ML19221B089
| ML19221B089 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| TASK-TF, TASK-TMR NUREG-75-087, NUREG-75-087-05.3.2, NUREG-75-87, NUREG-75-87-5.3.2, SRP-05.03.02, SRP-5.03.02, NUDOCS 7907120403 | |
| Download: ML19221B089 (18) | |
Text
NUREG 75/087 fgR R e c y,%
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U.S. NUCLEAR REGULATORY COMMISSION hh' #) STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTION 5.3.2 PRESSURE-TEMFERATURE LIMITS REVIEW RESPONSIBILITIF Primary - Materials Engineering Branch (MTEB)
Secondary - None I.
AREAS OF REVIEW l.
Pressure-Trperature Limits The following pressure-temperature limits imposed on the reactor coolant pressure boundary (RCPB) during operation and tests are reviewed to assure that they provide adequate safety margins against nonductile behavicr or rapidly propagating failure of ferritic components of the RCPB, as required by General Cesign Criterion No. 31:
Pressure-tem;erature limits for preservice hydrostatic tests.
a.
b.
Pressure-terperature limits for inservice leak and hydrostatic tests.
c.
Pressure-terperature limits for heatup and cooldown operations.
d.
Pressure-te perature liriits f ar core operation.
II.
ACCEPTANCE CRITERIA 1.
Applicable Re,ulations, Codes, and Basis Documents Appendices G and H of 10 CFR Part 50 descrit:e the conditions that require pressure-temperature limits and pro /ide the general basis for these limits. These appendices specifically require that pressure-te perature limits must provide safety margins at least as great as those recommended in the ASVE Boiler and Pressure Vessel Code (hereaf ter "the Code")Section III, Appendix G, '7rotection Against Non-ductile Failure," during heatup, cooldown, and test conditions. Appendix G to 10 CFR Part 50 also requires additional safety margins whenever the reactor core is critical (except for low-level physics tests).
The Code,Section III, specifies fracture toughness testing recuirements for ferritic raterials. Appendix G of the Code provides a basis for determining allowable pressure-temperature relationships for normal, upset, and test conditions Welding Research Council (WRC) Bulletin 175, 'PVRC Recommerdation on Toughness Require-ents for Ferritic Materials," provides the detailed technical basis for the code requirements.
d USNRC STAND ARD REVIEW PLAN Standard rew**se p6ene are propered for the guidence of the OMice of Nucteer Reector Raouwi >o sterf responobio for the review of applications to coastruct end operate neceoer power plente These documente ere made ovestable to the htW se part of tt6e Commeseeon o poncy to inform the nucseer industry and the generes pubhc of regulatory precedures and posec e Standard review eN..e are not subetstutee for requietory guideo or the Commise.on a requistione and compsience en% theen.e not required The standard.ecew piec sect.one are keyed to Revie.on 2 of the Stenderd Formet and Content of Sciety An pysie Reporte for Nuc6eee Power Ptente Not est sections of the Stones.e Formet here e correspondmg review pian Pubsiehed standeed recew piene wtN be revised periodicativ es oppropriate to eccornmodete commente end to rettect new info.metion and emperient.e Commente and suggestione for improvernent will be cons.dered end abound be ser:t to the U $ Nu ear Regulatory Commeos.o10ft6ce of Nucteer Reacter Regunet.on weehengton D C 20MB r-3 a,/
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2.
, Technical Bases a.
The principles of linear elastic fracture mechanics (LEFM) are used to determine safe operational conditions. The basic parameter of LEFM is th? stress intensity factor, K, which is a function of the stress state and flaw configuration. An g
analytical method is used to detemine the effects of real or postulated flaws.
The minimun K that can cause failure is defined as the critical stress intensity y
factor, Kk, and is the material property us<>d in this rethod. The K of the k
raterial is either directly measured as a function of terperature, or is conser-vatively estimated, using information from other fracture toughness tests.
b.
The Code specifies the maximum Kg, as a function of temperature, that can be assured for the specific material, based on results of tests on the raterial used.
This value is called KIR, reference stress intensity factor. The Code also pro-vides rules for calculating the K. including definitions of postulated flaws, and specifies the safety factors to be applied. The acceptance criterion is that the K of the raterial must always be higher than the K; calculated.
IR c.
Direct measurement of the K as a function of temperature is expensive and tire k
consuming and requires more sample material than is usually available. Ccerela-tions between the K determined directly and results of simpler fracture tough-ness tests are not exact, but may be used if appropriate allowances are made for variations in material behavior and data scatter. The Code gives ;alues of K IR as a function of terperature relative to a conservative determination of the nil-ductility transition temperature (tiDTT) of the raterial. This reference temper-ature, RT DT, is determined for the ferritic raterials of components for which t
operating and testing limit curves must be calculated. The effects of radiation on the fracture toughness of the material in the beltline region of the reactor vessel is accounted for by adjusting the RT;DT of the affected material upward.
f The amount of upward shift depends on the composition of the steel (especially its copper and phosphorous content), the neutron fluence, and the temperature of irradiation. Conservative predictions of the effect of radiation on the RT ;DT f
based on data in the literature are factored into the original limit curves. The continued conservatism of these predictions throughout plant life is verified by a mandatory material surveillance program described in Appendix H to 10 CFR Part 50.
d.
The Code specifies the stress corporents that rust be used for the K; calcula-tions, and the factors that rust be applied to each to provide adequate safety rargins. The Code, by reference to WRC-175, specifies the expression to use for calculating the K, using the applied stresses and the postulated flaw georetry.
g Although calculations are usually made by a computer, curves are provided in the Code to facilitate the use of conservative hand calculations if desired.
3.
Pressure-Temperature Requirements The requirements for the pressure-terperature limits are as follows:
a.
Pressure-Temperature Linits for Preservice Hydrostatic Tests During preservice hydrostatic tests (if fuel is not in the vessel), the Kyp must be greater than the K caused by pressure. The expression used is:
y 5.3.2-2 147 106
K = K (pressure) < K g
g IR b.
Pressure-TeTperature Limits for Inservice Leak and Hy1rostatic Tests During performanc.
,,f inservice leak and hydrostatic tests, the K must be gg greater than 1.5 times the K caused by pressure. The expression used is:
g K = 1.5 K (pre sure) < K g
g IR c.
Pressure-Temperature Limits for Heatup and Cooldown Onerations At all tires during heatup and cooldown operations, the K must be greater than gp the sum of 2 times the K caused by pressure and the K caused by thermal g
g gradients. The expression used is:
2K (pressure) + K (therral)
<K K
=
g g
g IR d.
Pressure-Temperature Limits for Core Operation At all times that the reactor core is critical (except for low power physics tests) the temperature must be higher than that required for inservict hydro-static testing, and in addi n, the pressure-terrperature relationship shall provide at least a 40 F ma-
)ver that required for heatup ard cooldown operations.
III. REVIEW PROCEDURES The reviewer will select and errphasize material from the procedures described below, as nay be appropriate for a particular case.
1.
Preliminary Safety Analysis Report (PSAR)
Actual operating limit curves cannot be deternined at the PSAR stage, because the fracture toughness and other required tests have not been perfonted on the actual material that will be used. Typical curves, with temperatures shown relative to the RTNDT, and the basis for determining the curves are reviewed and corpared with the acceptance criteria described in II above.
2.
Final Safety Analysis Report (FSAR1 The limits in the plant Tecnnical Specifications will be shown using real tempera-ture.
These curves and their bases are reviewed to determine acceptability ir the following areas; a.
The limiting RT has t:een properly deternined, and radiation ef fects are NDT included in a conservative manner.
b.
Limits are shown for all required conditions, c.
The linits proposed are consistent with the acceptance criteria descrit'ed in II above, d.
The procedures for updating the limit curves, in conjunction with schedu' tests on material surveillance specimens, are well defined and included in the Technical Specifications.
3.
Acceptability Determination Methods kk The reviewer evaluates each limit curve for acceptability by perforning check cal-cu.ations using the simplified methods referenced in the Code and WRC Bulletin 175 that have been verified by the Materials Er.gineering Branch to yield conservative 5.3.2-3
values. These methods are described in detail by exurples below, and the curves ner.essary to perfom the calculations are included herein as Figures 1, 2 and 3.
Figure 4 is an example of acceptable limit curves developed by this method.
a.
Preserv.ce Hydrostatic Tests The preservice hydrotest at 1.25 design pressure corresponds to the standard Code component hydrotest usually perforred in the shop, but in this case it is the hydrotest for field welds, so it may involve the entire reactor coolant system.
The Code reconmends that component hydrostatic tests be run at a temperature no lower than RTNDT + 60 F, but also recorrends that system tests should have more stringent requirements. The MTEB position is that the minimum temperature for the preservice test, if fuel is not in the vessel, be detemined using the methods of Code Section III, Appendix G, using less stringent factors.
First, the RT f the vessel material must be determined. This is defined by f40T the Code for new plants, and is essentially a corservative value of the NDTT as determined by drop weight test. Guidelines for estimating the RT If th*
NDT prescribed tests have not been run are covered by Branch Technical Position - MTEB No. 5-2, " Fracture Toughness Requirements."
The reference temperature and the toughness of the material at any temperature is a function of the differencc between the RT of the material and the temperature NDT of interest. The Code provides a curve (Fiqure G-2110.l) for the allowable calculated stress intonsity factor (KIR) as a function of the temperature relative to RT NDT' The Ccde also provides a recomended basis for calculating K, including recom-g mendations for assumed flaw size and shape, and appr priate front and back surface correction factors. Because the assumed flaw size is proportional to the wall thickness, t (flaw depth = 0.25 t and lengtn = 1.5 t), the K expressions are g
simplified to multiplers that are a function only of wall thickness and stress level. These factors, M for membrane stresses and M f r bending stresses, are B
provided in graphical form in Figure G-2114.1.
The criterion recorrended by ffTEB can be expressed as for the shell region.
Kg<KIR To get K, the stress level and wall thickness must be known. The pressure for g
the hydrostatic test is 1.25 times the design pressure, so either of two simple methods can be used to approximate the membrane stress accurately enough for this purpose:
stress = 1.25 times the Code allowable (S )
m stress =
where P is the test pressure and r is the vessel radius. As an example, assume a vessel with a design pressure of 2500 psig, made of steel with an S m
147 108 5.3.2-4
of 26,700 psi, and a ninimum yield strength of 50,000 psi. The stress for the 9
preservice hydrotest is then 26,700 x 1.25 = 33,400 psi, or (1.25) (2500) (951, 33,400 psi, for a vessel with a radius of 95 inches and a wall g
thickness of 9 inches.
The next step to determine the factor to apply to this stress to obtain K;.
Figure G-2114.1 (reproduced here as Fig. 1) provides several curves, depending on the ratio of the stress level to the yield strength of the material. In this case, the stress level is 33,400; the yield strength is conservatively assured to be 50,000 so the curve for a ratio of.7 should be used. (A ratio equal to or higher than the actual ratio must be used for conservatism.) For a 9-in.
thick vessel ( d = 3), the value of M fron Figure G-2114.1 is 2.94.
The K for g
this case is then:
K = (M ) (Membrane Stress) y K = (2.94) (33,400) = 98,300 p s i, i ri.
7 From Figure G-2110.1 (reproduced here as Fig. 2), a temperature of at least RT NDT
+ 120"F is necessary for a K of this level.
y If, for example, an criginal RT f 40*F is assumed, the required temperature NDT is then 40 + 120, or 160"F.
b.
Inservice Leak and Hydrotest The temperatures for the inservice leak and hydrotest, performed at operating pressure and about 1.1 operating pressure, respectively, are calculated in essentially the sare way.
The differences are that a factor of 1.6 must be applied to the calculated K to provide extra margin, and the stress levels are y
lower, so the value of M is taken from a lower ratio curve.
Using the same vessel as an exa: ole, with a normal operating pressure (P ) of g
2250 psi, the membrane stress for the leak test can be approximated as:
operating pressure x allowable stress design pressure or x 26,70G = 24,000 psi This is about half of the minimum yield strength, so the M is taken from the 0.5 ratio curve, and is 2.87.
The calculated K that must be assured is then:
y K = (1.5) (M ) (Membrane Stress) g or K = (1.5) (2.87) (24,000) = 103,500 psi /1T 7
From the K curve, a temperature of about RTNDT + 125 F is required. As this IR is an inservice test, the RT would probably have been increased from its NDT original value of + 40 F by some shift caused by radiation. Assume this shift is 100 F, thus the temperature for the leak test must be at least:
40 + 100 + 125 = 265'F The inservice hydrotest temperature (at 1.1 P ) is determined in exactly the g
se e way, and requires a minirum terperature of about RTNDT + 13 N, or 2731.
147 109 5.3.2-5
Heatup, Cooldown, and Nornal Operation For nomal operation, which includes upset conditions and startup and shutdown procedurcs, operating limit curves must be provided that show the maximum per-missible pressure at any terperature from cold shutdown conditions to full pressurization conditions.
Reactor vendors have developer 1 computer codes to perform the necessary calcula-tions, because thermal stresses must be included, and hand calculations of even moderate sophistication are very time consuming. WRC Bulletin 175 includes a set of curves derived from computer programs that can be used to approximate the K caused by therr:al stresses, as a function of wall thickness and rate of temperature change. Pressure-temperature curves developed using these approxi-mations agree fairly well with those determined using much more rigorous pro-cedures, and can be used with confidence to evaluate the proposed operating limits given in Technical Specifications. These curves require the calculation of only 3 to 5 points. Either allowable pressure at a given tercerature, or allowable temperature at a given pressure can be calculated. It is usually more convenient to calculate allowable minimum temperature, so this rrethod will be used in the example.
Using the same reactor vessel as in the previous example, and a rate of terrperature change of 50'F per hour, calculations of required temperatures for several pressures are illustrated..The curves for thermal effects given in WRC Bulletin 175 are very conservative, thus no additional margin need be applied to the K fron thernal stress, but a factor of 2.0 is used on primary g
stresses. The basic expression is then:
? K (membrane) + K; (themal)
K 1
I IR K (membrane) is calculated exactly as in the previous exemples. K (thermal) y 7
for a 9-in. thick wall, at 50 /hr is about 12,000 psi /Iri from Figure 4-5, WRC Bulletin 175 (reproduced here as Fig. 3).
Thus, for a pressure of 2250 psig, a membrane stress of 24,000 psi, and M of 2.87, the basic expression is given by KIR > (2)(24,000)(2.87) + 12,000 = 150,000 psi /Tni-From the K curve, a temperature of RTNDT + 158 F is required. With an IR RT f 140 F, the temperature required for operating pressure at a heatup or NDT cooldown rate of 50 /hr is then 140 + ISS = 298 F For a pressure of 1/2 of operating (1125 psig), the membrane stress is 1/2 of that at operating pressure, or 12,000 psi.
The M can be taken from the 0.5 "- ratio curve in Figure 2114.1, so is again m
cy 2.87.
K 1 (2)(12,000)(2.87) + 12,000 = 81,000 psi /Iri-
~
IR From the K curve, the mir,; mum temperature is RTNDT + 100 F, or 140 + 100 =
yp 240 F.
5.3.2-6 147 110
The same calculation for a pressure of 1/5 operating pressure (450 psig and 4800 psi stress) is similar, but in this case the stress is less than.1 of the yield strength, so the M (from the.1 ratio curve) is only 2.82.
KIR > (2)(4800)(2.82) + 12,000 = 39,000 psi /iri-The K curve shows tnat the minimum temperature is RTNDT + 0*F, or 140 F.
7p Three noints on a 50'/hr operating limit curve for this vessel at this tire in its service lifetime have thus been calculated:
Pressure Min. Temperature (psiqL (Fahrenheit) 450 140 1150 240 2250 298 A smooth curve drawn through these points will very closely approximate the results using more rigorous rrethods.
d.
Core Operation Apper. dix G,10 CFR Part 50, specifies pressure-temperature limits for core operation to provide additioral rr.argin during actual power production.
The pressure-temperature limits for core operation (except for low power physic.s tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minicum pressure-temperature curve for heatup and cooldown calculated as described in the preceding section. The minirum temperature for the inservice hydrostatic test for the vessel used in the preceJing example was 273*F. A vertical line at 273'F on the pressure-tempcrature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve as determined in the preceding section :onstitutes the limit for core operation for this example.
The ir. formation required to evaluate the adequacy of the terrperature limits can all be put on one figure. The temperature limits calculated in the preceding sections along with material data used are shown in Figure 4.
Acceptable limit curves for several typical plar.ts have been developed to e.
facilitate the evaluation of acceptability of proposed limits. These are included as Figures 5 and 6 of this review plan. Thcse are based on the sirrplified procedures described above, and are slightly more conservative than curves developed by more rigorous corputer calculations. Curves presented in plant Technical Specifications are considered acceptable if they are as con-servativa as these reference limit curves If they are based on more rigorous analytical rrethods, as recorrended by the Code or WRC Bulletin 175, they will be considered acceptable if the variation from the reference limit curves is not more than 10'F.
147
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5.3.2-7
If the proposed limit curves are more than 10*F less cons 'rvative than the reference limit curves, detailed bases 'nd calculations rest be submitted by the applicant for review. To be acceptable, all bases and analytical expressiens useu must be in accordance with Appendix G, 10 CFR Part 50, and the proposed curves must agree with check calculations made by the Materials Engineering Branch using these bases and expressions.
IV.
EVALUATICN FINDINGS The reviewer verifies that sufficient inforration has been provided to satisfy the require-r:ents of this review plan, and that the completeness and technical adequacy of his evalu-ation will support the following kind of concluding statement, to be included in the staff's safety evaluation report:
"Thc pressure-temperature limits to be imposed on the reactor coolant system for all operating and testing conditions to assure adequate safety margins against nonductile er rapidly propagating f ailure are in conformance with established criteria, codes, and standards acceptable to the Regulatery staff. The use of operating limits based on these criteria, as defined by applicable regulations, codes, and standards provides reasonable assurance that non6uctile or rapidly propagating failure will not occur, and constitutes an acceptable basis for satisfying the applicable requirerents of General Design Criterion 31."
V.
REFERENCES 1.
10 CFP Part 50, Appencix A. General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary."
2.
10 CFR Part 50, Appendix G. " Fracture Toughness Requirements."
3.
10 CFR Part 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements."
4.
ASME Bo;.cr and Presso.e Vessel Code,Section III, including Appendix G, " Protection Against Nonductile Failure," American Society of Mechanical Engineers.
5.
WPC Bulletin 175, "PVP,C Recorrendation on Fracture Toughness," Welding Research Ccuncil.
6.
Branch Technical Position MTEB 5-2, " Fracture Toughness Requiren.nts for Older Plants," appended.
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BRANCH TECHNICAL POSITION - NTEB NO. 5-2 FRACTURE TOUGHNESS REQUIREMENTS A.
Background
Current requirements regarding fracture toughness, pressure-temperature limits, and material surveillance are covered by the ASME Code and Appendices A, G, and H to 10 CFR Part 50.
The purpose of this branch technical position is to sumarize these requirements and provide clarification, as necessary.
Since many of these requirements were not in force when the plant was designed and built, this position also provides guidan;e for applying these requirements to older plants. Also included is a description of acceptable procedures for making the conservative estimates and assumptions for older plants that may be used to show compliance with the new requirements.
b.
Branch Technical Position 1.
Preservice Fracture Toughness Test Requirements.
The fracture toughness of all ferritic materials used for pressure-retaining components of tne reactar coolant pressure boundary shall be evaluated in accordance with the requirements of Section III of the ASME Code, as augmented by Appendix G, 10 CFR 50.
The fracture toughness test requirements for plants with construction permits prior to August 15, 1973 may not comply with the new Codes and Regulations in all respects. The fracture toughness of the materials for these plants must be assessed by using the available test data to estimate the fracture toughness in the same terms as the new requirements. This must be done because the operating limitations imposed on old plants must provide the same safety margins as are required for new plants.
1.1 Detemination of RT or Msel MaMals NDT Temperature limitations are determined in relation to a characteristic temperature of the material, RT!$DT, that is establishea from results of fracture toughness tests. Both drop weight NOTT tests and Charpy V-notch tests must be run to determine the RT The NDTT NDT.
temperature, as determined by drop weight tests (ASTM E-208) is the RT if, it 60'F NDT above the NDTT, at least 50 ft-lbs of energy and 35 mils lateral expansion are obtained in Charpy V tests on specimens oriented in the weak direction (traverse to the direction of maximum working).
In most cases, the fracture toughness testing performed on vessel material for old':r plants did not include all tests required to determine the RT in this manner.
NDT Acceptable estimation methods far the most comr:on cases, based on correlations of dat6 from a large number of heats of vessel material, are provided for guidance.
(1) If dropweight tests were not performed, but full Charpy V-notch curves were obtained, the NDTT for SA-533 rrade B, Class 1 plate and weld material may be assumed to be the temperature at wnich 30 ft-lbs was obtained in Charpy V-notch tests, or O'F, whichever was higher.
(2) If dropweight tests were not performed on SA-508, Class II forgings, the NDTT ray be estimated as the lowest of the following temperatures:
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b 5.3.2-14
(a) 60 F.
9 (b) The temperatures of the Charpy V-notch upper shelf.
(cj The temperature at which 100 f t-lbs was obtained on Charpy V-notch tests if the upper-shelf energy values were above 100 ft-lbs.
(3) If transversely-oriented Charpy Veotch spr-imens were not tested, the tempera-ture at which 50 ft-lbs and 35 mils LE would have been obtained on traverse specimens may be estimated by one of the following criteria:
(a) Test results from longitudinally-criented specimens reduced to 657 of their value to provide conservative estimates of values expected from transversely oriented specimens.
(b) Temperatures at which 50 ft-lbs and 35 mils LE were obtained on longitudinally-oriented specimens increased 20 F to provide a cnnservative estimate of the temperature that would have been required to obtain the sane values on transversely-oriented specimens.
(4) If limited Charpy V-notch tests were performed at a single temperature to confirm that at least 30 f t-lbs was obtained, that temperature may be used as an esti-mate of the RT provided that at least 45 f t-lbs was obtained if the specimens g
were longitudinally-oriented. If the mininum value obtained was less than 45 ft-lbs, the RT may be estimated as 20 F above the test temperature.
t4DT 1.2 Estimation of Charpy V Upper-Shelf Energies For the beltline region of reactor vessels, the upper shelf toughness must be adequate to accommodate degradation by neutron radiation. The original minimum shelf energy rust be 75 f t-lbs for vessels with an estinated end of lif e neutron fluence (> 1 MeV)
I9 of 1 x 10 and over. A value of 70 f t-lbs is considered adequate for material for vessels that will be subjected to lower fluences.
If upper-shelf Charpy energy values were not obtained, conservative estimates should be made using results of tests on specirens from the first surveillance capsule removed.
If tests were only made on longituoinal specimens, the values should be reduced to 65% of the lonqitudinal values to estimate the transverse properties.
l.3 Reporting Requirements Fracture toughness information required by the Code and by Appendix G, 10 CFR 50, must be reported in the FSAR to provide a basis for evaluating the adequacy of the operating limitations given in the Technical Specifications. In the case of older plants, the data may be estimated, using the procedures listed above, or other methods that can be shown to be conservative.
2.
Operating Limitations for Fracture Toughness 2.1 Required Pressure-Temperature Operating Limitations As rcquired by Appendix G 10 CFR 50, the following operating limitations shall be determined and included in the Technical Specifications. The basis for determination
- 5. 3. 2-l 5 147 119
shall be reported, and is the responsibility of the applicant, but in no case shall the limitations provide less safety margin than those determined in accordance with Apper.d'x G,10 CFR 50, and Appendix G to Sectica III of the Code.
(1) "inimum temperatures for performing any hydrostatic test involving pressurization of the reactor vessel after installation in the system.
(2) Minimum te peratures for all leak and hydrostatic tests performed af ter the plant is in service.
(3) Maximum pressure-minimum temperature curves for operation, including startup, upset, and cooldown conditions.
(4) Maximum pressure-minimum temperature curves for core operation.
2.2 Recommended Bases for Or-ating Limitations 2.2.1 Leak and..ydrostatic Tests (1) It is recorrended that no tests at pressures higher than design pressure be conducted with fuel in the vessel.
(2) Tests at pressures less than design pressure shculd be conducted at temperatures calculated according to Appendix G of Section III of the Code for the beltline region (including conservative estimates of radiation damage, see Section 3.0 below) if the maximum calculated primary stress in no otner region of the vessel exceeds 1.25 5 during the test, and the RT of the beltline is assumed to be riDT at least 30'F above that of the higher stressed regions. If primary stresses are calculated to be over 1.25 S in any region during the test, the RT f
m f4DT the vessel must be assumed to be at least 50 F higher than that of any region where the calculated primary stresses are over 1.25 S '
m (3) Alternatively, a fracture mechanics analysis, with technical justification for all assucptions and bases, may be made to determine the minimun test tempera-ture.
In no event shall the minimun temperature be lower than that resulting from calculations for the beltline region in accordance with Appendix G of the Code.
2.2.2 Heatup and Cooldown Limit Curves Heatup and cooldown pressure-temperature limit curves may be determined using simple fstresscalculations,usingthemethodgiveninAppendixGoftheCode.
The effect of therral gradients may be conservatively approv.imated by the procedures in Appendi.x G of the Code or from Figure 4-5 in WRC Bulletin 175.
Calculations need only be perforced for the beltline region, if the assurred RT fiDT the beltline is at least 50'F above the RT ;DT or all MgW sWssM mgions.
f Alternatively, more rigorous analytical procedures may be used, provided that the intent of the Code is rret, and adequate technical justification for all assumptions and bases is provided.
5.3.2-16
2.2.3 Core Operation Limits To provide added margins during actual core operation, Appendix G, 10 CFR 50 requires a minimum temperature during core operation, and a 40 F margin in temperature over the pressure-temperature limits as determined for heatup and cooldown in 2.2.2 above.
The minimum temperature, regardless of pressure, is the temperature calculated for the inservice hydrostatic test according to 2.2.1 above.
2.2.4 Upset Conditions The pressure-temperature limits described in 2.2.2 and 2.2.3 above are applicable to upset conditions. Normal operating procedures mus+ permit variations fron intended operation, including all upset conditions, without e(ceeding the limit curves.
2.2.5 Emergency and Faulted Conditions It is recognized that the severity of a tran;ient resulting from an emergency or faulted condition is not directly related to operating conditions, and resulting temperature-stress relationships in the reactor coolant boundary components are primarily system dependent, and therefore not under direct control of the operator.
For these reasons, operating limits for emergency and faulted conditions are not a requirement of the Technical Specifications.
The SAR must present evaluations of the continued integrity of all vital components during postulated faulted conditions. It is recomended that such es -luations 9
be made in a realistic manner, avoiding grossly overconservative assumptions and procedures, and clearly show that margins against loss of integrity are calculable and adequate.
2.3 Reporting Requirements The Technical 5pecifications must include the operating and test limits discussed above, and the basis for their determination. The Technical Specifications must also include information on the intended operating procedures, and justify that adequate raryins between the expected conditions and the limit conditions will be provided to protect against unexpected or upset conditions.
3.
Inservice Surveillance of Fracture Tuughness The reactor vessel may he exposed to significant neutron radiation during the service life. This will affect both the tensile and toughness properties. A matcrial surveillance program in conformance with Appendix H, 10 CFR 50, must be carried out.
3.1 Surveillance Progra,a Requirements The minimual requirements for the sarveillance program are covered by Appendix H, 10 CFR 50.
It is strongly recoended that consideration by given to the desira-bility of additienal surveillance methods, such as the inclusion of CT, DWT, DT, or other specirens to provide the capability of redundant test methods and analytical kk7
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5.3.2-17
I9 procedures, particularly if the estimated neutron fluence is over 2 x 10, Or the toughness of the vessel material is marginal.
The selection of material to be included in the surveillance program should be in accordance with ASTM E-185-73, unless the intent of the progeam is better realized by using more rigorous criteria. For example, the approach of estimating the actual RT nd upper shelf toughness of each plate, forging, or weld in the beltline as a riDT function of service life, and choosing as the surveillance materials those that are expected to be most limiting, may be preferable in some cases. This would include consideration of the initial RTriDT, the upper shelf toughness, the expected radiation sensitivity of the material (based on copper and phosphorous content, for example) and the neutron fluence expected at its location in the vessel.
3.2 SAR Requirements The adequacy of the surveillance program cannot be evaluated unless all pertinent information is included in the SAR.
Information requested for beltline materials includes the following:
(1) Tensile properties.
(2) DWT and Charg " t t results used to determine RT t4DT '
(3) Cnarpy V test results to determine the upper shelf toughness.
(4) Composition, specifically the copper and phosphorous content.
(5) Estimated maximum fluence for each beltline material.
(6) List of materials included in the surveillance program, with basis used for their selection.
3.3 Surveillance Test Procedures Surveillance capsules must be removed and tested at intervals in accordance with Apperdix H, 10 CFR 50.
The proposed renoval and test scnedule shall be included in the Technical Specifications.
3.4 Reporting Requirererits All information used to evaluate results of the tests on surveillance materials, evaluation methods, and results of the evaluation should be submitted with the evaluation report. This should include:
(1) Original properties and corpositi as of the materials.
(2) Fluence calculations, including original predictions, for both surveillance specimens and vessel wall.
(3) Test results on surveillance specimens.
(4) Basis for evaluation of changes in RT nd upper shelf toughness.
riDT (5) Updated prediction of vessel properties.
3.5 Technical Specification Changes Changes in *he operating and test limits reconnended as a result of evaluating the properties of the surveillance material, together with the basis for these changes, shall be submitted to the Directorate of Licensing for approval.
5.3.2-18