IR 05000010/1988002
| ML20149M600 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/19/1988 |
| From: | Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20149M594 | List: |
| References | |
| 50-010-88-02, 50-10-88-2, 50-237-87-40, 50-249-87-39, IEB-87-002, IEB-87-2, IEIN-87-024, IEIN-87-24, NUDOCS 8802260152 | |
| Download: ML20149M600 (12) | |
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' U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-010/88002(DRP);50-237/87040(DRP);50-249/87039(DRP) Docket Nos. 50-010; 50-?37; 50-249 License Nos. DPR-02; DPR-19; DPR-25 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 f Facility Name: Dresden Nuclear Power Station, Units 1, 2, and 3 Inspection At: Dresden Site, Morris, IL Inspection Conducted: December 9, 1987 thru January 29, 1988 Inspectors: S. G. Du Pont P. D. Kaufman 'd ' -[(F([[ Approved By: M. A. Ring, Chief ,
Reactor Projects Section 1C Date Inspection Summary liispection during the period of December 9,1987 thru January 29, 1988 { (Report Nos. 50-010/88002(DRP); 50-237/87040(DRP); 50-249/87039(DRP)) Areas Inspected: Routine, unannounced safety inspection by the resioent inspectors, of operational safety verification; followup of evct,ts; r..onthly surveillance cbservation; licensee event reports followup; I.E. Bulletin followup: I.E. Information Notice followup; management meetings; and report review.
Results: Of the 8 areas inspected, no violations or deviations were identified in 7 areas; two violations were identified in the remining (Corrective actions previously taken failed to prevent a repetitive area: violation - Paragraph 3.b; failure to perform Type B & C Local Leak Rate Test surveillances within the 2 year frequency requirement Paragraph 3.c).
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DETAILS 1.
Persons Contacted Commonwealth Edison Company .
- E. Eenigenburg, Station Manager J. Wujciga, Production Superintendent
- C. Schroeder, Services Superintendent L. Gerner, Superintendent of Performance Improvement T. Ciesla, Assistant Superintendent - Planning D. Van Pelt, Assistant Superintendent - Maintenance J. Brunner, Assistant Superintendent - Technical Services J. Kotowsk', Assistant Superintendent - Operations R. Christensen, Unit 1 Operating Engineer G. Smith, Unit 2 Operating Engineer
- E. Armstrong, Regulatory Assurance Supervisor W. Pietryga, Unit 3 Operating Engineer J. Achterberg, Technical Staff Supervisor R. Geier, Q.C. Supervisor D. Sharper, Waste Systems Engineer D. Adam, Radiation Chemistry Supervisor J. Mayer, Station Security Administrator D. Morey, Chemistry Supervisor D. Saccomando, Radiation Protection Supervisor
- E. Netzel, Q.A. Superintendent R. Stols, Q.A. Engineer
.*J. Williams, Regulatory Assurance The inspectors also talked with and interviewed several other licenset employees, including members of the technical and engineering staffs, I reactor and auxiliary operators, shift engineers and foremen, electrical, ! mechanical and instrument personnel, and contract security personnel.
- Denotes those attending one or more exit interviews conducted on January 29, 1988 and infonnally at various times throughout the inspection period.
2.
Operational Safety Verification (71710 and 71707) The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period from December 9, 1987 to January 29, 1988.
The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to observe plar.t equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to "erify that maintenance reauests had been initiated for equipment in need of maintenance.
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The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan.
The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls.
During the inspection, the inspectors walked down the accessible portions of the systems listed below to verify operability by comparing system lineup with plant drawings, as-built configuration or present valve lineup lists; observing equipment conditions that could degrade performance; and verifying that instrumentation was properly valved, functioning, and calibrated.
The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection period.
The review consisted of a verification for accuracy, correctness, and compliance with regulatory requirements.
The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste sb'pments and barreling.
l 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 procedures.
The following systems were inspected: Unit 2 Control Rod Drive Feedwater Unit 3 l High pressure Coolant Injection Control Rod Drive Feedwater No violations or deviations were identified in this area.
3.
Followup of Events (92700)
i During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR S0.72. The inspectors pursued the e9ents 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 , l were conducted within regulatory requirements and that corrective l 3ctions would prevent future recurrence.
The specific events are as I follows: l
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6.
During December 1987, Units 2 and 3 experienced four events that challenged operations. On two occasions Unit 3's Recirculation Motor-Generator units developed control oil leaks that required reduction in power and single recirculation loop operation to repair the leaks.
In addition, on two occasions, air leaks developed in the Unit 2 drywell, requiring reduction in pcwer, de-inerting and entry into the drywell to repair the leaks. One of the leaks was from an air-operated valve and was isolated by closing an equipment drain line valve. The other leak which occurred on December 24, 1987, was due to a failure of the air line to the Main Steam Isolation Valve (MSIV) 10 solenoid.
Both of the drywell air leaks were corrected and the unit was re-inerted within 24 hours of exceeding the Technical Specification of 4% oxygen.
These events demonstrated that many of the improvements initiated by the licensee in 1987 had effective results. All of the events noted above required cooperation and planning between operations, maintenance and radiation protection to resolve the problems without sr.utting down the units. Additionally, both of the power reductions on Unit 3 demonstrated the effectiveness of the feedwater system modifications and testing conducted in September 1987.
The inspectors found the licensee's actions relating to the above events to be effective and demonstrating good management involvement.
b.
During Unit 2's Average Power Range Monitor (APRM) gain adjustments
! on December 18, 1987, the APRM channels 3 and 4 were placed into ' ! b pass per surveillance DIS 700-17.
Earlier, on November 13, 1987 / Intermediate Range Monitor (IRM) 16 had been placed in bypass, due
to erratic behavicr. Although IRM 16 had a caution card attached i to the IRM panel and a plastic caution plate attached to the control panel alerting the operator of the Technical Specification require-l , ment to maintain rr.re than two channels of companion IRM/APRM ' channels, per Reactor Protection system (RPS) channel. The operator placed APRM 3 and 4 in bypass on December 18, 1987 at 3:05 p.m., with IRM 16 also in bypass reduced the number of operable APRM downscale IRM high-high trip functions in RPS Channel B to only one, which is in violation of Technical Specification 3.1.A. with the reactor mode switch in the "Run" position.
This violation was not discovered until 9:20 a.m., on December 19, 1987.
This is similar to a previous violation (50-237/67026-01) in that on August 18, 1987, the number of RPS channel B APRM/IRM companion trip functions had also been reduced to only one.
The previous occurrence, documented in an LER (Unit 2-87022) described several corrective actions, including having the plastic caution plates on the control room panel as an operator aid in alerting them to the Technical Specification requirements and
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However, several other contributing factors existed such that the previous corrective actions did not prevent this recurrence. These are the continued inadequate walkdowns of the control room panels by control room personnel W management, the lack of ready identification of bypassed IkM and APRM channels by control room personnel and the continued spiking of IRMs which results in bypassing IRMs. The licensee is reviewing the following corrective actions to prevent further recurrence:
A cover for the IPM & APRM channel bypass joysticks that will alert control room personnel of bypassed channels is currently being reviewed.
A review by General Electric demonstrated that the APRM downscale trip provides no safety benefit and, as such, Dresden is considcring submitting a Technical Specification amendment.
Since the previous corrective actions to the Auguac 1987 event l failed to prevent a further recurrence, this is considered a i violation of 10 CFR 50, Appendix B, Uiterion XVI (237/87040-01), c.
As part of corrective actions to an earlier event, reported in Unit 3 LER 87-018, the licensee discovered additional problems within the i surveillance tracking system, in that, Type B and C leak rate tests required by 10 CFR 50, Appendix J were misclassified as due on a refueling outage basis rather than on a fixed-interval basis.
I This problem identified in the tracking program resulted in a -{ licensee concern as to whether Unit 3 was still in compliance with 10 CFR, Appendix J, since the refueling outage was being rescheduled from January 16, 1988 to March 26, 1988.
While reviewing past Local Leak Test (LLRT) records for Type B and C testing performed during the previous Unit 3 refueling outage, the licensee discovered that 41 individual LLRTs, of containment , l penetration boundaries and containment isolation valves, were currently in excess of the 2 year limit required by 10 CFR 50, Appendix J.
Failure to perform Type B and C Local Leak Rate Tests l within 2 years is a violation of 10 CFR 50, Appendix J (249/87039-01).
The licensee has submitted as requested for exemption from this requirement to allow continued operation of Unit 3 until the refueling outage.
d.
On January 6,1988, at approximately 7:15 a.m., with both units around 95% power, the low, medium and high range Separate.
Particulate, Iodine and Noble Gas (SPIh3) 2/3 chimney monitor was declared inoperable due to little or no flow in the sample line to the monitor from the chimney. With the SPING inoperable, the requirement is that a 2/3 chimney noble gas grab sample be taken once per shift in accordance with Technical Specification Table
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Upon further checking, the licensee determined that the ' sample line from the chimney to the SPING had frozen up, thus ! preventing the licensee from obtaining a grab sample using the normal l method. The licensee contacted the State of Illinois to see if it I would be possible to get a grab sample from their sample line, j however, that line was also frozen.
Thelicenseethenelectedto ' climb the chimney with the temperature at about -10 F, to install a temporary line in order to obtain a grab sample and avoid a dual unit shutdown. The licensee was successful in obtaining a grab sample at 2:41 p.m.
The resident inspector monitored the licensee's actions throughout the entire event and noted that the grab sample was only made possible by the ccoperative team work between all departments. Had it not been for the licensee's perseverance both units would have commenced a reactor shutdown at 3:15 p.m.
. ' Two violations were identified in this area.
4.
Monthly Surveillance Observation (61726) The inspectors observed surveillance testing required by technical specifications for the items listed below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified esing the testing were properly reviewed and resolved by appropriate-aagement personnel.
The inspectors witnessed portions of the following test activities: Main Steam Isolation Valve Timing (MSIV) (DOS 250-2) Standby Liquid Control System Pump Test (DOS 1100-1) 101 MSIV Closure Testing Local Leak Rate Testing for Electrical Penetrations (DTS-1600-4) Unit 3 Local Leak Rate Testing Procedure Bellows Seal Penetrations (DTS-1600-2) No violations or deviations were identified in this crea.
5.
Licensee Event Reports Followup (93702) Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to detennine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been acccc.plished in accordance with Technical Specifications.
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(Closed) 82058-01: Small Leak in Reactor Water Cleanup Line 2-1201 B-8"A Due to Intergranular Stress Corrosion Cracking. The Supplemental Report < was issued to provide the root cause of the event and corrective actions taken to prevent recurrence.
The cause was determined to be intergranu-lar stress corrosion cracking (IGSCC) at the heat affected zone of the weld. A section of pipe 24 inches long was replaced with type 304 dual grade stainless steel, which is less susceptible to IGSCC.
i Unit 3 (Closed) 83002-01: Leakage From Test Tap Valve 3-1299-13 Due to Seat Damage of the Valve. The Supplemental Report was issued to provide the , cause of failure and update the corrective actions.
Failure has been ' attributed to steam cutting the valve seat and galling of the pipe cap.
The valve has now been totally removed and replaced with a welded pipe cap.
! (Closed) 85018-01: Reactor Scram on High Flux Resulting From Turbine Control Valve Closure Due to Personnel Error. The Supplemental Report was issued to provide corrective actions as a result of a SNED study regarding low SDV air header pressure.
Larger air regulators will be installed on both unit's SDV air heatr.rs. This work is planned for the upcoming 1988 refueling outages.
(Closed) 66018-01: Spurious Group V Containment Isolation Due to Design Deficiency.
TFc Supplemental Report was issued to provide the cause of the event and corrective actions taken.
Cause was determined to be differential pressure spikes and/or noise generated by an annubar flow , instrument inttalled on the isolation condenser ondensate return line.
Corrective actions involved a modification (two second time delay) to the
isolrion circuity.
t (Closed) 87020-01 Spurious Group V Containment Isolation Due to Design ' Deficiency.
The Supplemental Report was issued to provide the cause of + the event and corrective actions taken to prevent recurrence. Cause , was determined to be differential pressure spikes and/or noise generated by an annubar flow instrument installed on the isolation condenser condensate return line. Corrective actions involved a modification , (two second time delay) to the isolation circuity.
The preceding LERs have been reviewed against the criteria of 10 CFR 2, Appendix C, and the incidents described meet all of the following requirements.
Thus no Notice of Violation is being issued for these ' items.
' a.
The event was identified by the licensee, ' b.
The event was an incident that, according to the current enforcement policy, met the criteria for Severity levels IV or V violations, c.
The event was appropriately reported, ! - .
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prevent recurrence within a reasonable amount of time), and e.
the event was not a violation that could have been prevented by_ the licensee's corrective actions for a previous violation.
Unit 2 LER 87035-00 (Closed) Insufficient Number of 0 parable Reactor Protection System (RPS) Channels Due to Personnel Error.
Review of this event is documented under Paragraph 3b of this report.
Unit 3 LER88001-00(Closed) LLRT Testing Intervals Discovered in Excess'of 24 Months on an Individual Test Basis Due to Procedural Deficiency.
Review of this event is documented under Paragraph 3c of this report.
Two violations were identified in this area. One violation associated with LEP. 237/S7035-00 and one violation associated with LER 249/88001-00, 6.
IE Bulletin Followup (TI-25026)
The following IE Bulletin was reviewed by the Resident Inspectors to - determine if: (1) the licensee's written response was submitted within , ' the time limitations stated in the bulletin, (2) the written response ] included all information required to be reported (3) the written response included adequate corrective action commitments based on information presented in the bulletin and the licensee's response, (4) ' licensee management foro rded copies of the written response to the i requiredonsitemanagemer.. representatives,(5)informatinndiscussedin the licer.see's response was accurate, and (6) the corrective action taken was as described in the response.
< (0 pen) IE Bulletin 87-02, Revision 0 (237/87002-BB; 249/87002-BB): "Fastener Testing to Determine Conformance With Applicable Material Specifications." The NRC Resident Inspectors participated with the licensee's Quality Control Supervisor in selecting the following. fastener i samples from the stores current inventory at Dresden to be tested: Safety Related Material Item # Size Description Mterhi SI# a DS1 1 1/4"-8 Stud A193B7 760057 DS2 1/2"-12x1 Hex bolt A307 760445 Nut A1942H 500E56 ' , DS3 5/8"-18x3 Hex bolt A193B7 778H76 l Nut A1942H 760G54 054 7/8"-9x7 Hex bolt A325TY1 761B37 i
Nut A1942H 530410 ' !
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?. . . ' DSS 5/8"-18x5 Hex bolt A325TY 761E36 Nut A1942H 760G54 DS6' 5/8"-11x8-Hex bolt SAE 5 762C84 Nut A1942H 766H82 DS7 7/8"-9x4 1/2 Hex bolt A193B7 no code Nut A1942H 530410 DS8 1 3/8" x Stud A19387 no code 9 1/4 Nuts A1942H no code DS9 3/8"-16 x Hex bolt SAE 5* 604E93 1 1/2 A1942H 508A31 0510 4"pipe.
Hex bolts-See note 507H68 clampbolts Non-Safety Related Material DN 1 1/2"-13x3 Hex bolt SAE 5 531519 Nut 41942H 533579 DN2 5/8"-11 x Hex bolt SAE 5 531533 1 1/2 Nut 533581 DN3 3/4"-10x4 Hex bolt SAE 5 531556 Nut 533476 DN4 1"-8x4 Hex bolt SAE 5 534884 Nut 5 3583 DNS 1 1/4"-7 x Hex bolt SAE 5 776349 6 1/2 Nut 776350 CN6 1 1/4"-7 x Hex bolt SAE 5 531468 2 1/2 Nut 776350 DN7 3/4"-10x7 Stud
255280 Nut 2H 255281 DN8 1"-8x6 Stud
538278 Nut 2H no code DN9 1 1/4"-8x6 Stud B7 538683 Nut 2H 538685 DN10 1 1/4"-7 x Stud 4 1/2
- SAE Fastener Specification J429 Grade E This IE Bulletin will remain open pending evaluation of test results by the NRC's Office of Nuclear Reactor Regulation (NRR) and review of the licensee's bulletin response.
i No violations or deviations were identified in this area.
7, IE Information Notice Followup (92701) The following IE Information Notice (IEN) was reviewed by the Resident Inspectors to verify (1) that the information notice was received by 'i licensee mnerement, (2) that a review for appliedbility was perfornec, and (3) that if the information notice was applicable to the facility.
' applicable actions were taken or were scheduled to be taken.
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/ . . . (Closed)IEN87-24: "Operational Experience Involving Losses of
, Electri:31 Inverters."
Inverters are used at Dresden in uninterruptable power supplies (UPSs) for the Essential Service System (ESS) and the process computer. Dresden ESS UPS and computer UPS inverters are installed in appropriate environments to meet the manufacturer (Cyberex Inc.) specifications.
The computer UPS is a 100 KVA.Cyberex model and the ESS UPS for each unit's ESS bus is a 35 XVA Cyberex model. The manufacturer specifications do not set any humidity limits, but does specify that ambient temperature should not fall below 32 *F or exceed 104 'F at full load.
The ESS UPS cabinets are located in the Auxiliary Equipment Room, which is temperature controlled at 80 + 5 'F.
The I computer UPS room is cooled by outside air and modification M12-2/3-84-6 ' is in progress to add a heater to the room. Alarms exist in the control , room to monitor for abnormal UPS conditions, and procedures in place to ' assist and direct the operator into the proper operation of these systems.
, No vioistions or deviations were identified in this area.
8.
Allegation (AMS-No. RIII-87-A-0138) ' Background
, On October 16, 1987, Region III received information concerning the
spent fuel pool at Dresden Unit 1.
The concern, pertained to potential " ] degradation of the fuel racks or the pools concrete structure as a result
of stagnant conditions in the pool for over two years.
By letter dated November 10, 1987, Region III forwarded the allegation to ) the licensee.
The licensee responded to the issue on December 10, 1987.
l In addition, a Region III Facilities Radiation Protection inspection ! was conducted into essentially this same issue during the period of .! November 2 through 24, 1987.
The results of that inspection were docu-mented in Region III Inspection Reports No. 50-237/87036; 50-249/87035.
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i NRC Findings a ] The above Region III inspection found that the licensee was already in f
the process of dealing with the possibility of microbial influenced
corrosion (MIC) of Unit 1 fuel pool structures and equipment by routinely
collecting fuel pool dip samples and taking material coating samples from
j the pool / fuel surfaces for isotopic analyses.
, < Conclusion The licensee's response to this matter, dated December 10, 1987, was reviewed by the NRC Region III staff and found to be consistent with the i i above Region III inspection performed into this area during November
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. . . l l l 1987. Additionally, the licensee initiated a hydrogen peroxide treatment - program to the pool on November 30, 1987, to reduce the microbial population. Also, a coupon surveillance program is being implemented to > monitor any long tenn corrosion effects of hydrogen peroxide on the carbon steel storage racks. Monitoring progress of the pool cleanup program will be reviewed during a future inspection as documented by Open Items 247/87036-02; 249/87035-02.
Thus, this allegation is considered closed.
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Management Meetings (30702) a.
On December 11, 1987, the Deputy Regional Administrator and members of the regional and resident staff met with Comonwealth Edison Company officials in the Region III office to review the interim findings of the licensee's investigation into Dresden Unit 3 feedwater piping system vibrations. The investigetion stemed from a feedwater systen transient on August 7, 1987. Based on the extensive feedwater system testing between September 4, 1987 and October 6, 1987. The licensee's current conclusion is that the transient was caused by flow induced oscillations.
b.
On January 19, 1988, an Enforcement Conference was held with . Comonwealth Edison Company staff at the NRC Region III Office in Glen Ellyn, Illinois. The enforcement meeting was to discuss the Unit 2 primary containment deinerting event of November 28 and 29, , i 1986, at the Dresden facility.
Inspection details of the event are
addressed in Region III Inspection Report 50-237/87005.
The enforcement discussions covered apparent violation of two Technical Specification LCOs, root causes, licensee corrective actions, and application of Technical Specification Section 3.0.A.
c.
On January 22, 1988, a meeting was held in the NRC Region III Office at the request of Comonwealth Edison (CECO). NRC Region III personnel, Messrs. W. Forney, M. Ring, R. Mendez, and D. Jones met , with CECO representatives Messrs. N. Smith and C. Sargent to discuss 10 CFR 50, Appendix J, Type B and C, Leak Rate Testing (reference CECO exemption request letter dated January 10, 1988 from Silady to
' Murley), and to sumarize the results of the reviews of the test intervals at other Ceco plants.
Problems with the Local Leak Rate ' Test Intervals were found to be limited to Dresden Units 2 and 3.
However, the potential for problems at other CECO plants was also discussed. The meeting ended with Ceco comitting to do further ' l investigation of the methodology utilized in Local Leak Rate Testing '
and to submit a letter sumarizing the discussions held in the l Region III meeting on January 22, 1988, and results of the further i j investigation.
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On January 27, 1988, the NRC Region III Administrator and members of l 'the regional staff met with the licensee at the Dresden station to review the progress of the improvement programs initiated to improve the performance and material condition of the station.
' 10. Report Review During the inspection period, the inspectors reviewed the licensee's i Monthly Operating Report for November and December,1987. The inspectors , confirmed that the information provided met the requirements of Technical .
Specification 6.6.A.3 and Regulatory Guide 1.16.
11.
Exit Interview (30703) The inspectors met with licensee representatives (denoted in Paragraph 1) . on January 29, 1988, and informally throughout the inspection period and > summarized the scope and findings of the inspection activities, j The inspector also discussed the likely infcrmational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such , documents / processes as proprietary.
The licensee acknowledged the findings of the inspection.
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/ Commonwer"5 Edison Attachment 3 - e p w.. -- .,a em - . \\ J Adeen Reply to Pow Cmce Boa 757 v Q Chcago, bro 6 60690 0767 December 10 1981 . Mr. A. Dert Davis Regional Administrator U.S. Nuclear Regulatory Cormiission < Region III ' 799 Roosevelt Road l Glen Ellyn, IL 60137 Subject: Dresden Station Unit 1 Status of the Unit 1 Spent Fuel pool NRC Docket No. 50-10 Reference: Letter froto C. E. Norelius to Cordell Reed dated November 10. 1987 concerning allegation RIII-87-1-0138
Dear Mr. Davis:
The referenced letter requests Commonwealth Edison to respond to an allegation concerning the Dresden I spent fuel pool. The concern pertained to potential degradation of the fuel racks or the concrete structure as a result of stagnant conditions in the pool.
The attachment to this letter addresses these concerns as well as efforts to assure there is no impact en the spent fuel assemblies contained in the pool.
- As indicated in the evaluation, prior to receipt of the allegation.
the potential for microbial influenced corrosion (MIC) had been previously identified, an action plan had been developed, and a program was underway to address these concerns.
Although Commonwealth Edison does not consider the safe storage of Dresden I spent fuel to have been compromised, the Company is actively , pursuing resolution of these concerns as discussed in the attachment.
' When the results of post-cleaning examinations are available, a supplemental report will be provided which will identify any further
corrective actions which may be planned at that time.
Please contact this office, should further information be required.
< Very truly yours.
. j . A. Silady
Nuclear Licensing Administrator l Attachment , , cc: M. Grotenhuis - NRR ! NRC Resident Inspector - Dresden l C.
E. Norelius - RXII l
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- . . ATTACHMENT . CECO EVALUATION.OF Alt,ECATION RI!_I,8]-A,-0 Q8 The reference letter cites an allegation (RIII-87-A-0138) regarding ' the Dresden Unit I spent fuel pool.
The alleger was concerned that the pool
had remained stagnant for over two years and that this might be having an effect on the integrity of the fuel racks and the concrete structure itself.
Commonwealth Edison was asked to address procedures or practices in Place to assure the integrity of the fuel racks in the Unit I fuel pool and the concrete structure itself.
BACKGROUND The Unit I fuel pool has been stagnant since December of 1983.
At that time, the cooling and filtration system was taken out of service due to operational problems.
Since the spent fuel in the Unit 1 pool had been discharged for a number of years and had therefore cooled down considerably, it did not generate enough decay heat to require operation of the fuel pool heat exchangers.
The stagnant conditions in the fuel tool have since led to the growth of micro-organisms and the growth of these micro-organisms has led to the concern over microbial influenced corrosion (MIC) of pool structures and equipment. MIC is a relatively new issue within the nuclear industry.
Dresden Station personnel were not aware of the MIC phenomenon at the time that the decision was mede to take the cooling and filtration system out of service.
' CORRECTIVE ACTI_ONC (IN PROGRESS AND PLANNED) Once the potential of MIC was recognized, an effort was made to determine the extent of the problem and to obtain possible solutions.
Dr.
Daniel Pope from Bioindustrial Technologies, Inc. was contracted to perform , an on-site examination of the Unit I spent fuel pool on June 9, 1987.
Dr.
Pope's results showed high levels of bacteria.
As a result of Dr. Pope's examination, an action plan was developed to reduce the bacterial population using hydrogen peroxide as a biocide.
This method of treatment was successfully performed at the San Onofre Nuclear Power Plant. An initial 72 hour treattent of approximately a 100 ppm hydrogen peroxide was proposed.
A residual level of about 10 ppm hydrogen peroxide concentration will then be maintained as needed after the initial treatment has been completed.
Addition of hydrogen pesoxide to the pool began on November 30. 1987. Once the hydrogen peroxide has performed its function of microbial population reduction, there may be a residue of bacteria remnants in the pool that will have to be vacuumed out.
Dresden is in the process of contracting the necessary services to vacuum the bottom of the pool following initial hydrogen peroxide treatment.
To monitor any long term corrosion effects of hydrogen peroxide on the carbon steel storage i racks, a coupon surveillance program is being implemented, I l
p....._ __ -...... -....................._....... ......... ..._............ . . ' . -2-j . In addition, a program of visual surveillances of the fuel assemblies is being implemented to monitor any possible long term affects of hydrogen peroxide on the fuel assemblies. Visual examinations of selected fuel bundles will be performed one month, three months, six months and one year after initial treatment with hydrogen peroxide has been completed.
Annual inspections will then be made after the initial-one year inspection program has been completed. Both General Electric and Advanced Nuclear ruels believe that any long tera effect on the fuel of such low level concentrations of hydrogen peroxide will be minimal.
An On-Site-Review of the hydrogen peroxide addition program has been performed.
Also, procedures are in place for surveillar.ce of the carbon steel coupons and the fuel assemblies. The station will install a portable demineralizing system following the current cleaning program.
In addition, evaluation of several longer term options (such as a new, permanent demineralizing syr, tem) are in progress.
SAFETY _ SIGNIFICANCE Currently, there is no evidence of loss of integrity of the fuel assemblies, fuel racks or concrete structures.
This conclusion is based on the results of a visual examination, performed utilizing an underwater television camera, of a limited number of fuel assemblies and pool structures.
A more detailed assessment of the structural integrity of these components will be performed following successful cleanirig of the pool water and structural surfaces when examinations can be better conducted.
Following this detailed assessment, appropriate corrective actions will be taken if necessary.
Long term monitoring of the fuct, fuel racks and concrete structures will be assured by the implementation of a new Dresden Administrative procedure, specifically addressing the maintenance and surveillance of Dresden Unit 1 in its non-operational status.
This administrative procedure, which is currently in draft form, is being developed as a part of the Unit I decommissioning effort.
As a reJult of a telephone discussion with Peter Erickson (NRR reviewer of tre Dresden Unit 1 possess-but-not-operate license amendment) in February of 1986. Commonwealth Edison had an analysis performed to address I the impact of a loss of cooling for the spent fuel pool. Additionally, it was decided to have other accident scenarios involving loss of inter-fuel assembly spacing analyzed (e.g. Possible effects of a seismic event). RUS Corporation was contracted to perform these analyses of various spent fuel , ' storage pool accident scenarios.
The NU3 study showed that the pool k-effectives would be below the NRC criteria specified for spent fuel , ) storage under accident conditions including wet, dry, and optimal moderation.
The study further concluded that the dose at the site boundary would be below the EPA's protective action guide under accident conditions.
Although the NUS studies were not undertaken to resolve potential i problems resulting from microbial influenced corrosion, the accident scenarios studied do provide additional assurance of pool safety margins under a number of scenarios involving substantial changes in geometry and coolant conditions in the pool.
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Corconwealth Edison therefore believes that safe storage of the
Dresden Unit I fuel is assured. Ceco further believes that any additional corrective actions resulting from the post-cleaning examinations of the fuel and pool structures, could be implemented in a timely manner without compromising safe storage of the Dresden Unit I fuel.
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