ML19329E016

App 4A of Rancho Seco PSAR, Answers to Questions. Includes Revisions 1-4

ML19329E016
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Site:	Rancho Seco
Issue date:	10/31/1967
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APPENDIX 4 4A ANSWERS TO QUESTIONS LO W

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assumed. The crack arrest temperature through the thickness

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of the wall was developed on a stress-temperature coordinate system. The actual quench-induced, stress-temperature con-dition through the thickness of the wal.1 at several times during the quench was developed and plotted (Figures 4A.1-4 and 4A.1-5).

The maximum depth at which the material in the vessel wall would be in tension or at which the stress in the material would be in excess of the threshold stress for crack initiation (5-8 ksi) was determined by comparison of the plots.

Comparison shows that a crack could propagate only through the inner 35 percent of the wall thickness if a crack initiation threshold of 5-8 ksi is applicable, and further that a crack could propagate only through the inner 43 percent of the wall thickness if a crack initiation threshold of zero were assumed.

The foregoing method of analysis is essentially a stress analy-sis approach which assumes the worst conceivable material properties and a flaw size large enough to initiate a crack.

Actually, the outer 83 percent of the vessel wall is at a tem-perature above the RTT (NDTT + 60 F) when credit is taken for the neutron shielding, and for the original RTT profile through the wall thickness.

The analysis is conservative in that it does not deny that cracks can be initiated, and in that it assumed a crack from 1 to 2-ft long to exist in the vessel wall at the time of the accident. Therefore, it can be concluded that, if a crack were present in the worst location and orien-T tation (such as a circumferential1y oriented crack on the inside of the vessel wall), it could not propagate through the vessel wall.

A fracture mechanics analysis was conducted which assumed a con-tinuous surface flaw to exist on the inside surface of the vessel wall. The criterion used for the analysis is that a crack cannot propagate when the stress intensity at the tip of the crack is below the critical crack stress intensity fac tor (KIC). Using conservative values of K (f r fully irradiated TC cold 302-Grade B steel KIC equals 30,000 psi) 3 and the method of Emery 4 to calculate stress intensity factors, K, in the I

variable thermal transient stress field, it was found that the crack propagating energy is below that required for crack propa-gation when the crack reaches a depth of less than 3 in, or 35 percent of the wall thickness.

4A.1.1 The geometry of the plate and the cooling method assumed in the

analysis, ANSWER The analysis assumed a long cylindrical section which was insulated on the outside and subjected to a uniform flow of constant tempera-ture (90 F) cold water flowing past the inner wall of the reactor vessel and outer wall of the thermal shield.

For general dimen-sions of the thermal model and flow patch description, refer to Figure 4A.1-1.

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Docket 50-312 Amendment No. 1

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'a QUESTION THERMAL SHOCK ON REACTOR VESSEL 4A.1 With regard to thermal shock on the primary system components, induced by operation of the emergency core cooling system (ECCS),

provide details of an analysis which indicates that the reactor vessel can accommodate without failure the rapid temperature change at the end of its design life. The analysis should consider both the ductile yielding and the brittle fracture modes of failure, and should include the following specific information:

(See these fur-ther specific items following the answer to Question 4A.l.)

ANSWER The state of stress in the reactor vessel during the loss-of-Refer to coolant accident has been evaluated for an initial vessel tem-4.3.1.1 perature of 603 F.

The inside of the vessel wall is rapidly subjected to 90 F injection water at the m.aximum flow rate obtain-able. The results of this analysis show that the integrity of the vessel is not violated.

The assumed modes of failure are ductile yielding and brittle fracture.

The modes of failure are considered separately as follows:

a.

Ductile Yielding k \\s,)

The criterion for this mode of failure is that there shall be no gross yielding across the vessel wall using the minimum specified yield strength in the ASME Code,Section III. The analysis con-sidered the maximum combined thermal and pressure stresses through the vessel wall thickness as a function of time during the safety injection. Comparison of calculated stresses to the material yield stress indicated that local yielding may occur in the inner 14.7 percent of the vessel wall thickness.

b.

Brittle Fracture Since the reactor vessel wall in the core region is subjected to neutron flux resulting in embrittlement of the steel, this area s

was analyzed from both a transition temperature and a fracture mechanics approach.

The results of the two methods of analysis compare favorably and show that pressure vessel integrity is not lost.

The criterion used in the transition temperature analysis is that a crack cannot propagate beyond any point where the applied stress is below the threshold stress for crack initiation (5-8 ksi) or when the stress is compressive. 1, 2 This approach involves making the very conservative assumption that all of the l

vessel material could propagate a crack by a low energy 7-~

absorption or cleavage mode.

End-of-life vessel conditions were

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0133

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4A.l.3 The initial temperature of the vessel as a function of time delay in injecting the cold water, ANSWER Tne reactor vessel wall is protected agains.t radiation heating from the hot reac' tor core by three solid barriers:

(a) the core shroud, (b) the core support barrel, and (c) the thermal shield.

Each of these barriers is separated by a steam gap so that the reactor vessel is in a sense insulated from the hot core.

In addition, the core barrel assembly and the thermal shield have considerable mass, i.e.,

63,750 lbs and 47,500 lbs respectively, that must be heated before the reactor vessel wall is affected. The arrangement of these barriers is shown on Figures 3.2-59 and 3.2-60 of the PSAR.

Calculations show that the reactor vessel wall temperature will not 3

increase as a function of time during the first several hundred seconds of an LOCA.

The various component temperatures at 500 see and at 1,400 sec are:

Temperature Temperature Component (at 500 sec), F (at 1400 sec). F Core shroud 731 1,770 Core support barrel 579 770 Thermal shield 576 582 Reactor vessel 576 576 3

4A.I.4 The ef fect of axial temperature gradient in the vessel, during fill-ing with cold water, on the total scress intensity, ANSWER Figure 4A.1-3 shows the temperature profile through the vessel wall when the core flooding water impinges on a section of the vessel wall considering an abrupt line of demarcation between fluid and steam.

The use of such an abrupt line of demarcation between fluid and steam is conservative.

The conduction of heat through the ves-sel produces the gradual temperature change as shown on the isotherm plot on Figure 4A.1-3.

This temperature distribution has been 9

analyzed using the Seal Shell Computer Program, and the results of this analysis are shown as a stress profile on Figure 4A.1-3.

This stress profile shows that the worst stress condition is remote from the line of demarcation between fluid and steam, and that the axial conduction has more than of fset any adverse influence of the uncooled portion of the vessel wall. Therefore, the original anal-ysis, assuming a long cylinder subjected to a uniform quench, has presented the worst condition because the effect of the axial grad-ient will locally decrease the stress produced by ECCS operation in the LOCA.

4A.1.5 The effect of a circumferential1y nonuniform cooling of the vessel shell, by the cold water entering the vessel through the injection nozzles, on the stresses and distortion in the vessel, comEG 2

T)OJJU 0134 4 A -4 Amendment 1

[A The cooling method assumed in the analysis is as follows:

a.

The metal in the vessel wall and thermal shield is cooled by conduction.

b.

The heat transferred to the fluid is by forced convection.

4A.l.2 The heat transfer coefficient used, its experimental basis, and the degree of conservatism involved, ANSWER Theanalygisusedawaterfilmheattransfercoefficientof3,000 Btu /hr-ft -F.

Using the classical (text book) approach, 5 the water film heat transfer coefficient was calculated to be about 900 Btu /hr-2 ft

-F.

However, when the water film heat transfer coefficient reaches 2

a value of 2,000 to 3,000 Btu /hr-ft -F or more, the heat transfer properties of the metal, i.e.,

the metal conductivity, will govern the heat transfer rate, and consequently the shape and variation of the temperature profile through the thickness of the vessel wall with time (reference Figure 4A.1-2).

The experimental basis and degree of conservatism for the use of a water film heat transfer coefficient of 3,000 Btu /hr-f t2-F is as follows:

a.

The most severe condition that could possibly be postulated would be to quench the cylindrical portion of the vessal in a quench

( (j S1 tank. Much experimental work has been done to determine the V

water film heat transfer coefficient for this condition. 6,7,8 Using Reference 6, the water film heat transfer coefficient (f) is calculated as follows:

f = 2H k g

f = 2 x 4 x 277 2

f = 2,216 Btu /hr-ft _y where:

f = water film heat transfer coefficient, 2

Btu /hr-ft _p Hg = Grossman's Severity of Quench

(= 4 in violently agitated water) k = thermal conductivity of the material, 2

Btu /hr-ft -F/in. (= 277 for SA302GB) b.

The comparison of our assumed water film heat transfer coefficient to the coefficient as calculated by Reference 5 yields a conservative ratio of 3.32, and a comparison to the

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water film heat transfer coefficient, calculated by Refer-

'V ence 6, yields a conservative ratio of 1.35.

0135 o n-n Amendment 1 4A-3

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4A.l.ll The value of the yield stress used as the failure criterion in the

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ductile yielding analysis.

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ANSWER The analysis used the minimum yield strength values as a function of temperature, as listed in Table N-424 of Section III of the ASME Code.

The values of yield strength for SA 302, Grade B, are as follows:

Temperature, F Stress, psi 100 50,000 200 47,150 300 45,250 400 44,500 500 43,200 600 42,000 REFERENCES 1

Pellini, W. S. and Puzak, P.

P.,

Practical Considerations in Applying Laboratory Fracture Test Criteria to Fracture Eafe Design of Pressure Vessels, NRL 6030.

2 Pellini, W.

S. and Puzak, P.

P., Fracture Analysis Diagram Procedures for the Fracture Safe Engineering Design of Steel Structures, NRL 5920 3

La nd erma n,

E., Yanichko, S.

E., and Hazelton, W.

S.,

An Evaluation of Radiation Damage to Reactor Vessel Steels Using Both the Transition Temperature and Fracture Mechanics Approaches, WAPD.

4 Emery, A.

F., " Stress Intensity Factors for Thermal Stresses in Thick Hollow Cylinders," Journal of Basic Engineering, March 1966.

5 Hsu, S.

T., Engineering Heat Transfer, Van Nostrand, 1963, pg. 301, 6

Grossman, M.

A.,

Elements of Hardenability.

Austin, J.

B.,

Heat Flow in Metals, ASM Publication.

O Russell, T.

F., Russell's Tables.

9 OO-WAPD-TM-398.

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QUESTION With respect to the brittle fracture mode of failure provide the 4A.2 following additional information:

4A.2.1 The assumed distribution of the initial NDT temperature through the plate thickness. State also the experimental basis for this assump-tion and the degree of conservatism involved.

ANSWER The distribution of NDTT through the plate thickness was assumed to be a constant value of +10 F.

The +10 F was assumed because the B&W Material Specification requires that material in the core region will have, as a maximum, an initial NDIT of +10 F at a depth below the surface equal to 1/4 T.

The use of a constant value through the thickness of the plate is conservative when consideration is given to the recent work from Lehigh University, 1 B&W, 2 and others. 3 From the references cited it is found that, for all practical pur-poses, the NDTT at 1/4 T is the same as the NDTT at 1/2 T for plates in the thickness range of 8 to 12 in.

Our analysis is conservative in that it did not consider the benefit which could be gained by considering the enhanced properties which exist at the surface.

From References 1 and 2 the NDTT at the surface would be expected to be -50 F.

[

The AEC, with the cooperation of Industry, is at present engaged in a pro' gram of material characterization which will further substanti-ate the data presented here.

4A.2.2 The assumed time-integrated neutron flux (nvt) at the reactor ves-sel inner diameter.

ANSWER The assumed time-integrated neutron flux (nvt) at the reactor ves-sel inner wall is 3 x 1019 n/cm2 (E > 1 Mev).

This value is stated in the PSAR Section 4.1.4.1, page 4.1-8.

4A.2.3 The profile of the NDT temperature shift through the thickness of the plate.

ANSWER The NDTT profile at the end of Station life was assumed to be a constant value of 250 F through the thickness of the reactor vessel wall.

This value was stated in the PSAR Section 4.1.4.1, page 4.1-8.

The use of a constant value for NDTT shif t is very conservative be-cause the analysis did not consider the beneficial effect which can 4 of the material to be realized by considering the self-shielding radiation damage.

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REFERENCES I

Strunk, S.

S.,

Pense, A. W.,

and Stout, R.

D.,

The Properties and Micro-structure of Spray-Quenched Thick-Section Steels, Welding Research Council Bulletin No. 120, February 1967.

(Appendix A)

B&W Da ta on SA-302GB Ma teria l. (Appendix B) 3 Naval Reactors Program Data (Classified).

4 NRL Memo No. 1731, p. 15 - 21.

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0148 Amendment 1 4A-8

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a QUESTION An estimate of the effect of an initial vessel temperature higher 4A.3 than that assumed in the analysis on the extent of yielding and deformation of the vessel.

ANSWER The analysis used an initial vessel wall temperature of 603 F.

A sensitivity analysis considering various initial wall tempera-tures, up to 1,500 F, has also been completed. The results indi-4 cate that 31 percent of the material in the inner portion of the vessel wall thickness has yielded at 1,500 F.

The analysis with an initial wall temperature of 603 F indicated that 14.7 percent of material in the inner pertion of the vessel wall thickness had yielded.

Thus, the sensitivity analysis indicated an increase in the ductile yielding of 16.3 percent when tne initial wall temperature was assumed to be 1,500 F.

QUESTION An estimate of the maximum allowable pressure stress, when com-4A.4 bined with other stresses present in the vessel, which could be tolerated without failure.

ANSWER The maximum pressure that B&W considered was 600 psi. This is based on the fact that the core flooding tanks will not operate until the reactor vessel pressure is at or below 600 psi. This

([d.h internal pressure would only increase the depth of ductile yield-ing from 14.7 to 17.5 percent of the wall thickness.

QUESTION.

An estimate of the mdximum neutron flux exposure (nyt) of the 4A.5 vessel that could be tolerated without vessel failure.

ANSWER The analysis considering the brittle fracture mode assumed the conservative approach in that the material would behave in a com-pletely brittle manner, and thus the lower threshold stress was used for comparison with the imposed stresses. Therefore, the analysis as performed by B&W is insensitive to increased flux levels.

' I' pnnLC 0149 Amendment 1 4A-9

QUESTION The effect of potential local penetrations present in the vessel 4A.6 cladding, exposing the base metal to the coolant, on the results of the analysis.

ANSUER Our analysis did not consider the beneficial effect of cladding.

In regions where local penetrations in the clad surface are postulated to be potential occurrences, the actual temperature profile across the thickness will be virtually unchanged (because of the small difference in conductivity and the small thickness of clad), and the stresses at these points will be as they were originally calculated.

QUESTION The number of thermal shock cycles, induced by ECCS operation, 4A.7 that the vessel could withstand at the end of its fatigue life.

ANSWER B&U does not consider the ECCS operation as a cyclic occurrence.

However, plastic deformation (ductile yielding) might safely be repeated without the integrity of the vessel being violated.

If ECCS operation should occur when the vessel is in the brittle region, then further operation of the unit would be prohibited until an exhaustive examination of the vessel has been completed.

QUESTION Experimental data on the thermal shock effects in thich plates 4A.8 under stress, tested below the NDT temperature.

ANSWER The demonstration of the adequacy of the reactor vessel to accom-modate the thermal gradients, developed upon injection of emer-gency coolant following a loss-of-coolant accident, is a unique application of fracture mechanics and analysis involving stressed plates, thermal gradients, crack triggering by quenching, transi-tion temperature gradients, and notch geometries.

Data relative to the individual parts of this problem are avail-able.

This data exists in the form of the Robertson Gradient Tests, routine practice in cuenching heavy section shell forgings, and the transition temperature correlation work carried out by Pellini and Puzak at NRL.

Also there is extensive uork which is being conducted in the fracture mechanics field by such research establishments as ORNL, Westinghouse Research, and Universities.

All of this data was valuable in developing the conservative methods which were used in the analysis as presented.

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QUESTION An evaluation of the capability of the safety injection nozzles 4A.9 and accumulator piping to withstand the transient.

ANSWER B&W is considering the effect of ECCS operation in the analysis of the safety injection nozzle and accumulator piping.

As soon as this analysis is completed, the results of the analysis will be presented.

QUESTION An evaluation of the effects of this transient on the core barrel 4A.10 and other internals with regard to assuring that distortion would not restrict the flow path of the emergency core coolant.

ANSWER A detailed analysis of the effects of emergency core coolant flow on the reactor internals has not been performed.

However, pre-liminary analysis and previous similar experience indicate the following:

The reactor internals are constructed of Type 304 stain-less steel, and therefore are not subject to brittle fracture at temperatures of interest (some loss of impact strength has been observed at about -320 F).

Further, the material is sufficiently ductile that many quenches of

( (g-g the expected magnitude can be withstood without initiation s

of a crack, or propagation of an assumed existing crack.

Consequently, thermal shock f racture of the internals is not considered credible.

The reactor internals are being designed to conservative stress and deflection limits, so that failure or large deformations of the internals due to blowdown loadings will not occur.

A further degree of conservatism is provided by coolant inlet flow deflector vanes in the region of the emergency coolant inlets.

These vanes are attached to the core support shield, and will prevent that shield from approaching within about 5 in. of the vessel ID in the region of the emergency coolant inlets.

A (d'

(

, 0151 Amendment 1 4A-ll

03 QUESTION Provide a detailed outline of the research program required to 4A.ll verify the analysis methods on thermal shock effects in thick plates under stress below the NDT temperature.

Identify any other area related to the pressure vessel and piping thermal shock problem that requires a research and development program for proof-of-principle, and outline the required program.

ANSWER For safety analysis purposes, B&W does not believe a research program can significantly affect the conclusions obtained by the methods used in the analysis of the thermal shock effects on the reactor vessel caused by the actuation of the ECCS due to an LOCA.

However, as part of industry's continuing effort to improve the detailed knowledge of material behavior under all conceivable conditions, B&W has included this subject on the agenda for the PVRC meeting held January 16, 1968.

B&W does not consider that any area related to the pressure ves-sel and piping thermal shock problem requires a research and development program for proof-of-principle.

Y

~

0152 g

J 4A-12 Arendnent 1

QUESTION Thermal Shock 4A.12

~

(DRL 4.1) With regard to thermal shock on reactor components, induced by operation of the emergency core cooling system (ECCS), provide details of an analysis which indicates that the reactor vessel and reactor internals can withstand the rapid temperature change at the end of their design life.

The analysis should include both the ductile yielding and the brittle fracture modes of failure.

4A.12.1 The brittle fracture analysis for the vessel (DRL 4.1.1) should assume an initial crack size just below the critical crack size corresponding to the stresses present during normal operation and transients.

Since the initial crack is most likely to exist in a weld or a heat affected zone, the analysis should consider two cases:

a circumferential crack, and a crack parallel to the axis of the reactor vessel.

The details of the analysis should be provided including specific information on:

(a)

The critical stress intensity factor (KIC) assumed, and the basis for its selection, (b) The assumed time-integrated neutron flux C'I, ',

(c) The value of residual stresses assumed in (nyt) at the reactor vessel inner diameter, the base metal and the weld areas, (d) The initial crack geometry and size assumed in the analysis, (e)

Equations used to correlate crack size with the calculated stress intensity factor (K ).

y ANSWER (a)

This question was answered in the reply to Question 8.11.1 of the Florida Power Corporation PSAR (Docket No:

50-302 and

-303.)

(b)

This question has already been answered in Question 4A.2.2 in Amendment 1 to the Rancho q

Seco PSAR.

(c)

This question has already been answered in Question 4A.l.8 in Amendment 1 to the Rancho Seco PSAR.

(d)

This question was answered in the reply to Question 8.11.2 of the Florida Power Corporation PSAR (Docket Nos. 50-302 and

-303.)

(4 ~

(e) This question was answered in the reply to Question 8.11.3 of the Florida Power Corporation PSAR (DocketcNos. 50-302 and -303.)

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Amendmen't 2 4A-13

- {p zg,g y v."; a j

4A.12.2 The details of the ductile yielding mode of

'}-

(DRL 4.1.2) analysis for the vessel should include the following information:

(a) The geometry of the plate and the cooling method assumed in the analysis, (b)

The heat transfer coefficient used, its experimental basis, and the degree of con-servatism involved, (c) The initial temperature of the vessel as a function of time delay in injecting the cold

water, (d) The effect of axial temperature gradient in the vessel, during filling with cold water, on the total stress intensity and the dis-tortion of the vessel, (e) The temperature profiles and the calculated thermal stress profiles through the thickness of the plate for several times during the cold water injection transient, (f) The magnitude of the axial dead load stresses in the vessel, (g) The magnitude of the stresses in the vessel shell due to potential simultaneous seismic

loading, (h) The value of the yield stress used as the failure criterion in the ductile yielding

)

analysis.

ANSWER (a) This questi^- has already been answered in Question 4A._.1 in Amendment 1 to the Rancho Seco PSAR.

(b)

This question has already been answered in Question 4A.1.2 in Amendment 1. to the Rancho Seco PSAR.

(c)

This question has already been answered in Question 4A.l.3 in Amendment 1 to the Rancho Seco PSAR.

(d) This question has already been answered in Question 4A.1.4 in Amendment 1 to the Rancho Seco PSAR.

(c) This question has already been answered in Question 4A.l.6 in Amendment 1 to the Rancho Seco PSAR.

(f) This question has already been answered in Question 4A.1.9 in Amendment 1 to the Rancho Seco PSAR.

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01.54 4A-14 Amendment 2

A (g) This question has already been answered in I

1 Question 4A.1.10 in Amendment 1 to the Rancho Seco PSAR.

(h) This question has already been answered in Question 4A.l.ll in Amendment 1 to the Rancho Seco PSAR.

4A.12.3 Based on the analyses for the vessel provide:

(DRL 4.1.3)

(a) An estimate of the maximum acceptable initial temperature of the vessel that could be tol-erated without failure of the vessel, (b) An estimate of the maximum neutron flux exposure (nyt) of the vessel that could be tolerated without vessel failure, (c) An estimate of the maximum allowable pressure stress, when combined with other stresses present in the vessel, which could be tol-erated without failure.

ANSWER (a) This question has already been answered in

. Question 4A.3 in Amendment 1 to the Rancho Seco PSAR.

l(-

(b) This question has already been answered in s

(

Question 4A.5 in Amendment 1 to the Rancho Seco PSAR.

(c)

This question has already been answered in Question 4A.4 in Amendment 1 to the Rancho Seco PSAR.

4A.12.4 Evaluate the capability of the piping, safety (DRL 4.1.4) injection nozzles, and vessel nozzles to with-stand the transient.

ANSWER This question has already been answered in Question 4A.9 in Amendment 1 to the Rancho Seco PSAR.

4A.12.5 Evaluate the effects of this transient on the (DRL 4.1.5) core barrel and other internals with regard to assuring that' distortion would not restrict the flow path of the emergency core coolant.

ANSWER This question was answered in Question 4A.10 in Amendment 1 of the Rancho Seco PSAR.

jn 0155 MJ CC07I Amendment 2_

4A-15

4A.12.6 Current status of the fracture mechanics analysis of the

).

thermal shock problem.

ANSWER This problem was evaluated using two different analytical techniques and presented in the answers to Question 4A.1 in Appendix 4A.

One of these techniques was based on ductile yielding data relative to the propagation of flaws in reactor vessel steels. The other was based on a fracture mechanics analysis of the problem. Both of these methods predicted consistent results which indicate that the reactor vessel would not crack through its thickness as a result of this thermal shock.

During the ACRS review of other reactor applications in January 1968, a third method of analysis was proposed. This proposed method can be found in ASTM STP-381. While it is 3

felt that the evaluation presented in the PSAR adequately demonstrates that a crack will not propagate through the vessel wall as a result of the thermal shock, this third evaluation was undertaken using the method suggested. The preliminary results of this third method of ane!ysis confirm the results of the evaluation presented in the PSAR Lf demon-strating that the crack will not propagate through the wail of the vessel.

The assumptions in the original fracture mechanics analysis

'S (Question 4.A.1) and in the third method of analysis differed

,)

primarily in that, in the original fracture mechanics analy-sis, the critical stress intensity factor was considered to be a variable and residual stresses were considered to remain constant.

In the third method the critical stress intensity factor was considered to be a constant, and the residual stresses were considered to vary.

iT q 7 71 __

vvvi u O}.50 4A-16 Amendment 3

b QUESTION Discuss the full power radiation environment with respect Q

4A.13 to corresponding damage thresholds for the control rod (DRL 4.3) actuators and the primary loop pumps and pump motors.

Con-sider the N-16 activity, the fission product activity in coolant, and the radiation streaming contributions.

ANSWER The dose to primary loop pumps and motors from all gamma and 7

neutron sources is computed to be about 2 x 10 rads at the end of 32 effective full power years. Of this total dose approximately 60*/. is from N-16, 20*/ from fission and corro-3 sion products, and 20*/. from streaming through primary shield penetrations.

With the exception of the lubricating oil all materials in i

the pumps and motors are rated as being capable of withstanding at least 108 radsexposugebeforeexhibitinganysignsof radiation damage. At 10 rads the lubricating oil shows about a 10*/. increase in viscosity. However, this oil will be replaced every 5 to 10 years resulting in a maximum exposure to the oil of about 5 x 106 rads.

The radiation damage thresholds for all materials in the control rod actuators have not been identified at present.

The correlation of damage thresholds with radiation levels will be determined, and the design will take cognizance.

co n}

0'L57 F"

wo u I J Amendment 3 4A-17

QUESTION Provide a tabulation of all the nuclear pressure vessels in 4A.14 the Class I (seismic design) systems in the facility. The (DRL 4.4) tabulation should include a notation of whether the vessel design is complete. the stage of fabrication of the vessel, and the extent to which each of the vessels will comply with each of the 34 supplementary criteria in " Tentative Regulatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressure Vessels", issued by AEC Press Release No. IN-817, dated August 25, 1967.

For each vessel, provide a discussion that represents the reason why total compliance is not feasible for each criterion not met in its entirety.

ANSWER The following will provide information relative to the status of design and fabrication of the nuclear steam supply system components fabricated by The Babcock & Wilcox Company 3

and their compliance with the AEC Supplementary Criteria, a.

Nuclear Pressure Vessels - Class I (Service Design)

Class I equipment in the Rancho Seco Nuclear Generating Station is defined in Appendix 5A of the PSAR. Vessels in the reactor coolant system are designed and classified in accordance with appropriate and existing codes as listed in Section 4.1 and Table 4.1-9 of the PSAR. Ves-S sels in the auxiliary systems are designed and classified

)

in accordance with appropriate and existing codes as listed in Section 9 of the PSAR.

The nuclear pressure vessels in the Class I (seismic design) systems in the facility are tabulated below:

Vessel Design Complete Status of Fabrication Reactor Vessel Yes Material ordered Steam generators Yes Material ordered Pressurizer Yes Material ordered Core Flooding Tanks Yes Material ordered 3

Control-Rod Drive No Purchase Orders Pressure Housing Not Yet Placed Decay Heat Coolers No

~

Letdown Coolers No Makeup Tank No Purification Demin-No eralizers Purification Filters No Seal Return Coolers No c.-,.

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0158 4A-18 Amendment 3 l

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b.

Supplementary Criteria The AEC's " Tentative ' Regulatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressure Vessels",

which were issued by AEC Press Release No. IN-817 (August 25, 1967), have been reviewed by Atomic Industrial Forum ad hoc group and the ASME, as well as several industrial concerns. The outcome of comments by these organizations is awaited.

Further reference is made to the B&W Company 3

comments forwarded directly to the AEC on this subject.

Specific comments applicabic to the B&W furnished Rancho Seco vessel are contained in B&W 1etter to Dr. Harold Price dated April 2, 1968.

Excluding the equipment which is being fabricated by The Babcock & Wilcox Company (reactor vessel, pressurizer, steam generators and core flooding tanks), no purchase orders have been placed for equipment.

For other equipment, a detailed answer to this question prior to selecting vendors and placing purchase orders is consider'ed premature.

Following is a tabulation of the Criteria in compliance

~~'

and notations on those Criteria not in compliance.

.v Criterion B&W Compliance Comments 1.10 Classification of No Letdown Cooler -

Vessels Class C 1.11 Conditions for Design Yes 3

1.12 Certification of Yes S tress Reports 1.13 Conditions with Yes Unspecified Design Rules 1.14 Vessel Owner's Respen-No Not practical sibility for Inspection

~

1.15 Manufacturer's Respon-Yes sibility for Quality Control 1.16 Vessel Fabrication No Do not literally Report ccuply with weld g i

l repair records.

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0159 7C

~~

Amendment 3 4A-19

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Criterion B&W Compliance Comments 1.17 Boundary between Yes Vessel and Piping 1.20 Vessel Material No Use design Property Improve-consideration ment and clean 11-ness and prop-erty requirements in lieu of Criterion 1.21 Material Test Coupons Yes 1.22 Nondestructive Exami-No Do not agree nation of Reactor Ves-technically, sel Plates 1.23 Nondestructive Exami-No Impractical nation and Repairs of technically -

Material follow code requirements.

1.24 Examination of Reactor No Criterion Vessel Bolts impractical -

3

)

use better j

method.

1.25 Ductile Brittle Transi-No No excess tion Properties material, not practical.

1.26 Exclusion of Repairs Yes in Bolting Material 1.30 Fracture Mechanics No Not on all materials.

1.31 Design for Cyclic Yes Loading 1.32 Bolting Design Require-No Not applicable

~

ments on small con-nections.

1.33 Earthquake Load No Use loads when specified in 1.34.

1.34 Design Conditions -

No No fatigue analy-Combination Loading sis for earth-s g _,17 yke - impracticai j Of.iSC) 4A-20 Amendment 3 1

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. ( 'T Criterion B&W Compliance Comments

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1.35 Computer Programs Yes 1.36 Environmental Effects No Not practical.

1.37 Design for inspecta-No As welded clad-bility ding surface.

1.38 Attachments to Reactor Yes Vessel 1.39 Reactor Vessel Core No Do not agree Support technically.

1.40 Chemical Analysis No Code require-of Weld Wire ment considered adequate.

1.41 Cutting Plates No Not practical.

3 1.42 Welding Qualification No Code require-Procedure Requirements ment considered adequate.

/

2 1,43 Precautions for Welding No Do not agree

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technically.

s tx, 1.44 Welding Requirements No Do not agree technically.

1.50 Final Inspection and No As welded clad-Examination ding.

1.51 Nondestructive Examina-No No approval of tion and Responsibilities procedures.

1.60 Hydrostatic Testing No Not practical.

Requirements 1.70 New Materials Yes nn-77 O f g.,

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i i

Amendment 3 4A-21

-w.

G

.I QUESTION Submit Certified Code Design Specifications for component 4A.15 parts of the Class I systems as required by the ASME Code (DRL 4. 5)

Section III, paragraph N-141 (passed 6-23-67).

ANSWER Refer to B&W proprietary topical report CS(F)-3-22-T 3

submitted separately by SMUD.

(Deleted) e-on~,n UV>/U 0.5.0?*

4A-22 Amendment 3 j

Docket 50-312

\\/

Amendment No. 1 February 2, 1968 APPENDIX B B&W DATA TYPICAL NDTT DATA FOR SA-302 B PLATE MODIFIED TO CODE CASE illo PARAGRAPH 1 1.

Material from Shell plate 6-1/4" thick Chemistry B.Q.

C

.18 6-1/4 hours, B.Q.

Aust.

1675-1725F 6-1/4 hours, Mn 1.08 Aust.

1600-1650F B.Q.

P

.005 Temper 1175-1225F 6-1/4 hours, F.C.

Lab S.R.1100-1150F, 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, S

.012 Si

.24 Cr

.16 NL

.44

[Q

(

All specimens longitudinal to final Mo

.45 rolling direction.

Cu

.17 A. As Temoered Procerties - Surface Charny Tests Test Temp.

Ft-Lhs Lat. Exn.. Mils Est. 4 Shear 50

$1 48, 62, 68 30, 30, 30 61,100 100 65 70 0F 25 100,100 52, 39, 60, 61 25,, 80, 90

-30F

-30F 69, 85 94 50

-60F 53, 67 38, 47 13, 30

-90F 19, 33, 48 15,, 27, 33 3,

5, 10 W

l B. As Tempered - Just Below 1/4T Pronerties Charny Tests

~

Test Temo.

Ft-Lbs Lat. Ex Mils Est. 4 Shear 30 50 25, 40,

+40F 48, 63,69 40

,4 5, 25, 30

+10F 55, 56, 60 43

,7 10

-20F 30, 33, 34 25

,26

-40F 10, 20, 30 7,

,21 0, -

5 k.-

oua 00070 -

Amendment 3

~<r 7

1 2.

Material from Shell plate 9-3/4" Thick Mn-Mo-Ni Plate (A533B)

Air Cool from 1675-1725F Quenched from 1675-1725F Tempered from 1200-1225F, Air Cool Stress Relieved 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> 1100-1150F, Furnace Cooled A. As Tempered Pronerties - Surface Charov Tests Test Temn.

Ft-Lbs

-80F 7,

9,

-50F 10 22 37

-20F 28, 45,, 45

+10F 35,, 60, 62

+40F 70, 83

+300F 134,134 B. As Temnered - Just Below 1/kT Pronerties Cherny Tests Test Temn.

Ft-Lbs.

-40F 16 18, 14, 29

-20F 11 0F 25, 20, 28

+10F 47, 50, 55, 38, 42, 43

, 32, 37

+40F

+300F 130,131 C. As Temnered - Just Below 1/2T Pronerties i

Charny Tests l

Test Temn.

Ft-Lbs.

-40F 11

-20F 13, 16, 18 38' 46, 4, 35, 40, 42

+10F 21 33 3

+40F

+300 F 127,120 s

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UUvvU Q,jf}il Amendment 3