
. adequate. However, coordination and completion.of work on the ACAD system appeared weak. * Some weaknesses were noted in the tont~nt of surveillance procedures. * *
Engineering and Technical Sypport Newly written design change packages for the Unit 3 refuel outage appeared goo *
Safety Assessment and Quality Verification Weaknesses in the co~rective action process as discussed in inspection report 237/92009; 249/92009 continue to be evident.*
DETAILS l~
Persons Contacted C. Schroeder, Station Manager*
L..Gerner, Technical* Superintende*nt
- J. Kotowski,.Production Superintendent D. Van Pelt, Assistant Superintendent, Maintenance J. Achterberg, Assistant Superintendent, Work Planning
- G. Smith, Assistant Superintendent, Operations M. Strait, Technical Staff Supervisor
- R. Radtke, Regulatory Assurance Supervisor
- Denotes those attending the exit interview conducted on May 22, 1992 *
. The inspectors al so talked. with. and interviewed several.. other l i censee employees during the course of the inspectio *
- Licensee Action.on Previously Identified Items (92701. 92702)
a~
(Closed) Unr*solved Item.(237/91010-04(DRP)):
Safety Relate Pressure Switch Calibration Deficienc Upon review of. the licensee's corrective actions, associated with violation 237/91016, con~erns were raised as to wheth~r the reactor building ventilation (RBV) system had been capable of performing its
- intended function, i. e.., isolate the secondary containment under a s~stained loss of valve operator air pressur *
To resolve this concern, the licensee performed Special.Procedure*
(SP) 92-2-34 on reactor building inboard isolation damper, A03-5741B. * The test determined the lowest motive system air pressure under which the damper would close was 35 psig. Three of the total eight secondary containment isolation valves (four per unit)
. had as-found pressure switch settings of less than 35 psi By design there are two inlet as well as two outlet isolation valves
.in series with each other. -The redundant isolation valve to each of the isolation valves with low as-found pressure switches was above the 35 psig actuation pressure. Therefore, at least one out of the two secondary containment isolation valves on both the irilet and outlet ventilation piping would have sufficient motive air pressure to close *. This matter is considered close (Closed) Violation (237/90022-0l(DRP); 249/90022~0l(DRP)):
Failure. to perform a 10 CFR 50.59 safety evaluation associated with the installation and use of a portable air sample pump *. The sample pump, used to obtain daily drywell air samples, provided a containment leakage path in excess of the Technical Specification containment leak rate requiremen Corrective action to the violation incltided rebaselining the Final Safety Analysis Report (FSAR}, performing an evaluation of the
- cont-ainment air sample -system, revising the temporary system
alteration administrative procedure, revising the safety
_
evaluation administrative procedures and performing root cause evaluations of the recent 10 CFR 50.59 evaluation The inspector reviewed the licensee's co~rective actions and determined two were outstanding at the end of the inspection period. These were:
1)
A conunitment to rebaseline the Dresden licensing basis. The effort will culminate in a total revision of the FSA The licensee is committed to complete.this effort hi 199 ).
A commitment to perform an engineering evaluation of the conta~nment air sample syste The *valuation has been performed and the 1 itensee is in the process of drafting a *
letter to the NRC proposing how the system will be use Completion of these two correcttve actions is.c~nsidered an open item (237/92010-0l(DRP)).
Also, similar to the situation discussed in in~pection report 237/92009; 249/92009 sections 6 through 9, some of the other corrective actions were not completed by the committed date. Th failure to implement these corrective actions by the dates *
committed re-emphasizeQ the weakness in the licensee's ~orrective action process as discussed in inspection report 237/92009;
249/9200. (Closed) Op~n Item (249/89011-02(DRP)):
Corrective actions to ensure against radioactive releases when using the isolation condenser for extended time periods without offsite powe The *
licensee revised the appropriate operating procedures to specify
- the use of contaminated demi nera 1 i zed water for she 11 side makeup as the least preferred sour~e. In addition, long term actions to
- provide additional clean demineralized water capacity are scheduled for completion at the end of the next refueling outages for Unit 2 and Unit 3. Completion of this long term corrective action is considered an open item (249/92010-02(DRP)). (Closed) Unresolved Item (237/88010-0l(DRS}; 249/88012-0l{DRS)):
Potential effects of a fire originating in Unit 1 on Units 2 and In a memo dated March 31, 1992; from *Byron Siegel, NRC, to Thomas Kovach,.cECo, the NRC concluded that CECo adequately.
responded to and resolved the concerns raised regarding the propagation of a Unit 1 fire. Therefore, this unresolved item is considered close * (Open) Unresalved Item (237/91022-03(DRP}):
Performance of a core*
spray pump surveillance with the suction source isolated in July, 1991. rhe event was caused by an inadequate surveillance.
procedure which indicated that a manual valve at the condensate storage tank was locked.open when it was-locked closedL-Also, the
applicable critical drawing in the control room, M-35, showed the
- valve locke~ ope The inspector determined neither the
... _
surveillance procedure or the critical drawing had been revise The only outstanding matter is for the inspector to determine the safety significance of the. inadequate procedure/drawing as it-relates to operator actions when using the emergency-operating pr~cedure *
-
No viol at ions or deviations were -identifie ;
Licensee Event Reports Followup C92700l The iollowing licehsee event reports were reviewed to ensure that reportability requirements were met, and that corrective actions, both immediate and to prevent recurrence, were accomplished in accordance with the Technical Specifications:
- *
.
' (Open) LER 237/92010, Source Range Monitor Calibration Test Frequency Technical Specification Requirements Not Met due to Management Deficiency The inspector reviewed th~ corrective actions associated with the change in surveillance frequency and determined no corrective actions addressed the inadequate review in the frequency change for the surveillance. Therefore, the root cause of the event was not addresse The inspector met with the licensee to discuss this matter and the licensee committed to review the matter furthe The LER ~emains open pending further licensee cor~ective action (Open) LER 237/91026, Valve Closures during 125 VDC Ground Checking due to Procedure Deficiency *
While performing the followup to the unplanned ESE actuation on May 7, 1992, of the HPCI drain valve to the torus during ground checking activities the inspector determined corrective actions to this LER were incomplet The licensee was to review the LPCI, HPC I, and core spray logics to determine the effects of circuit
.
de-energization on the system The review was to be performed by November 1, 1991, with the results to be provided to the operations*staff for inclusion into the ground checking procedure Discussions with the technical staff on May 18, 1992, indicated the review had not bee~ complete *
The inspector informed the licensee of the situation and the need
- to revise the LER corrective action No violations or deviations were identified in this are *- Operational Safety Verification C71707l The inspectors reviewed the facility for conformance with the license and with regulatory requirement On a $ampling basis the inspectors observedcontrol room activities for proper control room staffing and coordination of *
plant activities. - Operator adherence to procedures or Technica Specifications and operator cognizance of plant parameters. and alarms was observe Electrical power configuration was confirmed. Various logs and surveillance records were reviewed.
for accuracy and completeness. *
Observations included:
On April 21, 1992, a control room licensed operator wai briefly inattentive to dut The inspector followed the licensee's investigation into the event ana corrective a~tions taken and co~siders the issue ~losed.*
On April 10, 1992, while attempting to insert rod M-4 from step 16 to step 14 on Unit 2, the reactor operator inserted the rod to step 1 After discussion with Qualified Nuclear Engineers, the rod was repositioned to the desired positio *. During review of the event, the licensee determined that M-4 was ~ fast rod, which resulted in the rod inserting the two additional steps, and that personnel error was not a contributing factor to the even The inspector followed *
the licensee's review of the event and concurred with the finding When Unit 3 was in the process of.shuttirig down to repair a
. leaking reactor recirculation pump seal the operating crew failed to N/A or ~hange five prerequisites to the general
- shutdown procedure. Therefore, the operators were using the
. procedure without the prerequisites signed~ Once identified to the crew corrective actions were taken to revise the *
procedure and/or N/A the steps. Though th~ procedure was not properly ~nnotated or revised, the shutdown activities were adequately controlled. Operations management discussed the failure to use the administrative controls available to
.deal wit~ proced~ral problems ~ith the operating crew and discussed the situation with other crew While reviewing the center desk operator log book, the.
inspector noted an April 2, 1992, entry where a non-licensed operator attempted to clear the 2/3 diesel generator day tank hi level alarm at the request of the shift enginee To clear the alarm the operator drained the tank below a licensee imposed 40 inch minimum level. Also, the inspecto observed the Unit 2/3 diesel generator day tank below th licensee's ~dministrative limit later in the inspection period. Subsequently, operators identified the Unit 2 day tank below the administrative limit that same da Th draining evolution and having the two day tanks below the administrative limit will be further reviewed by the
inspector for safety *significance along*with any other *
contributing factors. This matter is considered unresolved (237/92010-03(DRP)) pending completion of the inspector's
revie *
While reviewing the actions taken by operators when an onsite emergency power source or an offsite normal power source i~ lo~t the inspe~tors identified a discrepancy in licensee actions versus. those stated in Technical Specification 3.0.b. The licensee considered verification of redundant equipment on the division not affected by the loss of the power source to bnly be at the 4 KV level.. The*
eq~ipment below the 4 KV level should ~lso be ccinsidered:
The licensee was informed of this philosophy conflict and was reviewing their posture on this at the end of the inspection perio *
Control room operator log entries continued to be of good content.. Weaknesses continue in shift engin~er log keepin h. *.On a routine basis the inspectors toured accessible areas of the facility to assess worker adherence to radiation protection
.
controlS and the site security plan, housekeeping or cleanl 1ness,.
and control of field activities in progres *
Observations included:
While reviewing the diesel generator fuel supply syste~s, the inspector identified the Unit 3 diesel generator fuel transfer system is used to supply fuel to the 2/3 diesel fire pump and a heating boiler. Technical Specification 3.9.C states nThere shall be a minimum of 10,000 gallons of di.esel fuel supply on site for each diesel. n Immediate*
discussions with technical staff personnel indicated that the quantity of fuel required to operate the 2/3 diesel fire pump and the heating boiler was not factored into an..
administrative control to ensure the *Technical Specification fuel requirement was being maintained. Subsequent review by the licensee determined, although an administrative contro was not in place, adequate volume was being maintained since a minimal quantity of fuel was* required for operation of the fire pump and heating boiler. This minimal quantity could easily be compensated for by the.fuel in the diesel
.
generator day tank that was not hicl uded in the ca 1 cul at ion.
to determine the minimum Technical Specification
. requiremen An additional inspector concern was the affect
-of a malfunction/failure in the diesel fire pump or heating boiler systems on the fuel transfer system's ability to provide fuel to the Unit 3 diesel,generator day tank. The control of adequate fuel in the Unit 3 diesel generator fuel supply tank and the affects of a malfunction/failure in the diesel fire pump and heating boiler systems will be reviewed further in the next inspection report perio *
Housekeeping deficiencies were observed around the Unit 2 safety related 4160 KV electrical buse Once identified the material was remove However, the licensee's housekeeping efforts were not effective in resolving the problem prior to the inspector's observation Walkdowns of select engineered safety features (ESF) were performe The ESFs were reviewed for proper valve and electrical alignm~nts. Components were inspected for leakage, lubrication,
abnormal corrosion, ventil~tion and c6oling water supply availability. Tagouts and jumper records were reviewed for accuracy where appropriat The ESFs reviewed were:
Unit 2
Emergency Electrical Busses 23-1 and 24-1
250 VDC Batteries Unit 3
Core Spray System
Unit 3 Diesel Generator
Secondary. Containment (Reactor Bui.ldfog Vent System)
Emergency Electrical Busses 33-1.and 34-1 No v*olations or deviations were identified in this are s;
. Plant Startup from Refuel (71711).
During the inspection period~ the inspector monitored activities of control room operators and other operations *support personnel during the start up, approach to critically,. heat up, synchronization, and powe operation of the Unit 3 reactor. The inspector verified that the control rod withdrawal.sequence and rod withdrawal authorization were available and all surveillance requiring completion prior to start up were successfully perfor:-med.. The inspector also verified that the unit*
start up was being.performed in accordance with technically adequate and*
. approved procedures', and the start up was conducted in accordance with the requirements of the Technical Specification No deviatitin~ or v~olati-0n~ were identified in this area.
. 8
- Monthly Maintenance Observation (62703)
Station maintenance activities were observed to verify that they were
- conducted in accordance with _approved procedures and work packages, *
regulatory or industry guidance, and in conformance with Technical Specifications limiting conditions for operation The *inspectors*
verified that approvals were obtained prior to work initiation,-that*
quality c*ontrol inspections occurred, that appropriate post-maintenance functional tests or calibrations were performed, that maintenance personnel were qualified, that parts and materials used were properly certified; and that proper radiological and fire prevention controls were implemente The status of outstanding jobs was also reviewed to ensure that appropriate priority was assigned to maintenance of safety.;..
related equipment which could affect system performanc.
.
The following maintenance activities were-observed and reviewed:
- Unit 2
Repair and Troubleshooting of IRM 16 *
Reactor Recirculation Master Flow Controller Troubleshooting
Atmospheric Containment Atmosphere Dilution (ACAD) Repair Unit 3
Area Radiation Monitor Check and Calibration,
.
SP-92-2-40, Control Rod Drives Ball Check Valve Reseat
3B Feedwater Regulating Valve Repair
3B Reactor Recircul~tion Pump Seal Replacement
2/3A Standby Gas Treatment System Filter Change Out
3B Condensate Booster Pump Rebuild
..
Hydrogen Seal Oil Pressure Regulating* Valve Rebuild/Adju_stment Inspector observations were:
a.*
Special test procedure SP-92-2-40 required only the documentation of prerequisite actions f6r the first control rod drive teste *The procedure did not provide a mechanism to document the completion of prerequisites for subsequent control rod drive This was considered a weakness in the t~sting procedur The inspector noted that a substantial period of the licensee' administrative limiting condition for operation (LCOs) action statement was necessary to repair the ACAD syste The licensee
. performed a lessons learned to reduce the time in the action statemen The inspector will follow the effectiveness of these actions in further maintenance activities involving ad~inistrative LCO.
No violations or deviations were identifi~d.
7~
Monthlv Surveillance Obseryatjon C61726l The inspectors observed required surveillance testing and verified procedural adherence, test equipment calibration, Technical.
Specification action statement adherence, and proper removal and restoration of affected components. The inspectors rev1e~ed completed surve11 lance packages to ensure that results confonned with Techn i.ca 1 *
Specification and procedure.requirements, that there was independent verification of the results, that proper s1gnoffs occurred, and. that any test deficiencies were appropriately dispositione *
. The inspectors witnessed portions of the following test attivities:
Unit 2
- ** *
DOS 1300-01, DOS *.1400-01, DOS 1400-02, DOS 1400-05, DOS 1500-06,
. DOS 1500-10, Isolation. Condenser Five Year Heat Removal C~pability Test Core Spray system Pump Test with Torus Avaiiable*
Core Spray System Valve Operability Check Quarterly Core Spray System Pump Test with Torus Available for the In-Service Test (ISi) Program LPCI System Pump Operability Test with Torus Available
Quarterly LPCI System Pump Operability Test with Tor.us Available for the In-Service Test (IST) Program Unit 3
DOS 500-10, Turbine Stop Valve Closure Scram Circuit Functional Test.
.
DIS 0700-03, Source Range Monitor Calibration Common
DFPS 4123~07, Unit 1 Fire Pump Annual Capacity Test
DOS 6600-01, Diesel Generator Surveillance Test
DTS 6600-02, Diesel Generator Fuel Consumption Test Inspector observations were; On May 16, 1992, the inspector observed the performance of DOS 1300-01, *a1solation Condenser Five Year Heat Removal Capability Test", 6i the Unit 2 isolation condenser. Technical Specification 4.5.1.c states *the system (isolation condenser)
heat removal capability shall be determined once every five**
years." Dresden Station FSAR, Section 4.6.1 states "... the - *
isolation condenser was designed for a cooling rate of 252.5 x 10(6) Btu/hr". Section 4.6.3, states *The capacity of this system (isolation condenser), equivalent to the decay heat rate after 5 minutes, will absorb decay heat as it is produced.* With no
- makeup the water* stored above the isolation condenser tubes is
- depleted in 20 minutes.*
The inspector reviewed the procedure prior to performance and discussed the inethod of determining the h.eat removal capacity with the technical. staff. The inspector was concerned that the. method *
used would not correlate to the heat removal capacity jn the des.ign basis situation. During the test, the isolation condenser
.shell side temperature did not stabilize, and there is a question as to whether or not the clean demineralized water temperature and flow used in the procedure calculations were accurat The adequacy of the test performed is considered an unresolved ite (237\\92010-04(DRP)) On April 16, 1992, the inspector observed the performance of DFPS 4123-07 ori the ~nit 2/3.diesel fire pum Step 45 in the.
procedure required the operator to verify that the relief valve on the discharge of the pump opened at greater than or equ~l to 150 psi. The procedure step provided no document*tion for th~
actual pressure the relief valve opened and provided no upper acceptance limit. The inspector informed the fire marshall of these procedural weaknesse The inspector noted that the Source Range Monitor (SRM)
Calibration Procedure (DIS 0700~03) was si~nificantly changed to
- re-align the entire SRM chassis. This procedural change
represented a more thorough calibratioh*of the source range monitor integral components, thus increasing instrument accuracy
.and overa 11 performanc No violations or deviations were identifie.
Events Followup {93702)
During the inspection period, several events occurred, some of which required prompt notification of the NRC pursuant to 10 CFR 50.7 The inspectors pursued the events onsite with the licensee and with NRC **
official In each case, the inspectors reviewed the accuracy and
. timeliness of the licensee notification, the licensee's corrective actions and that activities were conducted within regulatory requirements; The specific events reviewed were: April 9,1992 - Declaration of an.Unusual Event on Unit 3 when the high pressure coolant injection system developed a severe oil leak requiring reactor shutdown during refuel startup testing. Within four hours the unit was in cold shutdown.and the Unusual Event terminate April 10, 1992 - Unplanned engineered safety feat~res (ESF) *
actuations of the standby gas treatment syste While performing surveillance procedure DIS 7500-1 "Standby Gas Treatment Auto Actuation Survefll ance", the train "A" standby,gas treatment.
y system automatically started twice. The first start was attributed by the licensee to personnel error when the wrong fuse
. was pulled to prohibit the automatic start of the train during surveillance testin *
Following recovery from this automatic *start, the license continued the surv~illance. Subsequently, *with.a simulated initiation signal present the reactor btii1ding ventilation failed to isolate and train *e* of the standby gas treatment system failed to start. Train *A* automatically started on.
a low flow signal from the *e* train per design. All*
isolation signals were reset, and the tests repeated twice, with no further fa i1 ure The licensee established an investigation team and the inspector will review their results as followup to the future LER on this even April 19, 1992. - Unplanned ESF ~f the lriw pressure coolant injection (LPCI) syste While performing Unit 3 DOS 1500-1, *LPCI Valve Operability Te~t,* the LPCI minimum flow
valve, 3-1501-13A, automatically closed while cycling the LPCI test line valve~ 3-150l-38 The 1 icensee determined*
the flow transmitter in the LPCI loop, upstream of 38A valve, had received a high LPCI flow signal closing the 13A valve. The piping between the LPCI test line valves 3-1501-38A & 20A is normally empty. Therefore, when the 38A valve opened the keep-fill system filled the empty space cre~ting an instantaneous high flow signal which tripped the minimum flow valve. This event was not reported within the time *
constraints of 10.CFR 50.72 and the root cause of not initiating such a report is discussed in inspection report 237/92009; 249/92009, section May 8, 1992 - The licensee entered into the emergency operating procedures for Unit 2 when a reactor operator identified an.
increase in control room torus/level indication above ~2.5.inche Level had risen from a normal level of -2.9 inches to -2.0 inches *
- over the. previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee identified th~ HPCI
system torus return line valve was open, and the valve was immediately closed. Also, a pumpdown of the torus was initiate When torus level reached -3.7 inches the pumpdown was secure Due to an equipment malfunction the high torus level annunciator failed to alarm in the control room when -2.5 inches was reache A work request was issued to repair the high level alarm in the control room, and a procedure change was generated to include the HPCI valve in the grotind check procedur *
Licensee inv~stigation into the event revealed that on May 7, 1992, a ground check on the HPCI valve DC bus was performed, but the procedure fa i.l ed to identify the HPC I torus return line valve, 2301-28, on the bu As a result, the solenoid to the valve actuator--was de-energize Due. to.. a-pre'.""existing
- *
e. I *sticking* of the valve actuator the valve did not 'open until_ two shifts late The licensee validated through test that the ground check activities would cause the valve to open._ *Following the te*st the shift engineer informed the NRC via the ENS of an unplanned ESF
(opening of the 2301-28 valve) durj."g the testing but failed to *
inform the NRC of the first unplanned ESF actuation of the 2301-28 valve on May 7, 199 * *
The failure to report the first actuation was a violation of 10 CFR 50.7 However, the violation was categorized at Severity Level V and it is not being cited because the criteria specified in Section VII.B.l of the *General Statement of Policy and Procedures for the NRC Enforcement Actions,* (Enforcement Policy, 10 CFR Part 2,* Appendix C (1992), were satisfied *.
The root cause of the violation was an incomplete understandfog by the shift engineer that calling in the second actuation did not cover the first actuation als The pre-existing "sticking* condition of the 2301-28 valve actuator was identified by the licensee on April 20, 199 Further evaluation is.necessary by the inspector to determine whether this "sticking* condition affected*HPCI operability. This matter is consider_ed unresolved (237 /92010"-05{DRP)) pending completion of that revie May 11, 1992 - The licensee determined that three piping supports in the Unit 2-high pressure coolant injection {HPCI) system were*
overstressed beyond FSAR allowable stress limits. The licensee evaluated the stresses on the piping supports and determined they were operable under the criteria ~tated in Generic Letter 91-1 May 11, 1992 - A controlled shutdown of Unit 3 was initiated to allow for replacement of the 38 Recirculation Pump No. 1 Sea The unit was off-line at 0215 on May 1 Indications of seal*
failure occurred on May 10, prompting a power decrease to minimum recirculation flo *
May 14, 1992 - The licensee determined that a section of th hydrogen monitoring system *on both units was not being leakrate tested. Adequate compensatory.measures were taken for th operating unit and the section of piping in question was tested-prior to mode ascension oil the shutdown uni One non-cited ~iolation and no deviation~ were identified in this are Design Changes Review (37700)
The i~spectors reviewed the documentation for the following modifications including the design basis, engineering evaluations,
- ...
-"-.
. *-
.
.
~--..
engineering-calculations, installation instructions, and te.sting and found them to be complete and accurate with few discrepancies. The design changes reviewed were:
Modify EDGSW Piping and Install New Check. Valve
- *
Rep 1 acement of Power Cab 1 e ;feeds tQ Motor Contro 1 Center
{MCC) 388-2.. *
- .
- ..
.... ** *. *. *
Replace Wide Range Torus Level 'Indicators..
,Upgrade MSIV Air *Accumul~tor Suppor.ts
Replace Feedwater Flow Transmitters
. * MSL Low Pressure Switch Time Delay Modifications Inspector observations were~ Modification documentation includirig th~ ~esign basis~ engirteering evaluations, engineering calculations, installation instructions,.
and testing were complete and accurate with.one exceptio *
.
.
.
.
.
The time dela_x setpoint for the MSl low pressure switch modification for Unit 2 was calculated to be 270 +/- 5 msec in order to ensure that the FSAR requirement of 500 msec was not exceeded when added to the existing time delays inherent in the**
syste However, when the maximum setting of 275 was added to the existing baseline delay of 226.81 msec, the FSAR requirement was exceeded by 1.81 msec. Subsequent investigation revealed that the baseline delay value stated in the modification letter was incorrect and the total delay time was well below the FSAR requirement. A revised modification memorandum was issued reflecting the correct values.**
- Design dat~ and engineering calculations for mo~ifications originated by corporate engineering were not included in the documentation sent to site engineering for review and
implementation unless specifically requested. This is a weakness in the program in that.information necessary for decision makfog is not readily available to site management. The licensee agreed t~ consider applicable corrective measure Minor discrepancies were communicated to the licensee and corrections were made in a timely manner..
No violations. or deviations were identified in this are.
Safetv Assessment and O~ality Verification C40500l Procurement On April 3, 1992, a portable terminal interface board and two.
.
relays were replaced in the 2/3 Chimney Sping syste The parts, originally.classified as regulatory-related, were ordered and installed under the non safety-related classification. During
- review of the evolution, the inspector determined that the parts classification change was due to a memo issued to the Dresden.
Station from CECo corporate engineering, dated April 4, 1992, which stated Regulatory Guide<h97 A*nstruments are non safety-rel ated subject to the requirement$ impos.ed by their various categories. * Subs*equent review by *the 1 icensee determined that
- since the 2/3 Chimney Sping 1s.used for environmental monitoring, and environmental monitoring equipment is classified as regulatory-related, the.parts,;classif.ication should have remained regulatory-related. A work request was written to temporarily remove the interface board and relays to allow a regulatory-related review to be performed on the parts. *
Further inspection into this matter is necessary to fully assess the weakness in the parts classification process that did not assure proper parts dedication. *This matter is considered an unresolved (237/92010-06(DRP)) pending further inspectio Dresden Management Improvement Activities Dresden Management Action Plans (DMAP)
A large number of the licensee's improvement efforts are*
implemented through DMA These plans encompass such tasks as procedure ~pgrades; communication of management and performance expectations, plant enhancements and improvements, with specific actions for each applicable organizational discipline. A project group comprised of a coordiriator, several supp~rt personnel, and several consultants facilitate the DMAP effort. This group i responsible for coordinating the development of' action*
plans, tracking implementation progress, and produting status report *
The DMAP program is tracked on a database on the licensee's mainframe compute Each entry into the DMAP database identifies the responsible manager/supervisor, the plan identifiers, completion status, source of th~
commitment/action plan, action plan objective, step identifier, description~ responsible persons, schedule/actual finish date, status, and cognizant department.*
. Presently, a monthly status report is produced identifying
- the total number of steps in the action plan population,
. total number of steps completed, total com~leted last month, total due this month, total due* next month, total steps open, and total steps overdu The licensee has selected five of the DMAP areas for accelerated attehtion. These include procedure process and control, work planning and control, control room processes,
material condition, and modification process imp~ovement Accelerated attention is intended to achieve significant, visible results in a short period of tim * Future inspection reports will evaluate the effectiveness of the DMAP syste '.
. 2. *
Work Planning and Control One of the DMAP activities is upgrading the work planning,
proces The work planning organization is comprised of an Assistant Superintendent of Planning and Scheduling, work planning supervisori and t~o planning group The outage planning *
group includes one senior outage planner, a lead planner (operations background), and two computer personnel per unit. In addition, a contractor has been ~etained to assist with unit twQ outage planning efforts. The daily planning group has two licensed operators (one SRO and one RO),
~echanical, electrical, and instrument maintenance coordinators, a technical specification surveillance coordinator, and a*computer ~erson. Each operator in this
- group is unit assigned and is responsible for coordinating the respective unit activities. The organization lacks a radiation protection presence..
The licerisee is implementing a new system of daily planning for routine Jnd non-routi~e work.* The new system is being developed around a rolling 12 week equipment.schedule.*
Within this, a 28 day surveillance schedule of activities is include From this 12 week schedule, a two weet look.
ahead, a a*ne week 1 oak ahead, and the current 3 day ro 11 i ng schedules are produced. A new database program, MOTS,
.automatically downloads from other mainframe computer programs such as the Technical Specification surveillance and the lubrication program to identify some of the
- activities.for the daily schedule. The effectiveriessof the new system will be assessed following further development
and use of the new system by the 1 i cense *
No violations or deviations were identifie.
Temoorary Instruction CTI> 2~15/112 Followup (Closedf TI 2515/112, Licensee Evaluations of Changes to the Envir~ns around licensed Reactor Facilities:.The TI instructs the inspector to evaluate the method licensees use to identify changes to the environs around reactor facilities and whether ide~tified changes are incorporated into the.Updated Final Safety Analysis* Report (UFSAR).
. '1
. 16
Station personnel stated that no program was in place to identify these changes *for the sole purpose of updating the UFSA Instead one
.
specific person was assigned to update the UFSAR and to identify changes around the facility. Changes were identified through information
- *
conveyed by public media, word of mouth, and review of licensee document The only periodic review follows the national 10 year *
census, to identify population changes *. There were no changes that the inspectors were aware of that have riot been identified by the license No violations or deviations were.identified *
. 1 Unit 2 Turbine Building Concrete Cracking Following* the observation by regional inspectors of cracking concrete in the Unit 2 turbine building (refer to inspection report 237/91033:
249/91036) a regional specialist was di~patched to re~iew the concrete cracking situation. The inspector observed spalled concrete in the Unit
- 2 turbine building walls near column line 43-E below.floor elevatiori 561 '-6" a Jong with. the associated floor damag Cracking on the Unit 3 *
turbine building floor was also observe The floor cracks were along old construction joints between an 8 inch concrete floor slab in the middle of the turbine building and 54 inch concrete floor slabs on either side which are adjacent to the Units turbines. There appeared to be no structural signifi~ance to the floor cracks since the eight inch slab spans the north-south direction between steel support beams and does not extend to the 54 inch slabs. The licensee's an~lysis that the present bearing areas/beam seat areas were *
adequate and the proposed cosmetic repair appeared acceptabl *
Ho~ever, beam seat spalling could become a structutal problem if the root cause of the apparent continuing movement or vibration causing the
. spalling is not identified. Until su.ch time as this is determined and corrected, the reduced bearing areas could lead to higher stresses on the remaining intJct concrete under the beams, which might lead to more spalling in the futur As stated in inspection report 237/91033:
249/91036 the Jack of tracking of this issue is considered a ~~akness~
No violatirins or deviations were identifi~.
Meetings and Other Activities (30702)
a~
On May 14, * 1992, *the Regional Administrator and. staff met with the Head of CECo and his senior corporate staff at the Dresden training building. The meeting topics included current NRC perspectives at various CECo nuclear sites and; licensee
.presentations on current status of operations performance review*
from external soutces, material condition assessments at Dresden, human performance evaluation initiatiVes at Dresden and diesel generator reliability at the Zion Statio....
'
. y On May 28, 1992, the Director of Nuclear*Reactor Regulation, the Regional Administrator and an aid to the Chairman-of the NRC toured the Dresd~n site. Afterwards the gentlemen met with CECo senior management. * The topics of discussion were present station.
challenges, operations performance reviews from internal and external to the company, the progr.ess of the.. vul nerabi 1 i ty assessment team, site engineering initiatives and diesel generator reliability at the Zion Statio *
1 Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or bot Two items disclosed during _the.inspection are discussed in* Paragr.aph. * J Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations~ or deviations. Unresolved items disclosed during the inspection are.*
discussed in Paragraphs 4, 7, 8, and 1 *
1 Violations For Which A "~otice of Violation" Will Not Be Issued The NRC uses the Notice of Violation to formally document failure to meet a legally binding requirement. However,. because the NRC wants to encourage and support licensee's initiatives for self~identification and correction of problems, the NRC will not issue a Noti.ce of Violation if the requirements set forth in 10 CFR 2, Appendix C, Section VII.B.l or VII.B.2 are met. A violation.of regulatory requirements identified during the inspection for which a Notice of Violation will not be i~sued *
i~ discussed in f~ragraph sj 1 Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) during the inspection period and at the c~nclusion of the inspection
- period on May 22, 199 The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in nature. Also, the licensee disagreed with the inspector's conclusion that the operator discussed in paragraph 4.a. was inattentiv *
- ,1
Commonwealth Edison 1400 Opus Place Downers Grove, Illinois 60515 October 2, 1992 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Subject:
Reference:
Dresden Nuclear Power Station Units 2 and 3 Response to Notice of Violation *
.
- Inspection Report 237/92009;249/92009 NRC Docket Numbers 50-237 and 50-249 A. B. Davis letter to Cordell Reed dated September 2, 1992, transmitting NRC Notice of Violation
- Enclosed is Commonwealth Edison's Company's (CECo) response to the Notice of Violation (NOV) which was transmitted with the reference letter and associated wit Inspection Report 237(249)/92009. The NOV cited one Severity Level Ill violation and six Severity Level IV violations requiring a written response. Our response to these violations is provided in Attachment A. Attachment B provides our response to additional information requested in the reference cover letter pertaining to the status of the Integrated Reporting Program and effectiveness of prior corrective action *
If your staff has any questions or comments concerning this letter, please refer them to Denise Saccomando, Compliance Engineer at (708) 515-728.
Very truly yours,
/!-t;.~J;;,-
T.J. Kovach
Nuclear Lice*nsing Manager Attachments cc:
A. B. Davis, Regional Administrator Region.Ill B. L. Siegel, Project Manager, NRR
W. G. Rogers, Senior Resident Inspector, Dresden 150054 w~LJ0070/14
} '7ff'
OCT 58 ZNLD/2170/1
\\.
\\ ATT AC HM ENT A RESPONSE TO NOTICE OF VIOLATION NRC INSPECTION REPORT 50-237/92009; 50-249/92009 VIOLATION: A Technical Specification 3.5.a.5 states, in part, that from and after the date that the low pressure coolant injection (LPCI) subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless it is sooner made operabl Technical Specification 3.5.a.8 states, in part, that if the requirements of 3.5.a cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the cold shutdown condition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Contrary to the above, from January 4, 1991, until August 10, 1991, for periods
.
greater than seven days, reactor operations continued with the LPCI system inoperable, in that, reactor recirculation valve 2-202-SA, a motor operated valve (MOV) required to close to ensure LPCI injection into the reactor vessel following a loss-of-coolant accident, was incapable of performing its safety function due to an incorrectly set torque switch on the MOV, and the licensee did not initiate an orderly shutdown of the reactor and place it in cold shutdown within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> REASON FOR VIOLATION:
On August 7.,.1991, valve 2-202-5A was discovered to be unable to close against differential pressure when Operations personnel attempted to close it as part of the routine pump restart sequenc The root cause of the SA valve being inoperable was the misinterpretation of the VOTES trace, due to valve stem anomalies which were not apparent to the valve testers. This led to the incorrect setting of the valve's closing thrust value. Because
- of limitations in the VOTES testing program, including the methodology, training, and software, the person performing the valve testing was not equipped to properly evaluate the reactor recirculation discharge valve data. *
An analysis shows that the LPCI System was capable of performing its safety function even though the SA valve was incapable of fully closing as require ZNLD/2170/2
'
- VIOLATION A: (continued)
CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED:
Commonwealth Edison Company's (CECo) Nuclear Engineering Department (NED)
reanalyzed the VOTES trace. A new zero coordinate was established for the trac This new zero marker allowed for a higher torque switch setting than previously analyzed. On August 10, 1991, the torque switch setting was corrected, and subsequently the valve was returned to servic NED has re-analyzed VOTES traces for all six CECo nuclear stations and identified no other similar anomalies of VOTES dat CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATION:
The vendor has trained CECo Station MOV Coordinators to use the new VOTES software. Should any uncertainty in data interpretation exist, the Station MOV Coordinators will contact NED for proper dispositio The Station MOV Coordinator has revised Dresden Maintenance Procedure (DEP)
040-10, "VOTES System Operating Procedure," to include enhanced independent review requirements and thrust window acceptance criteria.
DATE OF FULL COMPLIANCE:
Full compliance was achieved when the MOV 5A torque switch setting was corrected and the valve was returned to servic ZNLD/2170/3
- VIOLATION: B 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstance Contrary to the above, as of May 1, 1992, engineering development and submittal of thrust values to support the motor operated valve testing program at Dresden, an activity affecting quality, was not prescribed by any procedure. Also as of May 1, 1992, no procedure existed to describe corporate engineering required actions when a condition adverse to quality was identified within the motor operated valve progra REASON FOR VIOLATION:
CECO's MOV Program was developed to ensure that all MOVs in safety-related systems will perform design basis functions. CECo's implementation of the requirements of Generic Letter 89-10 is described in CECo's GL 89-10 MOV Program Document. This document describes the necessary procedures,
instructions and administrative controls to support the MOV Program. To facilitate CECo's setting of MOV torque switches, the Corporate MOV Group defines a
"window" within the MOV design limits called a target thrust window. In the MOV Program Document, MOV-WP-107 provides the methodology for generating a target thrust window. MOV-WP-107 was initiated in January 1991. This target thrust window methodology was reviewed by the NRC during the Byron MOV ProQram Inspection and found to be generally consistent with the Intent of the Generic Lette However, MOV-WP-107 did not have any provision for requiring documentation of the assumptions used in generating target thrust windows. Additionally, the MOV Program Document did not contain specific instructions for required actions when a condition adverse to quality was identified within the MOV Progra Generation oftargetthrust windows does not represent a design activity. The
purpose of the target thrust windows is to assist the Station in setting MOVs within desiqn basis limits. Because generation of target thrust windows is not a design activity, the instructions in MOV-WP-107 did not have the level of specificity for
- documentation of assumptions and resolution of conditions adverse to quality
.
contained in other design activities undertaken by CECo's engineering organization.
.. *
ZNLD/2170/4
'*" "
- !
VIOLATION B: (continued)
CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED:
All previously generated tar9.et thrust windows for the six CECo nuclear stations were reviewed and found to be within the specified design limit,
On May 7, 1992, a MOV Group technical guidance document was issued that incorporates the methodology from MOV-WP-107 with the additional requirements for documentation of assumptions when generatin~ target thrust windows. CECo personnel performing target thrust window evaluations have been formally trained on the technical guidance. Since issuance of the technical guidance, target thrust windows have been generated consistent with its requirements. The technical guidance will be incorporated into the next revision of CECO's Nuclear Operations Directive on MOVs, NOD M *
A VOTES Test Evaluation Checklist has been issued to the six CECo nuclear stations for their use when evaluating diagnostic tests on MOVs. The checklist was initially used during MOV testing that was performed at CECo's stations early in 199 CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATION:
The VOTES Test Evaluation Checklist will be incorporated into a ENC Technical Information Document (TIO). In addition to the information that is already contained in the checklist, the TIO will provide extensive guidance to the sites for their use during both static and dynamic testing and will serve as a technical reference document for all the sites for VOTES testing. The draft TIO has been issued and is currently in the review cycle. The final TIO will be issued by February 1, 199.
)
CECo has redefined the corporate MOV Program structure to better facilitate communication and to better define.the responsibilities of the different cognizant parties. AGL 89-10 Project Team has been formed with a Corporate Team Leade.
The Team will allow for more efficient transfer, evaluation, and dissemination of MOV *
test data between the sites and corporate offic DATE OF FULL COMPLIANCE:
Full compliance was achieved with the issuance of the Corporate MOV Group technical guidance documen *
ZNLD/2170/5
VIOLATION: C 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that measures be established to assure that conditions adverse to quality, such as*
deviations and non-conformances, are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of managemen Contrary to the above, on August 20, 1991, corporate engineers identified that the zero points on 17 motor operated valves (MOVs) deviated from their previously selected values, the thrust valves on 4 MOVs did not conform to the thrust windows provided to Dresden station, and the causes and corrective action taken for these non-conformances were not documente.
REASON FOR VIOLATION:
Investigations into the failure of the LPCI 2-202-5A valve showed that the MOV thrust setting for the 5A valve was in error due to a previously unrecognized problem with zeroing the VOTES trac After the VOTES testing problem was identified, the Corporate MOV Group reevaluated all 39 MOV VOTES tests that were performed during the Fall 1990 refueling outage at Dresden. All 39 MOV's thrust settings, except for the SA valve, were found to be within design limits and were acceptable. During the review process, the engineers rezeroed 17 of the 39 MOV VOTES traces. Rezeroing means setting a new zero reference point on the data trace. The zero reference point is used to evaluate MOV thrust output. The rezeroing of 17 VOTES traces did not cause any of the MOVs to have thrust values outside of design limits. Two of the 39 MOVs had as-left thrust values o.utside of the established target thrust windows; however, these MOVs were still set within the design limits. The engineers determined that no deficiencies existed for the 39 MOVs, except for the SA valve, and that no conditions adverse to quality existe CE Co acknowledges that. the corporate engineering review of the MOV tests performed by Dresden was not adequately documented. Also, no overall procedure existed within CECo's engineering organization (ENC) for disposltioning conditions adverse to quality. Individual programs and a~lvities within the engineering
- organization are covered under procedures and instructions which provide specific requirements for identification and resolution of conditions adverse to qualit ZNLD/2170/6
. VIOLATION C: (continued)
CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED:
CECo's engineering organization (ENC) is Implementing an Integrated ReportinQ Program (IRP) to address conditions adverse to quality. ENC will utilize the station's IRP process as the vehicle to address conditions adverse to quality. The goal of the program is to ensure that identified problems are documented, evaluated, and resolved in a timely manner with appropriate management review based on the safety SiQnificance of the issue. The ENC IRP will also provide a mechanism for tracking issues and ensure that required station and NRC notifications are performed. Implementation of the ENC IRP is ongoing in phases. Full implementation will be completed by December 31, 199 Pending implementation of IRP, any MOV with an as-left setting outside of the original target thrust window will be documented in the station Nuclear Tracking System for dispositio.
CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATION:
No further action beyond implementation of IRP is planne DATE OF FULL COMPLIANCE:
Full compliance will be achieved with implementation of the IR ZNLD/2170/7
- VIOLATION: D 10 CFR Part 21.21 (a)(1 )(i) requires, in part, that each individual, corporation, or other entity subject to the regulations In this part adopt appropriate procedures to provide for evaluating deviation *
Contrary to the above, as of February 5, 1992, the licensee failed to adopt appropriate procedures to provide for evaluating deviations. Specifically, Dresden Administrative Procedure OAP 2-8, "Deviations," did not provide sufficient guidance to ensure the evaluation of deviations involving software programs, methodologies, and trainin *
REASON FOR THE VIOLATION:.
Three separate screenings for Part 21 applicability were performed on the reactor recirculation valve failure. In August 1991, the Operating Engineer initially screened the valve failure to close, and concluded a Part 21 evaluation was not required. The LEA was forwarded to the Assistant Technical Staff Supervisor who completed a reportability screening of the event and concluded a Part 21 evaluation was not required. Additionally, in September 1991, the Ori~Site Review Committee reviewed the event investigation and proposed corrective actions. Their investigation concluded the incorrect torque switch setting was the result of an inappropriate zeroing of the VOTES trace and that a Part 21 evaluation was not require The need for Part 21 notification was not recognized by the reviewers due to a deficiency in OAP 2-8, "Deviation Reporting." The OAP provided guidance on the identification of hardware related issues but lacked adequate guidance to allow site personnel to properly identify software, methodology or training Part 21 issues. *
CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED:
Dresden performed a Part 21 evaluation and concluded that a defect existed associated with VOTES training and issued a Part 21 notificatio OAP 2-8, "Deviation Reporting," has been revised to include specific guidance on the identification of non-hardware defect In May 1992, the Technical Staff.Supervisor issued a letter to all Onsite Review Participants clarifying Part 21 reporting requirements, specifically covering the requirement to report non-hardware defect ZNLD/2170/8
VIOLATION D: (continued)
CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATION:
The Onsite Review Reference Manual that is used by Onsite Review Participant was revised to include detailed information of Part 21 reportabilit Dresden Training Department developed training materials on this event and Part 21 reporting criteria. Training of appropriate personnel was completed by July 31, 199 To ensure corrective action effectiveness, the Corporate Part 21 Coordinator will review Dresden Deviation Reports issued from August to December 1992 for Part 21 applicability. The review will be completed and a report will be issued by January 31, 199 DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:
Full compliance was achieved with the issuance of the revision to OAP 2-ZNLD/2170/9
- VIOLATION: E 1 O CFR 50, Appendix 8, Criterion V requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. and shall be accomplished in accordance with these instructions, procedures, or drawing.
Dresden Administrative Procedure (OAP) 9-11, "Procedure Usage and Adherence," Revision 3, step C.2.e(3), requires, in part, that if unexpected responses occur then certain applicable information will be documented on a procedure comment supplement, Form 9-ll OAP 9-11, Revision 3, step C.5.o(4), requires, in part, that if other than direct observation is utilized, then The initials of the person performing the ob$ervation must be included with the initials of the person actually performing the ste OAP 7-14, "Control and Criteria for Locked Equipment and Valves," Revision 2, step 8.3.a, requires, in part, that if plant conditions require a locked valve to be positioned in a manner other than that indicated on the locked equipment checklist, the valve may be unlocked and repositioned either by an approved procedure or an outage checklist. If a valve is to be unlocked without a corresponding procedure or an outage checklist an operator is required to be in continuous attendanc Contrary to the above: On March 7, 1992, during performance of Dresden Operating Surveillance (DOS) 6600-03, the unit 2/3 emergency diesel generator failed to transfer to unit 3 power when expected. Form 9-fla, Procedural Comment Supplement, was not completed in accordance with OAP 9-11, Revision 3, step C.2.e(3), although the unexpected system response required it to be use On March 7, 1992, during performance of DOS 6600-03, the test leader failed to document the initials of the individuals actually performing the surveillance steps as required by OAP 9-11, Revision 3, step C.5.o(4).
- On March 20, 1992, the requirements established In OAP 7-14, Revision 2, step 8.3.a, were not implemented when the standby. liquid control storage tank air
sparge inlet valve was opened and unlocked. Personnel did not use an approved procedure or outage checklist, and an operator was not in continuous
attendance. *
ZNLD/2170/10
\\
- VIOLATION E: (continued)
REASON FOR THE VIOLATION:
Dresden Station acknowledges that personnel did not adhere to the administrative requirements due to the lack of awareness of the appropriate procedures. Past training on new and revised administrative procedures was inadequat CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED:
OAP awareness training was provided based upon the initial training matrix to station personne *
To enhance overall awareness of administrative procedures, Dresden rnanagement has validated and controlled a matrix of administrative requirements for which each station position, both management and bargaining group is responsible. Station personnel have been provided with a matrix of DAPs for which they are responsibl Requirements for a periodic review of the required DAPs by station personnel have been established. Additionally, a process has been developed to ensure that revisions to DAPs are evaluated for identification of necessary training with respect to that revisio CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATION:
To ensure the effectiveness of the corrective actions for this violation, NuClear Quality Verification will perform an effectiveness review by December 31, 199 DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:
Full compliance was achieved with the completion of OAP awareness training..
ZNLD/2170/11
VIOLATION: F 10 CFR 50. 72(b)(2)(ii) requires, in part, that the licensee notify the NRC as soon as practical and in all cases, within four hours of the occurrence of any event or condition that results in.manual or automatic actuation of any engineered safety feature (ESF).
Contrary to the above: On March 14, 1992, the high pressure coolant injection (HPCI) suction valve unexpectedly opened during the unit 3 integrated leak rate test when high drywall pressure provided an ESF actuation signal to the HPCI syste Shift operations management failed to recognize the valve opening as an unplanned
~ ESF actuation, and did not report it until March 18, 199.
On April 19, 1992, at 6:40 p.m., the low pressure coolant injection (LPCI)
minimum flow valve unexpectedly closed twice during the performance of surveillance DOS 1500-1 on unit 3. Shift operations management did not.
- recognize the closings of the minimum flow valve as an unplanned ESF actuation, and did not report the closures until April 20, 1992, at 10:16 REASON FOR THE VIOLATION:
Dresden acknowledges that ENS notifications were untimely. Plant personnel experienced difficulty when determining if an event warranted notificatio Specifically, for the March 14, 1992, event the Shift Engineer did not consider the event re(l<?rtable because the operation of the motor operated valve was not spurious and the intended function was accomplished, that is, the valve went open. On April 19, 1992, the Shift Engineer did not consider the event to be reportable because only the minimum flow valve closed and the Shift engineer thought that the LPCI suction and discharge valves were only the ESF components in that syst~m.
.. CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED:
A detailed guidance document on reportability (Tenera Reportable Event Decision System (TR EDS)) was procured and customized for CECo's operation.. TREDS consists of flowcharts to guide the operator through various types of events and direct the operator to more detailed guidance information to assist in making a reportability determination for a particular event. For ESFs this guidance includes recognition of an ESF event, clarification of preplanned evolutions, a:nd when ESF systems/components are properly removed from servic TREDS has been incorporated into a controlled CECo Reportability Manual providing *
identical guidance to all six nuclear stations. This manual has been issued for
implementation at the six CECo sites by December 31, 1992...
A lesson plan on the use of the Reportability Manual has been developed and
. training was_ conq_!Jcted at Dresden. Full implementation at Dresden was completed by August 31, 199 *
ZNLD/2170/12
- VIOLATION F: (continued)
CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATION:
In conjunction with the implementation of the CECo Reportability Manual, Dresden has developed a single procedure, OAP 2-28 "Reportability determination and Event Notifications," which outlines the station process for making reportability determinations and notification PATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:
Full compliance was achieved when ENS notifications were made on March 18, 1992, and April 20, 1992.
ZNLD/2170/13
VIOLATION: G OAP 10-02, "10 CFR 50.59 Review Screening and Safety Evaluation," Revision 5,
, step F.l.c(5}, requires in part, that the 1 O CFR 50.59 safety evaluation/screening preparer refer to Checklist 5 (Worksheet) for supplemental help in filling out the 50.59 safety evaluation/screening form. Checklist 5 asks if safety related circl1its are isolated and separated from non-safety related circuit Contrary to the above, Checklist 5 was not used by the preparer on March 19, 1992, when performing a safety evaluation for the installation of measuring and test equipment under a temporary alteration to monitor voltage on the auxiliary compartment of ESF 4160 V AC bus 34-1. The temporary alteration provided an indirect interface between Class 1 E electrical equipment and non-safety measuring and test equipmen REASON FOR THE VIOLATION:
In March 1992, a chart recorder was installed*on the non-safety related po.rtion of the ESF Bus 34-1 to monitor voltage performance. The chart recorder was connected to a circuit that contained a fuse, thus protecting the safety related side of the circui However, the appropriate 50.59 che~klist was not use Individuals involved were not aware of and did not review Checklist 5 of OAP 10-2
"10 CFR 50.59 Safety Evaluation/Screening Worksheets Electrical Issues," when performing the safety evaluation. A part of these worksheets asks if safety related circuits are isolated and separated from non-safety related circuit Contributing to this event was the fact that the training provided on the OAP 10-2 requirement was not sufficient to ensure adequate awareness of the checklist requirement.
.
Additionally, OAP 10-2 was deficient in that cross reference to the checklists was not included on the 50.59 form to prompt the prepare CORRECTIVE ACTIONS TAKEN AND RESULTS ACHIEVED:
The Dresden Station Technical Staff Supervisor distributed a memo to all 1 O CF.59 Safety Evaluation Screeners and Evaluators referencing this event and the need to use the Safety Evaluation/ Screening Worksheets when screening or reviewing safety evaluation..
- *
ZNLD/2170/14
VIOLATION: G (continued)
CORRECTIVE ACTIONS TAKEN TO AVOID FURTHER VIOLATION:
OAP 10-2 was revised to require safety evaluation screeners and evaluators to document their screening/evaluation of plant design changes on a checklist. This
.
- checklist requires the user to provide a written negative confirmation of design issues.
included on the worksheet Appropriate training on OAP 10-2 was provided by the Technical Staff Supervisor to personnel who would be responsible for screening and approving 1 O CFR 50.59 evaluation DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:
Full compliance was achieved with the issuance of a memo which reinforced the use of the Safety Evaluation/Screening Workshee \\
ZNLD/2170/15
ATTACHMENT B RESPONSE TO NOTICE OF VIOLATION NRC INSPECTION REPORT 50-237192009; 50-249/92009 As requested in the reference cover letter, the following provides additional information pertaining to the status of the Integrated Reporting Program and our effectiveness review of prior corrective action INTEGRATED REPORTING PROGRAM*
The Integrated Reporting Program (IRP) is controlled by Dresden Administrative
.
Procedure OAP 02-27, "Integrated Reporting Process." IRP was implemented on August 19, 1992, using corporate guidance, as well as the lessons learned from Zion and Braidwood's experience with the program. To date, three other processes have been incorporated into IRP. These are Deviation Reports (DVRs), Radiation Occurrence Report (RORs) and Personnel Contamination Events (PCE).
Prior to implementation, the following training was conducted for site personnel:
Station tailgates along with security gate house handouts were used to inform the station about the IRP progra.
An all-station presentation was conducted on August 19, 1992, with emphasis on an overview of the IRP process and how each individual is involved in the proces Special focus was placed on the individual initiating a Problem Identification Form (PIF) when a problem Is identified. A follow-up presentation is currently being schedule Thirty minute briefs on the IRP process were given to individual work groups to aid in their understanding of the IRP proces Licensed Shift personnel received IRP training during their continual operator training progra IRP uses a Severity Level Matrix to classify problems from the lowest (Level 4) to highest (Level 1) impact. Formal root cause analysis techniques are used to investigate severity level 1 through 3 events by formally trained personnel. Thus far, 97 PIFs have been initiated of which 13 have been assigned to the Level 3 category. The remaining 84 are Level 4 items, which are evaluated for an approximate root caus Our current plan is to trend Causal Factors using a keyword index, and to trend over time. The more significant event/problems will be trended with the intent of identifying global causes and responses. Level 4 items will be trended to identify issues that may require a full root cause investigation. We estimate that there will be sufficient information in the database to perform meaningful trending by the end of the 1 st Quarter 199.
To ensure ~ffectiveness.of the.program ir:nplt:t111entation.a c;:orpo~~te rev.iew '!f the IRP program will be conducted dunng October 1992. A station effectiveness review program for IRP will be implemented by the end of the 1st Quarter 19.9 ZNLD/2170/16
AlTACHMENT B (continued)
CORRECTIVE ACTIONS EFFECTIVENESS REVIEW The Quality Programs and Assessment Group has reviewed the actions and effectiveness of past corrective actions as they relate to NRC violations from June 1991 to May 1992. Corrective actions from a total of 18 cited and 8 non-cited violations were reviewe Effectiveness of the corrective actions was evaluated using (1) direct followup by observation of the activities and discussions with effected personnel, (2) review of Field Monitoring Reports, and (3) review of event reports (Deviation Reports, Licensee Event Reports).
- The evaluation criteria applied to the effectiveness of the corrective actions was (1)
timeliness of corrective action implementation, (2) personnel awareness of the violation and the committed corrected actions and (3) evidence of event recurrenc This review concluded:
Some commitment dates were exceeded early on, but performance has improve Affected personnel were aware of specific actions taken in response to the violatio Field Monitoring Reports and direct observation did not indicate evidence of potential repeat violation The scope of corrective actions could be expanded in subparts of 5 of the 26 violations reviewe A final report of the corrective actions effectiveness review has been issued to upper station management. This report included specifics for each violation reviewed and recommendations that may further enhance the program enhancement Finally, Dresden will continue to assess the effectiveness of its corrective actions program through IRP trendin *
ZNLD/2170/17