ML20058M405
| ML20058M405 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/28/1993 |
| From: | Beach A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| References | |
| EA-93-137, NUDOCS 9310050205 | |
| Download: ML20058M405 (4) | |
Text
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NUCLEAR REGULATORY COMMISSION f ID
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611 RYAN PLAZA DRIVE, sulTE 400 4,{,$l V
AR LINGToN. TEXAS 760118064 SEP 2 81993 Docket:
50-298 License: DPR-46 EA:
93-137 Nebraska Public Power District ATTN:
Guy R. Horn, Vice President - Nuclear P.O. Box 98 Brownville, Nebraska 68321
SUBJECT:
SUPPLEMENTAL INFORMATION REQUESTED DURING ENFORCEMENT CONFERENCE Thank you for your letter, dated August 30, 1993, in response to our request communicated to you during the enforcement conference of August 13, 1993.
Part of your response clarified an inconsistency between your oral presentation at the enforcement conference and P.e information provided in Licensee Event Report (LER)93-011 reiat ve to the March 10, 1993, testing of the secondary containment.
This incons1::tency was not revealed until questions were asked by NRC. We wisn to empnasize the need for frank and open discussions at enforcement conferences or other meetings in which plant problems and issues are being discussed with the NRC and expect that issues such as this will be brought to NRC's attention without prompting.
We have no further questions at this time regarding the subject issues, and no further response is required.
Sincerely,
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Bill Beach, Director j
Division of Reactor Projects cc:
Nebraska Public Power District ATTN:
G. D. Watson. General Counsei P.O. Box 499 Columbus, Nebraska 68602-0499 Nebraska Public Power District nTn.
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D.O. Box 499 Columbus. Nebraska 53602-0400 1
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N5braska Public Power District +
Nebraska Department of Environmental Controi ATTH:
Randolph Wood, Director P.O. Box 98922 Lincoln, Nebraska 68509-8922 Nemaha County Board of Commissioners ATTN: Larry Bohlken, Chairman Nemaha County Courthouse 1824 N Street Auburn, Nebraska 68305 Nebraska Department of Health ATTN: Harold Borchert. Director Division of Radiological Health 301 Centennial Mall. South P.O. Box 95007 Lincoln Nebraska 68509-5007 1
Kansas Radiation Control Program Director
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Nebraska Public Power District bec to DMB (IE45) bcc distrib. by RIV:
J. L. Milhoan Resident Inspector Section Chief (DRP/C)
Lisa Shea, RM/ALF, MS: MNBB 4503 MIS System DRSS-FIPS Section Chief (DRP/TSS)
Project Engineer (DRP/C)
RIV File Senior Resident Inspector - River Bend SRI - Fort Calhoun W. L. Brown, GC G. F. Sanborn, EO J. Lieberman, OE, MS: 7-H-5 i
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Nebraska Public Power District b'ce ;_to DMB;.;(IE45).
bec distr o. by RIV:
J. L. Milhoan
- esident Inscector Section Chief (DRP/C)
_isa Shea RM/ALF, MS: MNB8 4503 MIS System
- RSS-FIPS Section Chief (DRP/TSS) ro;ect Encineer (DRP/C)
RIV File Senior Resioent Inspector - River Bend SRI - Fort Calhoun
... Brown. GC G. F. Sanoorn. EO ieberman. OE, MS: 7-H-5 l1
i COOPER NUCLEAM STATION
. Nebraska Public Power District
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=u NSD931033 August 30, 1993 U.S. Nuclear Re$ulatory Commaission Actancion: Document Control Desk Washin6 ton, DG 20555 Subj ect:
Information Requested During August 13,
- 1993, Inspection Report 93-17, Enforcement Conference Cooper Nuclear Station NRC Docket No. 50-298, i.icense No. DPR-46 Gentismen:
During the August 13, 1993 Enforcement Conference to discuss Inspection Report 93-17, held at the Region IV Offices in Arlington, Texas, Nebraska Public Power District (District) constitted to provide additional information on several topics. The followin6 information is provided in accordance with 10CFR50.4(b) in response to this commitment.
1.
Did the Corrective Action Program Overview Group (CAPOG) review various operating experience type of documents such as General Electric Service Information Lotters, NRC Information Notices, etc.,
as a part of its review of past documentation?
The CAPOG review of past documentation did not include a systematic review of operatin6 experience documents (e.g., NRC Information Notices).
no review was directed towards the traditional CNS corrective action pro 5 ram docu-nts such as Nonconformance Reports, Deficiency Reports.
Maintenance Work Requests. Operability Determinations, Operability Evaluations, and Radiological Safety Incident Reports, as well as certain NRC-related correspondence, including Bulletins and Generic 14tters. It should be noted that during the review of specific subjects by the CAPOG, some past operatin5 experience document responses were examined; however, previously stated, a systematic review was not performed.
In as consideration of the subjects discussed at the Enforcementiconference, a review of selected documents received over the last two years will be performed.
2.
Re5ardin5 secondary containment testing, the District stated that the Aptil 7, 1993, test verified that the March 11, 1993, test was acceptable; yet. CNS personnel increased the negative _ differential pressure of the :tadwaste Building from 0.1 inches to 0.2 inches wg to pass the March test. Was there any other vorac done on buildings or their penetrations af ter March 11 and before the April 7.1993, test that could affect the results? Why would CNS have been able to pass ~the test on March 11 without a ne5ative 0.2 inches wg Radwaste building pressure?
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Documen: Coetrol D Kk
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On the morning of March 11, with standby Saa treatment running,' an effort vaa made to establish the test " configuration" the same as during the l'.191 Outa6e. As an element of this troubleshooting effortj the system etagineer (a member of the troubleshootin5 team) for the Heatin6e ventilation and Air Conditioning (HVAC) system was dispatched to ensure enat plant buildin6 differential pressures vers in their normal bands.
The Turbine Building differential pressure was verified to be in the normal band.
However, the system en5 neer found the HVAC differential 1
pressure for the Radwsste Buildin5 to be outside its normal band and, at about 0930 and with assistance from Operations personnel, : adjusted it from -0.1 inches wg to approximately -0.20 inches wg.
The effect of the chan6e in' Radweste buildin5 differential pressure on Secondary contairusent differential pressure was imsediately noticed by Operations i
personnel in the Control Room and communicated to the troubleshooting j
team leader (the Flant Engineering Superviser.)
While waitin6 for Technical Specifications required wind conditions (wind
-i speed of 2 to 5 mph), the Plant Engineering Supervisor reviewed the implications of the observed change in Reactor Building differential pressure when the Radwaste Building differential pressure was altered.
A review of the USAR, Chaptar XIV, was conducted to determine the design basis requirements for the ventilation systems under accidene conditions.
It was concluded that no concerns existed with regard to the refueling accident, but that prior to startup the anomaly would,have to be investigated further due to the potential that the Radweste Building l
ventilation system may not be available for the Design' Basis LOCA scenario. Acceptable meteorological conditions existed that evening and the test was satisfactorily completed at approximately 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br /> on March 11, 1993.
j On April 7,
1993, subsequent to modifying (SORC MWR 93-1204 for EWR 93 021) the fuel pool cooling rupture seal drain line to install a loop seal, a test of Secondary containment was performed. his test was i
to serve as an acceptance test for the loop seal installation. Die flow throu6h the empty loop seal was determined to be less than 150 cfa. With this leaka5e factored in, the Acceptance Test Summary and Conclusions determined that the cost on March 11 did, in fact, verify operability of i
Subsequently, investiBation by the Enforcement Issues Investigation Team of maintenance performed on Secondary Containment between March 11 and April 7, revealed minor maintenance activities which should not have si5nificantly contributed to differences in the pressure retaining boundary of the secondary containment. This investigation was conducted.
by review of NWRs. and review and discussion of all Deviation from Outage Guidelines (DOGS) which were generated for all maintenance activities affectin5 contairusent. It can only be concluded from the available data that Secondary containment repair activities prior to the March 11, 1993, j
test sufficiently reduced in-leakage such that secondary containment j
limits were met without the loop seal installed on March 11, 1993.
I i
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Document Control Desk August 30, 1993 Page 3 of 4 j
3.
Regarding RHR-Mov-M017 and M018 has NFFD looked at truperatura concerns j
as well as elevated pressure on downstroom piping as a result of the j
laakaBer What was the setpoint of the relief valves that would mitigate the prussure increase. how often are they checked, and has proventive maine-n==ra been performed on them?
1 The design temperature of the RHR shutdown coolin5 suction piping is 350 degress Fahrenheit.
The maxianus measured temperature on the j
dischar6e sida of the RHR-MOV-MOl7 valve body while the leak was present l
1 was approximately 150 d*5ross Fahrenheit. The normal temperature in the piping during initial shutdown cooling opnations is 200-250 degrees rahrenhait. Therefore, no design parameters were excoeded and the pipin5 ints5rity was not challen5ed due to the observ.d leak.
The reliaf valve downstream of the RHR-M07-M017 and M018 valves is a Dresser Model 1970C with a relief setpoint of 150 psig and a m1M== flow capacity of 35 gym (as compared to the measured leak rate past the valves of 0.35 SPs). The relief valve was tested per FM 03693 uncti 4
/1, and it was then included in the IST Program, which assures testire o c least ones per five year cycle. It van last testad satisfactorily on iiarch 29, 1993.
4.
Concerning Secondary Containment testing data, resolve any apparenc difference between ths LEE submitted for the leak rats test failure and Investi ating Team the infortsation presented by the Enforcement Issues 5
regarding the conduct of leak rate testing on March 10, 1993.
NRC IR 93-17 (at page 5) indicates that an undocumented Secondary containment leak rate test was conducted on March 10.
LEE 93-011 discusses testing on March 10, but does not discuss documentation of the casting. The Enforcement Issues Investigation Team reviewed the Control Room and Shift Supervisor Logs for that data and found no record of an official test.
Initial interviews of personnel could not definitively establish whether a cost was conducted on March 10, 1993.
Because an entry into SP 6.3.10.8 would have been logged for an official test, soma personnel indicated that perhaps a lineup for troubleshooting purposes (which would not be loE5ed) may have occurred.
As a result, the Team initially concluded that no " testing" was conducted on March 10, 1993.
However, based upon subsequent interviews and a review of the control Room strip chart for Reactor Buildin6 differential pressure, the Enforcement Issues Investi5ation Team now concludes that indeed a cast wgg attempted on the evenin5 of March 10, but the test was apparently performed without initiating SP 6.3.10.8.
Late in the evenin5 on March 10, Standby Gas Treatment was started and the Control Room scrip chart for Reactor Building differential pressure was annotated
" Containment leak test."
'nThen test parameters revealed that the acceptance criteria could not be achieved, the test was apparently
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terminated.
1 In an= mary, LER 93-011. submitted on May 12, 1993, is accurate in its description of the testing performed on March 10, as clarified herein, i
Donsesnt Control Dock August 30, 1993 Paga 4 of 4 S.
Regarding the valve lineups to supporc int *5 rated leak rats tests (ILRTs),
a)
When was the last system valvo lineup performed for the H /02 2
analyzer?
The inst valve lineup was performed on June 28, 1993, por CNS Procedure 2.2.60.A.
b)
Have there been other previous instances noted where a valve was found to be mispositioned during the conduct of an integrated leak rate tast?
A. review - of IIATs conducted since 1976 revealed only 'two other instances of mispositioned valves and one case where a relief valve had been temporarily removed and a blank flan 6e installed.
The mispositioned valves were found and correctly positioned either during protest lineups, or durin5 the performance of SP 6.3.1.3, ILRT, and did not affect the test results. It should be noted that SP 6.3.1.3 contains a comprehensive listing of primary containment interfacing valves that are required to be ali ned in preparation 5
for the IIRT.
Should you have any further questions concernin5 this submittal, please contact No.
Sin rely, O.-~
G Horn Vice President - Nuclear CRH/ dis /ya i
NRC Re5 onal Offica cc:
Region IV Arlington, Texas I
l NRC Resident Inspector Cooper Nuclear Station i