ML20207K035

Forwards Util Which Transmitted Corrected Pages to SG Replacement Outage Startup Rept.Subject Ltr Was Inadvertently Not Sent to NRC Dcd,As Required by 10CFR50.4

ML20207K035
Person / Time
Site:	Byron
Issue date:	03/04/1999
From:	Levis W COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BYRON-99-0033, BYRON-99-33, NUDOCS 9903170025
Download: ML20207K035 (2)

Text

f

~'

W

, Commonwealth Edivin Company

)

2 a 9) ton Generating Station 4 iso North German Church Road j ll) ton,11. 61010979 i

]

j Teimse s1n j

March 4,1999 I I

LTR: BYRON 99-0033 File: 2.01.0700 l

l U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 1 Facility Operating License No. NPF-37 NRC Docket No. STN 50-454 l

Subject:

Steam Generator Replacement Outage Startup Report - Corrected Section 2.0  ;

Reference:

Commonwealth Edison Company letter, " Steam Generator Replacement Outage Startup Report Byron Station Unit 1," dated June 3,1998 The attached letter was inadvertently not sent to the NRC Document Control Desk as required j by 10 CFR 50.4. The letter transmitted corrected pages to the referenced report, which was originally sent to the Document Control Desk.

Should you have any questions concerning this matter, please contact Mr. Brad Adams at (815) 234-5441, extension 2280.

Regectfully,

/Y O' A &

William Levis Station Manager Byron Station b' h/

T Attachment KLG/JL/cib 3g0101 cc: Regional Administrator- NRC Region lll NRC Senior Resident inspector- Byron Station 9903170025 990304 PDR ADOCK 05000454 P ppg A l'nicom Compan}'

t__

l T

}

Attachment Commonwealth Edison Company 1.etter to James E Dyer (USNRC), " Steam Generator Replacement Outage Startup Report-Corrected Section 2.0," dated January 6,1999

)

I l

i l

1 1

l l

l l

f P:\99bylus\990033. doc

r

, Commonw calth 1.dison Company

.. 11)run C.cnerating Station 4 450 North German Church Road I)) ron,11. 610109794 Tel H 15-231-5 H 1 January 6, 1999 l

LTR: BYRON 98-0324 FILE: 2.01.0700 j

James E. Dyer Region III Administrator U.S. Nuclear Regulatory Commission Region III 801 Warrenville Road Lisle, IL 60532-4251 Byron Station, Unit 1 Facility Operating License No. NPF-37 NRC Docket No. STN 50-454

SUBJECT:

Steam Generator Replacement Outage Startup Report - Corrected Section 2.0

REFERENCE:

Comed letter, " Steam Generator Replacement Outage Startup Report,"

dated June 3, 1998.

In the referenced letter, we submitted the Startup Report to summarise the startup and power escalation testing required as a result of the steam Generator replacement modification for Byron Station Unit 1. Subsequent to the submittal we noticed several errors in Section 2.0 of the report, " Core Testing". The errors are administrative in nature that occurred during the transcription of data from source documents to the report and do not affect any of the report conclusions, qw2m5o1o Va p:\98byltrs\980324.wpf A Unicom Company

. . P;g3 2

's

• St ca Genarstor R: placement Please replace Section 3.0 of the original report in accordance with the attached update instructions. We apologize for any inconvenience this may cause you. Should you have any questions regarding this letter, please contact Joseph Langan at (815)234-5441, extension 2871.

Respectfully k& ==

William Levis Station Manager Byron Station Attachment WL/JL/cb cc: Byron Project Manager - NRR NRC Senior Resident Inspector - Byron Station Office of Nuclear Facility Safety - IDNS l

t l

p:\98byltrs\980324.wpf

)..'-

=

f 1

- l

).

l Attachment Corrected Pages to Byron Station Unit 1 Cycle 9 Startup Report l l

l 1

1 l

- l 1

1 l

I 1

l e

i 1

I P:\98byltrs\980324.wpf

.f .

. Updato In3tructions:

M Insert Title Page Revised Title Page Section 2, pages 2 through 9 .New Section 2, pages 2 through 9 l

p:\98byltrs\980324.wpf I

Comed Byron Nuclear Power Station Unit 1 Cycle 9 Startup Report December,1998 Revision I t

i 4

G@MITa C3 ~7 lof l

1.0 Introduction Commonwealth Edison conducted a comprehensive test program following replacement of the Unit 1 Steam Generators (S/Gs) that demonstrated that modified structures, systems, and components perform satisfactorily in service. The test program outlined in this report summarizes events and testing performed during the first heatup and increase to 100% power with Byron 1 Replacement Steam Generators (RSGs). The testing scope included sequencing of special tests (SPPs) and station surveillances to satisfy requirements of the modification.

The Byron Unit 1 Cycle 9 core includes a feed batch of 77 fuel assemblies manufactured by l Westinghouse. The new fuel region incorporates Integral Fuel Burnable Absorber (IFBA) rods with a B-10 loading of 1.5X with a 100 psig backfill pressure, The 1.5X IFBA rods have been used in previous cycles, but unique to Cycle 9 is the reduction of backfill pressure from 200 psig to 100 psig. In addition, enriched annular blankets are used on all feed assemblies (6" top and bottom). Table 1.1 contains characteristics of the Byron Unit 1 Cycle 9 core design.

The Cycle 9 reactor core achieved initial criticality 3/8/98, at 0150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.

The Unit 1 Main Generator was synchronized to the grid 3/9/98 at 0609 hours0.00705 days <br />0.169 hours <br />0.00101 weeks <br />2.317245e-4 months <br />.

Power escalation testing, including testing at full power, was completed 4/4/98.

Table 1.1 Byron Unit 1 Cycle 9 Core Design Data

1. Unit 1 Cycle 8 burnup: 433 EFPD

2. Unit 1 Cycle 9 design length: 410.2 EFPD Region Fuel Type Number of Enrichment Cycles Burned Assemblies w/o U-235 9A VANTAGE + 24 4.0 2 9B VANTAGE + 16 3.6 2 10A VANTAGE + 36 4.4 1 10B VANTAGE + 40 4.2 1 11A VANTAGE + 44 4.0 0 11B VANTAGE + 32 3.8 0 11C VANTAGE + 1 1.6 0 i

l l

I i

1  !

i i

\

~

..- ,- y i

2.0 Core Testine 2.1 Control Rod Drop Time Measurement This test is performed prior to each fuel cycle with T., greater than 550 degrees F and with all reactor coolant pumps in operation (per Technical Specification 3/4.1.3.4). - Due to the potential for marginally increased reactor coolant system flow impacting control rod drop times, a verification of acceptable rod drop times was required as part of replacement steam gene:ator testing The individual full-length shutdown and control rod (RCCA) drop time from the fully withdrawn position is required to be less than 2.7 seconds from the beginning of decay of the stationary gripper coil voltage to dashpot entry.

I All BIR08 RCCA drop times satisfied this acceptance criteria. Table 2.1 summarizes the results 1 of drop time measurements from BIR08. -In addition, a comparison of drop times to previous

{

cycles is provided. Based on this information, there where no changes in the RCCA drop times  !

due to Steam Generator Replacement.

)i 2.2 Zero Power Physics Testing Zero Power Physics Testing (ZPPT)is performed at the beginning of each cycle as specified by ANS/ ANSI-19.6.1, " Reload Startup Physics Test for Pressurized Water Reactors." A summary of the Startup Physics Test results is contained in Table 2.2. All test results were determined to be acceptable.

j 2.3 Power Escalation Testing Power Escalation Testing is performed during the initial power ascension to full power for each cycle and is controlled by 1/2BVS XPT-3. Tests are performed from 0% through 100% with major testing plateaus at approximately 30%,75%, and 100% power. Significant tests included:

Core Power Distribution at 25%,46%, 60%, and 98% power. l e

Reactor Coolant Delta-T Measurement at 60% and 100% power.

Hot Full Power Critical Boron Concentration Measurement (100%).

Reactor Coolant System Flow Measurement at 60% and_100% power. l

]

2.4 Core Power Distribution Core power distr:bution measurements were performed during power escalation at low power 3

(<30%), intermediate power (40-75%), and full power. Measurements are made to verify flux symmetry and to verify core peaking factors are within limits. Data obtained during these tests q

are used to check calibration of Power, Range NIS channels and to calibrate them if required.  ;

Measurements are made using the Moveable Incore Detector System and analyzed using the j INCORE 3-D computer code. l 1

i 2

I

Following successful completion of the low power flux map, power was increased to the 50%

S/G testing plateau (calorimetric power of 46%). At this power level, the maximum corrected Fy was greater than Fy"", but less than Fy". This was the result of the core load pattern and INCORE constants, not the result of S/G replacement. This resulted in the allowable power {

increase limited to 66% power. Reactor power was subsequently stabilized at 60.4% and a core i power distribution flux map obtained. The corrected F, at this power level was less than Fy"",

and the unit was released for increase to full power.

Results of the core power distribution measurements at 25%, 46%, 60%, and full power are l {

shown in Tables 2.3, 2.4, 2.5, and 2.6, respectively.

2.5 Full Power Loop Delta-T Determination l

The purpose of this test is to determine the full power Delta-T for each Reactor Coolant loop in order to recalibrate any loop with significant change. This procedure is applicable in MODE 1 l

and is performed above 95% Rated Thermal Power (RTP) after each refueling outage. Results {

are contained in Table 2.7. I l

l l

A 3

1

\

Table 2.1 Unit 1 RCCA Rod drop Time Comparison  !

1 B1R06 B1R07 #1 B1R07 #2 B1R08 Startup l

D 02 1.53 D42 1.555 D42 1.520 D-02 1.54 B 12 1.5 B-12 1.530 B-12 1A00 B-12 1.51 M-14 1.535 M-14 1.545 M-14 1.520 M-14 1.55 P-04 1 A05 P 04 1.505 P-04 1 A05 P-04 1.52 844 1.40 B-04 1.510 B-04 1.475 B-04 1A0 D-14 1.53 D-14 1.575 D-14 1.510 D-14 1.55 P-12 1A65 P-12 1 AGO P 12 1.475 P-12 1.40 M-02 1.555 M-02 1.580 M-02 1.535 M-02 1.50 G-03 1.51 G 03 1.535 G 03 1.515 G-03 1.52 C-00 1A75 C-00 1.520 C 00 1 A70 C-00 1.5 J-13 1.53 J-13 1.565 J-13 1.520 J-13 1.57 N47 1.5 N47 1.510 N47 1A85 N-07 1.52 C-07 1A75 C-07 1.515 C-07 1 A65 C-07 1.51 G 12 1A05 G-13 1.520 G-13 1A85 G-13 1.52 I N40 1.5 N-00 1.525 N40 1A70 N-00 1.5 J-03 1.5 J-03 1.525 J-03 1 A00 J-03 1.54 E 03 1.55 E-03 1.585 E 03 1.535 E-03 1.56 C-11 1.515 C-11 1.500 C 11 1.500 C-11 1.52 l L 13 1.53 L-13 1.550 L 13 1.520 L-13 1.53 N-05 1.525 N-05 1.520 N-05 1.520 N-05 1.52 C 05 1.485 C-05 1.520 C-05 1A05 C 05 1.40 E 13 1.51 E-13 1.575 E-13 1.510 E 13 1.54 N-11 1.505 N-11 1.525 N-11 1 A00 N-11 1.5 L 03 1.54 L-03 1.545 L-03 1.535 L-03 1.55 H-04 1.525 H-04 1.530 H-04 1 A00 H-04 1.5 D-08 1.495 D-08 1.535 D-06 1.480 D-08 1.51 H-12 1.5 H-12 1.515 H-12 1

1 A00 H-12 1.51 i M-08 1.505 M-08 1.525 M-08 1 A75 M-08 1A3 H-06 1.5 H-06 1.525 H-06 1.405 H-06 1.5 H-10 1.49 H-10 1.560 H-10 1 A80 H-10 1.52 F-06 1.525 F-08 1.555 F-08 1.505 F-08 1.5 K-08 1.52 K-08 1.535 K-08 1A00 K-08 1.51 F 02 1.55 F 02 1.560 F-02 1.505 F 02 1.52 B-10 1.515 B 10 1.525 B-10 1 A00 B-10 1.51 K 14 1.53 K-14 1.540 K-14 1.500 K-14 1.54 P 06 1A0 P-06 1.500 P-06 1 A00 P-06 1.52 j B-06 1.40 B 06 1.505 B-06 1A75 B-06 1.5 I F-14 1.55 F-14 1.605 F-14 1.535 F-14 1.50 P 10 1.515 P-10 1.525 P-10 1.405 P-10 1.52 K 02 1.54 K-02 1.580 K42 1.530 K-02 1.57 H-02 1.51 H-02 1.510 H-02 1.510 H-02 1.52 B-08 1.5025 B-08 1.515 B48 1 A85 B-08 1A0 H-14 1.5 H-14 1.520 H-14 1 A05 H-14 1.52 P-08 1.515 P 08 1.545 P 08 1.520 P 06 1.56 F-06 1 A75 F-06 1.530 F 06 1.505 F-06 1.51 F-10 1.515 F-10 1.580 F 10 1.525 F-10 1.53 K-10 1.475 K-10 1.545 K-10 1 A00 K-10 1.52 K-06 iA65 K-06 1.510 K-06 1 A70 K46 1A7 D-04 1.505 D-04 1.525 D 04 1 A85 D-04 1.5 M-12 1 A05 M-12 1.615 M 12 1.500 M-12 1.52 D-12 1 A85 D-12 1.625 D-12 1 A80 D-12 1.5 4

Table 2.1 Unit 1 RCCA Rod drop Time Comparison B1R06 B1R07 #1 B1R07 #2 B1R08 Startup M44 1.47 M44 1.406 E04 1.400 Mod 1.5 He 1.525 He 1.575 He 1.505 He 1.52.

Average 1.508 Average 1.540 Average 1.498 Average 1.521 Std. Dev. 0.02306 Std. Dev. 0.03046 Std. Dev. 0.01935 Std. Dev. 0.02563

+2 Sigma _ 1.554 +2 Sigma 1.600 +2 Sigma 1.537 +2 Sigma 1.572

-2 Sigma 1.462 -2 Sigma 1.479 -2 Sigma 1.459 -2 Sigma 1.469 Total 0.03 Sigma Total Average 1.516 Total +2 Sigma 1.58 Total -2 Sigma 1.46 5

l

i

]

1 I

Table 2.2 BIR08 Startup Physics Test Results {

Review Acceptance Parameter Predicted Measured Difference Criteria Criteria ARO Critical Boron 1437 ppm 1397 ppm 40 ppm 50 ppm 1000 pcm Critical Cs with Reference 1304 ppm 1265 ppm 39 ppm 1500 pcm N/A l Bank Fully Inserted 1

Differential Boron Worth -8.32 -8.37 0.6% 10% of design N/A pcm/ ppm r / ppm

. ARO ITC -2.814 -2.82 0.0 2 pcm/F of N/A pcm/F m/F design value ARO MTC -1.134 -1.14 0.0 N/A Within Tech pcm/F ro i/F Spec 3.1.1.3 4 Reference Bank 1109 pcm 1102.7 pcm -0.57% 510% between $15% between I (Shutdown Bank B) measured & measured & j Worth design design  !

Control Bank A Worth 309 pcm 254.6 pcm -54.4 pcm s15% ors 100 530% or 5200 pcm of design pcm of design Control Bank B Worth 789 pcm 818.3 pcm 3.71 % $15% or $100 530% or 5200 l pcm of design pcm ofdesign  ;

Control Bank C Worth 686 pcm 648.1 pcm -5.52% $15% or $100 530% or5200 I pcm of design pcm of design Control . Bank D Worth 573 pcm 551.5 pcm -21.5 pcm 515% or $100 530% or5200 pcm of design pcm of design Shutdown Bank A 233 pcm 227.8 pcm -5.2 pcm $15% ors 100 530% or5200 l pcm of design pcm of design Shutdown Bank C 448 pcm 444.4 pcm -3.6 pcm $15% or $100 $30% or5200  ;

pcm of design pcm of design {

Shutdown Bank D 450 pcm 445 pcm -5.0 pcm 515% or$100 530% or 5200 l pcm of design ocm of design I Shutdown Bank E 534 pcm 503.5 pcm -30.5 pcm 515% or $100 530% or5200 pcm of design pcm of design Total Rod Worth 5131 pcm 4995.9 pcm -2.63% $10%between >90% of the j measured & predicted sum 1 design of bank worths i

6

t' Table 2.3 l Core Power Distribution Results l 25% Power k Plant Data I Map ID: Byl0901 Date ofMap: 3/11/98 f

{

Cycle Burnup: 0.9 EFPD l j Power Level: 24.9 %

Control Bank D Position: 153 steps

)

l INCORE 3-D Results Core Average Axial Offset -0.139 l Tilt Rations for Entire Core Height: Quadrant 1: 1.0008 Quadrant 2: 1.0107 ,

Quadrant 3: 1.0197 Quadrant 4: 0. % 88 Maximum corrected Fxy: 1.8674 Fxy""- 1.930 Table 2.4 Core Power Distribution Results 46% Power -

Plant Data -

Map ID: Byl0902 1

Date ofMap: 3/13/98 '

Cycle Burnup: 1.51 EFPD Power Level: 45.9 %

Control Bank D Position: 172 steps INCORE 3-D Results Core Average Axial Offset .795 Tilt Rations for Entire Core Height: Quadrant 1: 0.9933 Quadrant 2: 0.9922 Quadrant 3: 0.9964 Quadrant 4: 1.0182 Maximum corrected Fxy: 1.9556 Fxy*"- 1.930 Max. Nuclear Enthalpy Rise Hot Channel Factor: 1.7248 Nuclear Enthalpy Rise Hot Channel Factor Limit: 1.9178 7

Table 2.5 Core Power Distribution Results

, 60% Power l Plant Data Map ID: Byl0903 Date of Map: 3/17/98 Cycle Burnup: 3.42 EFPD Power Level: 60.4 %

Control Bank D Rod Position: 185 steps INCORE 3-D Results Core Average AxialOffset -0.619 Tih Ratios for Entire Core Height: Quadrant 1: 0.9998

)

Quadrant 2: 0.9982 l Quadrant 3: 0.9893 )

Quadrant 4: 1.0226 j Maximum corrected Fxy: 1.7216 Fxy"": 1.930 l Max. Nuclear Enthalpy Rise Hot Channel Factor: NA Nuclear Enthalpy Rise Hot Channel Factor Limit: NA l l

Table 2.6 Cort Power Distribution Results Full Power Map Plant Data Map ID: By10904 Date of Map: 3/30/98 l Cycle Burnup: 14.6 EFPD l Power Level: 97.9 %

Control Rod Position: 219 steps 1 INCORE 3 D Results Core Average Axial Offset -2.681 Tilt Ratios for Entire Core Height: Quadrant 1: 1.0017 Quadrant 2: 0.9833 Quadrant 3. 1.0026 Quadrant 4: 1.0124 Maximum corrected Fxy: 1.6760 Fxy"": 1.930 Max. Nuclear Enthalpy Rise Hot Channel Factor: 1.5421 Nuclear Enthalpy Rise Hot Channel Factor Limit: 1.7107 8

T Table 2.7 Full Power Loop Delta-T Loop That Tcold Full Power Previous Delta-T Cycle Delta-T A 610.2 553.0 57.2 59.1 B 609.5 551.9 57.6 60.6 C 610.6' 553.1 57.5 60.8 D 611.2 552.7 58.6 62.1 9

L