ML19351C821
| ML19351C821 | |
| Person / Time | |
|---|---|
| Issue date: | 08/21/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1770, NUDOCS 8010080147 | |
| Download: ML19351C821 (13) | |
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(UE DATE: August 21, 1980 i
LXRTIFIED:
9/3/80 3 3lN MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON REACTOR OPERATIONS 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) for a power level increase for the Fort Calhoun Station, Unit No. 1.
The proposed power level increase would raise the plants power 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 specia'l 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 handouts and documents associated with the meeting. There were no written statements or requests for time to make oral statements received from members of the public, however a " Regulatory Response Plan" pre-l pared by the AIF Subcommittee 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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Morning Session - Ft. Calhoun Station, Unit No.1 - Power Increase Fort Calhoun Station uses a pressurized water reactor, whose nuclear steem supply system was designed by Combustion Engineerin~g.
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 MWt. The 5.6% power level increase requested will allow operation at the 1500 MWt level.
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 conc.ete 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 oMrating license was issued on 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 nation.
Ft. Calhoun produced about 73% of the electricity used by Omaha Public Power District customers in 1979.
The power level increase requested by Ft. Calhoun is from the current licensed level of 1420 MWt.to 1500 MWt. 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 sorne preliminary investigations into the possibility of increasing the
'. O power level above 1500 MWt, but has no definitive plans for an application beyond 1500 MWi. 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 detennined. There has been no further steam generator tube degradation to date. The rasults 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
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vibration or water hammer 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 corrosion 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.
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- Model changes between Cycles 5 and 6 for physic, transient, thermal-hydraulic, and large break LOL4 models reflect the change in fuel vendor from Combustion Engineering in Cycle 5 to Exxon Nuclear in Cycle 6.
Both the methodology used by 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 bt~nup demonstration pr, gram 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 method 6f post-accident hydrogen control at the present time for Ft.. Calhoun. However, this conclusion is still subject to a generic review.
As a result of the TMI-2 incident, Ft. Calhoun has implemented all Category "A" procecures and modifications. Those items specified as Category "B" items
are proceeding on schedule to me.e. the January 1,1981 deadline at this time, although problems could arise concerning equipment procurement.
(Category A and B requirements refer to de Lessons-Learned requirements).
At the conclus. ion of this portion 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 Subcommittee Chainnan 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 Comittee does not object to the power level increase.
Afternoon Session Role & Responsibilities of the NRC Resident Inspector - S. 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 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 information to NRC.
The immediate actions the RI takes for significant reactor events 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.
Corm 1unicates his views to Operations Center or Regional Office if time and circumstances allow.
If time and circumstances do not allow, he exercises his 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 incident 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 licensee 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.
M. r t 7...
o Special Low Power Test Procrams - 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 to continue the program beyond the NTOLs.
His general impression of the program was that there are benefits to be derived from the special low power test pro-gram in the areas of operator training and gaining additional information on a plant. Most utility people he talked with (operators, plant managers) felt the program was worthwhile, but could be scaled down.
Tech. Spec. Changes - 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 a ene hour time limit to achieve hot standby once a limiting condition of operation is exceeded.
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 tire period may be too short to satisfy cirterion (a).
1 A complete transcript of the meeting is on file at the NRC Pubite 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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hderal Reghter 8 Vcl. 45. No.139 / Thursds).,. / W.15 -4 VVWe'es hav detrain the passengrrs es a Isot resort. -
signals in train movements cnd wT to what is required by the above quoted De Board feels that the ir9y of the insure that all crewmembers involvec rdes and w:uld enly lend 13 confusion.**
BART train operator to uw'e the anderstand when the s!gna! mode is go.s-sa.sk copies of Safety Board lead portion of the t s5 from the changed. (See 45 FR 37919. June 5,19N ren ts an asscable withe'ut charge. u long burning cars. the quki.wss In which the ICG in response states that its s2 ;.mited supptes last. Copies of Scard smo)e and fire spead and the inability operating rules 12.12(i) and 633 are recommendataan letters. and responses or related c.orresponden:e. are aho proeded to later evacuate the minima! number of applicable in respect to the aed.: erit frat of charge. Ar requests for copies must be ces ered by the recommendstion and p:ssengers are protlen which should be addressed and this requires redsfon tbst complience with these rules would ' !L "L'4 ident. Led by recommendation or seport number. Address requests to: Pubhc of the emergency guldehnes.De have prevented the unfortunate event.
Board's March 21 letter further sugFests He applicable rules are:
3 that the 13 recommendstions the Board Rule 12. HAND MAC AND LANTERN Muit p!e copies of Safety Board reports me) be purchased from the Naoonal l
made to BART.Nos R-79-12.R-79-13
'EIGNALS Technical trJormeton service.U.S.
cnd R-79-4: through R-79-5, should sus r y-sv. e
,,,,, uw sua Department of Commerce. Spring 5 eld.Va.
= ews one c!so be considered in a review and gN * "" ** *'
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eevision of the AFTA guidelines.
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(49U.SC ts03le)(2).1 sos)
De Safety Board asked ATTA to reconsides its re:ponse oflaet October
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a==a y,,g,,e g, yhher' 30 to recomicendation R-79-53. Tbe 85,5",,5g,y,",",I, redem/AesisterLialson @er.
Board also asked to be apprised of any N,g,g, - g e==
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l corrective actions voluntarily taken by 8,5p y*T*l;,",,,,,""8 AFTA members because of their hearing susen coot aem
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(f the problems encountered by BART 5O am during the isnuary 17.19 9. fire.
lo lis detailed May 1 response. APTA anrme an = '====a.
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cxpresses reF'et in beving interpreted Rule 12(i). Hani fbg and lantern signals DNIO recommendation R-79-53 as specifically prescribed in Rules 12(ej through 12;g) must urposes descibed be used for the Ung or tantern signah may be Advlsory Committee on Reactor l
dealing with the AFTA document, other hand f Safeguards, Subcommittee on Reactot Aforing People Safely. In addition to used for other urposes proviang euch this document. AFTA notes that work is alsnah an un entood by au memben of as Opushna; Wehg under way to develop an Emergency De ACRS Subecmmittee on Reactor
(([Tl[,,$^'.* Employees gfving Procedures Document and several Operations will hMd a meeting on sonstant lockout for them
, transit des,gn and improvement guideline manuals These items will eggnah must locate themsehes so es to be blonday. August 41980 in Room 1046, i
address the problems encountered in the plairdy seen and give them so as to be clearly 1717 H St NW. Washing I
BART and other transit authority enderstood ne atmost care must be Notice of this mee: ting was published amercised to avoid taking signals that may be June 20.1980.
accidents and emergency evacuations.
blended for other trains or engines.
In accordance with the procedures AFTA will also provfde the Salety
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Board with a list of voluntary corrective when beding or shoving epe or cars, outlined in the Tederal Register on actions taken by its transit members f[.g',f[j,'*,",'
,gp\\oy October 1.19 5444 FR 56408) oral or writlen statements may be presented by because of the problems disclosed by siven mut be construed as a stop signay members of the public, recordings will j
the investigatien of the BART accident-except when empte>ee on lea 6rs car hu be permitted only during those portions AFTA reports that it is working with eentrol of air brakes. or movement is being the Urban Mass Transportation controUed by ra dio communication, or way la of the meeting when a transcript la being Admirdstration to develop a Content known to be clear.
hept, and quer,tions may be asked only Rear view mirror must not be used fo' by members of the Subcommittee.its Cuidelines forRol/ System Sofety observing hand s!gnata.
consultants, and Staff. Persons desiring Pregram Plans and a Safety Information R"
to make oral statements should notify Reporting and Ans!ysis System.These
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e or the Designsted Federal Employee as far plans will be used by the transit
[shing a tnin. engine, or can, co.npfete in advance as practicable so that a
industry to revise and update their tructions must be given or continuous individual system safety programa.
en&o contact must be maintained When appropriate arrangements can be made APTA is also conducting a Rapid backing or pushing a train. engine or cars.
to allow the necessary time during tce Transit Conference. June 16-19.1980,in the & stance of the masement must be meeting for such statements.
l San Francisco, which includes many specited and movement must stop in one-The entire meeting will be open to half the specified distance unless editional public sitendance except for those workshops related to passenger safety.
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R-60-22.from the Illinois CentrolGulf instruct 2ons are received If the instruction
- sessions during which the Subcomm!!M Rollroad Co. (/CGJ.fune Jr,19&2-
,,",,not nYu" oven"nnut finds it necessary to discuss proprittary
,et no int informatiort.One or more closed g
Response is to a recommendation issued stop imme&ately and not be resumed until
'eessions may be necessary to discuss
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May 23 following Investigation of an the mhunderstanding has been ruolved.
accident which occurred at a CAF reso contact has been rutond, oc auninforms' ion. Sunshine Act i
I Industries plant at Mobile. Ala last communication by other means hu been E.emption 4.) To t e extent practicable.
October 3 when an ICG railroad established these closed sesstons will be held so as switchman, performing switching duties, An enstne must not be moved in response to minimize inconvenience to members to raio communication unti! positive oggepggg attegance.
was killed and the yard foreman was Id'otiDe* tion is established as provided is a
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{ fen as ed 1CG to establish and implement o
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ICG states. It is our firm beli,-f that procedures, covered by appropriate any additionalOperating ales as Monday. August 4.1990 4
opersting rules, which will prevent the use of a mixture of hand and radio recommended, would not add anything e.30 un. until the concluton of Butness g
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EryJs FJderAl Register / Vol. 45. No.1?o / %ursday, July 17, 1980 / Notices During the inittal portion of the Septetraer u m 3, and la due to expire IV meetir*g. the Subcominittee. along w!th on August:4.1M.
He Deense prohl:ts e storage'of d
any ofits consultants who may be waste byproduct materist at le D
present. will enchange prel;minary licensee's facility for mere than three (3) view: regarding matters to be Condition 6 of the license states:"Any months because the packeged weste in considered during the balance of the package containing waste byproduct storsge is unsheltered frtr: the weather.
meeting meterial shat be transferred to an Observetions during various inspections ne Subcomm!ttee wi!) then hear authorized land burial facility not mon have observed that the packages in presentations by and hofd discussions then three (3) months fotowing receipt storage since October tirra, are with representatives of the NRC Staff, d the padage."
deteriorating %e June 27.1980 the Omsha Public Power District, their On June 9,1979, e shipment of inspection revealed tha' during the consultants, and other interested approximately sixty (60) 55 gsDon druma repackoging operation the licensee has persons regard;ng the Omaha Public contsining waste byproduct material found that most of the packages opened Power Datrict request for a STRETCH was returned to the Lcensee by Chem.
oo not meet current burial site criteria.
power incresse for the Ft Calhoun Nuclear. Barnwell. South Carolina, due ne continued storage of this waste t{P{ Packa qd ese cptyp i t to t, t.
m, rou, the at so a le azard h
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F er lnformation regarding items to I Office conducted on June 20.19'9 and riku$
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raw be discussed, whether the meeting has November 27.19 9. revealed this waste been cancelled or rescheduled. the and waste coUected subsequent to June packages In view of this and in the interest of the public health and safety, i
Chairman's ruling on requests for the 19*9, was stillin storage at the it has been determined that no prior cpportunity to present oral statementa licensee e facility in violation of not2ce as provided in to CFR 2.201 and and the time allotted therefor can be Condition 6 of the license. On December to CFR 30.61 la required and that obtained by a prepaid telephone call to 4.11F9. the beensee egned to seek a pursuant to 10 CFR 2.202(f) this Order la the cognizant Designated Federal license amendment to repackage all affective immediately*
Employee, Mr. Richard K. Major waste on hand, so it would comply with (telephone 202/634-1414) between 315 authorized land burial facility V
c m. and 5 pJn.. EDT.
requirements.De licensee further In view of the foregoing and pursuant I have determined, in accordance with agreed to promptly carry out the to the Atomic Energy Act of1954, as Subsection 10(d) of the Federal repackaging following such an amended, and the regulations in to GR Advisory Committee Act, that it may b.e amendment.ne license was Parts 2,20, and 30, it is hereby ordered necessary to close some portions of this appropriately amended on January &
that:
meeting to protect proprietary 1960. An inspection en February 6-10, 1.ne licensee exunine and, where 1
l Information.ne authority Ior such 1980,indipated the licensee was not necessary, repackage all waste closure is Exemption (4) to the Sunshine prepand to implement the procedune byproduct material in his possesalon.
Act,5 U.S.C. 552b(c)(4)-
required by the licence smendment observing all Commission rules.
Dated. july 14. taso, authorizing the repackaging.He regulations and license conditions, ao lcha C. Hoyle, licensee ag eed to cease repackaging that the waste complies with authorised Ah/sorycommitteeha pen Oficer, until be could comply. On April 1,1980, burial site requinments.
la a meeting at the Commission's Region 2.Following examination and..
tra n an.nm re.on er,
! Office, the licensee agreed to properly appropriate repeckaging, the licensee aus.o caos ruo.ai.as implement procedures for repackaging transfer all waste byproduct material sently in his possession to a land the waste and tranfer all waste to an
{urial facility authorized to receive the 15yproduct niaterial Ucense No. 37-t4000=
authorized land burial facility by July 1, 0tj 1980.
meterial, not later than July 18,1980.
3.The licensee shall not receive Apptird Health Physfes,Inc.; Order III additional waste byproduct material undliterns 1 and 2 abon have been Requiring Prompt Repackaging of On June 26,1980 the licensee Informed M aste Byproduct A aterlat and Prompt the Commission's Region I O!! ice that he completed.
- 4. With respect to the waste Transfer To Authorized Land Bur;al would be unable to repackage and byproduct material currently in the transfer all waste to an authorized land on.
n a
b burial facility by July 1.1980. An
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Applied Health Physica,Inc.,2984 inspectim on Juny 27,1980, revealed
. accond sentence, thus permitting the I
!adastrial Boulevard. Bethel Park.
that the repackagmg of the waste
- transfer of packaged waste bypmdoet Pe nn syh anla 15102 (the " licensee") is byproduct material was being condocted material to any authorized waste burial the holder of Byproduct Material in compliance with the Commission's facility provided the waste is packaged
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1.f canse No. 37-14500-011ssued by the regulations and li:ense conditions. no in ucordance with the requirements of
% clear Regulatory Commission (the inspection revealed that the licenses th6 facility.
E "Comm!ssion").The license authorizes would be unable to complete the
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the company to receive and possess repeckaging by July 1,1980, due to a YI g(
pckage containing waste byproduct lack of required material (30 gallon De licenses, or any othee person who I
material for the purpose of transfern'na drums) and the fact that the repackaging has an interest affected by this Order t! e packages to authorized land buriaf operation bad not been started.until seay, within twenty.five days of the data f:cihties.%e license was luued as June 23,1980, of this Order, request a hearing. A o
ATTACHMENT B ATTENDEES ACRS W. Mathis, Chairman H. Etherington, Member Emeritus D. Moeller, Member J. Ray, Member R. Major, ACRS Staff, Designated Federal Employee NRC STAFF OMAHA PUBLIC POWER DISTRICT R. Clark K. Morris P. Wagner W. Jones T. Novak J. Gasper R. Lobel F. Franco S. Bryan S. Stevens W. Anderson D. Skovholt EXXON NUCLEAR A. Berkow T. Krysinski G. Owsley PICHARD LOWE & GARRICK T. R. Robbins LEBOEUF LAMB LECHY & MAC RAE M. Tabor COMBUSTION ENGINEERING R. Mills F. Carpentino R. Bradshaw
detWT(67M9 TENTATIVE SCHEDULE ACRS SUB00'C;;TTEE MEETING ON REACTOR OPERATIONS POWER LEVEL INCREASE WASHIN3 TON, DC AU3UST 4, 1983 APPR0XIMATE TIME FORT CALH3'JN - POWER LEVEL INCREASE I.
Chairman's Opening Remarks 8:30 a.m.
II.
Licensee Presentation 8:40 a.m.
A.
Brief Plant and Site Description (any changes in population density)
B.
Licensing and Operating History 1.
Control rod guide tube wear problems?
ii. % Steam Generator Tubes Plugged iii.
Results of feedwater pipe cracking survey iv.
RPV fracture toughness requirements - changes in operational limits v.
Benefits from demonstration assembly program vi.
Experience operating under new 1500 MWt Tech. Specs.
(brief description of Tech. Spec. changes) vii.
Status on degradation of reactor coolant pump studs viii.
Future plans beyond STRETCH - length of operating cycle ix. Other items of safety significance C.
Overview on Power Increase D.
Cycle 6 Core Design E.
Power Increase Methodology Changes (if any)
F.
Transient / Accident Analysis 1.
Small Break Analysis
- 11. Why must validity of the Exxon rod bowing model for Ft. Calhoun be reassessed before Cycle 77 j
111. Current status of Control Room Habitability Study iv. Others - Brief description for the record BREAK 10:30 a.m.
REACTOR OPERATIONS APPROXIIGTE TIME III.
NRC Presentation 10:45 a.m.
A.
Introduction B.
SER/EIS C.
Reload 6 Safety Evaluation D.
Staff Comment on ACRS Letters - Generic Issues E.
Regulatory Experience LERs, Serious Incidents During Life of Plant, Radiation Exposure, Others F.
Brief Status of Changes as a Result of TMI-2 Procedures Plant Modifications IV.
Caucus V.
Meeting with the Licensee and Staff VI.
Recess of Morning Session 12:00 noon AFTERNDON SESSION:
Discussions with NRC Staff A.
Role and Responsibilities of NRC Resident Inspector 1:00 p.m.
l Increase in Direct Inspection Effort Incident Response B.
Discussion with Staff on Special Low Power Test Programs 2:00 p.m.
C.
Discussion of Proposed Revisions to Technical Specifications 3:00 p.m.
Achieving Hot Standby in One Hour ADJOURNE NT 4:00 p.m.
ATTACHMENT D LIST OF HANDOUTS AND RELATED DOCUMENTS 1.
Safety Evaluation and Environmental Impact Appraisal By The Office of Nuclear Reactor Regulation Supporting Facility Operation at 1500 MW For Facility Operating License No. DPR-40, Omaha Public Power District, Fort Calhoun Station, Unit No. 1, Docket No. 50-285.
2.
Regulatory Response Plan, Prepared by AIF Subcommittee on Regulatory Interaction, February 1980.
3.
Presentation to the Advisory Committee on Reactor Safeouards on Power Level Increase For the Fort Calhoun Station, August 4, 1980.
Omaha Public Power District.
4.
Presentation Slides:
Fort Calhoun Station, Unit No.1 Power Level Increase (Stretch Power) 1420 MWt to 1500 MWt - P. Wagner, NRC Staff S.
Presentation Slides:
Role of NRC Resident Inspector - S. Bryan, NRC I&E 6.
Presentation Slides: The to Hot Standby Condition - D. S.Skovholt, NRC 7.
Meeting Attendance List (Sign in Sheets)