ML19263D336

Response to NRC 790313 Request for Addl Info Re 1979 Reload 4,Cycle 5 License Application.Responses Pertain to Reevaluation of Transient Events & Reactor Startup Physics Test Program

ML19263D336
Person / Time
Site:	Cooper
Issue date:	03/22/1979
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Ippolito T Office of Nuclear Reactor Regulation
References
NUDOCS 7903270472
Download: ML19263D336 (11)

Text

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GENER AL OFFIM Nebraska Public Power District ' ' ' fE"$3837"M1U#** ""'

March 22,1979 Director, Nuclear Reactor Regulation Attention: Mr. Thomas A. Ippolito, Chief Operadng Reactors Branch No. 3 Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Reload 4, Cycle 5 Reload Licensing Submittal -

Request for Additional Information Cooper Nuclear Station NRC Docket No. 50-298, DPR-46

Dear Mr. Ippolito:

Your letter of March 13, 1979, transmitted a request for additional information relating to Nebraska Public Power District's January 31, 1979 Reload License Application. Enclosed please find the District's response to these four questions. These questions were informally received from the Staff February 23, 1979.

As stated in our application, the current refueling schedule still allows for plant startup on April 27, 1979; therefore, approval of the reload submittal is respectfully requested prior to that date.

Should you require additional inIormation, please do not hesitate to contact me.

Sincerely yours, kt

. I. Pilant Director of Licensing and Quality Assurance IMP /jw

srs22/3 Enclosure j

I i

79032704 9

Mr. Thomas A. Ippolito March 22, 1979 Page 2 STATE OF NEBRASKA)

)

PLATTE COUNTY )

Jay M. Pilant, being first duly sworn, deposes ar.d says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to suitnit this information on behalf of Nebraska Public Power Distr':t; and that the statements in said application are true to the ha* if his knowledge and belief.

b My'M. P11 ant Subscribed in my presence and sworn to before me t..is 22nd day of March, 1979.

j NOTARY PUBpC My Commission expires October 14, 1980.

EMEAL BITART Stats et setranta MARLYN R. HOHNDoAF

_ thCesent 4 oct, t4,1380

ATTACHMENT I REQUEST FOR ADDITIONAL INFORMATION COOPER NUCLEAR STATION UNIT 1, RELOAD 4 Question 1 The staff stated in Section 6.2.2 of its safety evaluation of the Generic Reload Fuel Application (which you have referenced in your reload application) that " Additional Data should be submitted by GE to the staff for review, to justify the conservatism of the GEXL correlation for the second and subsequent cycles of operation of the retrofit 8x8 bundles, when local peaking factors may increase sufficiently to cause non-conservative CPR calculations." Your reload submittal has not addressed this issae. Accordingly, we request you provide either directly or through reference, adequate information which speaks to this concern. Your response should include:

The extent to which individual heater rods are instrumented in steady-state critical power tests for the retrofit fuel design.

For each test bundle provide measured and predicted results in tabular form for the various test conditions.

Provide trend plots (measured critical power / predicted critical power vs.

hIN, G, P, critical power, test bundle) .

Maximum R-Factor for each test bundle (new and old R-Factor definitions) .

Thermocouple locations (rod-by-rod and axially).

Spacer-grid locations.

Provide power and heat flux for all plots of tranrient CPR cases.

Response

General Electric Company will respond to this request on a generic basis. This response will be contained in a letter, R. E. Engel (GE) to Robert L. Tedesco and Darrel G. Eisenhut (NRC), " Additional Info rma tion , 8x8R Fuel GETAB R-Factors", to be issued 3-31-79.

Question 2 The staff stated in its evaluation of the Generic Reload Fuel Application that it is acceptable to reanalyze only a relative few limiting transient events as part of reload safety analysis. The basis for the selected events stems in part from previous (e.g. FSAR) analysis results (consequences). Furthe rmo re , the relative consequences of all of the anticipated transient events considered (in the FSAR) is based on specific equipment performance characteristics and reactor protection system characteristics. It is not known whether the proposed reduction in the low pressure main steam line isolation valve setpoint (from 850 psig to 825 psig) will cause the currently non-limiting pressure regulator failure event to bece.e a limiting event. We require that you 7:ceide sufficient information that shows the pressure regulator failure event remains non-limiting, even with the proposed setpoint change. An accep ;able response would be to provide the transient ACPR (for each fuel type) for a pressure regulator failure transient analysis which models the proposed technical specification change.

Response

The proposed reduction in the main steam line low pressure isolation setpoint from 1850 psig to 1825 psig will not increase the severity of the pressure regulator failure opening event. This trar.sient is initiated by a sudden reduction in pressure demand signal. This signal causes the turbine control and bypass valves to begin opening resulting in a rapid drop in core pressure. This pressure decrease results in increased void formation. The rapid insertion of negative reactivity due to the increased void content results in a rapid power decrease. The rapid pressure / power decay is terminated by the closure of the main steam isolation valves. Lowering the setpoint will actually result in a larger power decay before the transient is terminated, the resulting peak heat flux will be lower and the CPR response of this transient will actually be improved (reference SAR Figure XIV-5-8).

Question 3 Provide the ACPR result for a fuel loading error consisting of mislocating an 8x8R bundle in a 7x7 cell.

Response

The NRC has accepted the reload submittal format contained in NEDO-24011A, Rev. 0 which provides for reporting only the most limiting bundle loading error.

In the event the NRC imposes new criteria which will make the mislocated 8x8R bundle limiting, the District will provide the results of the analysis of this event.

Question 4 It is the staff's position that adequate start-up physics testing be performed following each plant refueling in order to assure that the core conforms to the design. i.e. that the actual (measured) reload core configuration is consistant with the analyzed reload core configuration. The staff currently has a study underway for the purpose of generically establishing requirements for minimum BWR start-up physics test programs. Although this effort is not yet complete, we have concluded at this juncture that, in order to be acceptable, BWR start-up test programs, must include each of the following (or acceptable equivalents):

A. A visual inspection of the core including a photoL;aphic or videotape record.

B. A check of core power symmetry-by checking for mismatches between symmetric detectors.

C. Withdrawal and insertion of each control rod to check for criticality and mobility.

D. A comparison of predicted and measured critical insequence rod pattern for nonvoided conditions.

In view of the importance the staff places on the above four BWR start-up physics program elements, we request that you provide a commitment to include them (or acceptable equivalents) in the Cooper Station Unit 1, Reload 4 start-up program.

Additionally, in order that we may adequately assess the characteristics of the Cooper Station Unit 1 Cycle 5 start-up test program, we request that you provide the following information:

A description of the core loading verification (inspection) procedures to be followed for the core refueling including the number of independent checks to be made of a) the actual core loading, b) the intended core loading and c) the consistency between the two.

A description of each start-up physics test (including those indicated above).

The acceptance criteria and basis for each test (including those indicated above) which provides assurance that the actual core confo rms to the design.

The actions to be taken for each test (including those indicated above) whenever the acceptance criteria are not satisfied.

Response

Nebraska Public Power District will perform the four start-up physics tests described in Attachment II in the Cooper Nuclear Station Reload 4 start-up p rog ram.

ATTACHMENT II COOPER NUCLEAR STATION REL0s 4, CYCLE 5 START-UP PHYSICS TEST PROGRAM

CCRE LOADING VERIFICATION I. pURp0SE The core loading verification is performed after the core is fully loaded to assure that the core is constituted as per the design loading plan.

II. DESCRIPTION The core loading verification consists of the following three distinct phases:

A.

A scan of the core with a television camera is =ade by operations personnel located on the fuel handling bridge. The fuel asse:bly serial number Control and Root via orientation is observed and co==unicated to the phone. The Control Root verifies the correctness of the information (or asks for a reread). The person located in the Control Roce has a ec=puter generated =ap of fuel assembly serial numbers versus location. The personnel on the fuel bandling bridge =ust correctly identify the serial nu:ber and orientation before moving to the nett fuel assembly. The person located in the Control Room will or.ly verify cor:setness (or ask for a reread);

he does not co==unicate the gerial number to the personnel on the fuel handling bridge. .

B.

Concurrently with the above verification a video tape is =ade and the appropriate blocks of a blank core =ap are filled in with the fuel assembly serial numbers and orientations by,a representative of the Reactor Engineering Staff. The Reactor Engineering repre-se:tetive is viewing the verification on a monitor located r~e=otely from that used by the personnel on the fuel handling bridge and is also in continuous contunication with both the fuel' handling bridgc and the Control Room. He must agree with the fuel assembly serial number and orientation as identified by the personnel on the fuel handling question.

bridge or he calls for a reread of any fuel asse=bly in He also marks the beginning of each row scanned with voice identification on the video tape for subsequent review. The core cap produced during the video caping is then checked against a computer generated core cap of the design loading for correctness.

C.

After the whole core has been scanned and taped, the video tapes are reviewed by a three =an team (different personnel than partic-ipated in the above two phases) as a final check. Teat menber ene is responsible for operating the tape recorder and observing fuel assembly ,erial numbers; member one also serves as a quality con-trol check 'or member two's activities. Tean cember two views the tape and calls out the fuel asse si,.

in sequence. serial numbers and orientation Tea. ec=ber three writes down the called fuel assembly serial numbers in the appropriate block of a blank core nap. Member three also indicates on the core cap that the orientation is cor cet after member two verifies that the spring clip location is acceptable (orientation as it should be). After cach rcw is revieued and the fuel assembly serial numbers entered on the map, member three rec .

back the recorded fuel assembly seri:1 numbers and member two verifies these nu=bers against a cceputer zenarated core map of the design ionding.

III. ACCE?TANCE CRITERIA The core e.ust be leaded according to the design loading pattern (or a fuel vendor approved variation thereof to acco==odate dis-charged leaking fuel asse=blies).

IV.

ACTIO :S TO BE TAFE:i IF CRITERIA NOT SET The core will be rearranged to conform to the design leading pattern areasthe and of core loading verification will be redone for the affected the core. -

CO:ITROL RCD CRITICALITf A 'D MOSILITf CHECR I. PL*RPOSE This test is performed to assure that all fuel asse=blies are properly loaded and that all control rods are operable.

II. DESCRIPTIO:I This test will be performed af ter the core loading verification has been co=pleted; this will provide assurance that an inadvertent crit-icality will not occur due to fuel assemblies being =islocated (it is very unlikely that a fuel leading error could result in a situation where criticaliev cos'.d be achieved with a single control red being withdrawn). Perfor:ance of this test will assure that there are no fuel asse=blics loaded so as to affect the move =ent of a control red.

If =ove=ent of a control rod is affected, it could be caused by such things as a rotated fuel asse=bly, a fuel asse=bly not being properly seated, et2. Each control rod in the core will be withdrawn and in-serted tc assure that it can be =oved' with normal drive pressure. Also, the nuclear instrumentation will be monitored during the =ovement of each control red to verify suberiticality.

IIT. ACCEPTA2:CE CRITERIA Each ce= trol rod must be exercised (fully withdrawn and inserted) one at a tL=e to verify cobility under nor=al drive pre'ssure (nor=al drive pressure is plant dependent); suberiticality will also be verified.

I7. ACTIONS TO 3E TAKEN IT CRITERIA NOT MIT For those control rods that will no" move under nor=al drive pressure, appropriate repairs or adjustments will be made so that the nor=al drive pressure criterien can be =et. If criticality was to be achieved by the withdrawal of a single control rod, the concrol rod would be inserred and all further startup activities would cease; the fuel loading verification would be redone and no further red cove =ents vould occur until che situation was satisfactorily resolved.

INITIAL CRITICALITY PREDICTION

1. PURPOSE This test is performed to cenpare the predicted and reasured insequence critical red pattern for nonvoided conditions.

II. DESCRIPTION The initial critical control rod pattern in the cold Xenon free con-dition will be predicted at the beginning of each fuel cycle using data supplied by the fuel vendor. Control rod wor:hs will be deter =ined frca rod worth curves provided for the shutdown cargin de: castration.

The reactor will be brought cri:ical utili:ing the ccatrol red sequance deter =ined by the fuel vendor such that the withdrawal of the analyti-cally strongest control red will occur before criticality is at:sined.

A =oderator te=perature correction (supplied by the fuel vendor) will be applied if criticality eccurs at grea:er than SS 7. The actual and predicted critical red pat: erns will be ec: pared.

III. ACCEPTE CE CRITIRIA Acceptance criteria of +0.5% ak/k and -1.5% will be applied. The

+0.5% ak/k criterion is taken from the see:icn of the Technical Spec-ifications describing reportable occurrences (Reactivity Anomalies).

The -1.5% ak/k criterien is arbitrarily set so as to be consisten: with the 2% ak/k window allowad for the steady sta:e pcuer reactive ano=aly check (11% ak/k). .

17. ACTIONS TO BE TAKEN IF CRITERLA NOT SET ,

If criticality occurs 0.5% ak/k before the predic:ed'cri:ical position, actions vill be taken to deter =ine the cause and the event will be reported as per Technical Specificatica require ents. If criticality occurs more than 1.5% ak/k after the predicted cri:ical positica, actions 2111 be taken to deternine the cause of the event. In both cases, this actica would probably censist of requesting the fuel vendor to recheck the red worth curves provided and if necessary, have the fuel vender renormali:e his ccre simulator calcula:icns so that more accurate rod worth curves can be provided. Startup would continue and all ther=al li= ins would be closely monitored to assure cenpliance udth all Technical Specification Liciting Conditions of Cperation.

, TI? SY}CE 2Y CHECK I.  ?"?}0SE This test is performed to measure the total TI? syn =etry and to compsre the readings of individual sy==etric pairs of T1?'s.

II. DESCRI?TIO:i Typically, during a refueling outage, fuel is both removed and shuf-fled to acco==cdate the, core loading plan for the subsequent cycle.

Also, local power range conitors (LPRM's) are replaced as necessary due to their having reached end of life conditions. This test deter-nines the cagnitude of the TI? system and neutron flux asynnetrics af ter the core has been rec =nstituted according to the loading plan.

The reactor will be above 75% thernal power and operating with an ec:an: sy==e:ric rod pattern for the performance of this test.

Addi-ticnally, the reactor core should be at steady-state conditions es-re::ially free of .Tanon transients and enpected to rc:ain at stcady-state conditi=ns for the duration of the tes:ing. Da:a vill be taken u-4:ing the TI? syste= hardware and statisticall; analy:ed using

=eticds provided by the fuel vendor.

Ill. AC.TJ2:CI CRITERIA 1 criterion of 12% will be applied to the total T1? reading uncertainty; this 12% criterion translates into a criterion of 10.55," for total TI?

ins-- -=

uncertainty (randos noise and geometric cc=ponents). These values were deter =ined utilizing General Electric developed =ethods for deternicing the total precass co puter uncertainty.(NE00-20310). A cri: erica of 23% will be applied to the individual sy=:etric pairs; this was deter ' ed by =altiplying 6.5% by /2 by 2.5 standard deviations.

The 6.5% value is the total TI? instru=ent uncertaint'y that corresponds to a total TIP reading uncertainty of 8.7%; 8.7% total TI? reading un-cer:ain:y is that value used in reload licensing submi::als.

I',' . ACTIC::S TO 3E TAKE': IF Ch.AIA NOT SET Startup activities would continue and all thermal limits would be closel-

/ =onitored to assure co=pliance with all Technical Specifica-tion Li= icing Conditions of Operation. The responsible test engineer vould exa=ine the test data and analysis r.ethod carefully for errors.

If no enors are found, the TI? system hardware would be checked with e=phasis initially being placed on the checkout of TIP axial calibra-tion and align =ent. If faults are found, they would be corrected and the tests repeated. Another thing that could be checked if no faults are discovered would be the fuel channel location history for those channels located around the affected TI?'s; channel bowing due to non-unifor: neutron flun levels can affect the water gap dimensions around the TI? detector. This would in turn affe:: the power calculated for one =c=ber of a sy==etric TIP pair; analysis by Cenaral Electric has denenstrated that the probability is high that thermal liaits will be af f ected conservatively by such TI? asyn=e: ries. If no hardware faults or calculational errors are focnd, operation would continue and all ther=al liaits would be closaly onitored.