ML20010H309
| ML20010H309 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/18/1981 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20010H308 | List: |
| References | |
| NUDOCS 8109240326 | |
| Download: ML20010H309 (14) | |
Text
TABLE 4-1 HYDROSTATIC TEST PRESSURE / TEMPERATURE LIMIIS FLANGE BELTLIN E PRESSURE LIMITS PRESSURE LIMIT l4)
BELTLI N E TEM PERATU RE TEMP.
RATE CHANGE 0{F/HR)
(
50'F 100*F 150 F 200'F 250'F
< 120'F 0
848 950 1162 1600 300 20 805 904 1108 1529 40 762 858 1054 1858 2294 60 721 812 1001 1390 2194 80 678 766 948 1322 2096 100 637 721 896 1256 1996 HEATUp PRESSURE /TEMPERATURELIMITS FLANGE BELTLINE PRESSURE LIMITS PRESSURE LIMIT (4)
BELTLINE TEM PERATU RE (SEE NOTE 3 WHEN REACTOR IS CRITICAL)
TEM P.
RATE OF 0{F/HR)
CHANGE
(
50*F 100*F 150 F 200*F 250*F
< 150'F 0
632 709 868 1196 1875 300 20 598 670 818 1124 1755 40 56 4 631 769 1054 1642 60 531 593 722 987 1534 80 498 556 675 922 1832 100 465 519 631 860 1335 i
C00LDOWN PRESSURE / TEMPERATURE LlHITS 1
i FLAN GE l
BELTLINE PRESSURE LIMITS PRESSURE LIMIT (4)
BELTLINE TEM PERATU RE (SEE NOTE 3 WHEN REACTOR IS CRITICAi.)
RATE OF CgANGE i F/HR) 0 50'F 100 F 150'F 200*F 250*F
< 150'F
(
0 632 709 868 1196 1875 300 20 600 674 827 1153 1795 40 568 640 787 1090 1717 60 537 605 747 1039 1642 80 505 571 707 988 1568 100 474 537 668 938 1495 CRITICALITY TEMPERATURE LIMITS: THE MINIMUM TEMPERATURE AT WHICH A CRITICALITY (EXCEPT FOR LOW POWER PHYSICS TESTING) CAN TAKE PLACE IS 90*F NOTES I
- 1. DURING A NS$$ HYDROSTATIC TEST THE PRIMARY SYSTEM WATER TEMPERATURE CAN EXCEED 212 F WITHOUT ENTERING THE TECHNICAL SPECIFICATIOR POWER OPERATION DEFINITION FOUND IN SECTION 1.2.1.
l HYDROSTATIC TESTS ARE PERMISSIBLE ONLY WITH THE REACTOR Su8 CRITICAL.
- 3. TABLES ARE APPLICABLE FOR A FLUENCE LEVEL UP TO 4.5 X 10 ' NEUTRONS /CM g, g g,,y 1
2
- 3. HEATUP/LOOLDOWN PRESSURE LIMITS ARE APPLICABLE F TEMPERATURES 40'F HIGHER WHiN THE REACTOR IS CRITICAL.
4 FLANGE PRESSURE LIMITS ARE NOT DEPENDENT UPON THE TEMPERATURE RATE OF CHANGE.
- 5. ALL PRESSURE LIMITS ARE IN P984 I
8109240326 810918 l
PDR ADOCN 05000 l
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1 APPENDIX A - Big Rock Point Pressure / Temperature Operating Limits -
General Considerations SPECIFIC ASSUMPTIONS Certain fundamental assumptions must be made upon which operating limits' definitions are based. A 0*F unirradiated reference temperature for Big Rock Point reactor vessel beltline weld metal is assumed. This assumption ensures that the wcld metal will always be controlling, Thus, O'F is employed. The reference to temperature in the unirradiated condition for the SA-336 flange material is assumed to be 30*F.
This assumption is based upon recommendations contained in the Branch Technical Position associated with the US Nuclear Regulatory Commission Standard Review Plan 5.3.2.(5)
It is understood that specification requirements for the material required be at least 30 ft/lb of Charpy energy at 10*F.
It is assumed that the SA-336 is similar enough to SA-508 for the Branch Technical Position to apply.
It is also assumed that the methodology of Appendix G of the ASME Code applies. The Code indicates that its methodology applies to materials of yield strength of 50 KSI or less. Section 4.1.5.3 of the Big Rock Final Hazards Summary Report indicates that the vessel plate has a yield strength of 56.9 KSI. A review of additional data in Amendment 8 of the FHSR indicates that lower values for shell course and weld material may be more appropriate. With no additional considerations, it is judged that for all reasonable intents, the plate, FHSR and weld material reflect characteristics which the 50 KSI guideline is set up to ensure.
Other assumptions of note are that the vessel beltline materials are subject to membrar n stresses only and that the vessel flanges see no effective in-crease in the reference temperature due to irradiation.
SPECIFIC DATA A.
GEOMETRIC AND OPERATING DATA Re.ctor Vessel ID 106 Inches Beltline Wall Thickness 5.25 Inches Operating Pressure 1,350 Psig l
Pressure Measurement Error i 15 Psi l
Temperature Measurement Error i 5'F B.
FLUENCE / REFERENCE TEMPERATURE SHIFT Table 1 is taken from WCAP-9794. In view of the fact that the unirradi-ated reference temperature for the weld metal is taken as O'F, the weld 8
metal is controlling.
It has been determined that the fluence (n/cm >
i 1 MeV) at the 1/4 T vessel wall location will be 69 percent that of' the l
fluence at the vessel inner diameter and the fluence at the 3/4 T location i
will be 27 percent of that at the vessel inner diameter.(6) It is postulated that the revised curves should be applicable until the maximum nu0981-0185a-43 O
a 2
fluence at the vessel' inner diameter is 4.5 x 10" n/cm 2 > 1 MeV.
It becomes clear from Table 1 that near fluences of 3.11 and 1.22 x 10" n/cm* (the associated 1/4 T and 3/4 T fluences), the fluence reference temperature shift is a very flat 135'F as determined by the evaluation of three surveillance capsules. Therefore, heatup/cooldown/ inservice leak test limits will be drawn for an inner vessel wall fluence of 4.5 x 10"-
n/ca' > 1 MeV and for a reference temperature shift of 135*F which for the controlling weld metal implies an irradiated reference temperature of 135'F at the 1/4 T and 3/4 T locations.
C.
THERMAL GRADIENTS DUE TO HEATUP/C00LDOWN RATES The thermal gradients for the Big Rock Point reactor vessel due to uniform heatup and cooldown rates are shown in Table 2.
It was determined that temperature independent thermal conductivity and diffusivity properties resulted in insignificant differences from temperature dependent proper-ties. Therefore, temperature independent properties were employed which, in turn, imply that Table 2 applies for either heatup or cooldown tempera-ture change rates.(7)
D.
VESSEL FLANGE LIMITATIONS Paragraph IV.A.2 of 10 CFR 50, Appendix G, requires that, when operating pressure exceeds 20 percent of the preservice hydrostatic test pressure, the temperature of the stressed regions of nozzles, flanges and other areas of structural discontinuity shall be at least 150'F above the rsference temperature of the material of concern.
' aragraph IV. A.3 of the same reference says that when the water level is eithin normal range for power operation and the pressure is less than 20 percent of the preservice system hydrostatic pressure, the minimum permissible temperature is 60'F above the reference temperature of the material associated with closure flange region highly stressed by bolt preload. This criteria is when the core is critical.
l l
It is further noted that Paragraph IV.A.3 requires that except for l
criticality below 20 percent preservice system hydrostatic pressure (above l
paragraph), the allowable temperature at a given pressure with the core l
critical will be at least 40'F higher than with the core noncritical l
(Paragraph IV.A.2 of 10 CFR 50, Appendix G).
l l
It is expected that the 150'F above the reference temperature requirement will be changed to.120*F above the reference temperature for normal operations and 90*F above the reference temperature for inservice leak and hydrotesting. The unirradiated flange area material is assumed to be controlling and a reference temperature for the SA-336 material is taken as 30*F.
Twenty percent of preservice hydrotest is taken as 300 psig.
nu0981-0185a-43
,.--_.. - _ _ -, _ _ _ _ ~ _. _. _ _ _ _ _ _ - _ _. _ _. _ - _ - _ - -.
l
3 TABLE 1 REFERENCE TEMPERATUPE SHIFTS i
Fluence 30 Ft/Lb Material Type Capsule Type
> 1 MeV x 10" n/cm Trans Temp Shift Base Metal Wall 119
.15 0
Wall 127
.71 60 Wall 125 2.27 120 Shield 122 2.30 70 Shield 124 10.70 150 Weld Metal Wall 119
.15 55 Wall 127
.71 135 Wall 125 2.27 135 Shield 122 2.30 130 Shield 124 10.70 170 Heat Affected Wall 119
.15 0
Zone Wall 127
.71 25 Wall 125 2.27 110 Shield 122 2.30 80 Shield 124 10.70 160 i
l 9
4 l
l i
nu0981-0185a-43 i
m t
4 h
TABLE 2 HEATUP/bOOLDOWNTEMPERATUREGRADIENTS 1
Temperature Change Rate AT AT AT,,x 774 3/4
- F/Hr
- F
- F
- F 0
0 0
0 20 2.71 4.92 5.19 40 5.42 9.84 10.39 60 8.12 14.76 15.58 80 10.83 19.68 20.77 100 13.54 24.60 25.96 AT
= Absolute value of temperature difference between inside of reactor 1/4 vessel wall and a point one-quarter of the way through the vessel wall.
AT
= Absolute value of temperature difference between inside of reactor 3/4 vessel wall and a point three-quarters of the way tLrough the vessel.
wall.
T,
= Absolute value of temperature difference between inside of reactor vessel wall and a point on the vessel outer diameter.
l f
1 i
l l
i N
n l
l nu0981-0185a-43
5 APFENDIX B - Big Rock Point Pressure / Temperature Operating Limits - References 1.
10 CFR Part 50, Appendix G, " Fracture Toughness Requirements" This appendix is now being subjected to extensive revision to apply to BWRs. Drafts of this appendix have been subject to extensive comment.
The appendix provides the following:
A.
Definition of terms such as reference temperature, beltline, etc, by reference to other documents.
B.
Material testing requirements.
C.
Mandatory imposition of ASME Code, Appendix G, methodology and corre-lations in deterrining allowable operating limits.
D.
Operating temperature restrictions which may be controlling even with respect to ASME Code, Appendix G, calculated pressure / temperature limits. The flange and nozzle criteria are contained in Appendix G of 10 CFR 50.
2.
ASME Boiler and Pressure Vessel Code,Section III, Subsection NB 2331,
" Test Requirements and Acceptance Standards," 1980 Ed Subsection NB 2331 defines exactly what a " reference temperature" is.
The definition reflects a need for both drop weight and Charpy testing.
3.
S E Yanichko, S L Anderson, R P Shagan and R G Lott, " Analysis of Capsule 125 From the Consumers Power Company Big Rock Point Nuclear Plant Reactor Vessel Radiation Surveillance Program (WCAP-9794)," Pittsburgh, PA, Westinghouse Electric Corporation, September 1980. Topical Report on EPRI Research Project 1021-3 This report provides the analysis results for dosimetry and material
(
properties.
It tabulates the results from the 125 capsule along with the results from previous capsules to provide trend curves for the prediction of radiation damage throughout the vessel life.
l 4.
C Z Serpan and H E Watson, " Mechanical Property and Neutron Spectral 1
Analysis of the Big Rock Point Reactor Vessel," Nuclear Engineering and
(
Design, Volume II, No 3, April 1970 This report contained all of the surveillance capsule work done on Big Rock Point before the 125 capsule.
5.
US Atomic Energy Commission Regulatory Standard Review Plan, Directorate l
of Licensing, Section 5.3.2, " Pressure-Temperature Limits" This document offers a very methodical approach for drawing pressure-temperature limits. Most of the important curves from the ASME Code, nu0981-0185a-43 A
+--
t 6
f 1
Appendix G, are reproduced and sample pressure-temperature limits for a
" typical" PWR plant are provided. Associated with this Standard Review Plan is a Branch Technical Position paper. This paper provides guidance for determining reference temperatures and other material properties for older plants where a full complement of drop weight and Charpy test samples does not exist. Because of a lack of drop weight specimens and somewhat limited Charpy data, Big Rock Point is obliged to default to some of the criteria posed in the Branch Technical Position paper.
6.
Letter from S L Anderson (Westinghouse Electric Corporation) to R B Jenkins (Consumers Power Company), " Relative Variation of Fast Neutron (E > 1.0 MeV) Flux Within the Big Rock Point Pressure Vessel," Dated January 22, 1981 This letter provides the fluence attenuation factors used for the reactor vessel wall. These factors will result in the determination of reference temperature shifts associated with the various wall locations.
7.
P J Rashid, " Analysis: Big Rock Point Reactor Vessel Thermal Transient,"
Appendix to Letter From R B Jenkins to D P Blanchard (JENK 41-77) Dated September 23, 1977 This analysis provides the thermal gradients through the vessel wall at various uniform heatup and cooldown rates. The temperature difference provides an indicator of thermal stress.
e nu0981-0185a-43
7 APPENDIX C - Big Rock Point Pressure / Temperature Operating Limits -
ASME Code,Section III, Appendix G, Format GENERAL Pressure / temperature limits are based upon the ASME Code,Section III, Appendix G, methodology. Appendix G is stated in the ASME Code to be-I nonmandatory. However, the Standard Review Plan uses it.and 10 CFR 50, Appendix G, invokes it as, law. This appendix provides the results of im-provements in the state of the art in linear elastic fracture mechanics as far as pressure vessels are concerned.
It provides the following:
1.
An empirical data fit which correlates a reactor vessel material fracture toughness with operating temperature ano the reference temperature. This fracture toughness is a material property.
2.
A postulated material defect in the vessel waM This defect is postu-4 lated to be a sharp crack of a depth one-quarter of the reactor vessel wall thickness.
3.
Correlation factors to transform stresses into stress intensities. These factors convert pressure stress to a pressure stress intensity and thermal stress to a thermal stress intensity. The correlation factor for pressure is based upon wall thickness and stress level and that for temperature is based upon wall thickness and the assumed crack size.
- 4.
Additional safety factors to be applied to the pressure membrane stress intensity before it is added to the thermal gradient stress intensity for comparison with the vessel material fracture toughness. An additional safety factor of 2.0 is applied to pressure stress during normal opera-tions. That factor is reduced to 1.5 for leak testing.
NOMENCLATURE
(
T
= Temperature at crack location of concern
'F.
t T,
= Temperature at reactor vessel inner diameter at the vessel beltline
- F.
AT
= Temperature difference between reactor vessel wall inner diameter and gjg the one-quarter wall location
- F.
AT
= Temperature difference between reactor vessel wall inner diameter and 374 the three-quarter wall location
- F.
j l
I AT
= Total temperature difference across vessel wall
'F.
R
= Mean radius of reactor vessel at beltline - 55.63 in.
t
= Vessel wall thickness at beltline - 5.25 in.
nu0981-0185a-43 i
,.---,--,--.--,.---~,~.--,.,~.a,--
.~. - -
~
8 RT
= Reference temperaturc (nil ductility transition) - 135'F for case of ndt interest.
P
= Er.isting pressure psig.
P,
= Allowable pressure psig.
M,
= Membrane stress intensity factor.
M
= Thermal gradient stress intensity factor.
g Ib K
= Reference stress intensity factor - InY /In.
ir K,
= Stress intensity factor associated with pressure membrane stress -
g IbP 6.
K
= Stress intensity factor associated with thermal gradient stress -
it Ib in'
/in.
P,
= Pressure measurement error ~ assume 15 psi.
T,
= Temperature measurement error - assume 5*F.
COMPUTATION'sL PROCEDURE The pressure / temperature limit computational procedure is set up to arrive at an allowable pressure given a measured temperature and heatup or cooldown rate. Pressure / temperature limits for inservice testing are based upon scaling heatup curves. These allowable pressures are based strictly on beltline considerations. These beltline considerations are then compare'd with nozzle and flange region considerations where stresses are high and fluences negligible.
The beltline heatup and cooldown curves are arrived at through the following process:
1.
Select a heatup or cooldown temperature change rate.
2.
Calculate the reference stress intensity factors for various tempe ctures associated with this heatup or cooldown rate.
3.
Calculdte the pressure which ensu.a. i e equality of the sum of the pres-sure (with safety factor) and thermal gradient stress intensities with the reference stress intensity.
nu0981-0185a-43 A+-
w m
I 9
4.
The result is the allowable pressure at that temperature for that heatup or cooldown rate.
5.
Changing the heatup or cooldown rate, repeat the above steps.
nu0981-0185a-43
~
10 APPENDIX D - Big Rock Point Pressure / Temperature Operating Limits -
Evaluation EQUATION DEVELOPMENT The yield strength of the SA 302B steel at 650*F is not less than 37.3 KSI.
The maximum design pressure for the reactor vessel is 1,715 psi. Therefore, the maximum pressure membrane stress is:
PR = (1,715) (55.63) = 18.17 KSI <.5 (37.3) KSI t
5.25 therefore, per Fig G-2214-1 and M,= 2.2 per Fig G-2214-2 M =.26 The expression for K 18 Written Per Fig G-2210-1 as:
ir K
= 26.78 + 1.223 EXP [.0145 (T-RT
+ 160)] KSI 6 y
ndt The expression for the stress intensity factor balance is:
K 2 2K, + Kgg (Eq 1) ir y
i j
2 2Mm_PR + M AT t
or Pa = [E
~H T,
][t/2MmR]
h' t
f ndt* "t' M and the inclusion of error terms The substitution of values for RT m
results in the following expressions for K and the allowable pressure:
ir K
= 26.78 + 1.634 EXP [.0145T] KSI /55 (Eq 2) ir
)
and l
l Pa = 21.45 K
- 5.58AT
- 15 Psi (Eq 3) g I
For heatup, the "T" in Eq 2 is the measured temperature implied to exist at the beltline minus AT For cooldown, the "T" in Eq 2 is the measured j
374 temperature plus AT Were AT1/4, AT3/4 and AT all taken at the given gj4 nu0981-0185a-43 l
11 temperature change rates per Table 2 of Appendix A.
For inservice leak testing, the factor safety of 2 in Eq 1 is replaced by a 1.5.
Since it is assumed that such testing is done only during heatup, the heatup results can be scaled.
BELTLINE LIMITED-PRESSURE / TEMPERATURE LIMITS The heatup, cooldown and inservice test pressure /remperature limits are provided in Tables 3, 4 and 5, respectively. Before these limits can be incorporated into curves employed as operating limits, additional considera-tions need to be considered. The flange-controlled limits posed by 10 CFR 50, Appendix G, must be imposed. Similarly, the critical vs noncritical ILaitations must also be imposed.
It is also noted that there are curves for various heatup and cooldown rates.
It is not practical to consider all temperature change rates. For simplicity, the following considerations are assumed:
1.
A heatup rote of 40*F per hour is all that is achievable with the core noncritical working from pump heat. Therefore, for a noncritical condition, two heatup rates O'F and 40'F per hour are offered for noncritical and O'F and 100*F for critical.
2.
It is not known what cooldown rates are possible.
It is assumed that O'F and 100*F per hour can be provided for consideration and interpolation allowed between them.
3.
It is assumed that inservice testing will be done only after heatup and that the test will be conducted after a very low heatup rate. Therefore, an inservice test curve will be drawn assuming a hypothetical " isothermal heatup" rate of 0*F per hour.
I l
l I
nu0981-0185a-43
12 TABLE 3 HEATUP PRESSURE / TEMPERATURE LIMITS BELTLINE CONTROLLED Temp Change Allowable Pressure at T Beltline - Psig.
Rate "F/Hr 50 100 150 200 250 0
632 709 868 1,196 1,875 20 598 670 818 1,124 1,755 40 564 631 769 1,054 1,642 60 531 593 722 987 1,534 80 498 556 675 922 1,432 100 465 519 631 860 1,335 TABLE 4 COOLDOWN PRESSURE / TEMPERATURE LIMITS BELTTINE CONTROLLED Temp Change Allowable Pressure at T Beltline - Psir Rate "F/Hr 50 100 150 200 250 0
632 709 868 1,196 1,875 20 600 674 827 1,143 1,795 40 568 640 787 1,090 1,717 60
' 537 605 747 1,039 1,642 80 505 571 707 988 1,568 100 474 537 668 938 1,495 1
nu0981-0185a-43
\\
7 13 TABLE 5 INSERVICE TEST PRESSURE / TEMPERATURE LIMITS BELTLINE CONTROLLED Temp Change Allowable Pressure at T Beltline - Psia Rate
'F/Hr 50 100 150 200 250 i
0 848 950 1,162 1,600 20 305 904 1,108 1,529 40 762 858 1,054 1,458 2,294 j
60 721 812 1,001 1,390 2,194 80 678 766 948 1,322 2,096 l
100 637 721 896 1,256 1,996 l
t I
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