ML17212A336

Forwards Component Integrity Section,Matls Engineering Branch Draft SER Input Re Reactor Vessel matls,pressure-temp Limits,Reactor Vessel Integrity & Pump Flywheel Integrity. Request for Addl Info Encl

ML17212A336
Person / Time
Site:	Saint Lucie
Issue date:	06/30/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
References
NUDOCS 8107140453
Download: ML17212A336 (22)

Text

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Docket File.:

50-389 Dr'. Robert E. Uhrig, Vice President Advanced Systems and Technology Florida Power 5 Light Company P. 0. Box 529100 Miami, Florida 33152

Dear Dr. Uhrig:

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SUBJECT:

DRAFT SAFETY E'VALUATION REPORT FOR REACTOR VESSEL MATERIALS, PRESSURE-TEMPERATURE LIMITS, REACTOR VESSEL INTEGRITY AND PUMP FLYWHEEL INTEGRITY The Component Integrity Section, Materials Engineering Branch, Division, of-'ngineering, has reviewed the Final Safety Analysis Report for St. Lucie Unit No. 2.

Based on our review of this information, we have prepared our input to the Safety Evaluation Report (Attachment 1).

In this s';afety Evaluation Report we have identified areas for which sufficient information has nest,been submitted to determine compliance with or justification for an exemption to Appendices G and H, 10 CFR Part 50.

The areas where sufficient information has not been provided will remain.open items unti-l.='Florida Power 5 Light Co.

provides the, necessary

,information;-= The specific information required to resolve these open items is contained in Attachment 2.

Responses to the enclosed request should be submitted by July 14, 1981; If you cannot meet this date, please inform us within seven days after receipt of'his letter of the date you plant to submit your responses.

Please contact Mr. Nerses (301-492-7468), St. Lucie 2 Project Manager, if you desire any discussion or clarification of the enclosed report.

Sincerely, 0rtglnd Iign& bV Robert L Tedeloo Robert L. Tedesco,'Assistant Director for "Licensing,'

Divi sion of Licensing

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 m>s Q1 Docket File.:

50-389 Dr. Robert E.,Uhrig, Vice President Advanced Systems and Technology Florida Power 5 Light Company P. 0. Box 529100 Miami, Flor ida 33152

SUBJECT:

DRAFT SAFETY EVALUATION REPORT FOR REACTOR VESSEL MATERIALS, PRESSURE-TEMPERATURE

LIMITS, REACTOR VESSEL INTEGRITY AND PUMP FLYWHEEL INTEGRITY The Component Integrity Section, Materials Engineering Branch, Division of Engineering, has reviewed the Final Safety Analysis Report for St. Lucie Unit No. 2.

Based on our review of this information, we have prepared our input to the Safety Evaluation Repor t (Attachment I).

In this Safety Evaluation Report we have identified areas for which sufficient information has not been submitted to determine compliance with or justification for an exemption to Appendices G and H, 10 CFR Part 50.

The areas where sufficient information has not been provided will remain open items until Florida Power 8 Light Co.

provides the necessary information'.

The specific information required to resolve these open items is contained in Attachment 2.

Responses to the enclosed request should be submitted by July 14, 1981. If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plant to submit your responses.

Please contact Mr. Nerses (301-492-7468),

St. Lucie 2 Project Manager, if you desire any discussion or clarification of the enclosed report.

Sincerely,

Enclosure:

As stated Robert L. Tedesco, Assistant Director for Licensing Division of Licensing cc:

See next page

FLORIDA POWER AND LIGHT COMPANY ST.

LUCIE NUCLEAR STATION UNIT 2 DOCKET No. 50-389 MATERIALS ENGINEERING BRANCH, COMPONENT INTEGRITY SECTION 5.3. 1 Reactor Vessel Materials General Design Criterion 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," Appendix A, 10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and testing, conditions the boundary behaves in a nonbrittle manner and the probability of rapidly pro-pagating fracture is minimized.

General Design Criterion 32, "Inspection of Reactor Coolant Pressure Boundary," Appendix A, 10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed to permit an appropriate material surveillance program for the reactor pressure boundary.

The Construction Permit for the St.

Lucie 2 (hereinafter SL-2)

PMR nuclear reactor was issued on May 2, 1977.

The Edition and Addenda of the ASME Code applicable to the design and construction of any nuclear reactor is specified in Section 50.55a of 10 CFR Part 50.

Based on the Construction Permit date, this section of the Code of Federal Regulations requires that the SL-2 reactor vessel be purchased to the requirement of at least the 1971 Edition of the ASME-Code',

Summer 1972 Addenda.

The SL-2 FSAR states that the reactor vessel was designed, fabricated,

tested, inspected, and stamped according to the 1971 ASME Code, Summer 1972 Addenda.

Therefore, the applicant complied with the explicit requirements of Paragraph 50.55a (c) (3),

10 CFR Part 50.

Appendix G, "Fracture Toughness Requirements,"

and Appendix H, "Reactor Vessel Material Surveillance Requirements,"

of 10 CFR Part 50, specify the fracture toughness requirements for the ferritic materials of the reactor coolant 5-1

pressure boundary during normal operation, testing, maintenance, and antici-pated transient conditions.

5.3. 1. 1 Evaluation of Com liance to A

endix G

10 CFR Part 50 Based on our review of the applicant's submittal that describes the extent of compliance of SL-2 to Appendix G, 10 CFR Part 50, we have determined that the requirements of Appendix G have been met except for Paragraphs III.B.4, III.C.2, IV.A.1, IV.A.3 and IV.B.

These items were not identified by the applicant or were not mentioned in FSAR.

Paragraph III.B.4 of Appendix G, 10 CFR Part 50 requires that individuals performing fracture toughness tests shall be qualified by training and.

experience and shall have demonstrated competency to perform the tests in accord with written procedures of the component manufacturer.

The SL-2 FSAR does not mention the training or experience of the personnel performing the fracture toughness tests.

In order to determine whether the applicant has complied with the requirements of Paragraph III.B.4 of Appendix G, 10 CFR Part 50, the applicant must describe the training and qualifications of all individuals who performed fracture toughness testing.

Paragraph III.C.2 of Appendix G, 10 CFR Part 50 requires, in part, that materi-als used to prepare test specimens for the.reactor vessel beltline region shall be taken directly from excess material and welds in the vessel shell course(s) following completion of the production longitudinal weld joint, and subjected to a heat treatment that produces metallurgical effects equivalent to those produced in the vessel material throughout its fabrication process in accordance with paragraph NB-2211 of the ASHE Code.

The SL-.2 FSAR states that the sample welds were not made using the same base metal plates as in the reactor pressure vessel beltline area.

The sample welds were produced using the same weld process, weld wire, and flux material as those used in the reactor vessel beltline.

However, the SL-2 FSAR does not indicate whether the test welds were subjected to heat treatment that would be metallurgically equivalent to those produced

.in the vessel material 5-2

throughout its fabrication process.

In order to evaluate the applicant's degree of compliance to this paragraph of Appendix G, 10 CFR Part 50, the applicant must indicate whether the heat treatment given to the sample welds results in metallurgical effects equivalent to those produced in the beltline vessel welds.

Paragraph IV.A.1 of Appendix G requires that a reference temperature,

RTN0T, be determined for each ferritic material of the reactor coolant pressure boundary and that this reference temperature be used as a basis for providing adequate margins of safety for reactor operation.

The value of RTNpT is defined in the ASME Code as the higher of either (a) the nil ductility temper-

ature, as determined by the dropweight test, or (b} a temperature of 60'F less than the temperature at which 50 ft-lb energy and 35 mils lateral expansion is

achieved, as determined by the CVN impact test.

The applicant has complied with these requirements except for ferritic welds and heat affected zones.

The applicant has reported the RTN>T'or the beltline welds, but has not supplied data to substantiate the RTNpT values.

In order to determine whether the beltline welds have been tested to the extent specified in Paragraph IV.A.1 of Appendix G, the applicant must provide CVN impact data and the drop weight NOT temperature for each beltline weld.

The applicant has also not supplied RTNpT data for any reactor coolant pres-sure boundary ferritic welds outside the beltline region.

The applicant must determine the RTNpT for all ferritic reactor coolant pressure boundary (RCPB) welds.

The applicant must identify the method for determining the RTN>T. If the method is different than that required by Paragraph IV.A.1 of Appendix G, the applicant must supply CVN impact data and drop weight data to substantiate the method.

The applicant must report the RTN>T of all ferritic RCPB welds outside the RV beltline, which may be limiting for operation of SL-2.

Paragraph IV.A.3 of Appendix G, in part, states that materials for piping, pumps and valves shall meet the requirements of NB-2332 of the ASME Code.

The ASME Code requires CVN impact and drop weight testing to determine the RTN0T for all piping pumps and valve materials greater than 2-1/2 inches in nominal 5-3

wall thickness.

The ASHE Code requires CVN impact testing at a temperature lower than or equal to the lowest service temperature for all piping, pumps and valve materials less than or equal to 2-1/2 inches in nominal wall thick-ness.

The piping must meet the CVN impact requirements of Table NB-2332-1 of the Summer 72 Addenda to the 1971 ASME Code.

The SL-2 FSAR states that the RTNOT values of piping for the Spray Nozzles (Code Nos.

M-9213-1 and M-9213-2), Let Oown Drain Nozzles (Code Nos.

M-9214-1, M-9214-2, M-9214-3 and M-9214-4) and Orain Nozzle (Code No. M-9215-1) are estimated at O'.

However, the applicant has not provided any information or data to substantiate this estimate.

In order to demonstrate these materials comply with ASME Code requirements the applicant must provide (a) all CVN impact and drop weight data from the materials, and (b) data and analysis from the literature which indicates the piping meets the ASME Code requirements identified above.

Paragraph IV.A.3 requires, in part, that material for bolting and other fasteners meet the fracture toughness requirements of Paragraph NB-2333 of the ASME Code.

The Summer 72 Addenda to the 1971 ASME Code requires CVN impact of three specimens at 40'F or lower for all bolting material greater than 1 inch nominal diameter.

The CVN test results must exceed 25 mils lateral expansion.

All RCPB bolting materials have met this requirement except the Pressurizer Manway Nuts (Number C-5364) and Pressurizer Manway Studs (Number C-5365).

For the Pressurizer Manway Nuts the CVN MLE (mils lateral expansion is 18 - 31 mils at +10'F.

For the Pressurizer Manway Studs the CVN impact energy absorbed was 54-57 ft-lbs at 10'F but the CVN MLE was not recorded.

To demonstrate compliance with Paragraph IV.A.3 of Appendix G the applicant must provide material data and analysis which indicates that Pressurizer Manway Studs and Nuts would meet ASME Code Requirements at 40'F test temperature.

Paragraph IV.B of Appendix G, 10 CFR Part 50 requires in part, that all beltline welds have a minimum unirradiated CVN impact test upper-shelf energy of 75 ft-lb unless it can be demonstrated by appropriate data or analyses that lower CVN impact test upper-shelf energies still provide adequate margins for 5-4

deterioration from irradiation.

The SL-2 FSAR does not give any CVN impact test data for the beltline welds.

The applicant has 'indicated the copper and phosphorus content for each beltline weld.

For beltline weld metal to meet the upper-shelf requirements of Paragraph IV.B of Appendix G, the applicant must either show that they possess 75 ft-lb of CVN impact energy at some test temperature; or by using Regulatory Guide 1.99, Revision 1, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials,"

show that a lower value of initial upper-shelf energy will still provide an end-of-life upper-shelf energy of 50 ft-lb at the quarter thickness vessel wall location.

To demonstrate compliance with Paragraph IY.B of Appendix G, the applicant must show that the affected weld metals would have attained the minimum required upper-shelf energy if this weld material had been tested at higher temperature.

To provide an acceptable basis for an exemption from the requirements of Paragraph IV.B, the applicant must provide data, information from the literature, and/or analyses to demonstrate that a margin of safety equivalent to that of the requirement has been attained.

h 5.3. 1.2 Evaluation of Com liance to A

endix H

10 CFR Part 50 The materials surveillance program at SL-2 will be used to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region, resulting from exposure to neutron irradiation and the thermal environment as required by General Design Criterion 32, "Inspec-tion of Reactor Coolant Pressure Boundary."

Under the SL-2 surveillance program, fracture toughness data must be obtained from material specimens that are representative of the limiting base,

weld, and heat-affected zone materials iri the beltline region.

These data will permit the determination of the conditions under which the vessel can be operated with adequate margins of safety against fracture throughout its service life.

5-5

The fracture toughness properties of reactor vessel beltline materials must be monitored throughout the service life of SL-2 by a materials surveillance program that meets the requirements of Appendix H of 10 CFR Part 50.

Me have evaluated the applicants'nformation for degree of compliance to these requirements.

Based on our evaluation we conclude that the applicant has met all the requirements of Appendix H, 10 CFR Part 50 with the exception of paragraph II.B.

Paragraph II.B. of Appendix H, 10 CFR Part 50 requires that a reactor beltline material surveillance program comply with ASTM E 185-73 except where modified by Appendix H, 10 CFR Part 50.

ASTM E 185-73 requires the following data for each capsule:

(1)

Actual surveillance materials in each capsule; (base metal:

heat

number, plate identification number; weld metal:

weld wire, heat of filler material, production welding conditions, and plate material used to make weld specimens).

(2)

Chemical composition of each test 'specimen.

(3)

The lead factor-for each capsule with respect to the vessel inner wall.

The SL-2 FSAR identifies the base metal test material, but does not identify the weld metal test material.

Furthermore, the lead factor for each capsule with respect to the vessel inner wall (paragraph 5 above) was not identified.

To demonstrate compliance with Paragraph II.B of Appendix H, 10 CFR Part 50, the applicant must supply the data detailed above.

Before we can evaluate the compliance of SL-2 to Paragraph II.B of Appendix K, 10 CFR Part 50, the applicant must supply the information listed above for each surveillance capsule and surveillance specimen.

5"6

5.3. 1.3 Conclusions for Com liance with A endices G and H

10 CFR Part 50 Based on our evaluation of compliance with Appendices G and H, 10 CFR Part 50, we conclude that the applicant has met all the fracture toughness requirements of these Appendices except for the following:

Paragraphs III.B.4, III.C.2, IV.A.l, IV.A.3, and IV.B of Appendix G and paragraphs II.B of Appendix H.

These items will remain open in our safety evaluation until the applicant submits the requested information.

Appendix G, "Protection Against Non-Ductile Failure,"Section III of the ASME Code, will be used, together with the fracture toughness test results required by Appendices G and H, 10 CFR Part 50, to calculate the pressure-temperature limitations for the SL-2 reactor vessel.

The fracture toughness tests required by the ASME Code and by Appendix G of 10 CFR Part 50 provide reasonable assurance that adequate safety margins against the possibility of non-ductile behavior or rapidly propagating fracture can be established for all pressure-retaining components of the reactor coolant boundary.

The use of Appendix G,Section III of the ASME Code, as a guide in establishing safe operating procedures, and use of the results of the fracture toughness tests performed in accordance with the ASME Code and NRC regulations, will provide adequate safety margins during operating, testing, maintenance, and anticipated transient conditions.

Compliance with these Code provisions and NRC regulations constitutes an acceptable basis for satisfying the requirements of General Design Criterion 31.

The materials surveillance

program, required by Appendix H, 10 CFR Part 50, will provide information on material properties and the effects of irradiation on material properties so that changes in the fracture toughness of the material in the SL-2 reactor vessel beltline caused by exposure to neutron radiation can be properly assessed, and adequate safety margins against the possibility of vessel failure can be provided.

5"7

Compliance with Appendix H, 10 CFR Part 50 assures that the surveillance program constitutes an acceptable basis for monitoring radiation induced changes in the fracture toughness of the reactor vessel material and satisfies the requirements of General Oesign Criteria 32.

5.3.2 Pressure-Tem erature Limits Appendix G, "Fracture Toughness Requirements,"

and Appendix H, "Reactor Vessel Material Surveillance Program Requirements,"

10 CFR Part 50, describe the conditions that require pressure-temperature limits for the reactor coolant pressure boundary and provide the general bases for these limits.

These appendices specifically require that pressure-temperature limits must provide safety margins for the reactor coolant pressure boundary at least as great as the safety margins recommended in the ASME Boiler and Pressure Vessel

Code,Section III, Appendix G, "Protection Against Nonductile Failure."

Appendix G, 10 CFR Part 50, requires additional safety margins whenever the reactor is critical, except for low-level physics tests.

The pressure-temperature limits imposed on the reactor coolant pressure boundary during the following operation and test activities are reviewed to ensure that they provide adequate safety margins against nonductile behavior or rapidly propagating fai lure of ferritic components as required by General Design Criterion 31:

(1)

Preservice hydrostatic tests, (2)

Inservice leak and hydrostatic tests, (3)

Heatup and cooldown operations, and (4)

Core operation.

Appendix G, 10 CFR Part 50, requires the applicant to determine the initial reference temperature, RTN0T, for all ferritic pressure-retaining materials in the reactor coolant pressure boundary.

Paragraph IV.A.2.a of Appendix G

specifies the procedures that are to be used to construct the pressure-temperature 5-8

limits for the ferritic components in the reactor coolant pressure boundary.

These procedures include definition of the initial reference temperature, RTNDT for the ferritie material s and consideration of the change in initial RTNDT due to neutron irradiati on.

As Secti on 5. 3. 1 of thi s safety eval uati on discusses, the applicant has provided acceptable methods to define initial RTNDT except for the wel d material.

Until the appl icant has suppl ied the information requested in Section 5.3.1, we cannot complete our evaluation of the pressure-temperature limits.

The pressure-temperature limits to be imposed on the reactor coolant system for all operating and testing conditions to ensure adequate safety margins against nonductile or rapidly propagating failure must be in conformance with established criteria, codes, and standards acceptable to the staff.

The use of operating limits based on these criteria, as defined by applicable regu-

lations, codes, and standards, will provide reasonable assurance that nonductile or rapidly propagating failure will not occur, and will constitute an acceptable basis for satisfying the applicable requirements of General Design Criterion 31.

5.3.3 Reactor Vessel Inte rit We have reviewed the FSAR sections related to the reactor vessel integrity of St.

Lucie Unit 2.

Although most areas are reviewed separately, reactor vessel integrity is of such importance that a special summary review of all factors relating to reactor vessel integrity is warranted.

We have reviewed the information in each area to ensure that it is complete and that no inconsistencies exist that would reduce the certainty of vessel integrity.

The areas reviewed are:

(1)

Design (SER 5.3.1)

(2)

Haterials of construction (SER 5.3.1)

(3)

Fabrication methods (SER 5.3. 1) 5-9

(4)

Operating conditions (SER 5.3.2)

We have reviewed the above factors contributing to the structural integrity of the reactor vessel and conclude that the applicant has complied with Appen-dices G and H, 10 CFR Part 50, except for the following items:

Paragraph III.B.4, Appendix G:

The applicant has not identified how the individuals performing the fracture toughness tests were trained and qualified.

Paragraph III.C,2, Appendix G:

The applicant has not shown that the tests specimens representing welds in the beltline region were heat treated to achieve equivalent metallurgical effects as the welds in the reactor vessel be 1tlinc.

Paragraph IV.A.l, Appendix G:

The applicant has not supplied CVN impact and drop weight data to substantiate the RTNDT values for each beltline weld.

Paragraph IV.A.3, Appendix G:

The applicant has not provided CVN impact and drop weight data to define the RTNDT for seven sections of pipe.

Paragraph IV.A.3, Appendix G:

The applicant has not provided data to demon-strate that the CVN impact test results at 10'F will meet minimum ASNE Code requirements at 40 F for pressurizer manway studs and bolts.

Paragraph IV.B, Appendix G:

The applicant has not provided data on CVN impact data on the unirradiated beltline welds to demonstrate the welds will have CVN

'pper shelf energy absorbtion of 75 ft. lbs.

Paragraph II.B, Appendix H:

The applicant has not identified the surveillance capsule materials and locations to demonstrate compliance with ASTM E 185-73.

Until the applicant has supplied the information necessary to complete our evaluation of compliance with Appendix G and H, 10 CFR Part 50, we cannot complete our evaluation of the structural integrity of the SL-2 reactor vessel.

5-10

5.4.1.1 Pum Fl heel Inte rit General Design Criterion 4, "Environmental and Missile Design Bases," of Appen-dix A, 10 CFR Part 50, requires, in part, that nuclear power plant structures,,

systems, and components important to safety be protected against the effects of missiles that might result from equipment failures.

Because reactor coolant pump flywheels have large masses and rotate at speeds of approximately 1200 revo-lutions per minute during normal operation, a loss of flywheel integrity could result in high energy missiles and excessive vibration of the reactor coolant pump assembly.

The safety consequences could be significant because of possible damage to the reactor coolant system, the containment, or the engineered safety features.

Adequate margins of safety and protection against the potential for damage from flywheel missiles can be achieved by the use of suitable material, ade-quate design, and inspection.

The flywheels have been fabricated from SA-543 Grade B, Class 1 steel.

This steel is of a quality which is suitable for improved fracture toughness.

The flywheel material has been produced by a process that will minimize flaws and improve fracture toughness, and has been cut, machined,

finished, and inspected in accordance with Section III of the ASME Code and Regulatory Guide 1.14.

The applicant has indicated that each pump flywheel will be inspected per the recommendations of paragraph C.4.b of Regulatory Guide 1. 14.

The reactor coolant pump has been designed for a speed 125% that of the normal synchronous speed of the motor (approximately 1500 rpm).

However, the minimum speed for failure is estimated to be much higher than 125% of operating speed for flywheels of the design used at SL-2.

The applicant has stated that the minimum fracture toughness of the flywheel material at normal. operating tempera-ture is equivalent to a dynamic stress intensity factor (KId) equal to or greater than 100 ksi in.

Me conclude that the reactor coolant pump flywheels in SL-2 possess a margin of safety against flywheel missiles equivalent to that recommended in 5-11

Regulatory Guide 1.14.

Compliance with Regulatory Guide 1.14 will provide a

basis acceptable to the staff for satisfying the requirements of General Design Criterion 4.

5-12

St.'ucie 82 Oocket 850-389 I

Request for Additional Information-OL FSAR-251.0 Materials En ineerin Branch -

Com onent Inte rit Section 251,2 Provide data on the qualifications of the personnel performing the fracture toughness tests that indicates the personnel were qualified by training and experience and had demonstrated competency to perform the tests in accord with written procedures of the component manufacturer.

If the above information cannot be provided or if the information provided does not comply with the requir'ements for training of personnel, state why the information can not be provided and identify why the methods used for training are equivalent to those required in paragraph III.B.4 of Appendix G.

251. 3 Identify the heat treatment given the test specimens for the welds in the core beltline region.

Indicate whether this heat treatment gives equivalent metallurgical effects as the actual core beltline welds.

251.4 For Melds Inside the RV Beltline Re ion a)

Provide CVN impact curves and the NDT temperature for each beltline weld.

b)

If sample material cannot be provided to the requirements of paragraph III.C.2 of Appendix G, provide:.

])

CVN impact test curves from samples which were produced using the same process, post weld heat treatment, electrode type, flux type and weld manufacturer as the SL-2 RV

beltline welds to demonstrate the welds would meet the fracture toughness requirements of paragraphs IY.A.1 and IYB of Appendix G, 10 CFR Part 50.

Sufficient CVN impact test data must be provided to assure the data is a conser-I vative representation of the RV beltline weld.

2)

CVN impact test data for RV beltline weld materials from each weld metal test per paragraph NB-2400 of the ASME Code.

1 For each weld adjacent to a nozzle, a flange and a shell region near a geometric discontinuity in the RV indicate the RTNDT and the method for determining the RTNDT. If the method for determining the RTNDT is different than that required by paragraph IV.A. 1 of Appendi x 6, provide data to demonstrate that the method is conservative.

For RCPB welds outside the RV indicate the RTNDT of the welds that are limiting for operation.

Indicate the method for determining the RTNDT. If the method for determining the RTNDT is different than that required by paragraph IV.A.1 of Appendix G, provide data to demonstrate that the method is conservative.

For Spray Nozzles (Code Nos.

H-9213-1 and M-9213-2),

Letdown Drain Nozzles (Code Nos.

M-9214-1, H-9214-2, M-9214-3, and M-9214-4) and Drain Nozzle (Code No. H-9215-1) piping:

. a)

Provide all CVN impact and drop weight test data from the material.

b)

If there were insufficient tests performed to meet the fracture toughness requirements of NB-2332 of the Summer 72 Addenda to the 1971 ASHE Code, provide CVN impact data and drop w i ght test data from other material which has been procured to the same material specification and heat treated to a metallurgically equivalent condition as the above identified piping.

The applicant must analyze the data to conservatively demonstrate that the SL-2 piping would have met the Summer 72 Addenda fracture

toughness requirements.

The applicant must also provide heat treatment information from the SL-2 piping and the additional piping which demonstrates the piping were heat treated to a metallurgically equivalent condition.

251:8 Provide a conservative demonstration for Pressurizer Hanway Nuts (Number C-5364) and Pressurizer Manway Studs (Number C-5365) that the material when tested at 40 F or lower will meet or exceed 25 mils lateral expansion.

Lower bound CVN curves for SA-193 Gr. 8-7 and SA-540 Gr. 8-24 materials are considered acceptable methods for extrapolating the CVN impact data from the test temperature to 40'F.

In addition, demonstrate that the metallurgical condition of the materials used to generate the lower bound curves for SA-193 Gr. 8-7 and SA-540 Gr. 8-24 materials are equivalent to the metallurgical condition of the SL-2 material.

This can be accomplished by providing the heat treatment information for the material used to generate the lower bound curves and for SL-2 Pressurizer Hanway Studs and Nuts.

251. g RCPB Heat-Affected Zones a)

Provide CVN impact data for all RCPB heat-affected zones I) in the RV which were not fabricated using submerged arc or covered electrode weld processes; and 2) outside the RV which were not fabricated using submerged arc or covered electrode weld processes and are limiting for RV operation.

b)

If the CVN impact data in B. 1 cannot be provided, estimate the

'TN>T from samples which duplicate the RCPB heat affected zones.

The samples'ase material must be fabricated from material that has been procured to the same specification as the SL-2 material and heat treated to the same requirements as the SL-2 material.

The sample's weld metal shall be produced using the same process,

post weld heat treatment, electrode type, flux type and weld manufacturer as the SL-2 weld.

The materials surveillance program uses six specimen capsules that should contain reactor vessel steel specimens of the limiting base

material, weld metal and heat-affected zone material.

To demonstrate compliance with Appendix H, 10 CFR Part 50, provide a table that includes the following information for each specimen:

a)

Actual surveillance material; b)

Origin of each surveillance specimen (base metal:

heat

number, plate identification number; weld metal:

weld wire, heat of filler material, production welding conditions, and plate material used to make weld specimens);

c)

Test specimen and type; d)

Chemical composition of each test specimen.

Provide the location, lead factor and withdrawal time for each specimen capsule calculated with respect to the vessel inner wall.

I

. LUCIE Dr. Robert E. Uhrig, Vice President Advanced systems and Technology Florida Power 8 Light Company P. 0.

Box 529100 Miami, Florida 33152 Harold F. Reis, Esq.

Lowenstein,

Newman, Reis, Axelrad 5 Toll 1025 Connecticut

Avenue, N.

W.

Washington, D. C.

20036 Norman A. Coll, Esq.

McCarthy, Steel, Hectory 8 Davis 14th Floor, First National Bank Building Miami, Florida 33131 Mr. Martin H. Hodder 1131 N.

E. 86th Street Miami, Florida 33138 Resident Inspector St. Lucie Nuclear Power Station c/o U. S. Nuclear Regulatory Commission 7900 South A1A Jensen Beach, Florida 33457