ML19322C577
| ML19322C577 | |
| 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 |
| Shared Package | |
| ML19322C578 | List: |
| References | |
| TASK-TF, TASK-TMR NTFTM-790724-03, NTFTM-790724-3, NUDOCS 8001170903 | |
| Download: ML19322C577 (12) | |
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SEP 2 71979 MEMORANDUM FOR:
Richard C. DeYoung, Deputy Staff Director, NRC/TMI Special Inquiry Group FR0f1:
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 D0R 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 frorr 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 summary of major items which fall outside of the nonnal review process but were raised 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.
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Office of Nuclear Reactor Regulation Harold R. Dento, Director
Enclosures:
As Stated 8001170 %
INDEX OF SIGNIFICAi4T OC0iiEE ISSUES DURING REVIEW KEY:
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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 f
I-3 Supplement No. 3 - July 10,1973 II -
" Safety Evaluation of the Oconee Nuclear Power Station Units 2 and 3" Docket Nos: 50-270/287 - July 6,1973 II - 1 Supplement No.1 - August 3,1973 II - 2 Supplement No. 2 - October 1,1973 II - 3 Supplement No. 3 - January 29, 1974 INDEX 4
I page 2 Core Power Level l
H 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
! 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 Pirap Replacement i
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. 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 4
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 System Reliability I page 45 Post Accident Hydrogen Control I page 49 Anticipated Transients Without Scram 4
I page 51 Diverse Reactor Trip for ECCS I page 52 100% Load Rejection 1
I page 53 Onsite Power Reliability i
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 and Steam Generator Damage I-3 Fuel Densification, Unit 1 4
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. II page 3-9 Loss of Intake Canal Weir II page 4-6 Positive Moderator Temperature Coefficient II page 4-8 Core Mapping 11 page 4-8 Zenon Oscillations 11 page 4-9 Fuel Densification II page 4-11 CRD Motor Extension Tube Defects II page "-14 CRD Mechanism Damage (Dry Scram) 11 page 4-15 Prepressurized Fuel II page 5-1 Vessel Internals and Steam Generator Damage 11 page 5-1,'d Loose Parts Monitor II page 5-3 Vibration Measurements on Reactor Internals 11 page 5-7 Reactor Vessel Materials Surveillar.ce Program II page 5-8 Flood Line Flow Restrictor II page 6-5 Steam Generator Subcompartment Overpressure II page 7-2 ECCS Reflooding 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 Il page 7-32, 45 NPSH for ECCS and Spray Punps II page 7-34 Non-Class I Equipment Failure II page 7-36 Auxiliary Service Water II page 7-37 Anticipated Transients Without Scram 11 page 7-38 High Energy Line Ruptures II page 7-45 Post Accident Hydrogen Control i
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4-II page 9-1 Vent Radiation Monitors l
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 i
j II-2 Pump Overspeed II-2 Core Mapping 11-2 Steam Generator Subcompartt.1ent Overpressure 4
II-3 Fuel Densification, Unit 3.
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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 failure 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 thermal shield during flow conditions.
Resolution:
B&W made extensive modifications 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 Power Company Oconee Nuclear Power Station, Unit 1, Supplement No. 2, December 19, 1972.
Safety Evaluation of the Oconee Nuclear Power Station Units 2 and 3 - Docket Nos. 50-270/287, July 6,1973 - Section 5.2.1.
. 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 assunptions were incorporated with regard to gap conductance and other physical parameters. The staff concluded that densification effects on the integrity of the 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.
. Issue:
==
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 which 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 l
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 -
Docket Nos. 50-270/287, July 6,1973 - Section 7.1.11.
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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 pump seal failure which duaped 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 pemanent increased capacity was planned.
References:
Operating reports t
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... - 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 Keowee 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 Power Company Oconee Nuclear Power Station Unit 1 - Docket No. 50-269, December 29, 1970 - Section 8.4.
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' Issue:. Control rod drive motors
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Description:==
Oconee 1 experienced burnout of control ' rod drive stepping motors.
Resolution: Control rod drive motors were replaced by a more advanced improved -
model and performed satisfactorily.
R eferences:
Operating Reports 4
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- Issue: -Punp lube oil fires
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Description:==
On at least two occasions Oconee suffered fires 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 providing sump overflow capacity and the second was subsequent to the fix as a result of maintenance error.
Resolution: Overflow from sumps was collected in barrels by way of installed piping. Procedures were instituted to prevent overflow valves from being left closed.
References:
Operating Reports l
. ITEN'2 RESPONSES It would be difficult, after the fact, to wed any of the 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 Oconee Nuclear Power Station Units 2 and 3" - Docket 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 Oconee plant, Unit 1, the power operated relief valve was opened and failed to close.
The block valve was closed and could not be reopened against system pressure. The plant continued operation on pressurizer heaters and sprays. The PORV was removed and exanined.
Operating reports should provide information on the resolution of this problem.
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