ML19322C180

Results of Oconee 1 Hot Functional Testing,Internals Vibration Monitoring Program

ML19322C180
Person / Time
Site:	Oconee
Issue date:	02/15/1973
From:	DUKE POWER CO.
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ML19322C181	List: ML19322C180
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NUDOCS 8001090593
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{{#Wiki_filter:[ RESULTS Of OCONEE I E T FUNCTIONAL TESTING INTERNAI.S VIERATION MONITORING PROGRAM 1 INTPODUCTION 1.0 _. 1, " Prototype Vibration Measurement Program for Reactor h EAU-10038, Rcy. Internn'.c (177-Fuel Assc=bly Plant)," outlined the program to be impic- [ mented rt Oconce Nuclear Station during hot functional testing (HFT) of Unit 1. During the period November 1972 to January 1973, this program a was successfully carried out. Thc prr"ose of this report is to summari=c the results of the vibration monitoring program to confirm the design of the reactor interaals as preocutnd in BAU-10051, Rev. 1, " Design of Reactor Internals and Incore Instrur.nt Nozzles for Flow-Induced Vibration." 2.0 COLLECv.ON OF DATA The rceponse of the reactor internals and incore inctrument nozzles was In addition, there l 'noaitored by 40 strain gages and 12 acceleromeccrs. rere pr2ssure cells attached to the core support shield and thermal shield i ?ressure sensing to deteruine the static and dynamic pressure variations. lines ; cre nico connected to an incore instrement neszie and mating guide irbe to provide data for calculating the cross-flov velocitics in the vcaccl lower head. Figure 1 shows the instrument locations. Strai gages were placed on two incore instrument nozzles, on two incore . :str ient guide tubes (one gussette_d and one non-f,ucor.tted), two th.ermal

.hicle lower support bolts, six thermal shicla support pads, at two "occtions on the plenum cylinder betueen the outlet holes, on one Accelerometers surveillance specimen holder tube, and on one chroud tube.

=re placed on the thermal shield, inside a guide tube to record the External actions of the flou distributor, and on the core screen. accelerometers were placed on the vessel support skirt, and on a vessel icad stud. The accelerometer on the screen was utilized to conitor for icose parts, and those external on the reactor vessel were used for Seven pressure transducers were located around the outside ref eren.cc. didcaterof ^ the thermal shield, and four around the outside diameter of the corc support shield.. Pressure sensing lines were placed, four each in an incore instrument nozzle and four each in an incore guide tube. All of the sensor output signals and the conditioning electronics were s in r.ood uorking order uhen the reactor coolant pumps were initially operated. For the duration of the prcoperational testing period, the scncers and the data acquisition system continued to uork with very l Some data channels, however, did become good reliability and accuracy. inoperable during the test. At the end of HFT, there were 14 channels (uostly strain gages) that were inoperative. The failure of these signal channels did not, however, prevent the obtaining of sufficient data to assess the structural adequacy of the internals. 4

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~ During RFT a total of 280 hours trere accumulated (logged) with four reactor coolant pumps operating (full flow conditions) at conditions 0 of approximately 530 F and 2155 psig. Consequently, the thermal shield which had the lowest measured structural frequency of 12 Hz, was sub-7 jected to more than 10 cycles. The reactor internals were also subjected to an additional 280 hours of conditions other than full flow operation. This was sufficient to accumulate an additional 107 cycles of vibration on the thermal shield. Outputa of the sensors recorded on magnetic tape were continuously monitored by an on-line computer. 3.0 RESULTS Tabic 1 ccmparcs the stresses and deflections measured during the hot functionni testing vibration monitoring program with the predicted and allowa* ole values as presented in BAU-10038. As indicated by the large , ratio of allowable values to ucasured values, a substantial acceptance - margin c::ists for all components. It is noted that the predicted responses of the various components were calculated using conservative criteria. Coaccquently, the generally lower values of measured response are not unexpected. 4.0 POST iiOT FUNCTIONAL TESTING INSPECTION Follouing the conclusion of HFT and the concurrent vibration monitoring program, the reactor internals were removed from the vessel and the post-test inspection program was implemented. The purpose of this 1 prograu was to visually inspect all major internals, surfaces and/or parts for any indications of distress, loose parts, cracking, fretting, or distortion as a result of HFT. i The results of this inspection program indicated that the reactor internals sustained no structural dausge as a result of EFT. It was determined that no deterioration that might affect the structural integrity of the internals had occurred. Daly expected, minor indi-cations, of fretting and vcar were observed on metal-to-metal con; tact surfaces"; 'c'.g.., internals vent valve seats. It wau discovered, however, prior to UFT, that the surveillance specimen holder tubes required more torque than anticipated to rotate them for interaals innertioa into the reactor vessel. The same situation existed { after HFT. This cituation was invcotigated as part of the inspection routL,e and it was determined that bearing and tube misalignment were the cauce of the increased torques. Corrective actions ucre taken: (a) The surveillance tubes were straightened to improve alignment. Eccause of schedule, the identical surveillance tubes for Oconee 2 ucre inatalled on Oconee 1. The Oconee 1 tubes will be used on Oconee 2. (b) The installation procedure -was modifiei to improve the alignment. l

.s Thiscituationdidnotaffectthystructuraladequacyofthesurveillanca tubco. 5.0 cc::ct.LSIONS .; a Evaluation of the measured vibrational responses and the direct visual examiration of the reactor internals has revealed that the structures experienced extremely low stresses during operation and sustained no structural damage. All data indicate that the internals will perform in the intended manner with no adverse deterioration occurring over the extended life of the unit. S e e e s% e S O I l e 0 (8, e \\*9 = y e e g ,e 9 k ~*1= a-

^ s. VIEIMTION TEST RESULTS ~ $$asurEdValues -; Predicced Accepeance Ratio of A11ouable Component ' gjs penk-to-Peake2 Values _ Cr_i,t_eria __To Measured Qcah-to-Peakf7)I ei es. Nozzle <100 psi <100 psi 3,000 psi 5,800 psi 758.0 Guide Tube: Gossetted <100 psi <100 psi 2,050 psi .7,000 psi >70.0 Non-Gussetted <100 psi <100 psi 2,700 psi 8,600 psi >86.0 Flow Distributor .002 in.. .003 in. .011 in. _j .025 in.; 8.3 4 8 E Surveillance Specimen i Italder Tube 250 psi 300 psi 10,000 psi 13,500 psi 45.0 i Upper Thermal Shield Support 200 psi 300 psi 4,150 psi 13,500 psi '45.0 Lower Thermal Shield Bolt 760 psi 3,900 psi 3,500 psi 7,900 psi 2.0 I e a g t i j. K .h

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DUKE P 0 W E'R C0MPANY 1CONEE NUCLEAR STATION UNIT 2 DOCKET NO. 50-270 LICENSE NO, DPR-47 STARTUP REPORT ( SUPPLEMENT 1 i i NOVEMBER 7, 1974 l k +- l -. = -. -., +, -

~ s l 51.0 SUPPLEMENT 1 i On August 9,1974, the Startup Report for Oconee Nuclear Station, Unit 2 was e submitted. This report addresses unit startup and power escalation testing through 0500 hours, June 24, 1974. At that time, all or part of several power escalation tests remained to be completed. This supplement is prepared and. submitted with regard to the period after the above date and prior to 2400 hours, October 31, 1974. During this period, no power escalation tests were completed. e i / 1 e 4 i e i s t G S1-1 l. .m p

RE-ESCALATION TO POWER Table of Contents Section Pg A

1.0 INTRODUCTION

A1.0-1 A2.0 TESTS AND OPERATIONS A2.1-1 A2.1 FUEL RELOADING A2.1-1 A2.2 CONTROL ROD DRIVE DROP TDfE TEST A2.2-1 A2.3 REACTOR COOLANT PUMP FLOW TEST A2.3-1 A2.4 LOOSE PARTS MONITORING SURVEILLANCE TEST A2.4-1 A2.5 INITIAL CRITICALITY AFTER FUEL RELOADING A2.5-1 A2.6 NUCLEAR INSTRUMENTATION OVERLAP A2.6-1 A2.7 "ALL RODS OUT" CRITICAL BORON CONCENTRATION ^ A2.7-1 A2.8 CORE POWER DISTRIBUTION A2.8-1 A

3.0 CONCLUSION

S _ A3.0-1 f f f i j e f i i i I S \\, l A-1

A

1.0 INTRODUCTION

On January 4, 1974, during power escalation testing at 75 %FP, a malfunction in the switchyard caused a turbine / reactor trip. Following this trip, a foreign object was detected in the bottom of the reactor vessel. After evaluations by Duke and the Babcock & Wilcox Company, and with Atomic Energy Commission concurrence, power operation was resumed. While at 15%FP, a seal failure in Reactor Coolant Pump 2B2 occurred. Since repairs required the unit to be shut down for an extended period of time, it was decided to remove the fuel from the reactor and inspect the reactor vessel interior and the vessel internals. Due to this incident, further unit' operation was delayed until May 23, 1974. This section of the report covers the events starting with fuel r,eloading until unit operation at 75 %FP under the initial power escalation procedure. The specific tests or operations reported in this section are listed below: (a) Fuel Reloading (b) Control Rod Drive Drop Time Test (c) Reactor Coolant Pump Flow Test (d) Loose Parts Monitoring Surveillance Test (e) Initial Criticality After Fuel Reload (f) Nuclear Instrumentation Overlap (g) All Rods Out Critical Boron Concentration (h) Core Power Distribution Each section of this report addresses a specific test or operation which was conducted. The analyzed results are presented and comparisons made to pre-dicted values, to previously measured values and to the safety criteria limits. e 6 4 e A1.0-1 i-k'

V A2.0 TESTS AND OPERATIONS A2.1 FUEL RELOADING Fuel reloading was initiated with the insertion of fuel assembly 2C52 into the core on March 25, 1974, and was completed with the loading of 2C53 on April 2, 1974. The final configuration of the core at the conclusion of fuel reloading is identical to that of the initial fuel loading and is depicted in Figure 2.0-1. Neutron count rate was monitored throughout the fuel reloading sequence, utilizing the unit's two nuclear instrumentation source range channels, NI-l and NI-2, and two temporary incore detectors. Independent calculations and plots of inverse neutron count rate ratio were maintained from the, output of these detectors. While loading the first several fuel assemblies; dif-ficulties were encountered with the response of the two temporary incore detectors and fuel loading was halted and an evaluation made. Investigations showed that gamma saturation was the cause of the unusual response and the detectors were replaced with fission chamber type detectors which could operate in the presence of a high gamma field. After replacement of these detectors, no further problems were encountered with the temporary incore detector system. Fuel reloading of Oconee Unit 2 was completed in nine days; except for the difficulty with the temporary incore detector system and minor equipment problems, fuel reloading proceeded in an orderly manner. I e I i I A2.~ 1-1.

A2.2 CONTROL ROD DRIVE DROP TIME TEST The purpose of the Control Rod Drive Drop Time Test was to verify the inte-grated, functional trip capability of the Control Rod Drive System and to determine for each control rod assembly, the total elapsed drop time from the initiation of a trip signal until the control rod assembly was three-fourths inserted. This test was conducted at various combinations of reactor coolant flow, pressure and temperature as follows: Test Condition Flow Pressure Temperature 1 No Flow 3,350 psig i 400 F' i 1700 psig 2 One Pump Each Loop 3,350 psig i 400 F 11700 psig 3 Four Pumps 2155 1 30 psig 532 1 10 F At each condition, Control Rod Groups 1 through 7 were driven, sequentially, to the fully withdrawn position. A manual trip of all control rod drives was then initiated and simultaneously a time signal was provided to the data logging equipment. As each control rod assembly reached the three-fourths insertion position, a second time signal was provided to the data logging devices from a switch located on each control rod drive's position indicator tube. The total elapsed time from the initiation of a trip. signal until three-fourths insertion was then determined for each control rod drive frcm the data acquired. The test was repeated a second time for all control rods at each test condition. The total drop time for each rod by core location in milliseconds was re-corded along with the date, time, number of reactor coolant pumps operating, reactor coolant flow, reactor coolant temperature, and reactor coolant pressure. I i An analysis of the rod drop times for the three test conditions stated above is presented in Table A2.2-1. As can be seen, the fastest dropped rod was B-08 and the slowest dropped rod was M-09, except during one drop where 0-09 was the slowest and M-09 was the second slowest. The rod drop times were well below the maximum acceptance criteria stated in Section 4.7 of the Technical Specifications, of 1.66 seconds at full flow and 1.40 seconds at no flow conditions. Also, as would be expected, all drop times under flow conditions were longer than under no flow conditions. l A2.2-1 i

Control Rod Drive Drop Times For the Fastest and Slowest Control Rod at Each Test Condition. Unit Drop Fastest Rod Slowest Rod i Condition Number Location Time (sec) Location Time (sec) 1 1 B-08 1.132 M-09 1.194 1 2 B-08 1.129 M-09 1.194 2 1 B-08 1.185 0-09(1) l'. 2 5'l 2 2 B-08 1.192 M-09 1.256 3 1 B-08 1.248 M-09 1.324 3 2 B-08 1.235 M-09 1.307 f 1 Note (1): The next slowest control rod in this case was M-09 at 1.247 seconds. i 1 } 5 I e i .i ~. Table A2.2-1

A2.3 REACTOR COOLANT PUMP FLOW TEST The purpose of the test was to determine the functional capabilities of the Reactor Coolant System and the reactor coolant pumps for hot flow test conditions after unloading and reloading the core and to verify that the e reactor coolant flow was approximately the same as measured earlier during the performance of the Reactor Coolant Pump Flow Test, Section 3.1. Reactor coolant flows were determined for the pump combinations listed below by means of both the unit computer and loop flowmeter AP cells. Run Pump Operating Number Al A2 B1 B2 1 X 2 X X X X For each pump combination, five sets of steady-state data were read from the computer, and the indicated flows, temperatures and pressures were averaged and these average values used to calculate properly compensated flow values. From the loop flowmeter AP cell indications, flows were calculated as follows: h Flow = C AP g g Vs Where: Cf = Flow Meter Coefficient = 397,100 AP = Indicated AP Vc = Specific volume at reference conditions (68 F, 14.7 psig) Vs = Specific volume at system conditions Table A2.3-1 gives the minimum and maximum flow rates for these two different pump combinations, along with the measured flow rates obtained during the performance of this test during zero power physics testing and during re-escalation to power. It can be seen that the measurements of flowrate are within the acceptance criteria and within less than 5 percent of each other. I i O e G I l l A2.3-1

... e. 4. Comparison of Reactor Coolant Flow Minimum Maximum Measured Measured Acceptable Acceptable Flow Rate Flow Rate Flow Rate Flow Rate During ZPPT After Fuel Reload Deviatian p,,p 6 6 Combination

(10 1bm/hr)

(10 lbm/hr) (10 1bm/hr) (10 lbm/hr) (%) Q A1, A2,.B1, B2 138.4 151.4 146.6 142.0 3.1 5

B2 None None 34.5 32.9 4.6 D-L

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A2.4 LOOSE PARTS MONITORING SURVEILLANCE TEST The purpose of Loose Parts Monitoring Surveillance Test was to verify that no loose objects or non-characteristic noises were present in the Reactor Coolant System during unit re-escalation to power. The test was conducted at various combinations of operating reactor coolant pumps, Reactor Coolant System pressure and temperature, and power levels as follows: , Test Pump Pressure Temperature Power Conditions Combinations (Psig) (OF) (%FP) Cold A1, A2, B1, B2 < 500 < 250 0 Al + A2, B1 + B2 Al + B1 + B2 Hot A1 + A2 + B1 + B2 2155 500-532 0 Power Al + A2 + B1 + B2 2155 579 40 Al + A2 + B1 + B2 2155 579 75 For each pump combination listed above, a minimum of five minutes of test data was required at cold and hot conditions, and a minimum of ten minutes of test data was required at power conditions. In addition, data were also taken during pump starts and coastdown. The technique of analysis was to listen to each of the 11 active channels which were connected to the Loose Parts Monitoring System and to verify that no unusual noise's were heard. Also, the frequency content of the noises heard from each accelerometer was checked using a spectrum analyzer. Data from all four pumps, both steam generators, and the top and' bottom of the reactor vessel were recorded during pump starts and coastdowns. Analyzed test data at all test conditions indicated no noises which could be characterized as loose objects in the bottom of the reactor vessel or the top of either steam generator. I i l o A2.4-1

A2.5 INITIAL CRITICALITY AFTER FUEL RELOADING Criticality was achieved on thy 23,1974, at 1337 hours at reactor coolant 0 conditions of 532 F and 2155 psig. All control rod groups were fully withdrawn except for Group 7 which was 59 percent withdrawn and the critical boron concentration was 1600 1 10 ppm. The procedure for the approach to criticality is briefly described below: (a) Control rod group withdrawal: Groups 1-4 100% withdrawn Group 8 100% withdrawn Group 5 100% withdrawn Group 6 100% withdrawn Group 7 75% withdrawn (b) Deboration from 1854 ppm 3 to 1680 1 50 ppmB using a feed and bleed rate of approximately 70 gallons per minute. (c) Deboration from 1680 1 50 ppm 3 to the critical boron concentration using a feed and bleed rate of approximately 20 gallons per minute. Control Rod Group 7 inserted to maintain criticality is required Throughout the approach to criticality, two independent plots of the inverse neutron count rate ratio were maintained. At the end of each reactivity addition, count rates were taken from each startup range neutron detector. The ratio of the initial average neutron count rate to the average neutron count rate at the end of each reactivity addition was the value plotted. As indicated above, deboration was carried out in essentially two steps. First, deboration from an initial boron concentration greater than 1800 ppmB to 1680 +/- 50 ppmB was accomplished using a 70 gpm letdown and makeup rate. This step brings the Reactor Coolant System boron concentration to within 65 ppmB of the predicted critical boron concentration (1615 ppmB). Boron concentration changes during this period were determined by taking reactor coolant samples at least every 30 minutes. The second deboration step was accomplished at a lower makeup and letdown rate of.20 gpm. Again, Reactor Coolant System baron samples were taken at least every 30 minutes, and in-verse multiplication plots were maintained versus boron concentsation and versus time. The above procedure was followed with the exception that during the second deboration, just prior to ' criticality, Control Rod Group 7 was inserted and deboration stopped in order to check an indicated high core flood tank level. Af ter. resolution, criticality was achieved by the withdrawal of Control Rod Group 7 to 59 %wd. The plots of inverse neutron count rate ratio versus-time, boron concentration, and Control Rod Group 7 position are shown in Figures A2.5-1, A2.5-2, and A2.5-3, respectively and utilize data from each of the source range channels. A projection of the boron criticality end point before the withdrawal of Control Rod Group 7 was 1612 ppmB as shown in Figure A2.5-2. A2.5-1

During the deboration cycle, some discrepancy between measured boron samples were observed. Reanalyses of all measurements that were ambiguous were performed and excellent agreement was obtained. A plot of Reactor Coolant System boron concentration versus time is shown in Figure A2.5-4. In summary, criticality was obtained in an orderly manner. Analyzed results indicate good agreement between predicted and measured criticality endpoints. o 6 I D i e \\ s A2.5-2 r-

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..r.mt = t _.._. g5,

:

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l== = t

t =: cc 4 -- i.. --t-n r== t=J=. =

U2 M t=+-- p==j= =:_n{szt:0 _L_.== 4

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- - - ' ~ .:- ninn H---* ~~ in

;

.:n.;m.

==.. 4 an . ; =.. ". n tur.*4 =rr F..=..=..... p =.. _. _._...~. t.......

n. =... n...i. f.=....

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n.. In. n "r... !n..

h .... I. n... .t.=.. .d g ..f N u 8

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u = -aa nl:l=

mt=

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. n t a..t=nul

n er

• n.p:

L*n =

-.!

.~

- * * ~ * - '*:~ In-n.2 - :: = . :.p:

dC

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g g =.... n..:. n.P.=... =....+tn"_. =.... =....

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[ d

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.x n n.

>* O

j== n.nd:. =n r

nst utn.;= tn: 2:.;

t

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k J'" U Ch! Ei!!kk(b!EE Ib

---*- $2 2 03.C NN NI N'-I NhfN N

h: n:.:-== n --t ~-

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t

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t

::1:.

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• 4 C

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d. hit Nii!I: iiNINE : 0 NNfN'..h i fi

!l ga.

..n=: w:$Mg.331.;_=_:. r=

O C C .n nj=a

n ru an

.;; un C O

n.

nn L ::

n .u nn un u U U_ f. u O.gw nn ::.= n ../

n. ;

r-- -r=:nc~~- -~"tn : O sj== u

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n= :n- / a.

n.a gg u g 4

+ - - j: n n ut=- A O t-.n =

@ n=:

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;

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t'

: t, n

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• =

11

tc - r l

t"

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r at.. .. n hn "".::-

S

": n n n n m % C,S, r -- i

n j#n4 aln n4n nIu : nni =r-l=n ain.. : =b===+f=-- n=m s

n 9 :: =n r+=r r~

n

= m e =.p:=.. =.a..p" : a-ahn - nt:f%n.... : fu..... nk:.m =t=(c{n.. .. m== ..ub n.........t--..=.:. . B= ..:.v as O g a

~t3 w=

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e.. : m. :=..n..

. en. mi.l.ii.s_ 1' - cc si. Hi.i. i.s.

u.. ;. :.

i. . t. t, "4 s

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a... ::.:t. r.

t.t.u.. n..:.:.: n.: .I N5!5! i'dliP 4'i !!ll.

li!

iddj i:l!! iii!h:" iil!j.. Eh!! illilii}[difiUNk5 !!.'i.i Hii Fj:i

lin ifi iiji!!;

il:i.

i!i@i iii}ii ii.:.]i.i. : IE. m.i. i.ih-iipi.E.. il. n 'l.F' .. i .i y n 4 nyn:: ::5i.= : nf=uj n ,.n ajn.....:ln ... e n nl;n onqh. r[w ga -..w..

.n:mn ng "n ene rc

.EI!I!!!MhU n!!hi.NI@ ! di iMi 'dli!I4i!!bi!! iIdii IIIE!$IIIIE ' UIi!!E W/T 'uopesyI ping as.taAuI d Figure A2.5-2

e ....,...h =...................... {u 3_. _.J.....j;...._.........q.. _3..+ + -

r. :

t= =;g: ;nT,._. _.= n_+.d =.-- _..,...i-in 4:-!=.,;- n!=.... Ei:7 =;i !"-=Hii".@:=. t=-. = fiIi;w g--5 5 -.. _..___.fhrt : - E 'E-V.;g +."@ = FinMigi ._.. s........... . m.- + ..........n..i._n__n_.

.-.. _. ~.... --- -. ~. - _ +:-

. =. _..._..I _L. :+-~ .r.._-. ,_..'.n...l..r =.... tar =... +... - ..t.=._ t , ag ..u tu n Inn c:.;;2 nntrn '=~"~ ~~;"- C C2g="~U *g:^^2F inn: L. u_g= ns:ug g g =:n..=;=g=1Z ~ "~2 ~ ~ ~ ~ - s "~ W ~ # D* - l 3 m....au =:nn .=2=

;

=:n= un=r nut =:-s T- - -- ~*: ~ ~. " - - O ~ - - * - ~. t- ~ ~ * ~ ~ W

r~~.in r.

n:

pp nt=

:

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-+-*

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=n =:rr

~~ - 4 ..a m n.= p4 = n"..... =.. _. = r:c : x: nn..... n... - =;.

t=

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.....=;=.

l M ..r4:m--- =r-n-H M --~: r n --: --~ ' = := .=n 2 ..5. g M ._. !=.. _[_.n g ~ d = +nn 2 :n._. _.......... -I.....-"-- a t= ~o Z- _....... =r= CG C _t-- 3 nu ' ' - --=;== = = = nn[===jn:== - ~ ~ - 4 H c ** n an; n an; = a.. :.= = t===

. :_n u:=

tun

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t

r r= :.n t= =:=: ntn-

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SEN5b NhE5 IEN N5I5fibEhNN5E! h5 !!bEI '55-f E_. . MN y-t:= g ~ y 2M22b =g... " ~- I2

• O

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2..~,. "..~. ~_2?...E..

...1..._ oa =..t.-' 7. ;r=. ..4.. _w__ g 2.

2...

go Fl=iri JEj=i4Hja =i!!= Eiiiiir lig-jp= =j"" =i-{ipipiti!"2 =Ef2E =gs y.0 g =

• i n ;g: glii-jii:E i-%i

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hs -g"i-ti

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:= = ja..

u tn.. ..ntur m.;= 4 0

- =

==t==. n, p<

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==

=t

=

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nan;

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rn"-

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a..

. =g =

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l=== = = = = u==n t -=

==: "....i_.- --- m

=a=

- + - nu =

2.1=

tn~ =-t

n. g..=......b...=.1..

C t=:= ...t_.. g _.n o g_...........n..

g..............

a. .,c., o g ..;7....n ...t...... ............. _...=._...n _I... ._.t..... 3.... .....n. y 9 ..4....

== .n,.= ~ ml 7 x..i2. _.gu.l.m n .= p,... x.;. _. =/s,;? r= =.:t =n,i=n - nh-. m.

t

t=

=

n

m. ;. _....., _.

-m = =.. _.. o r-n -m-u r:r m. = r-- r .u m _3_ JE+C T-2 ~~ t EE...f!E... +:1:k...s .....T....i2 5:iE2 9i@is istis =EtEi- =Eti=i==I.=- r o _.y. - n-

zii

!E i:EC

= =- -= c u,n y== :,._.;_n:n4==n===gtm =l:-n rn:== rw===:==. ;g=an==h= =nnj==- + 2 u:= =- .4..

= e u t==

o ,s - t= ~ r-- --- :n C u u mr =n . ~ -.... _-. t:--l--+-.=.;=.= _=__tu n.... .u.,_: ar -

t a= :n

=-

=._ A / = =.

w...

5 nt= IN 5!E4EYN 5IM5S555" EWE:E:i: LEE MEE MIME E"E'i: _ ~.. ~ =t.un

4) O*

-- m = ~n nat= un;=... n = = =n muy p=_..c.=:.. 2.x unlur = tuu,=Tn n =n.= =tu r - - + - - - t= .n- =

.:n-an..r== u..

n uoa 2.,

===t== tan

==h~.: _..... __. ;.. =i= =.... -=.._.;.;

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nu a=

e a n t==n t --

i=.a"l==:

2.- = - m:

m s:m

n m

na-n.n-- Nil!PMME"!i55EI-~ - ME"4SEEl5i5MfEi5ti"diEi+5 tit5ii@!i!E i 5EM5i!! i MiE'85ii5fEiiEYiii"if5iiM954F" WDii Ehii$5Ei i 9ilECiii!!~EE:SiE%E1-. __ G d g/I *uopecTI ping esaanuI t I 25 Figure A2.5-3

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• n uninnl=unn'

-.=1-. ~ ~ ---t== . nt= t== , _n,=3 .7 n;-+. - +. - t- =t"nin n t n .. pn;;-_;=g4n;. ;; ;Ihr

• =.nr: = t---*-r =t /i.:r-tr.

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==t=

=trj nn;=="- ~~ 1.n nn =n= nnh= =* nn =t nn "-- t _.. nutun an{=. .:*...= nn;;;n=nr r-t = rirr .:n := ~ =;:- e

unzuuun;nn

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t

.- 00. =.... -... "... *g"; ;. g".;.=..

=...

..;n.; 9 . ; ;.: = n-- n-gn., n.;d-.yE_-._..._~_.-- ~6 ~ ~ + c

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• + - - =$=

O g; y g

=."}u

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-- t --- -r. = =r==t- :

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= n ::

k h. 37-

:nn =*

• =!=

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== nn+= nninn nn+=

n::::= nng in u : =... ann

. ir -----t.#._ m_ 4 = t-~=. _ d cr ....r='= .._u.

W ZU Q

k _"u.... ~={=n

t=- un = =r

--r . =+_-*r ' n1== **._.i. -*~ t -.a._. J. _-

=t=

nr t== t--*: '=t= .nr M-- nnann n-I nntun

=

~- =!nuln r-

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• u=

x :n pnf==.-=.i =. ::g-t=

..t

g eo

= =

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~ 'E_.-

O ~E2= CIC nut"... " ".. nn:r',- =

t

g,, g g .,._.7_ kg

==t..__1 .. _.t = = t.= -?:. =tn - -,.= =tn.. ..t2 -it-t--- - - - ^ =;nny} b... u.

nr =t=

nu+= U

=7.

=t =j=n-t a H

t

-:.+.:

t

.i

• .zt=

M > ad ~ ~ M.._- b __.g.._..... _.. f w.. E

=t= - nin-

ra = C'E'~ rr t = ' rirr+-W -if..

h[ ... _.._.i-n=t==

f;t

==n2== t= -1= t---t=. 2-*=

wen-0 F(.= t urin==n

=t: u'

tr rnul

~- =".i n Y~t-nl: -~

=...=- '- ""t un = E^~- -

~< t= = uk 8 = +n-- t-n ::nt _. --+- 4 8

n nnt

- - n~ n .. n * =:

_

-- t=n nn**n n

=:nn un+I=n.

I-+- e

• d n +---

~._-*.

- - - - - - - + - - 4u .r-t--r

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nn

=:!n=

=r-m C t. - -.4= +.- n =IC: W.: nttnr=?n:t=tnn C r_ o 4

=t=

t

m-t= / tr =t= = = t= i. nrt: :-- _r=t._ c,U j gt- ..n 3._.n pan; = ..._..;. _. + _ _ _ +. _. +_ = 2;-n _nn;=_ r...._.r=.. ;== ku aminn un = nnr=. - - ~... O =* = ra_ a3 6 2__ y_ U . _.._ t r-t rat.._. ~i'*11. [ .._4 ..i.L. --*d~~~ I =7.i; : g A __ gu

=n4 u.

.r.=; n= r=..=t. :- r-- _e ~ * - =.ij =. t =;nn oo g...-.,._ q M -. L._-~+ _._.y. q.;.;... _ ._ gg. =_= fE t-

t-

.._t n ..1_ -p g-

=;

g. ~.

-,w

• .-4.i+.+

, _,,, _~_____ i'"* .-.1*t .i --i..f.- t=t= n

=

~~ ~ nut *th #F"r=- nn;=t=t - .i-- ---+ ._.i y ;;7g;;n;.y q;,.,.,4;;n.gg-gt,____..--__gt_...- 7 3-y =, 4; i

s_P **? ----t-+V,P~-++*-* +---+'**C1%

,,,,;;II.C_ n-+----*~-f3 ^ * * * - ~ h'-- n.i,_;g==;n-.=t=;;;

==tng:=:t.;. =n. _.= ..=

==t=......,.. mdd 'uoT2v22uacuoc no2og say d Figure A2.5-4

A2.6 NUCLEAR INSTRUMENTATION OVERL\\P Technical Specifications state that prior to operation in the intermediate nuclear instrumentation range, at least a one decade overlap between the source range and intermadiate range must be observed. This means that before the source range count rate equals 105 cps the intermediate range,9 must be on scale. If the one decade is not observed, the approach to 10 ampere on the intermediate range cannot be continued until the situation has been corrected. In addition, the following number of nuclear instru-mentation channels must be in operation for the test program to continue beyond initial criticality. Channels Available Minimum Operating Source Range NI 2 2 Intermediate Range NI 2 2 To satisfy the above overlap requirements af ter criticality was reached, core power was slowly increased until the intermediate range channels came on scale. Detector signal response was thus recorded for both the inter-mediate and source range channel. This was repeated for two more decades until the source range signals approached 106 cps. The results of the nuclear instrumentation overlap at 532 F and 2155 psig are shown in Table A2.6-1. Examination of this table shows that the overlap between the source and intermediate range is constant at an average value of 2.33 decades which is well above the minimum of one decade overlap stated in the Technical Specifications. I I m G 9 J -A2.6-1

= N Summary of Nuclear Instrumentation Ove.rlap Measurements Taken after Fuel Reloading Data Source Range Indication Intermediate Range Indicationi Average SR Average IR Overlap Set NI-1 (cps) ' NI-2 (cps) NI-3 (amps) NI-4 (amps) Indication Indication (Decades) 1 4.50 x 10 4.50 x 10 1.00 x 10-1.00 x 10' 4.50 x 10 1.00 x 10-2.35 4 -11 -II -1 2 3.75 x 10 3.60 x 10 8.00 x 10 8.00 x 10 3.68 x 10 7.63 x 10 2.31 -10 -II -10 3 3.75 x 10 3.25 x 10 8.00 x 10 8.00 x 10 3.50 x 10 7.63 x 10 2.34 2.33 (ave) h Note (1): Overlap is obtained between the Source and Intermediate y llange by using the average indications in the equation belov. Overlap = ( 6 - Log CR ) + ( Log I - 11 ) A

A2.7 "ALL RODS OUT" CRITICAL BORON CONCENTRATION O The "all rods out" critical baron concentration was measued at a temperature test plateau of 5320F. The measurements were made with Control Rod Group 7 partially inserted, but the measured boron concentration was adjusted to the "all rods out" cor.dition using the results of rod worth measurements to determine the reactivity worth, in terms of boron concentration, of the inserted control rods. The results are tabulated in Table A2.7-1. These results show that the measured boron concentrations compare quite favorably to the predicted results. e i .D T ..) A2.7-1

4 i t j e i All Rods Out Critical Boron Concentration j _- Measurement Af ter Fuel Reloading ppa boron i n 4 ( I' Moderator Predicted ~ Measured i Temperature Results Results f. 532 F 1622 1632 i l i 5 i 3 t i f f j 4 l f ' -y F L - Table 1A2.7-1 g-4

== g g a w ty.. w y ma-we< g. p 9 y

A2.8 CORE POWER DISTRIBUTION Core power distribution measurements were taken as required during the power re-escalation test program. As a minimum, core power distribution was required once per eight hour shift to check minimum DNBR and maximum linear heat rate. In order that comparison could be made to previously measured core power distributions, data were also taken for the following control rod patterns, o core power levels, and xenon conditions. Case Power Control Rod Group Position (%wd) Equilibrium Number Level (%FP) 1-5 6_ 7, 8_ Xenon 1 15 100 75 00 35 No 2 40 100 75 00 35 Yes/(3-D) 3 75 100 100 75 100 Yes/(2-D) ~ Each of the above cases were analyzed to determine the radial core power distribution and the maximum axial times radial peak. A comparison of the radial core power distributions'obtained during initial power escalation and during power escalation af ter fuel reloading is presented in Figures A2.8-1, A2.8-2, and A2.8-3. As can be seen, good agreement between the two measure-ments exists for the three cases, with a maximum deviation on peak radial assemblies of 3.6 percent. The maximum axial times radial peak (total peaking factor) also gave similar results when axial imbalance was approxi-mately the same. During the power re-escalation program, minimum DNBR and maximum linear heat rate were recorded at various power levels and extrapolated to 102 percent full power. From the data recorded during the re-escalation period, a minimum DNBR of 2.54 and maximum linear heat rate of 13.83 kw/ft were ob-served when extrapolated to the LOCA limit which is well within the acceptance criteria of a minimum DNBR of 1.55 and a maximum linear heat rate of 17.30 kw/ft. A summary of minimum DNBR as a function of power level is given in Figure A2.8-4. In summary, good agreement was obtained on radial and total peaking comparison when taken under equivalent conditions. Minimum DNBR and maximum linear heat rates measured during the re-escalacion program were all well within the acceptance criteria of 1.55 and 17.3 kw/ft, respectively. I i Y / A2.8-1 l

CO GARISON OF !EASURED PADI AL CORE POWER DISTRIBUTICNS OBTAINED DURING I:iITIAL POWER ESCALATICU AND CURING POWER ESCALATI0d AFTER FUEL RELOAD AT 15 %FP CONDITIONS. A. "easured Results Cbtained During Initial Power Escalatien Control Pcd Group Positions Core Power Level 15 %TP Cps 1-4 100 % wd Boron Concentration 1350 ppm Cp 5 100 % wd Core Burnup 07 EFPD ~ Cp 6 To % ud Axial Tebalance +0.4 %FP Cp 7 00 % wd Max Quadrant Tilt -0 75 % Cp 8 30 % wd Xenon Equilibriun Ho Yes or No 3. Measured Results Obtained During Power Escalation After Fuel Reload Centrol Rod Group Positions Core Power Level 15.h %FP Cps 1-4 100 % wd Boron Concentration 1397~ ppm Cp 5 N % wd Coro Burnup 11.0 EFPD Cp 6 77 % wd Axial Imbalance -2.4 %FP Cp 7 02 % ud Max Quadrant Tilt +1.9h % Cp 8 3h % ud Xenon Equilib. ium 30 Yes or No 1.08 1.28

1. 32 _1.h2 1.17 1.30, 1.37 0,.9h g

1.K 1.30 1.25 1.38 1.20 1.37 1.35 0.86 30 1.h0 1.26 1.22 1.08 1.02 0.79 e

1. h 1.hk 1.31 1.26 1.08 1.00 0.80

.2h 1.29 1.00 0 97 0 72 0.h6 1. 1.30 0.96 0 93 0 72 0.h5 .09 1.02 0.8h 0.69 1. 1.02 0.82 0.62 4 92 0.83 0.53 i 0. 0.78 0.53 w.u6 i 0.z N 1 F b s X.XX Conditions (A) X.XX Conditions (3) Figure A2.8 COMPARISON OF MEASURED RADIAL CORE POWER DISTRIBUTIOTiS OBTAINED DURING INITIAL POWER ESCALATION AND DURING POWER ESCALATI0ii AFTER FUEL RELOAD AT ho %FP CONDITICUS. A. Measured Results Obtained During Initial Power Escalation Control Rod Group Positions Core Power Level 39.8 pp Cps 1-4 100 % wd Boron Concentration llol ppm t ~ Cp 5 100 % vd Core Burnup 4.3 EFPD Gp 6 75 % ud Axial Imbalance -4.6 %FP ':p 7 00 % wd Max Quadrant Tilt +1.16 % GP 8 35 % wd Xenon Equilibriun Yes Yes or No B. Measured Results Cbtained During Power Escalation After Fuel Reload Control Rod Group Positions Core Power Level h0 %FP Cps 1-4 100 % wd Boron Concentration 1208 ppm Cp 5 100 % wd. Core Burnup 12.6 EFPD Cp 6 75 % ud Axial Imbalance -9.6 %FP Cp 7 00 % wd Max Quadrant Tilt -1.25 % Cp 8 35 % vd Xenon Equilibrium Yes Yes or No 1. T 26 1.26 ~1.36 1.10 1.29 1.37 0 90 1.02 1.2k 1.27 1.36 1.15 1.32' 1.32 0.83 ~ O 1.37 1.27 1.23 1.08 1.02 0.80 1 1. 1.41 1.30 1.25 1.07 0 99 0.78 '1x25 1.28 0 98 0 98 0.Th 0.48 1.K., 1.30 0 98 0 95 0 73 0.h6 lx07 1.0h 0.8h 0 70 1.Q 1.05 0.88 0.66 h5 0.8h 0.59 0. 0.80 0.5h 'G. 57 0.Q ? 9 b l 1 X.XX Conditions (A) J X.XX Conditions (B) Figure A2.8-2

COMPARISON OF MEASURED RADIAL CCRE POWER DISTRIBUTIC:IS OBTAINED DURING INITIAL POWER ESCALATION AND DURING POWER ESCALATIOil AFTER FUEL RELOAD AT 75 %FP CONDITIONS. A. Measured Results Obtained During Initial Pcwer Escalation Control Ecd Group Positions Core Power Level 75 %FP Cps 1-4 100 % wd Boron Concentration 1275 ppm Gp 5 100 % wd Corc Burnup 11.4 EFPD Cp 6 100 % ud Axial Imbalance _o.o %FP Cp 7 75 % wd Max Quadrant Tilt e.? ~4 Cp 8 100 % wd Xenon EquiliSrf um Yes Yes or No B. Measured Results Obtained During Power Escalation After Fuel Reload Control Rod Group Positions Core Power Level 75 %FP 100 % wd Boron Concentration 1277 ppm Cps 1-4 Gp 5 100 % ud Core Burnup 15.0 EFPD Cp 6 100 % wd Axial Imbalance -6.9 %FP Cp 7 76 % wd Max Quadrant Tilt +1.h % Cp 8 E % wd Xenon Equilib rium Yes Yes or No 1.31 1.17 1.05 _1.16 1.07 1.28_ 1.35 ,0.86 g 1.$ 1.19 1.06 1.19 1.08 1.30 1.3h 0.86 s 6 1.18 1.12 1.18 1.09 1.11 0.87 1 1. 1.20 1.1h 1.18 1.09 1.11 0.88 '1x10 1.18 1.01 1.08 1.12 0.63 1.1% 1.20 1.02 1.09 1.03 0.63 'l 04 1.0h 0.87 0.80 1. 1.05 0.86 0.70 h89 0.79 0.58 09 0.80 0 56 i N 01 2 i 0. s I r A X.XX Conditions (A) X.XX Conditions (3) Figure A2.8-3 w =

- ~, A2.7 "ALL.t0DS OUT" CRITICAL BORON CONCENTRATION The "all rods out" critical boron concentration was measued at a temperature test plateau of 5320F. The measurements were made with Control Rod Group 7 partially inserted, but the measured boron concentration was adjusted to the "all rods out" condition using the results of rod worth measurements to determine the reactivity worth, in terms of boron concentration, of the inserted control rods. The results are tabulated in Table A2.7-1. These results show that the measured boron concentrations compare quite favorably to the predicted results. I \\ ~ l l 7 l j e A2.7-1 r.~ l

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3.0 CONCLUSION

S Fuel reloading commenced March 25, 1974 and power re-escalation was completed on May 30, 1974 when unit startup was returnr. to the initial power escalation procedure. Initial criticality af ter fuel reloading was achieved on May 23, 1974 at 1337 hours and within seven days the unit was operating at 75 %FP. A summary of each of the tests or operations performed during re-escalation to power is given below: (a) Fuel Reloading Fw ? reloading of Oconee Unit 2 was completed in nine days; and except for difficulties with the temporary incore detector system early in the fuel loading sequence, fuel reloading proceeded in an orderly manner. (b) Control Rod Drive Drop Time Test The measured rod drop times for the slowest control rod were observed in core location M-09. The measure time of 1.324 seconds at full flow and 1.194 seconds at no flow were well below the acceptance criteria stated in Section 4.7 of the Technical Specifications, of 1.66 seconds at full flow and 1.40 seconds at no flow conditions. Also, as would be expected, drop times under flow conditions were longer than under no flow conditions. (c) Reactor Coolant Pump Flow Test (. The measured reactor coolant flowrate during re-escalation to power provided adequate margin to both the maximum and minimum allowable flowrates. Compari-son to.he reactor coolant flow measured during Zero Power Physics testing and during re-escalation to power agreed within less than 5 percent. (d) Loose Parts Monitoring -Surveillance Test Analyses of test data at all test conditions indicated no noises which could be characterized as loose objects in the bottom of the reactor vessel or the top of either steam generator. (e) Initial Criticality After Fuel Reloading Initial criticality after fuel reloading was achieved on May 23, 1974 at 1337 hours at reactor coolant conditions of 532 F and 2155 psig. All control rod groups were fully withdrawn except for Control Rod Group 7, which was at 59 %wd and the critical boron concentration was 1600 + 10 ppm. (f) Nuclear Instrumentation Overlap r The results of nuclear instrumentation overlap at 532 F and 2155 psig showed that the overlap between the source and intermediate range was constant at an average value of 2.33 decades, which is well above the minimum of one l decade overlap required by the Technical Specifications. o g., s, A3.0-1

(g) "All Rods Out" Critical Boron Concentration The "all rods out" critical boron concentration was measured at 532 F and 2155 psig to be 1632 ppm which compares quite favorably to the predicted value of 1622 ppm. 3 ~ (h) Core Power Distribution A comparison of measured data obtained during initial power escalation and during power re-escalation after fuel reloading resulted in good agreement between radial and radial times axial peaking when taken under equivalent conditions. Minimum DNBR and maximum linear heat rates measured during the re-escalation program were well within the acceptance criteria of 1.55 and 17.3 kw/ft, respectively. (. I 5 f r 4., A3. 0-2.-

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