JAFP-95-0291, Ja FitzPatrick Cycle 12 Startup Testing Rept
| ML20085N532 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/23/1995 |
| From: | Harry Salmon POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JAFP-95-0291, JAFP-95-291, NUDOCS 9506300232 | |
| Download: ML20085N532 (7) | |
Text
.
James A.MtzPetrick Nuclear Power Plant P.O. Box 41 l
Lycoming New York 13093 315 342 3840
- > NewWrkPbwer Harry P. Salmon, Jr.
g ggg Resident Manager June 23, 1995 JAFP-95-0291 I
i United States Nuclear Regulatory Commission Document control Desk Mail Station P1-137 Washington, D.C.
20555
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 CYCLE 12 START-UP TESTING REPORT Gentlemen:
Please find enclosed the Cycle 12 Start-up Testing Report for the James A. FitzPatrick Nuclear Power Plant, which is submitted to you in accordance with the reporting requirements of Section 6.9. A.1 of the Plant Technical Specification.
We trust you will find this information satisfactory.
- However, should you desire more information, please contact Mr. David Burch at (315) 349-6311.
Very truly yours, HARRYP. SALMON,j HPS: DEB:rfh Enclosure cc:
U.S. Nuclear Regulatory Commission Region 1 NRC Resident Inspector WPO for RMS Headquarters Distribution G. Rorke D. Burch Document Control Center F. Edler /of-JAFP File's ~
RMS (JAF).
f00007 gg63 32 950623
{I P
CK Osooo333
/
James A.
FitzPatrick Cycle 12 Startup Testing Report
)
Cycle 12 operations commenced March 21, 1995.
The startup testing
)
program was completed April 27, 1995 after rated power with equilibrium conditions had been achieved and Traversing In-core J
Probe (TIP) axial alignment checks were complete.
The test program was conducted in accordance with Reactor Analyst Procedure (RAP) l 7.3.30, " Cycle Startup Reactor Physics Test Program. " Listed below is a summary of the test results.
1.
Core Loadina and Verification:
A full core offload /onload was performed during the refuel outage in which three hundred fifty six bundles were relocated, two hundred four bundles discharged and two hundred four new bundles loaded.
The cycle 12 fuel bundle inventory is:
Fuel Type Enrichment Number Installed GE8 3.36 35 Reload 8 GE8 3.39 21 Reload 8 Gell LTA 3.02 4
Reload 9 GE10 3.22 56 Reload 9 GE10 3.24 88 Reload 9 Gell 3.56 72 Reload 10 Gell 3.59 80 Reload 10 Gell 3.56 32 Reload 11 Gell 3.59 16 Reload 11 Gell 3.80 152 Reload 11 Siemens Atrium 1CA LTA 3.39 4
Reload 11 I
The final core loading pattern was verified to be correct in I
accordance with RAP-7. 2. 4,
" Reactor Core Funl Verification" using an underwater television camera and video recorder. The videotape was independently verified by Quality Assurance personnel.
2.
Control Rod Drive Mechanism Tests:
Prior to startup, Surveillance Test ST-20N,
" Control Rod Exercise / Timing / Stall Flow Test" was performed on all control rod drives to demonstrate that each control blade was coupled to its drive and that the stroke times of each drive were within specification.
Control rod scram time testing was performed on all control rod drives prior to reaching 40% of rated core thermal power.
Test results were:
Notch Tech Spec Limit Averace for all rods 46
.338 sec .326 sec 38
.923 sec .742 sec
-4 24 1.992 sec 1.488 sec 04 3.554 sec 2.606 sec The average of the scram insertion times of the three fastest operable control rods for all groups of four control rods in a two by two array were less than the maximum times allowed by the Technical Specifications.
3.
Shutdown Marain Test:
Initial criticality for Cycle 12 was achieved on March 21, 1995.
Shutdown margin was demonstrated to be 1.06% Ak using the in-sequence critical method.
This value exceeds the Technical Specification requirement of 0.38% Ak (R = 0%Ak for cycle 12).
4.
Control Rod Seauence:
The control rod withdrawal sequence was prepared and loaded into the Rod Worth Minimizer (RWM) in accordance with RAP-7.3.32,
" Reduced Notch Worth Procedure."
Prior to reactor startup, ST-20A, " Rod Worth Minimizer Functional Test" was performed to demonstrate RWM operability.
5.
SRM Performance Check:
Source Range Monitor (SRM) functional testing was performed prior to startup to demonstrate operability of the instruments. During reactor startup, an SRM/IRM (Intermediate Range Monitor) overlap check ias performed to demonstrate that each IRM was on scale before any SRM exceeded the rod block setpoint.
6.
IRM Performance Check:
IRM functional testing was performed prior to startup to demonstrate operability of the instruments.
During reactor startup, an IRM/APRM (Average Power Range Monitor). overlap check was performed to demonstrate that each APRM channel was on scale before any IRM exceeded the high IRM rod block setpoint.
7.
Reactivity Anomalv Check:
A comparison between the predicted and actual control rod density was performed at 99.7% rated core thermal power and 98.2% rated core flow.
The actual rod inventory (notches inserted adjusted for reactor thermal hydraulic conditions) was 436, compared to a predicted full power, full flow value of 318 notches.
This value is within the Technical Specification allowable 11% Ak bounds of 47 to 589 notches inserted.
8.
Core Power Distribution Measurements and LPRM Calibrations:
j
. The core power. distribution was monitored throughout the power j
' ascension using 3D-Monicore software in conjunction with TIPS and the Local. Power Range Monitors (LPRMs).. LPRM Calibrations i
were performed at 50%, 75% andi100% of rated power. ;The 50%~
and 75%: calibrations utilized the manual data entry feature of-l 3D-Monicore due to the inability of 3D-Monicore to obtain TIP.
l data in real-time on the plant computer system..This; problem was corrected by the installation of more random access memory in the computer, defragmenting the computer's hard. drive and i
changing a system timing call used in the routines which move i
TIP data through the computer system.
Core parameters were maintained within Technical Specification (Core Operating Limits Report) allowable values throughout the startup, i
9.
' Core Power Symmetry Calculations:
i Bundle power symmetry was checked at 50%, 75% and 100% rated core thermal power.
The calculated values were-Test Plateau Maximum Difference Averace Difference 50% power 11.89%
1.24%
75% power 11.70%
1.30%
100% power 11.95%
1.27%
10.
Manual Heat Balance:
A comparison of the power calculated by computer heat balance and by a correlation to turbine first stage pressure was performed at 25%, 50%, 75% and 100% rated power.
The results were:
Test Plateau 1st Stace Pressure Heat Balance 25% power 26.8%
24.8%
50% power 52.4%
52.5%
75% power 73.8%
74.6%
100% power 101.5%
99.3%
11.
i During scram time testing, when control rod insertions and withdrawals were performed, LPRM response testing was conducted on all operable detectors.
This test verified that each operable detector is connected to the correct flux amplifier.
Nine LPRM assemblies were replaced during the outage.
A TIP response test was conducted at 50% power, which i
verified that each TIP tube is connected to the correct LPRM assembly.
12.
Core Flow Evaluation:
A core flow evaluation was performed at 100% rated power per RAP-7.3.7, " Core Flow Evaluation and Indication Calibration."
,The indicated Elcw of 76.4 Mlb/hr matched the value of 76.4
.Mlb/hr calculated using the procedure.
13.
Pfetermination of Rated Drive Flow:
A' rated drive flow calculation was performed at 100% power and the results show that 33.665 M1b/hr drive flow produces the rated core flow, 77 M1b/hr.
14.
TIP System Checkout:
TIP checkouts were performed in accordance with RAP-7.3.14, l
" Traversing Incore Probe System."
The gains of the three TIP l
flux amplifiers were adjusted such that the channel responses in the reference location (LPRM assembly 28-29) were within
'l 15% of one another.
Axial alignment of the machines was also demonstrated to be satisfactory.
15.
Core Thermal Hydraulic Stability:
l Data was acquired from the APRMs and select LPRMs at 25% and 75% rated thermal power in accordance with ST-SS,
" Neutron Instrumentation Noise Monitoring."
This information will i
serve as baseline data for the operating cycle when Technical l
Specifications require performance of ST-5S.
l 16.
APRM Calibrations:
l Numerous APRM calibrations were performed throughout the startup.
1 l
l l
l L
l l
l i
i u
f:,
i 4
",j -
g I
d b.!
'i:
g-JAFNPP J
.-d,
- a-3.3 (cont'd) 4.3 (cont'd) y],. '
l C.
Scram insertson Twnes C.
Scram insertion Tmes i.
. i.
4-1, 12- ; The merage scram inoesdon amo, bened on.'Ste d>.
1.
After esda refuehng outage, all operable rods shed be 5'
3.
onergization d the scram p504 valve solenoids as time ~
scram time tested from the fully withdrawn poedson with eq zero, d all operable control rods in the reactor power the nuclear system pressure above 950 psig (with
-i; operation condition shall be no greater then:
saturation temperature). This testing shall be ccivipi.ied l
?
Wba N
m.
d scram h 9
' i]
ControlRod Average Scram testing below 10% power, the RWM shall be operable.
l
[.}
Notch Position hTime
.e l
Obsened (seconds) a!
.g.
1 46 0338 l;
e 38 0.923
-[
'l 24 1.992 i, -
9 04 3.554 i;
4 e '
!f
!i
.A
. [6 c+j e:,
l ).'
i' n
)I 1
.;-i
.4 "I
Y I
9 d.
i,
' t-.
1
+
4j j.
- l j;
d I,
- no.f. iss gs n
L m
ms m
m2 m
i-
- e vn sa L
a a
m
.m
-~rrr rr
~
e.
-r
+-r+-mm--
eu w L
m
~%e_+- *y A-
\\
JAFNPP 4.3.C (cont'd) 3.3.C tcont'd) 1 2.
At 16-week intervals,10 percent of the operable control -
The. average of the r, cram insertion tiines for tho three l
2.
festest operable control reds of all groups of four control rod drives shaR be scram timed above 950 psig.-
i
'f rods in a two-by-two arrey shot be no greater then:
Whenever such scram time measurements are made, an.
evaluation shaN be made to provide reasonable assurance
}
Control Rod Average Scram that proper control rod drive performance is bemg
}
Notch Position Insertion Tome mamtamed.
Observed (Seconds) v 46 0.361 38 0.977 24 2.112 e
04 3.764 I
i I
3.
The maximum scram insertion time for 90 percent 3.
AN control rods shall be determined operable once each operating cycle by demonstrating the scram discharge l
insertion of any operable control rod ehell not exceed volume drain and vont valves operable when the scram 7.00 sec.
test initiated by placing the mode switch in the SHUTDOWN position is performed as required by Table 4.1-1 and by verifying that the drain and vent valves:
Close in less that 30 seconds after receipt of a a.
signal for control rods to scram, and b.
Open when tia, scram signal is reset.
l Amendment No f,[,[,[ 1/5. 203 6
i 96
=
.