ML20118B060

Post-Refueling Startup Test Summary Rept,Cycle 4

ML20118B060
Person / Time
Site:	Perry
Issue date:	09/22/1992
From:	CLEVELAND ELECTRIC ILLUMINATING CO.
To:
Shared Package
ML20118B059	List: ML20118B060 PY-CEI-NRR-1554, Forwards Perry Nuclear Power Plant Unit 1 Post-Refueling Startup Test Summary Rept,Cycle 4. Rept Submitted Based on Installation for Cycle 4 of Fuel W/Different Design than Fuel Used in Cycle 3
References
NUDOCS 9209290347
Download: ML20118B060 (20)

Text

Y.

Attachment 1

. PY-CEI/NRR-15$4 L Page: 1 of 20 THE CLEVELAND ELECTRIC ILLUMINATING COMPANY PERRY tiUCLEAR POVER PLANT UNIT 1 POST REFUELING STARTUP TEST

SUMMARY

REPORT CYCLE 4 ,

1

+

U 9209290347 920922 PDR ADOCK 03000440 PDR 1

m. - ..._ < _ . - _ - _ . _ . - . . _ - - . . - ~ . _ . . __ - - _ __ _ _ . _

n Attachment 1-

. PY-CEI/NRR-1554.L Page: 2 of 20;  !

1.1 INTRODUCTION

This report presents _ a int ary of the results from the post- refueling startup tests which were conducted in preparation

- for Cycle 4 at Unit 1 of-the Perry Nuclear Pover Plar.t. This report is submitted pursuant to Technical Specification 6.9.1.1.

1.2 PLANT DESCRIPTION L

The Perry Nuclear Power Plant is operated by the Cleveland-Electric Illuminating Company.(CEI) and is located near Lake Erie in Lake County, Ohio. Unit i has a Boiling Water Reactor (BVR) nuclear steam supply system as designe' and supplied by the General Electric Company (GE) and designated BVR/6, with a Mark III containment. The_ balance of plant was designed by Gilbert Associates, Inc., Reading, Pennsylvania, as architect-engineer.

The rated core thermal power is 3579 MVt with a gross electrical output of 1250 MVe. The turbine is an 1800 rpm tanuem compound, .

six flow, reheat unit consisting of one double flow high pressure stage in' tandem with three low pressure stages. The generator is a direct coupled 60 Hz, 22 KV, three-phase unit with a water cooled stator and hydrogen cooled rotor.

1.3

SUMMARY

OF ACTTVITIES DURING REFUELING OUTAGE 3 The third refueling outage at Unit 1 of the Perry Nuclear Power Plant egan L

on March 21, 1992 (main generator off-line) and was completed on June 13, L 1992 (main generator on-)ine). The outage duration vas 84 days.

Key activities of this refueling outage were: the performance of a core shuffle, replacement of 204 fuel bundles (out of 748), replacement'of 12 control rod blades _(out of 177), the installation of over_50 design changes,-

the completion of a large number _of_ corrective and preventive maintenance-activities, and the performance of-required Technical Specification-surveillance tests .

Start Date 21 March, 1992 Stop Date - 13 June, 1992 Duration _~84-days

'Vork Ordert .1662 Design Changes 58 Surveillances 481-Repetitive Tasks 1109 l-b-

t

- - . , . . - - , . , , , . . m-.,mr, p ., ,,E.% , , , _ ,,, . -y,._.

^

Attachment 1 PY-CEI/NRR-1554 L Page: 3 of 20

2.1 DESCRIPTION

OF DIFFERENCE IN FUEL DESIGNS Unit 1 at the Perry Nuclear Power Plant used the following General Electric fuel designs for cycle 3:

GE6B-PBSIB176-4GZ-120M-150-T 140 bundles, GE6B-P8 SIB 219-4GZ-120M-150-T 64 bundles, GE8B-PBSOB301-5GZ-120M-150-T 136 bundles, GE8B-P8 SOB 301-7GZ-120M-150-T 136 bundles, GE8B-P8 SOB 320-9GZ-120M-150-T 104 bundles, GE8B-P8S0B322-707-120M-150-T 168 bundles.

and the following General Electric fuel designs for cycle 4:

~

GE6B-P8 SIB 176-4GZ-120M-150-T 5 bundles, GE8B-P8 SOB 301-5GZ-120M-150-T 136 bundles, GE8B-P8 SOB 301-7GZ-120M-150-T 135 bundles, GE8B-P850B320-9GZ-120M-150-T 104 bundles, GE8B-P8 SOB 322 7GZ-120M-150-T 164 bundles, GE10-P8SXB306-10GZ2-120M-150-T 68 bundles, GE10-P8SXB306-11GZ3-120M-150-T 135 bundles.

Complete descriptiens of these General Electric fuel designs are given in GESTAE II, General Electric Standard Application for Reload Fuel. The major differences in the GE10 fuel relative to the GE8 fuel include:

1) An increase in the number of gadolinium rods in the bundle.

(10 or 11 vice maximum of 9 previously)

2) Use of a large central vater rod instead of two small vater rods.

3) Fever fuel rods per bundle.

8 (60 for GE10 vice 62 for GE8)

4) Use of an interactive channel with machined flow trippers.

5) Axial enrichment zoning of Uranium. _

6) Ferrule spacer design and resulting improved bundle flev.

7) Expanded and offset lattice design.

A l

l 5 [

- . ~ _ .

-Attachmentll PY-CEI/NRR-1554 L-Page: 4 of 20 3.1 POST-REFUELING STARTUP TEST PROGRAM During refueling operations and the subsequentLreturn'to pc ar, activities vore controlled under normal administrative programs rather than a separate formally defined post-refueling startup test progra .

These administrative programs cover areas of normal operation surb e Design Changes / Post-modification Testing Post-maintanence Testing Technical Specification Surveillances Inservice Inspection Special Nuclear: Material Control Periodic and Special Tests Computer Software Modification Radiation Control

~

The acceptance criteria for these tests were derived from the requirements of the administrative programs- .The reactor conditions for conducting the tests vete g. tided by requirements in the appropriate-administrative program.

3.2 POST-REFUELING STARTUP TEST REPORTS As required by Technical Specification 6.9.1.2, this report addresses each of the startup tests identified in USAR Subsection 14.2.12.2. Each' test was evaluated by the Responsible System Engineer who determined-whether the test was impacted by any refueling activity. Those tests 4

determined not to be impacted by any refueling activity are listed in

.dble 3.2-1.

For those tests which were impacted by refueling activities, this report lists:

1. A description of the measured values of the operating conditions or characteristics obtained during the test program and a-comparison of these values with design predictions and specifications;

2. A description of any corrective ections that were required to obtain satisfactory operation;

3. Any additional specific detailsLrequired in license conditions ' based on other-commitments..

b

_ _ . . . . . . _ . - ~ . _ . - _ . _ . - - . . . , _

4.

Attachment 1-PY-CEl/NRR-1554 L Page: 5 of 20 3.3 POST-REFUELING STARTUP TEST REPORTS -- SUMMARIES 14.2.12.2.3 Test Number 3 - Fuel Loabing Test Objective The purpose of this test is to load fuel safely and efficiently.

i Discussion Fuel unloading and loading vas conducted in Operational Condition 5 under 101-9, " Refueling." Fuel movement followed a predetermined plan in accordance with a Fuel Movement Checklist per FTI-D09, "Use of the Fuel Movement Checklist." A core shuffle (vice an offload / reload) was _

performed to minimize fuel movement.

Fuel movement was performed in several phases. First, fuel was offloaded 0 from control-cells whose control rod drivesLor blades were to be replaced. After these bundles were offloaded, the control rod drive and-blade work commenced. Control rod drives were replace to enhance performance during the upcoming cycle. Control rod blades were replaced to ensure that the depletion limits sould not be exceeded during the cycle. During the-time at which any control rod was withdrawn and/or removed, no fuel was loaded into the. reactor. By the completion of the control rod verk, all control- rods were fully inserted. The core shuffle was then started. Discharge fuel was removedLfrom-the core, new fuel inserted, and reload fuel moved within the reactor vessel. All fuel moves involving the reactor vere analyzed for shutdown margin, within a-minimum analytical. margin of 1% delta-k/k. This ensured tne acceptance criteria of at least 0.38% delta-k/k vas met for all core configurations.

g Attachment 1 PY-CET/NRR-1554 L Page: 6 of ?O 14.2.12.2.4 Test Number 4 - Full Core Shutdown Margin Test Objective The purpose of this test is first to demonstrate the reactor is suberitical +hroughout the fuel cycle with any single control rod fully withdrawn and second to determine quantitatively the shutdown margin of the as-loaded core.

DISCUSSION Full core shutdown margin and reactivity anomaly were demonstrated to be within their Technical Specification require.nents during Cycle 4 startup (reactor startup Number SS). SVI-B13-T0001, "Insequence Critical _

Shutdovn Margin Calculation," vas performed at 0% -- ir and 143 degiaes F moderator temperature. The minimum shutdown margi.. ,r Cycle 4 was measured to be 1.18% delta-k/k, compared to the Technical Specification minimum of 0.38% delta-k/k. The measured teactivity anomaly at the beginning of cycle was 0.245% delta-k/k, compared to the Technical Specification maximum of 1.0" delta-k/k.

u

+

Attachment 1-

. PY-CEI/NRR .1554 L Page: 7/ of 20 14.2.12.245 Test Number 5 - Control Rod Drive System 1

Test Objective p The purposes of the control rod drive system tests are to demonstrate that the control Rod Drive (CRD) svatem operates properly over the full range of

reactor coolant temperatures and pressures from ambient to operating, and to determine the initial operr, ting characteristics of the entire CRD system.

Discussion 1 Control rod- insert /vithdraw timing was perf ormed in accordance with PTI-C11-P0005, "CRD Speed and Drift Testing," on all control. rods. This evolution was performed following completion of core alterations and before the initial criticality of Cycle 4. All rods were verified to have insert  !

and withdraw times of 40 to 60 seconds. All tests were performed at atmospheric pressure and various temperatures.

h Twenty-tvo control rods vere friction tested per PTI-C11-P0003, " Control _ Rod

Friction Testing." Ten of these control rods had their. associated control 1 rod drive mechanisms replaced during the outage and the other 12 were control rod blade replacements. The absence of control rod blade interference for l l the remaining 155 control rods was confirmed by an evaluation of the results-of the control rod insert /withdrav timing and the scram timing tests.

I control rod scram timing was performed for all control rods in accordance with SV1-C11-T1006, " Control Rod Maximum Scram-Insertion Time." Control rod.  !

i maximum scran . insertion times vere determined in accordance with Technical Specification 4.1.3.2.b for those rods those associated control rod drive mechanisms were replaced, those control .is with blades replaced, and those  !

- rods whose hydraulic co-trol units had-maintenance.during the outage. This 1 testing was performed at rated temperature-and pressure prior to entering l Operational Condition 1. Scram timing was performed on the remaining rods in I the core per Technical Specification 4.1.3.2.a prior to exceeding 40% of

rated thermal power. _ - Buf fer times (the sloving dovn time at position (X) .

during a scram) were not measured because Perry has not experienced a control rod drive with an unacceptable buffer time in previous tests. j All rods vere ' fast

• per Technical Specification 3.1.3.2 on their first attempt, with the exception of control rod 54-27 vhich_vas determined to be i

'slov' operable'in accordance with Technical Specification 3.1.3.2.a. A vork order was generated to replace the scram solenoid pilot valve for rod $4-27.

The control rod was scram timed following replacement _of the scram solenoid pilot valve and determined to be ' fast' prior to exceeding 40% rated thermal-power.

Ganged rod timing Vas not performed bscause all control rods satisfied the individual insert /vithdraw timing requirements. _,

No other hydraulic testing was performed on the control rod drive system as there were no changes to the system that would have affected these parameters.

The timing tests were considered physics tests in USAR 14.2.12.2.5.

.._ . . _ _ . . _ . . _ . _ _ _ . _ . . . _ . . . _ . . . . . _ . - - . - _ . . . _ _ . ~ _. _ _ ._ .._ _ - ~._

4 A t tachmen t -~ 1 PY-CEI/NRR-1554 L Page 8 of 20 14.2.12.2.6- Test Numb r 6 - SRH Performance rand Control Rod Sequence Test Objective The purpose of this test is to demonstrate _that the neutron sources, Source Range Monitor (SRN) instrtu4tentation and rod withdrawal sequences provide l

adequate information to achieve criticality and increase power in a safe and efficient manner. j l

Discussion s SRM !! vas inoperable for reactor startup. This SRM vas identifled as a I potential Technical Specification impact on operation and was identified for corrective maintenance. The other three SRMs indicated greater than 0.7 cps l vith a signal to noise ratio greater than two to'one for approach to 1

criticality in accordance with 101-1, " Cold Startup."

This test van considered a physics test in USAR 14.2.12.2.6.

n n

i W

i i

-, - , , ,w, y gf,,..y- . . - , . _ _ , _ 3- - , ,  ;',- -

r # - <-- v- . . .

- . - . . .- _ -- . - . .-- ~.. . - . . . . .- . . - ...-. -..-- . .. . - - - - _ .-..- ... - -, - -

. .l Attachment 1 =l PY-CEI/NRR-1554 L -

- Page - 9 of 20 14.2.12.2.7 Test Number 8 - Rod Sequence Exchange l

Test Objective {

The purpose of this test is to perform a representative sequence exchange of control rod patterns at a significant power level.

Discussion Because cycle 4 utilizes a control Cell core fuel 'oadiiig pattern, the i reactor operates in the same sequence for the entire cycle. Therefore, no_ control rod sequence exchange was performed.

4 l

4 4

e s

9

?

4 1

A

' 1 em, . n~ , - - e -niE, a w -

- , - * - -- -c- er ei -- - -

• g e -we s+ wr er

_ . _ . . - . . ~ _ . .-.- _ .,.. _ ._- ..

. - . . . _ . _ - - . . . . _ _ - - _ . - ..=._....,._.,.--.

Attachment 1 PY-CEI/NRR-1554 L Page: .10.of 20-14.2.12.2.8 Test Nuraber 10 - Intermediate Range Monitor Perfor, .nce Test Objective

, The purpose of this test is to' adjust the Intermediate Range Monitor (IRM) system to obtain an optimum overlap vith the Source Range Monitor (SRM) and Average Power Range Monitor (APRM) systems.

Discussion All eight IRMs were confirmed to have a half decade overlap with both the ,

SRMs and the APRMs during performance of 101-1, " Cold Startup."

This test was considered a physics test in USAR 14.2.12.2.8.

i l

I I

i h-g l

t i

I~'

I.--

l r.

- ., ,, ,, . - . - . . . . _ - , . , - , . _ . . _ . -,s. ... .. ,, _.., ,_.

Attachment 1 PY-CEI/NRR-155'. L Page: 11 of 20 14.2.12.2.9 Test Number 11 - LPRM Calibration Test Obje,tive c

The purpost. of this test is to calibrate the Local Power Range Monitoring (LPRM) system and to verify the LPRM fl"x response.

Discussion LPRM flux response was verified with the reactor power between 8% and 35% of rated power in accordance with PTI-C51-P0001, " Verification of Proper LPRM Crnnection." The testing procedure involves moving an adjacent control rod past each LPRM and cbserving the appropriate change (significant in magnitude and correct in direction) in the LPRM reading. Two of the LPRMs did not _

respond; votk requests were initiated. The remaining 162 LPRMs responded satisfactorily.

Near 38% reactor pover, Process Computer Program 0D1, "Whole-Core LPRM Calibration and Base Distribution," vas performed. Based on this program, the LPRM teadings are luternally adjusted by the Process Computer to be used in the core power distributica and theaval limits calculation. At this time, the gain adjustment fattors (GAFs) vere not manually adjusted. Although the GAF array ranged between 0.811 and 1.287 (outside the desired 10% band), the LPRMs outside the desired range were not considered it operable because the Process Computer and Average Power Range Monitor (APRM) calibration correct for any varintion. In addition, the GAFs vere not adjusted because of the non-steady state power conditions and non-equilibrium xenon conditions.

After full power was reached, SVI-C51-T5351, "LPRM Calibration," was performed to manually adjust the LPRM GAfs. The 12 new LPRMs installed during RF03 were set with an initial calibration current. of 700 microamps. 1 Eight of the 12 LPRMs were outside the SVI's acceptance criteria and required ,

calibration per the SVI. ,

This test was considered a physics test in USAR 14.2.12.2.9.

_ _ _ _ _ _ _ _ _ _ _ _ _ _.__ . __.__._____________________.___.______.___._________________.__________________._____________________.___.]

Attachment 1 PY-CEI/NRR-1554 L Page: 12 of 20 14.2.12.2.10 Test Number 12 - APRM Calibration Test Objective The purpose of this test is to calibrate the Average Power Range Monitor (APRM) system.

Discussion During startup following retueling, the APRMs were calibrated several times in accordance with SVI-C51-T0024, "APRM Gain and Channel Calibration" as the reactor was brought to full power. The. APRMs were calibrated to read within 2% of actual core thermal power on the following dates at the identified power levels. _

% Rated

_ Date CTP i

14-June-92 27 16-June-9? 38 17-June-92 38, 59  ;,

21-June-92 58 22-June-92 93, 94, 100 Technical Specification and fuel varranty limits on APRM scram and rod bloch vere not exceeded. The startup APRM scram functions vere checked in the SVI-C51-T0030 series, "APRM Channel Calibration."

This test was considered a physics test in 14.2.12.2.10.

n l

F l

I i

l

-Attachment 1  :

PY-CEl/NRR-1554 L '

Page: 13 of 20 14.2.12.2.11 Test Number 13 - NSS Proce a Computer Test objective The purpose of this test is to verify the performance of the NSS Process Computer and on-line NSS computer programs under plant operating conditions.

Discussion The Traversing In-Core Probe (TIP) tubing undervessel vas removed / Installed during the refuel outage per_Vork Order 91-3460. The TIP alignment vas checked, after installation of the tubing, per ICI-C-C51-5, "TIP System Hechanical Drive System Calibration."

The NSS software for Cycle.4 vas updated and tested via Computer . .

ProgramfModification Request 92-09 in accordance with PAP-0506, " Computer ,

Access and Software Control." The Pericdie NSS Core Performance Log-(P1) calculations for thermal limits vere compared to those from an independent I

copy of the NSS software to establish the NSS software's acceptability for Technical Specification application. The GE program Backup Core Thermal Limits Evaluation (BUCLE) is no longer used as the backup thermal limit-calculation method for Perry, therefore, no BUCLE resulta were compared during the startup tests for Cycle 4. The LPRM calibration factors calculated by 0D1 were verified by manual calculations. ,

In preparation for the above comparisons, the' input data and intermediate .

calculations vere evaluated. At the end of Cycle 3, records were generated for the NSS database for fuel constants, exposura andcisotopics,.LPRM exposures, and control rod exposures. While the plant;vas shutdown for -

reactor refueling,-r.he NSS database was updated to reflect the new core design-and a statistically significant number of entries in. selected arrays were checked for reasonableness. During the_ reactor startup from 0% to_25%

of rated thermal power, the process computer-values for control rod _

-positions, LPRM_ readings, and thermal: hydraulic parameters were compare'd against reading = from appropriate plant sensors. While slightly less than-25% of rated thermal power, the process-computer calculation of reactor' thermal power was compared against a manual calculation per FTI-305, " Core Heat Balance," and verified to agree to within 1% of rated thermal power.

Actual agreement was vithin 0.5% of rated thermal pover. After validation of.

the core thermal power calculation, the Computer Outage Recovery Monito, program (0D15) was. initiated to start the Ten-Hinute-Core Enecgy Incre ent

. program (P4) which vas checked via manual calculations. Reactor poverevas increased and the Whole-Core LPRM Calibration and Base Distribution program- '_

(0D1) was performed to calibrate the LPFMs for the process computer. The'

process computer calculations for'TIP normalization factor, LPRM calibration constants, and~LPRM substitute values were verified against appropriate-manual' calculations. Prior. to actually physically adjusting the amplifier gains on the LPRMs, the power, time, flow, and run flags-ot the P1 program vere checked for reasonableness; and, the thermal limits from:P1 were-compared against'those from the independent source code listing. The results -

of the independent source code are identified as P1-PCjr in the Ecllowing-

-tables. The results of that comparison are as follows:

d 1

_ . . . _ _ _ . _ _ . - _ _ _ _ . - m--.-_._- _, ._ , _ _ _ . _.

4

)

l Attachment 1 v .

PY-CEI/NRR-1554 L Page: 14 of 20 SOURCE P1- P1-PCj r TEST CONDITIONS: ]

Date 14-JUN-92 14-JUN-92 4 Core Thermal Power (MVt) 1319.3 1315.8 Core klov (Hlb /hr) 35.39 35.43 Core Inlet Subcooling (Btu /lb) 32.04 -31.93 Reactor Dome Pressure (psia) 982.6 982.6 NSS SOFTVARE RESULTS:

MCPR Location: 33-20 2.370 2.376 '

Difference (%) 0.252 LHGR Location: 25-30-5 (kV/ft) 5.74 5.74 Difference (%) _

0.000 MAPLHGR Location: 19-34-8 tkV/ft) 4.92 .4.89 Difference (%) 0.613 The Thermal Data in Specified Fuel Bundle program (0D6) calculation of.

l thermal. limits ves compared against-the corresponding P1 calculations'vith no differences being identified. At this point _the OD1, P1-and OD6 NSS programs were considered operational. A second confirmatory check of these programs ,

was performed at reactor conditions closer to rated core power and rated core flow. In this second check the pover, time, flov, and run. flags of-the P1 program were again checked for reasonableness; and, the thermal limits from o P1 uere again compared against those from the independent source code. The results of this comparison are as follows:

SOURCE P1 PI-PCjr .

TEST CONDITIONS:

Date 24-JUN-92 24-JUN-92 Core Thermal _Pover (MVt) 3577.5_ 3574.4 Core Flov (Mlb/ht) 90.51 90.39 Core Inlet Subcooling.(Btu /lb)

. 23.85_ 23,88 Reactnr Dome' Pressure (psia) 1039.6 1039.6 ,

i -

NSS SOFTWARE RESULTSt.

. MCPR Location: 33-42 1.366 1.366  ;

Difference (%) ~0.000 LUGR Location: 35-48-12 (kV/ft) 12.61 12.60.

Difference (%) 0.079 NAPLHCR Location: 33-48-12 (kV/ft) 10.64 10.63J Difference (%). 0.094__

The generator and core energy accumulation features of the Pi program vere checked. After:P1 vas considered operational, the daily generator and: core 1 energy. accumulation and -the 'Lf401 exposure accumulation features of theLP2 program were checked. -After the. Daily _ Core Performance Summary program (P2) was considered operational, the total generator and core energy accumulation and the_conttol rod exposure accumulation features of the Monthly Core Performance Summary program (P3) vere checked. Finally, the:LPRM drift

- detection and reset feature of the Drif ting LPRM Diagnostic program (PS) was checked. At this point, all programs were considered operational.

E yi h 4 -t, v n ,4 e w -4m,' + 4,,,L,.,,, r- , , . _, ,:.n.- e, ,..c,%.- -r..,-a,.m, +,,,,,--o  %,,, w w ,r,1.,w, y, iy .' -c, -,.=n-e. -

sl

-Attachment 1 l

. :PY-CEI/NRR-1554 1.

Page: 15 of 20 14.2.12.2.15 Test Number la :- Core Power Distribution i

Test Objective I The purpose of this test is to determine the reproducibility of the Traversing In-Core Probe (TIP) system readings.

1 Discussion j The licensing basis TIP uncertainty for -the.HCPR -Safety Limi t :is '8.7% - 1 (standard deviation), as identified in Technical Specification 3.3'.7.7.b.

This value is based on-a model uncertainty of 4.6%, LPRM uncertainty,of 3.4%,

and a TIE measurement uncertainty of 6.6%. The TIP measurement uncertainty is based on 6.4% geometry uncertainty and 1.2% random noise uncertainty.

FTI-A16,-" Total TIP Uncertainty," vas performed in an octant symmetric control rod pattern at approximately 100% of rated thermal. power'to determine the TIP measurement uncertainty. The TIP measurement uncertainty was determined to be 1.513%. A comparison to the 6.6% value for.TIP measurement uncertainty assumed in the licensing basis shoved the TIP reading reproducibility to be neceptable.

This test was considered a physics test in USAR 14 2.12.2.15.

4 A

l.

I l

l l-l i

f I.

I i

==

I

-Attachment 1 _

PY-CEI/NRR-1554 L Page: 16 of 20 14.2.12.2.16 Test Number 19 - Core Pertormance l 1

-Test Oblective The purposes:of this test are to evaluate the core thermal pcVer and core

flow and to evaluate the following core performance parameters:

1. Maximum Linear Heat Generation Rate (MLHGR).

2. Minimum Critical Power Ratio (MCPR).

3. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).

Discussion The fuel thermal limits were monitored on a shiftly basis under the Technical-Specification Rounds Instruction. The core thermal power.vas continuously monitored and controlled under 101-3, " Power Changes." During post refueling testing, there were no fuel thermal limit violations and no steady-state operation above the more limiting of rated thermal power or the bounding licensed load line. The following lista the average thermal power (CMVTA) and the most limiting values for each of the thermal limits for every day in June that reactor power was above 25% of rated thermal power.

CMVTA Date (MVt) MFLCPR MFLPD MAPRAT 14-JUN-92 1133 0.891 0.436 0.637-15-JUN 1130 0.792 0.394 0.571 16-JUN-92 1262 0.816 0.521 0.684

j. 17-JUN 1782 0.810 0.650 0.791 18 -JUN-92 1533 0.781 0.565 0.704-19-JUN-92 1378 0.813 0.391 0.555 20-JUN-92 1785 0.811 0.585 0.713 21-JUN-92 2064 0.761 -0.570 0.705 JUN-92 3346 0.875 0.914 0.929 23-JUN-92 3572 0.895 0.951 0.961 -

24-JUN-92 3576 0.888 0.916 0.925 25-JUN-92 ^3576 0.922 0.874 0.883 JUN 92 3576 0.914 0.867- 0.875 27-JUN-92 3576 0.907 0.869 0.877-L 28-JUN-92 3562 0.919 0.873- 0.881

( 29-JUN 3576- 0.906 0.871 0.879 L 30-JUN.92 3577 -0.924 0.870 0.879 P

This test was considered a physics test in USAR 14.2.12.2.15.

L' h

Attachment 1 PY-CE1/NRR-1554 L Page 17 of 20

14. 2 .12. 2 .1 FI Ten t thimber 21 - Cov e Power Vold Hndo Rospnnse Te3t Obj ee t ive The purpose of this test is to measure the stability of the core power-void dynamic response and to demonsttae that its behavior is within specified limitr.

Discussion This test was not perf ornied during the return to power since the lov pressure drop spacers used in the nev GE10 fuel increases reactor stability. This stability is bounded by the older fuel designs in the reactor.

This test was considered a physics test in IJSAR 14.2.12.2.18.

~

Attachment 1

. PY-CEI/NRR-1554-L Page 18 of 20 14.2.12.31 Test Number 35 - Recirculation System Flow Calibration

Test Objective

! The purpose of this test is to perform complete calibration of the installed recirculation system flow instrumentation.

Discussion The jet pump flov instrumentation was checked at steady state conditions with

. core flov indicative of an operating cor6 flov for the cycle. As part of the the NSS software verification via Computer Program / Modification Request 92-09

~

under PAP-0506, " Computer Access and Software Control " the measured core flov vas verified against an established core flov/ drive flow correlation.

The comparison indicated good agreement and no adjustment of the instrumentation was performed.

At approximately 83% of rated core flow, FTI-A13. " Core Flow Calibration,"

vas performed and the instrumentation was calibrated becauce the measured core flow did not shnv good agreement with the previously established l correlation as required bf FTI-A13, " Core Flow Calibration." The gain adjustment factors for the summers vere determined from-the ERIS compbter program. Following the adjut trent of the gain on the B jet pump loop summer, the gain adjustaent f actora vere recalculated using EBIS and vere determined to be within 1%.-

Date 09 Jul 92 15 Jul 92 Jet Pump Loop Flow At Recorder B33-R612A 44 ~44 Summer D33-K611A Measured 44.00 43.58 Calculated 43.49 43.62 Gain Adjustment Factor 0.993 0.997 Jet Pump Loop Flov B:

Recorder B33-R612B 44 42 Summer B33-K611B Measured 43.59 41.73 Calculated 42.59 41.80 Gain Adjustment Factor 0.984 1.001 Total-Core Flows Recorder. B33-R613 87- 85.5 Summer B33-K613 Heasured 87.16 84.88-

, Calculated 86.08 85.42 Gain Adjustment Factor 0.993- 1.003 Except for the Gain Adjustment Factor, the values above ar i Mlb/hr.

The electronics for the APRM flow-blas instrumentation Slibrated-in SVI-C51-T0030, "APRM Channel-Calibration for 1C51-K605," - 1es and the flov transmitters vere calibrated in the SVI-C51-T0029, "AFRH Flov Reference Transmitter ID33-N014 and11B33-N024 Calibration," series. The flow blased APRM system was verified to be within the estab.lished core flow / drive flow correlation in SVI-C51-T0026, "ADRM Flov Biased Pove -Flov Verification."

[

Attachment 1

, PY-CEI/NRR-1554 L u

Paget 19 of 20 Table 3.2-1 l The folloving tests from USAT 14.2.12.2 vere evaluated by the Responsible System i Engineer and determined not 17 have been impacted by any refueling activities.

14.2.12.2.1 Test Number 1 - Chemical and Radiochemical 14.2.12.2.2 Test Number 2 - Radiation Monitoring l 14.2.12.2.12 Test Number 14 - RCIC System 14.2.12.2.13 Test Number 16A - Selected Process Temperatures j 14.2.12.2.13.1 Test Number 16B - Vater Leve] Reference Leg Temp, 4- re i 14.2.12.2.14 Test Number 17 - System Expansion i 14.2.12.2.17 Test Number 20 - Steam Production Startup Test 14.2.12.2.19 Test Number 22 - Pressure Regulator 14.2.12.2.20.1 Test Number 23A - Feedvater System t!

14.2.12.2.20.2 Test Number 23B - Loss af Feedvater Heating 14.2.12.2.20.3 Test Number 23C - Feedvater Pump Trip 14.2.12.2.20.4 Tc t Number 23D - Maximum Feedvater Runout Capability  :

14.2.12.2.21 Test Number 24 - Turbine Valve Surveillance ,

14.2.12.2.22.1 Test Number 25A - Main Steam Isolation Valves Function Tests 14.2.12.2.22.2 Test Nuaiber 25B - Full Reactor Isolation 14.2.12.2.22.3 Test Number 25C - Main Steamline Flov Venturi Calibration -

14.2.12.2.23 Test Number 26 - Relief Valves 14.2.12.2.24 Test Number 27 - Tuthine Trip and Generator Load Rejection 14.2.12.2.25 Test Number 28 - Shutdown From outside the Control Room  ;

14.2.12.2.26.1 Test Number 29A - Recirculation Flow Control - Valve Positien Control 14.2.12.2.26.2 Test Number 29B - Recirculation Flov Control - Flov Loop 1 14.2.12.2.27.1 Test Number 30A - One Pump Trip 14.2.12.2.27.2 Test Number 30B - RPT Trip of Two Pumps 14.2.12.2.27.3 Test Number 300 - System Performance .

14.2.12.2.27.4 Test Number 30D - Test Deleted  !

14.2.12.2.27.5 Test Number 30E - Recirculation System Cavitation 14.2.lt.2.28 Test Number 31 - Loss of Turbine-Generator & Offsite Power i 14.2.12.2.29 Test Number 33 - Dryvell Piping Vibration 14.2.12.2.30 Test Number 34 - Vibration Heasurement i 14.2.12.2.32 Test Number 36 - Isolated Reactor Stability 14.2.12.2.33 Test Number 70 - Reactor Vater Cleanup System 14.2.12.2.34 Test Number 71 - Residual Heat-Removal System 14.2.12.2.35 Test Number 74 - Offgas System ,1 14.2.12.2.36 Test Number 99 - Emergency Response Information System 14.2.12.2.37 Test Number 100 - Integrated HVAC 14.2.12.2.38 Test Number 113 - Service Vater-System 14.2.12.2.39 Test Number 114 - Emergency Closed Cooling System 14.2.12.2.40 Test Number 115 - Nuclear _ Closed Cooling System Test Numbec 116 - Turbine Building Closed Cooling System 34.2.12.2.41' 4 14.2.12.2.42 Test Number 117 - Emergency ServD.e Vater 14.2.12.2.43 Test Number 118 - Circulating Vater System 14.2.12.2.44 Test Number 119 - Suppression Pool Cleanup System 14.2.12.2.45 Test Number 120 - Feedvater System ,

14.2.12.2.46 Test Number 121 - Extraction Steam System 14.2.12.2.47 Test Number 122 - BOP Piping Expansion and Vibration 14.2.12.2.48 Test Number 123 - Concrete Temperature Survey 14.2.7~.2.49 Test Number 124 - Main and Reheat Steam System 14.2.12.2.50 Test Number 125 - Condensate System i

. - , - - , _ .y . . . . ,, , , . , , _ , , . . _ . , _ ~ _ _ , . - , , , , , . . _ _ . . . , _ . _ , . . . . _ , , _ . . . , . . , . . . , , , . . , . . _ - . ~ . _ . _ _ .

1 Attachment 1 PY-CCI/tJRR-1554 L Page 20 of 20 14.2.12.2.51 Test 14 umber 126 - Main, Reheat Extraction and Hisc. Drains 14.2.12.2.52 Test 11 umber 127 - LP/ IIP lleater Drains and Vents 14.2.12.2.53 Test 11 umber 128 - Condensate Demineralizer System 14.2.12.2.54 Test tiumber 129 - Steam Seal System 14.2.12.2.55 Test 14 umber 130 - Condenser Air Removal System 14.2.12.2.56 Test tiumber 131 - Offgas Vault Refrigeration System 14.2.12.2.57 Test 14 umber 132 - Turbine Plant Sampling 14.2.12.2.58 Test tiumber 133 - 1.oose Parts Monitoring System )

)

\

t 4-- . - - . - * - - ..-w...,..e.-.. m.~,-,...,.m-.,.,,,,,,'-,.mw_,,,--,.dy..7,v.,. ,.wm.,,,nr w sei---e m e m e- gw-+-.--W-w+w++gnt.w - -m-+ --e* -