ML20209J004
| ML20209J004 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/04/1985 |
| From: | Churchman C Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20209H967 | List: |
| References | |
| PSE-SE-Z-024, PSE-SE-Z-24, NUDOCS 8511110322 | |
| Download: ML20209J004 (20) | |
Text
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-2-024 TITLE: DELETION OF TIP UNCERTAINTY, TEST NUMBER 16 Date: NOV 4 1985 1.0 PURPOSE The purpose of this Safety Evaluation is to document the results of the evaluation performed to ensure that the deletion of Test Number 16, TIP Uncertainty, will not adversely affect reactor safety.
2.0 SCOPE The scope of this Safety Evaluation is the adequacy of Hope Creek's power ascension test program as it concerns the testing of the traversing incore probe (TIP) system.
3.0 REFERENCES
- 1. Regulatory Guide 1.68, Revision 2, August 1978
- 2. Hope Creek Final Safety Analysis Report (FSAR),
Chapter 14
- 3. General Electric Startup Test Specification, 23A4137, Revision 0 4.0 DISCUSSION Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 5.y requires that the incore neutron flux instrumentation be calibrated as necessary and proper operation verified. The ability of the TIP system to obtain flux traces is demonstrated during power ascension testing of the process computer. Test Number 16, TIP Uncertainty, determines the uncertainty of TIP-system readings at several reactor power / flow conditions. It is proposed that Test Number 16 be deleted.
Total TIP system uncertainty is composed of a geometric component and a random noise component. The geometric PSE-SE-2-024 1 of 3 8511110322 851106 34 PDR ADOCK O
component is due to off-center placement of the TIP tube within the LPRM instrument tube, bowing of the instrument
' tube, and water gap dimensional variations.
The random noise component is due to electronic noise in the TIP circuitry and boiling noise in the reactor.
Total TIP uncertainty is obtained directly in Test Number 16 by comparing TIP traces taken at symmetric core I positions. The random noise component is measured by making repeated TIP runs at the common instrument tube
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location with each detector. The geometric component is calculated by statistically subtracting the random noise component from the total TIP uncertainty.
Measurements of TIP uncertainty during power ascension testing at previous plants have always been well below the acceptance criterion of 6.0%. TIP uncertainty data from several recent plant startups (Attachment 1) illustrate this point. This data includes results from two different types of TIP detectors. Specifically, data from Leibstadt is from a gamma TIP detector. All other data is from a thermal neutron TIP detector. The average values of geometric, random noise, and total TIP uncertainty from these plants are 1.85%, 1.02%, and 2.17%, respectively. Only one plant measured a total uncertainty of greater than 3.2% (Kuosheng, at Test
-Condition 3 which was subsequently reduced after correcting alignment errors in TIP axial positioning) and the highest total uncertainty measured at 100% power (Test Condition 6) was 2.65%.
Results from special tests of gamma TIP detectors at Edwin I. Hatch Nuclear plant (Attachment 2) indicate that use of the gamma TIP detectors reduced the core average nodal power asymmetry (which is TIP uncertainty plus actual core flux asymmetry) by about 33%. Hope Creek will'be installing gamma TIP detectors. At other plants which have installed prototype or pilot production gamma TIP systems, the reduction was between 11% and 56%.
Improvements in the minimum critical power ratio of'5%
are typical following gamma TIP installation.
5.0 CONCLUSION
Because total TIP uncertainty at plants using thermal neutron TIP detectors has always been well below the acceptance criterion of 6% and because Hope Creek will uso gamma TIP detectors which further reduce TIP uncertainty, Test Number 16, TIP Uncertainty, can be PSE-SE-Z-024 2 of 3
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, deleted. This will not adversely affect any safety system or the safe operation of the plant. An unreviewed safety question does not exist and no changes to the Technical Specifications are required.
6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS Revision to Hope Creek's FSAR and startup test procedures shall be made to reflect the deletion of Test Number 16, TIP Uncertainty, as discussed above.
8.0 ATTACHMENE
- 3. TIP Uncertainty Startup Data 7' , EPRI Report NP-540, Special TIP Detector Measurements at Edwin I. Hatch Nuclear Plant, Unit 1, Prior to End of Cycle 1 9.0 ' SIGNATURES Originator h d:L. d #
/s- 1//c//rs' Verifiee JK fr / S4' V gV ha'%e Group Head (or SSE) b be, /t4 :i/4/85 Systems Analysis Group Head b .( . J P6 11 / 5 Site Engineering Manager hd hu. d W b / 4 S' D' ate PSE-SE-2-024 3 of 3
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. ATTACHMENT 1 TABLE 1 - TIP UNCERTAINTY STARTUP DATA TIP UNCERTAINTY (%)
GEOMETRIC RANDOM TOTAL
- HANFORD-2, TC3 2.87 1.42 3.20 HANFORD-2, TC6 1.80 1.43 2.30
'LASALLE-1, TC3 1.24 1.18 1.71 LASALLE-1, TC6 2.16 1.54 2.65
..LEIBSTADT, TC3* 2.55 0.68 2.64
-LEIBSTADT, TC6* 1.57 0.83 1.'78 FUKUSHIMA-6, TC3 1.50 1.00 1.80 FUKUSHIMA-6, TC6 1.30 1.10 1.70 CHINSHAN-l', TC3 2.51 1.21 2.79 CHINSHAN-1, TC6 2.40 0.61 2.48 CHINSHAN-2, TC3 1.20 0.88 1.49 CHINSHAN-2,'TC6 0.98 0.59 1.14 CAORSO, TC2 (25% POWER) 1.40 1.14 1.81 CAORSO, TC2 (43% POWER) 1.60 0.98 1.88 CAORSO, TC3 (53% POWER) 1.97 0.95 2.19 CAORSO, TC3 (49% POWER) 1.37 1.10 1.76 CAORSO, TC6 (97% POWER) 2.29 0.73 2.40 CAORSO,-TC6 (97% POWER) 2.21 0.94 2.40 KUOSHENG-1, TC3** 4.80 0.78 4.86
~ KUOSHENG-1, TC6 2.17 0.86 2.33 SUSOUEHANNA-1, TC3 0.78 1.46 1.66 SUSOHEHANNA-1, TC6 1.08 1.08 1.53 SUSQHEHANNA-1, TC6 0.90 1.08 1.41 AVERAGE 1.85 1.02 2.17
- Leibstadt has gamma TIP detectors, all other plants have Thermal Neutron detectors.
- TIP a:iial-positioning was incorrectly aligned.
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.. 1 ATTACHMENT 2 ,
V SPECIAL TIP DETECTOR MEASUREMENTS AT EDWIN 1. HATCH NUCLEAR PLANT UNIT 1 PRIOR TO END OF CYCLE 1
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EPRI NP 540 *
(Research Project 130-3)
Final Report September 1977 Prepared by i Nuclear Energy Systems DMelon Ge9 erat Electne Company l 175 Curtner Avenue j San Jose. Cat fornia 95125 l
Principat investigator K. W. Burke i
Prepared for Electt!c Power Research Institute 3412 Hi! Mew Avenue Palo Alto, California 94304 Project Manager Robert N. WNtetet
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I FOREWORD U Power distributions In operating belling water reactors (BWM s) are calculated by the process computer using informatoon from traversing In. core probes (TIP's).
The current TIP system design consists of ionisation chImbers sensitive f(Te thermar neutron tissions. These detectors indicate power asymmetries for core locations where the actual power distribution is thought to be symmetric. The indscated asymmetries can be attributed to sensitivity of the detector response to water pap verlations and detector positroning. These Indicated asymmetries can result in conservative thermal. hydraulic limits which tend to !* duce reactor operating flexibility. ~
With the encouragement and cooperation of Geo?gla Power Company and Soumem Company Services, a cooperative research e1 ort was deve! aped by Genere!Elecesc
' Company and EPMIas an entension of MP 130. Nuclear Reactor Core Benchmark Data. To carry out this research effort. a series of mea surements was performed at the Hetch 1 Nuclear Power Plant prior to the end of Cycle t and dunng the refuelong outage which followed. The measurements consisted of: 1. tests with three deferent i types of traversing on. core probes, and 2. gemme scans to determine both fuelled
. and bundie power distributtons et the end of Cycle 1. In the measuroments w9th the vanous TIP s, expenments were conducted with born fast neutron and gamme eensi.
eve detectors to see If either of these would beless sensMive to geometric vanakonsin me water gaps between BWM fuel bundles. The gamma scens were performedin -
l order to obtain detailed bundle power measurements. These were needed to ben.
chmark calculations wrth the TIP data collected from the three types of detectors. The gamma scan results were also needed to enlarge the data base lot qualtfreeben of nuclea'anair :ss methods, b
Similar gamme scans were conducted at Quad Cities 1 Nuctant Power Stetton in January 1976. These are reportedin EPMINP 2td. Osmma Scan Measuremente et
, Quad Cities Nuclear Power Station Unit 1 Pollowing Cycle 2. As it approached the end of Cycle 2. the Quad Cities reactor was operatedin en all-rods out condition.
The Hatch t core had a substantialinventory of controlrods at the end ofits first cycle. Hence. these data willprovide a more severe test of the nuclear analyets methods.
The measurements with the speelel gamme. thermalneutron. and fa st neutron senti.
Hve TIP's are reported hetern. A second report (NP S t t) willcontain resuRs kom the tueIrod and bundle gamma scens. A third report (NP $0 t) wlR conta rn compensene of bundle powem interred from both the gamma sensitive TIP and thermet neutron
- sensitive TIP with those deduced from the gamme scan data. Fins!!y. a fourth report (NP 662) wil document core design and operating data for Hatch 1 dunne Cyose t.
This last report us Intended foi use by those who wish to modelme hat:h t operedag history for spualdication as a test of nuclear analysis methods.
Robert N. Whitesel EPMI Project Manager E .
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, ABSTRACT
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This report presents results. Conclusions, and doscussion of a special TIP (Treverning in Core Probe) testperformcd at the Hatch 1 reactor. The purpose of the test was to provice power distribution data to support resolution of the aDDerent thermal neutron TIP estmmetty problem and to provide detailed qualification data for process computer programs and BWR core anetyeis methods.
Full core power distribution data were obtained using three General Electric test detectors, i.e., thermal neutron 71p (stendard production TIp). gamma TIP, and a fast neutron TIP. Apparent asymmetries measured with the gemme TIP were a lector Of two lower than stymmetries indicated by the thermal neutron TIP.
Although the sata analysis including gamme scan evaluation is not complete. the G gamma detector appears to be a euttabic replacament for the thermal neutron detector.
Data obtained for the last neutron dotcctor were estremery norsy and of limited ucoturnosc in tho sna!ysic of thic datoctor's performance. However, the latt neutron data did Indicate both a higher asymmetry and a higher dependency en void traction then the gamme TIP.
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- 1. INTRODUCTION f Power distributions calculated by the process computo in operating bolling wate' reactors (BWR) using information from the thermal noJtron traversin2 in core p'obe (TIP) indicate asymmetnes for symmetric (about the in-core axis of symmetry) core locations The magnitudes of the asymmetries va y with each reactor and its operating history. Analyticat and experirnentat investigations reves! that most of the indicated asymmetne: are het real power f asymmetnes, but rather are incorrect readirigs caused by ther; ia! neutron signal sensitevity to variations in weter gap
- thickness and detector positioning. Because the TIP asymmetry translates directly into a power distribution uncertainty, a reduction in TIP asymmetry will result in a reduction in power d strioution uncertainty used in thermal
- simite evaluations.
A possible solution to the TIP asymmetay problem is to replace the thermal neutron TIP with an lontration chamber which is less sensitive to water gap thickness variatio9s and in-core ptocement va'istions.Two possible TIP replacements that are less affected by geometric consideration 6 are the gamma detector which responds to prompt and delayed gammas and the fast neutron detector which responds to fast neutrons it should be noted that for the purposes of this report the detector sensitive to thermal neutrons will be referred to as " thermal TlP." the
- detector sensitive to gamma radiation will be called " gamma TIP." and the detector which responds to f ast neutrons
, will be called " fast TIP."
The purpose of the Hatch 1 TIP test was to provide power distribution data to support resolution of the apparent thermal Tip asymmetry problem as weil as to provide data for quellfication of selected process computer models inrough comparison with the gamma scan measurements. Specifically, objectives of the Hatch 1 TIP test included tne following-
- 1. Determine apparent asymmetries for each of the three test TIPS and provide asymmetry data for com.
parison with gamma een results.
b 2. Establish a vold dependency difference between the thermal Tip algnal and each of the gamma and feat TIP e30 nals for the Hatch 1 core.
- 3. Investigate the effects of delayed gammas on the response of the g amma detector to local and full core changes in power.
- 4. Determine the reproducibility characteristic Of the gamma and fast detectors compared with the thermal detector.
S. Evaluate the performance of the fast TIP in a BWR environment fo a period of several days N w ve,ernee tne oore in the enteestmat roe.en between fuel twndies roterred to two as tne =s er see t
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SUMMARY
' b' The ma,or result of the Hatch 1 TIP test program is the conclusion that the gamma sensitive detector is a potentially viable replacement for the thermal rieutron detector to solve the apparent TtP asymmetry problem. Test
' dets indicato tno gamma signat has farless apparent asymmetry than olths' the therma' or the f ast neutron TIP. Also.
the gamma signal reasires less correction to nodai powers thai the therma! neatron signal. Conversely, the faat neutron TIP does not appear to be a suitable repfacement for the thermat neutron detector eue primarily to its strong vold fraction dependency and tack of reduction In apparent asymmetry (comps'od to a thermai neutron TIP)
=' Another potential probtem witn the fast neutron Tip is signal degessation with increased exposure Although these problems might be solved through e5d!1iontI det4 Hec design work or fast detector electronics. cabing. and coating l
rnster.af, there is no incentive to pursue f ast detector development at this time. Therefore. future efforts. especially Comparison of Tip data to gamma scan results, should be concentiated on confirmation of gamma detector performance.
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0F RESULTS V
- 1. The Samme Tip indicated integral asymmetries which are about 1/2 the asymmetries indicated by the
' thermal or fast TIP. Average apparent integra! 4 symmetries indicated by the gamma, thermal. and f ast
, TIP's were 2.5. 6.8, and 7.3*.. respectivsty. Average apparent nodal asymmetnes indicated Dy the tnroe test TIP's were 5.5,8.2, and 12.29. respectively.
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- 2. Fast TIP data were exceedingly nolsy,i.e., signal to-noise ratios of approximately 5 to 1. Fast T1P data j est 3 was too noisy to be considered reliable fordata analysis Most of the noise is tnought to have doen '
caused by the unshielded detector cable being pulled through a rough T!P tube. l l
- 3. ', Signal to signal retto analysis for the Hatch 1 core indicates that the thermal TIP signal can be related to the gamma TIP algnal by the following correestion:
(TIP) , = (TIPh (a + N)
+ where v represents nodal four-bundio average vold fraction Coefficients for data sets 1,2, and 3 are as follows.
Data Get a b ;
1 1.13 - 0.366 2 1.11 - 0 313 3 1.11 - 0.325 G --
- 4. The thermat TIP signal can also be related to the f ast TIP signal as a f unction of vold fraction. However, the thermet to fast relationship is much more dependent on void [raction than the thermal to gamma dependency. For example. the RMS d!fference between the therma' Tip and gamma TIP signals was about 9.6% while the RMS difference between the thermal TIP and f ast TIP signals was about 31%.
- 5. The gamma TIP signal requires less correction to indicate nodal powers than either the thermal TIP eignal or the fast TIP algnal.
. 6. The gamma TIP detector is sensitive to both photons from fission products and prompt gammaa from the fission reaction. As a consequence, the gamma TIP is not as responsive to instantaneous power
^ variations as the thermal TIP. Gamma TIP measurements made d uring and following control rod move-ment indicate that after a local power change a steady state signal is achieved within -12 rninutes.
This response time, although slower than the thermal TIP,is sufficiently rapid to avoid any additional monitoring uncertainty, ,
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- 7. ~ The gemma TIP signal does not respond to neutron flux depreestons caused by fuel rod spacers or LPRM's as well se the thermal TIP. Because these signal depressions are weed for TIP amiel alignment, a new exist alignment procedure would have to be developed before a gamma TIP system could be implemented at a BWR.
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- 8. . The gamma TIP alonel has the same degree of reproduelbility (measure of the TIP's ability to accurately duplicate an axial power distribution) as the thermal TIP signal and is more reproducible than the feat '-
TIP eignal. Reproducibility analysis indicates the standard devlations in the eneans for the gamma, thermal, and fast TIP's are 0.78,0.66, and 8.2%, respectively. The large deviation for the fast TIP la due primarily to nolee picliup by the unshleided Tip osble.
- 9. The fast TIP sensitivity decreased by about 15% during the 2. day period that the detector was left
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- 4. CORE SYMMETRY ANAt.YSit h
l'u Figures 13.14,15. and 16 present results of symmetry analyses performed on nodes 3 to 22. The bottom and top t.o node, we,e e.ciudad from anai,.is cau.e et un,easone.iy a2.eni..ied .rro,. due io ei.e, .ign.i grad.e9ts at oore bottom and top. These noces were omitted to d *fe'estiste betwoe, true asymmetif5l dse to water gas var.at:ons and detoctor positioning and st ght s%Ifts in ax!af af.gnment. Apparent Integra! TlP asymmetry for three data sets is as follows:
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Set 1 Set 2 Set 3 Average Detector (%) (%) (%) (%)
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Thermat 8.57 6 70 6 93 4.76 Gemma 2.77 2.64 2.33 2.64 Fast 9.92 4.65 *
- 7.28 Apparent noda! TIP asymmetry for three data sets is es follows:
Date d'o b&
~ Data Data Set 1 Set 2 Set 3 Avere04 Deteetor (%) (%) (%) (%)
Thermal 8.07 8.20 S.35 S.21 Gamma 4.44 5.78 4.32 8 SS Past 12.30 12.12 - 12.21 t.
Table 1 contains detailed results of coro. average symmetry analysis.
As can be seen above, the gamma sigelIndicated about 1/2 of the saymmetry lodicated by the thermelneu-q tron TIP.This is expected since the gamma detector output is less sensitive to water gap goometric verlations than the thermal neutron detector. Actual Hatch 1 core asymmetries are being determined by fuel bundle gamma geen measurement e.t. icNits will be subsequently compared to the above anstyeis.
Asymmetries indicated by the fast TIP were about the same or even larger inen asymmetries indicated by the thermal neutron TIP. The fast TIP was expected to produce lower asymmetnes since it is less effected by water gap geometric considerations than the thermal TIP.The major cause of fast TIP asymmetries probably is versation in the detector cable noise levels Reproducibility analysis shows feistively high uncertainty due to excesolve nolee. See Figure S.12 for a good example of an excessively noisy fast TIP signal. Fast neutron algnal for location $$ 21 data set 1, had en unexplained electronic shift during TIP traverse.Thus, symmetric emalysis for this location le not in.
l cluded for this partleuler string.
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- 9. TIP SIGNAL REPRODUCISILITY ANALYSIS
%* Three sets of data were taken to determine signaltrace reproducibility,These data are presen B 10 through 8-18. Tne reoreducibility test involved repeated traversing of the some in core lo sarne probe to check the cons <stency of the TIP output.The first reproducibility data set (p to B 12) was reco ded on one graph per detector. Tnus, the individual tra os were not discernib Note the high noise levet and signat spikes,on the fast TIP traces. Results of ana!ysis for the data sets (nodes 3 to 22) are as foffows:
- Reproducibility Test Average Detector Set 2 (%) Set 3 (%) (%)
Thermal 0 66 0 64 0.66 Gamma 0.96 0.56 0.76 Fast 6.95 6.40 6.18 The numbers in the above table are the standard deviation in the means of the nod input 6 traces and the average nodal value for the 6 traces.
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GENERAL ELECTRIC COMPANY TECHNICAL ANALYSIS
_ HOPE CREEK GENERATING STATION TEST NUMBER 16 - TIP UNCERTAINTY JUSTIFY TEST DELETION OBJECTIVE:' ,
Regulatory Guide 1.68 (Revision 2; August 1978), Appendix A,
- paragraph 5'.~y requires that the incore neutron flux instrumentation be calibrated as necessary and proper operation verified. The Traversing Incore Probe-_(TIP) system is one of
.several incore neutron / gamma flux instrumentation. systems. It
.provides gross core power distribution information for several applications. TIP system operability is demonstrated during preoperational_ testing and.during power ascension testing of the process computer. Test Number 16, TIP Uncertainty, determines the uncertaihty of the TIP system readings'. It is prop 6hed to delete Test Number 16.
DISCUSSION:
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TIP system operability is demonstrated during preoperational-testing of the TIP hardware and electronics and during_ power ascension testing when the process computer undergoes the
-dynamic system-test case. During the latter testing, the process computer program OD-1 is used in conjunction"with the TIP system to_ provide information on the gross core power distribution. Test Number 16 is a separate test performed later in the _ power ascension test program. , It provides a-measure of the uncertainty in TIP system data.
Uncertainty in TIP indication effects the accuracy of LPRM calibrations, thermal limits calculations, operating recommendations, etc. For Test Number 16, the acceptance criterion. states that total TIP uncertainty shall be less than 6.0%.
- Total TIP uncertainty is comprised of geometric and' random noise' components. Geometric uncertainty results from the off-center placement of the TIP tube within the LPRM instrument tube, bowing of the instrument tube, and water gap dimensional variations. These geometric differences cause the thermal neutron TIP detectors to indicate flux levels different from the values ideally obtained by an axial scan down the center of the water gap. A measure of'this uncertainty is obtained by comparing data from symmetric TIP locations and correcting for random noise uncertainty.
Random noise uncertainty is caused by neutron, electronic and
. boiling noise in the reactor._ This uncertainty is determined by comparing data from repeticive scans in the common instrument tube by each TIP detector.
1 r_ _ _ _ _ . _
Measurement of these uncertainties at the beginning-of-life of an initial core, during power ascension testing, provides the best measure of TIP uncertainty caused by these effects because the fuel bundle power asymmetry is at a minimum. Results from previous plant startups show that measured total TIP asymmetry has always been well below the acceptance criterion, 6%.
Detailed analysis of 45 TIP sets from eight plants for power levels ranging from 18% to 100% and core flow from 33% to 105% l showed that the average total TIP uncertainty was 3.8%.
Results from more recent power ascension testing of 8 plants, summarized in Table 1, show that the average values of the geometric uncertainty, random noise uncertainty and total TIP uncertainty were 1.85%, 1.02%, and 2.17% respectively.
Geometric uncertainty has been reduced at plants which have replaced the thermal neutron TIP detector with a gamma flux sensitive detector (the type which will be used at Hope Creek Generating Station). Gamma TIP detectors sense the gamma flux in the water gap between fuel bundles and because of the relative insensitivity of gamma flux to water content between the fuel bundles and detector, the variation of TIP signals resulting from TIP tube orientation and water gap geometry differences is minimal.
Relative performance measurements at one plant near the end of cycle 1 show that the gamma TIP detector reduced the core average total power asymmetry (total TIP uncertainty plus bundle power asymmetry) by approximatley 33% (Reference 1). At other plants which have installed prototype or pilot production gamma TIP systems, the reduction was 11% to 56%. Five percent improvements in the minimum critical power ratio (MCPR) are typical following gamma TIP installation (Reference 2). It is expected that Hope Creek Generating Station will have similar results.
l CONCLUSION:
I L Based on the test results from previous plant startups, TIP J uncertainty for the Hope Creek Generating Station is expected
! to be much less than the limiting value of 6%. In addition, the geometric component of TIP uncertainty will be reduced by the use of gamma sensitive TIP detectors. TIP system operability will be demonstrated during preoperational testing of the TIP hardware and electronics and during power ascension testing of the process computer. In view of these considerations, it is concluded that deletion of Test Number 16, TIP Uncertainty, does not adversely affect any safety related systems or the safe operation of the plant and as such does not involve an unreviewed safety question.
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REFERENCES:
- 1. - K'. . W. Burke, "Special TIP Detector Measurements at Edwin -
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,,k I._ Hatch Nuclear Plant Unit.l' Prior to End of Cycle 1," -
Electric Power.Research Institute (EPRI), September 1977 (EPRI NP-540).
- 2.. Station Nuclea'r Engineering Manual," General Electric
. Company, September-1983 (NEDO-24810B).
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ATTACHMENT l TABLE'l - TIP UNCERTAINTY STARTUP DATA TIP UNCERTAINTY (%)
GEOMETRIC RANDOM' TOTAL. .
- HANFORD-2, TC3_
2.87 1.42 3.20 HANFORD-2, TC6 1.80 1.43 2.30 LASALLE-1, TC3- 1.24 1.18 1.71 LASALLE-1,~TC6 2.16 1.54 2.65
, 'LEIBSTADT,-TC3* 2.55 0.68 2.64
. LEIBSTADT, TC6* 1.57 0.83 1.78
'FUKUSHIMA-6, TC3' 1.50 1.00 1.80 FUKUSHIMA-6, TC6 1.30. 1.10- 1.70
. CHINSHAN-1, TC3 2.51: 1.21 2.79
-CHINSHAN-1, TC6 .2.40 0.61 2.48 CHINSHAN-2, TC3 1.20 0.88 1.49 CHINSHAN-2, TC6 0.98 0.59 1.14
- CAORSO, TC2-(25% POWER) 1.40 1.14 1.81 CAORSO,1TC2 (43% POWER) 1.60 0.98 1.88 CAORSO, TC3 (53% POWER) 1.97. 0.95 2.19 CAORSO,-TC3_'(49% POWER)- 1.37 1.10 1.76
,. CAORSOk.TC6.(97% POWER) 2.29 0.73 2.40 CAORSO, TC6 (97% POWER) 2.21 0.94 2.40-KUOSHENG-1,;TC3** 4.80 0.78 4.86 KUOSHENG-1,?TC6 .
2.17 0.86 2.33
' SUSOUEHANNA-1, TC3 10.78 1.46 1.66 SUSOHEHANNA-1,: TC6 1.08 l~. 0 3 -1.53
' SUSOHEHANNA-1,.TC6 0.90 1.08 1.41
-1.85 AVERAGE. 1.02 2.17 Leibstadt has-gamma TIP detectors, all other plants have Thermal Neutron detectors.
- LTIP axial positioning:was-incorrectly aligned. [
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2 i 3 l 4 4 5 l 6 timanasarr( l l(22)l l_ l I l l l (2) Perfoon Test 5, tialmg of 4 alowest control l l l l l l l rode, to conjection with espected merame l 1 l ObeBical and medioC21emical l x ,
x lx ! x l lx , I u l
l 2 l mediatica Mesourement lx lx lx l lx l l lx I l (3) Dynamic system Test case to be cumpleted l 3 l hel Ioading lx l l l l l l l betiaca test constione 1 and 3 l l l 4 l h11 core shutdown Margia lx l l 1 l l l l l l 5 l Control and prive -lE lx lx(2) llx(2) lx(2) l g lx(2) l l (4) Af ter recirculation pump tripe (natural l 6 l SRM Performance l3 l l l l l l l l l circulation) i e i 1858 hrfomance l l la I l l l l l l l 9 I a. net calibration l lx la i la l I lx l l (5) netween e0 and so percoat themal power, l 13 l AMet Calibration l 1x lx lx la l lx lx l l and near 100 percent core flaw l is l process camputer lx lx lx(38 l lx l lx l l l 1 32 l acIc I lx lx l l l l l l l (6) nas ru amaut capability 6 ancare m ap i 13 I urci l lx l l lx l l l l mmbeck must have already be = empleted l 14 I selected process Temp I lx l l la int 4l ll lx(4) l l l ** l unter savet met ses Temp l lx i I lx l l lx l l (7) menetor power between so ans se percoat i 85 I system supension lx lx lx .I la l l la l l l"!"*"- " ;' ! l ! l M l l Z l ; (s) masctor power betumen 45 and 65 percent l 37 I core serfomance l l lx la lx la la la I i I is l steam production l l l l l l l l l l (v) me.ctor power between 75 and se percent '
l ** I core tw-void made mesponsel l l l l la la l l l l 28 l Praesiere angulator l l lx lx la lx lx lx l l (10) At nazimum power that will not cease scram l 2: I sees sre-setooles change. l lx la lx la la lx lx l l l 2: I roes sre-toes tw usatine l l l l l l 1 la(5) l l (11) h rform betwo.= test comattiene 1 and 3 l as I reediater pump Trip l l l l l l l latel l 4 l 28 l Mas FW mueout capab111ty l l l l l l lu(Il l (12) heactor poemt between 40 and 55 percent l
l 22 l Turbine velve surveillance l l ln(II in(10)l i 23 l MSIV runctional Test l 23 l Me1V rull Zaolation l l l (913llE(12)lxIII lx g g gg(13)lll l g
l (13) meactor power between 60 and e5 percent l l 1 l l gg(20)l
-l lx l l 24 l me11ef wolves l l lE(20)gn l la(20)l l l (14) setween test conditione 2 and 3 l 25 l Tubine Trip 6 Load l l l lx(15lla(16)l l l gaggygg g l l nejection l l l l l l l l l l (15) Cenerator load rejectica, within bypase
- l 26 l shiAdawn outside cac l l l lx l l l l l l valve capacity l 27 l ascirculation Flow contro! l l l la(143l l lxt tei g l l l 2e l macir m e Pump Trip l l l l la l l la l l (16) menctor power between es and to percent l 2s l arr Trip-Two rumpe l l l l lx(193l l l l at core flow 195 percent - turhLee trip l
l 2e l mectro syste performance l l l lz lx lx l 13 l l l 2e I maciro pump nuaback l l l l lx l l l l (17) toad rejection l
i 20 l macire are cavitation l l l l lx l l l l l l 38 l toes of Of fette Pwr l l la l l l l l l l (le) setween test condition 5 and 6 i:
l 31 l pipe vibratica l lx lx la lx l l lx { l l 29 l accarc Flow Calibrettom I l' l l lx l l lx l l (19) >SOE power and p5 core flow, and garformed l 32 l InsCRI l lx l l l l l l l l before Turbine Trip & Imad sejectica l D l *isa l l l lx l l 1 lx(2838 l 1 34 I orywell a steam Tmmel l lx la l la l l tz l (20) check saw est pointe during major scram l cooling l ,
I l l l l l l l l l teste l no,E cassa i 35 l Geseous madweste l l lx l la l lx l l GEast aATessG STATscas l 3e i sac 3 performance l 1 l l 13 l g ll lx l (all' hrformed dering cooldown from test l fissAt SAFETY AssAtysas atPOa7 l AO l Confirmatory In-Plant Test l l l 13 i l l l l l con & tion 6 rs&a3/7 . (22) The test number correlates to rsAs section TEST SCHEDULE AND CONDITIONS 14.2.12.3.x ediere s to the indicated test number. .
i FseumE 14 2& a - se anos u __
HCGS FSAR- 10/84 r
- d. Acceptance Criteria Level 1: l
- 1. There shall be no evidence of blocking of the displacement of any system component caused by thermal expansion of the system, t
- 2. Inspected hangers shall not be bottomed out or have the spring fully stretched.
- 3. The position of the shock suppressors shall be such as to allow adequate movement at operating _
temperature. t
- 4. The piping displacements at the.establis.hed transducer locations shall not' exceed the limits specified by the piping designer, which are based on not exceeding ASME Section III Code stress values. These specified displacements will.be used as acceptance criteria in the appropriate startup test. procedures.
s
.12.3.16 TIP Uncertainty-d
- a. Ob ive 1
The test objec is to demonst e the reproducibility of TIP s em readings.
- b. 5rerequisites The co is at steady-state power level wit eq brium xenon, so as to require no rod motio l- ange in core flow to maintain power level during l acquisition by the TIP system.
l t
14.2-171 Amendment 8 l
. Test Method
. Core power distribution data are obtained duri the power ascension test program. Axial pow distribution data are obtained at each TIP cation. At intermediate and higher pow r 1 els, several sets of TIP data are ob ined to det mine the overall TIP uncertainty
- 2. TIP data re obtained with the actor operating with a sy tric rod pattern d at steady-state conditions. he total TIP certainty for the test is calcul ed by ave ging the total TIP uncertainty dete ined om each set of TIP data. _
The TIP uncertain is ade up of random noise and !
geometric component
- 3. Core power sym try is a o calculated using the .
l TIP data. A asymmetry, determined from the analysis, 11 be accounted e in the calculat ns for MCPR.
- d. Accepta e Criteria Leve 2: l The total TIP uncertainty shall be within the spe fied limits required in the GE startup test specificatio . -
14.2.12.3.17 Core Performance i
- a. OEjective The test objective.is to evaluate the principal thermal and hydraulic parameters associated with core behavior.
- b. Prerequisites The plant is operating at a steady-state power level.
i 14.2-172 Amendment 8