IR 05000341/1988006
| ML20197F246 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/26/1988 |
| From: | Cooper R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20197F236 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-341-88-06, 50-341-88-6, NUDOCS 8806100250 | |
| Download: ML20197F246 (22) | |
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U.S. NUCLEAR REGULATORY COMMISSION .
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REGION III
Report No. 50-341/88006(DRP)
Docket No. 50-341 Operating License No. _NPF-43 Licensee: Detroit Edison Company 1
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2000 Second Avenue Detroit, MI 48226 Facility Name: Fermi 2 Inspection At: Fermi Site, Newport, Michigan Inspection Conducted: February 6, 1988 to March 31, 1988 Inspectors: W. G. Rogers M. E. Parker P. R. Pelke T. D. Reidinger S. Stasek G. M. Nejfelt Approved By:
CP C $dh R. Cooper, Chief ggfgg '
Projects Section 3B Date Inspection Summary Inspection on February 6, 1988 to March 31, 1988 (Report No. 50-341/88006(DRP))
Areas Inspected: Action on previous inspection findings; operational safety verification; maintenance; surveillance; LLRT observation; LER followup; safety evaluation; NUREG 0737 actions; Nuclear Operations Improvement ;
Program; followup of events; startup test preparation; regional requests; allegations and management meeting Results: Five violations were identified (Paragraphs 3, 4, 6, and 9).
One unresolved item was identified (Paragraph 11),
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DETAILS
. Persons Contacted Detroit Edison Company
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F. Abramson, Operations Engineer
- P. Anthony, Licensing
- S. Catola, Vice President, Nuclear Engineering and Services
- D. Gipson, Plant Manager
- L. Goodman, Licensing J. Green, Systems Engineering R. Kelm, Director, Nuclear Security J. Leman, Director, Plant Safety, Nuclear Production
- R. Lenart, General Director, Nuclear Engineering
- L. Lessor, Advisor to Plant Manager R. May, Superintendent, Maintenance and Modification G. Ohlemacher, Principal Engineer, Licensing
- W. Orser, Vice President, Nuclear Operations / Plant Manager J. Pendergast, Compliance Engineer J. Plona, Operations Support Engineer E. Preston, Assistant Director, Plant Safety T. Randazzo, Director, Regulatory Af fairs B. Sheffel, Nuclear Production, Technical Engineering ISI
- F. Svetkovich, Technical Engineer, Nuclear Production
- B. R. Sylvia, Group Vice President, Nuclear Operations
- G. Overbeck, Director Nuclear Training
- J. DuBay, Executive Assistant
- R. Stafford, Director, Quality Assurance ;
- W. Tucker, Superintendent, Operations U.S. Nuclear Regulatory Commission
- M. Parker, Resident Inspector
- W. Rogers, Senior Resident Inspector
- P. Pelke, Project Inspector
- T. Quay, NRR/ LPM
- H. Walker, Reactor Inspector (DRS)
- R. Knop, Chief, Projects Branch 3
- R. Cooper, Chief, Projects Section 3B
- A. Davis, Regional Administrator
- M. Virgilio, Deputy Division Director (DRP)
H. Richings, NRR/I&C R. Emch, NRR/ Technical Specifications
W. Brooks, NRR/ Reactor Systems J. Mauck, NRR/ Technical Specifications
- Denotes those attending the exit meeting on March.30, 198 # Denotes those attending the management meeting on March 29, 198 Denotes those attending the management meeting on February 17, 198 The inspectors also interviewed others of the licensee's staff during this inspec+1o . . . .
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. Action on Previous Inspection Findings (92701)
. (Closed) Open Item (341/86026-08): Reactor Engineering Documentation of Events Evaluation. The licensee has provided additional guidance on what is requ',r'ed from the cusnift reactor engineer in terms of documentation. Sections 6.1.2 and 6.1.11 of Procedure 51.000.10 provide the guidance. Alse, the reactor engineering log has been formalized. This matter is considered ~
close (0 pen) Unresolved Item (341/87026-05): UF3AR accuracy. The licensee has committed to review all safety evaluations to ensure that they were incorporated into + 2 UFSAR by August 31, 198 (0 pen) Violation (341/87038-01): Failure to apply valve locking devices. This item was previously reviewed in Report No. 50-341/8704 The licensee conducted another complete set of walkdowns and included an independent verification. Systems vital to the safety of the plant were walked down in their entirety with exception of valves inside the Drywell and High Radiation Areas. Results of the second walkiown include All valves requiring locking devices in accordance with POM 21.000,14 had the correct locking mechanism installed except P41-F111 which had a lead seal installed instead of a lock and chain specified in 21.000.14. A lock and chain were subsequently installe There were no valves with locking devices installed that were
.iot required to have them installed either by P0H 21.000.14 or by their respective system operating ,r;-ocedures (SOPS).
Valves with missing label plates are being processed in accordance with Plant Order EFP-104 A work request was written for valves missing handwheel j DER 87-536 was written for two valves which were in the field !
but not on the drawings (3/4 inch valve / test connections).
The valve line-ups in several SOPS equire updating because I they do not indicate the type of locking device to be used on the locked valve in the syste Several functional operating sketches require updating to l indicate the locked valve statu Surveillance Procedure 24.203.07 contains a valve verification sheet on which two Core Spray valves are locked open; however, no locking requirement exists in the locked valve guidelines or the SO Many valves within the SOPS are required to be locked, but do not require locking per the locked valve guideline .
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The inspector reviewed the completed checklists used to perform the
. second set of walkdowns. No problems were identified. During the current LLRT outage, the licensee will walk down systems which were previously inaccessible. This item will remain open penaing completion of these walkdowns plus further walkdowns by the inspecto (Closed) Open Item (341/87050-01): Unfiltered leakage criteri This matter was reviewed by NRR and was considered acceptable. The memo dated February 18, 1988, from Theodore. Quay to Charles Norelius documents this revie No violations or deviations were identifie . Operational Safety Verification (71707)
The inspectors observed control room operations, reviewed applicable logs and conducted discussions with centrol room operators during the perio The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the reactor building and turbine building were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenanc The inspectors, by observation and direct interview, verified that the I physical security plan was being implemented in accordance with the station security pla ,
The inspectors observed plant housekeeping / cleanliness conditions and i verified implementation of radiation protection controls. During the i inspection, the inspectors walked down the accessible portions of the l Standby Liquid Control and High Pressure Coolant Injection Systems to '
verify operability by comparing system lineup with plant drawings, as-built configuration or present valve lineup lists; observing equipment conditions that could degrade perfo"mance; and verified that instrumentation was properly valved, functioning, and calibrate These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under Technical Specifications,10 CFR, and administrative procedure During these operational observations the inspectors ascertained: That on February 24, 1988, shift personnel reacted properly to a potential fire in the controlled air compressor room. It was later determined that there was no fir l That licensed operators were not performing Technical Specification action statement 3.5.1.e properly. On February 4, 1988, one of the core spray system differential pressure instrument channels (621-N004A) was providing a reading different than the other channel. The channel was declared inoperable, out-of-specification log entry 88-0109 initiated, and action 3.5.1.e designated as
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applicabl The action statement requires the instrument channel be
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returned to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or determine the differential pressure locally at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Prior to exceeding the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> operators began verifying the differential pressure locally on a situational required surveillance logsheet, Attachment 10 to Procedure 24.000.02, "Shif tly, Daily, Weekly and Situational Requir ed Surveillances." This is the normal mechanism the operators use '.o document such matters. The logsheet specified that the local verification be done every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This was inconsistent with the Operations Engineer's instructions to do the check every eight hour Howe.ver, the operators did not do the local verification within the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> action statement on nine separate occasions between February 7 and 18, 1988. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> requirement was exceeded by 27, 12, 12, 15, 13,1, 23, 30 and 5 minutes on February 8, 9, 10, 11, 12, 14, 15, 16 and 18, respectively. Six different NASS's and eight different reactor operators improperly initialed that this action statement was performed or reviewed satisfactorily. On February 19, IS88, the inspector brought this to the attention of the onshift NAS The NASS changed the periodicity of the check to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and initiated DER 88-0265. Subsequent review of differential pressure instrument E21-N004A revealed that the instrument had always been operable and these different gauge readings between core spray headers had been observed at other BWR's and a General Electric SIL was written on this matter. Also, the rounds operator was checking the differential pressure during normal shif tly rounds even though this was not considered meeting the action statement requirements under the licensee's administrative program. All readings on the local differential pressure instrument were acceptable. Therefore, a ;
Technical Specification Section 3 violation did not occu However, failure to implement the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodicity as stated on the situation required surveillance logsheet is considered a violation (50-341/88006-01(DRP)) of 10 CFR 50, Appendix B, Criterion V for failure to accomplish activities affecting quality in accordance .
with written approved instruction ] On February 8, 1988, the inspector noted that the control room copy of the Emergency Response Information System (ERIS) Users' Manual was Revision 0. The graphical computer presentations in tFis manual were Revision 22 (i.e., Containment Integrity - Bar; and SPDS Overview). However, the actual displays in the Safety Parameter l
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Display System (SPDS), for the equivalent displays, were Revision 2 The issuance and control of manuals is expected to be accoinplished in accordance with NOIP 11.000.131, Section 5.16.1, Section 5.16.1, Revision 4. However, no violation was issued, because the differences that were verified between the Users' Manual and actual computer displays were changes in format onl No other violetions or deviations were identified in this are .
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. Monthly Maintenance Observation (62703)
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Station maintenance activities on safety-related systems and components listed below were observed to ascertain that they were co ned in accordance with approved procedures, regulatory guides ano industry codes or standards and in conformance with Technical Specification The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating tha work; activities were accomplished usir.g approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplisheo by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemente Work requests were res.ewed to determine the status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may effect system performanc The following maintenance activities were observed or reviewed to scme extent:
fire p>netration sealing of CTC-2 for electrical wiring associated with the primary containment monitcring syste calibration and check of post accident recorder B21R623 diesel fire pump alternator replacemen EDG 13 fuse holder replaceman Significant observations / reviews of these maintenance activities were:
, On February 18, 1988, while witnessing the calibration of post accident recorder B21R623A :nder work request B5500108, the <
inspector noted chat two wires had been lifted during performance 1 of the activity but had not been recorded or independently verifie Following completion 'f the calibration the wires were relanded l without indarendent verification. The inspector questioned the I&C technicians as to whether a lifted wire / jumper logsheet should have been used. The I&C personnel responded that no logsheet had be ,
'prov'ded with the wuck package. The inspector reviewed the wort request and noted that the job had been classified as safety-reltte r P0M 12.000.080, Conduct of Electrical Field Activities, Paragraph 7.1..ates in part, ""An Interim Alteration Checklist"
- should be used for documenting the determination and retermination of electrical connections as well as other interia alterations that 1 become necessary as part of-testing, investigation, repair or l replacement . . ." and in the same procedure Paragraph 7.5.1 states, 1
"An independent second check of restoring to normal shall be pe 'ormed 4 l for all interim alteratians performed under work orders designated as J d
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Safety-Related on the Work Order Package Attachment A." Regulatory
. Guide 1.33 requires administrative procedures be established for equipment contro POM 12.000.080 provides assurance that instrumentation is wired properly following a wiring chang Technical Specification 6.8.1 identifies Regulatory Guide 1.33 procedures as required to be established, implemented and maintaine The failure of I&C personnel to properly adhere to Administrative Precedure 12.000.080 is considered a violation (50-341/88006-02(DRP))
of.Technica1 Specification 6. Following completien of naintenant.e on the post accident recorder, the inspectors verified that this system had been returned to service properly, b. The inspector reviewed completed work order 034 B02288 on replacement of a fuse holder in Emergency Diesel Generator No.13's control pane During review the ir.spector noted that Block D, Part 7, for "out of specification log cleared" under the "return to service" portion of the work request had beer signed by someone other than a NASS or NS The block clearly designates that the signature is for the NASS/NS The inspector brought this matter to licensee management attentio The licensee agreed that this was not a correct practice and discussed the matter with the f ndividual involve The inspecter also discussed this matter with the individual and asce-tained:
He did not normelly sign off on work request His intent was to help reduce an administrative burden on the NASS/NS He had previously held a SR0 license.
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He stated that he would never do this agai Given the isolated nature of this situation, the corrective actions are considered sufficient and this matter 's closed, Followup activitic.s associated with diesel fire pump maintenance vere completed during this inspection period. In Inspection Report 50-341/88003, an inspector witnessed the replacement of the alternator on the diesel fire pump. The replacement alternater was
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from a different manufacturer and hac' a differt:nt part numbe However, in the licensee's Spare Parts Reference System (SPRS)
both the old and new alternators had the same stock numbe Nuclear Engineering Procedure 6.12 "Engineering Evaluation of Onsite Material," provides the guidelines for whether components are acceptable for use and how to evaluate the componen Step 7.4.2.1 states that if the material is already evaluated and ,
, approved by Nuclear 2,igineering it :ay be issued without Material '
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Engineering concurrence if the stock item is listed in SPR .
Therefore, the inspector requested the licensae provide the original engineering evaluation which accepted the replacement alternator (3603857-RX) as a suitable replacement for the old alternator (A0012301JB).
Af ter investigation the licensee reported that no evaluation had been performed. The licensee related that in 1985 DEC0 purchasing agents contacted Cummings Diesel, the pump vendor, about spare part Cummings stated that they could supply rebuilt alternator The original alternator was not from Cummings. The purchas?qg agent procured the Cummings alternator and updated a materials management system (MMS) form showing the Cummings alternator (3603857-RX) as a replacement for alternator A0012301J8. MMS is a Detroit Edison wide spare parts inventory system and is not part of the 10 CFR 50, Appendix B, Criterion VIII controls program on quality related equipment. However, at that time any change to MMS automatically changed SPRS without engineering review i'or components other than Therefore, in SPRS for stock number 453-9200, two part numbers were disp'aye On January 8, 988, maintenance personnel initiated a Request on Stores (ROS) for stock number 453-9200 tc perform maintenance activities. The 3603857-RX alternator was provided for the maintenance. When the ROS was reviewed by the Material Engineering Group, it was only to verify that 3603857-RX matched stock number 453-9200 in SPR Once this situation was identified, Materials Engineering personnel performed an engineering evaluation and found the new alternator (3603857-RX) acceptable on Febrt'3ry 11,198 The next day DER 88-0239 was written regarding this situation. The failure of the licensee to procerly maintain SPRS is a violation (50-341/88096-03(DRP)) of 10 CFR 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts and Component No other violations or deviations were identified in this are . Monthly Survei!1ance Observation (61726)
Tht inspectors observed surveillance testing on Fuel Pool Ventilation Exhaust Radiation Monitor, Division 2, Channel D per Procedure 64.020.108
.equired by Technical Specifications and verified that: testing was performed in accorcance with adequate procedures, test instrumentation was calibrated, limiting conditions for operation were met, removal and restoration of the affected components were accomplished, test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the indivb.aal directing the test, and any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The inspectors also witnessed portions of the following test activities:
Quarterly intrusion detection testinn on March 3, 198 _ _ _ - _ - _ _ _ _ - _ _ - _ - - _ _ _ _ _ _ _ _ _
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Procedure 42.302.04, Calibration and Logic-System Functional Test of l
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Procedure 44.010.109, IRM Functional Test Division I Bartor calibration on steam flow channel K8150 using calibration procedure C-32-N003D on March 17, 198 Procedure 44.160.001, Fire Detection Operability and Functional _ Test for zones 4 and Nu violations or deviations were identified in this are . Observation of Local Leakage Rate Testing (LLRT)
The inspector witnessed Type B and Type C locat leak rate testing (LLRT)
of the following containment isolation devices during the inspection period and verified that each test was conducted in accordance with approved procedures, test equipment was properly installed and calibrated, personnel conducting the tests were sufficiently t.nowledgeable of test methodology, and test results were properly reviewed against established acceptance criteri f Type B Test Component Containment penetra*'.on N Flange X-21CA Type C Tests Valve Containment Penetration N E51-F019 X-227A P44-F282B X-34A E11-F016B X-390 E11-F001B X-210A The inspector reviewed the following test procedures associated with the LLRTs witnessed, and verified that each appeared adequate to properly control those specific test .401.200 local Leak Rate Test, Type B- Ceneral 43.401.218 Local Leak Rate Testing For Relief Valve Flanges 43.401.300 Local Leak Rate Test, Type C- General 43.401.335 Loccl Leak Rate Testing For Panetration X-34A 43.401.420 Local Leak Rate Testing For Penetration X-227A '
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43.401.345 Local Leak Rate Testing for Penetration X-39B 43.401.404 Local Leak Rate Testing for Penetration X-210A
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The inspector independently calculated leak rates for the subject
. valves using the data acquired from each test and compared each result to that obtained by the tes: engineer as well as to the specified acceptance criteri '
The results of the type B test performed for the flange of penetration -
X-210A indicated unacceptable leakage. The licensee indicated further troubleshooting and possible rework of the flange prior to reperformance of a LLRT would be require The type C tests for penetration X-227A, X-34A and X-210A exhibited ,
leak rates well within acceptable limit Valve E11-F016B at penetration X-398 failed its type C tes The other valve, E11-F021B, in the penetration previously failed its test. Both valves have been scheduled for rewor During the water test of valve E11-F001B at penetration X-210A on March 17, 1988, the inspector noted that test personnel were using their own personal watches to calculate the actual measured leakage in ml/ mi The acceptance criterion for total maximum ~ leakage was 568 ml/ min. In this particular case, the watch was analog with no second hand. In response to the inspector's concern, the licensee immediately obtained calibrated stop watches for test personnel in the field. Additionally, the use of calibrated stop watches was not procedurally prescribed. The licensee committed to revise appropriate procedures and to review previous LLRT tests where non-calibrated watches were use Failure of the'LLRT test procedure to include provisions to assure that adequate test instrumentation was available and used is a violation of 10 CFR 50, Appendix B, Criterion XII (50-341/88006-04(DRP)). Licensee Event Reports Followup (92700)
- Through direct observations, discussions with licensee personnel, and
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review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective ,
action was accomplished, and corrective action to prevent recerrence hao been accomplished in accordance with Technical Specification (Closed) LERs 87-051-00, 87-051-01, and 87-051-02: Differential Pressure Actuation of the Emergency Equipment Cooling Water (EECW) system due to a design deficiency. On September 29, 1987, Division I of the EECW.and Emergency Equipment Service Water (EESW) systems actuated on a low differential pressure signal in the Reactor Building Closed Cooling Water (RBCCW) system. This occurred when a third RBCCW pump was put into service in preparation for switching pump The RBCCW has three 50% capacity pumps (two running and one idle). In order to perform pump mairtenance and to equalize run times for the three pumps, the pumps are periodically rotated, the idle pump started and one of the previously running pumps stopped. During pump rotation a low differential oressure transient occurs (approximately six seconds) which causes unnecessary actuatiuns * EEC _ - - _ - _ _ _ _ _ - _ - _ __ - _-_ _-__ _ _ _ _ _ _ _ _ _ _ _ _ _ __ - ____ _ __ _ _ _ _ = _ - - - _ _ - _ _ _
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The licensee implemented minor modification. (PDC-7701) to change the time
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settings for agastat time delay relays P44M805A, P44M806A, P44M807A, P44M805B, P44M8068, and P448078 from 1.5 to 15 second Sequence of Events Test SOE No. P4400-87-03 was developed to test the new 15 second
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time delay. However, during performance of the test on November 11, 1987, EECW Divisions I and II automatically initiated during RBCCW pump rotation. The licensee determined that PDC-7701 did not take into account a. logic seal-in citcuit located between the differential pressure switches and the agastat relays. EDP-8126 was then issued on December 12, 1987, to place 11 second time delay relays between the differential pressure switches and the real-in logic. Work Request PN21 001A1107 implemented EDP-8'2 SOE Nc l'4400-87-04 successfully tested the new logic and was completed on December 29, 1987. All three RBCCW pumps were rotated without automatic initiation of-EEC During this inspection period the inspectors reviewed the following deviation report:
DER 88-0271 Configuration control on ATWS time delay setting The inspector confirmed through record review that the ATWS relays had the appropriate time delay setting No violations or deviations were identified in this are . Safety Evaluations During the inspection period the inspectors reviewed select safety evaluations performed by the licensee. The safety evaluations were:
88-041 Alternate decay heat methods88-003 RCIC operability with valve E51-F095 out of service No violations or deviations were identified in this are l 9. TMI Item Review (Closed) TMI Item I.B.1.2: Organization and Management. NRR has acc pted the licensee's organization in SSER-3 (Pages 22-2 and 22-3)
and through the issuance of Amendment No.11 to Operating License No. NPF-43 dated October 22, 1987. The inspector reviewed
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Figure 6.2.1-1 (Offsite Organization) and Figure 6.2.2-1 (Unit' l Organization) of the Technical Specifications and Section 13.1 of i the UFSAR. The inspector determined that the actual onsite ;
organization is not the same as that specified in the Technical Specifications. Examples ici de: l Placement of the Independent Safety Engineering Group under the Nuclear Safety Review Group Chairman (previously under the Director Nuclear Quality Assurance).
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Establishment of the Director Nuclear Training (previously the
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two positions of Director Nuclear Training-General and Director Operator-Training).
The General Supervisor Nuclear Fuel reports directly to the Vice President Nuclear Engineering and Services (presiously reported to the Director Nuclear Engineering).
Reorganization under the Director Nuclear Quality Assuranc Director of Nuclear Security position now exists under the ,
Plant Manager (previously Superintendent Services),
i Elimination of the Assistant Operations Engineer from the Unit Organizatio The inspector determined that the licensee has not submitted a change request to NRR to amend the Administrative Controls Section of the Technicel Specification This is a violation of Technical Specifications 5.2.1 and 6.2.2 (50-341/88006-05(DRP)). i
' (0 pen) TM1 Item II.K.3.28: Verify Qualification of Accumulators on Automatic Depressurization System (ADS) Valve NRR's review and acceptance of the functional capability of the accumulators for the ADS valves is discussed in the following documents:
i Chapter 22 of the SER '
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l NRR to Detroit Edison letter dated December 5, 1986
Detroit Edison to NRR letter dated February 6,1987 NRR to Detroit Edison letter dated April 6, 1987 Detroit Edison to NRR letter-dated May 26, 1987 ;
j NRR to Detroit Edison letter dated February 22, 1988 la the SER conclusion transmitted by the February 22, 1988 letter, NRR concluded that the licensee has adequately verified. functional capability of the accumulators for the Fermi-2 ADS valves and thereby satisfies the requirements of TMI Item II.K.3.2 Each of the five ADS safety relief valves (SRVs) are supplied with i an accumulator and check valve arrangemen Each accumulator l receives pneumatic pressure from Division I of the Primary i
, Containment Pneumatic Supply (PCPS) System which is normally fed !
from the nitrogen supply system. Division I of the Non-Interruptable Control Air System (NIAS) is available as a backup {
supply for Division I of the PCPS System. Additionally, each division of PCPS has a qualified connection outside of the secondary containraent to permit bottled nitrogen to be supplied as an additional backup, 12 l
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Technical Specification 4.5.1.d.1 requires the performance of a
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channel functional test of the PCPS Icw pressure alarm system at least once per 31 days. This requirement is implemented by-POM 44.030.011. The inspector reviewed completed surveillance packages for September 2,1987 and October 3,1987 to verify implementation of P0M 44.030.01 Technical Specification 4.5.1.d.2.c requires performance of a channel calibration of the PCPS low pressure alarm system and verification of an alarm setpoint of 80 5 psi on decreasing pressure at least once per 18 months. This requirement is implemented by POM 44.030.012. The inspector reviewed documentation for the July 25,1986 completion of this surveillanc Operation of the PCPS is covered in P0M 23.406, "Prinary Containment Nitrogen Inerting and Purge System." This item will remain open pending completion of a PCPS walkdown by the inspecto . Review of the Nuclear Operations Improvement Plan (NOIP)
The NOIP is documented in DECO letters dated May 9,1986, and April 22, 1987. The incomplete and ongoing NOIP objectives have been incorporated into the annual Fermi Business Plans. A review of select NOIP objectives follows.
. NOIP 11 - Include NOIP objectives in Annual Work Plans of Nuclear Operations Management Team mambers as part of individual performance appraisal process. The inspector verified that each NOIP goal was assigned to an individual or individuals through their Annual Work Plan The lowest tier Annual Work plan to which ar, individual NOIP goal was delegated was a function of the nature of the objective and the o*ganization to which it was assigned. Implementation of this NOIP objective will be further assessed as part of the assessment of other
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NOIP 14 - Develop and implement a scram reduction program utilizing techniques found successful by INPO and other industry groups. The May 9,1986 DECO letter established a December 1986 due date for this objectiv In 1986, this objective was assigned via the Annual Work Plan I of the Plant Manager and delegated down three organizational tiers. The 1987-1991 Fermi Business Plan stated that the scram reduction program was to be developed and implemented by May 1987. A DECO internal memorandem dated June 11, 1987, describes various elements to be incorporated into
the Scram Reduc. tion Program. The memorandum further stated that a plar,'.
procedure giving an overall program description and formally documenting the elements would be completed by July 30, 1987. As of February 1988, .
the scram reduction procedure war in draft form. The 1988-1992 Technical l Engineer Business Plan continues to carry the scram reduction program I with a commitment to refine the program by December 1988. However, this objective is no longer carried in the 1988-1992 Fermi Business Plan. The
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inspector was also informed that the scram reduction program will be a subset of the plant performance program which is under development. The f
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scram reduction program will be reviewed in future inspections to verify
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. procedure implementation and the incorporation of INPO guidelines, BWR
- Owners Group recommendations, and interfaces with the plant testability progra . Followup of Events (93702)
During the inspection period, the licensee experienced several events, some of which required prompt notification of. the NRC pursuant to 10 CFR 50.72. The inspectors pursued the events onsite with licensee and/or other NRC officials. In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the licensee was taking prompt and appropriate actions, that activities were conducted within regulatory requirements, and that corrective actions would prevent future recurrence. The specific events are as follows:
February 14, 1988 CCHVAC shifted to recirculation mode while troubleshooting a rad;ation monito * February 23, 1988 HPCI steamline isolation during surveillance testin * February 26, 1988 All emergency diesel generators declared inoperable and an Unusual Event declare February 28, 1988 Five gallon sulfuric acid spil *
February 29, 2988 Valve failures during local leak rate testing reveal leakage in excess of 0.6 L * February 29, 1988 Loss of shutdowr, cooling twice during RPS bus transfer * March 10, 1988 Moisture Separator Reheater (MSR) Damag *
March 11, 1988 Six of ten safety relief valves failed testing on'their setpoin *
March 11, 1988 CCHVAC shifted to recirculation mode when a radiation monitor failed upscale during maintenance activitie March 13, 1988 Failure of two Hydraulic Control Units (HCUs) to depressuriz )
- March 15, 1988 Missed shiftly surveillance tes * (
March 17, 1988 Containment pressure transmitter C71-N050C found isolate March 18, 1988 Bomb Threa _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _
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March 18, 1988 Loss of shutdown coolin .
March 20, 1988 Automatic start of Division 1 emergency diesel generators following core spray logic functional testin The more significant activities or insptetor findings were: On March 24, 1988, the inspector witnessed setup and preliminary performance of a special test on Control Rod Drive System hydraulic Control Units (HCUs) 10-11 and 34-51. The licensee was conducting the test to determine the root cause for these HCU accumulators not properly discharging during a scram which occurred on March-13t It was observed that prestaging for the test' appeared less than adequate to assure the testing was accomplished in the most efficient manner possible. Examples included equipment required for the performance of the test, such as special pipe fittings and electrical power cords not being readily available. The time expended to obtain the required equipment once the test was underway resulted in significant delays in the conduct of the test. Also, when procedural steps were reached in the test requiring the opening of HCU piping to connect a hose, the test was temporarily halted so workers involved could leave the job site, proceed to the Radiation Protection office, and sign-in on the appropriate job-specific Radiation Work Permit (RWP). After the appropriate hoses were connected to the system, a verification was to be made ensuring the HCU accumulator had been draine However, when the accumulator 2 isolation valve was opened, a continuous trickle of- water was observed. Since this was not foreseen in the procedure, testing j
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was then terminated while a temporary procedure change was develope Further testing will be conducted during the next inspection period, On March 17, 1988, while performing valve lineups for local leak *
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rate testing, the rack valve for a transmitter was found closed with '
no apparent reason for it being closed. The licensee is presently A
conducting an investigation as to how and when the valve was close This matter is considered an Unresolved Item (50-341/88006-06(DRP)). I
! On March 18, 1988, the licensee received a bomb threat. All i unessential personnel were evacuated from the protected are i A search was conducted and no bomb was found. All personnel returned to their duty station Following plant shutdown for the LLRT outage, the licensee discovered !
damage to the Moisture Separator Reheaters. The licensee removed the MSR internals by the use of cutti.1g and grinding equipment. The internal structure consisted of primary distribution plates which contained stub tubes fixed transversely over the whole surface of 1
the plate and a double layer of secondary perforated distribution l plates end moisture separator pads which consist of a 6 inch layer !
of stainless steel wire mesh in a stainless steci frame, j!
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During the inspection period, inspectors entered the East and West
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Moisture Separator Reheaters to inspect the remaining piping and support elements. Upon inspection of the remaining horizontal tubes and reheating tubenest, it vias revealed that they were corroded to a degree that it would effect the efficiency.of the conversion of cold reheat steam to hot reheat steam via the reheating tubenes An inspection of the removed internal primary distribution plates-revealed extensive cracking of'the stub tube plates in both a vertical and horizontal plane. The secondary perforated distribution plates appeared to have suffered similar damage as experienced by the primary distribution plates. The moisture separator pads in general suffered minor damage to the first one inch layer of steel wool. At least three of the moisture separator pads support franes experienced frame distortien. A number of the stub tubes are not present on the cracked and deformed plate When the primary distribution plates were damaged, a large number of stub tubes separated from the plates.
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The licensee is performing inspections of the feedwate* system to locate and remove all the loose stub tubes. Some of the stub tubes have been found in the reactor feedwater pump suction strainer Subsequently, discussion between Region III management and licensee senior management took place. Presently, the licensee is preparing a letter discussing their corrective actions to this matte e. Personnel in the Technical Specification Improvement Group identified that a portion of the circuitry associated with starting the emergency diesel generators had not been tested. The personnel initiated a deviation. report and informed the plant manager of the situation. With this information in hand the emergency diesel generators were declared inoperable and a plant shutdown was started from 81% power in accordance with Technical Specification 3.8.1. An Unusual Event was declared by the licensee, as required by their emergency plan. The unit was placed in Cold Shutdown the next day within the time constraints of the appropriate action statemen Later the next day, the Division 1 emergency diesel generator circuitry was satisfactorily teste .
I The portion of the circuitry that had not been tested dealt with intermediate relay contacts from the degraded grid relay. These contacts actuate the electrical breakers to the diesel load bus, the load breakers on that bus, and the EDG start relay. A number l of these components had been tested over the year However, there j are three relay sets (undervoltage, loss of power, degraded grid)
that actuate these components, and each relay set had not been i
i independently tested. This matter will be reviewed more fully when the LER is issue . . . _ _ _ - _ _ _ _ - _ _ _ - _ _ _ _ _ _
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. On March 13, 1988, at 0700, the operating day shift accepted
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control of plant. activities from the night shift and started a scheduled 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift. The unit was in cold shutdown and two weeks into a planned six week outage. During the outage, shift personnel had been placed on 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts for the outage duration. By 0830, surveillance procedure 24.000.02, "Shiftly, Daily, Weekly, and Situational Required Surveillances,"
Attachment 6, was completed by the patrol nuclear supervising operator (NS0). Procedure 24.000.02, Attachment 6, documents the shiftly Technical Specification channel checks for 17 different Technical Specifications which were applicable for the day in question. Attachment 6 is normally performed by the patrol NS0 and reviewed by the nuclear assistant shift supervisor (NASS).
Upon completi(n of the surveillance, the patrol NSO was dispatched to the RHR con ) lex by the NASS for the rest of the shift. The patrol NSO ano the NASS forgot that the shif tly channel checks were due during the afternoon of March 13, 1988, and 24.000.02, Attachment 6, was not performed. The oncoming shift at 1930 satisfactorily completed the channel checks at 0056 on March 14, 198 The missed shift check was identified the next day by the individual who collects the completed surveillance tests for the Technical Staff. No mechanism on the shift assured completion of the surveillance. Once the missed surveillance was identified, a deviation report was written and the matter was brought to station management's attentio >
Failure to perform the afternoon sniftly channel checks by 2330 on March 13, 1988, which was the maximum allowed time interval for performing the surveillances without entering into a Technical Specification action statement is censidered a violation (50-341/88006-07(DRP)) of Technical Specifications 4.4.9.2.1, 4.4.9.2.2, 4.4.9.2.3, 4.5.3.1.b., 4.7.2.a., and 12 other This apparent violation appears to be indicative of a breakdown in the overall appreciation of Technical Specifications as exhibited by successful performance of required surveillance within the specified time limits. These same elements were apparent in the 50-341/87026 violation and as such the corrective actions taken in response to that violation were inadequate to preclude the current violatio No other violations or deviations were identified in this are . Startup Test Preparation
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The inspector attended preparation meetings and evolution walkdowns for completion of the shutdown from outside the control room demonstratio ,
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1 Regional Requests
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During the inspection period, the inspector continued to pursue the regional request dated September 24,~1987, dealing with preventive maintenance activities associated with GE AKF-2-25 field breaker The inspector ascertained: The licensee had scheduled _ the breaker inspection at an "every other refueling outage" frequency instead of "every refueling" as stated in the Information Notic No five year breaker teardown procedure or scheduling requirement exist The breaker inspections are scheduled for the current outage.
, The discrepancy between the licensee and Information Notice requirements was brought to the attention of maintenance personnel, and DER 88-0290 was initirte The inspector will review this matter more fully upon resolution of DE by the license '
14. Allegation Followup
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(Closed) Allegation (RIII-87-A-0099)
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During July 1987, the NRC received several anonymous allegations cencerning ;
Fermi 2 warehousing activities. These concerns were immediately reviewed by the Resident Inspectors and subsequently followed-up by the Project Inspecto '
J Concern No. 1 *
Procedures in use at the Radiological Controlled Area (RCA) third floor
issue station were out of date. These included POM 12.000,62, Revision 8, in use, but Revision 9 was issued February 3,1987; POM 66.000.31, ,
Revision 4, in use, but Revision 5 was issued February 3,1987; and '
POM 91.000.30, Revision 5, in use, but Revisica 7 was issued May 22, 1987.
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NRC Review
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Initial review by the resident inspector determined that information copies of the three procedures located in the third floor issue station '
were out of date. However, these procedures are not controlled procedures necessary for performing routine warehouse practices but are .
general radiological procedures available for information onl The l licensee immediately removed and destroyed these procedures and !
indicated that the procedures were made'available to warehouse personnel l in the RCA for info mation only. The supervisor indicated that he I would evaluate having centrolled procedures or determine an appropriate i
mechanism to maintain the "information only" copies up to date,
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all procedures located in the third floor issue station were current
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the "information only" procedures curren .
Concern No. 2 It was also noted that the consumable log was not being maintained and'
that consumables were being issued without being logge NRC Review The resident inspector noted that only two entries had been made in the Controlled Material Accountability Log provided on the RCA third floor issue station, and that the last entry was made on May 27, 1987.
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Licensee personnel indicated that- they did issue consumables frequentl Therefore, it was assumed that the log was not being maintained. The '
issue station maintained a bound log for issue / return of alcohol and acetone to supplement the accountability log. The licensee indicated th personnel would be reminded to log all consumable materials as required by procedur Only miscellaneous consumable materials were issued from this station ("snoop," grease, teflon, tape, etc.) and not safety-related consumable parts. This station does, however, issue welding rod which is not required to be logged under the Controlled Material Accountability Log. Weld rod is controlled under a separate
, procedure, and its accountability and control was maintained by a wald material requisition form. The licensee subsequently removed all alcohol, acetone, and miscellaneous consumables from this statio This allegation was substantiated; however, the licensee no longer issues consumables from this statio .
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No procedures were available for issuing lube oil fro 'he second floor i lube oil statio '
- NRC Review
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The inspector observed the second floor lube oil station and reviewed licensee controls for the issue of lube oil. Control of issuing lube l oil was by the use of a log and governed by procedure POM 12.000.5 i The log listed the type of oil issued, tne person to whom the oil was j issued, the container number that the oil was issued in, the date the oil was issued, and the date the container war returned. The inspector reviewed the log, determined that the log was being properly used, and determined that this was an adequate method to issue lube oil from the second floor lube oil station. Warehouse ptrsonnel responsible cor the station apoeared adequate;y trained in the use of the log. On February 12, 1988, the Assistant General Supervisor indicated that the-a i
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lube oil station is in the process of being converted from an inprocess )
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material control area to a warehouse storage area and will be governed 1 by Procedure DOM 12.000.29. Subsequently, the inspector verified that POM 12.000.29 was being used for the issuance of lube oil. The concern regarding the lack of procedures for issuing lube oil was not substantiated and the inspector concluded that there.are adequate controls in place for the issuing of lube oi Concern No. 4 The alleger indicated that the Assistant General Supervisor, who reports to the General Supervisor of Nuclear Materials, was notified and no action ,
was take SAFETEAM had also been notified on June 29 or 30,1987, and '
to date (July 2, 1987) no action had been take NRC Review On February 2, 1988, the project inspector contacted SAFETEAM to determine if there was an existing file which matches the above concerns and the June 29 or 30, 1987, time period. SAFETEAM identified file No. 6337 which the inspector subsequently reviewed. The SAFETEAM concern matched Concern No. I and was received by SAFETEAM on June 29,.1987. SAFETEAM initiated an investigation on June 29, 1987, and comoleted the investigation on July 14, 1987. The results of the investigation are consistent with the NRC findings for Concern No. However, SAFETEAM's fihal letter forwarding the results of the investigation to the individual providing the concern was not isseed until October 21, 1987. -The inspector also determined that the individuals who filled the positions of the Assistant General Supervisor and the General Supervisor of Nuclear Materials, at !
the time of the allegation, no longer work at Fermi Concern No. 5 ;
t One of the computerized spare parts systems, "Spare Parts Reference System" (SPRS), had inaccurate inventory quantities for specific parts, j NRC Review t
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This concern was previously identified by the NRC during an NRC Maintenance Survey conducted at Fermi 2 oa April 20-24, 1987. This concern was reviewed as part of the followup to the Maintenance 9 Survey and is addressed in Sections 3.b(10) and (11) of Report No.
341/87028. The inspector determined that the failure of SPRS to 1
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correctly reflect the quantity of spare parts in stock did not pose ;
a significant safety issue, only that a delay could result in '
3 obtaining a par i t l Concern No. 6 ,
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Access to the main warehouse was not properly controlled and personnel (visitors) were not logged i i
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NRC Review
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Ttis concern was previously identified by the NRC during a routine inspection on April 7-11 and 21-25, 1986, documented in Report No. 341/86011 (0 pen Item No. 341/86011-08). This concern was subsequently reviewed and closed in Section 2.e. of Report No. 341/87028. The inspector determined that access control was adequate to protect items from_ visiting personne Conclusions Review of the anonymous allegations received in Juiy 1987 Gid not identify any issues that would have an impcet on the public health and safety, nor were there any such issues made known to or identified by the inspectors during the inspectio No violations or deviations were identifie . Management Meetings On February 17, 1988, the licensee met with NRC Nuclear Reactor Regulation personnel in White Flint, Maryland regarding LER 88-00 After discussion of the event and the particular Technical Specifications involved, it was resolved that: NRR would provide regional personnel with documentation that the SRM/IRMs need not be declared inopersble following a reactor scram solely on the surveillance tests not being curren . The licensee would submit a Technical Specification change requesting the use of the REFUEL mode to test SRM/IRMs, a !
NRR would review the change for generic applicabilit . The licensee would keep as much of the SkM/IRM circuitry surveillantes current at power as possibl On March 29, 1988, the licensee met with NRC, Region III management in Glen Ellyn, Illinois. The meeting was the second mor,thly meetin On March 31, 1988, NRC regional and headquartecs personnel met with Canadian government representatives while on tour at the Fermi 2 l sit The meeting was to provide Canadian personnel with an understanding of the NRC's role and to discuss matters of concern to the Canadian visitor . Un.esolved Items i
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Unresolved items are matters about which more information is required in
order to ascertain whether they are acceptable items, violations, or deviations. An unresolved item disclosed during the inspection is discussed in Paragraph 1 . Exit Interview (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1)
and informally throughout the inspection period and summarized the scope and findings of the inspection activities. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspectio The licensee did not identify any such documents / processes as proprietar The licensee acknowledged the findings of the inspectio '
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