ML19345B550
| ML19345B550 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/12/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1770, NUDOCS 8012010611 | |
| Download: ML19345B550 (8) | |
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MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON REACTOR OPERATIONS 3/
Washington, D. C.
August 4, 1980 The ACRS Subcommittee on Reactor Operations held a meeting on August 4,1980 in Room 1046, 1717 H St., N.W., Washington, D. C.
The purpose of this meeting was several fold. The first object was to discuss the application of the Omaha Public Power District (OPPD) fcr a power level increase for the Fort Calhoun Station, Unit No. 1.
The proposed power level increase would raise the plants pcwer level from 1420 MWt to 1500 MWt, an increase of approximately 5.6".
A second area of discussion with the NRC Staff was the role and re- -
sponsibilitites of the NRC resident inspector relating to an increase in the direct inspection effort and incident response. A third area addressed by the Staff was a discussion of special low power test programs; finally, there was a discussion of a proposed revision to technical specifications.
Speci-fically, the issue of achieving hot standby in one hour was discussed. Notice of this meeting was published in the Federal Register on Thursday, July 17, 1980.
A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B, and the schedule for the meeting is included as Attachment C.
A complete set of meeting handouts has been included in the ACRS files. Attachment D lists the hardeuts and documents associated with the meeting. There were no written statements or requests for time to make oral statements raceived from members of the public, hcwever a " Regulatory Response Plan" pre-pared by the AIF Subcomittee on Regulatory Interactions dated February 1980 was distributed for the general information of the Subcomittee. The meeting was entirely open to the public.
The Designated Federal Employee for this meeting was Richard Major.
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'~ Mornino Session - Ft. Calhoun Station, Unit No.1 ; Power Increase Fort Calhoun Station uses a pressurized water reactor, whose nuclear steam supply system was designed by Combustion Engineering. The A-E for the plant was Gibbs & Hill, Inc. General Electric Company provided the turbine generator for the station. The plant was designed for 1500 MWt and is currently licensed to operate at 1420 K4t. The 5.6% power level increase requested will allow operation at the 1500 K4t level.
1 Ft. Calhoun's site is located approximately nineteen miles north of Omaha, Nebraska and is located five miles south of the town of Blair, Nebraska, which has a population of approximately 6800 people. The plant is located adjacent to the Missouri River, which is used for the condenser cooling water for the unit.
The containment building is a prestressed, post-tensioned concrete structure with a wall thickness of approximately four feet with a 1/4 inch thick leak tight carbon steel liner, with a design rating of 60 psig.
Ft. Calhoun's full power operating license was issued or August 9,1973. Com-mercial operation began on September 26, 1973.
For the first nine months of 1979, the Ft. Calhoun Station produced 94.6% of its rated capacity which earned Ft. Calhoun the number one ranking for capacity factor among U.S. nuclear power plants. By year's end, Ft. Calhoun achieved a 91.6% capacity factor ranking, placing Ft. Calhoun third in the nat-lon.
Ft. Calhoun produced about 73% of the electricity used by Omaha Public Power District customers in 1979.
The pcwer level increase requested by Ft. Calhoun is from the current licensed level of 1420 Kdt.to 1500 P4t. The 1500 MWt power level was used for the pur-pose of equipment design and accident analyses in the Ft. Calhoun FSAR. OPPD is doing some preliminary investigations into the possibility of increasing the
ey power level above 1500 M4t, but has no definitive plans for an application beyond 1500 KWt at the present time.
A review of the operating history reveal the steam generators are performing well. There have been no steam generator tube leaks to date; only three tubes in steam generator "A" were plugged due to some wall thinning.
The cause of the thinning has not been detemined. 9,ere has been no further steam generator tube degradation to date. The results of feedwater nozzle inspections were that no crack-like or unacceptable code discontinuities were revealed.
piping supports for the main and auxiliary feedwater lines were inspected; no damage due to vibration or water hanner was observed.
Following the 1980 refueling outage, leakage of reactor coolant was discovered during a cold pressure test at the gasket joint between the pump casing and cover of a reactor coolant pump. Subsequent inspections revealed corrosien damage had occurred to a number of closure studs on three reactor coolant pumps. Damage to the closure studs is believed to have been caused by boric acid attack.
Insulation had completely enclosed the shank area of the studs creating a corrosive environ-ment. After pump repair, new pump insulation was fabricated and installed so that the shank area of the studs was left exposed. Additional instrumentation has been installed so that any leakage between the pump casing and cover can be de-tected. OPPD looked for other similar problem areas in the plant following this occurrence, no other problem areas were discovered.
The Ft. Calhoun core is composed of 133 fuel assemblies with a 14 x 14 rod matrix.
CE designed fuel composes approximately two-thirds of the core, the other third of the core (the newest) contains Exxon-designed fuel.
In order to achieve the increased power level, the core flow rate is kept constant but the core operates at a higher. linear heat generation rate.
h n+ Model changes between Cycles 5 and 6 for physic, transient, thermal-hydraulic, and large break LOCA models reflect the change in fuel vendor from Combustion Both the methodology used by Engineering in Cycle 5 to Exxon Nuclear in Cycle 6.
Combustion Engineering and Exxon are standard methodologies; no new methodology was employed for the Cycle 6 power level increase.
ft. Calhoun is participating in the DOE high fuel burnup demonstration program designed to achieve higher uranium utilization. The program has as its goals to develop and demonstrate a fuel management scheme which allows the reduction in the amount of uranium required to produce a given amount of energy and to further reduce uranium requirements by extending fuel exposures significantly beyond current practice.
Ft. Calhoun is currently operating in the sixth cycle with rated power defined as 1420 MWt. The Cycle 6 Technical Specifications were derived at 1500 MWt; there have been no problems operating with these Technical Specifications. The only Technical Specification change for operation at 1500 MWt is a change of the rated power definition.
Mr. Wagner of the NRC Staff explained to the Subcommittee that the Staff had completed its review of the Ft. Calhoun power level increase, documented in a July 9,1980 SER, and concluded there were no objections on the part of the Staff to Ft. Calhoun being operated at 1500 MWt.
Ft. Calhoun uses a hydrogen purge system to remove hydrogen in the containment resulting from a postulated LOCA. The Staff has accepted this mathed 6f post-accident hydrogen control at the present time for Ft. Calhoun. However, this conclusion is still subject to a generic review.
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w,.- As a result of the TMI-2 incident, Ft. Calhoun has implemented all Category "A" procedures and modifications. Those items specified as Category "B" items are proceeding on schedule to meet the January 1,1981 deadline at this time, although problems could arise concerning equipment procurement.
(Category A and B requirements refer to the Lessons-Learned requirements).
At the conclus. ion of this pcrtion of the meeting, the Subcommittee did not object to the NRC Staff's plan to license Fort Calhoun Station, Unit No. 1 to operate at a power level of 1500 MWt. The Subccamittee Chairman will state this conclusion to the full Committee during the 244th (August 1980) meeting.
He will propose that a memorandum be sent to the NRC's Executive Director for Operations from the Executive Director, ACRS, stating the Connittee does not object to the power level increase.
Mr. Etherington raised a general question on ;tretch power applications.
Although his concern did not apply to the present Ft. Calhoun stretch ap-plication, he noted several past cases where stretch power could not be justified based on analyses similar to those presented in the FSAR. These stretch power applications were therefore requested on ', ass conservative e
criteria, which were acceptable to the NRC Staff, but had never been re-viewed by the Committee.
Mr. Etherington suggested that the Reactor Oper-ations Subcommittee recommend to the full ACRS that when new design bases are used in a stretch power application that the subcommittee responsible for the particular vendor design involved, initially review the case and 1
provide guidance to this subcommittee.
Mr. Etherington felt it was never the Committee's intention to have the Reactor Operations Subcommittee review reductions in design margin which properly belong under the cognizance of a particular vendor subcommittee.
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i AfternoonSessier$
Role & Responsibilities of the NRC Resident Inspector - 5. Bryan, NRC (I&E)
Mr. Bryan of the Office of Inspection and Enforcement made a presentation on the role and responsibilities of the NRC Resident Inspector (RI). He noted there was not less than two inspectors at any site, and not more than one inspector per unit at multiple-unit sites. The direct observation and independent assessment f
duties of the RI, center on safety systems surveillance tests, technical speci-fication and operating paramater checks, maintenance overviews, adherence to operating procedures, control room observation and plant tours, management activities, and transient reviews. The roles of the RI during events is to assess the adequacy of licensee's actions and safety significance of the event and secondly to provide the infomation to NRC.
The immediate actions the RI takes for significant reactor ever.ts are:
Obtains status of plant, details of incident, licensees assessment and plans.
Assesses plant conditions and licensee's actions.
When the Onsite Technical Support Center is manned, provides above informa-tion to Operations Center and mans phone or gathers data as directed.
Resident acts as NRC's senior representative on' site until more senior staff arrives.
Uses his best judgment in assessing licensee actions.
Comunicates his views to Operations Center or Regional Office if time and -
circumstances allow.
If time and circumstances do not allow, he exercises hk best judgment in requesting or suggesting licensee actions.
If licensee disagrees, resident must convey concerns to Operations Center for subsequent decision and orders.
.. ~. Resident does not have authority to order licensee actions or to take operational actions.
In the event of offsite events:
Resident proceeds to site and establishes phone contact with Operations Center or Regional Office.
Identifies subsequent phone contact schedule.
Obtains details of incicent and assess safety significance (radiation surveys conducted as needed) requests help as he determines the need.
Obtains action plan of responsible authorities. Conveys observations and conclusions to ifcensee of other officials present, if absent, resident assumes authority.
Holds until NRC team or other teams arrive.
Represents NRC until team arrives. He is directed by Operations Center or Regional Office until then.
Scacial Low Power Test Programs - N. Anderson, NRC Division of Human Factors Safety Mr. Newton Anderson of the NRC's Division of Human Factors Safety discussed the special low power test program completed at North Anna 2 and Sequoyah, with the Subcommittee. He noted there has been no final decision on whether or not te continue the program beyond the NTOLs. His general impression of the program v.as that there are benefits to be derived from the special low power test pro-gram in the areas of operator training and gaining additional infonnation on a pl ant. Most utility people he talked with (operators, plant managers) felt the orogram was worthwhile, but could be scaled down.
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. ~ - - Tech. Spec. Chances - D. Skovholt, NRC Division of Safety Technology Mr. D. Skovholt of the NRC's Division of Safety Technology discussed a change to plant operating Technical Specifications contained in an April 10, 1980 generic letter to all licensees. The change was 3 one hour time limit to achieve hot standby once a limiting condition of operation is cyceeded. Responses from some plants indicate that a reactor trip may be required to meet this requirement.
The NRC Staff is reviewing this topic. The criteria the Staff will use to choose a time to hot standby will be to insure it is (a) long enough to minimize system instabilities and assure the operator is not rushed, but (b) short enough to assure expeditious action is taken. The Staff has concluded that in some cases the one hour time period may be too short to satisfy cirterion (a).
A complete transcript of the meeting is on file at the NRC Public Document Room at 1717 H St., N.W., Washington, D. C. or can be obtained from Alderson Reporters, 300 7th St. S.W., Washington, D. C.
(202)554-2345.
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