ML19283A536
| ML19283A536 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/13/1966 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 8008280581 | |
| Download: ML19283A536 (60) | |
Text
V.
THIS DOCUMENT CONTAINS POOR QUALITY PAGES D!
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s DOCKETED USAEC 6
FEB 9 1966 >
3 REGULATCtf/ '
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ANNUAL REPORT YEAR 1965 CDJ"
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Al;17UAL P2PCRT OF STATIO:i CFE2ATIO :
FOR Ti!3 YF.\\R 1965 Janugry 13, 1966
DRESDEM NUCLEAR FCNER STATICN AfNUAL REFCRT I.
INTRODUCTION This fourth annual report is submitted in ecmpliance with paragraph 3.c.(2) of Utilization Facility License DFR-2, as aeanded, and covers, operation of Dresden Nuclear Power Station during the year 1965.
II.
SUltARY OF OPERATIONS A.
Scope of Operations Operation of the plant continued from the preceeding year and the longest continuous run of 137 days uas terminated on January 28, 1965. At this time the plant was shut doun to repair a len!: in the turbine cross-under pipe betueen the high pressure exhaust and the intermediate pressure inlet.
The third partial reactor refueling outage was performed from March 23, 1965 to May 29, 1965.
During this outage the reactor contcincent vessel was successfully leak tested.
The plant continued operation to the end of the year with a total of 11 shutdcwns during the year. Additions to and changes in facility design were made in the off-gas system filtering and instrumentation, as described in paragraph 7, page 32.
Additional equipment includes a prototype control rod worth minimizer which was installed.
Seven shipments totaling 168 spent Type I fuel assemblies were made for the U.S.A.E.C. (Savannah River Operations Office) to the Chemical Processing Plant of Nuclear Fuel Services, Inc. at West Valley, New York.
B.
Shutdouns The plant was shut down eleven times during the yes'r as indicated on Table 1.
Five of these shutdowns were forced outages:
one to repair a leak in the turbine cross-under pipe, one to repair a leak in the intermediate pressure turbine flange, two to repair leaks in a high pressure drain line, and one caused by loss of transmission lines caused by a tornado.
There were six scheduled outages during the yaar:
one initiated in preparation for refueling but instead the plant was returned to service for two weeks, one for refueling, three for turbine overspeed tests and one to investigate the malfunction of reactor control red drive J-2, which was later valved out of service in its fully-withdrawn position.
C.
_ Load Restrictions The load restrictions imposed during the year are listed in Table 2.
The major portion of these restrictions were due to fuel depletien prior to refueling.
TABLE 1
_OpERATIm PEUFO'OtANCE 1965 No. of
_ Off Systen On System Outar.e Date Tirm Da te Tima Cutage Itours Renson For Outage 54 1/28/65 9:00 p.m.
1/31/65 11:27 p.m.
74 Urs.
27 Min.
Repair Icaks in cross-under pipe.
55 3/14/65 10:45 p.m.
3/15/65 12:40 a.m.
1 Ilr.
55 Min.
Unit shutdown for overhcul.
System required unit back in service.
56 3/28/65_
8:58 p.m.
5/29/65 9:11 p.m.
1487 IIrs.
13 Min.
Third partial refueling and turbine overhaul.
57 6/18/65 10:55 p.m.
6/20/65 8:22 a.m.
33 lirs.
27 Min.
Repair I.P.
flange Icak.
53 7/2/65 5:42 p.m.
7/2/65 5:50 p.m.
O Min.
Turbine overspeed test.
I 59 7/2/65 11:10 p.m.
7/2/65 11:16 p.m.
6 Min.
Turbine overspeed test.
60 7/3/65 11:10 a.m.
7/5/65 8:38 a.m.
45 Urn.
20 Min.
Turbine overspeed test.
Unit kept out of service to repnir condenser tube leqk.
61 10/7/65 9:23 p.m.
10/11/65 1:26 a.m.
76 IIrs.
3 Min.
Repair leak in II.P. turbine dra in line.
62 10/11/65 2:53 p.m.
10/17/65 6:45 a.m.
135 llrs.
52 Min.
Repair leak in II.P. turbine drain line.
63 10/17/65 10:46 a.m.
10/18/65 5:05 a.m.
18 Hrs.
19 Min.
4.0 hrs. to withdraw control rod J-2.
14 hrs. 19 min. to repair condenser tube Icak.
64 11/12/65 2:50 p.m.
11/13/65 9:45 p.m.
30 11rs.
55 Min.
Storm damage lost n.11 lines from station.
Total Outage llours For Year 1903 I!rs.
53 Min.
~
TABLE 2 LOAD RESTRI,CTIONS FOR 1955 Reduction from Maximum Date
_ Capability of 210 f ue Condition January 1 - March 9 65 Minimize fuel failure.
January 5 95 Cross-under Icak investigation.
January 7 120 U. P.
exhaust crosa-under piping leak repair.
January 24 120 Repair cross-under leak.
March 10 - 16 66 - 70 Fuel depletion.
March 17 - 20 70 - 75 Fuci depletion.
March 21 - 20 75 - 80 Fuel depletion.
Itay 30 - June 3 100 Incore stabilization.
June 4 - June 10 50 Incore calibration.
June 11 - June 17 10 Incore calibrat' ton.
July 19'- July 24 1-5 Circulating water temperature.
July 25 - July 30 5 - 10 circulating water temperature.
July 31 - August 1 2-5 Circulating water temper'ature.
August 6 - August 9 1-2 Circulating water temperature.
August 15 - August 21 2-6 Circulating water tempt ture.
August 24 - August 26 2
Circulating water temperature.
TABLE 2 (Cont.)
n, educt,i,on frem Maximum Date Capabili_tv of 210 I:.Je Condition October 7 100 ll. P.
turbine steam leak.
October 11 60 II.P. turbine steam leak in 1" drain line.
October 27 - October 28 20 Incore calibration.
november 6 25 IIcater vent line repair.
November 14 - November 18 20 "B" secondary steam generator tube leak.
December 9 - December 31 15 "C" recirculation pump flange Icak and diccharge valve bypass leck, a
V e
t S
FIGU"E 1 PLANT ELECTRICAL LOADING YEAR 1965 DRESDEN NUCLEAR PO';((FR STATION
" Iv$1$1 "$t ndN5ak$kg (2)
'$tuef[cNdroIYob"an[$o"n$cncr' I
6 10.
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i>: !cpa I$P lan o leak 31 )
c r i P ub D is(
11.
Lo of 'i'ra mission Facilities (31) 6
___ _ _ _ y' 200 200 2
fa 10 0 100 O
IGO S
A 10 0
- V f verage Ioad f 7 7
' W 99
- +
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t00 10 0 l
/-l d
/
40 40 0
3
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1 7 05-4 Md FE0 t0AQ A?Q MAY JUD3 JUL AUG SEP
.O C T NOV D"C
III.
DIS _CUSSION A.
_Operatine Experience 1.
Generction The total reactor operating (critical) time during the year was 6960.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the total power for the period was 138,149 MWD (thermal).
The gross electrical generation during the year was 1,018,344.32 tr.id; not generation was 962,307. 8 M'a.
As of December 31, 1965 the total gross generation, since
~
commencement of pouar operation, April 15, 1960 was 5,125,954.57
! C.'H.
2.
S e rm,s a.
At 11:10 a.n. on July 3, the reactor scranrad from high flux during turbina overspeed tests. conducted at the request of the Gancral Electric Cc=pany.
The reactor was critical, operating at 604.6 If.!t (210 nue).
There ware 55 rods withdraun and one rod withdraun, 10 notches.
b.
At 2:50 p.n. on Novenber 12, the reactor'scratmed froa high flux due to loss of all electrical transmission lines causing a turbine overcpeed condition similar to the test conducted in July.
The reactor was critical, operating at 617 IDt (200 IUc).
The loss of the trancaission lines uas caused by tornado damage.
3.
_ Incidents There were no incidents during the year uhich compromised the safety of continued operation.
4.
control nod Drives a.
_ Control Rod Drive Ooeration 1.
On June 11, during a routine instrument response test,
~
C-9 control rod blade was inserted from position 10 to 9.
Instrument response was normal.
In attempting to withdraw the blade frca position 9, the rod inserted to position 8 and renained there.
Further attempts to withdraw the blade were unsuccessful.
The drive responded norcally in the insert direction.
The withdraual inpulse would cause the drive position indicator to show movenant froa position 6 to position 7 and back to position 8.
It would not continue to withdrew beyond position S.
9 9
. a.
Control Rod Drive Operation (Cent.)
Numerous checks were made to clear the malfunction without success.
It was concluded that the locking collet was not unlocking, possibly due to a foreign particle preventing collet movement. 'The drive was valved out of service at position 8.
~
2.
During the shutdown on October 11, 1965 for turbine repair, control rod J-2 would not insert with normal actuating hydraulic pressures, (reactor pressure +
200 psig. and reactor pressure + 15 psig.).
This drive (J-2) did not indicate abncreal performance during the October 8 scram and fricticn tests.
The drive insert actuating pressure was increased to 275 psig, and after several attempts to insert, went to indicated position 91.
The insert actuating pressure was increased to 300 psig. in order to incert the drive to the fully inserted position 0.
Numerous tests were made on October 12 and 13 without disclosing a conclusive cause of failure of J-2 to insert.
On October 14 a General Elcetric representative arrived on site to consult and confer on the J-2 problem.
Consideration of all test results indicated that the prob 1cm is in the control red mechanisn and not in the external hydraulic oc electrical control systems.
Control rod J-2 is valved out at position #12.
Reference is made to letters of October 14 and Octob2r 27 to Mr.
Roger S. Boyd, Chief Research and Fower Reactor Safety Branch, U.S.A.E.C.
b.
Control Red Drive Tests All control rod drives were scram and friction tested on January 29, March 29, and April 24 and 25.
The data obtained from the above tests were satisfactory on all drives.
Scram and friction tests also were conducted on October 8.
Attempts to scram (using the normal scram solenoid salves)
B-7, B-4, and E-1 on accumulator number 15 vere unsuccessful.
Investigation revealed that a netal chip had lodged on the scram solenoid valve diaphragm, rendering the valve inoper-ative.
The scram solenoid valve was replaced and all three drives scrauaad without further difficulty. Had the accumulator been given a scram signal during operation, the scrat backup solenoid valve would have scrammed the drives on accumulator number 15.
c.
Insrection Prior to shutdown for refueling, two drives were selected for removal and inspection.
The following table shows the drives removed and reasons for their selection.
c.
Inspection (Cont.)
Serial fiumber Orive Reasons for Removal 1274 G10 Drive had not been removed from reactor since drive modification in February,1951.
Drive failed to withdraw beyond position 7 on June 7, 1964.
Inves-tigation revealed that a jammed orifice in the ASCO valve to botten of drive prevented normal drive operation.
The orifice was replaced and the drive operated normally.
Drive failed to withdraw nornally on August 28, 1964. Foreign particles or other cause of high friction suspected.
1312 G5 Drive had not been removed frca reactor since drive-codification in February, 1961.
Exhibited higher than average friction pressure during friction tests.
Exhibited shorter than average time in buffer during scram tests.
Drive operated normally after exercising.
All drives ware scran and friction tested un 1: arch 29,1965.
These tests served to deternine if there were any additional defective drives prior to drive removal.
The friction and scram test data obtained were satisfactory on all drives tested.
A fluorescent dye pene. rant (Zyglo) exanination was cade on cocyonents of the two drives. A penetrant (ZL-2), developer (ZP-5), and an ultraviolet light were used for this inspection.
The following components of the two drives were subjected to the fluorescent dye penetrant examination:
Index tube (5850288)
Piston head assembly (1920554)
Shuttle piston (865B398)
Stop pisten (115AE600)
Collet assembly (693CS27)
Roller mount assembly (932C149) a.
Weld b.
Spud c.
Anti-rotational roller d.
Guide roller Roller hcusing c.
Spring (lllA3293)
Spring vasher (l'+5AS454)
Guide plug (S555397)
Drive housing welds (5850289)
c.
Insocetion (Cont.)
The results of individual drive inspections are listed below.
1.
Drive Mur.ber 1274 (Cell C10) - Failure to Uithdraw The outer bushings and piston scals en the drive piston were heavily scored and one segment of the two piece bushings had failed.
Failure of the bushing and scoring of the seals was more than likely caused by foreign material between the main piston and the inner
~
tube surface.
The foreign material in this location was also probably the cause of the malfunction on August 2S, 1964.
A cracked shoulder was found on one segcent of the three picec tangentially cut inner piston seal 05.
All other conponent parts were visually inspected and found to be in good condition.
The dye ponctrant inspection revealed no cracks.
2.
Drive Humber 1312_(Cell _G5) - H_istory of Micher _than Average Friction Three of the four stop piston three segment seals had a corner broken off on one segm nt.
One segment had failed on the three segment tangentially cut inner piston seal #4 of the main piston (this is the buffer scal).
Excessive leakage through these broken seals accounts for the history of lower than average time in huffer on scram tests of this drive.
Numerous shallow scratches were found on the surface of the index tube.
Some of these scratches were caused by the collet fingers riding on the index tube surface, although the majority were probably caused by foreign material trapped between the outer (scratched) surface of the index tube and the chromed inner surface of the collet assembly.
The chrome plating on the collet assembly did not appear to be marred or damaged.
The foreign material between the aforementioned surfaces is believed to have been the cause of the high frictien history of this drive.
Visual inspection and dye penetrant checks revealed no defects, cracks, or damage to any of the other parts of this drive.
_ 10 -
Insp'ection (Cont.)
c.
In su=marizing the results of this inspection, the following significant items were noted:
1.
Although two ralfunctions were experienced on driva 1274 (Cell G10), the drive was capable of.being inserted, and the malfunctions were rapidly corrected.
2.
No cracks or defects w2re found on either drive by visual inspection or dye penetrant tests.
3.
No trends were noted in regard to failure of component parts.
4.
Failure of scee seals was not unusual, since both drives have been in service for core than four years, and failure of seals did not cause a malfunction of the drives.
5.
Systen Stability Tasts a.
Turbina Overspeed Tests The turbine overspeed tests were proposed by General Electric Company and similar tests were previously conducted on the SEUN reactor.
Evaluations perforced by both General Electric and Ccmmonucalth Edison indicated that the flux transient following a generator trip would be only a little core than one second in duration and would be of no thermal significance as far as the fuel is concerned.
General Elcetric estiaated that the peak neutron flux follouing a generator trip at full pouer would be about 130% and that the corresponding peak heat flux would be only 105%.
Both General Electric and Commonwealth Edison were confident that no adverse affects would be suf fered from these tests.
The peak neutron flux actually experienced during the full power test was within 10% of the estimated value and was well below the peak neutron flux experienced at SENN.
No burn out ratio (BOR) problem was expected, and none was experienced.
The BORs as calculated by the Nuclear Engineer, are given in Tables 3, 4, and 5.
The initial and final BOR during the transient was calculated from the test data to be about 1.8.
The three turbine overspee/ (generater trip) tests were initiated oy opening the rain 138 KV oil circuit breaker af ter all auxiliary poner had been transferred to transforcer Number 12.
Special high speed recording instrument's were used to documsnt the tests.
The transient perforcance is shown in Tables 3, 4, and 5, and the control rod pattern for the tests is shown in Figure 2.
The reactor scrammed on high neutron flux af ter the 210 IOc trip.
__. I'1'. '
TABLE 3
_120 MWe 'IURBINE OVERSPEED TEST 5:42 p.m., JULY 2, 1965 NO REACTOR bCRAM Initial Peak Vallev Final MWe 120 0
0 0
~
Psf (1b/hr. x 10+6) 1.37 1.50 0.87*
1.33*
Ssf (1b/hr. x 10+6) 0 0
0 0
Reactor Recirculation 26 27.2 24.4 26 Flow (1) (1b/hr. x 10+6)
Reactor Pressure psi. (2) 1,000 1,013 1,000 Reactor Power MNt 379.6 343.6 Burnout Ratio 4.8 4.3 4.4 Out of Core. Pyreent Po'.ar 2
61 96 55 63 4
56 88 50 57.5 5 (3) 58 95 50 58 6 (4) 58 95 50 58 Incore, RE-108 (3), Percent Licanna Hagt Flux D
28.5
'5 30 C
40.5 72~
43.5 B
40.5 64.5 42 A
18 30 18 Incore, RE-109-B (5) 47 72 39 48 Generator Excition - 505 acps., const. voltage = 14,200 Generator Overspeed - 1930 rpm.
(1)
Brush recorder - A Ssg 2h P (2)
Sanborn recorder, (3)
Trace recorder (4)
Sanborn recorder, (5)
Brush recorder All others are from plant recorders
- Bypassed to condenser.
- TABLE 4 172 MWe TUR3Ih?, OVERSPEED TEST 11: 00 p.m., July 2, 1965 NO REACTOR SCRAM Initini Peak Valley Final MWe 172 0
0 0
Psf (1b/hr. x 10+6) 1.28 1.42*
0.91*
1.23*
Ssf (1b/hr. x 10+6) 1,03 o
o o
Reactor Recirculation Flow (1) (1b/hr. x 10+6) 26 26.8 23.6 25.6 Reactor Pressure psi. (2) 998 1,008 1,000 Reactor Power MJt 568.6 320.6 Burnout Ratio 3.3 2.9 4.2 Out.>of fore,, Percent Power 2
81.5 106.5 48 61 4
80 107 48 59 5 (3) 79 108 58 80 6
81 111 46 63 6 (4) 80 113 68 80 Incore, RE-108 (3), Percent License Heat Flux D
37.5 49.5 21 25.5 C
63.7 78 33 45 B
66 88.5 24.5 46.5 A
31.5 40.5 15 21 Incore, 109-B (5) 70 95 56.3 64 Generator Excition - 730 amps., const. voltage = 14,200 Generator Overspeed - 1975 rpm. no turbine trip (1)
Brush recorder - A Ssg sd[ P.
i (2)
Sanborn recorder, (3)
Trace recorder (4)
Sanborn recorder, (5)
Brush recorder i
All others are from plant recorders.
i
- Bypassed to condenser, j
TABLE 5 210 MWe TUR3INE OVERSPEED TEST 11:10 a.m.,
July 3, 1965 REACTOR SCFRf, TURBINE TRIP, RECIRCULATION PUMP TRIPS Initial _
Peak Valley Final MWe 210 0.
0 0-Psf (iblhr. x 10+6) 1.59 0
I Ssf (1b/hr. x 10+6) 1.37 0
0 i.
Reactor 7.ecirculation all pumps tripped i
Flow (1) (1b/hr. x 10+6) 26.1 27.2 24.8 6.0 i
Reactor Pressure psi. (2) 1,000 1,011 decreasing i
Reactor Pcuer MWt 684.6 0
Burnout Ratio 2.4 1.8
-Out of Core, Percent Power E
2 97 118 0
Ii 0
l.
4 96 140 5 (3) 94 136 O
l' 6
96 124 0
l:.
I 6 (4) 94 143 0
1:
Incore, RE-108 (3), Percent Power License Heat Flux
'{
0
~.l D
51 69 9
e,o C
75 84 0
j S8 O
h B
75.-7 108 Q
0 2.'
A 36 51 t.
Incore 109-B (5) 83.7 115 0
Generator Excition - 880 amps., const. voltage = 14,200 jj Genera tor Overspeed - 2014 rpm., turbine tripped.
(1)
Brush recorder - A Ssg P, final value estimated.
3 (2)
Sanborn recorder, (3)
Trace reccrder (4)
Sanborn recorder, (5)
Brush reccrder
(
All others are from plant recorders.
y e.
I r.
~}l 3
FIGUPS 2 CONTROL CCD PATTE2."S DURIKO II:E
_ TURBIT? CVERSPCED TESTS ABCDEFGHJK ABC D ~E F GHJ K iol 10 to 10 9
e 2
3 9
8 1
2 Y[4 Yf5' e
10 8
8
-2 '
8 10 8~
8 7
3 2
9 2
7 2
2 9
2 73 FS 9 1
6 10 9
2 6
to 5
3 9
10 5
1 9K
' 'O 10 2
9 2' 3
4 O,N 2
9j2 2
?
4 3
to e
e 3
to 8
- f ff.:
e l
^
3 2
10 2
2 1l 19 2
l to i
10
-l 120 We 172 We 5:42 p.m., July 2 11:00 p.m., July 2 ABCDEF GHJK ABC DEFGHJK 10 10 10 9
8 3
3 l
9
'2
?)
e 10 8
?
8 e
/
7 3
2 9
2 'If-7 6 10 g;
9 3
6 5
3-9 %
TU to S
4
f;-
2 9
2 3
4 3
in r
W Z
e 3
l I
2 3
1.0 2
I to I
(
l 210 I"le 11: 10 a.m., July 3 Y
15 -
5.
Systen Stabili_ty Tasts (Cont.)
b.
Recirculation Pumn Trio Tests Recirculatica pump trip tests uere conducted in order to obtain data necessary to anticipate reactor and system response likely to be experienced during pump trips for Cycle 4.
The following pump trip tests were conducted:
~
Test Number of Number D' _a Pumos Trinped Pcuer frJe 1
July 1 1
110 2
July 6 1
160 3
July 6 1
160 4
July 6 2
160 5
July 7 2
110 The transient performance nad the control rod patterns in use during the pump trip tests are shoun in Tables 6 through 10 and Figure 3 respectively.
Some of the data was obtained from high speed recorders, but most of it was obtained frem plant recorders.
The reduction in recirculation flow follouin3 a recirculation pump trip causes a decrease in the primary steam flow.
The governor and dual cycle attempts to maintain the original steam flow to the turbine by picking up secondary load.
The first and last tests were conducted with zero secondary flev.
In the second test, conducted with the governor controllin3, the secondary picked up so much that the final electrical generation uns slightly greater that. it was initially. To insure that there uculd be no ECR or carry-over problem, the subsequent tests were conducted with the secondary load limit controlling and with the operator reducing the secondary steam flow via the secondary load limit during the tests.
The primary drum level was lowered previous to the simultaneous tripping of two pumps, and the operator held down the primary drum level swing by manual feedwater control to insure that the turbine would not be tripped due to high drum level.
The initial and final BORs for these tests were 3.0 or greater, based on incore 108.
The data as exhibited in Figurc 4, shows that the shed in primary steam flou can be reasonably well predicted frca the shed in reactor recirculation flou.
A cicser exanination indicates that there may be a second order effect due to the secondary steam flou.
Cases with secondary steam flow show a slightly smaller shed in primary steam flou than cases without any secondary steam flow.
The mechanism causine the second order effect is probably one of. increased sub-ecoling folleuing a secondary pick-up.
TABLE 6 "B" RECIRCULATION PDiP TRIP TEST TRANSIENT PERFOR:'ANCE AT 110 M1.'e_
5: 15 p.m., July 1, 1965, Tes t #1 Af ter Trip After
_ Initial Transient Steadv Loop Closed Number of Pumps 4
3 3
3 Number of Loops 4
4 4
3 MWe 112 97 92 i
Psf x 10 lb/hr.
1.38 1.2 1.14 Ssf x 10+
lb/hr.
0 0
0 0
Primary Drum Level (South)
- 3.6
+ 0.9
- 3.6 Secondary Drum Level (A)
+ 1.7
- 7.3 0
Secondary Drum Level (B)
+ 9.4
-1 ReactorRecirculggion Flou lb/hr. x 10 25.8 20.0 20.0 Reactor Temperature F.
B 532 532 532 532 A,C,P 532 532 532 532' Pfvt F.
352 343 340 Sfwt F.
359 353 353 Out of Core, Percent Power 1
53 42 47 47 2
58.5 I:6 51.5 51.5 3
57.5 45 50 50 4
54 44 48.7 48.7 5
54.5 43 48 48 6
55.5 43 49 49
. TABLE 6 Cont.
Af ter Trip After Initial Transient Steady Loco Closed
_Incore, R2-100, Perce f t Licence IIent Flux D
24.9 18 21 C
36.3 27.9 32.3 B
39 30 34.5 A
18 14 16 109-E No Data
TABLE 7 "B" RECIRCUIATION PUMP TRIP TEST TRANSIENT PERFORMANCE AT 160 MWe GOVERNOR CONTROLLING 5:06 p.m., July 6,1965, Test #2 Af ter Trip After Initial Transient Steady Loop Closed Number of pumps 4
3 -
3 3
Number of loops 4
4 4
3 MWe 161 154 163 168 Psf (1b/hr. x 10+5) 1.50 1.36 1.34 Ssf (1b/hr. x 10' 6) 0.66 0.87 1.02 Primary Drum Level (South)'
- 5.1
- 1.98
- 3.6 Secondary Drum Level (A) 0
+ 3, -5 0
0 Secondary Drum Level (B)
+ 0.5
+ 9, -8 0
+1 Flow lb/hr. x 10'gion Reactor Recircula 26.0 19.7 20.7 Reactor Temperature F.
B 519 490 470 A,C,D 519 511 504 0 F.
391 395 400 Pfut Sfwt F.
395 397 400 Out of Core, Percent Power 1
70.5 55 71 72.5 2
3 4
71 57 70.5 72.5 5
6 63 49 58
. TABLE 7 Cont.
Af ter Trip After Initial Transient Steadv Loop Closed Incore, RS-108, Forcent__Liconse _Hea t_ Flux D
38.6 27.8 37.5 39 C
60 45 60.7 63 B
60.7 48 63 65.2 A
30 24 31.8 33.4 109-B 72.5 63.1 70
IABLE 8 "B" RECIRCULATION PIJMP TRIP TEST TRANSIENT PERFOR'3ANCE AT 160 MWe SECONDARY LOAD LIMIT CONTROLLING 6:06 p.m.,
July 6, 1965, Test #3 Af ter Trip After Initial Transient Steadv Loon Closed Number of pumps 4
3-3 3
Number of loops 4
4 4
3 MWe 160 147.5 153 154 Psf (Ib/hr. x 10+6)
- 1. 49-'
1.35 1.33 Ssf (1b/hr. x 1C+4) 0.58 0.68 0.75 Primary Drum Level (South)
- 6.2
- 1.56
- 3.7 Secondary Drum Level (A)
+1
+ 2, -2 0
0 Secondary Dreut Level (B)
+1
+ 9.5, -6.5 +1 0
Rea:cor Recircul,-* ion c' low lb/hr. x 10+6 26 19.78 20.1 Reactor Temperature F.
B, 519 498 483 A, C, D 519 514 510 0
Pfwt F.
390 387 390 Sfwt F.
394-392 394 Out of Core, Percent Power 1
69 52 55 57 2
3 4
79 55 66 66.5 5
6 58.5 45 51 54
21 -
- TABLE 8 (Cent.)
Af ter Trip After Initial Transient Steadv Loco Closed Incore, RE-108, Percent Licenac Heat Flux D
38.2 27.3 34.8 34.9 C
58.5 43.5 55.5 57 B
59 45.8~
57.7 59 A
28.5 22.5 28.2 28.8 109-B 72.5 63.1 71.9
TABLh'9 "B" AND "C" RECIRCULATION PUMP TRIP TEST TRANSIENT PERF0FF.ANCE AT 160 ige SECONDARY LOAD LIMIT CONTROLLING 10:50 p.m., July 6, 1965, Test #4 Af ter Trip After Initial Transient Steady Loco Closed Number of Pumps 4
2 2
2 Number of Loopa 4
4 4
2 MWe 161 142 146 137 Psf (1b/hr. x 10+6) 1.47 1.16 1.25 1.12 Ssf (1b/hr. x 10+6) 0.6 0.76 0.73 0.83, 0.77 Primary Drum Level (South)
- 7.6
- 0.72
- 3.6 Secondary Drum Level (A) 0
+ 1.7, -5.2 0
0 Secondary Drum Level (B) 0
- 6.5, +3.6 0
0 Reactor Recirculation Flow (1b/hr x 10+6) 25.9 17.5 13.76 0
Reactor Temoerature F.
B, C 520 495 463 A, D 520 509 500 Pfut F.
390 386 383 Sfwt F.
394 388.
383 Out of Core. PerceEt. Power 1
68 40 62 59 2
3 4
07 21 61 58 5
6 61 35 50 53'
- TADLE 9 Cont.
Af ter Trip After Initial Transient Steadv Loco Closer!
Incore, RE-103, Percent Lig_ense IIcat Flux D
40.5 19.5 33 30 C
60 31.5 57 51 B
57 33 55.5 52.5 A
24.8 15 25.5 24.7 109-3 65 47.3 65
TABLE 10 "B" and "C" RECIRCUlf,TIO?! FU:.T TRIP TS_ST T:VC;SIE::T P3T5CF/I!C3 AT 110 INe SsCG1!"IRY LCAD LIMIT CO:,TROLLI?:3 5:12 p.m., July 7, 1965, Test 45 Af ter Trip After Initial Transient Steady
_ Loop Closed Number of Pumps 4
2 2
2 Number of Loops 4
4 4
2 M'Je 108 E4 85 75 Esf (lb/hr. x 10+5) 1.35 1.04 1.06 0.95 Ssf (1b/hr. :.1C+6) 0 0
0 0
Primary Drum Level (South)
- 6.9
- 0.48
- 3.6 Secondary Drum Level (A)
+3
- 7.3, :5.7
-1
-1 Secondary Drun Level (3)
+ 0.3
+ 10, -8
- 0.5
- 0.5 Reactor Recirculation Flow (1b/hr x 10+6) 25.8 17.0 13.7 Reactor Te:cperature F.
B, C 533 529 529 527 A, D 533 529 529 527 Pfut F.
350 335 326 Sfwt F.
359 358 358 358 Out of Core, Percent Power 1
48 27 39 35 2
48 28 39 35 3
44 27 37 33 4
4S 30 40 36 5
44 25 35 31 6 (high speed) 44 25 32 39
TABLE 10 Cont.
After Trip After Initini Transient Stendv Loop Closg Incore, RE-108, Percent L_fyn..e Heat Flux D
25.5 13.5 17.3 15.8 C
26 20.2 29.2 26.2 B
26 21 29.2 25.5 A
16.5 9
13.5 12 109-B 42 33.5 41 i
e FIGUPI 3
_ CONTROL ROD PATTE".QJC,R _ THE O
_RECIRCULATIC'i PUMP TRIP TESTS ABCD EFGHJK ABC DEF GHJ K 10 10 10 10l 9
8 3
9~
8 1
8 8
?;:
8 to 8
6 @f 7[s 6
10 7
3 2
9
)3 7
1 3
9 3 iff G
10 fff 9
6 10 If 9
~
//
10 5
9 hl f(4 l10 5
3
?
4
_lh$
2 9
2 3
4 Y$l_
rY" 6 l1 l
a ll 3
jo g _.'
7 3
10 6
g 3
to 2
l 1
1.0 2
l In l
10 Test # 1 Tests # 2, 3, 4 ABCDEF GHJK ABC DEFGHJK 10 to 10 9
8 3
3 l
9 8
8 ff';
f,'?
8 to 8
7 3
2 9
2 2.!?
7 6
10 74 N7 9
3 6
5 3.
9 Ff-10 5
l 4
2 9
2 3
4 k
p rg 2
3 3
to 2
l 10 l
Test 4 5
t-
. FIGURE 4 FRIMARY STEAM FLOW SHED vs.
RECIRCULATION FLOW SHED a
40 -
July, 1965 after pump trip.
[
July, 1965 af ter loop closed.
M January, 1961 General Electric pump tests.
30 --
Slope = 0.625
~
N Y4 Q
~
20 --
D 1
5
~
u 4
1 b
m.
-a v
to __
2
~~
3 I
0-8 I
i l
0 10 20 30 4l0 50
- 7. Shed.in Reactor Recirculation Flow 1, 5 - 110 FMe, No Secondary Flow 2, 3, 4 - 160 FNe, Sece Secondary Flow O
O
28 -
5.
Iystem Stability Tests (Cont.)
c.
Transient Perf ot eance of Reactor and Associated Eculomant During Tornado Caused Shutdown The unit tripped out at 200 MWe when all the transmission lines (138 and 34.5 KV) were lost due to _the_ tornado on November 12, 1965.
The sequence of events and transients experienced were essentially the same as the full pouer generator trip test conducted on July 3,1965.
The only significant difference was the temporary loss of auxiliary power, a short time after the reactor scrammed, which cut in the emergency steam condenser and produced other minor phenomenon.
The reactor and its associated equipment functioned properly and a safe shutdown condition was easily achieved.
The transients experienced are depicted in Table 11.
The turbine was tripped on overspeed and normal shutdcwn procedures were follcued.
The exhaust hood temperatures had been running at 95 F. for the turbine end and 100 F. on the generator end prior to the outage.
Ia cdiately af ter the trip, both temperatures dropped to 65 F.
This was due to the loss of steam flow to the condenser while cooling water was still being supplied by the power generated as the turbine was dropping in speed.
The hood temperature then began to increase at an average 0
rate of 73 F. per hour until it reached 2230 F.
(3.3 psig.)
at 5:00 p.m. at which time the southuest rupture disc on the L. P.
turbine ruptured.
During startup on November 13, 1965 it was discovered that the steam seal dump valve to "A" extraction was stuck open allowing steam to enter the L.P.
turb ine.
It is believed that this valve was stuck cpen at the time of trip, allowing heating of the turbine exhaust hood.
The rupture disc was replaced by 3:20 a.m. on November 13, 1965.
The neutron flux increased on the average from 89 percent to a peak of 137. All six of the power range instruments saw this increase in flux.
The heat flux also increased by about 11%.
Incore instruments, though calibrated to ceasure heat flux, actually measures neutron flux.
Since the fuel has a long thermal time constant the heat flux at the c'ladding surface is _a function"of byth neu_ tron
~
~
flux and time during transients.
By analog computer studies of similar flux responses it was determined that the rise in heat flux was a small fraction of the rise in neutron flux.
. c.
Transient Performance of Reactor and Associated Equipment During Tornado Caused Shutdoun (Cont.)
Such a small increase in heat flux is possible, since the flux spike was only one or two seconds in duration, and the fuel has a time constant sf about 8 seconds.
The reactor scrammed when the neutron flux passed 1207.
on the out-of-core instruments and other scram signals were later initiated by (a) loss of oil pressure, (b) loss of auxiliary AC power, (c) the closure of primary steam valves and sphere isolation valves, (d) low condenser vacuum.
The actual scram terminated the flux increase and shut the reactor down to a level of about
- 67. from which it further decayed.
The flux spike lasted only about one or two seconds. All rods were inserted except C-9 and J-2 which were valved out of service.
After shutdown the reactor vessel temperature decayed very slowly until at approximately 'S:40 p.m. when cooling water was introduced into the reactor by the emergency primary feed pump.
This resulted in a rapid decrease in the temperature of the vessel bottom until at 9:50 p.m.
this 0
temperature reached a lov of 200 F.
The mid-section and flange temperature decay rates were not affected by this cooling water and the bottem to mid-point temperature differential approached 180 F.
The nominal limit for this differential is 1500 F.
The nominal limits for the temperature differentials were not even approached during the period of transients immediately follouing the scram.
Consultation with General Electric Company indicated that exceeding the differential temperature limit by 300 F. would not damage the vessel or shorten its lifetime.
Figure 5 shows the control rod patterns in use before and after the outage.
TABLE 11 REACTOR AND TURBINE-GENEPATOR
~
TRANSIENT DATA
(% Initial)
Final Initial Peak (Approx)
MWe 200 0
(0)
O tr.Je 617 685 (111) 7 Turbine Speed (RFM) 1500 1950 (109)
O Pri. Valve Position 86%
(0)
O Bypass Valve Position 0
1007.
(-)
0 Reactor Pressure 1007 1016 (101.3) 0 Neutron Flux (uua) 89 137 (154)
O Pri. Steam Flow 1.61 1.76 (111)
O Emergency Cond. Temp.
Stm. Inlet (CF.)
132 515 (390)
Water Storage 86 216 (251)
Exhaust Hood Temp. ( F.)
95 223 (235) 95 Cond. Vacuum (In. H )
28.3 28.9 (102)
O g
Pumps in Service 4
4 0
BOR 2.5 2.2 (89)
. FIGURE 5 CONTROL RCD PATTERNS BEFORE, DURIFG, AND AFTER UNIT OUTAGE DUE TO TORNADO, NOVEMBER 12, 1965 Rated Scrammed A B C.D E F G.H J K ABC DEF GHJ K 10 11 10
-- - l./ /,
~
9 8
4 5
to 9
8
/
'f W 5
8 Ic 10 8
/ 6' 4;, / Tf A
/
7 4
2 10 3
7 / b A '/, /0 ?'
9)'
/,
6 5
3 1
11 6 / S 7, '), ' 6!.'
7/ :; Z 5 t1
_1 3
s 5 'M',',Q
,) 5
'%,' 9- /7 4
3 ic 4 4't #$ % l ' I,E : 6 ',',
j y,j [ /
' l9 f;
3 10 1c 3
/
10 5
3 2
E' 41
//-@J1' 2
12
'///,l,
/
I ti i
y,'
11/12/65 11/12/65 2:40 p.m.
2:50 p.m.
h".le 200 FMe 0 Withdrawn 50 rods Withdrawn 1 rod 7 notches 8 notches Residual Residual
_ Critical Power ABCDEF GHJK ABC DEFGHJK 10
. 'A 7
10 11 9
8 9
8 2
10 8
8 10 7 30 3'?
7 7
7 10 2
6 6
2 2
11 5
S 11 2
2 4-4 2
10 2
3 3
7 10 7 !10 2
2 10 2
2 l
l l11 11/13/65 11/13/65 4:26 p.m.
11:20 p.m.
INe 0 IF.le 110 Withdrawn 36 rods Withdrawn 58 rods 3 notches Residual
. A.
Opera ting Exnerience (Cont.)
6.
Control Red Blades During periods of operation control rods have been verified for blade foll'oving on a weekly basis.
Monthly control rod worth tests were conducted after the refueling outage.
Prior to the refueling these tests were terminated to eliminate local flux peaking and thereby minicize fuel failures.
During each startup control rod patterns for criticality have been
~
~
predicted and all blade following verified.
7.
Changes in Facility Design a.
Off-cas Svstem Additions and Changes 1.
Fi_l te r ing At the request of the General Electric Company, additional particulate filters were installed during the refueling outage on the off-gas system to provide test data to aid in the design of the Unit 2 cff-gas system.
The new filter box uas installed at the base of the ventilatien stack and is divided to provide tuo flou channels. The north flow channel contains two MSA filters and the south channel contains two Cambridge filters. Flow is directed through either channel by means of a slide valve.
The slide valve is operated by tuo cables brought out through the stac!: below the access door.
2.
Of f-cas Instrur.entation Channes About two months af ter filing the September 18, 1964 request for modificati'on of the license specifications on maximun permissibic stack gas release rates, it was ascertained that certain changes, also sought, in the type of off-gas instrumentation and in the measuring method could properly be made under the license so long as there was reasonable assurance that the substitute system would provide the same nonitoring and centrol function as did the original system.
The changes uere cenpleted in May, 1965 and the substitute system was placed in service following the refueling outage, b.
Relocation of Off-Site Monitorine Station The Cc=nonwealth Edison Company sold a parcel of property which. included the Flainfield of f-site monitoring station.
b.
Relocation of Of f-Site Monitoring Station (Cont.)
On July 22 the station uss coved to substation J-59 which is located approximately one half mile north of the previous site, on the east side of State Route 59.
8.
Personnel Radiation Exposure Personnel exposures to radiation were within limits specified in 10 CFR Part 20.
9.
Liquid Poison Svstem The liquid poison system was operative at all times during the year.
The boron poison was sampled on January 31, ISy 25, and October 9.
There were no conditions which would indicate a loss of boren frca the solution tank.
Boren concentrations in the reactor water rcmained low throughout the year.
10.
Thoria Red Removal fren Tvne II Fuel Assemblies Removal of thoria rods frem Type II fuel assemblies was started on November 4, 1965. The method for removing the nine thoria rods from the 49 rod assemblies was by removal of the upper tie plate.
This method was selected af ter our experience with removing nine thoria rods frem each of tuo 64 rod assemblies, PF-ll and FF-12 and one Type II assembly by cutting the upper tie plate with an end milling cutter.
The cutting method was found to be very time consuming.
The upper tic plate of each Type II fuel assembly was removed by first cutting and removing the locking wires from the eight tie rods.
The nuts on each tic rod were then removed.
The tie plate was taken off, and the springs removed.
The nine thoria rods from each assembly were then transferred to one of three containers depending on the position and condition of the rods.
Thoria rods from known leaking fuel assemblies were transferred to an 88 tube container located inside of a single element encapsulation can that was modified for this purpose. A second 88 tube container was used for the storage of broken thoria rods or broken rods which could not be positively identified as to material content (UO2 or Th02).
Thoria rods from non-leaking fuel assemblies were placed in 34 tube containers. Uranium rods that had to be removed from damaged fuel assemblies were placed in 23 tube containers.
Scrap generated during this work was accumulated in the bottom " catcher" of the fixtures in which the work was performed or in GI cans.
Upon ccepletion of the work the scrap uns separated en an underwater table and placed in scaled 5" aluminum cask liners.
Fuel scraps were placed in two containers and wire, springs, nuts, bolts, etc., were placed in a third awaiting clearance from the U. S. A.E. C. (Savannah River Operations Office) as to other containment for storage and shipment. All fuels were carefully inspected to insure the complete removal of thoria rods.
G
10.
Thoria Rod Renoval from Type II Fuel Assemblies (Cont.)
Nuclear criticality in numerous containers was evaluated by both Commonwealth Edison and General Electric Company nuclear engineers.
The procedures followed for this work were evaluated and approved by the Commonwealth Ed.ison Company Safety Review Board prior to starting the work.
The Fuel Building ventilation system was revised temporarily to provide air flows across all work areas such that any possible gas released from damaged fuel would be carried away from personnel in the work areas. A ventilation air activity monitor and alarm system was also installed.
The regular Fuel Building area monitor alarm trip point was lowered to alarm at 5 mr/hr. above normal background as an added safety measure.
All thoria rods were re=oved frem the 107 Type II fuel assemblies by December 9, 1965. A total of 26 leaking or damaged Type II ascembly remaindsrs were then verified and encapsulated.
This work was ccepleted and locations of til fuels in storage were verified on December 21, 1965.
11.
Radioactive Uaste Disposal Release of radioactive liquid waste was accoeplished in batch quantities at controlled release flow rates according to established procedures.
The contribution to the activity of dilution water was always maintained within the limits specified in the applicable federal regulations.
The average contribution to the unidentified activity in the water utilized fer radio-active liquid waste dilution during the year was calculated to be.0273 uc/ml (27.3 uuc/ liter) compared to an average limit of 0.1 uc/ml (100 uuc/ liter) f.or unidentified mixtures containing no radium 226 or radium 228 as specified in CFR Part 20.
On January 2,1965 a concrete and steel cask containing 12 incore chambers, miscellaneous scrap, sludge and sweepings, and an insignificane unknown quantity of UO2 pellet material was shipped by truck to Nuclear Fuel Services, Inc. for permanent burial.
The activity of the shipment was 40 curies.
Solid radioactive wastes were stored on-site pursuant to 3
License DPR-1.
A shipment consisting of 507 f t of dry radio-active waste with a total activity of 158 millicuries was sent by truck to nuclear Fuel Services, Inc., a subsidiary of W. R. Grace and Co., West Valley, New York, on August 26, 1965.
A second shipment of 1019 f t3 of dry radioactive waste with an activity of 280 millicuries was also sent to Nuclear Fuel Services, Inc. by truck on August 31, 1965. A third shipment of 1803 ft3 of dry radioactive vaste with an activity of 140 millicuries was sent to Nuclear Fuel Services, Inc. by truck on September 2, 1965.
11.
Radioactive Waste Disposal (Cont.)
Concentration of noble fission products in the stack discharge to atmosphere was ' maintained well within license limits of 700,000 microcuries per second.
The average activity release rate for the year while the plant was operating was approxi-mately 24,800 ue/second.
A total of 168 Type I fuel assemblies were shipped to the Chemical Processing Plant of Nuclear Fuel Services, Inc., West Valley New York. All of this fuel had been removed from the reactor in the first re fueling outage (November,1962).
These assemblies were shipped in the N.F. S.
rail shipping cask.
Each shipment contained 24 assemblies for a total of seven shipments.
Table number 12 identifies the shipment number, date, and fuel shipped.
This fuel is the property of the U. S. A. E.C. per contract number AT(38-1)-315 and was shipped for and upon instruction from it's Savannah River Operations Office.
u 9
9 0
TABLE 12 SPENT FUELS SHIPPED TO NUCLEAR FUEL SERVICES FOR CHEMICAL PROCESSING, A. E. C.
LATCH #1 Shipment Number, Date, and Fuel Identity Number 1
2 3
4 5
6 7
Assemblies
_6-11-65 6-30-65 7-16-65 8-3-65 8-16-65 9-2-65 9-22-65 1
A-130 A-127 A-143 A-133 A-125 A-119 A-124 2
A-152 A-137 A-157 A-162 A-142 A-136 A-158 3
A-182 A-179 A-211 A 1.67 A-171 A-217 A-224 4
A-237 A-239 A-222 A-168 A-186 A-230 A-225 a
5 A-238 A-242 A-226 A-253 A-208 A-2' A-231 E
6 A-240 A-243 A-229 A-254 A-221 A-249 A-234 7
A-260 A-271 A-232 A-265 A-223 A-269 A-236 8
A-264 A-278 A-248 A-300 A-244 A-270 A-263 9
A-267 A-279 A-251 A 365 A-245 A-277 A-273 10 A-276 A-291 A-280 A-368 A-246 A-282 A-274 11 A-288 A-296 A-295 A-381 A-261 A-283 A-289 12 A-347 A-305 A-311 A-413 A-262 A-292 A-29 7 13
'A-372 A-310 A-358 A-436 A-275 A-298 A-349 14 A-407 A-417 A-371 A-437 A-281 A-301 A-354 t
%.f.
TABLE 12 (Cont.)
Shipment Number, Date, and Fuel Identity Number 1
2 3
4 5
6 7
Assemblics 6-11-65 6-30-65 7-16-65 8-3-65 7-16-65 9-2-65 9-22-65 15 A-424 A-422 A-378 A-451 A-286 A-356 A-355 16 A-426 A-440 A-397 A-476 A-306 A-373 A-370 17 A-427 A-442 A-398 A-481 A-353 A-376 A-419 18 A-435 A-443 A-403 A-483 A-404 A-383 A 432 19 A-439 A-494 A-406 A-484 A-433 A-453 A-441 20 A-445 A-495 A-498 A-488 A-450 A-454 A-444 EI 21 A-457 A-496 A-499 A-504 A-456 A-486 A-447 22 A-458 A-497 A-500 A-505 A-487 A-508 A-452 23 A-459 A-510 A-501 A-513 A-502 B- '1 A-503 I
24 B-20 A-512 B-3 B-21 A-506 B-22 A-511 Tota l To Date 24 48 72 96 120 144 168
- 12.
"C" Recirculation Loco Cn December 9 the "C" recirculation loop was taken out of service to replace the recirculation pump bowl gasket.
Af ter the repairs, a hydrostatic test revealed a small leak on the discharge valve bypass line.
The loop is still out of service as of December 31, 1965.
- 13. _ Tests a.
Temnerature and power Coeffi~cients of Reactivity Reactivity measurements were made af ter refueling operations in a xenon free state.
Temperature coefficient measurements were made at approximately 150, 250, 350, and 420 F.
Figure 6 is a plot of the calculated and experimentally determined tamperature coefficient of reactivity as a function of temperature.
Reactivity measurements were again made during heating and pressurizing at tuo temperature levals in a xenon free state during startup on October 16, 1965 to determine the effect of exposure on the reactor tecperature coefficient of reactivity.
The coefficients as exhibited in Figure 6 appear to have decreased since startup,and may 0
even be slightly positive at temperatures below 200 F.
The objective of performing the power coefficient of reactivity is to demonstrate that the reactor's pouer respcuse to small reactivity steps is well damped and self-limiting.
Ucutron flux levels were decreased to approximately 3 decades below heating level at which time a step increase in reactivity was inserted.
Heutron flux levels ware allowed to increase until they reached equilibrium or near equilibriua.
levels which indicated that the excess reactivity inserted had been absorbed by the fuel and moderator coefficients, b.
Fuel Sipning and Leaker Detection Fuel sipping was started on April 3 and completed on April 8.
Thirteen Type II, two Type I and FF-ll fuel assemblies were definitely established as leakers by " sipper" techniques.
Nine Type II's and fourteen Type I's were established as prime suspects.
The core location of all known leakers and suspects as established by sipper technique is indicated in Figure 7.
FIGURE 6 T_EliP"RATURE _C,C_EJFICIf3_GF RSACTIV_IP,
~
2 0 ---
X X
j u,
-6 Estimated values based oa g
hmaanerofrefueling.
n
' cs M
.-e VM asured values sulsequent x
s
/\\ to refueling, May 28-29, 1965 s -
Measured values, October 16, 1965 x
x k -
-14 ~~
-16 l
l j
q G
100 200 300 400 500 600 MODEPATOR TEITE"ATURE F.
- 40 FIGURE 7 LOCATION OF DEFECTIVE FUELS LEAKERS A!!D SUSPECTS BY TYPE 26 25 24 23 3'
3' 37 10 22 S1 21 S
9 i
Lo Sy to so 20 Lo l
S9 19 3
L>
so So 18 Si 7
S?
S2 I'o 17 16 1.1 33 s3 15 S2 L9 L3 L9 6
14 L1 13 5
S9 L?
S2 12 11 L2 4
10 S2
._01_
3 L?
L2 L2 08 07 2
06 si si S 3___
05 I,2 h
y 04 S1 03 02 01 A
B C
D E
F G
H J
K 51 52 53 54 55 56 57 58 59 60, 61 62 63 64 65 66 67 68 69 70, 71 72 73 74 Symbol Description yrl:c r S1 Suspect Type I 9
S2 Suspect Typs I' 14 L
Leaker Type ?.7-11 1
p L1 Lea.:er Type I 2
L2 Leaker Type II 13 39
41 -
13.
Tests (Cont.)
c.
Shutdeun Margin Checks Margin tests conducted on April 23, and 24,1965, shown in Figures 8, 9, and 10 indicated that the refueled core met the license requirements and that the margin was greater than 1.1 percent but less than 1.8 percent Zi k.
The 1.1 percent check was successfully performed but the 1.8 percent check could not be performed since critical was attained during withdrawal on the third rod.
Critical checking in the region of the critical E1 zone, Figure 8 indicated that the D-10 type control rod is likely to be stronger than the C-9 type.
Preliminary checks of potential criticals in the scuthucst zone were made en April 26, 1965 to determine the ccmbination of control rods uhich uould likely lead to a successful method of determining the absolute value of the
~
margin in the vicinity of the strongest roo.
d.
_ Critical Testinn of Refueled Core Critical checks uere performed subsequent to general margin testing af ter refueling to determine the value of the margin in the vicinity of the strongest rod in the core.
The method normally used during ref eling, Figure 17, does not establish the absolute value, as previously pointed out, but only that the license requirement is more than adequately fulfilled.
The first phase of this test consisted of determining the telative worths of th; strongest peripheral rods by withdrawal of single rods on fissica chanber response in the vicinity.
The second phase required the withdraual of the strongest rod and sufficient adjacent rod to attain a critical state and a subsequent calibration of the critical rod making use of period measurements and suberitical multiplication techniques.
The third and final phase consisted of a demonstration that the margins in all other regions were greater than that in the region of the test.
In-vessel safety system and lou level neutren monitors were used in the vicinity of all critical tests for data and local core protection during testing.
The results of such tasting are exhibited in Figures 14 through 17 and demonstrates that the minimum critical margin with the strongest rod withdraun is of the order of 1.45 percent.
- d.
Critical Testine of Refueled Core (Cont.)
The uhole core was critically tested on April 27, 1965 (prior to head replacement) by rod withdrawal in the normal sequence shown in Figure 18.
The reactor went critical on the 21st rod as contrasted to the 17,.37, and 44th rods required during initial critical testing in cycle 1, 2, and 3, as exhibited in the following tabic.
Cycle 1 Initial Core Item Leading
_C_ycle 2
_ Cycle 3
_ Cycle 4 Year 3/12/60 1/13/63 5/12/64 4/2/65 Number Refueled 440 192 96 200 Enrichmant of Fuel Installed 1.5 2.5 1.83 1.83 1.5 2.34 Initial Critical Rods 17 37 +
44 +
20 +
0 Reactor Uater Temperature 67 F.
76 F.
.100 F.
67 F.
c.
Automatic Dispatch System A number of tests were rade to determine the feasibility of applying Autematic Dispatch System operation to Dresden Unit 1.
These tests were limited to the load range
~
resulting from operation of the secondary steam control valves.
(110 MRe to 210 trJe).
The unit load was increased and decreased at a rate of 2 MUe per minute and 5 HNe per minute to insure there would be no adverse affects on the reactor system.
The unit load was then cycled with the governor controlling at a rate of 13 MWe per minute frcm 135 Fue to 185 MWe, dropped at 11 MWe per minute to 165 MWe and increased at 18 MUe per minute to 192 h"Je.
System components were stable and no operating problems were observed.
This load cycling was a more severe test than can be expected from A. D. S. ope ra tion.
The maximum load change that can be commanded by the A. D. S. is 7.3 Kle (3.57.) per minute, with normal operation being approximately 5 K.te per minute.
Throughout the tests the governor was not observed to be sticking and responded well to load change impulses.
Auxiliary equipment responded well and no operating problems were encountered.
FIGURE 8 SHUTDOWU MARGIN A!'D CRITICAL CHECKS SUIS' RY A
April 23, 1965 Rod positions during critical check 10 12 8
6 7
0 ABC_DEFGHJK ABC DEF GHJ K 10 12 12 10 12 12 9
12 12 9 ~
7 8
8 7
7 6
6 5
5 4
4 3
3 I
1 5:48 p.m.
5:55 p.m.
Suberitical Critical 1.17. Margin Cheek ABCDEF GHJK ABC DEFGHJK 10 1:
10 10 9
10 9
12 8
8 7
7 6
6 5
5 4
4
~
3 3
2 2
I I
l 6:10 p.m.
6:15 p.m.
Critical Subcritical Note:
Bottom figures indicate that D-10 is stronger than C-9.
- 44 FIGURE 9 SHUTDOWN MARGIN O'ECK?
PERIPHERAI, SUMIAR'{
a-ABCDEFGHJK ABC DEFGHJ K 10 I
10 12 12 1
9~
9 12 12 8
8 7
7 6
6 5
5 4
4
~
3 3
2 2
12 12 I
I 12 12 4-23-65 4-24-65 5:45 P.M.
12: 17 A. M.
North side South side ABCDEF GHJK ABC DEFGHJK 10 10 9
12 9
12 8
8 7
12 7
~
12 6
6 5
5 4-12 4
12_
3 3
2 12 2
12 1
I 4-24-65 5: 41 A. M.
7 : 11 A. M.
West side East side
FIGURE 10 NORI!AL SUBCRITICAL l'ARGI" CHEC"S PERIPHERAL ZONE CHECXS ABCDEFGHJK ABCDEFGHJ K 10 x
x 10 x
x 9
x x
9~
x x
8 8
7 7
6 6
S S
4 4
3 3
2 2
I i
1.17.F!argin Check 1.87. F:argin Check CENTRAL ZONE ABCDEF GHJK ABC DEFGHJK 10 1
10 l
9 9
8 8
7 x
x 7
6 6
S S
4 4
3 3
2 2
I I
]
1.97. Margin Check
FIGURP. 11 PERIPEEFAL ZONE - CRI'"ICAL "J.2073 April 26, 1955 ABC.DEFGHJK ABC DEF GHJ K 10 10 i
9 9
8 8
7 7
6 6
5 5
3 4
12 4
12 l
3 3
p 2
2 I
I 6:34 P.M.
7:08 P.-M.
Suberitical Critical 178 Seconds ABCDEF GHJK ABC DEFGHJK 10 10 l
9 l
9 8
8 7
7 6
6 5
12 5
4 4
4 12 3
l 3
3 3
2 2
I I
l 7:22 P.M.
7:50 P.P.,
Cricical-Crititel 135 Secends 105 Ss::.n:!s
7IGURS 12 PERIPHERAL ZONE - CRITICAL CHECKS ABC.DEFGHJK ABC DEF GHJ K 10 10,
s 9
9 8
8 7
7 6
6 5
5 4
12 12 4
12, 2
3 3
12 2-2 I
I 8:00 P.M.
8:18 P.M.
Suberitical Suberitical ABCDEFGHJK ABC DEFGHJK 10 10 9
9 8_
8 7
7 6
6 5
5 4
12 3
4 l
3 5
3 2
2 I
i l
8:25 P.M.
Critical 85 Sccands
I
- FIGURE 13 PERIPHERAL ZO!G - C'tITICAL CHECKS ABCDEFGHJK ABC DEFGHJ K 10 10 i
9 9'
8 8
7 7
6 6
5 5
4 4
3 3
2 3
3 2
12 1
12 1
12 9:48 P.M. Suberitical C-2 @ 2 10:23 P.M. Suberitical 9:50 P.M. Critieni C-2 @ 3 7 Seconds ABCDEFGHJK ABC DEFGHJK 10 10 l
9 l
9 8
8 l
7 7
6 6
5 5
4 4
3 6
3 2
12 2
1 12 I
10:36 P.M. Crs.cical 90 Secer.ds
PIGUP2 14 STUCK ROD MAROIN - CRITICAL TESTIN0 RE0 ION OF STROF0EST CONTROL' ROD April 27, 1965 ABC,DEFGHJK ABC DEFGHJ K 10 12 10 l
9 9
9.
8 8
12 7
7 12 6
6 5
5 l
4 4
3 3
2 2
I I
1 4/28/65 8:45 P.M. Critical C-9 @ 9 120" period 12:59 A.M. Suberitical 9:18 P.M. Suberitical C-9 @ 8 10:18 P.M. Critical C-9 G10 58" period 10:52 P.M.
Critical C-9 @ll 52" period 11:15 P.M. Critical C-9 G12 45" period ABCDEF GhJK ABCDEFGHJK 10 12 10 9
12 9
5 5
8 12 8
7
~
7 6
6 5
5 4.
4 3
3 2
2 1
i 1:30 A.M. Suberitical l'- 4 5 A. M. Suberitical
FIGURE 15 CRITICAL CHECKIt:G IN REGICITS OF STR0!!CEST RCD 5 April 27 6 28, 1955 Relative Strength Check of Strongest Critical Chee': Involvine Strencest Red Rods and Criticci Check A B C D E F/G 3 J K ABC DEF GHJ K Critical Suberitical N
10 lb
( 12 10 SV 10 9
TM
)
K p '~
9 ~
'12 I
s 8[72 6
/ 12 8
\\
/
7; 12 l
\\
2 7
6
/
x 6
5 w
5
^
4 12
\\
1:
4
\\
7X j
3 12 /
J l12 3
/
12h 2
12
)
2 j l2 g
l
't 12
)
12 I
a(
12 /
l Q
Critical All Regions Suberit cal Except D-12, C-9 Region Initial Core, Cycle 01 Critical Zones Corearisons Central Zone Pargin Checks ABCDEF GHJK ABC DEFGHJK Suberitical uberitical W ( C 10 10 d
s 9
5 5
i
/
5 9
/
\\
5 8
8 x
7
~
l 7
/ 12 12 t
6 6
12'
( 12) 5 5
L 12 <
\\ 12) 4 c
4 i2 12'M 3
/
3 2
5 5
/
2 I
vN suberitical I
k 1.
All regions suberitical, All Regions Suberitical third refueling.
2.
Similar regions critical during initial core, cycle 01 testing.
FIGURE 16 STROMGEST ROD, SU3C'1ITICAL MARGIN C'4ECK SUBCRITICAL MARGIN 1.45 1 0.27.
47"
+ 0.1 --
89" 188" K = 1.0 Critical 1590" 0.0
^
- 236" e
- 0.1
-106" A B C D E F G H J K
- 0. 2 - -
'= 12.
10 12
- 0. 3 --
- 83 9
C 4=6.5 8
7
- 0. 4 --
License 6
Requ ireraent 5
- 0.5 --
17.
M=2.9 E!
3 E
- 0.6 --
2 e M=2.25 5
1 g
- 0.7 --
C U
- 0.8 --
si a:1
- 0. 9 -
- 1. 0 --
M=1.35 Symbols:
- 1.1 ~~
C = Critical Rod H = Multiplication
- 1.2 --
" = Period in seconds
. M=1.15
- 1.3 --
- 1.4 -
- 1.5 l
l j
l l
l l
1 l
O 1
2 3
4 5
6 7
8 9
10 11 12 C-9 NOTCH FOSITION
FIGURE 17 STROI: GEST ROD, SUBCRITICAL l'\\PGIN CHECK
+0.099 +0.115
+0.0396 +0.073 Ke=1.0 Critical 8,
[0 11 jy 1
/ s 9
/\\
5 A
G^
Notch Position of 7
/\\
Calibrating Rod Suberitical !!argin 3
A A
0 Strongest Rod D-10 Withdrawn y
Minimun !!argin With Strongest Rod Withdrawn 1.457.
e
FIGUP2 18 INITIf,L G OLE C0"E CRITICAL CYCLE d4 SEQUE"C2 "A" April 27, 1965 Criticap B C D E F G H J K 10 N
X 9
x x
x 8
x x
7 x
6 x
x 5
4 3
x 2
x 1
x 2:10 p.m.
20 Rods 5 notches withdrawn to critical 67 F. Reactor water temperature ABCDEFG HJK 10 x
9 x
5 x
x 8
x 0 <- - Suberitical 7
0 Critical Maintained During 6
x o
x Rod Insertions Indicated.
5 x
o x
4 0
3 x
x 2
x x
l X
em
- 13. Tests (cont.)
f.
_ Safety Valves The five spare drum safety valves were tested for relief pressure on the new safety valve test fixture.
The fixture uses nitrogen in place of steam, thus eliminating the effect of temperature on consecutive pops.
This fixture was described in the 1964 Annual Report.
All valves were set at their respective design pressure
+ 10 psi.
Relief pressures vere checked by repeated popping.
Two of the valves, BA 7917 and SS 9705, were sent to Joliet Statien and tested for relief pressure with steam under simulated operating conditions.
Results were comparable to those obtained on the test fixture.
The State Inspector observed these tests and approved the new test fixture for all relief valve testing and adjustment except for setting blowdown.
g.
Sphere Leak Rate Test The reactor containment vessel was pressurized with air to 20 psig. on May 18 - 19.
Two leakage rate measurements were then made during consecutive 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> periods.
Due to the usual variations in outside temperature experienced during such tests, and the lag experienced in the reference system, the pariods used for calculation of results ware carefully selected.
The leakage during the first 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (6:00 a.m. on 5/20/65 to 6:00 a.m. on 5/21/65 was 0.177 1 0.025%.
The second test performed during the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (6:00 a.m. on 5/21/65 to 6:00 a.m. on 5/22/65) gave a calculated leakage rate of 0.180 1 0.0257..
The error analysis was computed for the instrumentation caly.
Calcula tion of the leak rate for the above two tests using the NASA (TN-D-1731) equation resulted in values of.184% and.191%/
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the first and second test respectively.
The allowable leak rate of 0.5%/24 hours at 37 psig, corrected to 20 psig. using the conservative laminar flow extrapolation equation results in an allowable leak race of 0.294%/24 hours.
A controlled leak approximately equal to the design leak rate of.57./24 hours at 37 psig. was imposed on the conta inment immediately following the two 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> tests.
The controlled leak was maintained for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and hourly readings were taken.
The sphere temperature increased continuously during this test resulting in an erroneously high leak rate value due to the lag between the containment e
g.
Schere Leak Rate Test (Cont.)
air and the reference system air temperature. The Icak rate was 0.507./24 hours including the controlled leakage.
Subtracting the controlled leak value of 0.23%/24 hours resulted in a containment leak rate of 0.27%/24 hours at 20 psig.
All tests therefore established that the containment vessel leakage rate is well within the design specification.
The mechanical penetrations of the containment were checked for leakage with soap solution and the electrical penetrations were checked with an electronic Freon leak detector.
Six leaks were found in mechanical penetrations and no leaks were' found in electrical penetrations.
The leaks found were estimated by timing the rate of growth of 2-inch diameter soap bubbles in five cases and a separate measurement was nade en the sixth and largest leak (venti-lation exhaust valves) immediately following the containment 3
tests. The sum of the five small lea'cs was 295 f t /24 hours.
3 The leakage at the exhaust valves nas 805 ft /24 hours.
h.
Air Loc!<s All air locks, ventilating valves, and process isolation valves uere tested during the year as required by License DPR-2 and found to be within the licensed allowable leakage rate.
i.
Emergency Condenser Tests The emergency condenser was testcd after the shutdown on March 28.
Steam flow was approximately 400,000 lbs/ hour.
The north outlet valve (MO-101) was opened at 9:44 p.m.
Visual inspection of the vent indicated approximately 350 gallons of water carryover. The total water removed from the condenser was 4,300 gallons. MO-101 was closed at 10:15 p.m.
The condenser handled 20.6 Fut for the north coil. A spot check of the south coil indicated the same capacity as the north coil.
Operation of the condenser was satisfactory throughout the test.
The emergency condenser was placed in operation automatically due to the loss of the safety system voltage on November 12 when all transmission lines vere lost.
During the half hour of operation, 71,700 lbs. of water was evaporated and 37.1 FNt was rejected to the condenser.
This compares favorably with the test results obtained in March.
56 -
III. DISCUSSION (Cont.)
B.
License DPR-2 Table 13 lists the amendments to our License that were pending during the year or authorized.
Pertinent correspo_ndence partaining thereto are listed in the Correspondence References.
w 0
w e
M e
O
51 -
TABI-E 13 SUM!ARY OF LICENSE AIENDMENTS PENDIEG DURIi G 1965 Date
~ _ ~} Recuest Authorization General authorization to take credit for use of respiratory protective 9/16/64 2/4/65 equipment. (Last Submittal) Amend Appendix A, License DPR-2 revising the specifications of maximum permissible stack release rate to include 1131 and halogens. 9/18/64 Modification of above request. 10/21/64 Issuance of Special. Nuclear Material License SNM-847 authorizing use of Knapp Mill shipping casks. 11/27/64 1/12/65 Request to amend License DPR-2 to permit operation with 104 Type III-F fuel assemblies with recovable poison rod. 12/24/64 5/23/65 Renewal of Special Nucicar Material License SNM-225. 1/19/65 2/11/65 Request to amend License DPR-2 to permit operation of PF-10 with one rod removed. 4/14/65 4/28/65 Request to amend Special Nuclear Material License SNM-638 to allow shipment of Batch No. I fuel to NFS plant. 5/14/65 5/26/65 Request to amend Special Nuclear Material License SNM-847 to allow transport of high irradiated fuel in Knapp Mill casks. 5/14/65 8/9/65 Request to amend Special Nuclear Material License SNM-225 to remove incore fission chambers in air. 5/21/65 5/21/65 Request to amend Special Nuclear Material License SNM-638, to trat.sfer urania rods of Type II fuel. 12/23/65
- Correspondence References - 1965:
(1) Letter to A. E.C. dated January 19, 1965, requesting renewal and amendment of Special Nuclear Material License SNM-225. (2) Letter from A.E.C. dated February 4,1965, authorizing credit for use of respiratory protective equipment. (3) Letter and addendum to Exhibit 1 to A.E.C. dated March 2, 1965, for amendment of Appendix A of License DPR-2 to permit eperation with Type III-F fuel. (4) Letter to Commonwealth Edison Company dated March 23, 1965, authorizing the loading of 104 Type III-F fuel assemblies. (5) Letter to A.E.C. dated April 2,1965, inviting Mr. R. R. Maccary to observe the 20 psig. sphere leakage rate tests. Enclosed was a copy of the test procedure. (6) Lette r to A. E. C. dated April 14, 1965, requesting that Appendix A of . DPR-2 be amended to authorize the operation with Type FF-10 fuel assembly, codified. (Change No. 11) (7) Letter to Commonuealth Edison Company dated April 28, 1965, authorizing operation with Type PF-10, modified. (8) Letter to A. E. C. dated May 6,1965, regarding compliance with Section E. 6.b., Appendix A to License DPR-2. (9) Letter to A.E.C. dated May 14, 1965, requesting amendment No. I to A.E.C. License No. SNM-847, to transport an upgraded fuel load in the Knapp Hills casks. s10) Letter to A.E.C. dated May 21, 1965 requesting a=endment to Special Nuclear Material License SNM-225 - removal of irradiated incore fission chambers in air using cask shielding techniques. (11) Letter to A.E.C. and enclosed certificate dated June 10, 1965 in compliance with condition 8(c) of Special Nuclear FSterial License SNM-638. (12) Letter to Commonwealth Edison Company dated August 9,1965, amendment No. 1 to A.E.C. License No. SNM-847. (13) Letter to A.E.C. dated September 8, 1965, requesting amendment to License No. SNM-638 to permit shipment of Dresden Processing Batch No. 3. (14) Letter to A.E.C. dated October 14, 1965, explaining malfunction of control rod J-2. (15) Letter to A.E.C. dated October 27, 1965, further explaining malfunction of control rod J-2. O e
s Correspondence References - 1965: (Cont.) (15) Letter to A.E.C. and enclosed certificate dated June 10, 1965 in ccepliance with condition S(c) of Special Naclear Material License SNM-638. (16) Letter to Ccmmonwealth Edison Company dated August 9,1965, acendment No. I to A. E. C. License No. S131-847. (17) Letter to A.E.C. dated September 8,1965, requesting amendment to License No. SNM-638 to permit shipment of Dresden Processing Batch No. 3. (18) Letter to A.E.C. dated October 14, 1965, explaining malfunction of control rod J-2. (19) Letter to A.E.C. dated Cetober 27, 1965, further explaining malfunction of control rod J-2. e}}