ML19250B248
| ML19250B248 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Crane  |
| Issue date: | 09/27/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Deyoung R NRC - NRC THREE MILE ISLAND TASK FORCE |
| References | |
| FOIA-79-98, TASK-TF, TASK-TMR NTFTM-790724-03, NTFTM-790724-3, NUDOCS 7910290203 | |
| Download: ML19250B248 (14) | |
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SEP 2 71979 MEMORANDUM FOR:
Richard C. DeYoung, Deputy Staff Director, NRC/TMI Special Inquiry Group FROM:
Harold R. Denton, Director, Office of Nuclear Reactor Regulation
SUBJECT:
RESPONSE TO NRC/TMI SPECIAL INQUIRY GROUP REQUEST NTFTM 790724-03 You requested certain information regarding the review and operating history of the Oconee plants in your memorandum to me dated July 24, 1979. The enclosed responses address Items 1 and 2 and the second part of Item 3 relating to three significant events for which DOR assumed the lead responsibility from IE.
We understand that IE will separately respond to the remaining parts of Item 3.
Of the NRR staff who directly participated in the Oconee review, only Irving Peltier, Albert Schwencer and Mort Fairtile were available for background on Oconee on a "best-efforts" basis.
Because of the limited time available, these people relied heavily on their recall and infomation available in the safety evaluation reports without any assistance from technical reviewers who had participated in the Oconee review. The responses to Items 1 and 2 that were not documented in the safety evaluation reports relied completely on recall and therefore we cannot assure you of the completeness of these responses.
However, we have included in the enclosure an index of significant Oconee review issues for which the issue itself and the resolution are reasonably summarized in the SER's and their supplements.
The document and page numbers are given for each issue.
In addition, the enclosure includes a brief swnary of major items which fall outside of the nomal review process but were 'aised because of a generic concern or operating experience at the Oconee facilities.
For these items we have, where ever possible, however, cited references in readily available documents such as the SER and its, supplements.
D$w
' c.w Harold R. Dento Director
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Office of Nuclear Reactor Regulation
Enclosures:
As Stated 898 185 7910290 2_.03
INDEX OF SIGNIFICANT OCONEE ISSUES DURING REVIEW XEY:
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" Safety Evaluation of the Duke Power Company Oconee Nuclear Power Station, Unit 1" Docket No: 50-269 - December 29, 1970 I-1 Supplement No.1 - March 24,1972 I-2 Supplement No. 2 - December 19, 1972 I-3 Supplement No. 3 - July 10,1973 II -
' Safety Evaluation of the Oconee Nuclear Powr Station Units 2 and 3" Docket Nos: 50-270/287 - July 6,1973 II - 1 Suppl ement No.1 - August 3,1973 II - 2 Supplement No. 2 - October 1,1973 II - 3 Supplement No. 3 - January 29, 1974 INDEX I page 2 Core Power Level I page 6 Valley Diffussion Model I page 13 Incore Detectors I page 14 Xenon Induced Oscillations I pages 15, 17 Single Loop Operation I page 16 DNB Thermal Hydraulic Correlations I page 18 Incore Thermocouples I page 18 Prepressurized Fuel I page 22 Internals Vent Valves I page 23 Control Rod Drive Roller Nut Design I page 23 Unit 1 Primary Pmp Replacement
. I page 24 Once Through Steam Generator I page 30 Reactor Internals Vibration Monitoring I page 31 Loose Parts Monitoring i page 36 Penetration Room Ventilation System I page 39 ECCS Redesign to GDC 44 I page 39 ECCS Analysis I page 42 Core Flooding Tank Block Valves I page 44 pH Control of Containment Spray Solution I page 44 Reactor Building Cooling Systm Reliability I page 45 Post Accident Hydrogen Control I page 49 Anticipated Transients Without Scram I page 51 Diverse Reactor Trip for ECCS I page 52 100% Load Rejection I page 53 Onsite Po m r Reliability I page 53 Independence of ESF Buses I page 60 Loss of Component Cooling Water System I page 66 Dropped Fuel Cask Analysis I pages 66, 69 Spent Fuel Storage Filters I page 71 Operating Shift Size I page 76 ACRS Recommendations I-1 ECCS Interim Acceptance Criteria Evaluation I-2 Vessel Internals cnd Steam Generator Damage I-3 Fuel Densification, Unit 1 898 187
,. II page 3-9 Loss of Intake Canal Weir Il page 4-6 Positive Moderator Temperature Coefficient II page 4-8 Core Mapping II page 4-8 Zenon Oscillations II page 4-9 Fuel Densification II page 4-11 CRD Motor Extension Tube Defects II page 4-14 CRD Mechanism Damage (Dry Scram)
II page 4-15 Prepressurized Fuel II page 5-1 Vessel Internals and Steam Generator Damage II page 5-1,13 Loose Parts Monitor II page 5-3 Vibration Measurements on Reactor Internals Il page 5-7 Reactor Vessel Materials Surveillance Program II page 5-8 Flood Line Flow Restrictor II page 6-5 Steam Generator Subccmpartment Overpressure II page 7-2 ECCS Reflocding Analysis II page 7-5 ECCS Small Break Analysis II page 7-9 Core Flooding Tank Line Break II page 7-30 LOCA With Idle Reactor Coolant Pumps II page 7-32, 45 NPSH for ECCS and Spray Pteps II page 7-34 Non-Class I Equipment Failure II page 7-36 Auxiliary Service Water II page 7-37 Anticipated Transients Without Scram II page 7-38 High Energy Line Ruptures I: page 7-45 Post Accident Hydrogen Control 898 188
. II page 9-1 Vent Radiation Monitors II page 9-1 Charcoal Filters II page 10-1, 11-3 Refueling Accident II-l Fuel Densification, Unit 2 II-2 Operations at 2468 MWt II-2 Positive Moderator Temperature Coefficient 11-2 Pump Overspeed II-2 Core Mapping 11-2 Steam Generator Subcompartment Overpressure II-3 Fuel Densification, Unit 3.
m 898 189
ENCLOSURE ITEM 1 RESPONSES Issue: Flow induced failure of vessel internals
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Description:==
During preoperational testing of Unit 1, failure of instrument guide tubes at the bottom of the reactor vessel was experienced. The f ailure resulted in damage to the top tube sheet and tube ends of both steam generators.
The damage was major in one steam generator and minor in the second. The reactor core was not in place at the time.
It was also discovered that there was excessive movement of other internals such as the themal shield during flow conditions.
Resolution: B&W made extensive modificaticns to beef up the instrument guide tubes and the top tube sheet of the steam generators were machined to repair the damage. The thermal shield and its installation anchors were modified to reduce movement and wear. An extensive internals vibration program was conducted at B&W facilities to better understand the problem and to improve analytical models. The fixes were apparently satisfactory and the staff approved the modification.
References:
Safety Evaluation of the Duke Powr Company Oconee Nuclear Power Station, Unit 1, Supplement No. 2, December 19, 1972.
Safety Evaluation of the Oconee Nuclear Pomr Station Units 2 and 3 - Docket Nos. 50-270/287, July 6,1s'73 - Section S.2.1.
898 190
. Issue: Fuel densification effects
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Description:==
The phenomenon of fuel densification was discovered and resulted in a generic program to study the effects of fuel densification on fuel rod integrity and thermal behavior.
Resolution: B&W developed analytical models to calculate the effects of fuel densification and B&W reactor fuel pellets were modified in design to minimize the densification phenomenon. The effects analysis resulted in some reactor operating restriction on linear heat rate, flux imbalance, etc. that were more restrictive than previous restrictions. The staff approved the B&W analytical models with the provision that certain conservative assumptions were incorporated with regard to gap conductance and other physical paraneters. The staff concluded that densification effects on the integrity of tne fuel for at least the first fuel cycle were acceptable.
References:
Safety Evaluation of the Duke Power Company Oconee Nuclear Power Station Unit 1 - Docket No. 50-269, Supplement No. 3, July 10,1973.
Safety Evaluation of the Oconee Nuclear Powe Station Units 2 and 3 - Docket Nos.
50-270/287, Supplement No.1 & 3, January 29, 1974.
9
. Issue: High Energy line breaks
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Description:==
As a result of an anonymous letter to the ACRS,,the issue was raised that high energy line breaks outside of containment could either by direct pipe whip or jet impingment of by environmental effects such as pressure, temperature, flooding or moisture impair the operation of safety systems required to mitigate the consequences of the accident or cause the loss of function of safety systems.
Resolution:
The licensee made extensive modifications to the Oconee facility whicn included additional pipe restraints, methods for venting penetration rooms containing safety systems, etc. The corective measures were extended to include low and moderate energy systems for the protection against environmental effects.
The staff established criteria for the postulation of pipe break locations and the type of pipe breaks and acceptance criteria for the protection of safety related equipment. The licensee's corrective modifications were acceptable to the staff.
References:
Safety Evaluation of the Oconee Nuclear Power Station Units 2 and 3 -
Occket Nos. 50-270/287, July 6,1973 - Section 7.1.11.
Issue: Primary pump seal failure
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Description:==
As a consequence of loss of cooling water to primary pump seals, Oconee 1 suffered a primary punp seal failure which dumped a large quantity of slightly radioactive water on the containment floor. The liquid rad waste system was not adequate to process the volume of water and therefore, it had to be trucked to a reprocessing plant.
Resolution: Measures were taken to assure pump seal cooling and the licensee instituted design modifications to increase the capacity of the liqued rad waste storage and processing facilities. Temporary increased storage capacity was added to the facilities and long term permanent increased capacity was planned.
References Operating reports m
898 193 Issue: Onsite power
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Description:==
The Oconee onsite power system is the Keowee hydroelectric generators in combination with one dedicated Lee Steam Station gas turbine as back up during periods when the hydro station is down for maintenance. Following the review it was learned that a single failure or inadvertent closing of the water intake gate for the Keowe Station could make the Keowee hydro units unavailable for emergency onsite power.
Resolution: The applicant agreed to chain and lock open the intake gate to prevent inadvertent or accidental closing of the gate during nuclear power plant operation.
References:
Safety Evaluation of the Duke Powr Company Oconee Nuclear Power Station Unit 1 - Docket No. 50-269, December 29, 1970 - Section 8.4.
898 194 Issue: Control rod drive motors
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Description:==
Oconee 1 experienced burnout of control rod drive stepping motcrs.
Resolution: Control rod drive motors were replaced by a more advanced improved model and performed satisfactorily.
References:
Operating Reports e
898 195
s
. Issue: Punp lube oil fires
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Description:==
On at least two occasions Oconee suffered # ires inside containment resulting from lube oil for the main coolant pump motors overflowing the sumps and spilling onto hot reactor coolant piping. The first time was prior to prc/iding sump dverflow capacity and the second was subsequent to the fix as a result of mainte.unce error.
Resolution: Overflow from sumps was collected in barrels by way of instalTed piping. Procedures were instituted to prevent overflow valves from being left closed.
References:
Operating Reports ITEM 2 RESPONSES It would be difficult, after the f act, to wed any of tne issues and recommendations of NUREG-0560 and NUREG-0578 to the staff's review of the Oconee plants prior to July 1974. However, during the review and early operations of these plants, there was a general concern about the availability and reliability of auxiliary feedwater and the operation of power operated relief valves on the pressurizer.
A brief discussion of these two matters follows.
Auxiliary Feedwater The staff became concerned about the availability and reliability of auxiliary feedwater during review of the hydrology of the intake canal weir and its potential for failure subsequent to a loss of Lake Keowee water level and during the review of high energy line ruptures external to containment. Discussion and resolution of these concerns can be found in " Safety Evaluation of the Ccenee Nuclear Power Station Units 2 and 3" - Cocket Nos. 50-270/287 - Page 3-9, 7-36, and 7-38, Dated July 6,1973.
Power Operated Relief Valves On at least one occasion during operation of the Cconee plant, Unit 1, the power operated relief valve was opened and failed to close. The block valve was closed and could not be reccened against system pressure. The plant continued operation on pressurizer heaters and sprays. The PORV was removed and examined.
Operating reports should provide information on the resolution of this problem.
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UNITED STATES
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September 4, 1979
...s Dockets Nos. 50-269/270/287 P.EMORANDUM FOR:
Ocmenic B. Vassallo, Acting Director, Division of Project Managemer.t FROM:
Darrell G. Eisenhut, Acting Director, Division of Operating Reactors
SUBJECT:
D0R RESPONSE TO NRC/TMI SPECIAL I! ;UIRY GROUP (SIG) REQUEST FOR OCONEE STATION OPERATIONAL.ISTORY This mamcrandum is the DOR certion of the response to the request for infor-maticn from DeYoung, SIG, to Denton/Stello dated July 24, 1979, enclosed.
COR, after discussion with IE Region II, is responding to these items transferred to DOR by IE referred to in the second part of Item 3 of the SIG request. DPM: took responsibility for Items 1 and 2, IE Region II, is responding to tne first part of Item 3.
The second part of Item 3 asked for a description of the staff's handling of f
"significant" events and how the lessons learned from the events were constructivel; used to prevent future adverse consequences.
All significant ovents at an operating plant are normally reported to IE through a Licensee Event Report.
IE will transfer these events that require licensing action to resolve, ard these generally require additional review by DOR. Only three such significant events at Oconee were transferred to DOR:
Possible Use of Atypical Weld Wire in Reactor Yessel Welds, Turbine Building Flooding and
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Steam Generator Tube Failures.
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We were informed of the atypical weld wire problem by Babcock & Wilcox to the NRC dated August 4,197.b.hrough a Part 21 submittal Duke contacted IE the same day to report that weld material in the Oconee 3 reactor vessel may be different from the Mill Certifications.
IE transferred the review to DOR.
We were informed that as many as 12 reactor vessels could be involved that were nanufactured by 35W, four in Westinghouse sy:te-'s, seven in ESW systems ar.d one GE vessel.
ESW conducted an investigation of QA records at their plant to determine which vesseis tad atypical material. Tne investigation was ir. conclusive.
On August 14, 1978 a
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- ared asking all potentially affected plants to put more
e had ;reviously phcned
- serva:ive heatuo and cooldown curves into use.
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By August 23, 1975 all the licensees "ad responcec :na: B&W, er
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they were in use. B&W completed an in-reactor irradiation at Crystal River Unit 3 of the atypical material and submitted a preliminary report to NRC.
We have concluded that continued operation of these plants is acceptabl1 with the more conservative heatup and cooldown curves in ase.
We were informed of the Turbine Building flooding by Licensee Event Report R0-287/75-18 dated October 25, 1976 addressed to OIE, Region II. The event occurred during a main condenser inspection. The discharge water surface is at a higher level than the Turbine Building floor, the condenser manways were opened and a condenser circulating water discharge salve failed open, which resulted in a flooding path.
The FSARhad assumed a flood of about 1000 cfs from the condenser circulating intake pipe, w. ich was greater than the actual discharge. side flood.
The licensee propos:d installing protective walls around vital equipment in the Turbine Building and separating the Turbine Building from the Auxiliary Building by waterproofing and sealing the common wall between the buildings.
Duke's proposal was submitted in a letter dcted April 21,1977; the proposals were first discussed with the ONRR staff during, two meetings in November 1976.
Subsequent to Duke's April 21, 1977 let cer the modifications were performed under 10 CFR 50.59(a)(1).
Duke, in order to reduce th5 number of vital areas in their Station Security Plan proposed a Safe Shutdown Facility independent of the present shutdown capability.
This Safe Shutdown Facility would also serve to get the plant in a safe configuration I.
after either a flooding event or a fire. This facility is currently under construction and should be operable by the end of 1980.
This flooding event at Oconee was unique in that the plant had a heat sink water surface at a higher elevation than the Turbine Building floor level.
There is an ongoing review of all plants that started as a rewit of this flood in addition to the generic flood review caused by the event at Quad Cities.
The Oconee Unit No. I steam generators suffered recurring leaks that raised quest" over continued safe operation of B&W once-through steam generators.
One primary concern was how many simultaneous tube failures could be tolerated, say in the event of a main steam line break, and not exceed Part 100 doses at the site bounda ry..
A series of seven Licensee Event Reports dated between October 31, 1976 and April 27, 1978 were submitted by Duke describing 10 tube leaks, tube inspections ar.d tube removal or plugging operations. The first leak reported in October 31, 1976 occurred in SG 1 A, the remaining nine leaks occurred in SG 13.
Since April 27, 1978 another LER was submitted by Duke for a SG 1B leak, the LER was dated August 20, 1979.
The DOR staff held many meetings with Duke and B&W, sent many formal requests for information and prepared a Safety Evaluation dated October 4,1977.
This SE effected a reduction in the primary to secondary Technical Specification leak limit through a SG tube frcm 1.0 gpm 0.3 ;;. and found that <ce understood the mechanism of degradaticn and rate c' that.he SG could continue to operate tnroughout the inspectior, ds;racati:- s:
' ne rs al.
- xe submi::ec a 3afety Analysis datec September 9, 1977, which i cica:ed 7,a: _p to ten 53 ubes ccuid uncergo a dcuble ended rupture anc
- ::e:s se;aration of the encs during a main steam line break and that the cor; 8%
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quences would not exceed Part 100. The staff has not completed the evaiuation of this submittal. Our review indicated two separate degradation rechanisms, one an erosion /rSrrosion effect at the 14th support plate level and lane tube degradation.
The review resulted in Technical Specification changes for B&W operating plants in additico to the reduced tube leak limit
~at Oconee 1.
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D. G. Eisenhut, Acting Director Division of Operating Reactors Office of Nuclear Reactor Regulation
Enclosure:
Memo to HR0enton & VStello fm. RDeYoung dtd. 7/24/79 re: Request for Information cc:
RYollmer BGrimes
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WGam.~iill 4'
LShao JRMiller TJCarter WRussell RReid MFairtile RIngram VNoonan 898 200
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