IR 05000263/1989029
| ML20011E673 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/06/1990 |
| From: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Larson C NORTHERN STATES POWER CO. |
| References | |
| EA-89-248, NUDOCS 9002220145 | |
| Download: ML20011E673 (4) | |
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Docket No. 50-263 License No. DPR-22 EA 89-248 Northern States Power Company ATTN: Mr. C. E. Larson Vice President, Nuclear
, Generation-414 Nicollet Mall Minneapolis, MN.55401 Gentlemen:
L SUBJECT: ' NOTICF 0F VIOLATION-(NRC INSPECTION REPORT NO. 50-263/89029(DRP))
l This refers to the NRC inspection conducted during the period October 13 l
through November 27, 1989, at the Monticello Nuclear Generating Station, of
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activities authorized by NRC Operating License No. DPR-22. The inspection was conducted to review secondary containment and standby gas treatment (SBGT)
l system issues. The report docurenting the inspection was sent to you by I
- letter dated December 13, 1989.
The NRC concerns relative to the inspection findings were discussed with you and other members of your staff during an enforcement conference in the NRC Region III: office on December 21, 1989.
On October.13, 1989, NRCpersonneldiscussedwithyourstaffsecondary.
containment-integrt_ty problems that had been identified at the Duane Arnold l
Energy Center. Based on this discussion, your staff immediately initiated a
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special test of Monticello's SBGT system with the reactor building plenum exhaust fans secured. This test showed that the SBGT system, by itself, was not capable of maintaining a reactor building vacuum of at least 0.25 inches-of water, as required by Technical Specification 4.7.C.1.
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. After you detemined that the SBGT. system could 'not maintain the required negative pressure in the. secondary containment you concluded that your surveillance procedure was not adequate to identify significant leakage into the reactor building through several sets of ventilation dampers and through the reactor building railwcy airlock door gaskets. The surveillance procedure deficiency permitted operation of the-reactor building exhaust fans during the test and.resulted in masking the leakage through the ventilation dampers and-f the airlock door seals. This led you to believe that the SBGT system was E
functioning properly to maintain secondary containment integrity when, in fact, secondary containment was slowly degrading with time.
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Northern States Power Lompany
February 6, 1990 In decordance with the " General Statement of Policy and Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C (1989) (Enforcement Policy),
the violation described in the enclosed Notice has been categorized as a Severity Level III violation. A civil penalty is normally considered for Severity Level III violations. However, after consultation with the Director.
Office of Enforcement, and the Deputy Executive Director for Nuclear Materials Safety, Safeguards, and Operations Support, I have decided that the civil penalty will be mitigated in its entirety due to the extensive corrective actions undertaken by you once informed r,f a possible problem by the NRC.
In
addition to conducting an immediate test of the SBGT system, your corrective actions included a broad scope evaluation of the Station's HVAC and other
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safety system surveillance testing that went beyond the initial problem, as
.well as appropriate changes to your maintenance program for ventilation damper and door seal gasket integrity.
I also note that your past performance has been good as evidenced by your SALP 1 rating in Maintenance / Surveillance and a SALP 2 in Engineering / Technical Support. The other factors in the Enforcement Policy were considered but none warranted additional adjustment of the civil
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penalty.
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You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response.
In your response, you should docurer.t the specific actions taken and any additional actions you plan.to prevent recurrence. After reviewing your response to this Notice including your proposed corrective actions and the results of future
. inspections, the NRC will determine whether furthnr NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
Sincerely
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Original signed by C. E. Norelius for A. Bert Davis Regional Administrator
Enclosures:
1.
Inspection Report
No.50-331/89029(DRP)
See A,ttached Distribution OE (Fax)
OE:D (Fax)
DEDS (Fax)
k'. Troskoski J. Lieberman H. Thompson 01/31/90 01/31/90 02/01/90 (SEE ATTACHED CONCURRENCE)
RIII RIII RIII RIII RIII (6 uhW
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Northern States Power Company
i In a
Inspection Report
No. 50-331/89029(DRP)
See Attached Distribution
- OE (Fax)
OE:D (Fax)
DEDS (Fax)
W. Troskoski J. Lieberman H. Thompson 01/31/90 01/31/90 02/01/90 RIII R II RIII RIII RIII WM
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Schultz /gd oa /_ Gree ~ man (DRP)
Paperiello Davis 2.-54o O' > pt/P
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Northern States Power Company
February 6, 1990 i
Distribution I
cc w/ enclosure:
D. D. Antony, Plant Manager.
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. Northern States Power Company i
Licensing Fee Management Branch
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Resident Inspector, RIII Monticello
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Resident Inspector, RIII Prairie Island
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John W. Ferman, Ph.D., Nuclear
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Engineer, MPCA
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LPDR-SECY
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CA J. M. Taylor, EDO H. Thompson, DEDS-
J. Lieberman, OE
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A. B. Davis, RIII i
L. Chandler, OGC T. Murley, NRR
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J. Partlow, NRR Enforcement Coordinators RI RII, RIV, and RV F. Ingram, GPA/PA B. Hayes, 01 E. Jordan, AE0D
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M. Stahulak, RIII
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NOTICE OF VIOLATION Northern States Power Company Docket No.
50-263 Monticello Nuclear Generating Plant License No.
DPR-22 EA 89-248 During an inspection conducted on October 13 through November 27,1989, a violation of NRC requirements was identified.
In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1989), the particular violation is set forth below:
10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service, is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Technical Specification Limiting Condition for Operation (LCO) 3.7.C.1 requires secondary containment integrity to be maintained during all modes of plant operation unless specific exemption criteria are met.
In addition, Technical Specification Surveillance Requirement 4.7.C.I.c requires secondary containment capability to maintain at least a 1/4 inch of water vacuum under calm wind conditions with a filter train flow rate of not more than 4,000 scfm.
This shall be demonstrated at each refueling outage prior to refueling.
Contrary to the above, from June 1971 until October 1989, the licensee failed to establish an adequate surveillance test to demonstrate that secondary containment was capable of maintaining 1/4 inch of water vacuum under calm wind conditions with a filter flow rate of not more than 4,000 sefm.
The surveillance test was inadequate in that interaction with other ventilation systems was not properly considered.
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This is a Severity Level III violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201, Northern States Power Company (licensee) is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20555, with a copy to the Regional Administrator, Region III, U.S. Nuclear Regulatory Commission, 799 Roosevelt Road, Glen Ellyn, Illinois 60137 and a copy to the NRC Resident Inspector at the Monticello Nuclear Generating Plant, within 30 days of the date of the letter transmitting this Notice. This reply should be clearly marked as a " Reply to a Notice of Violation" and should include:
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(1) admission or denial of the alleged violation; (2) the reasons for the viola-tion if admitted; (3) the corrective steps that have been taken and the results achieved; (4) the corrective steps that will be taken to avoid further viola-tions; and (5) the date when full compliance will be achieved.
Where good cause is shown, consideration will be given to extending the response time.
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' Notice of Violation:-
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' adequate reply is not. received within the time specified in this Notice, an
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) order may. be issued to show cause why the license should not be modified,:
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suspended, or revoked or why such other action as may be proper should-not-!
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be taken.
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FOR THE NUCLEAR REGULATORY COMMISSION'
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. Bert Davis
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Regional Administrator l}
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Dated at Glen:Ellyn, Illinois-
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this :6th day of-February 1990 si r
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Report No. 60263/89029(DRP)
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Docket No. 50 263 ticense No. DPR-22 Licensee: Northern States Power Compary 414 Nicollet Mall
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Minneapolis, MN' 55401 Facility Name: Monticello Nuclear Generating Station Inspection At:. Monticello site Montice110 Minnesota inspection Conducted: October 13 through November 27, 1989
' Inspectors:
J. E. Hard P. B. Moore W. L.Il I
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Reactor Projects Branch 2 Date
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. Inspection Sumary Inspection on October 33 through November 27,1989_ (Report No. 50263/89029(DRP))
Areas Inspected: Tpecial unannounced ins;>ection by the resident inspectors or Results:y containment ~ integrity and stand>y gas treatment system issues.
seconsar The inspection revealed that the licensee took issediate corrective
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action to investigate the secondary containment issue, stopping work in
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progress which depended on secondary containment integrity. Corrective action was teken progtly. However, the surveillance procedure for the standby gas
treatment system l$ETS) was not adequate, so degraded secondary containment j
integrity any have existed for an extended period of time. One apparent
' violation was identified for not seeting the secondary containment integrity technical specification because of inadequate surveillance procedures.
(Paragraph r.a.).
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DETAILS 1.
Persons Contacted Plant Manager
- D. Antony, General Superintendent Engineering 8 Rad. Prot.
B. D. Day M. H. Clar,ityI,Assistent to the Plant Manager D. E. Nevinsk General Superintendent Operations W. J. Hill, General Superintendent, Maintenance
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security & Admin.
R. L. Scheinost, General Superintendent. Quality,ing
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5. J. Namer, Superintendent Operations Engineer A. E. Ward, Lead Production Engineer D. A. Scott, Senior Production Engineer
5. K. Peterson, Production Engineer
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Other plant personnel were also contacted by the inspectors.
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'In attendance at the exit meeting.
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2.
followup of Events
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Event Sumary On October 13 less, discussions between NRC personnel and the
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General Superintendent. Engineering and Radiation Protection were held to review the results of Standby Sas Treatment system ($4T$)
testing at the Ouane Arnold Energy Center.
(Duane Arnold personnel found a breach in their secondary containment which was detected during$GTStesting.) Based on this discussion, Monticello management ordered imediate cessation of control rod scree testing, which was in progress at the tiew untti these questions on $$TS testing and secondary containment in,teprity could be answered. A special test of
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the SGTS with the reactor building. plenum exhaust fans secured was
conducted on October 14 and was witnessed by the resident inspection
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staff.
($ee further discussion of the design features in Paragraph
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2.b.below.) This test showed that the1GT$ by itself was not I
capable of maintainint a reactor building vacas of et least 0.25 i
inch of water as spec' fled in Technical $pecification (78 4.7.C.1.
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The vacuum developed by 5GTS during the test was 0.1 inch)of
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The licensee made the required notlfications to the NRC within four i
hours as specified in 10 CFR 50.72 and by Licensee Event Report (LER)
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263/89023 within 30 days.
j Major efforts were undertaken by the licensee to measure the 'as-found*
l condition of the secondary containment from which hpothetical
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post-accident doses could be coeputed, and to detemine and correct e
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the sources of in-leakage 'to the reactor hvilding. Leakage from the L
reactor building into the reacter building exhaust plenum was i
detemined to be about 2600 cfm with the plenue exhaust fans and the
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56T5 operating. This is the amount of air flow through dampers which
bypassed the $GTS under these conditions. Leakage into the retetor j
building was shown to be principally through six sets of ventilation
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dampers which did not close properly end thropph the reactor building railway airlock door pasket material. Fo110 wing gasket replacement
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e retest with only 56T5 operating showed and repair of the dampers, lop a reactor building vacuum of 0.3 inch that the system would deve
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of water, well within the required 0.75 inch specified in TS.
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Deshn Bachround Secondary containment at Monticello is provided by the reactor building. The $6T5 is provided to exhaust the reactor building etmosphere to the offgas stack via a filter system when it is
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necessary to de so. This arrangement is designed to Ilmit ground level releases of radioactive materials in the event of accidental tech of two trains of $$TS releases inside the reactor building, ilding at 0.75 inch of water is required to maintain the reactor bu vacuum with respect to atmospheric, under calm wind conditions.
This ensures that even with significant wind, on the order of 40 mph exfiltration from.the reactor building will be limited.
Normal reactor building ventilation air exhausts to a reactor building plenum which is a large room attached te the reactor building but outside the secondary containment bour.dary. This plenum contains a variety of fans including three plenum exhaust
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fans plus other ventilation equipment.
I The plenum exhaust fans are only provided with normal power and cannot be operated under loss of-offsite-power conditions. When
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nomal power is available, two of these fans run continuously.
discharging to the reactor building vent, and maintaining the plenum
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room at a vacuum of 0.5 inch of water relative to atmospheric.
Since the SGT5 develops less than 0.5 inch of w ter vacuum in the reetter building (actually measured at 0.3 inch it follows that openings between reactor butiding and plenum wi I result in air flow toward the plenum even with 5GT5 operating. This flow of reactor
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building air which is not filtered, as is $6TS flow, and which.is discharged out thp reactor building vent rather than the offgas-stack is referred to as ' bypass flow.' As noted above, the es-found bypass flow was seasured at F600 cfm.
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Event Evaluation since plant startup. Monticello has tested the $$T5 with the reactor building plenum exhaust fans operating. Therefore, the true
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performance of $6TS and the condition of secondary containment may have been masked during this entire time. On October 14, when $6TS
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was tested without the plenum exhaust fans operating, inadequate vacuum (0.1 ind) was drawn indicating excessive leakage into the
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reactor butiding 6nd proving that this masking has been taking place.
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The Itcensee also measured the as found bypass flow (2600 cfs) on which offsite dose calculations were based. Discussion of the various dose calculations performed by General (16ctric Co. for the Itcensee follows.
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Dose calculations were perfcmed for the two design basis accidents
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for which credit.for $61$ is assumed in the tf5ARI the Loss of
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Coolant Accident (LOCA) and the Refueling Accident (REFA). Further, each accident dose was computed for the source tem discussed in the USAR and for the source ters from Regulatory Guide 1.3 (R.G.1.3).
An additional breakdown was made 50 doses could be computed assuming
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that bypass flow was present and also assuming that 5615 and secondary containment did not exist. This gave a total of eight different cases for which doses were computed.
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For these accidents the calculations showed that using the USAR source tem assumptions, doses at the Exclusion Area Boundary, in the Low Popuistion tone, and in the control room, all would be within regulatory guidelines. However. using R.S.1.3 source tem assumptions, the inhalation doses at all three locations would be i-L excessive for the LOCA as would the tsclusion Area Boundary inhalation dose for the REFA.
Further calculations were made to estimate how long after an accident the plenum exhaust fans could be left running without exceeding regulatory dose guidelines assuming the more restrictive source ters from R.G.1.3 and 2600 cfm bypass flow. These calculations by General Electric showed that the fans could remain runnine for at least 45 minutes following a LOCA. (Calculations for REFA are underway. No fuel movements will be pemitted until they are complete.) Based on these results, the licensee decided that manual actions to turn off the plenum exhaust fans after a LOCA were an acceptable short-tem fix to the question. A method was developed, tested, and procedura11:ed to accomplish this, e.
Corrective Actions ($ee LER 263/89029)
Sumery of Corrective Actions Taken:
l (1) All work requiring secondary containment integrity was stopped
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immediately upon identification of the concern.
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(t) ' The Secondary Containment Capability Test was changed to require securing the Reactor Building Planum Exhaust Fans during the test.
(3) All Secondary Containment isolation dampen were inspected; six sets of dampers were repaired.
(4) Access pistes were installed in ventilation ducts to permit inspection of damper vanes.
(5) The threshold weatherstrip gasket material for Reactor 8vilding I
Railway Airlock doors was replaced.
(6) After all inspections and repairs were completed, the revised Secondary Containment Capability Test was performed satisfactorily.
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l (7) Operatts procedures cere revised to require tripping the
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Reacter BW11 ding Plenum Exhaust Fans within 45 minutes of any l
event that requ5res initiation of Standby Ses Treatment System
for radioactive releases from within secondary containment.
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Sumary of Corrective Actions to be Taken
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(1) Correctiva maintenance testing procedures will be standardized to assure that secondary containment isolation dampers are fully functional. Procedures will be prepared, reviewed and
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approved by March 1, 1990.
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Preventive maintenance procedures will be written to requite l
inspection of material condition of secondary containment
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isolation dampers and to require corrective maintenance as degradation is discovered. Procedures will be prepared,
reviewed and approved by December 31, 1990.
r (3) Preventive maintenance procedures will be written to require i
inspection of the material condition of secondary containment
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airlock doors and to require corrective maintenance as
degradation is discovered. Procedures will be prepared,
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reviewed and approved by December 31. 1990.
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(4) Design changes are being considered to improve the Reactor-Building to Reactor But l ding Plenum boundary inte r and
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i to reduce bypass flow during secondary containmen lation
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conditions. This review will be complete by Decee6er 31, 1990.
l (5) All Technical Specification required surveillance tests will be i
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reviewed to ensure that performance of Safety Related equipment is not being enhanced by the operation of Non-Safety Related equipment. Tests will be reviewed by June 1, 1990.
(6) The design basis of secondary containment ~end Standby Gas Treatment systhms will be reviewed as part of the licensee's
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c6nfiguration management program on an expedited basis. This
review will be completed by December 1990.
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Exit Interview The inspectors met with D. Antony, Plant Manager, on November 27,1989, to discuss the scope and findings of the inspection. In addition, the inspector also discussed.the likely informational content of the
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inspection report with regard to documents or processes reviewed by the
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inspector during the inspection. The licensee did not identify any such documents or processes as proprietary.
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