ML20203F183
ML20203F183 | |
Person / Time | |
---|---|
Site: | Grand Gulf |
Issue date: | 06/16/1986 |
From: | Davant G, Megan Wright MISSISSIPPI POWER & LIGHT CO. |
To: | |
Shared Package | |
ML20203F123 | List: |
References | |
1C88-ST05, 1C88-ST5, NUDOCS 8607300120 | |
Download: ML20203F183 (185) | |
Text
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GRAND GULF NUCLEAR STATION UNIT 1 STARTUP TEST REPORT Test Number: I C88-STOS Test Condition: Heatup through TC-6 Test
Title:
BOP System Pining Thermai pypo,,an, uggaz Prngram Criteria
( ) Satisfied
( x) Level 1 Not Satisfied
( ) Level 2 Not Satisfied Report Prepared By: D% .
Date: BfI S/T b 1"!.L !;.;; L;; Ji;; -/GE @D&A Engineer 8 '
Results Reviewed By: N- MA) Date: 4 f24 ' [86 H M ar Test Group Leader n
2 ( __ #M Date: L !L?. c MP& startui Supervisor
. Date: 73 8
Results Accepted By: Date:
95RC "\ l/W V(/'
e 0
0607300120 060723 6 ADOCK 0500 PDR P
l l
B0P Piping Thermal Expansion IC88-ST05 Page 1 1.0 PURPOSE The purposes of this test were to monitor the thermal expansion of selected plant piping systems throughout the range of power operation; to verify that the piping systems were clear of l obstructions which could prevent free thermal movement; and to t ensure that spring hangers and snubbers were functioning properly. l These objectives were satisfied by the performance of detailed system walkdowns and the collection of displacement data for comparison to specific criteris.
2.0 ACCEPTANCE CRITERIA 2.1 At no time during the entire thermal cycle (i.e., cold to hot) shall there be any evidence of thermal blocking in the piping systems unless it is intended by design. If pipe blocking occurs, the test should be put on hold pending Nuclear Plant Engineering (MP&L) approval to continue testing.
2.2 ' Spring hanger movement shall remain within the hot and cold l
set points, and snubbers shall not become fully extended or
- retracted. If the spring or snubber limits were exceeded, the I
test should be put on hold pending Nuclear Plant Engineering (NP&L) approval to continue testing.
2.3 The measured thermal movements shall be within the specified expected or acceptable limits. The expected thermal boundary limits are based on 10.25 inch or 125 percent of the analytical value, whichever is greater, at 550*F Reactor Pressure Vessel reference temperature, unless otherwise specified.
If the measured thermal movements exceed the expected boundary limits, but are within the acceptable boundary limits, the testing may proceed but the data shall be evaluated. If the measured movements exceed the acceptable boundary limits, the test shall be put on hold pending Nuclear Plant Engineering (MP&L) approval to continue testing.
3.0 PLANT CONDITIONS 3.1 Non-Nuclear Heatup (NNH) Testing 3.1.1 First Partial Non-Nuclear Heatup Dates : Sept. 21 thru Sept. 25, 1982 Reactor Power : 0 MWt (0%)
Reactor Temp. Range : 100'F through 350'F Reactor flooded above Steamlines, MSIV's closed, Main Steam line drains open to provide steam line warning flow path.
1
BOP Piping Thermal Expansion '
IC88-ST05 Page 2 3.1.2 Second Heatup Dates : Oct. 13 thru Oct. 23, 1982 Reactor Power : 0 MWt (0%)
Reactor Temp. Range : 100*F thru 514*F Reactor flooded above steamlines, MSIV's closed, Main Steam line drains open to provide steam line warming flow path.
3.2 NUCLEAR HEATUP (NH) TESTING 3.2.1 Reactor Heatup to Rated Temperature Dates : Sept. 26 thru Oct. 11, 1983 Reactor Power Range : 0 thru $191 MWt (0% to $5%)
Reactor Press. Range : 0 thru 952 psig Reactor Temp Range : 100*F thru 535'F Reactor critical at less than 5% power, with turbine / generator off-line and turbine bypass valves closed.
3.2.2 SRV Actuations at 250 psig (1B21-F047D and F051D)
Date : October 3, 1983 Reactor Power : $191 MWt ( i 5%)
Reactor Pressure : 248 psig Reactor Temperature : ~390*F SRV actuations in accordance with Startup Test 1-B21-SU-26-H 3.2.3 Bypass Valve Operation Date : October 30, 1983 Reactor Power : $191 MWt ( 15%)
Reactor Pressure : 962 psig Reactor Temperature : 535'F
! Turbine Bypass Valve Positions : A-5% B-1% C-0%
3.2.4 RCIC Operation Date : November 1, 1983 Reactor Power : $191 MWt ( 15%)
Reactor Pressure : 942 psig l Reactor Temperature : 534*F RCIC Flow : 800 gym l
l l
BOP Piping Thermal Expansion 1C88-ST05 Page 3 3.3 POWER ASCENSION TESTING 3.3.1 Turbine Warmup Date : October 17, 1984 Reactor Power : C'4 MWt (16.8%)
Reactor Pressure : 951 psig Reactor Temperature : 507'F Turbine Front Casing Temp. : 298*F 3.3.2 RER Steam Condensing Operation Loop A Loop B Date : Oct. 27, 1984 May 21, 1985 Reactor Power : 732 MWt (19.1%) ~3373 MWt (88%)
Reactor Pressure : 951 psig ~1010 psig Reactor Temp. : 507*F ~ 522'F 3.3.3 25% Main Steam System Flow Date : November 15, 1984 Reactor Power : 950 MWt (24.8%)
Reactor Pressure : 956 psig Reactor Temperature : 506*F Main Steam Flow : 3.6 M1b/hr Feedwater Flow : 3.0 M1b/hr Feedwater Temperature : 173'F 3.3.4 50% Main Steam System Flow Date : January 2, 1985 Reactor Power : 2017 MWt (52.6%)
Reactor Pressure : 970 psig Reactor Temperature : 517'F Main Steam Flow : 8.2 Mlb/hr Feedwater Flow : 8.0 Mlb/hr Feedwater Temperature : 362'F 3.3.5 SRV Actuations at Rated Pressure (IB21-F047D and F051D)
Date : January 3, 1935 l Reactor Power : 2041 MWt (53.2%)
Reactor Pressure : 1000 psig Reactor Temperature :a-520'F SRV actuations in accordance with Startup Test 1-B21-SU-26-2 4
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BOP Piping Thermal Expansion IC88-ST05 Page 4 3.3.6 Main Steam Isolation Valve Leakage Control System Operation Date : January 5, 1985 Reactor Power : 0 MWt (0%)
Reactor Pressure : $29 psig Reactor Temperature :--250*F MSIVLC Systems running in accordance with Startup Test 1-000-SU-77-2 3.3.7 75% Main Steam System Flow Date- : April 11, 1985 Reactor Power : 2860 MWt (74.6%)
Reactor Pressure : 992 psig Reactor Temperature : 520*F Main Steam Flow : 11.9 Mlb/hr Feedwater Flow : 11.8 Mlb/br Feedwater Temperature : 389'F 3.3.8 100% Main Steam System Flow Date : May 13, 1985 Reactor Power : 3692 MWt (96.3%)
Reactor Pressure : 1014 psig Reactor Temperature : 525'F Main Steam Flow : 16.0 Mlb/hr Feedwater Flow : 15.6 M1b/hr Feedwater Temperature : 410*F 3.3.9 RHR Shutdown Cooling Operation Loop A Loop B Date : May 18, 1985 May 18, 1985 Reactor Power : 0 MWt (0%) O MWt (0%)
Reactor Pressure : 147 psig 114 psig Reactor Temperature : 349'F 329'F RHR loops in Shutdown Cooling in accordance with Startup Test 1-E12-SU-71-6 with Heat Exchangers partially bypassed Loop A Loop B Date : Oct. 13, 1985 Oct. 13, 1985 l
. Reactor Power : 0 MWt (0%) O MWt (0%)
j Reactor Pressure : O psig 0 psig i Reactor Temperature : 125'F 146*F RHR loops in Shutdown Cooling in accordance with TSTI 1E12-85-003-0-5 with full RHR flow to Heat Exchangers and reduced SSW cooling flow to minimize cooldown.
BOP Piping Thermal Expansion IC88-ST05 Page 5 4.0 RESULTS I
This test consisted of monitoring the thermal expansion of selected MSSS and BOP piping, throughout the Startup Test Program, from Mon-Nuclear Heatup through Full Power Testing. The results of the Non-Nuclear Heatup test phase were provided in an earlier interim
, test report, IC88-ST05, dated May 1983.
Throughout the course of testing, data was collected using the GETARS Transient Recorder data acquisition system. Remote
, instrumentation consisted of FRA-MAR high tensioned Lanyard Potentiometer displacement transducers and precision 4-wire RTD's with Validyne BA-332 and PT-174 signal conditioning amplifiers, respectively. The lanyard pots were the same instruments used in the piping vibration test program, described in Test Report IC88-ST04.
I Pipe displacement data was re:orded and trended during the first
, four plant heatups, in addition to over 200 separate data readings
- taken at rated reactor temperature during the test program. Table 4.1 summarizes the critical data collected at various plant operating conditions. For simplicity, data obtained at the intermediate reactor temperatures during the heatups are not provided here. Of the over 200 readings provided in Table 4.1, 51% experienced failures of the Expected Limits, while 24% of all data failed the Acceptable Limits. Additionally, 11% of the readings were judged to have been obtained at too low a system temperature to accurately ascertain acceptance. In addition to Table 4.1, the data is presented graphically in Appendix I with the criteria limits for ready visual comparison.
. In accordance with the requirements of the Bechtel Test Specification, 9645-M.275-0, criteria failures were referred to MP&L's Nuclear Plant Engineering (NPE) group for resolution. In several cases obstructions were concluded to exist and were actually located in the field. In other cases, engineering
! evaluations by NPE or Bechtel were performed to verify that no unacceptably high stress levels occurred. By the end of the program, all deficiencies had been satisfactorily resolved or accepted.
i
BOP Piping Thsreal Exprn: sics
, IC88-ST05 PAGE 6 TABLE 4.1 NON-NSSS PIPING THERMAL EXPANSION DATA l l SENSOR l MEASURED DISPLACEMENTS (MILS) l
) l ID l Hax l Max l 25% l 50% l 75% l 100% l Bypass l Turbine l l l NNH NH l Steam Flow Steam Flow Steam Flow Steam Flow Valve OPS Warmup j l Reactor Power (%) 0 55 l 24.8 52.6 74.6 96.3 55 16.8 l l Reactor Temperature (*F)l 1514 1535 l 506 l 517 520 525 535 507 1
lFeedwater (Containment)l l l I lB21G02601DLY l +581 l +887 l +1013 l +731 l +665 l +610 l l l i l 2601TT l 397* l 227* l 174* l 333' l 374* l 377* l l l l 2602DLX l +258 l +269 l +280 +275 +260 l +250
] l l l l l l 2602TT l 382* l 199* l 176* l 336' l 365* l 381* l l l
- l 2603DLX l +202 l -28 l -137 l -216 l -194 l -223 l l l l l 2604DLY l +416 l +1175 l +1348 l +770 l +642 l +641 l l l l l 2605DLX l -505 l +236 l +323 l +316 l +300 l +396 l l l 1 l 2606DLY l +465 l +488 l +560 l +418 l +402 l +357 l l l
- l l l l l l l l l l lFeedwater/ Condensate l l l l l l l l l lN19G00101DLZ l l l +86 l +63 l +244 l +204 l l l
, l 101TT l l l 95* l 113* l 112* l 126' l l l l 401DLX l l l +164 l +162 l +291 l +222 l l l
! l 401TT l l l 175* l 210' l 234' l 248* l l l l lN21G00101DLX l l l +80 l -136 l +58 l +24 l l l l 101T7 l l l 179* l 362* l 388* l 405* l l l i l l l l l l l l l l l
l Main Steam (Non Con- l l l l l l l l l l l tainment) l l l l l l l l l j l1C88SAIT l l 535' l 535* l 536' l 537' l 539' l l 532' l lN11G00101DLX -251 ' '
l l l l Removed froml Service ' ' '
l 1 l 101TT l l 537* l 535* l 536' l 537' l 538' l 535' l l
} l 102DLY l l -586 l -496 l -1305 l -603 l -556 l l -491 l
, l 102TT l l 536* l Removed froml Service e l l l i l l l 103DLY l l -394 l -376 l -471 l -386 l -415 l l -390 l l 103TT l l 534* l 534' l 535* l 536* l 536' 534' l l l
! l 104DLX l l -293 l -270 l -474 l -532 'l -178 -452 l l l l 105DLX l l -3548 l Removed fromlService l
- 1 I I I I l l l l l i
I
BOP Piping Tharmal Exprnaics IC88-ST05 PAGE 7 TABLE 4.1
\
NON-NSSS PIPING THERMAL EXPANSION DATA l SENSOR l MEASURED DISPLACEMENTS (MILS) l l ID Max Max l Bypass l
NNH l
NH l 25% l 50% l 75% l 100% l Turbine l l
l l l l Steam Flow l Steam Flow l Steam Flow l Steam Flow Valve OPS l Warmup l l R2ector Power (%) l 0 l $5 l 24.8 l 52.6 l 74.6 l 96.3 55 l 16.8 l lResctor Temperature (*F)l 5514 l 5535 506 l 517 520 525 535 507 IMain Steam (continued) l l l lN11G00106DLY [ l -674 l -590 l -558 l -548 l -598 l l -593 l 3 l 107DLY l l -602 l -437 l -573 l -496 l -490 l l -606 l l 108DLY l l -287 l -239 l -299 l -284 l -229 l l' -293 l l 109DLY l l -2407 l 4---Removed lfrom servicel l
l 301DLZ l l +1028 l +1056 l +859 l +1039 l +905 l +1015 l+1194 l l l 301TT l l 535* l 535' l 536* l 537* l 537' l 534' l 535' l l 302DLY l l -620 l -612 l -870 l -643 l -595 l -695 l -576 l l 401DLZ l l +1051 l +1292 l +825 l +1228 l INOP l +1041 l+1388 l l 401TT l l 538* l 539* l 540' l 541* l 541' l 538' l 538' l l 402DLX l l -2534 l -2689 l -1174 l -2621 l-2244 l -2524 l-2694 l l 403DLZ l l +155 l -53 l +130 l +331 l +22 l +188 +4 l l lTurb. Front Casing Tempi l N/A l 251' l 302* l 328* l 347' l l 298* l I l l l l l l l l
' l l SENSOR l Max l Max l RCIC l HSIVLC l l l l l l ID l NNH l NH l OPERATION l OPERATION l l l l l l Racetor Power (%) l 0 l <5 l <5 l 0 l l l l l HRenctor Temperature (*F)l .s;514 l [535 l 2 534 ~250 l l l l l I RCIC l l l l l l l l lE51G004SA8DLY l +260 l +1028 l +1003 l l l l l l l l SA9DLX l +14 l +60 l -125 l l l l l l i l SA9DLZ' l -26 l +1426 l -2071 l l l l l l l l SA10TT l 137' l 530* l 530* l l l l l l j lE51G00401DLY l l +1 l +11 l l l l l l 1
l 401TT l l 233' l 482* l l l l l l l l l l l l l l 1 I lMSIV Leakage Control l l l l l l l l l lE32G10901DLX l l -4 l l +196 l l l l l
- l 10901TT l l 66* l l 208* l l l l l l l l l l l l l l l 1 l l l l l l l l I l
) l l l l l l l l l l 1
L___ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _
BOP Piping i Tharmal Exp:niica 1C88-ST05 PAGE 8 TABLE 4.1
- NON-NSSS PIPING THERMAL EXPANSION DATA i
l SENSOR l MEASURED DISPLACEMENTS (MIIS) i l ID l Max l Max l 25% l 50% l 75% l 100% l RHR Shut- lRHR Shut-I l NNH NH l Steam Flow Steam Flow l Steam Flow Steam Flow down Cool #1 Down Coo 1#3 l Reactor Power (%) 0 <5 l 24.8 52.6 l 74.6 96.3 0 0 i Reactor Temperature (*F) 5514 .s535 l 506 517 l 520 l 525 349/329 l 125/146 ECCS Systems l l l l lE21G00201TT l 103* l 105* l l l l , l l l l 202DLL l+161 l +205 l l l l l l l lE22G00301TT l 103* l 102* l l l l l l l
- 302DLY l l+656 l +782 l l l l l l l I'
lE12G01501DLY l+770 l +931 l l l l l l l l 1501TT l 116' l 96* l l l l l l l l 1502DLX l +23 l +13 l l l l l l l ,
l 1502TT l 100* l 88* l l l l l l l
- l 1504DLZ l+149 l +151 l l l l l l l l 1505DLY l+690 l +745 l l l l l l l l 1602DLY l+669 l +659 l l l l l l l l 1602TT l 99* l 88* l l l l l l l I i l l l l l l l l lRWCU/SLC l l l l l l l l l lB33G02401DLY l-667 l -704 l l l l l l l l 2401TT l 510' l 518* l l l l l l l lG33G01101DLZ l -67 l -36 l l l l l l l l 1101TT l 516' l 536* l l l l l l l lB21G16301DLL l +27 l -4 l l l l l l l l l 16301TT l 113' l 107* l l l l l l l l l l l l l l l l l lRHR Interties l l l l l l l l l
! lE12G02101DLZ l-469 l +15 l INOP l -51 l -116 l -149 l +2101 l -112 l
' l 2101TT l 436' l 450' l 429' l 431' l 424* l 411' l 307* l 100* l lB33G023RB6DLX l+736 l +761 l l l l l +357 l -14 l l lE12G012RB7TT l 152' l 271* l l l l l 349' l 146' l l ~l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l i J
-l BOP Piping Tharmal ExpIncica IC88-ST05 PAGE 9 TABLE 4.1 NON-NSSS PIPING THERMAL EXPANSION DATA l SENSOR l MEASURED DISPLACEMENTS (MILS) l l ID lRHR Shut- l RHR Shut- l RHR Steam l RHR Steam l l l l l
. l down Coo 1#3ldown Cool #3l Cond. A l Cond. B l l l l l
! l Reactor Power (%) 0 l 0 l l l l l l l l Reactor Temperature (*F) 349/329 125/154 l l RHR/RCIC l 2
lE12G00901DLY l -2 l +85 l -21 l +6 l l l l l l l 0901TT l 74* l 138* l 72* l 77* l l l l l l 0902DLY l +18 l +136 l -49 l -4 l l l l l l 1001DLZ l -214 l -161 l l l l l l l l 1001TT l 320' l 144' l l l l l l l l 1002DLY l +462 l +127 l -54 l +277 l l l l l l 1002TT l 308' l 143' l 75* l 272* l l l l l l 1005DLX l -289 l -58 l l l l l l l l 1005DLZ l -2883 l +3 l l l l l l l l 1201DLX l +174 l +46 l 44 l -30 l l l l l
) l 1201TT l 319* l 125' l 378* l 75* l l l l l
! l 1202DLY l +299 l +106 l + 4 7 ', l +11 l l l l l
- i 1202TT l 315' l, 119* l 73* l 74* l l l l l 4
l 1203DLY l +53 l +1 l l l l l l l l 1204DLZ l -254 l -138 l l l l l l l j l 1301DLX l +122 l -26 l l l l l l l l 1301TT l 280* l 99* l l l l l l l l 1302DLZ l -17 l -52 l l l l l l l l 1303DLY l +114 l -2 l l l l l l l l 1401DLX l -180 l -21 l l l l l l l l 1401TT l 66* l 123' l l l l l l l l 1402DLY l +92 l +29 l l l l l l l l 1403DLX l +324 l +1 l l l l l l l l 1404DLZ l +68 l +95 l l l l l l l l 1901DLY l +30 l -25 l l l l l l l l 1901TT l 100* l 91* l l l l l l l l l l l l l l l l l lE51G00101DLY l -11 l -92 l l l l l l l
- 101TT l l 119' l 122* l l l l l l l l 102DLZ l l l +1 l +17 l l l l l l 102TT l l l 81* l 105* l l l
l l l l l 103DLY l l l -10 l -2 l l l l l l
BOP Piping Thermal Exp::naien IC88-ST05 PAGE 10 TABLE 4.1 NON-NSSS PIPI M THERMAL EXPANSION DATA l SENSOR l MEASURED DISPLACEMENTS (MILS) l l l ID l SRV @ l SRV @ l l l l l l l l l 250# Rated Press l l l l Raactor Power (%) l <5 53.2 l l l Reactor Temperature (*F)l ~ 390
~520 l SRV Discharge l l lB21G02301DLA l -247 l -256 l l l l l l l l 2302DLL l -62 l -185 l l l l l l l l 2303DLL l -68 l -207 l l l l l l l 4
l 2303TT l 231' l 308* l l l l l l l l 2401DLA l +257 l +703 l l l l l l l l 2402DLL l +13 l +44 l l l l l l l l 2402TT l 234* l 348* l l l l l l l l 2403DLL l +221 l +731 l l l l l l l l l l l l l l l l l l l l t l l l l l l l l l l l l l l l l l l l l l l l l l l I
I
BOP Piping Thermal Expansion IC88-ST05 Page 11 5.0 DISCUSSION 5.0.1 The remote instrumentation program had been organized to collect data with the plant in a variety of cold and hot conditions. Thermal displacements were calculated by the Startup group and compared to the criteria curves provided in Appendix I. Test failures were documented by a Test Exception, and referred to NPE's Piping Analysis Section. Their preliminary evaluations were provided back to the Startup group prior to the continuation of plant heatup or power ascension. In some cases additional piping walkdowns were required in order to locate potential obstructions. Where needed, formal analyses were performed at a later time, in conjunction with Bechtel Engineering to document the ultimate acceptability of the measured displacement. In this way a system of independent checks were enforced to ensure that proper engineering considerations were given to the documented failures.
5.0.2 To support the remote instrumentation measurement program, system walkdowns were periodically performed to identify actual and potential obstructions. During these walkdowns, interfereoces were identified by either direct observation or by finding evidence of crushed or deformed piping insulation. Additionally, areas of low clearance were identified as well. The actual or potential obstructions were dispositioned in a variety of ways. In a few instances, a plant design change was implemented, minor problem areas were either resolved by Maintenance Work Order (for trimming, cutting back, or otherwise modifying piping insulation or non-structural components) or by evalution indicating the acceptability of the obstruction. In conjunction with the visual observations, piping snubbers were checked to ensure that they were not fully extended or retracted and spring hangers were checked to ensure thay they were between their hot and cold settings. Actual spring and snubber readings were taken for numerous supports to baseline component and system operation, and to support the remote measurements.
Due to radiological considerations, such of the piping was inaccessible during high power operation, thus the walkdowns were limited to some extent. For example, drywell piping was walked down with the reactor at rated temperature, but with the power level below 5%. The Turbine Building piping, such as Main Steam and Feedwater was easily accessible up to 25% power, but at high powers only limited portions cou'Id be accessed. This condition most affected the Feedwater System where the process temperature varies directly with power level, l consequently a rated temperature Feedwater System walkdown could not be fully performed. In lieu of a rated i temperature walkdown, the sytem was checked after full
BOP Piping Thermal Expansion IC88-ST05 Page 12 power operation following a reactor shutdown. This walkdown paid particular interest in identifying any areas where there may have been evidence of prior problems such as crushed or deformed insulation or obviously damaged components.
5.0.3 Throughout the course of testing certain factors continually contributed to the numerous criteria failures experienced. Provided below is a brief discussion of these factors and their impact on the test.
a) Test criteria were found to be too conservative in relation to Code allowable stresses. The actual criteria were based on a specified tolerance band around the analytically predicted displacements.
The bands were selected very conservatively for ease of criteria development. Displacements outside the more liberal " Acceptable Limit" band did not necessarily indicate equipment problems or high stress levels.
Since over 98% of the " Acceptable Limit" failures were found to have adequate margins to the Code allowable stresses, a broader tolerance band would have been beneficial from a test performance / failure resolution point of view, thus expediting the test program.
b) The analytical displacements from which the criteria limits were derived were based on analytically assumed line temperature distributions. During testing, it became evident that the actual line temperature gradients were higher than assumed (particularly on isolated lines) so numerous failures were experienced simply because portions of the piping were substantially cooler than assumed for the given system configuration. Analyses using actual line temperatures generally found substantial margins to Code allowable stresses.
c) Inter-ties and connections between systems and major equipment provided variable boundary conditions which did not always correspond to those conditions assumed in the criteria development. An example of this situation was branch piping being independently subject to the thermal movement of the connected main line. This was particularly noticeable on RHR and RWCU system piping, both of which are heavily inter-tied to process piping such as Recirculation and Feedwater. Additionally, the temperature induced displacements of major equipment such as the RPV, turbine and heat exchangers provided variable boundary conditions which were not (and realistically could not be) thoroughly accounted for during the criteria development.
B0P Piping Thermal Expansion IC88-ST05 Page 13 d) True cold plant "zero displacement" conditions were difficult to obtain because the reactor was not always brought to an ambient condition (85*F) during shutdowns. On several occasions " ambient" data was obtained with the reactor at an elevated temperature, as high as 130*F. Consequently, the l cold-to-hot piping displacements often corresponded to varying degrees of pipe delta-T. Since the criteria assumed a relatively low ambient condition of 85'F, the criteria limits were based on higher than actual piping delta-T's, and it was often found that the piping moved less than expected. This situation, while leading to failures, was readily evaluated by NPE by simply pro-rating the criteria based upon the difference between the measured ambient temperature and the assumed 85*F ambient temperature.
e) Sensor failures contributed to an increased work load. Sensors were found to fail due to a variety of mechanisms, such as dirt accumulation, water spray, and physical damage sustained during periods of heavy traffic. Despite periodic functional l checks and calibrations, sensors failed at inopportune times, and were not readily replaceable. Hard up-scale or down-scale failures were easily identifiable (readings > 3.7 inches).
l Intermediate failures were not always immediately discernible. The sensor data was assumed to be accurate, unless proven to be faulty. Several i
instances of criteria failures occurred which could not be substantiated by either data from other
- sensors or by physical measurements of nearby spring I and snubber supports. In these instances, the sensor was functionally tested at the first accessible time. If the device was found to be faulty, it was replaced and data recollected, if possible, at the next test condition. In instances where the sensor was found to be operable, increased emphasis was placed on visual observation. Additional system walkdowns were performed to identify any obstructions, and in certain cases, snubber
! functional testing was performed to identify the i existence of potentially locked snubbers.
5.0.4 The following report subsections describe the various phases of the Thermal Expansion testing. The actual test data is provided in Table 4.1 in tabular form, and in l Appendix I in graphical form for comparison to the !
Expected and Acceptable criteria limits.~ As discussed ;
earlier, off-rated temperature piping data taken during )
plant heatup to rated temperature is not provided. The l
)
l
BOP Piping Thermal Expansion 1 I
1C88-ST05 Page 14 )
i off-rated data was used for trending purposes, and though l were compared to off-rated criteria during the program, I only the final rated temperature data are given. This l trending data proved very useful in the early detection j of some system problems. The only significant thermal j expansion related problem occurred during Non-Nuclear '
Heatup (NNH) and was attributable to not having a reliable trending curve. Refer to Section 5.1.2 for details.
5.0.5 Appendices II and III contain the conclusions of the NPE calculation results for Nuclear Heatup (calculation MC-Q1111-830027) and Power Ascension (calculation MC-Q1111-85039), respectively. These are provided as background information for this report. The Piping Analysis section of NPE independently calculated the piping expansions using the test data provided by Startup, and were responsible for the evaluations of the acceptability of the measured displacements.
5.1 Non-Nuclear Heatup (NNH) Testing 5.1.1 First Partial Heatup (NNH #1)
This test represented the second reactor heatup to rated conditions. The first heatup was performed in
, 1981 during the hot Recirculation Preoperational test. Pipe expansion was monitored at that time in conjunction with Special Test Procedure IC88-ST03, and is not discussed in this report. This second heatup was conducted in a similar fashion, with the reactor shutdown and the system heated by high speed recirculation pump operation. At the time of the test, the reactor was flooded, as were the Main Steam lines (with the MSIV's closed). The Main l Steam and RCIC Steam Supply lines were heated by opening a flow path through the steam line drains,
, and allowing hot, pressurized water to flow through I
them. This was found to be necessary during the j previous plant heatup to prevent fully contracting
- or extending variable supports. If the steamlines l were kept cold, the upward vertical reactor vessel
- growth would " pull" the lines upward, without the l naturally occurring offsetting downward thermal expansion of the steam lines.
During the test the following systems were remotely monitored at approximately 50*F increments of reactor temperature using the previously installed lanyard pots and RTD's:
BOP Piping Thermal Expansion IC88-ST05 Page 15 1
' Reactor Core Isolation Cooling (RCIC)
'Feedwater (FW) - inside containment
- Low Pressure Core Spray (LPCS)
'High Pressure Core Spray (HPCS)
' Low Pressure Core Injection (LPCI) Subsystems of RHR
' Recirculation System (RR)
- Main Steam (MS) - inside containment The Recirculation and Main Steam systems were specifically monitored in accordance with Startup Test Procedure 1-000-SU-17-H, as they are NSSS l Vendor supplied piping. The thermal expansion data l was obtained from the installed lanyard potentiometers and RTD's at the following reactor temperatures:
Reactor Temperature Date 166*F 9/21/82 189'F 9/22/82 258'F 9/22/82 293*F 9/23/82 322*F 9/23/82 346'F 9/25/82 On September 23, 1982, with the reactor at approximately 350*F, a complete drywell pipe expansion / vibration walkdown was conducted on the following systems:
'RHR 'RWCU 'LPCS *LPCI(RHR) *MS
'RCIC 'SLC 'HPCS ' Recirculation 'FW
\
l Additionally, the RWCU system was checked up to and
! including the non-regenerative and regenerative heat exchangers. The walk down identified numerous items, 6 of which required rework or increased attention. Most notable was the decreasing clearance between SRV IB21-F047A stainless steel air supply line and a structure I-beam. This clearance was to be visually checked every 15'F increase in temperature. However, the heatup was terminated prior to reaching rated reactor temperature due to an increasingly high drywell air temperature, resulting from inadequate temporary piping insulation and non-optimal performance of the drywell cooling systems. The reactor was cooled down and additional insulation was added on the hot piping systems, particularly in the vicinity of the RPV nozzles / shield penetrations. Also, additional work was performed on the drywell cooling system air flow balancing to force cooler air to the drywell hot spots.
t BOP Piping Thermal Expansion IC88-ST05 Page 16 5.1.2 Second Non-Nuclear Heatup (NNH #2)
Following the drywell work, the reactor was cace again heated up in 50*F increments. The same systems as before were monitored using the installed remote instrumentation. Data was collected at these reactor temperatures:
Reactor Temperature Date 197'F 10/13/82 244*F 10/13/82 320*F 10/13/82 i . 348'F 10/13/82 402*F 10/14/82 451*F 10/14/82 498*F 10/16/82 512*F 10/22/82 514*F 10/23/82 At each temperature hold point, several hours of
" soak" time were allotted for the piping systems to come to equilibrium before expansion data was taken
, from the remote sensors.
Following the heatup this data was forwarded to NPE for evaluation.
On October 14, 1982 with the reactor at
- approximately 350*F and again on October 20, 1982 with the reactor above 500*F complete drywell walkdowns were again performed on the following systems
'RHR *RWCU *LPCI *LPCI(RHR) *MS
- RCIC *SLC 'HPCS
- Recirculation *FW Numerous obstructions, actual and potential, were identified and removed, particularly following the rated temperature walkdown. Most of these had to do with insulation butting against equipment or structures, and many low clearance (potential) obstructions were flagged. The most critical 1
problem found was on the RWCU system, where snubber j Q1G33G002R12 was found to be fully retracted. A design change was immediately implemented to resolve this problem. In addition to walkdowns, selected spring hanger and snubber movements were recorded with the reactor above 500*F. These readings were taken by the plant's Inservice Inspection (ISI) group over a several day period. The readings were compared to the expected readings based on design data in I
i BOP Piping Thermal Expansion IC88-ST05 Page 17 order to identify gross failures. Due to the limited time available for this effort, not all the desired measurements were obtained before the reactor was cooled down. The remaining hanger data was then 1
scheduled for Nuclear Heatup.
It was during this phase of testing that the most serious criteria failures were uncovered. At the time of this test, the test criteria had been provided in the form of linear equations dependent
- on line temperatures. From these equations,
- criteria " envelopes" were constructed for trending purposes. While the test was in progress, an independent assessment of the criteria by Bechtel Engineering inoicated that a problem existed with the criteria equations, in that they were based on rated power operation (RPV at 535'F and Feedwater at 420*F), and were not applicable to the NNH lineup.
The criteria equations were re-evaluated by Bechtel and were replaced with more specific criteria curves, for NNH use specifically. These curves arrived on site after the unit had achieved rated temperature, so the test data was backfitted into the plots. During their review of the lanyard pot data, NPE determined that three feedwater points fell significantly outside the Acceptable Criteria. Sensors
- IB21G02603DLX, 2604DLY and 2605DLX were determined to be subject to an unidentified obstruction which severely limited free movement. The analysis by Bechtel indicated that the magnitude of the stresses and loads on the feedwater nozzles were significantly high, although within the ASME Code ultimate stress levels. In response to this concern, NPE ordered UT examinations of the 6 welds l between the affected feedwater piping and RPV i
safe-ends and the associated riser tee's. The UT examinations were performed after the unit had been cooled down and proved to be acceptable, but the source of the obstructions could not be located, even after numerous walkdowns following the reactor cooldown.
During conversations with Maintenance Engineering, it was later determined that the unidentified obstructions had been temporary insulation packed between the i
feedwater lines and the 6 RPV shield wall penetrations.
l It was determined that vertical RPV growth pulled the feedwater lines hard against the dense insulation, on top of the lines, thereby preventing normal movement of the lines. The temporary insulation was verified to have been removed immediately after the NNH testing and was replaced with permanent insulation covers, the design of which would preclude a repetition of the problem. To ensure that this indeed was the cause
B0P Piping Thermal Expansion IC88-ST05 Page 18 of the problem, additional temporary sensors were installed in the drywell for future monitoring, and an increased monitoring interval was to be used on the effected sensors.
5.2 NUCLEAR HEATUP (NH) 5.2.1 Heatup to Rated Temperature Under the 5% power license, the reactor was brought critical and heated up to rated temperature and pressure in September - October 1983. During this testing the turbine generator was off-line and the turbine bypass valves were generally closed.
Lanyard pot expansion data was collected at 50*F increments for the following systems, additional data was taken on the Feedwater piping inside the containment every 15*F.
I og
- RCIC
'RWCU
'FW
. *SLC
'LPCS
- EPCS
- MS(to turbine and bypass valves)
Data was taken and plotted against the Expected and Acceptable Criteria limits at the following reactor temperatures:
Reactor Temperature Date 205*F 9/26/83
, 243*F 9/26/83 1
304'F 9/27/83 341*F 9/27/83 404*F 10/2/83 449'F 10/3/83 499*F 10/8/83 535'F 10/11/83 As with the earlier heatups, NPE concurrence was obtained after their preliminary evaluations of the failed data points, prior to continuing the plant heatup. Their final acceptance of the data is documented in calculation MC-Q1111-830027.
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BOP Piping Thermal Expansion IC88-ST05 Page 19 On September 27, 1983, with the reactor at 340*F an informal spot checking was conducted of the drywell and steam tunnel piping. Twenty seven (27) items were identified, the most noteable being the identification of RWCU drain piping below the A and B Recirculation Pumps being hard against the floor, including the insulation around valve G33-F102.
Additional problems found were snubbers Q1B33G103R01 fully retracted, spring hanger E51G001H01 on the RCIC steam lines, bottomed out, and several instances of low to zero clearance between pipe insulation and whip restraints. Later, heatup walkdowns were conducted in the Turbine Building at 465* and 490 F reactor temperature. Several items were identified during each walk through.
The unit achieved rated temperature and pressure on October 11, 1983. Between this date and November 1, extensive piping inspections were conducted in the drywell, Containment, Auxiliary Building and Turbine Building. The drywell problems were particularly few thanks to the previous problem resolutions.
However, the Turbine Building steam lines were found to have numerous problems. This was not totally unexpected, as this heatup was the first time the complete Main Steam System had operated at temperature. The lanyard pot data had revealed the existence of a potentially locked Main Steam line snubber in the Turbine Building. NPE had evaluated the stress levels and found them to be acceptable.
Snubber functional testing ultimately determined that NIN11G001R01, an INC mechanical snubber, was locked up. This component was replaced with an equivalent PSA type mechanical snubber, at the first oppor:une shutdown. The rated temperature walkdown ioentified 70 items which the Startup group I
flagged as requiring rework or NPE evaluation.
Major items found and later resolved were:
i l *RCIC Steamline insulation hard against whip l restraint, along with several other related interferences.
l
,
- Main Steam spring hangers in the drywell crushing protective box for SRV solenoids.
'HPCS test connection hard against unistrut.
- Main Steam Lines in contact with grating.
During this test phase it became evident that the extensive walkdown efforts put forth during NNH testing significantly minimized the-time spent in the drywell while at power. Inspection and rework activities were reduced, thus attributing to significantly lower radiation dose rates received by plant personnel during NH testing.
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B0P Piping Thermal Expension IC88-ST05 Page 20 5.2.2 SRV Actuation - Reactor at 250 Psig In conjunction with initial SRV testing, Startup Test 1-B21-SU-26-H, the thermal expansions of two representative SRV tailpipe lines were remotely monitored during the 10 to 20 second valve pops.
Valves IB21-F047D and F051D (each tailpipe instrumented with 1 RTD and 3 lanyard pots) were opened and the expansion recorded on the high speed data acquisition system. The RTD's were found to quickly heat to approximately 230'F and the recordings of the piping movement show the sudden movement. However, due to procedural requirements, the valves were closed prior to the lines reaching equilibrium conditions. This contributed to several criteria failures, which NPE analyzed and found to i be acceptable.
5.2.3 Turbine Bypass Valve Operation Although the majority of the NH test phase was done with the bypass valves closed, some 5% power testing did require bypass capacity. Consequently, the piping between the bypass valves and the main condenser did experience brief high temperature I
operation. Lanyard potentiometer measurements taken on the common line upstream of the valves did not reveal any problems.
5.2.4 RCIC Operation In conjunction with rated reactor pressure RCIC system testing, 1-E51-SU-14-H, thermal displacement data was taken on the RCIC steam supply line to the RCIC turbine with the system in operation. Visual inspections were also carried out on the entire RCIC steamline, from drywell to turbine, combined with hanger and snubber readings for comparison to design settings. No significant anamolies were found for this mode of system operation.
5.3 POWER ASCENSION 5.3.1 Turbine Warmup The initial turbine warsups were monitored over a three day period by checking the Main Steamline displacement sensor data. The final set of data was obtained with the turbine front casing temperature at a steady value of 298*F. This data was supplied directly to NPE for evaluation prior to continuing
BOP Piping Thermal Expansion IC88-ST05 Page 21 further turbine loading and power ascension. While not required by procedure, the steam piping was visually spot checked for potential obstructions near the turbine, this was possible due to the relatively low dose rates existing with the reactor below 25%
power. No obstructions were found during this evolution.
5.3.2 RHR Steam Condensing On October 27, 1984, in conjunction with Startup Test 1-E12-SU-71-1, the RHR "A" subsystem was aligned in the Steam Condensing mode. Lanyard pot data'was collected and compared to the criteria.
The steam condensing mode represents the greatest heat exchanger heat load, and thus temperatures, for the system. Numerous failures were observed
- however, most being traced to significantly lower l line temperatures downstream of the RHR heat exchangers. Final NPE resolution and acceptance of the data is documented in calculation MC-Q1111-85039.
An optional RHR system piping walkdown could not be performed, nor could spring and snubber readings be taken due to the increasing radiation dose rates experienced near the various steam piping.
The B RHR subsystem was tested in June, 1985, while the system controls were being optimized for operation. The results of the B Loop Steam Condensing test were quite similar to the A loop, l in that the heat exchangers cooled the process flow to temperaturet significantly lower than assumed for the criteria. Consequently the line temperature distributions, as well as the actual heat exchanger
{
shell temperature gradients created numerous apparent failures and data points being to low in temperature to reasonably apply the criteria. Once again, no walkdown could be performed due to radiological accessibility.
5.3.3 25% Main Steam System Flow i On November 15, 1984 with the turbine generator
! on-line remote sensor data from the Main Steam and l Feedwater lines was collected at approximately 25%
l Main Steam Flow. The steam lines experience minor i temperature changes due to a slightly higher reactor I pressure and a hotter turbine front easing temperature, while the feedwater lines experience.
heating due to the feedwater heaters being put into service. For comparison purposes, the 25% flow i
. ~ . _ , _ - _ _ __. _ _ _
BOP Piping Thermal Expansion IC88-ST05 Page 22 condition, which approximates 25% power, resulted in a 180*F feedwater temperature, which is 70*F hotter than the temperatures experienced during Nuclear Heatup (without the heaters in service). Data l was collected with no significant problems other than '
several sensors failing the criteria. NPE was able to evaluate each failure, and provide acceptance of the data. Their final acceptance is documented in calculation MC-Q1111-85039.
On November 17, 1984 after the temperature on the feedwater lines had increased to approximately 300*F, the first high temperature walkdown of the Turbine Building Feedwater and Condensate piping was
, conducted. Except for the feedwater heaters, the bulk of the piping was accessible. High radiation in the vicinity of the various steam lines, limited access to the heaters. No significant problems, were discovered during this walkdown.
5.3.4 50% Main Steam System Flow Data was taken on January 2,1985 of the Main Steam, Feedwater and Condensate piping. The reactor was at 517'F, while the turbine front casing temperature was 302*F and the feedwater heater outlet temperature had increased to approximately 360*F.
As with the 25% data, several sensors had failed the criteria limits, but acceptance of most of these l
failures ws readily obtained from NPE due to the similarity between this data and the 25% data. The final NPE approval of all test results is given in calculation MC-Q1111-85039.
}
5.3.5 SRV Actuation - Reactor at Rated Pressure In conjunction with Startup Test 1-B21-SU-26-2, i
SRV's IB21-F047D and F051D were manually actuated at i rated pressure. This test was very similar to the 250 psig test performed during Nuclear Heatup, with l the exception of the higher resultant tailpipe 1
temperatures of 300 to 348'F. As with the earlier i test, the 10 to 20 second opening of the valves prevented true equilibrium conditions from being achieved. The criteria failures were reviewed and i accepted by NPE as satisfactory, these conclusions
- are formally documented in NPE calculation 4
MC-Q1111-85039.
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BOP Piping Thermal Expansion IC88-ST05 Page 23 5.3.6 Main Steam Isolation Valve Leakage Control System Operation The MSIV LC system was run in conjunction with Startup Test 1-000-SU-77-2 following a planned reactor shutdown.
i 3'
The reactor had been depressurized to near atmospheric pressure prior to system startup. The
- reactor temperature was below 250'F with the MSIV's 1
and the Main Steam Shutoff valves closed. The
- system was normally started up when the system i pressure interlocks cleared, and began to draw down the steamline pressures. The MSIVLC piping heated to approximately 208*F when the test data was obtained. This corresponds to near maximum system operating temperature. The data from the remote sensors was forwarded to NPE for evaluation. Their final acceptance is documented in calculation MC-Q1111-85039.
5.3.7 75% Main Steam System Flow On April 11, 1985 with the reactor at 74.6% of rated power, Main Steam, Feedwater and Condensate system j piping displacements were obtained and compared to J criteria. The following temperatures were measured:
i Reactor - 520*F Turbine Front Casing - 328'F Feedwater - 388'F The test data was consistent with the lower power data in so far as the number of failed sensors, the degree: of failure ranged from slightly better to worse.
The test data was, once again, provided to NVE.
l Their preliminary approval was obtained prict to increasing power, and their final approval !.s documented in the MC-Q1111-85039 calculation.
At this higher power condition, no visual
' inspections could be conducted, due to high area radiation doses.
5.3.8 100% Main Steam System Flow 1
The final check of the Main Steam, Feedwater and Condensate Systems thermal expansion was conducted
- on May 15, 1985. This test was run shortly after
BOP Piping Thermal Expension IC88-ST05 Page 24
, the unit reached the rated power test condition (between 95 and 100% power). With the reactor at 96.3% power and 535*F recirculation suction l temperature, the turbine front casing was measured to be 347'F, while the Feedwater temperature at the outlet of the final high pressure heaters was 405'F, 15'F short of rated. Data from the 21 remote ,
sensors on these systems were forwarded to NPE for 1
- evalution. Half of the points failed the " Expected" l displacement limits, while one third failed the l
" Allowable" limits. Additionally, sensor N11G0401DLZ was inoperable at the time of the test as proven following the next plant outage. NPE was abit to approve all test results (calculation MC-Q1111-85039) but lanyard pot N11G00403DLZ results
- indicated the existence of some obstruction. This sensor, located on the feedwater pump turbine steam i supply line had shown significantly low thermal i i movement throughout the program. Snubbers in the l vicinity of the sensor had been functionally tested
- previously, and were found to be operable. Since j this was the last opportunity to determine the source of the problem a final intense effort was put I forth to check for obstructions. Despite numerous walkdowns (with the reactor shutdown and again at low power) by Startup and NPE, no evidence of an obstruction was found.
Despite this condition, NPE analysis was able to show that the actual stresses on the piping were acceptable, and no further action was taken. The final approval is provided in NPE calculation MC-Q1111-85039.
In addition to the remote measurements, a full system walkdown on the Condensate and Feedwater was scheduled. In as much as radiologically possible, these systems were visually checked and selected spring hanger and snubber positions were recorded, with the feedwater system at 405'F. Except for the areas near the low pressure feedwater heaters, most of the Condensate System was inspected. However, only a very limited portion of the feedwater system
- piping was looked at. The piping near the process steam piping were in high dose rate areas and were rescheduled for inspection following a subsequent reactor shutdown. When the reactor was finally i shutdown following an inadvertent scram, the Feedwater and Main Steam systems were checked, with the plant in a cold condition. This inspection attempted to identify evidence of obstructions, such a
1
BOP Piping Thermal Expansion IC88-ST05 Page 25 l as crushed or deformed insulation, deformed pipe supports, etc. No problems were found during this effort, thus the walkdown was concluded on a satisfactory note.
. 5.3.9 RER Shutdown Cooling Operation i
4 On May 18, 1985 with the reactor shutdown, at 349'F, I
the RER A System was aligned in a Shutdown Cooling Mode.
! Due to the high heat capacity of the heat exchangers
! it was impossible to fully align the system with full RHR flow to the heat exchangers without
- severely violating the maximum allowable Tech Spec cooldown rate of 100*F an hour. To solve this, the heat exchanger bypass line was cracked opened thus limiting the flow to the heat exchanger and
- minimizing the cooldown rate. The result of this valving was that the heat exchanger did not achieve
, the full design operating temperature, consequently the pipe displacements were significantly different
- than expected for the given system temperature. The B RHR loop was tested in a similar manner. The test
- results were forwarded to NPE, who requested that additional data be obtained. The low system temperature created negligble thermal growth on several segments of piping, so the test results were not conclusive.
On July 29 and 30, 1985 a retest was performed with full flow RHR to the heat exchangers. This was possible because the reactor temperature was only at approximately 270*F, so the RPV cooldown rate was not as large as the earlier (higher temperature) test. Unfortunately, with the reduced inlet temperature, and high heat exchanger capacity, the discharge lines were very near ambient temperature and experienced little thermal growth. This data was also sent to NPE, and again they requested additional data at a higher outlet temperature. At this time, it was recognized that the heat exchanger capacities were so large, that normal system operation would not provide an opportunity to " heat" the discharge side of the heat exchangers. It was concluded that the only way to prevent on excessive l RPV cooldown was to decrease the Standby Service I Water (SSW) cooling flow to the tube side of the j heat exchangers. This would enable the RHR side from cooling down as much as it would under normal circumstances. -
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- BOP Piping Thermal Expansion IC88-ST05 Page 26 The third attempt was run on October 13, 1985 in 1
- conjunction with RHR System Technical Special Test .
Instruction 1-E12-85-003-0-S. This test aligned the '
system in the Shutdown Cooling Mode and then I manually throttled down on the SSW flow to the RHR heat exchangers. Then, full RHR flow was valved to the heat exchangers by closing the heat exchanger bypass valve. During the test, these extra steps took longer than expected and by the time the system was ready for data collection, the reactor temperature had decreased to 146*F for the A loop l test, and 125'F for the B loop test. Once again, the system temperatures were barely high enough to
- properly check criteria. Ensuing discussions between NPE and Startup concluded that the higher discharge piping temperature operating state desired for the expansion monitoring were not practically achievable in the Shutdown Cooling mode, and high temperature operation would not be achievable during
) normal system operation. In response to this j conclusion, NPE indicated that the data collected to this time showed that the system displacements were 4
acceptable, including those portions the piping i
which never achieved design temperature. In lieu of direct measurements, NPE personnel conducted an independent RHR system walkdown in an attempt to identify possible obstructions. Since only very 3 minor interferences were observed, none of which
! would impact the ability of the piping to freely expand, NPE accepted the RHR Shutdown Cooling data as having satisfied the test requirements. This subtest completed the BOP Piping Thermal Expansion Monitoring Program conducted as part of Special Test IC88-ST05.
- 6.0 PERTINENT DATA f 1. Appendix I - Displacement vs. Temperature Plots
- 2. Appendix II - NPE Calculation MC-Q1111-830027, Piping Thermal Expansion Evaluation During Nuclear Heatup
- 3. Appendix III - NPE Calculation MC-Q1111-85039, Piping Thermal Expansion Evaluation During Power Ascension Testing I
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PATA SMIT U6 Spee. 9645.M.275.0 r- PRE-OP. THRMAL MOpfTORING APPendia O Rev. 10 R E AT- m
, sheet 1 ef 6 0 Pre-ep Iso. N..__M-1ssa Rrv.s -
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. 1 I_
AP yendia 5 Rev. 10 1
PRE-OP. TIERNAL NoltITOR11tG I 0'** m-'n ss..t 4 9 .e & o Pr..o,i... s.. M-issu. Rtv ., ....... .. . .. - -
Lanyers No. N 11Goofo 3 .DW -i-s Stress Prob.No. 1,4 3.P. Bo I:"" =- "' 4 -
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Ise. No. 9 1654. Rev. ?
Pre -v. sE:f - -
No.M (GOOlO3DLY 3,anyard A =- h --4._
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spec. Mel-N-175.0 DATA SIBIT U.6 t-PRE-OF. MML MMMORING APP **di8 9 Rev. 10 pcuan MAT-m sheet 5 2er 6 o
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tre-*P Zee. No. M1G54 REV.2 _
Lamyerd No. N 11 G 00104_TRE ~== l = = -
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DATA SEET U.7 Spe c. 9643.M.275.0 f- PRI.op. TMrmy.AL McN1Tontwc Aprendin y gay,11 Power Ascension / System Heatups Sheet of .N Pre-or Iso. No. M.1654 Rev.2 Lacyste No. N I i G Lh.J I D 4 D LX .=. _ =s =='-__ l g.P. l 2 6 1 - ~' ~M =:r:rsa'- 200 Stress Prob.No. p- __
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DATA STEXT U.6 spec. 9645-M-275.0 3- PRE-0P. TERMAL IGITTORDIG APPendia U Rev. 10 occt'" "AT-un ss..t52 f Go Pre-op Iso. No. M-1ssia RN 7 . . m _. %
lanyard No. NilGooloM I)LY :
Sttoeo Prob.po. 12, 3.p. 'RRy =t= - -
- x - -- --
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i_ tut r2s _wt.i
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=
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=
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_ _ . _ _ . _ .f_.
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=--
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=
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r--
ER J
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e age =
s-
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4__._. .m
^
___ . __. _.__ .__. - Je_. _ _ 2400
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Expoeted Limits __
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Acceptable Lietts -------- __
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-600
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-200
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R.P.V. Temp 'T m r
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Stress Prob.5e. i_""*.= .
32 3,y.g 2 -
P_._ . .
-1200
- . rue.s_um...u_s. -
I._n._u_.=: _ u: nu urtes nur_must - _
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-800
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_ _ .. __. _ _ _.._ _ ..._ _ __ __.s...._ ._. _. .__ ._ .. ... _ . ._ ... .. __. _ . .. _. ._
-200
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Skeet h et 6 0 Pre-ep Ise. No. M-1654 D 9 - - _ _ _ . _ _ _ _ .
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