ML20039B660
| ML20039B660 | |
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| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 10/09/1981 |
| From: | Norris E SOUTHWEST RESEARCH INSTITUTE |
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Text
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SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road l
l San Antonio, Texas 78284 ANALYSIS OFTHE VESSEL MATERIAL SURVEILLANCE CAPSULES WITHDRAWN FROM LACROSSE BOILING WATER REACTOR DURING THE1980 REFUELLING by l
E. B. Norris l
l FIN AL REPORT l ~p SwRI Project No. 02-6208-001 for Dairyland Power Cooperative Lacrosse Wisconsin 54601 October 9,1981 Approved:
U. S. Lindholm, Director Department of Materials Sciences Dohk o!00h 9
s P.
.PDR
ABSTRACT The third set of reactor vessel surveillance capsules was removed from the Lacrosse Boiling Water Reactor during the 1980 refuelling out-The neutron flux results and the neutron embrittlement responses i
age.
l of the surveillance materials, although in good agreement with data from previous analyses, did indicate that the rate of embrittlement is less than previously projected. A revised reference transition temperature l
vs power generation curve was prepared.
f 11
TABLE OF CONTENTS Page LIST OF FIGURES iv LIST OF TABLES v
I.
SUMMARY
OF RESULTS 1
II.
EACKGROUND 3
III.
LACBWR SURVEILLANCE PROGRAM S
A.
Test Materials and Specimens 5
B.
Capsule Design and Loading Arrangements 5
C.
Flux Wires and Temperature Indicators 8
D.
Impact and Tensile Properties of Unirradiated l
Materials 8
1 IV.
TESTING OF SPECIMENS AND EVALUATION OF DATA 11 A.
Capsule Disassembly 11 B.
Test Equipment and Procedures 11 C.
Evaluation of Thermal Monitors and Flux Wires 12 D.
Impact and Tensile Test Results 17 E.
Check Chemical Analyees 31 V.
ANALYSIS OF RESULTS 33 A.
Reference Temperature Projections 33 B.
Material Toughness Projections 40 VI.
RE"ERENCES 43 APPENDIX A - SKETCHES AND DRAWINGS FROM ACNP66513 45 APPENDIX B - UNIRRADIATED CHARPY V-NOTCH AND TENSILE DATA 55 APPENDIX C - DISCRETE ORDINATE TRANSPORT ANALYSIS 59 APPENDIX D - TANH-FIT CHARPY CURVES 69 e
111
~~
LIST OF FIGURES Figure Page 1
Internal Themal Shield Showing Locations of 20 Radiation Capsules and 2 Dosimetry Wire Holders (12) 6 2
Charpy V-Notch Properties of Plate NP-1055 25 3
C%rpy V-Notch Properties of Plate NP-1054 26 4
charpy V-Notch Properties of Plate NP-1056 27 5
Charpy V-Notch Properties of LACBWR Weld Metal 28 6
Charpy V-Notch Properties of Standard Material 29 7
Transition Temperature Response of LACBWR Vessel Surveillance Materials to Neutron Irradiation (16) 35 8
Effect of Neutron Fluence on RTNDT of LACBWR Ves-sel Surveillance Materials 37 l
9 Comparison of Current and Revised Curves Relating the Raference Transition Temperature to Plant 39 Operation 10 Shelf Energy Degradation Projections 41
~
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iV
LIST OF TABLES Table Page I
Characterization Data on LACBWR Surveillance Materials 7
II LACBWR Surveillance Capsule Contents and Removal Schedule 9
~
III Initial Transition Temperatures for LACBWR Sur-veillance Materials 10 IV Su==ary of LACBWR Plant Operations up to 1980 Refuelling 13 V
Material Surveillance Capsule Dosimetry Results 13 VI Su= mary of Neutron Dosimetry Results 18 VII Irradiated Tensile Properties of LACBWR Surveil-lance Material 19 VIII Charpy V-Norah Data on Plate NP-1055 20 IX Charpy V-Notch Data on Plate NP-1054 21 X
Charpy V-Notch Data on Plate NP-1056 22 XI Charpy V-Notch Data on Weld Metal 23 XII Charpy V-Netch Data on Standard Material 24 XIII Effect of Neutron Irradiation on LACBWR Vessel Surveillance Materials 30 XIV Calculated Neutron Flux Density Lead Factors for LACBWR Vessel Material Surveillance Capsules 34 XV Initial Transition and Reference Temperatures for LACBWR Pressure Boundary Materials 38 4
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I.
SUMMARY
OF RESULTS l
i The analysis of the data obtained from the vessel material surveil-lance capsules removed from the Lacrosse Boiling Water Reactor (LACBWR) vessel led to the following conclusions.
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1.
The irradiated properties of the LACBWR primary pres-sure boundary materials appear to be adequate to pro-vide continued safe operation of the plant according to present day criteria.
2.
The LACBWR vessel plate NP1056 is predicted to control the reference transition temperature (RTNDT) for ap-proximately 17 effective full power years (EFPY) of operation. The projected peak value of RT DT at the N
vessel I.D. for the refuelling outages scheduled I
through 1986 (based on 40,000 MWDe power generation per year) are as follows.
Refuelling Total Power Effective Full 2-NDT Year (MWDe)
Power Years (dag F) 1981 378,000 6.3 125 1982 418,000 7.0 128 1983 458,000 7.6 131 1984 498,000 8.3 134 1985 538,000 9.0 137 1986 578,000 9.6 140 3.
The LACBWR weld metal is predicted to control the RTNDT of the primary system after 17 EFPY of operation. At a projected peak vessel fluence of 2.7 x 1019 n/cm2 (E >
1 MeV) after 20 EFPY of operation, the shift in RTNDT, as controlled by the weld metal, is predicted to be 175'F.
Since the initial RTNDT of the weld metal has been taken to be O'F, the value of RTNDT for the pri-mary pressure system after 20 EFPY of operation is projected to be 175'F.
4.
The above projections are based on 6RTNDT values deter-mined at the 30 ft-lb level and on the results of a two-dimensional discrete ordinates transport calculation of the energy and spatial distribution of the neutron flux between the reactor core and the vessel.
The transport analysis showed that assuming 40 percent voids in the steam separators would yield lead factors (ratio of capsule neutron flux to vessel neutron flux) that pro-vide conservative values of neutron flux incident on the LACBWR pressure vessel wall.
Based on an analysis of the dosimetry results from the ten specimen capsules and two vessel wall dosimeters removed to date, the LACBWR vessel is projected to receive a peak fast fluence (E >
1 MeV) of 1.35 x 1018 n/cm2 each EFPY.
.5.
At a projected peak vessel fluence of 2.7 x 1019 n/cm2 (E > 1 MeV) after 20 EFPY of operation, the Charpy shelf energy of.the vessel weld metal is predicted to be reduced to 50 ft-lb.
The Charpy shelf ener-gies of.the vessel beltline plates are predicted to range from 54 to 64 ft-lb.
6.
The values of RTNDT and toughness at the 1/4-thick-ness location in the vessel wall are substantially better than those summarized above because the fast neutron 1 flux and fluence at the 1/4 T is 80 percent of that at the vessel I.D. surface.
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II.
BACKGROUND For many years, the basis for defining a minimum safe operating tem-l perature of a pressure system had been the Fracture Analysis Diagram (FAD) developed by Pellini and Puzak.(1)* The FAD is keyed to the drop-weight nil-ductility transition (DW-NDT) temperature defined by ASTM Method of Test E 208.(2) The Fracture Transition Elastic (FTE) temperature, above which stresses in excess of yield are required to propagate a large flaw, is indexed at DW-NDT + 60*F.
Until recently,Section III of the ASME Boiler and Pressure Vessel Code had defined the minimum permissible pressurization temperature as 60*F above the higher of (1) the DW-NDT temperature and (2) the minimum temperature at which a set of three Charpy V-notch specimens, represent-ing weld metal and heat-affected zone as well as base material, meet the fracture energy requirements specified by the Code for the particular material. The Charpy V-notch requirements ranged from 15 ft-lb minimum for steels having a specified minimum yield strength less than 35,000 psi to 35 ft-lb minimum for steels having a specified minimum yield strength of 75,000 psi and above.
Currently, the allowable loadings on nuclear pressure vessels are determined by applying the rules in Appendix G, " Fracture Toughness Re-quirements," of 10CFR50.(3)
In the case of pressure-retaining components ude of ferritic materials, the allowable loadings depend on the reference stress intensity factor (KIR) curve indexed to the reference nil ductility temperature (RTNDT) presented in Appendix G " Protection Against Non-ductile Failure," of Section III of the ASME Code.(4) The RTNDT is defined by Sec-tion III of the ASME Code as the highest of the following temperatures:
1.
Drop-Weight Nil Ductility Temperature (DW-NDT) per ASTM E 208;(2) 2.
60*F below the 50 f t-lb Charpy V-notch (Cv) temperature; and 3.
60*F below the 35 mil C temperature y
The initial RTNDT must be established for all materials, including veld metal and heat affected zone (HAZ) material as well as base plates.
and forgings, which comprise the reactor coolant pressure boundary.
It is well established that ferritic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron fluences in excess of 1017 neutrons per em2 (E > 1 MeV).(5)
In addition to a general dependence on neutron fluence, it has been established that tramp elements, particularly copper and phosphorous, affect the radia-tion embrittlement response of ferritic materials. (6-8)
- Superscript numbers refer to references listed at the end of the text.
3
-In general, the only ferritic. pressure boundary materials in a nu-clear plant which are expected to receive a fluence sufficient to affect RTNDT are those parent materials and welds which are located in the core beltline region of the reactor pressure vessel. As a consequence, one or more heats-of these ferritic materials must be monitored for radiation-induced changes in RTNDT Per the requirements of Appendix H, " Reactor Vessel Material Surveillance Program Requirements," of 10CFR50(3) and
~
ASTM E 185(9) which describe.the current recommended practice.for mon-itoring the radiation-induced changes occurring in the mechanical prop-erties of ferritic pressure vessel materials.
l Allis-ChaLaers provided such a surveillance program for the Lacrosse Boiling Water Reactor (LACBWR). The encapsulated Cy specimens are located on the I.D. surface of the thermal shield where the fast neutron flux den-
- sity is approximately twice that at the' adjacent vessel wall surface.
Therefore, the increases (shif ts) in transition temperatures of the mate-rials in the pressure vessel are generally less than the corresponding shif ts observed in the surveillance specimens.
However, because of azi-muthal variations in nantron flux density, the capsule fluences lead the maximum vessel fluenc, b, varying amounts.
This report describes the results obtained from testing and evalu-ating the capsules removed during the 1980 refuelling outage at LACBWR.
The results obtained from capsules removed during the 1972 and 1975 re-fuelling outages were reported earlier (10,ll), but have been reevaluated as described in Section IV of this report.
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4
l III. LACBWR SURVEILLANCE PROGIaM The LACBWR surveillance program is described in detail in ACNP-66513.(12) Twenty test specimen capsules (10 Type A and 10 Type B) were placed around the periphery of the core on the inner surface of the thermal shield as shown in Figure 1.
The capsule supports are located in such a way as to center the capsules axially about the horizontal mid-plane of the core.
In addition, two vessel wall dosimeter capsules (which do not contain mechanical property test specimens) were placed between the vessel wall and the thermal shield to assist in the determination of the acceleration factor for exposure between the test specimen capsule and the vessel wall locations.
Capsules lA and 13 had been removed during the 1972 outage, and the results have been previously reported.(10) Capsules 2A, 73, 9A, and 93, along with the two vessel wall dosimeter capsules, were removed during(ll)the 1975 refuelling outage, and these results have also been reported.
This report covers the testing of specimens from Capsules 3A, 3B, 8A, and 83, removed during the 1980 refuelling outage, and an analysis of all LACBWR surveillance data.
A.
Test Materials and Specimens Each radiation capsule contains 22 Charpy V-notch specimens machined from the vessel beltline materials, 6 Charpy V-notch specimens machined from a " standard" material, and 6 miniature tensile bars machined from one vessel beltline plate.
The vessel beltline materials include 3 ASTM A 302 Grade 3 vessel plates (NP1054, NP1055, NP1056) and weld netal. The ASTM A 302 Grade B " standard" material, furnished by the Atomic Energy Co= mis-sion, Chicago Operations Office (AEC-COO), was characterized by Battelle Northwest Laboratories.
The available data on chemistries and heat treat-ments of these materials are given in Table I.
The NP1054 Charpy V-notch specimens were machined from 18 tested DW-NDT specimens.
The NP1055 tensile specimens and the NP1055 and NP1056 Charpy V-notch specimens were machined frem excess plate material. All specimens were oriented parallel to the rolling direction and were lo-cated at the upper or lower quarter-plate thickness.
The impact specimen notches were oriented perpendicular to the plate surface. Drawings show-ing the location of specimens within the sample plates and specimen ma-chining drawings are given in Appendix A.
B.
Capsule Design and Loading Arrangements The radiation capsules are 24 in. long and were fabricated from 1-1/4-in. Schedule 80 TP304 stainless steel pipe with welded closure plugs at both ends. The top closure plug, fitted with a cable and lif ting ring assembly, was installed after filling the capsules as described below.
All specimens were cleaned in acetone, arranged in a clean box in the proper groups for each capsule, then bundled and wired.
The flux wires and temperature indicators described in Section III.C were also inserted in the 5
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i FIGURE 1.
INTERNAL TilERMAL SilIELD SIIOWING LOCATIONS OF 20 RADIATION CAPSULES AND 2 DOSIMETRY WIRE IlOLDERS(12)
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TABLE I CHARACTERIZATION DATA ON LACBWR SURVEILLANCE MATERIALS A.
Chemistries Mat'l Heat Data Chemical Analysis (%)-
Ident.
No.
Source C
Mn P
S Si Cu-Mo NP1054 A5848*
(17) 0.19 1.25 0.009 0.016 0.20 0.47 0.47 l
NP1055 A5848*
(17) 0.19 1.25 0.009 0.016 0.20 0.47 NP1056 A5852*
(17) 0.20 1.30 0.008 0.022 0.20 NP1056 A5852*
(14) 0.22 1.35 0.007 0.018 0.22 0.11 0.52 Weld (14) 0.10 1.39 0.016 0.006 0.43 0.18 0.55 0.15 l
Weld (19)
Standard N31438t (18) 0.22 1.33 0.017 0.013 0.30 0.07 0.52 3.
Mechanical Prooerties at RT Elong.
Mac'1 Data Tensile Yield in 2 in.
Charpy V-Notch Ident.
Source (ksi)
(ksi)
(%)
at 10*F(ft-lb)
NP1054 (17) 82.8 60.6 31 91, 94, 77 NP1055 (17) 86.5 57.5 30 100, 98, 90 NP1055 (12) 87.2 64.1 28 92, 76 NP1056 (17) 83.5 57.5 31 89, 90, 82 Standard (18) 97.0 76.9 25 60, 58 C.
Heat Treatment LACBWR plates and tests were annealed at 1950-2050*F, then heated to 1725-1775*F, held 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per inch min. and water spray quenched to 500*F, then tempered at 1200-1250*F air cooled; tests were stress relieved at 1100-1150*F (held 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> min.).
Standard material plate was charged into a 1100*F furnace, heaced to 1650*F, held 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, water quenched to below 300*F, recharged into a 750*F furnace, heated to 1200*F, held ~4 hours and air cooled.
- Lukens Steel Company.
t U.S. Steel Corporation.
7
Charpy V-notch specimen assemblies at this time. Aluminum powder was packed to a depth of 0.88 in. in the bottom of each capsule,.the capsules were purged.with argon gas, the Charpy V-notch specimen assemblies were inserted, and more aluminum powder was packed around the Charpy V-notch specimen bundles. After the tensile specimen assemblies were inserted, all void space was packed with aluminum powder, and the capsules were weighed to assure that each contained approximately the same quantity of powder. The closure welds were made, then each capsule was subjected to a liquid penetrant inspection. A drawing illustrating a typical capsule arrangement is included in Appendix A.
A summary of the test specimen contents and the current removal schedule of each capsule is given in Table II.
C.
Flux Wires and Temperature Indicators Three types of flux wires are contained in the surveillance capsules.
These are 0.021-in. diameter pure iron wire, 0.020-in. diameter pure nickel wire, and 0.020-in. diameter aluminum /0.1% cobalt wire. One piece of each wire, approximately 1-1/2 in. long, is located in the V-notch area of each of the seven layers of Charpy specimens in the test specimen capsules.
Two vessel wall flux vire assemblies, each containing iron, nickel, and aluminum /0.1% cobalt wires, were placed 180' apart in the annulus be-tween the vessel wall and the thermal shield. These assemblies, which ex-tend the length of the core, were also fitted with a cable and lifting ring. A drawing of the assembly is included in Appendix A.
Four Charpy V-notch specimens in each capsule contain low melting point eutectic alloys inserted in a hole drilled in the end of the speci-mens. The four eutectic alloys and their melting points are:
Allov
. Melting Point, 'F 2.6 As, 97.4 Pb 554 2.5 Ag, 97.5 Pb 579 0.5 Zn, 99.5 Pb 604 Pure Pb 621 D.
Impact and Tensile Properties of Unirradiated Materials The tensile and impact properties of the LACHWR surveillance mate-rials in the unirradiated condition have been reported previously.(13)
The detailed test data are presented in Tables 3-1 through B-5 in Appen-dix.B.
A summary of the initial DW-NDT te=peratures and 30 ft-lb Charpy V-notch "fix" transition temperatures are given in Table III.
8
~
TABLE Il LACBWR SURVEILLANCE CAPSULE CONTENTS AND REMOVAL SCllEDULE Removal Schedule Surveillance Tensile
.(full power Capsules Charpy V-Notch Impact Specimens Specimens years *)
Type Quantity NP1054 NP1055 NP1056 Weld Standard NP1055 Weld 1.4t A
1 6
10 6
6 3
3 B
1 10 6
6 6
3 3
- 2. Sit A
2 12 20 12 12 6
6 B
2 20 12 12 12 6
6 VW**
2 6ttt A
2 12 20 12 12 6
6 B.
2 20 12 12 12 6
6 10 A
2 12 20 12 12 6
6 B
2 20 12 12 12 6
6 15 A
2 12 20 12 12 6
6 B
2 20 12 12 12 6
6 Standby A
1 6
10 6
6 3
3 B
1 10 6
6 6
3 3
- One full power year equals 60,200 Mwt.
Withdrawals to be made during the nearest schedul.ed refuelling outage.
t Removed during August 1972 outage.
- Vessel wall dosimeters.
tt Removed during May 1975 outage.
ttt Removed during November 1980 outage.
d TABLE III INITIAL TRANSITION TEMPERAn~tES FOR LACBWR SURVEILLANCE MATER 1ALS 30 ft-lb DW-NDT Charpy "Fix" Mac'1 Temperature (OF)
Transition Ident.
Surface 1/4e Temperature (oF)
NP1054 10 10-
-h71055 75 NP1056 40 50 30 Weld 30 S:andard 60 s
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4 1
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IV.
TESTING OF SPECIMENS AND EVALUATION OF DATA SwRI utilized the procedure for shipment of the LACBWR surveillance capsules which had been used for the two previous capsule shipments.
Four capsules, 3A, 3B, 8A, and 8B, were removed during an outage which began on November 9, 1980.
The capsules were then taken to SwRI for analysis. One capsule was opened at a time so that the contents could be examined, iden-tified, and placed in indexed receptacles to prevent mixing v?.th the con-tents of the remaining capsules.
A.
Capsule Disassembiv The first hot cell operation was to remove the closure plug at each end of the capsule with a bandsaw, locating the cut per assembly drawings contained in ACNP-66513.(12) The capsule shells were cut in two length-wise with a milling machine. The aluminum powder was sintered to a degree similar to that previously reported (ll), but the spectrens could be broken out quite easily without damage.
The wires binding the assemblies together were removed, and the con-tents were carefully laid out so that the dosimeter wires could be recov-ered and identified as to location within the capsule.
The specimens were cleaned in an ultrasonic bath, examined to determine the specimen identifi-cation number, and placed in an indexed receptacle. Those Charpy V-notch specimens containing low-melting eutectic alloys were examined to determine which temperature indicators had fused during the exposure period, the re-sults being described in Section IV.C of this report.
B.
Test Equipment and Procedures The tensile specimens were tested in a 22,000-lb servo-controlled tension testing machine equipped with a strain gage load cell. Accesso-ries include a set of elevated temperature extensometer arms which attach directly to the specimen gage section, an Instron strain gage extensometer, an electric laboratory furnace, and an X-Y recorder.
The calibration of the load cell was verified prior to conducting the tensile tests with an elastic proving ring traceable to the U.S. Bureau of Standards.
Tests were conducted on each material at the temperature of the upper knee of the Charpy curve and at 550*F.
Elevated temperature tensile specimens were instrumented with two thermocouples wired to the top and bottom of the gage section of the specimen.
The Charpy V-notch tests were conducted on an instrumented SATEC e
impact machine permanently installed in a warm cell.
The calibration of the machine had been checked with a set of USAMMRC standards less than one year previous to the dates of testing. Nonambient specimen tempera-tures were obtained with a liquid bath.
The procedure permitted the op-erator to remove a specimen from the temperature conditioning bath, place it on the anvil, and break it in less than five seconds.
11
Nonambient test temperatures (tensile specimens and Charpy condi-tioning bath) were measured with thermocouples made from calibrated wire and a laboratory potentiemeter which is periodically checked against stan-dard voltage sources traceable to the U.S. Bureau of Standards.
The flux wires were weighed on a Mettler laboratory balance ther counted with a Ge(L1) solid state detector and a 4084-channel Norland multichannel analyzer.
In addition to the unknowns, 60Co, 137Cs, and 54Mn standards were counted to determine the efficiency of the experi-mental setup as a function of y-ray energy.
C.
Evaluation of Thermal Monitors and Flux Wires Examination of the thermal monitors revealed that all of the 554*F melting point eutectic alloys had fused. The presence of sintered alu-minum powder made it difficult to assess the condition of the 579'F al-loy specimens, but none of the 604*F or 621*F alloy specimens had fused.
Therefore, it was concluded that the maximum temperature reached by the contents of the four capsules during the operating period of the LACBWR vessel was above 554*F and below 604*F.
The specific activities of each flux wire, corrected to the plant shutdown date of November 9,1980 (hereaf ter referred to as the time of removal--TOR), were determined.
The first step in the calculation of the neutron flux is to correct the specific activities at TOR, A(TOR), to in-finitely dilute saturated activities at a selected power level, As:
m=n A(TOR)/As " E (1 - e-AT ) e-AH" m=1 where:
decay constant for the activation product A
=
equivalent operating days at the selected power level Tm
=
for the mth operating period number of days from the end of the mth operating period tm
=
to TOR.
The daily load charts and operating susraries from the LACBWR monthly oper-ating reports covering the period from July 10, 1967, to November 9, 1980, were utilized to determine values for Tm at 165 MWth and for tm to the TOR date.
The plant operations were divided into 64 operating periods as sus-marized in Table IV.
The resulting saturated activities for each flux wire removed from specimen capsules are given in Table V.
12
I 1
TABLE W 1
SU:^!ARY OF LACBWR PLANT OPERATIONS UP TO 1980 REFUELLING 1
l l
30erattsg teactor t u.aleet Oecar time I
Fortod
? stet Ihut dmm Operatir.g
% et Output Operacial After Pef 104 m
s ort n,
m, m
<t,)
<e.,
l i
1 37-10-47 02 25-e4 233 26 3.15 a 638 J2-2$ e4 03-19-e4 20 2
O bl9-e4 3 b2 bee 7
66 1.4J 6.611 33-Z hea 04-21-44 26 3
34-21-44 04-2 bee 5
41 0.37 6.540 06-26-44 05-37-94 11 35-07-94 05-30-64 23 364 2.23 4.1-6 05-30-44 11-20-64 74 5
11-20-e4 12-26-94 J4 09 2..a 6,338 12-26-94 01-14-99 21 4
01-16-e9 01-31-69 17
. JO 0.79 4.303 31-31-09 03-27-49 55
- 3 3.45 4.238 33-27-99 04-33-99 3.-0 b49 04-29-e9 26 1
34-2 be 9 35-03-49 39s
- 2. 1 4,208 05-O be9 37-21-49 79 9
37-21-49 10-14-49 55 9.i.3 55.61 5.044 lble-e9 35-32-70 200 13 35-02-73 C6-16 70
.7 3.054 19.51 3.797 06-L S-7 3 07-02-73 16 11 37-02-70 09-02-70 62 5.673 34.36 3.727 09-02-73 39-05-70 3
12 09-05-70 10-31-70 56 9.276 50.19 3.662 10-31-70
&2-20-70 50 13 12 I0-70 01-20-71 31
.753 28.79 3.581 01-20-71 01-30-71 10 8
1.237 7.32 3.363 14 31-30-71 32-07-71 32-07-71 10-71 3
=
=
15 32-10-71 02-25-71 15 2.262 13.39 3.565 32-25-71 02-23-71 1
9 1.240 7.32 3.333 16 02-23-71 33-09-71 03-09 71 03-11-71 4
.7 33-15-71 09-O b 71 169 21.659 131.27 3.355 09-03-71 12-31-71 119 58 12-31-71 31-08-72 S
992 6.01 3.228 it-06-72 31-12-7*
5 19 31-l b 72 0 3-3 M 2 78 13.729 46.26 3.165 33-31 72 04-03-72 3
23 0+-O b 72 35-19-72 6
7.695 46.6e 3.096 35-19 72 06-17 *2 29 21
- 6-17-72 06-23-72 6
790 6.79 3.061 06-2b72 96-26-72 3
22 06-26-72 3 7-M-7 2 8
1.081 6.35 3.253 37-0 -72 07-07-72 3
23 37-07-72 37-15-72 5
940 5.82 3.339 37-15-72 37-2*-72 7
2.
07-22-72 38-19-72
- 5 e.340 26.-4 3.306 04-19 72 10-16-72
!6 25 10-14-72
.1-02-72 19 2.094 12.69 2.929 11-02-72 11-05-72 3
26 11-05-72 11-23-72 19 2.264 13.72 2.908 11-23-72 12-02-72 9
27
.2-02-72 3 M 1-73 61 9.331
-4.67 2.538 32-31-73 32-06-73 3
29 02-0 -73 O b30-73 5-3.599 52.12 2.731 33-30-13 09-26-73 59 165 3.48 2.689 29 36-26-73 0k30-73 06-30-73 37-33 73 3
6 a-0 2.67 2.640 30 37-)3-13 37-09-73 37-09-73 37-16 73 5
31 37-le-73 37-16-73 2
370 2.26 2.673 07-L6-73 J7-21-13 5
32 37-Z M 3 09-;0-73 51 7.234 h3.34 2.617 09-Lb73 09-LF73 3
33 39-ib 73 1 M 3-73 51 6.529 39.*7 2.563 I M 3-73 12-26-73 53 g
12-2 H 73 03-)1-71
$5 3,315 57.67 2.445 3 3 7 a 3 M 3-74 2
35 33-O hia 05-:6-7; 6.
13.090
+1.13 2.379 0 5-M 35-49-74 23
- 6 35-29-74 37-15-74
.7 s.311 39.-4 2.309 37-15 74 17-;6-72 1
37 07-16-74 04-25-74 a3 6,714 0.69 2.233
- 3-;5-7a 09-20-71
- 3 29 09-20-74 39-24-74 esA 2.93 2.224 19-Za-74 lb10-7 16
- 9
.:-i3-7
- 1-L3-75 71
- 234 56.59 2.27 31-ib'$
31-i4-75 t
-0 31-i.-75 J 2-L L-7 5 3.
. 351 29. e4
. 295
,2-i 75
..;7-75 3
- 1 32-L7 75 e-;6-75 33
- 9. 22 51 *
. 33a 2.-14-75
?*-21-75 5
2
% 7 '
35-09-75
-,611 15.75
- s 1-
- 5-O+
39-;1-75 1-13
9 1
TABLE IV'(CONT.)
Operettag 9eseter 14uivalent Decay Tiza Feriod Sates Shutdown operettas Power outpet operating After Period (s)
Start sese Oave Davs N "3 t )
Savs (T-)-
Its) 43 08-11-75 12 11-75 122 17,?u 107.34 1.795 12-11-75 12-12-75 1
44 12-12-75 02-13-76 73
- 10. " 1 65.88 1.7*1 02-23-76 08-11-74 170 45 05-11-76 08-11-76 4
5 0.05 1.347 08-15-76 08-17-76 44 04-17-74 11-03-76
??
10.923' 64.20 1.467 11-03-76 11-06-76 3
7 618 3.75 1.457 47 11 06-76 11-13-76 11-13-76 11-19-76 6
44 11-19 76 02-02-77 75 10.498 63.62 1.376 02-02 77 02-04-77 2
47 02-W 77 05-11-77 96 9.472 57.41 1.278 05-11-77 03-09-73 302 30 03-09-73 04-27-78 49 5.109 30.96 927 04-27-75 05-10-78 13 161 18.676 113.19 753 51 05-10-78 10-15-73 10-13-73 11-08-78 21 52 11-08-79
'1-20-78 12 565 3.4 720 11-20-73 11-24-78 4.
53 11-24-73 01-13-79 50 3.903 23.66 ese 01-11-79 01-22-79 9
62 4.6%
23.46 -
595 -
54 01-22 79 03-15-19 03-25 79 03-26-79 62 7
157 0.95 526 55 05-26-79 06 02-79 06-02-79 04-03-79 1
32 3.587 13.56 493 56 06-03-79 07-05-79 07-05-79 07-06-79 1
60 S.007 49.01 432 57 07-06-79 09-04-79 09-4 79 09-07-79 3
21 2.225 13.49 404 53 79-c7-79 09-2S-79 09-28-79 10-06-79
-4 113 15.434 93.54 232 59 10-06-19 02-01-40 02-01-40 02-04-40 3
62 8.488 51.44 217 60 02-04-80 04-06-40 04-16-40 04-30-40 24 52 7.391 44.79 141 61 04-30-30 06-21-40 06-21-80 06-23-80 7
42 5.055 30.64 92 62 W 23-40 03-09-80 08-09-80 05-19-80 9
7 323 1.56 76 63 38-15-60 03-25-60 08-25-80 09-91-40 9
64 09-03-40 1.1-09 -
67 9.426 31.07 0
Total 337.557 2.045.30*
- 2.045.50 davs - 1.76757 s 103 seconda.-
9 i
14_
TABLE V MATERIAL SURVEILLANCE CAPSULE DCSIMETRY RESULTS (Saturated Activity, dps/mg) 54Fe(n,p)54ya 58Ni(n,p)58 o Capsule 59co(n,y)60Co C
3AT
.7575E+08
.7027EM 4
.9055E+05 3AU
.7877E+08
.6974E+04
.9243E+05 3AV
.8592E+08
.7288E+04
.9334E+05 3AW
.9558E+08
.7274E+04
.9736E+05 3AX
.9532E+08
.5946E+04(a)
.9348E+05 3AY
.ll62E+09
.6915E+04
.9145E+05 3AZ
.8420E+08
.6725E+04
.9153E+05 3BT
.7271E+08
.6967E+04
.8804E+05 3BU
.7800E+08
. 4960E+04 (a)
.8944E+05 3BV
.8428E+08
.7089E+04
.9234E+05 3BW
.9066E+08
.6943E+04
.9195E+05 3BX
.8793E+08
.2689E+04
.9010E+05 3BY
.8248E+08
.6638E+04
.8866E+05 3BZ
.8542E+08
.6078E+04
.8586E+05 8AT
.7173E+08
.7018E+04
.8849E+05 8AU
.7697E+08
.7125E+04
.8319E+05 8AV
.8475E+08
.7303E+04
.9155E+05 8AW
.8751E+08
.7353E+04
.9148E+05 SAX
.8327E+08
.6948E+04
.8961E+05 SAY
.8113E+08
.6746E+04
.8803E+05 8AZ
.8028E+08
.6471E+04
.8589E+05 83T
.7747E+08
.7308E+04
.9270E+05 8BU
.7456E+08
.7375E+04
.9280E+05 8BV
.8917EM 8
.7641E+04
.9586E+05 8BW
.9005E+08
.7308E+04
.9525E+05 SBX
.9357E+08
.7627E+04
.9430E+05 83Y
.9404E+08
.7178E+04
.9313E+05 83Z
.9039E+08
.7025E+04
.9127E+05 Avg. =.8529E+08 Avg. =.7040E+04 Avg. =.9125E+05 (a) Values not used in co=puting averages.
15
The neutron flux density is given by :
As
=
N#
O where:
2 energy dependent neutron flux density (n/cm sec)
=
saturated activity (dps/mg target element)
A
=
s NO number of target atoms /mg target element
=
2 spectrum-averaged ' activation cross section (cm ),
3
=
In the analysis of the LACBWR neutron flux dosimeters, the neutron flux density calculations were based on the results ob,tained from the iron and nickel w1res.
(The bare cobalt wires were sensitive to thermal and epithermal flux as well as the fast flux.) The value of U was based on two spectra:
1.
A fission spectrum-averaged cross section.
This was utilized for reference only because much of the early surveillance program' data in the literature is based on the use of a fission spectrum-averaged cross section.
2.
'A calculated spectrum-averaged cross section. The DOT 3.5. two-dimensional discrete ordinates transport code was used to calculate the neutron flux densi-ties and spectra at various points of interest out-side the LACBWR core. This information was utilized to determine capsule lead factors (ratio of the neu-tron flux density at the capsule locations to the maximum neutron flux density incident on the pres-sure vessel wall) as well as the spectrum-averaged i
cross sections for the 54 e(n.p)S4 n and 58Ni(n,p)58 o F
M C
reactions.
Details of the DOT 3.5 analysis are pre-sented in Appendix C.
As discussed in Appendix C, the neutron flux densities and spectra at the surveillance capsule and vessel wall locations are dependent on the void concentration in.the steam separators since the steam separators are located between the reactor core and the surveillance capsules (and, of course, the vessel wall). This analysis indicated that an average void content of 30% may be reasonable, but that a conservative approach would be to assume an average void content of 40% because-this leads to
~
a higher calculated value of vessel wall neutron flux density.
16
l The flux values obtained for each capsule were multiplied by 1.768 x 108 seconds (the equivalent operating time at 165 MWth) to de-termine the corresponding values of neutron fluence. The neutron flux and fluence determinations obtained with each =ethod are sn-marized in Table VI.
The dosimetry analyses reported earlier for the six previous cap-sules(ll) were based only on the iron activities. Reevaluated fluxes and fluences for these capsules, using both the iron and the nickel data, are also included in Table VI.
It is of interest to note that the re-evaluated flux for Capsules LA and 13 agrees well with the SAND-II ca~cu-lation based on a 3WR spectrum reported earlier.(10)
D.
Impact and Tensile Test Results The results of tensile tests conducted on specimens removed during the 1980 outage are given in Table VII. Examination of the 550*F data in Table VII indicates that the plate and weld materials experienced a degree of radiation hardening similar to those in Capsules lA and 73.511)
This correlates with the results of the dosi=etry analysis since all six capsules received about the same fluence.
The Charpy V-notch i= pact data obtained on the specimens removed during the 1980 refuelling outage are given in Tables VIII through XII.
Charpy V-notch fracture energy transition curves were de3 eloped by a least-squares fit of each data set to the relationship:
Y=A+Btanh(
0) where:
Cy function (fracture energy or lateral xpansion)
Y
=
Cy test te=perature, deg F T
=
Interceptwhentanh(-
)=0 A
=
Slope 3
=
Temperature at transition midpoint, deg F T
=
o One-half of transition range, deg F.
C
=
The resulting transition curves, given in Appendix D, suffered from the lack of lower shelf data. The hand-drawn curves given in Figures 2 through 6 were used to define ARTNDT for each material. A su= mary of the transition te=perature shifts at 30 ft-lb, 50 ft-lb, and 35 mil lateral expansion, as well as irradiated upper shelf energies, is presented in Table XIII.
17
TABLE VI
SUMMARY
OF NEUTRON DOSIMETRY RESULTS Spectrum 5, E > 1 MeV (barns)
Capsule Neutron Flux Density,(a) Neutron Fluence,(a)
Type 54Fe (n, p) S4Mn 58Ni(n p)58Co Identification E > 1 MeV (n/cm2sec)
E > 1 MeV (n/cm2)
Fission (b) 0.113 3A 9.95 x 1010 0.113 3B 9.53 x 1010 0.113 8A 9.90 x 1010 0
0.113 8B 10.40 x 1010 Avg. = 9.96 x 1010 1.76 x.1019 DOT 3.5(c) 0.177 0.223 3A 6.15 x 1010 s
0.171 0.223 3B 5.90 x 1010 0.177 0.223 8A 6.01 x 1010 s
0.177 0.223 8B 6.32 x 1010 oo Avg. = 6.10 x 1010 1,og x 1019 0.176 0.222 1A 6.58 x 1010 0.176 0.222 IS 5.92 x 1010-Avg. =
6.25-x 1010 2.81 x 1018 0.168 0.214 2A 1.06 x 1011 0.168 0.214 7B 1.05 x 1011 Avg. = 1.05 x 1011 1.02 x.1019 0.184 0.231 9A 6.34 x 1010 g
0.184 0.231 9B 6.70 x 1010 Avg. = 6.52 x 1010
'6.32 x 1018 (a) Neutron flux densities and fluences subject to a i 15% uncertainty (10).
(b) Uf for fission-averaged cross section based on ASTM E 261.
5, E > 1 MeV = Ug/0.693.(5)
(c) 3, E > 1 Mev, based-on 40% voids in steam-separators, per Appendix C.
TABLE VII IRRADIATED TENSILE PROPERTIES OF LACBWR SURVEILLANCE MATERIAL Test Material Capsule Temperature UTS
.2% YS Elongation.
R.A.
No.
Identification Identification
(*F)
(ksi)
(ksi)
(%)
(%)
1 Weld 3A
+150 (a)
(a)
(b) 65.9 2
8A
+150 91.2 76.2 (b) 64.4 3
8A
+150 88.9 69.6 26.2 52.6 l
4 3A
+150 86.0 64.5 28.0 56.5 1
3 8B
+150 91.2 74.1 23.5 52.0
)
6 3B
+150 87.8 68.3 30.2 61.0 l
14 8B
+550 77.4 57.4 19.7 53.8 18 3B
+550 80.3 61.8 20.3 49.3 19 8A
+550 88.3 70.4 20.7 55.4 r.
20 3A
+550 34.1 68.1 (b) 61.1 21 3B
+550 89.3 72.5 20.4 53.2 22 7
8B
+550 88.8 69.4 21.3 57.0 7
NP-1055 3A
+200 88.8 68.4 25.3 67.8 8
3A
+200 88.6 69.2 24.9 67.4 9
8B
+200 91.1 70.9 24.5 66.8 10 3B
+200 91.7 71.0 26.2 68.2 I
11 8A
+200 93.4 72.9 23.4 68.2 12 8A
+200 91.9 71.6 24.5 67.8 13 8B
+550 91.7 66.5 22.0 61.1 15 3B
+550 89.8 65.5 22.2 64.6 16 8A
+550 91.1 67.6 20.6 60.5 17 3A
+550 85.5 63.2 26.4 63.9 23 8B
+550 90.6 67.3 20.6 60.5 24 9
3B
+550 90.3 66.8 21.2 63.1 (a) Extensometer malfunction.
(b) Specimen failed outside gage marks.
TABLE VIII-CHARPY V-NOTCH DATA ON PLATE NP-1055 (Removed from-LACBWR in 1980)
Lateral Fracture I
Capsule Temp.
Energy Expansion Appearance
-No.
(*F)
(ft-lb)
(mils)
(% shear) 8A
- 40 22.0 16 nil 83
- 40 18.0 14 nil 3A
- 40 21.0 17 nil 3B
-40 20.0 18 nil 8A-'
- 20 27.0 21 nil 8E
- 20 39.0 29 nil 3A
- 20 16.0 15 nil 3B
- 20 45.0 35-nil 8A 0
53.0-22 nil 8B.
0 9.5 19 nil 3A 0
43.0 20 5
3B 0
56.0 29
'10 SA 0
28.5 40 nil
+
~8B 01 22.5 10 nil 3A 0
~22.5 35 nil 3B 0
39.0 43 nil 8A
. + 20 36.5 29 5
8B
+ 20 15.0 15 nil 3A
+ 20 33.5 29 5
3B
+ 20 63.5 52 15 8A
+ 40-51.0 44 15 8B
+ 40 26.5 27 nil-3A
+ 40 76.0.
61 5
3B
+ 40 64.0 52 25 8A
+ 74 75.0 58 35 8B
+ 74 81.0 59 50 3A
+ 74 91.0 70 50 3B
+ 74 83.5 65 40 l.
'8A
+110 103.5 85 100 8B
+110 97.0 78 100 3A
+110 110.0 85 100 3B
+110 38.5 71 70 1
l SA
+150 96.5
~80 100 8B
+150' 112.5 90 100 3A
+150 112.0 77 100
,33
+150 101.0 82 100 8A
+200 102.5 87 100 83
+200 112.5 91 100 3A
+200 118.5 90 100 3B
+200 103.0 86 100 20 f
..._m
I TABLE IX-'
CHARPY V-NOTCH DATA ON PLATE NP-1054 (Removed from LACBWR in 1980)
Lateral Fracture
- Capsule
' Temp.
Energy ~
Expansion Appearance No.
(*F)
(ft-lb)
(mils)
(% shear) 8A
- 20 37.0 30 nil 3A
- 20 23.5 19 nil 8A
+ 40 70.5 58 10 3A
+ 40 74.0 67 25 8A
+ 74 84.0 65 40 3A
+ 74 64.5 53 30' I
8A
+110 102.5 82 100 3A
+110 94.0 82 90 8A
+150 107.5 89 100 3A
+150 108.0-89 100
~
8A
+200 107.0-90 100 3A
+200 102.0 88 100-I 9
r e
h i~
21 i
TABLE X CHARPY V-NOTCH DATA ON PLATE NP-1056 (Removed from LACBWR in 1980)
-Lateral Fracture
. Capsule Temp.
Energy Expansion
' Appearance No.
(*F)-
(ft-lb)
(mils)
(% shear) 8B
+ 74 35.0 25 nil 3B
+ 74 16.5 18 5
8B
+110 16.5
'22 nil 3B
+110 37.5 39 5
83
+150 41.0 45 40 3B
+150 41.0 43 20 8B
+200 46.0 48 20 3B
+200 74.0 65 65 8B'
+250 91.0 75 100-3B
+250 78.0 71.
95 8B
+300 88.5 79
'100 3B
+300 91.5 84 100 22
l TABLE XI CHARPY V-NOTCH DATA ON WELD METAL 1
(Removed from LACBWR in 1980)
Lateral Fracture Capsule Temp.
Energy Expansion Appearance No.
(*F)
(ft-lb)
(mils)
(% shear) 8A
+ 40
-15.0 16 20 8B
+ 40 14.5 16 nil-3A
+ 40 23.0 25 15 3B
+ 40 7.5 9
5 8A
+ 55 18.0 21 5
8B
+ 55 23.0 26
-25 8A
+ 74 30.0 31 30 8B
+ 74 35.5 30 25 3A
+ 74 26.0 42 25 3B_
+ 74 38.5 24 5
3A
+ 74 42.5 22_
25 8A
+110 44.5 44 65 8B
+110 47.5 45 65 3A
+110 24.0 28 20 3B
+110 45.0 44 100-3B
+110 31.0 35 60 8A
+150 55.5 51 95-8B
+150 54.0 54 80 3A
+150 56.5 55 10 3B
+150 53.5 54 95 8A
+200 55.5 58 100 8B
+200 56.0 56 100 l
3A
+200 60.5 63 100 3B
+200 52.5 52 100 l
I i
[
23 r
TABLE XII CHAB?Y V-NOTCH DATA ON STANDARD MATERIAL
(% moved from LACBWR in 1980)
Lateral Fracture Capsule Temp.
Energy Expansion Appearance No.
(*F)
(ft-lb)
(mils)
(% shear) 8A 0
14.5 14 5
83 0
14.5 14 5
3A 0
17.5 18 10 3B 0
16.0 14 5
l 8A
+ 40 22.0 22 10 8B
+ 40 20.5 20 L5 3A
+ 40 44.5 37 20 3B
+ 40 28.5 25 10 8A
+ 74 46.5 38 25 8B
+ 74 43.5 35 10 3A
+ 74 55.0 51 95 3B
+ 74 42.0 36 15 8A
+110 53.0 48 100 83
+110 52.0 47 100 3A
+110 69.5 58 100
~
3B
+110 56.0 51 100 8A
+150 61.5 61 100 8B
+150 66.5 64 100 3A
+150 74.0 64 100 3B
+150 69.0 67 100 8A
+200 65.0 59 100 8B
+200 71.5 60 100 3A
+200 81.0 66 100 3B
+200 72.0 62 100 24
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100 200 300 400 Test Temperature, deg F FIGURE 5.
CHARPY V-NOTCH PROPERTIES OF LACBUR WELD METAL 28 J
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t
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. y { J, l.
t,
j i, t
, i i
t 0
-200
-100 0
100 200 300 400 Test Temperature, deg F FIGURE 6.
CHARPY V-NOTCH PROPERTIES OF STANDARD MATERIAL 29
TABLE XIII EFFECT OF NEUTRON-IRRADIATION ON LACBWR VESSEL SURVEILLANCE MATERIALS NDT'*
1rradiated Haterial Neutron Fluence 30 ft-lb TT 50 ft-Ib TT 35 mit TT RT Identification n/cm. E > 1 HeV Increase, des F lucrease, deg F Increase, deg F Increase. des F Cv Shelf, ft-Ib 2
NP-1055 1.08 x 10I9 80 90 85 80 106 NP-1054 1.08 x 1019 nit 20 (b) ut!
105 NP-1056 1.08 x 1019 85 100 70 100 90 y
IJeld Metal 1.08 x 1019
!!O 95 110 llo 56 Standard 1.08 x 10I9 105 110 105 105 66 (a) ARTNDT determined at 30 f t-Ib level per ASTM E 185-79.
(b) Not determined. No lateral expansion data available for untrradiated material.
l l
l l
E.
Check Chemical Analyses The copper content of one broken Charpy V-notch specimen, representing each material type, was determined with'an X-ray fluorescent technique.(15)
The results, summarized below, agree well with those reported earlier, see Table I.
Material Identification Copper Content (wt. %)
Plate NP-1054 0.10 Plate NP-1055 0.14 Plate NP-1056 0.14 Weld Metal 0.14 Standard Material 0.07 e
e 31
V.
ANALYSIS OF RESULTS The analysis of the data obtained from the LACBWR vessel surveillance specimen data has the following goals:
1.
Estimate the period of time over which the properties of the vessel beltline materials will meet the frac-ture toughness requirements of Appendix G of 10CFR50.
This requires a projection of the measured reduction in C upper shelf energy to the vessel wall using y
knowledge of the energy and spatial distribution of the neutron flux and the dependence of Cy upper shelf energy on the neutron fluence.
2.
Determine the increase in RTNDT as a function of reac-tor pcwer generation. This requires a projection of the measured shift in RTNDT to the vessel wall using knowledge of the dependence of the shift in RTNDT on the neutron fluence and the energy and spatial dis-tribution of the neutron flux.
The energy and spatial distribution of the neutron flux for LACBWR was calculated as described in Appendix C.
The calculated lead factors for each of the LACBWR surveillance capsules are presented in Table XIV.
The vessel I.D. surface lead factors vary from 2.13 for specimen cap-sules 2A, 2B, 7A, and 7B to 1.37 for specimen capsules 4A, 4B, 9A, and 9B.
The neutron flux densitites at the vessel wall dosimeter capsule locations are very nearly equal to the maximum neutron flux density in-cident on the pressure vessel I.D. surface.
A.
Reference Temperature Proiections An independent program for evaluating the response of the LACBWR pressure vessel material to accelerated neutron irradiation was carried out by the Naval Research Laboratory (16) at the request of the Atomic Energy Commission and Allis Chalmers Manufacturing Company.
Figure 7, taken from Reference 16, summarizes the results obtained en specimens of LACBWR pressure vessel steels irradiated in the Oak Ridge Low Intensity Test Reactor at a controlled temperature of 550*F.
(The ASTM reference material referred to in Figure 7 is not the same heat of A302B steel which is being used as a reference material in the LACBWR surveillance program.)
The transition temperature shifts obtained on the corresponding ma-terials which have been removed from LACBWR have been added to Figure 7 for comparison.
The 30 ft-lb transition temperature shifts are plotted at fluences based on a fission spectrum-averaged cross section to be con-sistent with the procedures employed by the Naval Research Laboratory at that time. The results obtained from the LACBWR surveillance program show good agreement with the results reported in Reference 16.
33
)
TABLE XIV CALCULATED NEUTRON FLUX DENSITY LEAD FACTORS FOR LACBWR VESSEL MATERIAL SURVEILLANCE' CAPSULES Capsule Lead Factor (a)
Identification I.D. Surface 1/4 T(b)
M 4 T(c)
LA, 1B 1.62 2.02 1.60 2A, 2B 2.13 2.66 4.73 3A, 3B.
1.66 2.07 3.68 4A, 4B 1.37 1.71 3.03 5A, 5B 1.50 1.87 3.33 6A, 6B 1.62 2.02 3.60 7A, 7B 2.13 2.66 4.73 8A, 8B 1.66 2.07 3.68
~
9A, 9B 1.37 1.71 3.03 10A, 10B 1.50 1.87 3.33 V.W.
0.96 1.20 2.10
[
(a) Ratio of neutron flux density, E > 1 MeV, at the capsule location to the maximum incident on'or within the pressure vessel wall with 40% voids in the steam separators.
(b)~ 1/4 T is 1 in. within the pressure vessel wall. Neutron flux den-sity is attenuated'to 80% of that at the vessel I.D. surface.
(c) 3/4 T is 3 in within the pressure vessel wall. Neutron flux den-
.sity is attentuated to 45% of that at the vessel I.D. surfact.
34
g g4ty3 600_,ggg
- ggj g g ;ligig
- gj i,
j i
_ A302-8 STEEL PLATE Z
- O 6-lN. ASTM REFERENCE
-LACROSSE REACTOR-Z l
1.L
_ O 4-lN. NP 1056 l
'_, 500 3 "H AZ" NP 1056 276 1
Z 0 WELD METAL,NP IOS6
~
g m
- O 4-IN. NP 1057
- A 4-IN. NP 1058
- c. f.
U T
- 'd!N-O
- s:
Z 222 400 w
TREND FOR I.'
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g
~
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w
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t1.
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w 300
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b-
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~
TREND FOR U
550* F, LITR k
lll b' '", 'I gQQ IRRAOfATIONS
.l t
l m
LACBWR LnfR p.
b
..f?)
U Y
e i
d-Z
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o c.s.-?
cs m
100 56
.. cg.. ~/
. g.:.. :c // g) 0 0 O
6 s.-
O 0
0 ONRL 0-I II I'll I II IIII I II IllI I II II" O 1017 1018 1019 1020 NEUTRON FLUENCE l
( n/cm2>IMev, FISSION SPECTRUM)
FIGURE 7.
TRANSITION TEMPERATURE RESPONSE OF LACBWR VESSEL SURVEILLANCE MATERIALS TO NEUTRON IRRADIATION (16) 35
l I
Trend curves could be constructed on Fi method is to employ Regulatory Guide 1.99.(8)gure 7, but a more. appropriate-In Figure 8, the curve iden-(
l tified as " upper limit" is taken from Figure 1 of Regulatory Guide l.99, and the other two solid curves are computed for the LACBWR weld metal and l
Plate 1056 using the expression A = [40 + 1000(% Cu - 0.08) + 5000 (% P --
O.008)] [f/10 9]1/2 The plotted data points represent results obtained 1
from the 1980 surveillance capsules plus the reevaluated embrittlement data from the previous capsules (References 10 and 11). Figure 8 shows l
that the weld metal is the most radiation sensitive of the LACEWR vessel surveillance materials but that the measured transition temperature shifts l
fall below the calculated trend curve.
The LACBWR vessel plate material, on the other hand, appears to follow tne calculated trend curve quite well.
Averaging the dosimetry results obtained from the capsules tested to date indicates that the peak neutron flux, E > 1 MeV, incident on the pres-sure vessel wall is:
1 6.25 x 1010. 1,05 x }oll, 6.52 x 1010. 6.10 x 1010 4.30 x 1010 2
n/cm.sec
=
4 1.62 2.13 1.37 1.66 The maximum neutron fluence, E > 1 MeV, incident on the pressure vess21 wall per EFPY is 4.30 x 1010 n/cm2.sec x 3.15 x 107 sec/yr = 1.35 x 1018 n/cm2 The next step is to predict the RTNDT of the LACBWR vessel as a func-tion of reactor operation.
The values of initial (unirradiated) RTNDT-given in Table XV have been established for this purpose. As indicated by Table
~
XV, the minimum pressurization temperature was initially. controlled by ma-terials in the main steam and forced recirculation systems. (17) Because these materials are not exposed to a significant amount of neutron radia-tion, Plate NP-1056 became the controlling material within the first EFPY,
-since it had the highest initial RTNDT.
T The analysis of the data obtained from the 1975 capsules indicated that the weld metal would control the primary system RTNDT after 3 EFPY.
However, that analysis (ll) defined ARTNDT at the 50 ft-lb level (as re-quired by Appendix G of 10CFR50 in effect at the Ethne) which resulted in an overestimate of the weld metal ARTNDT because of the proximity of 50 ft-lb to the upper shelf energy. All ARTNDT values shown in Figure 8 were redetermined at the 30 ft-lb level per ASTM E' 185-79(9) as currently practiced by the NRC.
The projection of the value of RTNDT as a function of plant opera-tions is shown in Figure 9.
This projection is based on both the LACBWR weld metal and vessel plate trend curves of Figure 8 using the average pressure vessel fluence rate of 1.35 x 1018 n/cm2, E > 1 MeV, per EFPY.
Also included in Figure 9 is the RTNDT vs.-EFPY curve currently in the LACBWR Technical Specifications.
The reevaluated vessel material sur-veillance data indicate that (1) the embrittlement rate is less than pre-viously predicted, and (2) the vessel plate NP-1056 will control the pri -
mary system RTNDT for more than 15 EFPY of operation because of its high initial value of RTNDT*
36
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c-4 3 Sap *a2nse2adsal acua2ayau 30 ananasn[py 37 i
TABLE XV INITIAL TRANSITION AND REERENCE TEMPERATURES FOR LACBWR PRESSURE B)UNDARY MATERIALS A.
Main Steam and Forced Recirculation Materials (19)
Drop 20 ft-lb Charpy Heat
. Weight V-Notch Transition Component Material No.
NDT (OF)
Temperature (oF) 20-in. Piping A-335, P-11 B-2795
-30
-27 20-in. Piping A-335, P-11 B-3064
+10 20-in. Piping A-335, P-ll B-3080
+20 Pump Casing A-217, C-5 5-272
+30 20-in. Fittings A-217, WC-6 2-1823
>+10
+70 Valve Casing A-217, WC-6 C-842
+50 Roto-valve Casing A-216, WCB C-853
+10 Roto-valve Casing A-216, WCB C-861
+30 Roto-valve Casing A-216, WCB C-863
+10 Roto-valve Casing A-216, WCB C-903
+30 B.
Pressure Vessel Surveillance Materials Material Initial RTNDT Ident.
DWNDT 30 ft-lb 50 ft-lb 35 mil deg F NP1055
-75
-55
-65 0(a)
NP1054 10
-15
-5 10(b)
NP1056 50 15 90 55 50(b)
-20 30
-15 0(a)
(a)
Since DWNDT' tests were not run, RTNDT = 30 ft-lb Cy IT or 00F, whichever is higher.(18)
(b)
RTNDT is the higher of (1) DWNDT, (2) 600F below the 50 ft-lb C TT (increased by 200F because specimens are longitudinally yoriented), and (3) 600F below the 35 mil LE Cy TT (increased by 200F because specimens are longitudinally oriented).(18) 38
400 _
a= g - --
w
/'.
f' Current Technical Specificatione l Y'-
ij l
1 700 x
l - W-i
- f-
+
- i; e
ec o
i u
s i a 6% il i t ;itil o ig i
i
- 1 i
/i
.i T
i
.. g'I f
' j.
t /
.! ii i
p 1,,.-
y, NP-1056
,i
.-o g
i i ii h
,f i. m, m
o ii1
- iii*Ir I;f t:itii
'h.
i i
- It
! 11l #ii'
'lt' i;
if l I i
t i
px 100 =_i x1. - - - - - - - - -
. - -. = - - = - -
= - = - - = - -= =-
r-t
- = - - - - - -w y---. - -..
= _ -. - - _.. _. -. _
80
-~
.gg
_g =- _
,"j Weld Metal
- ~ ~-
o>
60 1
I h=
g g___--
--===; m-1. = =;-
g
_ 3 ; g y g = g.
==
=
=_
==
- 0 m
I
~
20 1
2 3
5 7
10 20 30 53 Effective Full Power Years at 165 MW t
FIGURE 9.
COMPARISON OF CURRENT AND REVISED CURVES RELATING THE REFERENCE TRANSITION TEMPERATURE TO PLANT OPERATION 39
B.
Material Toughness Projections Appendix G of 10CFR50 requires that the primary pressure boundary materials retain a Charpy V-notch upper shelf energy of at least 50 ft-lb through the life of the plant. The LACBWR pressure vessel plate materials have high C upper shelf energies and appear to be relatively insensitive y
~to radiation embritriement. The irradiated upper shelf energies for the LACBWR plate materials given in Table XIII are for longitudinally-oriented specimens, but if they are reduced by 35 percent as recommended in the NRC Standard Review Plan,(18) the values for material irradiated to 1.08 x 1019 n/cm2 (E > 1 MeV) range from 58 ft-lb for Plate NP-1056 to about 69 ft-lb for Plate NP-1055. The weld metal exposed to the same fluence has retained a Charpy shelf energy of 56 ft-lb.
The initial (unirradiated) values of Charpy V-notch upper shelf energies were not well established for the LACBWR vessel beltline mate-rials.
However, the trend bands for decrease in shelf energy given in Regulatory Guide 1.99(8) can be used to estimate the Charpy shelf ener-gies at the end of the 20 EFPY design life (E.O.L.) as shown in Figure 10.
Regulatory Guide 1.99 estimates that the shelf energies of the weld metal, O.15% 'Cu plate (NP-1055) and the 0.10% Cu plate (NP-1056) have been reduced by 29%, 24%, and 19%, respectively, at a fluence of 1.08 x 1019 n/cm2, 1
E > 1 MeV). At the peak 20 EFPY vessel fluence of 2.7 x 10 9 n/cm2 (E >
1 MeV),-the guide predicts that the shelf energies of these materials will have been reduced by 36%, 30%, and 24%, respectively. The projected E.0.L. shelf energies are:
(1)
Weld Metal:
Shelf at E.0.L. = 56 ( f - 0
= 50 ft-lb Cy 0
(2)
Plate NP-1055:
ShelfatE.O.L.*69(f[,
= 64 ft-lb Cy (3)
Plate NP-1056:
Cy Shelf at E.O.L. : 58
[
= 54 ft-lb The LACBWR material surveillance program does not include specimens representing the HAZ material.
However, the companion program conducted by NRL(16) did include HAZ specimens machined from Plate NP-1056. The re-sults obtained indicated that the radiation sensitivity of the HAZ material, as measured by the increase in the transition temperature, was similar to that of Plate NP-1056.
Since the transition temperature of the unirradi-ated HAZ material was nearly 100*F below that of the base plate, and since
~
the longitudinal Cy shelf energy was in the 90 ft-lb range (i.e., 60 ft-lb in the transverse direction) after being irradiated to 2.1 x 1019 n/cm2 (E > 1 MeV, based on a fission spectrum-averaged cross section), it is con-cluded that the properties of the HAZ material will not pose a problem to
~
LACBWR operations.
40
60 g n:
i l
a, n I-r
- si
.. l i
o h
i !.}....
l I Y
-- k I
li J g y q
- hd lU l
I a,
+
4 40 7--h
] [ j-
- ft,1,J.
~
hl !!-
g a.
l
... {;
t a.
.--+,
- inilE p,
i i
!g i
~
.~
g 7;,
b Ii
!,t.
.T u
- l
.r p
c q,;
{
h.,}IW n
-f-I,. :.W[
t
,, - r'l 3
2 20 h
1
+j..-
l
'~'
l 0.015% Cu Weld g
.a I
0.015% Cu Plate h[
'{
W e
I c
,f
'~
l
().010% Cu Plate I d
-~~
~~
-~"
10 p
i i
j
.j 3
] [h.
i
.j.
j.j.
7
-j T-'
' [
j p ;
r m
g a
1
!l i
I i
r r
c i
e i
a j[iIl i ll
~
~
ii, E..-
~T l
p pljp
~
l 6
i jt i,tt,
'i I
Code:
U-o r
i e
+
th y
e
..t e o
I t
9 i
v NP-1055 b
i I i o
l A NP-1056
{[j, o
A 4
I j b j
i j l
- Weld 3
. hd I
1 i
il l
- q 3 ;
-l l i'!;
.j i.
5
[I
}'
l l
ll i
,s.
I i dl
~l l
'I l l, t
l M
I i
I i
i r
2 2 x 1017 4
6 8 1018 2
4 6
8 1019 2
4 6
2 Neutron Fluence, n/cm, E > 1 HeV FIGURE 10.
SilELF ENERGY DEGRADATION PROJECTIONS
)
It should be pointed out that this analysis has been based on the max-imum fluence incident on the LACBWR pressure vessel I.D. surface. For the 4-in, thick pressure vessel wall, the fluence at the 1/4T position would be only 80 percent of that at the surface. Therefore, the results obtained on weld metal toughness from the 1980 capsules would be at a fluence equivalent to 10 EFPY at the 1/4T position.
Therefore, there appears to be no need to modify the vessel material surveillance capsule removal schedule at this time.
It is recommended that Capsules 5A, SB, 10A, and 10B be removed after approximately 10 EFPY.
1 O
e 42
VI.
REFERENCES
'1.
Pellini, W. S.,.and Puzak, P. P., "FractureLAnalysis Diagram Proce-dures for the Fracture-Safe Engineering Design of Steel Structures,"
NRL Report 5920, March 1963.
2.
ASTM E 208-69, " Standard Method for Conducting. Drop-Weight Test to J
Determine Nil-Ductility Transition Temperature of Ferritic Steels,"
1972 Annual Book of ASTM Standards, Part 31.
3.
Title 10, Code'of Federal Regulations, Part.50, " Licensing of Pro-duction and Utilization Facilities."
' 4..
ASME Boiler and Pressure Vessel' Code,Section III, " Nuclear Power.
Plant Components," 1974 Edition.
5.
Steele, L.
E., and Serpan, C.
Z.,
Jr., " Analysis of Reactor Vessel Radiation Effects Surveillance Programs," ASTM STP 481', December 1970.
-6.
Steele, L.10 " Neutron Irradiation Embrittlement of Reactor Pres-sure Vessel Steels," International Atomic Energy Agency, Technical Reports Series No. 163, 1975.
7.
ASME Boiler and Pressure Vessel Code,Section XI, " Rules for In-service Inspection'of Nuclear Power Plant Components," 1974 Edition.
8.
Regulatory Guide 1.99, Office of Standards Development, U.S. Nuclear.
Regulatory Commission, April 1977.
9.
ASTM E 185-79, " Standard Recommended Practice for Conducting Sur-veillance Tests for Light-Water Cooled' Nuclear Power Reactor Ves-sels," 1979 Annual Book of ASTM Standards.
10.
Norris, E.
B.,
" Analysis of the First Vessel Material Surveillance Capsule Withdrawal from Lacrosse' Boiling Water' Reactor,". Topical.
Report No.~ 1, SwRI Project 02-3467, March 23, 1973.
13.
Norris, E. B.,_" Anal, sis of the Vessel Material _ Surveillance Cap-sules Withdrawn from Lacrosse Boiling Water Reactor During the 1975 Refuelling," Final! Report, SuRI. Project 02-4074-001,' April 26, 1977.
12.
" Lacrosse Boiling Water Reactor--Reactor Vessel Material Surveil-lance Program for Evaluation of Radiation Effects," ACNP-66513, February 1966.
13.
Norris, E. B., " Tensile and Impact Properties of LACBWR Reactor Vessel Plate and Control Material," SwRI 1228-7-35, August 12, 1966.
43
i 14.
Oldfield, W.,
Wilshaw, T.
R.,
and Wullaert, R.
A., " Fracture Tough-ness Data for Ferritic Nuclear Pressure Vessel Materials; Task A Report on Statistical Analysis of Control Material Round Robin,"
Effects Technology Technical Report TR 75-39, May 1975.
15.
ASTM E 322, " Standard Method for Spectrochemical Analysis of Low Alloy Steels and Cast Irons Using an X-ray Fluorescence Spectrom-eter," 1974 Annual Book of ASTM Standards.
16.
Hawthorne, J.
R.,
et al, " Irradiation Effects on Reactor Structural Materials," Quarterly Progress Report, 1 August 1967 - 31 October 1967, NRL Memorandum Report 1833, November 15, 1967.
17.
Norris, E.
B.,
and Wylie, R.
D., " Transition Temperatures of Se-lected Materials from LACBWR Main Steam and Forced Recirculation Systems," USAEC Report SwRI 1228 P7-37, June 28, 1967.
18.
NRC Standard Review Plan, NUREG-75/087, November 24, 1975.
19.
" Lacrosse Boiling Water Reactor--Reactor Vessel Materials, Fabrica-tion and Inspection," ACNP-65534, May 10, 1965.
O e
44
O 9
i APPENDIX A SKETCHES AND DRAWINGS FROM ACNP66513 9
i i
i i
e 45
1
-l l
_Y h 0"
t v
h v
rolling ldirection weld for l drop weight n
break
\\
/
A A
~
Typical for 18 NDT drop weight plates total of 72 Charpy V-notch specimens CHARPY V-NOTCH TEST SPECIMENS MACHINED FROM BROKEN NDT DROP YEIGHT TEST SPECIMENS FROM LACBWR REACTOR VESSEL PLATE NP 1054 e
I 1Illfll 1
ll Q
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55
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e 5
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ao
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p e
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6 a
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TO t
a r
e OR r
u e
k NF q
a t
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e u
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P H e
R l
i
/
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/
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ra
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=4 3 C
l s
4 IlIl1
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[
4
/
/
/
/
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[
)
a
]>
l r
r r
r 152 Charpy - NP 1055
,A 3 1/2"
@7
+-
v 4
3 A
b E'
1/2a 132 Chcrpy - Weld b
i
~A
=
g u
13"
~
5
_,y 60 Charpy - NP 1056 save leftover NP 1056 1
ir NP 1056 9u 23" Charpy in the plate taken from upper quarter plane and lower quarter plane parallel to rolling direction as shown, with notch oriented as shown. Charpy in the weld between the two plates taken 6 per thickness, as shown.
Charpy V-rotch test specimens machined from LACBWR reactor vessel welded plates NP 1055 and NP 1056.
I WELDED PLATES NP 1055 and NP 1056
l
/
/
/
_/
/
/
/
/
NP 1057 10-1/2" 15 equal spaces b
neld ju-NP 1058
_I 15-3/4"
'i _'
17-3/4" Tensile Weld Specimens 15 tensile specimens in each quarter plane of weld (4 planes full thinness) 4 60 tensile taken,15 in each quarter plane of weld CHARPY V-NOTCH AND MINIATURE TENSILE TEST SPECIMENS MACHINED FROM LACBWR REACTOR VESSEL WELDED PLATES NP 1057 and NP 1058 I
rotal 144 Charpy from upper and K
i Iower quarterplane
\\
1/2"
\\
1/2" N~
[
Y v
V V
V V
V V
V V
V V
8 k
=
5 S
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0 5
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a.N rolling direction
\\ N f* N\\
,r g,
CHARPY V-NOTCH TEST SPECIMENS MACHINED FROM BATTELLE NORTHWEST STANDARD OR CONTROL MATERI AL (Ref. BNWL-CC-236)
216 S (REF)
/.394 1.0821.002 1.08 2t.00 2
.197 v
.x p -
i E
.394 t
- I97 l
6 1
/
A" DI A.X"B' D E E P AST TA ND ARD E-23 ONE HOLE SEE NOTES 1 & 2 MARK M/ T ERIAL
A-
' B-NO IDE N T IFIC AT IO N 001 N P.- l O 5 4 002 NP 1056 003 NP 1055 l
004 N P.1 0 5 5
.125
]
NOTE S:
- 1. E ACH SPECIMEN IS TO BE STAMPED ON SOTH ENDS INDIC A TI NG ITS RESPECTIVE MATERI AL I DE NT IF IC A T IO N.
- 2. STAMPING 15 TO BE DONE WITH LOW STRESS STAMP.
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O k
b APPENDIX B UNIRRADIATED CHARPY V-NOTCH AND TENSILE DATA e
i h
a 55
TABLE B-1 TENSILE TEST RESULTS ON HEAT NP1055 AT AMBIENT TEMPERATURE (Unirradiated)
Specimen Gage 0.2% YS UTS RA Elong.
No.
Diameter (in.)
(ksi)
(ksi)
(%)
(%)
1 0.506 64.1 85.3 70.9 26.7 2
0.505 62.8 85.8 70.8 27.5 3
0.250 64.7 87.6 72.5 30.9 4
0.250 63.6 88.1 74.9 27.7 5
0.250 64.2 87.6 72.1 27.3 6
0.250 65.2 88.6 71.3 27.6 i
TABLE B-2 a
RESULTS OF CHARPY V-NOTCH TESTS ON HEAT NP1055 (Unirradiated)
Absorbed Fracture 1.ateral Te g Energy Appearance Expansion pp)
(ft-lb)
(% Shear)
(mils) 70 120+
100 37 70 120+
100 93 40 120.0 100 99 40 120.0 100 90 10 120.0 100 92 10 97.5 60 76
-20 68.5 30 61
-20 89.5 25 73
-50 58.0 15 51
-50 41.5 10 37
-70 57.0 8
51
-70 47.0 8
42
-80 27.5 2
25
-80 38.0 2
33
-80 93 2
7
-80 23.5 2-12
-100 16.5 1
14
-100 11.5 I
11
l l
TABLE B-3 RESULTS OF CHARPY V-NOTCH TESTS ON BW REFERENCE HEAT (Unirradiated) l l
Abwehed F racture LLeral Energy Appearance Expanw m l
(it.lb)
(% Shear)
(nuls)
I 70 Mx.0 100 70 70 M7.5 100 70 40 M6.0 100 64 40 M7.0 100 59 10 74.0 95 55 10 76.0 90 65 20 47.0 30 48
- 20 59.5 35 50
-50 32.0 20 30 50 33D 25 32 60 37.0 15 27 60 30.5 iM 29 70 26 3 15 26 70 30.5 18 29 A0 14.0 10 I4 MO 19.5 A
19 K0 24.0 10 21 NO 27 3 10 24 iOO 9.5 5
9 100 15.0 5
15 TABLE B-4 RESULTS OF CHARPY V-NOTCH TESTS ON IELD METAL (Unirradiated)
Absorbed Fracture i steral k.mP-Energy Appearance Expansion O
(ft4b)
Fe Shear)
(mils) f 70 43.5 70 69 70 62.5 80 l
63 10
$ 1.5 60 54 10 47.5 65 49
-20 44 0 30 46
-20 25.5 25 26 50 22.0 20 23
-50 23.5 18 24
-80 7.5 5
9
-80 03 5
11
- 100 8.5 2
11
-100 10.5 5
7
TABLE B-5 RESULTS OF CHARPY V-NOTCH TESTS ON HEAT NP1056 (Unirradiated)
Absorbed Fracture Lateral Tm. g.
p Energy Appearance Expansion (ft.lb)
(% Shear)
(nuls) 150 53.5 95 51 150 58.0 90 52 130 54.0 90 50 130 59.0 90 55 110 57.0 90 50 110 63.5 90 54 90 -
54.0 90 45 90 56.5 80 50 70 53.5 60 St.
70 59.0 50 42 70 47.0 60 51 50 42.5 40 43 50 40.5 35 34 40 29.5 30 26 40 38.0 30 33 30 -
29.0 25 24 30 34.5 30
' 31 10 23.0 20 20 10 23.0 20 20
-10 19.5 10 15
-10 263 10 22 4
APPENDIX C DISCRETE ORDINATE TRANSPORT ANALYSIS
^
f i
t i
l
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I l
l i
59 r
[
APPENDIX C DISCRETE CRDINATE TRANSPORT ANALYSIS
- A.
Background
The LACBWR Safety Review Committee, at their meeting of October 4, 1973, questioned if the fast neutron irradiation of the vessel is pro-ceeding at a faster rate than envisioned in the design calculations given in the LACBWR Safeguards Report ACNP65544.
The committee's Recom-mendation #16 was to resolve this question of vessel NDT changes near the steam separators. This is a very important consideration because information on the neutron flux distribution is the primary link relat-ing surveillance capsule and pressure vessel material property changes.
An increase in the NDT temperature of the vessel steel may be ex-pected when the fast neutron fluence (E > 1 MeV) to which the steel is exposed exceeds a threshold value of approximately 1017 n/cm2, E > 1 MeV. Tce fast neutron flux intensity in the vicinity of the pressure vessel bocndary varies axially, radially and azimuthally. The only por-tion of the vessel which is expected to receive a fast neutron fluence above the threshold value for radiation damage during the design life is the vessel wall opposite the core, the maximum exposure generally occur-ring opposite the vertical center of the core. However, local perturba-tions in fast neutron flux within this region result from geometric as well as material differe.nces.
4 Cince it is difficult, if not impossible, to place the material surveillance capsules exactly at one or more points of maximum fast flux,-
the reactor design calculations should provide the lead factor (s) which relate the fast neutron flux at the surveillance capsule locations to the maximum fast neutron flux expected anywhere on the vessel wall I.D.
ACNP65544 indicates that the incident neutron flux on the pressure ves-sel'at the core centerline is expected to be 1.53 x 1010 n/cm2 sec > l-MeV, while that on each capsule is predicted to be 3.03 x 1010 Thus, a single acceleration factor of 1.98 is defined for all surveillance capsules.
4 The answer to DRL Question III-18 confirms that the effects of staam voids in the separator and downcomer regions were considered in calculating the neutron exposure of the pressure vessel. However, by coincidence, the estimated 15 percent increase in fast flux due to void distribution was cancelled by the 15 percent overestimate of fast flux obtained from the
- PlMG program because of the energy group structure selected.
The material surveillance capsules installed in the LACBWR vessel are located between the steam separators, but they are shielded from the core by the steam separators to varying degrees, as indicated by Figure 2 in ACNP66513, February 1966.
Therefore, all capsules are not located at the position assumed in the original reactor design calculations. For
m-example, the first two capsules removed (lA and 1B) are located about 5.5' from the true center between two steam separators. Examination of Figures 1 and 2 in ACNP66513 indicates that the vessel wall dosimeters (located outside the thermal shield) are also positioned between steam separators so that they cannot necessarily be relied upon to provide an experimental confirmation of maximum vessel wall neutron flux intensity.
It was recommended by SwRI that the original LACBWR reactor design calculation be reviewed to determine the limitations of the calculated neutron flux values contained in the LACBWR Safeguards Report. This re-view should also be directed to satisfy the recommendation in SwRI Topi-cal Report No. 1, " Analysis of the First Vessel Material Surveillance Capsule Withdrawal from Lacrosse Boiling Water Reactor," concerning the LACBUR flux spectra calculations. It was also recommended that should the original design calculations not be available, or prove to be of in-sufficient detail, the question of vessel NDT changes near the steam separators can be resolved by performing a new set of calculations using c spectral computer code such as P3MG at radial positions between the steam separators and through the steam separators.
After it was determined that the original design calculations were unavailable, a list of available computer codes was compiled. The two-dimensional Discrete Ordinates Transport code, DOT 3.5, was selected from the Radiation Shielding Information Center (RSIC) computer code collection.
A 40-group coupled neutron and gamma-ray cross section package, CASK, was also obtained from RSIC.
B.
DOT 3.5 Analysis The DOT 3.5 code was used to calculate the neutron spatial and energy distribution in the LACBWR vessel with 0 percent, 20 percent and 40 percent voids in the steam separators.
The results were compared far consistency with data obtained from the vessel material surseillance program neutron dosimeters to determine wnich level of voids was most appropriate.
In the performance of these calculations, the LACBWR vessel and internals were modeled two-dimensionally in a plane perpendicular to the vertical core axis. A one-eighth segment, with one boundary parallel to a compass point and the other 45' to the compass point, was taken to be representative be-cause of the symmetry involved. The boundaries of the core, steam separa-tors, thermal shield and pressure vessel were then described in R-9 coor-dinates, as shown in Figure C-1.
The core was subdivided into two regions, an inner region with the m
operating control rods inserted and an outer region with all control rods p
withdrawn.
Ti.e core materials within each region were homogenized over their respective areas, assuming that one-half of the shrouds were stain-less steal and the remainder were 7.ircaloy. The stainless steel steam separators, homogenized over the area they enclosed, and the stainless steel thermal shield were taken as a mixture of 18 percent Cr, 8 percent Ni and 74 percent Fe.
The pressure vessel was assumed to be 98 percent iron and the coolant as pure water.
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1.
$y, the maximum neutron flux density (E > 1 MeV) inci-dent on the vessel I.D. surface.
2.
$c, the neutron flux density (E > 1 MeV) for each sur-
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3.
3, the spectrum-averaged cross sectio'- for the 54 e(n,p)S4 n reaction at each surveillance capsule F
M location.
Then $m, the measured neutron flux density for each removed surveil-lance capsule, was calculaced using the iron dosimeter activities and the appropriate 3.
The azimuthal location of the maximum neutron flux density incident on the vessel I.D. surface, Cy, is affected by the void content in the steam separators, see Figure C-1.
With 0 percent voids, the azimuthal lo-cation of Cy is at the point on the vessel having the closest approach to the core boundary, as would be expected. The presence of voids in the steam separators, however, reduces the attenuation of neutron flux density by the coolant which in turn results in the moving of the peak flux density to a region behind the nearest steam separator as well as an overall in-crease in the calculated neutron flux density on the vessel wall.
The effect of void content in the steam separators on the lead fac-tors ($c @v) f r the surveillance capsules which have been removed is shown l
in Table C-1.
This effect is relatively small because the presence of voids in the steam separators has similar effects on both the capsule and vessel neutron flux densities.
A comparison of calculated (c ) and measured (7m) neutron flux len-e sities as a function of void content in the steam separators is presented in Table C-2 and Figure C-2.
Referring to the latter, several features are of interest:
1.
The assumption of 0 percent voids leads to completely inconsistent results between the measured and calcu-lated neutron flux densities at the locations from which capsules have been removed.
2.
There is an apparent difference in behavior between those capsules removed in 1972 and those removed in 1975. This might be explained in three ways:
(1) the average void content before the 1972 refuelling was lower than that after the 1972 refuelling; (2) the void-content varies from steam separator to steam separator; (3) the model may bs more consistent with
l t
I TABLE C-1 j
I
' NEUTRON FLUX DENSITY LEAD FACTORS AS A FUNCTION OF VOID CONTENT IN THE STEAM SEPARATORS 4
Capsule Lead Factor (a)
Identification 0% Voids 20% Voids 40% Voids
~
lA, 13, 6A,-6B 1.41 1.60-1.62 2A, 2A, 7A,-7B 2.15-2.28 2.13
-3A, 3B, 8A, 8B 1.51 1.67 l'.66 4A, 4B,'9A, 9B 1.46 1.49 1.37 5A,;5B, 10A, 10B 1.38 1.52 1.50 V.W. Dosimeter:
0.86 0.94 0.96 u
(a) &c/$y.
1 l
f I
g s t
f 1
m,
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TABLE C-2 COMPARISON OF CALCULATED AND MEASURED NEUTRON FLUX DENSITIES AS A FUNCTION OF VOID CONTENT IN Tile STEAM SEPARATORS Capsule Iron Dosimeter 0% Volds 20% Voids 40% Voids Ident.
ASAT, dps/mg 4c(a) 8(b) 4,(c) 4y(d)
&c d
im
&v
&c 4m 47 1A 7.705 x 103 4.18
.189 6.50 4.61 5.87
.183 6.71 4.19 8.48
.176 6.98 4.31 IB 6.902 x 103 4.18
.189 5.82 4.13 5.87
.183 6.01 3.76 8.48
.176 6.25 3.86 2A 1.196 x 104 6.37
.179 10.65 4.95 8.39
.174 10.96 4.81 11.16
.168 11.35 5.33 7B 1.109 x 104 6.37
.179 9.88 4.60 8.39
.174 10.16 4.46 11.16 168 10.52 4.94 9A 7.806 x 103 4.32
.198 6.28 4.30 5.47
.192 6.48 4.35 7.15
.184 6.76 4.93 3
4.32
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.192 6.56 4.40 7.15
.184 6.85 5.00 9B 7,902 x 10 3
2.54
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.160 4.45 4.73 5.00
.154 4.62 4.81 V.W.
4.466 x 10 V.W.
4.600 x 101 2.54
.168 4.36 5.07 3.46
.160 4.58 4.87 5.00
.154 4.76 4.96 Average $y, excluding 1A and IB 4.70 4.60 5.00 (a) Neutron flux density, n/cm /sec x 10-10, calculated with DOT 3.5 Code 2
1 (b) Spectrum averaged cross section, barns. for 54Fe(n.p)S4Mn from DOT 3.5 Code (c) Neutron flux densucy, n/cm /sec x 10-10, using ASAT and 6 2
(d) Maximum neutron flux density incident on vessel wall using $,and Icad factor from Table I
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(Before 1972 the center of the core was less heavily rodded and, thecafore, the power, on the average, was probably shifted to the core center and away from the vessel wall.)
3.
If it is assumed that the average void content was the same in all steam separators prior to and since the 1972 refuelling, an average value of 30 percent voids would provide for the most consistency between calcu-lated and ceasured neutron flux densities for all cap-
~
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4.
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cess of 40 percent would lead to inconsistencies sini-l lar to those noted in 1. above.
l l
C.
Su= mary i
One cannot expect to obtain perfect agreement between calculated and measured results since operations would be expected to vary and the dosin-i-
eters integrate the effect of operating variables over the exposure period.
Based on the calculations made, the conservative approach would be to as-sume an average void content of 40 percent in the analysis of the capsules removed during the 1975 refuelling outage because this leads to a higher value of vessel wall flux, as shown in Table C-2.
Assuming an incident flux of 5.0 x 1010 (E > 1 MeV) on the I.D. sur-face of the LACBWR vessel, the projected fast fluence after 20 full power years of operation is 3.2 x 1019 n/cm2 (E > 1 MeV).
This is very close to that estimated from the analysis of capsules lA and 1B (Topical Report No. 1, SwRI Project r?-3467, " Analysis of the First Vessel Material Surveillance Capsule Withdrawal from Lacrosse Boiling Water Reactor," March 23, 1973),
but more than three times that predicted in ACNP-65544.
Except for capsules 2A, 23, 7A and 73, the lead factors computed in this analysis are quite different from the value of 2 suggested in ACNP-65544. However, the neutron flux density at the vessel wall dosimeter lo-cations was computed to be nearly equal to the maximum value incident on the pressure vessel wall as planned.
The variation in flux density at the capsule locations predicted with the DOT 3.5 code are supported by the variations in activities of the neutron dosimeters contained in the sur-veillance capsules removed to date.
O J
l l
l APPENDLX D TANH-FIT CHARPY CURVES (CAPSULES 3A, 3B, 8A, AND 8B) d e
69
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