ML20062G935

Forwards Request for Addl Info by 790521 to Complete FSAR Review Re Matls Integrity,Reactor Fuels,Accident & Analysis, Emergency Planning,Geology/Seismology & Geotechnical Engineering

ML20062G935
Person / Time
Site:	South Texas
Issue date:	02/14/1979
From:	Parr O Office of Nuclear Reactor Regulation
To:	Eric Turner HOUSTON LIGHTING & POWER CO.
References
NUDOCS 7903080585
Download: ML20062G935 (24)

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UNITED $TATES g'

3 gi NUCLEAR REGULATOnY COMMISSION g

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JEB 141979 I

Docket Nos. STN 50-4987 and STN 0-499

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Mr. E. A. Turner

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Vice President

• I Houston Lighting and Power Company P. O. Box 1700 Houston, Texas 77001

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW 0F THE I -

SOUTH TEXAS FINAL SAFETY ANALYSIS REPORT (FSAR)

As a result of our continuing review of the South Texas FSAR, we find that we need additional information to complete our evaluation. The specific infonnation required is in the areas of materials integrity, reactor fuels, accident analysis, emergency planning, geology /

seismology and geotechnical engineering, and is listed in the Enclosure.

To maintain our licensing review schedule for the South Texas FSAR, we will need responses to the enclosed request by May 21, 1979.

If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to submit your responses so that we may review our schedule for any necessary changes.

Please contact us if you desire any discussion or clarification of the enclosed request.

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Sincerely, a'rr ef Light Water Reactors Branch No. 3 Division c. Project Management I

Enclosure:

As Stated cc w/ enclosure:

.See next page t

790308078[

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Mr'. E. A. Turner '

FEB 14 ISIS cc: Mr. D. G. Barker Mr. Troy C. Webb i

Manager, South Te ns Project Assistant Attorney General Houston L.ighting and Power Company Environmental Protection Div4 P. O. Box 1700 P. O. Box 12548 Houston, Texas 77G01 Capitol Station Mr. M. L. Borchelt Central Power and Light Company Mr. R. Gordon Gooch, Esq.

P. O. Box 2121 Baker & Botts Corpus Christi, Texas 78403 1701 Pennsylvania Avenue, N.W.

Washington, D. C.

20006 Mr. R. L. hancock City of Austin Director, Governor's Budget Electric Utility Department and Planeing Office P. O. Box 1088 Executive Office Building Austin, Texas 78767 41: W. 13th Street

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Mr. J. B.'Poston Assistant General Manager for Operations City Public Service Board P. O. Box 1771 San Antonio, Texas 78296 Mr. Jack R. Newnan, Esq.

Lowenstein, Newman, Axelrad & Toll 1025 Connecticut Avenue, N. W.

Washington, D. C.

20036 Mr. Melbert Schwarz, Jr., Esq.

Baker & Botts One Shell Plaza Houston, Texas 77002 k/

Mr. G. Hohmann Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pe.insylvania 15230 Mr. E. R. Schmidt i

NUS Corporation NUS-4 Research Place Rockville, Maryland 20850 Mr. J. H. Pepin

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Brown & Root, Inc.

P. O. Box 3 I

Houston, Texas 77001 l

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ad E?tCLOSURE REQUEST FOR ADDITI0tlAL IflFORMATI0ff FOR THE REVIE'.l 0F THE FSAR COR THE SOUTH TEXAS PROJECT, UtlITS 1 A*10 2

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121-1 121.0 MATERIALS ENGINEERIftG - fMTERIALS INTEGRITY 121.5 The inspection program.reauirenents, as detailed in Request i

(5.2.4)

No.121.1, have recently been revised to reflect information gained from recent inspection nrogram reviews. Therefore, 3

l Request No. 121.1 is now superseded by the following request.

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We still require that your inspection program for Class 1, 2 and 3 components be in accordance with the revised rules in 10 CFR Part 50, Section 50.55a, paragraph (g).

To evaluate your inspection program, the following minimum information is necessary for our review:

(1) A pruservice inspection plan to consist of the' applicable ASME Code Edition and the exceptions to the code requirements.

l (2) An inservice inspection plan submitted within six months of _ anticipated commercial operation.

i The preservice inspection plan will be reviewed to determine compliance with preservice and inservice inspection requirements.

The basis for the determination will be compliance with:

(1) The Edition of Section XI of the ASME Code stated in your FSAR or later Editions of Section XI referenced in the FEDERAL REGISTER that you may elect to apply.

(2) All augmented examinations established by the Connission when added assurance of, structural reliability was deemed necessary. Examples of augmented examination requirements can be found in staff positions on (a) high energy fluid systems in SRP Section 3.2, (b) turbine disk integrity in SRP Section 10.2,3, and (c) feedwater inlet nozzle inner L

radii.

Your response should define the applicable Section XI Edition (s) and subsections.

If any examination requirements of the Edition of Section XI in your FSAR can not be met, a relief request including complete technical justification to support your i

conclusion must be provided.

The inservice inspection plan should be submitted for review within six months of anticipated commercial operation to i

demonstrate compliance with 10 CFR Part 50, Section 50.55a, paragraph (g). Submittal at that time will permit you to incorporate Section XI requirements in effect six months prior

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l 121-2 121.5 to commercial operation and any augmented examination i

(5.2.4) requirements established by the Commission. Your response i

(Cont'd) should define all examination requirements that you determine are not practical within the limitations of design, geometry, and materials of construction of the compo,nents. 21-A provides detailed guidelines for the preparation and content of the inspection programs and relief requests to be submitted for staff review.

121.6 The response to Request No.121.3 cited the following references:

(10.2) 3.5-1 Shaffer, D.

H., S. C. Chay, D. K. McLain, and B. A.

Powell, " Analysis of the Probability of the Generation

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and Strike of Missiles From a Nuclear Turbine,"

i Mathematics Department, Westinghouse Research I aboratories, March 1974 for Steam Turbine Division Engineering, Westinghouse Electric Corporation (also Appendix 3.5-A, STP PSAR. Docket Nos. 50-498 and 50-499, Amendment 27 July 18, 1975).

3.5-2 Westinghouse Electric Corporation, " Missile Report for Turbines with 40 Inch Last Row Blades at Design, Inteme-diate and Destructive Overspeeds," (also Appendix 3.5-B, STP PSAR, Docket Nos. 50-498 and 50-499, Amendment 27, July 18, 1975).

3.5-3 Westinghouse Electric Corporation, "The Containment of Disc Burst Fragments by Cylindrical Shells " (also Appendix 3.5-C, STP PSAR, Docket Nos. 50-498 and 50-499,

(

Amendment 27, July 18, 1975); also published as ASME Paper No. 73-WA-Pwr-2 by A. C. Hagg and G. O. Sankey and as Scientific Paper 73-1E7-STGR0-P2, Mechanics Department, Westinghouse Research Laboratories, June 25, 1973, done for Steam Turbine Division Engineering, Westinghouse Electric Corporation by the same' authors.

3.5-4 "The Westinghouse Preservice Inspection and Test Progra'.s for Nuclear Turbine Rotors and Discs," West'inghousa Electric Corporation (also Appendix 3.5-D, STP PSAR, Docket Nos. 50-498 and 50-499, Amendment 27, July 18,1975).

3.5-9 Westinghouse Electric Corporation, " Turbine Disc and Rotor l

Integrity," Steam Turbine Information Section 17, l

CT-23989, Revision 1 (May 1977).

The contents in References 3.5-1, 3.5-2, 3.5-3, and 3.5-4 do not contain the specific infomation needed to evaluate the degree of l

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l 121.6 conformance of the STP turbines with SRP 10.2.3, " Turbine Disk (10.2)

Integrity." Reference 3.5-9 is not on file with the NRC.

(Cont'd)

Provide inf information necessary to shcw the degree of compliance with paragraph II " Acceptance Criteria," of SRP 10.2.3.

121.7 Provide a sketch of the STP 1 and 2 reactor vessels (including (5.2.3) dimensions) showing all longitudinal and circumferential welds, and all forgings and/or plates. Welds should be identified by a shop control number (such as a procedure qualification number), the heat of filler metal, type and batch of flux, and the welding process. Each forging and/or plate should be identified by a heat number and material specification.

121.8 Supply the following informatinn for each of the ferritic

_(5.2.3) materials of the pressure retaining components in the reactor coolant pressure boundary of the STP 1 and 2 plants.

i (1) The unirradiated mechanical properties as required by the testing programs in Section III of the ASME Code and Appendix G of 10 CFR Part 50 (test results to be presented should include Charpy V-notch, dropweight, lateral expansion, tensile, upper shelf energy T and RT If any of these properties have not been USIermined by)a*

N test method required by Appendix G of 10 CFR Part 50, state the actual test procedure used and/or the method used to estimate the test result together with a complete technical justification for the procedure used and tne associated test data.

(2) Identify the material (s) in the reactor coolant pressure boundary that will limit the pressure-temperature operating

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curves at the beginning-of-life.

For each reactor vessel beltlir.a weld, plate or forging provide the following additional information:

(3) The chemical composition (particularly the Cu, P and S i

content) and the maximum end-of-life fluence.

I (4) The relationship used to predict the shift in RT and l

percent decrease in upper shelf energy as a funcNn of t

neutron fluence.

1 (5) Identify the material (s) in the reactor coolant pressure boundary that will limit the pressure-temperature operating curves at the end-of-life.

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121.9 It is indicated in the STP FSAR Table 5.2-2, " Class 1 Primary i

(5.2)

Components Material Specifications," that SA 533 Grade A 2

Class 2 and SA 508 Class 2a steel might be used in the pres-surizer and sterm generators. These steels are low alloy i

ferritic steel having minimum yield strengths of 70 ksi and j

j 65 ksi, respectively. Appendix G to 10 CFR Part 50 s.tates that i

the adequacy of the fracture toughness of ferritic. steels having t

a minimum specified yield strength greater than 50 ksi shall be i

demonstrated to the Consnission on an individual case basis.

i We require that the information required by Note i c? Code Case 1528-3, later incorporated into the ASME Code as-paragraph G-2110(b), be provided in the FSAR if any high strength I

ferritic material is to be used in a pressure-retaining component of the RCPB.

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Sections 5.2.1.2 and 5.2.3.3.1 of STP FSAR reference a letter (NS-CE-1228, C. Eicheldinger, Westinghouse PWRSD Nuclear Safety, to J. F. Stolz,10/4/76) which presents preliminary fracture toughness data for base metal, weld metal and I. cat affected One metal for high strength ferritic materials used in the RCP8.

The information presented in this letter is insufficient to r

meet the requirements stated above.

Therefore, we require

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additional information to complete & review.

In this regard, Westinghouse subsequently issued a topical report (WCAP-9292 " Dynamic Fracture Tc1ghness of ASME SA 508 Class 2a ASME SA 533 Grade A Class 2 Base and Heat Affected l

Zone Material and Applicable Weld Metals") on this matter, which we are currently reviewing. A reference to this topical 4

report will provide an acceptable response to this request.

It should be noted, however, that if the topical report is b

found to be acceptable, it will only satisfy the generic requirements of Appendix G to 10 CFR Part 50. That is, the adequacy of the subject materials to be described by the K f

cursa of Appendix G to the ASME Code will have'been demon IR strated; however, the specific materials to be used in the STP-i nuclear plants must meet all the fracture roughness requirements.

i of 10 CFR Part 50, Appendix G.

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121.10 Some of the infomation required by the above Request Mos.121.7 4

(16.3/4) and 121.8 is in the STP Technical Specifications; however,.it is L

not readily apparent which unit the supplied data refers to.

Request Nos.121.7 and 121.8 are directed at both units and thus, I

the information that is required should be determined and' tabulated for each unit separately. Revise Figures 3.4-2 and 3.4-3 and Table 16.B3/4.4-1 to reflect this.

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(5.3.2)

. The pressure-temperature limit calculation methods given in the-Technical Specifications are those given in Topical Report J

(16.3/4)

WCAP-7924A. The NRC staff has reviewed and accepted this RSP report with the following exception. The evaluation stipulated that the method for detennining the shift in RT is not acceptable and.that an acceptable method must bh0 Included in 2

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the FSAR. Section 5.3.2.1 of the STP FSAR presented an I

alternate method of determining the shift in RT as a function j

of fluence. This method has been evaluated and Nund unacceptable.

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It is 'our position that all of the methods recommended in Revision 1 to Regulatory Guide l.99, " Effects of Residual Elements on Pre < Meted Radiation Damage to Reactor Vessel Materials" be used to evaluate radiation damage to the reactor vessel materials of STP Units 1 and 2.

Revise the FSAR accordingly.

k 121.12 Paragraph B 3/4.4.9.d of the STP Technical Specifications (16.3/4).

indicates that the pressurizer heatup and cooldown rates shall not exceed 200*F/hr. Westinghouse has notified NRC that the design hatup rate for their pressurizers is 100*F/hr and that the heatup rate in the Technical Specifications should not t

l exceed this value.

Paragraph a 3/4.4.9.d should be revised to incorporate this change (paragraph 3.4.9.2 of the Techicdl Specifications already state; the correct heatup and cooldown rates). Section S.7 of the Technical Specifications, when submitted in the STP FSAR, should also reflect this change.

i 121.13 Provide sufficient information to demonstrate that the materials (5.3.1.6) surveillance programs for STP 1 and 2 comply fully with all of the requirements of Appendix H to 10 CFR Part 50.

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"121.14 Confirm that all bolting and other fasteners, used in the RCPB (5.3.1.7) of STP Units 1 and 2, with nominal diameters exceeding 1 inch, meet the minimum requirements of 25 mils lateral expansion and 45 ft-lbs in terms of Charpy V-notch tests conducted at the preload temperature or at the lowest service temperature, l

j whichever is lower (10 CFR Part 50, Appendix G, paragraph IV.A.4).

j 121.15 STP FSAR Table 5.2-2 lists bolting material for tne STP Units 1

. i (5.2) and 2 reactor vessels as SA 193 Grade B7 and FSAR Section 5.3.1.7-t (5.3.1.7) lists the bolting mater'ial as SA 540 Class 3 Grade B24. Clarify

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this discrepancy.

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121.16 Confirm that the reactor vessel fasteners for Units 1 and 2 (5.2.4) will be inspected according to the requirements of Sections'.III (5.3.1.7) and XI of the ASME Code as supplemented by Regulatory Guide 1.65

" Materials and Inspections for Reactor Vessel Closure Studs."

121.17 Supply the tested ultimate tensile strength of the reactor (5.3.1.7) vessel fasteners, as indicated in Regulatory Guide 1.65.

121.18 STP FSAR Section 6.4.1.5.2 states that access to the reactor i

I (5.2.4) coolant pump flywheel has been provided to allow inspection in (5.4.1.5) accordanct with the requirements of Section XI of the ASME Code and the recomendations of Regulatory Guide 1.14.

This FSAR section also states that a description of the pump flywheel inspection program is presented in F9 R Section 5.2.4.

The program description is not in this referenced FSAR section.

Supply this informat$c.n.

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121.19 STP FSAR Section 5.4.!.5.2 states that certification reports are (5.4.1.5) available for all plates, from which the reactor coolant pump flywheels were fabricated, that demonstrate the acceptability of the flywheel material on the basis of the recommendations of Regulatory Guide 1.14.

Supply this information.

121.20 STP FSAR Section 5.4.2.2 states that any discrepancies between (5.4.2) the Westinghouse steam generator inspection program and that presented in Regulatory Guide 1.83, Revision 1 are discussed in the position paper in FSAR Section 3.2.

This position paper is r

not in the referenced FSAR section. Supply this information.

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ATTACHMENT 121-A i

GUIDAW.E FOR PREPARING PRESERVICE AND INSERVICE INSPECTION l

PROGRAMS AND RELIEF REQUEST PURSUANT TO 10 CFR 50.55a(g)

A.

Preservice/ Inservice Inspection Program Description j

This program covers the requirements set forth in 10 CFR 50.55a(g) and the ASME Sciler and Pressure Vessel Code Section XI, Sub-sections IWA, IWB, IWC and IWD.

The guidance provided in this enclosure is intended to illustrate the type and extent of infomation that should be provided for NRC review.

It also describes the infomation necessary for " request. for relief" of items that cannot be fully inspected to the requirements of ASME

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Section XI. By utilizing these guidelines, licensees can signifi-cantly reduce the need for having to respond to additional infomation requests from the NRS staff.

B.

Contents of the Submittal i

The infomation listed below should be ir.cluded in the submittal:

1.

For each facility, include the applicable ASME B & P V Code date j

and appropriate addendum date.

2.

The period and interval for which this program is applicable.

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3.

Include the proposed codes and addenda to be used for repairs, modifications, additions or alternations to the facility that might occur during this inspection period.

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4.

Identify the examinations that you have exempted under the rules of ASME Section XI. A reference to the applicable paragraph of the code that grants the exemption is satisfactory. The inspect-ion requirements for exempt components should be shown; i.e.,

visual inspection during a pressure test.

5.

Icentify the inspection and pressure testing requirements of the applicable Section XI requirements that are deemed impractical because of the limitations of design, geometry and material of construction of the components. Provide the infomation requested in paragraph C for the inspections and pressure tests identified.

l C.

Request for Relief from Certain Inspection and Testing Recuirements It has been the staff's experience that many requests for relief from testing requirements submitted by licensees have not been supported by adequate descriptive and detailed technical infomation. This detailed f

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' f information is necessary to document the impracticality of thi ASME Coda requirements within the limitations of design, georetry and materials of construction of componeats and to determine whether the t

use of alternatives will provide an acceptable level of quality and safety.

Relief requests submitted with a justification such as " impractical "

" inaccessible;" or any other categorical basis, require additional information to permit the staff to make an evaluation of that relief request. The objective of the guidance set forth below is to illustrate the extent of the information that is required by the NRC staff te make a proper evaluation and to adequately document the basis for granting the relief in the safety evaluation report. The NRC r

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staff believes subsequent requests for additional information and delays in completing the review can be considerably reduced, if this information is provided initially in the licensee's submittal.

For each relief request submitted, the following infonnation should be included:

i 1.

Identification the component (s) and/or the examination require-ment for which relief is requested.

2.

Number of items associated with the requested relief.

3.

ASME Code class.

4.

Identification of the specific ASME Code requirement that has been determined to be impractical.

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5.

Information to support the determination that the requirement is impractical; i.e., state and explain the basis for requesting relief.

6.

Identification of the alternative examinations that are proposed in lieu of Section XI requirements or to supplement partially j

performed Section XI examinations.

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Description and justification of any changes expected in the overall level of plant safety by performing the proposed

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alternative examinations in lieu of the ASME Section XI examina tion. If it is not possible to perform alternate examinations, discuss the impact on the overall level of plant quality and safety.

For inservice inspection provide the following additional information regarding the inspection frequency:

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State when the relief request would apply during the inspection period or interval; i.e., is the request to defer an examination.

9.

State when the proposed alternative examinations will be imple-mented, and perfomed.

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10. State the time period for which the requested relief is needed.

Technical justification or data must be submitted to support the relief request. Opinions without substantiation that a change will i

not affect the quality level are unsatisfactory. If the relief is I

requested for inaccessibility, a detailed description or drawing which depicts the inaccessibility must accompany the request. A i

relief request is not required for tests prescribed in Section XI e

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that do not apply to your facility. A statement of N/A (not applic-able) or none will suffice.

D.

Request for Relief for Radiation Considerations Exposures cf test personnel to radiation to accomplish the examinations prescribed in ASME Section XI can be an important factor in detennining wheths.r or under what condition an examination must be performed. A requot for relief must be submitted and approved similar to that requ ked for inaccessibility.

t We recognize that some of the radiation considerations will only be known at the time of the test. However, the licensee generally is eware, from experience at operating facilities', of those areas where relief is necessary and should submit as a minimum the following j

information with the request for relief:

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1.

Total estimated man-rem exDosure involved in the examination.

2.

Radiation levels at the test area.

3.

Flushing or shielding capabilities which might reduce radiation

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levels.

j 4.

Alternate inspection techniques propsed.

5.

Remote inspections considerations.

6.

Redundant systems or similar welds which can be inspected.

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Preservice and any inservice results of welds involved.

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Consequences if the weld failed.

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i 231-1 231.0 REACTOR FUELS j

231.1 The fuel performance code (PAD 3.3) used for the South Texas 1

(4.2.1.3) j safety analyses has recently oeen approved by us with some restrictions which are identified in a letter dated February 9, 1979, from J. Stolz, NRC to T. Anderson, Westinghouse. Provide an assessment as to whether those restrictions effect the results of the safety analyses presented in the FSAR. If so, revise the analyses accordingly and provide the results.

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231.2 Predicted cladding collapse times for South Texas 1 and 2 have (4.2.1.3) been calculated with the model as given in WCAP-8377. We have approved the use of this model, by letter to Westinghouse dated February 14, 1975, subject to provisions that no alternations

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were rade to the specified curves used as input to the model.

Provide assurance that these provisions have been satisifed.

231.3

. Provide the basis for the internal fuel rod gas pressure (4.2.1.3)

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criteria presented in the FSAR. Ne note that these criteria are the same as those approved in our review of WCAP-8963 (see l

letter from J. Stolz, NRC to T. Anderson, Westinghouse) in l

which an acceptable basis is provided. Therefore,a reference i

j to WCAP-8963 will provide an acceptable response to this reauest.

f Due to the restrictions to the fuel performance code (PAD 3.3)

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imposed at high burnups, as discussed in-Request No. 231.1 above, determine the effects of these restrictions on satisfying the i

l rod pressure criteria.

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i 231-2 231.4 In addition to the seismic and hydraulic loads on fuel assemblies, (4.2.1.4)

Westinghouse has determined that an asymetric horizontal load I

j will be imposed on the reactor core in the event of a rupture in the primary system piping. Provide the results of an analysis l

which shows that the S J Texas fuel assemblies can withstand this additional horizontal load. Similar analyses have been submitted on the North Anna, Sequoyah, Farley and Diablo Canyon dockets and these can be referenced for information. This analysis should be performed with the best available methodology and criteria for the South Texas design.

s We are currently reviewing a generic analysis method to account for the horizontal asymmetric load (see Task A-2, NUREG-0371, dated November 1978). If the methodology or criteria develooed j

from this task were to result in a need for a reevaluation on South Texas, at that time we would require that you perform such a reevaluation.

{'6. 2.1. 6) l.5 The full length absorber rods for South Texas ir. corporate a new e

Westinghouse design feature. The technical bases and evaluation of this new rod design are given in a topical report, WCAP-8846.

NRC has reviewed the noted topical report and found the rod i

design acceptable but noted that irradiated B C is leachable or 4

soluble in water. Thus, we require a routine surveillance program to assure that the reactivity invested in the absorber i

rods is not being lost through an unanticipated breach in the I

cladding. Submit your plans for such a surveillance program.

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i 231.6 A new requirement for routine fuel surveillance has been i

(4.2.1.7) established and is discussed in Revision 1 of Section 4.2 i

of the Standard Review Plan. The new requirement places special emphasis on surveillance when new design features are introduced. South Texas will be the first plant to use the Nestir.ghouse 14-foot 17x17 assemblies and these will differ i' rom the 17x1712-foot design, specifically in t.'1e holddown and lift-off forces and the grid-to-grid span length. Therefore,

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in light of the new recuirement, submit a description of the post-irradiation fuel surveillance program and a description of the on-line fuel rod failure detection methods to be used for South Texas Units 1 and 2.

231.7 Recent PWR experience has shown that fretting wear has occurred (4.2.2.2) between control rods and thimble tubes at a location associated with the fully withdrawn " parked" rod position. Provide assurance, either through fuel assembly inspection results.

(see item 231.6 above) or prototypical hydraulic flow tests k~i (see item 231.8 below), that fretting wear is not a concern in South Texas.

231.8 WCAP-8278 is used to demonstrate that the design methods for (4.2.3.1) predicting vibration amplitudes and fuel rod frettinn wear are conservative, based upon a 1000-hour flow loop test of a 12-foot 17x17 fuel assembly. Similar verification tests for the 14-foot 17x17 fuel have been previously discussed by Westinghouse and were to be completed in 1978. Provide assurance that fuel rod fretting for the 14-foot fuel assembly is not a concern in South Texas.

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231.9 For fuel rod bowing calculations, South Texas uses the model (4.2.3.1) given in WCAP-8691. However, we have been unable to complete 1

our review of WCAP-8691 because of several inadequacies (discussed in a letter from J. F. Stolz, flRC to T. 'M. Anderson, Westinghouse, dated June 19, 1978) and have requested Westinghouse to submit a new report to address those

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inadequacies. In an enclosure to the noted letter, we provided an acceptable interim method for treating the effects-(

of rod bow on CHF. Therefore, provide the results of a reanalysis of the effects of rod bow based on that interim j

method.

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231.10 In view of recent reactor operational experience, n control (4.2.3.6) rods (Al 0 -8 C burnable poison in St. Lucie 1 CErl-38(F),

23 4 Revision 1), the stability of irradiated B C in water appears 4

to be less than that expected from the previous referenced studies. Therefore, the corrosion ratas given in the South k.)

Texas FSAR would seem to be non-conservative. Provide justification for the stated corrosion rates by using either additional suitable references or recent experimental data.

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j 312-1 312.0 ACCIDENT ANALYSIS i

I 312.12 The ratio of spray ' additive flow to barated water flow (6.2.2 &

determines the pH of the spray solution. Provide the 6.5.2) magnitude of this ratio. What procedures will be used to verify (during pre-operational and routine testing) that the two #10w rates would indeed be maintained at this ratio?

312.13 It does not appear that heat tracing is provided in the (6.2.2 &

chemical additive tank, even though the NaOH solution may 6.5 2) have a concentration greater than 30 weight percent.

Provide your reason for the omission of heat tracing, or provide suitable design changes.

312.14 For the control rod ejection accident, provide curves (15.4.8) showing the primary and secondary system pressures versus

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313.0 EMERGENCY PLANNING i

313.1 Provide definitions of the phrases "STP operating limits" j

(1.0) and "STP procedure specifications" used in your plans.

Characterize these as to their relationship to Technical Specifications.

313.2 Provide a summary of how your plans relate to those of the (3.0) state and local agencies, especially with regard to emergency classifications, notifications required and immediate actions planned. Based on the results of our previous reviews of other

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plants, we suggest that preparation of a table using the following column headings would be most productive:

(1) Incident or Accident (2) Probable Means of Detection (3) Probable Emergency Class (plant or state names)

(4) Immediate Notifications Required (5) Immediate Actions Required (onsite and offsite) 313.3 Include provisions in your emergency plans fur appropriate (4.1.4.3)

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actions to cope with unplanned r21 eases of liquids from your site by any route. We do not agree that the only conceivable route is via the cooling reservoir; your drainage ditch to the Colorado River and clean drains are examples of other conceivable routes.

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,l 313-2 313.4 For each postulated accident in Section 4.2, indicate the (4.2)

State of Texas Incident Level (n.b. Appendix E of your plans, pp 4-5) to which the accident would be initially assigned.

If the consequences of an accident could result in its being classified as a Site or General Emergency, prescribe the t

emergency action levels at which it would surely be classified as either i Site or General Emergency.

(Note that emergency action levels are not projected doses and vice versa; see definitions in R.G. 1.101, Revision 1, Annex A.)

313.5 In case of an emergency on the night shift which incapacitates (5.2.1) the Watch Supervisor-Nuclear, state which on-shift personnel would assume the role of Emergency Director until higher authority can arrive at the plant.

313.6 Your prescription for notification of the state agencies should (6.1.5.2)

,I recognize the information content expected by the state (n.b.

page 13 of Appendix E of ynur Emergency Plan). In this regard,

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a najor objective of your coordination with the state should be to arrive at a clear understanding of the import of your initial classification of an emergency and the predetermined actions, if any, that should necessarily follow off-site.

In particular, your General Emergency (State Level A Emergency) could well be l

classified and announced even before a major release actually occurred, in which case some of the data expected by the state would not be available or would be of little value (e.g. curies released) in the immediate time frame. Revise your plan accordingly.

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313.7 Elaborate regarding the lock and key control for the emergency (7.2.2) e

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communications to the sheriff's department to show that the j

equipment would be accessible by others if the Watch Super-visor were incapacitated or otherwise unavailable in an emergency.

313.8 Describe the assumptions and models used to generate the (Appendix G) curves in Figures G-3 and G-4.

313.9 Emergency action levels for the guidance of emergency monitoring (6.4) teams should be included in your plans. These should be

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specified in terms of contamination levels in water, soil, vegetation and milk.

(n.b. U. S. DHEW draft recommendations, re: Accidental Radioactive Contamination of Human and Animal Feeds, 43 CFR 58790, December 15,1978.) In particular, provide the dose rate at three feet above pasture land at which you would recommend removal of grazing dairy animals from pasture to protect milk, following a major fission product release.

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361.0 GE0 LOGY / SEISMOLOGY 361. 8 Piezometric level declines (deep aquifer) experienced at the (2.5.C)

STP site generally exceed the estimates shown on PSAP.

Figure 2.5.1-15G. This figure indicates a projected six foot decline between January, 1973 and June, 1978. The actual decline, utilizing the data prewited in PSAR Figure 2.5.1-16D and FSAR Figure 2.5.C-22, ranges from

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seven feet at Piezometer 607 to 27 feet at Piezometer 604.

Eleven feet of decline has occurred directly beneath the plant site in a 21/2 year period (tiovember,1975 to June,1978). Discuss the impact of this rapid groundwater decline (especially directly beneath the structures area)

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on the subsidence estimates given on FSAR page 2.5.1-125.

Discuss the significance, if any, of this decline with respect to site safety.

,361.9 Figures 2.5.C-23 and 2.5.C-24 of the FSAR provide a plot

(>'2.5.C) of the combined horizontal strain - regional subt idence data obtained through June 1978 from your subsidence monitoring program. Submit' revised figures to include the data obtained during December 1978. We request that these figures be revised periodically during construction and throughout the life of the plant and be submitted in accordance with the horizontal monitoring frequencies (i.e., semi-annually and annually) sr

_.----..w-..u 361-2 361.9 suggested on Table 2.5.C-1 of the FSAR. In addition to 2.5.C)

Cont'd) submitting revised figures, provide appropriate discussions j

relative to the reasons for, and the safety significance of, the observed horizontal and vertical movements.

361.10 Provide a composite figure incorporating subsidence contours (2.5.C) derived from observations taken from the monuments shown on FSAR Figure 2.5.C-1 and piezometric levels derived from the g

deep aquifer piezometers shown on FSAR Figure 2.5.C-18.

The 1973 piezometric levels (PSAR Figure 2.5.1-16D) are to be used as a base for the piezametric decline. We request that revisions to this figure be submitted periodically to coincide with the submittal of the combined Horizontal Strain-Regional Subsidence Monitoring data described in Request No. 361.9 above. With each subsequent submittal, describe the suspected relationship between the observed subsidence and the deep aquifer piezometric level decline.

b61.11 The response to Request No. 361.1 includes a reference to (2.5.1) page 2.5.1-74a.

This page was apparently not submitted.

Correct this omission.

361.12 The Cambe Geological Service Map No. T-7, provided in response (2.5.1) to Request No. 361.1, indicates an additional site well not shown on PSAR Figure 2.5.1-32.

This previously unidentified l

well (Robbins #1) is located approximately one mile east of Well #16 (PSAR Figure 2.5.1-32).

Provide all pertinent data relative to the Robbins #1 well. Revise appropriate portions of the FSAR accordingly.

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t 361-3 361.13 Provide the following clarifications:

(2.5)

(1) On page 2.5.2-15, the references to Subsection 2.5.2.5 should be changed to Subsection 2.5.2.1, and the reference to Subsection 2.5.2.6 should be changed to Subsection 2.5.2.2.

(2) Table 2.5-23 in the FSAR should be updated to include earthquakes since 1969.

i (3) Paragraphs three and four on page 2.5.2-20 should be included in Subsection 2.5.2.6.

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362-1 362.0 GEOTECHNICAL ENGINEERIflG 362.1 Provide additional details regarding the vibroflotation densifi-(2.5.4.12) cation program for the Unit No. 1 Fuel Handling 9uilding.

(1) Identify the location of the nine 84-inch diameter casings installed for the Fuel Handling Building.

(2) Provide any fisld reports that discuss the densification g.,

program.

_ (3) Provide the chronology regarding the detection and treatment of the foundation disturbance.

(4) Provide any photos that may be pertinent.

_(5) Explain the difference between Unit 1 and Unit 2 procedures that avoided the recurrence of this problem.

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