ML20032A038
| ML20032A038 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0787, NUREG-0787-S01, NUREG-787, NUREG-787-S1, NUDOCS 8110280181 | |
| Download: ML20032A038 (75) | |
Text
NUREG-0787 Supplement No.1 l
Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power & Light Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation h< f<d *~ %,
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
1.
The NRC Public Documerg Room,1717 H Street., N.W.
Washington, DC 20555 2.
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l GPO Printed copy price: $4.25
l NUREG-0787 Supplement No.1 Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power & Light Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1981
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10/13/81 TABLE OF CONTENTS PaSe 1 INTRODUCTION AND GENERAL DISCUSSION.................
1-1 1.1 Introduction........
1-1
- 1. 7 Summary of Outstanding Issues 1-1
- 1. 8 Confirmatory Issues 1-2
- 1. 9 License Conditions......................
1-3 i
2 SITE CHARACTERISTICS 2-1
- 2. 2 Nearby Industrial, Transportation, and Military Facilities..
2-1 2.2.1 Nearby Industries...................
2-1 l
3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT AND SYSTEMS 3-1 i
3.7 Seismic Design.
3-1 3.7.2 Seismic System Analysis................
3-1 3.8 Design of Category I Structures 3-1 3.8.2 Steel Containment...................
3-1 3.8.5 Foundations......................
3-1 4 REACTOR...............................
4-1*
4.4 Thermal-Hydraulic Design...................
4-1 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5-1 5.3 Reactor Vessel........................
5-1 5.3.1 Reactor Vessel Materials 5-1
- 5. 3.1.1 Compliance With Appendix G, 10 CFR Part 50...
5-2 5.3.1.2 Compliance With Appendix H, 10 CFR Part 50...
5-7 5.3.1.3 Conclusions for Compliance With Appendices G and H, 10 CFR Part 50 5-9 5.3.2 Pressure - Temperature Limits.............
5-9 5.3.3 Reactor Vessel Integrity 5-10 5.4 Component and Subsystem Design................
5-12 5.4.3 Shutdown Cooling (Residual Heat Removal) System....
5-12 Waterford SSER #1 III
TABLE OF CONTENTS (Cont'd)
-Page 6
ENGINEERED SAFETY FEATURES.....................
~6 6.1 Engineerco Safety Features Materials.............
6-1 6.1.2 Organic Materials...................
6-1 6-1 6.2 Containment Systems 6.2.3 Secondary Containment Functional Design........
6-1 7. INSTRUMENTATIO!1 AND CONTROL.....................
7-1 7.3 Engineered Safety Features Actuation System 7-1 7.5 Safety-Related Display Instrumentation............
7-3
- 7. 7 Control Systems Not Required ~for Safety 7-4 13. CONDUCT OF OPERATIONS 13-1 13.3 Emergency Preparedness Evaluation......
13-1 13.3.1 Introduction.....................
13-1 13.3.2 Evaluation of Applicant's Emergency Plan.......
13-1 13.5 Plant Procedures 13-4 e
13.5.2 Operating and Maintenance Procedures.........
13-4 15-1 15 ACCIDENT ANALYSIS 15-1 15.3 Limitirg Accidents 15-1 15.3.1 5 team Line Breahs 15.3.2 Feedwater Systea 9fpe Breaks.............
_15-2 15.3. 4_ S tea.a Generatar Tube Rupture.............
15-3 15.3.5 Anticipdtad Transients without Scram.........
15-4 18 REVIEW BY THE ADVISORY COV.MITIEC CM REACTOR SAFEGUARDS.......
18-1 20-1 20 FINANCIAL QUALIFICATIONS 20.1 Business at Applicant 20-1 20.4 Rean -an t e f.sserm ce of Fur:cs 20-1 20.4.1 General 20-1 20-2 20.5 Coaci.o Waterford SSER #1 iv
i TABLE OF CONTENTS (Cont'd)
Page 22 TMI-2 REQUIREMENTS.....................,...
22-1 22.2 Discussion of Requirements 22-1 APPENDICES l
A CONTINUATION OF THE CHRON0 LOGY OF RADIOLOGICAL REVIEW A-1 B CONTINUATION OF BIBLIOGRAPHY B-1 C ACRS LETTER DATED AUGUST 11, 1981 C-1 LIST OF TABLES 20-1 Louisiana Power & Light Company and New Orleans Public Service Inc.
Pro Form Consolidated Statement of Capitalization, June 30, 1981..
20-3 20-2 Louisiana Power & Light Company and New Orleans Public Service Inc.
Pro Forms Consolidated Statement of Income for the Twelve Months Ended June 30, 1981........................
20-4 i
i Waterford SSER #1 v
J f
1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On July 9,1981, the Nuclear Regulatory Commission (NRC) staff issued a Safety Evaluation Report (SER) related to the operation of Waterford Steam Electric Station, Unit No. 3.
In that SER, the staff indicated certain issues where either further information or additional staff effort was necessc y to complete the review.
The purpose of this supplement is to update the SER by providing (1) our evalua-tion of additional information submitted by the applicant since the SER was issued, (2) our evaluation of the matters the staff had under review when the SER was issued, and (3) our response to comments made by the Advisory Committee on Reactor Safeguards in its report dated August 11, 1981.
Each of the following secticns of this supplement is numbered the same as the section of the SER that is being updated, and the discussions are supplementary to and not in lieu of the discussion in the SER.
Appendix A to this supplement is a continuation of the chronology.
Appendix B is an updated bibiliography.
- 1. 7 Summary of Outstanding Issues Section 1.7 of the SER contained a list of outstanding items.
This supplement addresses the resolution of a number of these items previously identified as open.
These are listed below, along with the section of this report wherein their resolution is discussed.
(1) Appendices G & H (5.3.1)
(2) Steam voiding in reactor vessel analysis (15.3)
(3) Feedwater line break analysis (15.3.2) becomes confirmatory (4) Loss of offsite power or tripping of the reactor coolant pumps during a main steam line break (15.3.1)
(5) Clarification of transient analyses with potential for fuel damage (15.2.1) becomes confirmatory (6) Pressure transient in shield building annulus (6.2.3)
(7) Site hazards (toxic gas) (2.2)
(8) Cold shutdown without leaving control room (5.4.3)
(9) Emergency feedwater control (7.3) becomes confirmatory (10) I & E Bulletin 80-06 (7.3)
(11) Single failure of control system (7.7) becomes a License Condition (12) Organic materials (6.1.2)
(13) Operating procedures (1.C tasks) (22)
(14) Valve position indication (7.5)
(15) Emergency planning (13.3)
(16) Thermal-hydraulic design (4.4)
(17) Plant Procedures (13.5)
At this time, there remain a number of safety issues that have not yet been resolved.
These will be addressed in a subsequent supplement to the SER. The following is a list of these items.
Waterford SSER #1 1-1 1
(1) Fi re protection ( 7. 4, 7. 5, 7. 7, 8. 3. 3, 9. 5.1, 9. 5. 2, 9. 5. 31 (2) Licensee qualifications (13.1, 13.2, 13.4)
(3) PSI /ISI (3.9.6, 5.2.4, 6.6)
(4) Emergency planning (13.3)
(5) Environmental qualifications (3.11)
(6) Seismic qualification (3.10)
(7) Reactor coolant pump shaft break analysis (7.1, 7.2, 7.3, 7.5, 15.2.3.1)
(8) Thermal-hydraulic design (4.4)
(9) Turbine missiles (3.5.1.3, 3.5.3)
(10) Indemnity requirements (21)
(11) Site hazards (explosions) (2.2)
(12) Q-List (17)
(13) TMI issues Operational Safety (I.T. 1 tasks)
Organization and Management (I.B.1.2)
Operating Procedures (I.C. tasks - long term)
Control Room Review (I.D.1)
Containment System Design (II.E.4.2)
ICC Instrumentation (II.F.2)
- 1. 8 Confirmatory Issues At the time of the SER issuance there were several issues which were essentially resolved to the staff's satisfaction, but for which certain confirmatory inform-ation had not yet been provided by the applicant.
Since that time, the staff has reviewed this information and as expected, have confirmed the prelimin-ary conclusions.
These issues are listed below with appropriate references to subsections of this report.
(1) Revised analysis of shield building (-5WG), (6.2.3)
(2) Reanalysis of Category I structure (3.7.3)
(3) Reevaluate foundation mat (3.8.5)
(4) Welding - justification of extrapolation (3.8.2)
At this time several issues remain for which the staff has not yet received the necessary confirmatory information.
These issues, which are listed below, will br addressed in a subsequent supplement to the SER.
(1) Testing the ultimate heat sink (2.4)
(2) Piping analysis (3.9.1, 3.9.2)
(3) Containment isolation actuation signal (7.3)
(4) Performance of PWP, relief and safety valves (22)
(5) Security plan (13.6)
(6) Sizing of primary safety valves (5.2.2)
(7) Shutdown initiation using safety grade equipment (5.4.3)
(8) Containment sump vortex test (6.3.3)
(9) Reactor coolant pump s W t reizure analysis (15.2.3.1)
(10) Diesel engine piping P _ 4. ), 9.5.5, 9.5.6, 9.5.7) 4 15.2.4.4)
(11) Boron dilution event (12) Emergency feedwe -
.1 ;l.3) s Waterford SSER #1 1-2
(13) Feedwater line break analysis-(15.3.2)
(14) Clarification of_ transient analyses with potential for fuel damage (15.3.1)
- 1. 9 Li-case Conditions In addition to those issues listed in the SER as requiring a license condition to ensure that NRC requirements are met during plant operation,.the staff has identified the follcwing license conditions:
(1) Single failure of control system (7.7)
(2) Periodic survey of local industrial and transportation activities (2.2)
(3) Recalculate the adjusted reference temperature for the weld' metal during plant life (5.3.1.2) i Waterford SSER #1 1-3
2 SITE CHARACTERISTICS 2.2 Nearby Industrial, Transportation, and Military Facilities 2.2.1 Nearby Industries In the SER the staff indicated the presence of toxic gas hazards due to the various industrial and transportation activities in the vicinity of Water-ford 3.
We also noted that the applicant was proposing chlorine and ammonia detectors in the control room outside air intakes and that they intended to rely on operator training for the detection and protection against all other toxic gases which could potentially reach the control room in the event of an offsite hazard.
NRC staff review of this topic has processed sufficiently to justify the following observations and findings.
Due to the extensive industrial and transportation activities involving many hazardous materials in the vicinity of the proposed plant, it is our judgment that several diverse means of protection are needed.
In particular, reliance on operator ability to detect toxic gas odors is questionable.
At least for some gases, either the lack of odor, or the possibility of saturation effects which inhibit olfactory response when the gas concentration is high, are reasons for seeking a more reliable means of protection.
Based on a series of meetings and discussions with the applicant, the staff believes that a diversi-fied approach is feasible and the applicant has indicated their willingness to make commitments in that regard.
In addition to providing chlorine and ammonia detectors, they have also committed to providing broad range toxic gas detectors for the remainder of the chemicals which have the potential for reaching the control room in sufficient concentrations for causing short term incapacitation of the control room operators.
Presently, the applicant is in the process of selecting broad range detectors suitable for installation at Waterford 3.
In addition to the toxic gas detectors, the applicant ha. committed to installing and maintaining a hot line communication system between the Waterford 3 control room and the Emergency Operations Center (EOC) located at the St. Charles Parish Courthouse.
The center maintains a private tie line to approximately 16 industries within a 5 mile radius of Waterford 3.
The system, currently operational, is designed to be activated by any one of the member industries (including Waterford 3) or the E0C itself in the event of impending or actual emergencies within the member organization facilities.
This provides a relatively high reliability emergency communication capability which will increase significantly the chances of an early warning to the Waterford 3 control room in the event of a toxic gas release in the vicinity.
Since new product lines and manufacturing processes could introduce toxic gas sources which are not within the present list of chemicals analyzed by the applicant, they have agreed to perform a periodic survey of the local industrial and transportation activities as well as obtain letters of agreement with the local industries so that the applicant would be informed of changes in toxic chemical inventories.
Waterford SSER #1 2-1
The staff believes that the aggregate of the above measures, if implemented effectively, will reduce the toxic gas risk to an acceptably low level.
It is our position that these measures, when provided by the applicant, will have to be reviewed and approved by the staff as a condition for issuing an Operating License for Waterford 3.
The staff is still evaluating explosion hazards and will report on its resolution in a future supplement to the SER.
A i
n Waterford SSER #1 2-2
3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT AND SYSTEMS 3.7 Seismic Design 3.7.2 Seismic System Analysis In the Safety Evaluation Report, the staff stated that the applicant agreed to perform additional dynamic analyses of the Category I structures to determine the effects of the appropriate ties between the various cantilever stick model representing the Category I structures supported on the common mat foundation, the effects of the inclusion of the torsional soil springs, and the effects of considering actual and accidental eccentricities for all mass points.
The applicant has, performed these analyses and provided the summary of the results for the staffsreview.
The results of the new dynamic analyses indicate no i
significant change in responses for the respective Category I structures.
Therefore, the staff considers this item resolved.
3.8 Design of Category I Structures 3.8.2 Steel Containment In the Safety Evaluation Report, the staff stated that the applicant had made substantial extrapolations from the data provided in the " Welding Research Council Bulletin No. 107," relating to the evaluation af in ;l stresses of cylindrical steel shells due to external loads, without p uviding adequate j
justification.
i l
At the request of the staff, the applicant has provided experimental results
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as justification for the extrapolations to i.h9 results provided in " Welding
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Research Council Bulletin No.107." These ne results, which justify the extrapolated values, are provided in the SHELLTECH Report 80-2, " Evaluation of Reinforced Openings in large Steel Pressure Vessels." The staff considers this item resolved.
(
3.8.5 Foundations In the Safety Evaluation Report, the staff stated that the applicant agreed to reevaluate the foundation mat !or changes in the value of the subgrade modulus.
This reevaluation would verify the location of the critical sections in the foundation mat, when used as in a rigid block, and would permit the evaluation of special reinforcement identified for these sections.
The applicant has performed three analyses which consider changes in the effects of the subgrade modulus.
The results of the analyses indicate that the design curves, which represent the mat capacity, envelop the design requirements imposed by the results of the three new analyses which consider the changes in subgrade modulus.
The staff considers this item resolved.
Waterford 55ER #1 3-1
4 REACTOR 4.4 Thermal-Hydraulic Design In the Safety Evaluation Report the staff required that the applicant perform the following:
1.
demonstrate the applicability of the CE-1 Critical Heat Flux (CHF) l correlation and the proposed limit value to the Waterford 3 fuel design; I
2.
supply the necessary information for the Core Protection Calculator (CPC) system; and 3.
perform the safety analyses which account for the new fuel design and power distributions in future cycles.
The staff is awaiting a commitment from the applicant to supply the CPC information by March 1982.
In addition, the applicant has not provided the revised safety analyses required by Item 3 above.
The effects of grid design and grid spacing on CHF and Combustion Engineering's justification of why the CE-1 correlation is applicable to the new Waterford fuel design were discussed in the SER.
Spacer grid configurations and spacing play an important role in CHF.
Researchers have studied the effects of spacer grids and spacings on CHF and many researchers have shown that spacer grids increase CHF with various magnitudes depending upon the types of grids.
The results have led to the conclusion that a well designed spacer grid does enhance downstream turbulent mixing and, therefore, increases CHF.
However, the effects of turbulent mixing decays downstream from the spacer and results in minimum turbulent mixing just upstream of the next grid.
This is the main reason that, for non-uniform axial power distributions (APDs), DNB generally occurs upstream of a spacer grid.
Combustion Engineering's non-uniform APD data have shown that the DNBR limit of 1.19 is conservative when the Tong-F factor is used in the CE-1 correlation.
This is due to the fact that the CE-1 correlation was developed from uniform APD test data and the Tong-F factor was then applied to the non-uniform data without optimization.
The staff recommended that the applicant optimize the F-factor so that the CE-1 correlation would better predict the data.
The applicant chose to preserve the present form of the F-factor and contends that the margin of conservatism is large enough to cover the unfavorable effects which arise due to larger grid spacings and the uncertainty of the effects on CHF due to the HID-1 grid design.
The staff has concluded that the DNBR limit of 1.19 is applicable for fuel assemblies with standard grid configurations.
Deviations from the standard grid configuration should be evaluated separately with due consideration that the beneficial effects from the CE standard spacer grids were included in the development of the CE-1 correlation.
In the absence of data to prove otherwise, Waterford SSER #1 4-1
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the staff has concluded that'the Waterford fuel assembly design may result in a reduction in CHF.
However, the staff believes that the reduction in CHF will be small.
Therefore, the staff will impose a 1 percent adjustment which will result in a DNBR limit of 1.20.
Summary This SER s'upplement addresses the following open items:
(1) the applicability of the CE-1 CHF correlation _and the proposed' limit to the Waterford fuel design; (2) the applicant's commitment to supply the needed CPC informatior, by March 1982; and (3) the applicant's revised safety analyses which account for the new fuel design and power distributions in future cycles.
Based on the information presented in this report the staff considers Item (1) above resolved.
The DNBR limit for the Waterford fuel design is now 1.20 Items (2) and (3) will be addressed in a future supplement to the SER.
Waterford SSER #1 4-2
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.3 Reactor Vessel 1
f 5.3.1 Reactor Vessel Materials l
General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure 7
Bounc ry," Appendix A, 10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to ensure that, when stressed under operating, maintenance, and testing conditions, the boundary behaves in a nonbrittle manner and the probability of rapidly rcopagating fracture is minimized.
General Design Criterion 32, " Inspection of Reactor l
Coolant Pressure Boundary," Appendix A,10 CFR Part 50, requires, in part, l
that the reactor coolant pressure boundary be designed to permit an appropriate material surveillance program for the reactor pressure vessel.
l 1
1 i
The fracture toughness requirements for the ferritic materials of the reactor coolant pressure boundary are defined in Appendix G, " Fracture Toughness I
Requirements," and Appendix H, " Reactor Vessel Material Surveillance Require-ments," of 10 CFR Part 50.
The staff has reviewed the materials selection, toughness requirements, and extent of materials testing conducted by the applicant to provide assurance j
that the ferritic materials used for pressure-retaining components of the reactor coolant pressure boundary possess adequate toughness under operating, maintenance, testing, and anticipated transient conditions.
The Waterford 3 reactor vessel, steam generators (primary side), and pressurizer were designed to the specifice', ions of the 1971 Edition of the ASME Boiler and Pressure Vessel Code (hereinafter, "the Code"),Section III, " Rules for Construc-tion of Nuclear Power Plant Components," including Addenda through Summer 1971.
The reactor coolant pressure boundary piping, pumps, and valves were designed to the specifications of the 1971 Edition of the Code,Section III, including Addenda through Winter 1971.
Based on the November 1974 construction permit date, a.ction 50.55a, " Codes and Standards," 10 CFR Part 50 requires that the ASME Code editions and addenda applied to the pressure vessels be no earlier than those of the Summer 1972 Addenda of the 1971 Edition.
Section 50.55a also requires that the ASME Code edition and addenda applied to the piping, pumps, and valves that are part of the reactor coolant pressure boundary r
(RCPB) shall be no earlier than those of the Winter 1972 Addenda of the 1971 Edition.
The design and construction of the RCPB components of Waterford 3 are, therefore, not in compliance with the requirements of Section 50.55a, 10 CFR Part 50.
However, the staff evaluated the applicant's RCPB materials to Appendix G of 10 CFR Part 50 which ensures that material properties are equivalent or superior to those specified in Section 50.55a, 10 CFR Part 50.
Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Materials Surveillance Requircments," of 10 CFR Part 50, specify the fracture Waterford SSER #1 5-1
toughness requirements for the ferritic materials of the reactor coolant pres-1 sure boundary during normal operation, testing, maintenance, and c.ticipated transient conditions.
5.3.1.1 Compliance With Appendix G,10 CFR Part 50 The staff has evaluated the information in the applicant's FSAR to determine the degree of compliance with the fracture f.oughness requirements of Appendix C, 10 CFR Part 50. -This evaluation indicates chat the applicant has met all requirements of Appendix G,10 CFR Part 50, except for Paragraphs III.B.1, III.C.1, III.C.2, and IV. A.1, for which.ths applicant has supplied stJficient information to justify exemptions.
The applicant has supplied additional information to demonstrate compliance with the re wirements in Patagraphs III.B.3, IV.A.3, and IV.B.
The staff evaluation of these paragraphs follows.
l l
Paragraph III.B.1 requires that the location and orientation of Charpy V notch (CVN) impact test specimens comply with the requirements of Paragraph NB-2322 of the ASME Code.
This paragraph of the Code specifies that the CVN impact test specimens for all plate and forging materials used for pressure-retaining parts of vessels and piping shall be oriented in a direction normal (transverse) to the principal rolling or working direction. The Waterford 3 impact test qualification program does not conform to this requirement because vessel and piping materials were CVN impact tested with longitudinally oriented specimens instead of transversely oriented specimens.
l To compensate for the effect of directionality on the CVN impact test, the applicant developed criteria for determining the transverse CVN impact value from the longitudinal CVN impact value.
The applicant provided data from Palo Verde 1 and EPRI Report NP-232, " Fracture Toughness Data for Ferritic Nuclear Pressure Vessel Materials." The staff has evaluated the data submitted by the applicant and considers the criteria for determining the transverse CVN impact value from the longitudinal CVN value sufficiently conservative to adequately characterize the fracture toughness of all plata and forging materials used for pressure-retaining parts of vessels and piping in Waterford 3.
Based upon the above evaluation, the staff grants an exemption to testing pressure-retaining materials in vessels and piping using transversely oriented CVN impact specimens per Paragraph III.8.1 of Appendix G, 10 CFR Part 50.
Paragraph III.B.3 requires that calibration of temperature instruments and CVN impact test machines complies with the ree"irements of Paragraph NB-2360 of the ASME Code.
Paragraph NB-2360 requires the calibration of temperature instruments and impact machines every 3 months and 6 months, respectively.
The applicant has indicated in an amendment to tne FSAR that the calibration of temperature instruments and impact machines used for testing of Waterford 3 RCPB materials were calibrated at 3-month and 6-month intervals.
Therefore, the applicant complied with the requirements of Paragraph III.B.3 of Appendix G, 10 CFR Part 50.
Paragraph III.C.1 requires that CVN impact tests be conducted over a temperature range sufficient to define the CVN test curves for all reactor vessel beltline material.
Ferritic steel plates, weld metal, and heat-affected zones are the materials located in the Waterford 3 reactor vessel beltline.
The Waterford 3 Waterford SSER #1 5-2
r i
l FSAR report includes CVN impact test data for all plctes, fer two weld materials, and one heat-affected zone in the RV beltline.
In addition to CVN impact data, the applicant has indicated that the ve:Je1 br:ltline welds were fabricated using the submerged arc weld process with Linde 0091 flux and MIL B-4 wire and also the shielded metal arc weld process with E 8018 electrodes.
Even th wgh only one heat-affected zone in the RV beltline has been CVN impact tested, the staff concludes, based on the weld processing identified by the applicant and other information and data reviewed by the staff, that the fracture toughness of post-weld heat-treated ferritic heat-affected zones in the Waterf ord 3 RV beltline region is equivalent or greater than that of the beltline plates.
Based on the conclusion that the base metal plates will be more limiting than the heat affected zene, the staff concludes that CVN impact curves for beltline heat-affected zones are not required.
Although only two weld materials were tested over a temperature range sufficient to define their CVN test curve, all other beltline weld metals were tested over a sufficient temperature range to determine the limiting weld metal in the belt-line region.
Based on the above, the staff grants an exemption to CVN impact testing all beltline materials over a temperature range sufficient to define their CVN test curves per Paragraph III.C.1 of Appendix G,10 CFR Part 50.
i l
Paragraph III.C.2 of Appendix G,10 CFR Part 50, requires, in part, that the base materials used to prepare test specimens for the reactor vessel beltline l
i region shall be from excess base plate from the vessel beltline re;'on.
Paragraph III.C.2 of Appendix G was not complied with in that mater 10s used to prepare weld test specimens for the reactor vessel were taken from simulated weldments prepareu from excess production plate.
However, the weld wire and flux materials used in the test specimens are the same as those used in the reactor vessel beltline.
After weld completion, the sample weldments were subjected to a heat treatment to obtain met-11urgical effects equivalent to
(
those produced during fabrication of the reactor vessel.
Based on our evalua-tion of this information, the staff concludes that although the same base material was not u;ed to prepare the test samples, an exemption from the specific requirements of Paragraph III.C.2 of Appendix G is justified because the same heat treatment, weld wire, flux, and welding process used in the vessel l
welds were used in the test specimens.
Since the weld toughness properties are determined primarily by heat treatment, weld wire, flux, and welding process, and not by differences in similar base materials, the use of weldment test specimens having the same weld wire, flux, and heat treatment as the vessel l
welds is sufficient to satisfy the requirements of paragraph III.C.2 of Appendix and provides acceptable justificat ion for an exemption to the exact requirements of Paragraph III.C.2 of Appendix G.
}
Paragraph IV. A.1 requires that a reference temperature, RTNDT, be detc.rmined l
I for each ferritic material of the reactor coolant pressure boundary and that this reference temperatum be used as a basis for providing adequate margins of safety for reactor operation.
The value of RT is defined in Paragraph NDT NB 2330 of the ASME Code as the higher of either (a) the nil ductility tempera-ture, as determined by the dropweight test, or (b) a temperature of 60 F less Waterford SSER #1 5-3
than the temperature at which Su ft-lb energy and 35 mils lateral expansion is achieved, as determined by ihe CVN impact test.
CVN impact test specimens from plates and forgings are to be oriented in a direction normal (transverse) to the principal rolling or working direction.
Th'. applicant has t.t the requirements of Paragraph IV. A.1 for all ferritic reactor coolant pre,sure boundary (RCPB) materials except the following:
(1) SA-105 base metal used in RCPB applications in valve bonnets, pump covers (lower flange of drive mount), and surge nozzle forging; (2) SA-182 Type 403 and ASTM A-276 Type 440 material used in RCPB applications in lower control element drive mechanism housings and upper control element drive mechanism housings, respectively; (3) All base metals used in the reactor vessel, the primary side of the steam generator, the pressurizer, and RCPB piping; (4) Weld metal outside the beltline region; and (5) Heat-affected zones outside the beltline region.
The applicant has supplied additional CVN impact data which had not been in the applicant's FSAR for SA-105 base metal used in pump covers (lower flange of drive mount) and for SA-182 Type 403 material used in lower control element drive mechanism hausings.
CVN impact data for the sutga nozzle forging is contained in Table 5.2-8 of the FSAR.
These data indicate that these materials will not be limiting for operation of Waterford 3.
Thr. applicant has indicated that ASTM A-276 Type 440 material was used for a vent valve seat in the u @ er control element drive mechanism housing, which is not a RCB application.
All base metal used in the reactor vessel (RV), the primary side of the steam generator, the pressurizer and RCPB piping was CVN impact tested with specimens oriented in the longitudinal direction rather than the transverse direction To compensate for the effect of directionality, the applicant has estimate.d the temperature at which 50 ft-lbs energy and 35 mils lateral expansion would be achieved if transversely oriented specimens had been tested.
If the minimum absorbed energy of three longitudinally oriented CVN impact specimens was less than 30 ft-lbs, the temperature at which 50 ft-lbs and 35 mils would occur for '.ransversely oriented specimens was estimated from lower bound CVN in. pact test data from Waterford 3 and 25 heats of material from San Onofre 2 which had been fabricated to the same material specification and heat treated to the same metallurgical condition as the material in Water-ford 3.
The applicant demonstrated that the materials in Waterford 3 and San Onofre 2 were equivalent by identifying tne heat treatment received by the materials.
fr m 1 ngitudi-The applicant us9d three other methods for estimating the RTNDT nally oriented CVN impact tests.
In NRC's evaluation af the applicant's compliance to Paragraph III.B.1,10 CFR vart 50, the staf f evaluated these Waterford SSER #1 5-4
methods and determined that the data submitted to substantiate the methods were sufficient to determine the transverse CVN impact properties from longitudinal CVN impact test results.
The applicant determined the ni' ductility temperature from drop weight test data for all ferritic reactor coolant pressure boundary base metals except those in the steam generator and the piping.
Paragraph NB 2330 of the ASME Code requires drop weight testing of all ferritic RCPB steam generator materials greater than 5/8-inch thick and of all ferritic RCPB piping material greater than 2-1/2 inches thick.
The staff has utilized data from 10 other nuclear facilities, and limiting data in Table NC 2311(a)-1 of the 1980 ASME Code to determine the generic lower bcund of the nil ductility temperature for all steam generator and piping materials which required drop weight testing.
The generic drop weight data, the CVN impact data previously identified, and the CVN impact data and drop weight data presented in the FSAR were utilized to determine the limiting base materials for operation of Waterford 3.
The applicant has not determined the RT f r any RCPB welds or heat-affected NDT zones outside the beltline region.
Howt /er, the applicant has indicated tnat all RCPB welds were made using submerged arc and covered electrode processes.
Based on the data from several other nuclear facilities presently under review which were fabricated using submerged arc and covered electrode processes, a conservative estimate of the RT f r these welds is 0 F.
The staff will NDT utilize 0 F at the RT for all RCPB welds outside the beltline region when NDT it evaluates the pressure temperature limits for operation of Waterford 3.
Based on the welding processes identified by the applicant and other data and i1 formation reviewed by the staff, the fracture toughness of the heat-affected zones outside the beltline region is equivalent or greater than the adjacent base metal.
Although the applicant did not determine tne RTNDT per Pang'aph NB 2330 of the ASME Code for each ferritic material, the critical RTND; f r operating, maintenance, and testing conditions has been determined based on additional information available in the literature and additional data supplied by the applicant.
Therefore, the staff will grant an exemption to Paragraph IV.A.1 of Appendix G.
Paragraph IV.A.3 requires, in part, that material for bolting and other fasteners meets the fracture toughness requirements of Paragraph NB 2333 of the ASME Code.
10 CFR Part 50, Paragraph 50.55a, requires pressure vessel bolting to be fabricated to the ASME Code editions and addenda no earlier than the Summer 1972 Addenda but no later than the Summer 1978 Addenda.
For bolting, the Code requires CVN testing of three specimens at a temperature no higher' than the preload - 7erature or the lowest service temperature, whichever is less.
All three specimens shall meet the ASME Code requirements of:
(1) 25 mils lateral expansion for bolting over 1 inch through 4 inches in nominal diameter; and (2) 25 mils lateral expansion and 45 ft-lbs absorbed energy for bolting over 4 inches in nominal diameter.
Waterford SSER #1 5-5
RCPB bolting and fasteners meet these requirements except for reactor coolant pump (RCP) casing studs and nuts, RV bolting (heat no. 18551), and pressurizer manway nuts.
These materials were tested to a Code earlier than the Summer 1972, which did not require CW impact tests at the lowest service temperature or the preload temperature.
The RCP casing studs and RV bolting (heat no. 18551) were fabricated from SA-540 B-23; pressurizer manway nuts from SA-193 B-7; and RCP nuts from SA-194 B-7.
In addition, the mils lateral expansion data for the 1.656-inch diameter steam generator and pressurizer manway studs were not reported.
T H e studs were fabricated to SA-540 Grade B24 requirements.
The applicant has estimated a minimum preload temperature of 60 F and lowest service temperature of approximately 160 F.
The bolting materials that did not meet the ASME Code requiraments for Charpy V notch impact testi7g wm e-106F.
Since the materials were tested at tested at temperatures of 0 F d
temperatures belc1 the tempe.
.e required by the ASME Code, compliance with ASME Code CVN impact requiremuts was demonstrated by extrapolating the balt-ing material CVN impact data from 0 F to 60 F.
The applicant has submitted in FSAR Figures 121.3-1 and 121.3-2 curves of mils lateral expansion vs. temperature and energy absorption vs. temperature from six additional heats of SA-540 Grade B23 material. These CVN curves are con-sidered a conservative representation of the effect of temperature upon CVN impact properties for Waterford 3 SA-540 Grade B23 materiale because the Waterford 3 materials were heat treated to a metallurgical condition equivalent to the six heats of SA-540 Grade B23 materials which were utilized to generate FSAR Figures 121.3-1 and 121.3-2.
These data indicate that if RCP casing stud material and RV bolting (heat no. 18551) material had been tested at 60 F, the minimum preload temperature, they would have complied with minimum ASME Code requirements.
CVN impact data from the Aerospace Structural Materials Handbook (Figure 3.0331) indicate that material with the same chemical composition and heat treated to a metallographic condition wivalent to the pressurizer manway nuts and RCP nuts have a relatively na. row transition zone and that at 60 F the materials would be tested at their upper-shelf energy level.
These data also indicate that the upper-shelf energy level for these materials is alove 50 f t-lbs.
The CVN impact data submitted by the applicant indicate that the pressurizer marway nuts and RCP nuts were tested in their transition zone and will have greater than 50 ft-lbs energy absorbtion at 60 c.
Therefore, the staff concludes that if pressurizer manway nuts and RCP nuts had been tested at 60 F, the minimum preload temperature, they would have comlied with minimum ASiE Code requirements.
The steam generator and pressurizer manway studs which did
. have the mils lateral expansion reported had absorbed energy values of 54 57 ft-lbs at a test temperature of 10 F.
A review of CVN impact data for RV closure head bolting in FSAR Table 5.3-11 indicate that SA-540 Grade B-24 material with 54-57 ft-lbs will have greater than 25 mils lateral expansion at 60 F.
Waterford SSER #1 5-6
Therefore, the steam generator and pressurizer manway studs CVN impact test results meet minimum Code requirements.
Although the CVN impact test results for RCP_ casing studs and nuts, RV bolting (heat no. 18551), and pressurizer manway nuts did not meet ASME Code requirements (in accordan: e with 10 CFR Part 50, paragraph 50.55a), additional data supplied by the applicant and contained in the Aerospace Structural Materials Handbook indicate these materials were tested to an earlier equivalent Code and an exemption to Paragraph IV. A.3 is not required.
Paragraph IV.B requires reactor vessel beltline materials to have a minimum CVN upper-shelf energy of 75 ft-lbs in accordance with Paragraph NB 2322.2 of the ASME Code.
This paragraph requires beltline base material to be tested with specimens oriented in the transverse and longitudinal directions. The applicant has determined the upper shelf for all reactor vessel beltline base l
metal with specimen oriented lcngitudinally.
The beltline base material is i
SA-533 Grade B Class 1.
The 12 west CVN upper-shelf energy absorption value identified in FSAR Table 5.2-6 is 138 ft-lbs.
According to Idaho Nati..71 Engineering Laboratories (INEL) Report EGG-FM-5313,
" Review of the Estimation of the Transverse Charpy V-Notch Shelf Value from the Longitudinal Value," the transverse CVN upper-shelf energy absorption may be censervatively estimated as 65% of the longitudinal CVN upp r-shelf energy absorption.
The staff has reviewed this report and considers
- El's conclusion acceptable.
Sixty-five percent of the lowest CVN upper-shelf i.ergy absorption value (138 ft lbs) is 89 ft-lbs, which exceeds the minimum requirements of Paragraph IV.ll.
Therefore, an exemption to h ragraph IV.B that r w ires the determination of the transverse upper-shelf anergy for base metal is fustified.
The applicant in FSAR amendments has submitted additional CVN impact data for each beltline weld material.
I ' tource of the data was archive sample material which had been prepared using the,ome weld procedure, heat of wire, and lot of flux and heat treated to the same metallurgical condition as the beltline welds.
These data indicate that all weld metals would have a CVN upper-shelf energy greater than-75 ft-lbs; therefore, the applicant's beltline welds comply with the requirements of Paragraph IV.B, 10 CFR Part 50.
The fracture toughness of the heat-affected zones in the RV beltline were discussed in Paragraph III.C.1.
Since their fracture toughness was considered equivalent or greater than that of the adjacent base metal or weld metal, no CVN impact curves for RV beltline heat-affected zones are required.
5.3.1.2 Compliance With Appendix H, 10 CFR Part 50 The mate.-ials surveillance program at Waterford 3 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 Waterford 3 surveillance program, fracture toughness data must be obtained from material specimens that are representative of the limiting base, weld, and heat-affected zone materials in the beltline region.
These data will permit the determination of the Waterford SSER #1 5-7
conditions under which the vessel can be operated with cdequate margins of safety against fracture throughout its service life.
The fracture tougnness properties of reactor vessel beltline materials must be monitored throughout the service life of Waterford 3 by a materials surveillance program that meets the requirements of Appendix H of 10 CFR Part 50.
The staff has evaluated the applicant's information for degree of compliance to these.equirements.
Based on this evaluation the staff concludes that the Clica:,c has met all the requiremants of Appendix H,10 CFR Part 50, with the excep m n of Paragraph II.B for wiich the applicant has supplied sufficient information to justify an exemptian.
Staff evaluation of the applicant's deviation from the requirements t,f this paragraph follows.
Paragraph II.B, Appendix H to 10 CFR Part 50, requires that reactor vessels constructed of ferritic materials, with a peak neutron fluence (E>l MeV) at the end of the design life of the vessel exceeding 1017 n/cm, shall have 2
th. 4r beltline regions monitored by a surveillance program complying with the
- p. ifications of ASTM Standard E 185-73 except as modified by Appendix H.
ASTH Standard E 185 requires that material placed in the surveillance capsules represent the material that may limit operation of the reactor during its life-time.
The selection of the base metal, heat-affected zone (HAZ), and the weld metal surveillance specimens is based on consideration of the fracture toughness properties of all of the beltline materials (RT and upper-shelf CVN energy)
NDT in the unirradiated condition, the chemical cecposition, and the neutron fluence to establish the limiting materials. The applicant's selection of the lower shell plate (M1004-2) as the limiting base metal and the selection of the HAZ (M1004-2) as the limiting HAZ is acceptable.
According to staff evaiuation, weld metal E 8018/ BOLA is the most limiting weld metal in the Waterford 3 beltline.
The Waterford 3 surveiliance program contains weld metal from 88114/0145.
Because the Waterford 3 surveillanca metal is not the most limiting weld metal, the applicant's surveillance prograni is not in full compliance with Appendix H, 10 CFR Part 50.
To have an acceptable surveil-lance program for U terford 3, the appiicant must use the following analysis for all weld metal removed from the capsule ar.d tested.
During the plant's life, the applicant must recalculate the adjusted reference temperatu:u (RTNDT) f r the weld metal based on the greater of the measured shift in RT as determined by impact testing of 88114/0145 weld metal and NDT the predicted shift in RT as determined by Regulatory Guide 1.99, " Effects OT of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials,"
for E 8018/ BOCA weld metal.
Although material from the most limiting weld seam is not contained in the Waterford 3 material surveillance program, the staff will grant an exemption to Paragraph II.B of Acpendix H, 10 CFR Part 50, because methods of analysis contained in Regulatory Guide 1.99, which will be used to determine the radiation induced change in fracture toughness of limiting weld metal, are conservative.
Waterford SSER #1 5-8
5.3.1.3 Conclusions for Compliance With Appendices G and H, 10 CFR Part 50 Based on the preceding evaluation of compliance with Appendices G and H, 10 CFR Part 50, the staff concludes that the applicant has met all the fracture tough-ness requirements of these appendices except for the following:
Paragraphs III.B.1, III.B.3, III.C.1, III.C.2, IV.A.1, IV A.3, and IV.B of Appendix G and Paragraph II.B of Appendix H.
The applir mt has supplied sufficient information to justify an exemption to Paragraphs II.B.1, III.C.1, III.C.2, and IV.A.1 of Appendix G and Paragraph II.B cc Appendix H.
The applicant has supplied additional information N Jemonstrate compliance to Paragraphs III.B.3, IV.A.3, and IV.B ef Appendix G.
Appendix G, " Protection Against Nonductile Failtre,"Section III of the ASME Code, will be used, t lether with the fracture toughness test results required by Appendices G and L. 10 CFR Part 50, to calculate the pressure-temperature limitations for the Waterfo;d 3 reactor vessel The fracture toughness tests required by the ASME Code and by Appendix G of 10 CFR Part 50 provide reasonab 4 assurance that adequate safety margins against the possibility of nond.sctile 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 pro-vide adequate safety margins during operating, testing, maintenance, and antici-pated trcnsient 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 Waterford 3 reactor vessel beltline caused by exposure to neutron radia-tion can be properly assessed, and adequate safety margins against the possi-bility of vessel failure can be provided.
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 Design Criterion 32.
5.3.2 Pressure-Temperature Limits Appendix G, " Fracture Toughness Requirements," sM Appendix H, " Reactor Vessel Material Surveillance Program Requirements," 10 Cr'R Part 50, describe the condi-tions that require pressure-temperature limits and provide the general bases for these limits.
These appendices specifically require that pressure-temperature limits must provide safety margins at least as great as those recommended in the ASME Code,Section III, Appendix G, " Protection Against Nonductile Failure."
Appendix G, 10 CFR Part 50, requires additional safety margins whenever the reactor core is critical, except for low-level physics tests.
WateNord SSER #1 5-9
The following pressure-temperature limits imposed on the reactor coolant pres-sure boundary during operation and tests are reviewed to ensure that they provide adequate safety margins against nonductile behavior or rapidly propagating failure 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.
Appendices G and H, 10 CFR Part 50, require the applicant to predict the shift in reference temperature due to neutron irradiation.
The shift in RT due NDT to neutron irradiation is then added to the initial RT to establish the NDT adjusted reference temperature.
The base plate and weld seam having the highest adjusted reference temperature are considered the most limiting materials for which the pressure-temperature operating limits are based.
In the case of Waterford 3, the most limiting material af ter 8 effective full power years and 32 effective full power years is beltline plate M-1004-2.
According to the staff evaluation, the proposed pressure-temperature limit curves for Waterford 3 are acceptable for 32 effective full power years.
After removal of the surveillance capsule, the pressure-temperature limits must be revised to reflect the actual neutron damage as determined from the results of the reactor vessel surveillance program.
The pressure-temperature limits to be imposed on the reactor coolant system fo" all operating and testing conditions, to ensure adequate safety margins against nonductile or rapidly propagating failure. Tust be in conformance with established criteria, codes, and standards to be acceptable to the staff.
The use of operating limits based on these criteria, as defined by applicable regulations, 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 Integrity The staff has reviewed the fSAR sections related to the reactor vessel integrity of Waterford 3.
Although most areas are reviewed separately in accordance with other review plans, reactor vessel integrity is of such importance that a special summary review of all factors relating to reactor vessel integrity is warranted.
I The staff has reviewed the information supplied by tha applicant to ensure that it is complete and that no inconsistencies cvist that would reduce the certainty of reactor vessel integrity.
The areas reviewed are:
(1) Design (SER 5.3.1)
(2) Materials of construction (SER 5.3.1)
Waterford SSER #1 5-10 li g
g i
e (3) Fabrication methods (SER 5.3.1)
(4) Operating conditions (SER 5.3.2).
The staff reviewed the above factors contributing to the structural integrity of the Waterford 3 reactor vessel and concludes that the applicant has complied with the required regulations, codes, and standaros except for the following.
Paragraph III.B.1 of Appendix G requires that CVN impact test specimens for all plate and forging materials used for pressure-retaining parts of vessels and piping shall be oriented in a direction normal (transverse) to the principle rolling or working direction.
Although CVN specimens for vessel materials were
'ested in the longitudinal direction rather than the transverse direction, the applicant's methods for correlating longitudinal CVN impact properties with tranverse properties are sufficiently conservative to justify an exemption to the CVN impact specimen orientation requirements of Paragraph III.B.1 of Appendix G.
Paragraph III.C.1 of Appendix G requires that CVN impact tests be conducted over a temperature range sufficient to define the CVN test curves for all reactor vessel beltline material.
Although only one heat-affected zone and two welds in the reactor vessel beltline were CVN impact tested, the applicant has provided sufficient weld processing inforration and CVN impact test values to adegaately. characterize the beltline mater ials and justify an exemption to the requirements of h agraph [II.C.1 of Appendix G.
Paragraph III.C.2 of Appenuix G requires that the base metal used to prepare test spoimens be taken from excess base metal from the vessel beltline region.
The sr.cimens for testing the vessel beltline welds were not prepared from exceos production plates.
The applicant, however, has supplied sufficient data to demonstrate that the weld specimens do represent the welds in the vessel beltline region.
Therefore, an exemption to Para 0raph III.C.2 is justified.
Paragraph IV.A.1 of Appendix G requires that a reference temperature, RTNDT' be determined per Paragraph NB 2330 of the ASME Code for each ferritic material in the reactor coolant pressure boundary.
Although the applicant did not deter-mine the RTNDT per Paragraph NB 2330 of the ASME Code for each ferritic material, the critical RT f r operating, maintenance, and testing conditions has been NDT determined based on additional information available in the literature and additional data supplied by the applicant.
Therefore, the staff has concluded that an exemption to Paragraph IV.A.1 of Appendix G is justified.
Paragraph II.8, Appendix H, requires that the material surveillance program comply with ASTM E 185-73.
The materials in the Waterford 3 surveillance program does not comply with all requirements in ASTM E 185; however, the materials that are in the program, together with methods for predicting radis-tion damage, provide sufficient information to conclude that an exemption to Paragraph II.B, Appendix H, is justified.
The staff has reviewed all factors contributing to the structural integrity of the reactor vessel and concludes there are no special considerations that make it necessary to consider potential reactor vessel failure for Waterford 3.
Waterford SSER #1 5-11
5.4 Component and Subsystem Design 5.4.3 Shutdown Cooling (Residual Heat Removal) System In the staff's SER it was identified that to place the plant into shutdown cooling, an operator had to leave the control rocm to restore power to the safety injection tank (SIT) isolation valves in order to isolate them and prevent their injection during depressurization. The staff had required that this function be performed from the control room, c, that other methods be available so that the plant could be placed into shutdown cooling with no need for an operator to leave the control room.
In response to staff question 211.94, the applicant has demonstrated that as an alternative to closing the SIT isolation valves, the SITS can be depressurized and the plant placed into shutdown cooling through use of the SIT vent valves.
These can be operated from the control room, and they meet staff requirements.
Therefore, this issue is satisfactorily resolved.
Waterford SSER #1 5-12
)
i l
6 ENGINEERED SAFETY FEATURES
. 6.1 Engineered Safety Features Materials 6.1.2 Organic Materials i
In the FSAR the applicant indicates that the coating systems used on exposed surfaces inside the containment have been qualified in accordance with ANSI N101.2, " Protective' Coatings (Paints) for Light Water Nuclear Reactor Contain-i I
ment Facilities," American National Standards Insitute (1972), and ANSI N5.12, i
" Protective Coatings (Paints) for the Nuclear Industry," American National i
Standards Institute (1974).
The applicant also stated that the protG tive
{
coating system for the containments are applied in accordance with Regulatory j
Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to-Water-Cooled Nuclear Power Plants."
t Based on our evaluation, the staff concludes that the protective coating systems and their applications are acceptable and meet the requirements of Appendix B to 10 CFR Part 50. This conclusion is based on the applicant having met the quality assurance requirements of Appendix B to 10 CFR Part 50 since the coating systems and their applications meet the positions of Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," and the quality assurance standards to ANSI.
j N101.2, " Protective Coatings (Paints) for Light Water Nuclear Reactor Contain-ment Facilities." Also, the containment coating systems have been evaluated as to their suitability to withstand a postulated design basis accident (DBA) environment.
The coating systems chosen by the arlicant have been qualified under conditions which take into account the postuiated DBA conditions.
The control of combustible gases that can potentially be generated from the organic materials and from qualified and unqualified paints is reviewed in SER section 6.2.5.
The consequences of solid debris that can potentially be formed I
from unqualified paints are reviewed in SER section 6.2.2.
6.2 Containment Systems 6.2.3 Secone.ary Containment Functional Design The staff reported in the Safety Evaluation Report that the applicant had provided an analysis that showed the shield building annulus is maintained at 3 negative pressure greater than 0.25 inch water gauge relative to the outside l
atmosphere throughout the design basis LOCA, thus ensuring no primary contain-1 ment out-leakage escapes unfiltered directly through the shield building.
-Staff review of the applicant's analysis, however, found at least one noncon-servative modeling assumption (i.e., nonadiabatic boundary condition at the shield building exterior surface) and a nonconservative assumption for the initial shield building annulus pressure (i.e., -8.0 inches water gauge (e.g.)
versus the Technical Specification limit of -5.0 inches w.g. ).
Furthermore, the staff was unable to conduct a confirmatory analysis because of insufficient l
l Waterford SSER #1 6-1
shield building ventilation system (SBVS) fan data. Because of this, the staff
= concluded.in the SER that acceptance of the applicant's conclusions regardiry the shield building annulus pressure follcwing the design basis LOCA is con-tingent upon a' confirmatory analysis.
j In Amendments 19 and 20 to the Waterford 3 Final Safety Analysis Report, the applicant provided both a~ revised shield building annulus pressure analysis as4uming an initial annulus pressure of -5.0 inches w.g. and the~SBVS fan data-required to complete our confirmatory analysis.
The' applicant's revised analysis.
again showed that the shield building annulus pressure never' rises above
-0.25 it:ch w.g..follos;ing a LOCA.
The staff's confirmatory; analysis similarly shows the annulus pressure remains below -0.25' inch w.g..Therefore, the staff concludes that this matter is resilved.
1
- l Waterford SSER #1 6-2
i 7 INSTRUMENTATION AND CONTROL 7.3 -Enginetred Safety Features Actuation System Emergency Feedwater (EF) Isolation By Main Steam Isolation Signal and EF Control The emergency feedwater actuation signal (EFAS) is employed to start emergency feedwater (EFW) pumps.
EFAS, including " feed-only good generator" -logic,-was described in the Waterford 3 SER. ~The EFW control and block valves are to be operated by a supplementary control system. The function of this control system has been described in Amendment 19 of the Waterford FSAR.
On a-func-
)~
tional basis this control scheme is acceptable.
Detailed engineering drawings (electrical schematics) have not been submitted to date.
The EFW valve control system is to be part of the plant's protective system and is to be designed in compliance with criteria enumerated in Section 7.1 of the Waterford 3 FSAR, applicable to emergency safety feature actuation signals (ESFAS).
The staff will~ confirm compliance with these criteria upon receipt of the drawings.
A reliability study of the EFW system has been performed by the applicant and submitted as Appendix 10.4.98, Amendment 13, November 1980.
This study did not include reliability analysis of the supplemental EFW valve control system.
A subsequent reliability study, which considers this control system, has not been submitted to the staff to date.
The staff will review this study.
This reliability study should demonstrate that the added complexity of the EFW valve control system does not reduce overall system reliability.
r A main steam isolation signal (MSIS) will, by design, cause closure of the EFW control and block valves.
This feature is to be overridden by the EFW valve control system, should the water level in the steam generator fall below a predetermined value.
The staff will confirm the acceptability of this inter-lock scheme upon receipt of the drawings.
Conformance with IE Bulletin No. 80-06:
Reset and Override of Engineered Safety Features The satisfactory resolution of this item is based on the applicant's response to Question 030.35 of June 12, 1981.
Except as discussed below, the applicant will modify the design such that all ESF components will remain in their emergency mode following ESFAS reset.
Confirmation that emergency safety feature components do not change state following a systems level emergency safety feature actuation signal, will be accomplished by testing, which will be incorporated as part of the preoperational test program.
The justification for the exceptions is as follows:
4
-Watarford SSER #1 7-1 s
(1) Evaluation of Maintaining SIAS Reset Scheme for Following Components:
Volume Control Tank (VCT) Discharge Valve, Boric Acid Pumos A & B, and Charging Pumps A, B, & AB.
The position of the VCT discharge valve is controlled automatically by the level in the VCT during normal operations.
An SIAS signal closes the valve.
SIAS reset will return the valve to automatic control, but only if the initiating conditions no longer exist, i.e., measured signals are within limits of setpoints.
The boric acid pumps A and 8 are normally in the automatic mode during operation.
In this mode, a preset blended boric acid solution is auto-matically introduced into the VCT via the boric acid pumps upon demand from the VCT level controller (see FSAR Subsection 9.3.4.2.1.2).
SIAS causes these pumps to run continuously and directly supply the charging pump suction.
SIAS reset will return the pumps to the automatic mode described above if the initiation signal no longer exists.
During normal operation, one charging pump is in operation.
The letdown flow control valve is controlled to maintain a balance between letdown flow rate plus reactor coolant pump bleedoff rate and charging flow rate.
Operation of the two standby charging pumps and the letdown control valve is controlled by the pressurizer level control program (Figure 5.4-8 of FSAR).
An SIAS will cause pumps A and B to start.
If pump AB was running prior ot SIAS, it will continue to run.
An SIA5 reset will cause all charging pumps to revert to the automatic mode described above if the initiation signal no longer exists.
Hence, one charging pump will stay in operation.
The only accident analysis which assumes charging pump operation is the Small Greak LOCA Analysis, which assumes operation of one charging pump.
Following reset of SIAS, a minimum of one charging pump will remain in operation.
Inadvertent reset will not obviate the safety analysis since reset will not occur if the initiation signal still exists.
If the charging pumps had restored reactor pressure above the low pressure trip setpoint, a subsequent reduction of reactor pressure below the setpoint following reset will reinitiate SIAS.
As described, SIAS will close the letdown line and start Charging Pumps A and B.
If the system is not reset, then, two or possibly three charging pumps will discharge into the RCS, with letdown isolated (until the operator manually reopened), independent of pressurizer level control.
This would greatly increase the possibility of going solid in the pres-surizer and possibly lifting the safety valves on inadvertent SIAS initia-tion or for events which could lead to a rapid recovery of reactor pressure.
Hence, reset on the system level (upon SIAS reset) rather than on the recovery equipment level, is considered acceptable for this system.
(2) Evaluation of Maintaining SIAS Reset Scheme for Control Room fssociated (Habitat) Areas Exhaust Fan Bypass Dampers 0-18 (SA) and (SB).
These dampers are normally closed.
A SIAS will open the dampers, which will trip the exhaust fans E-34 (SA and Sil).
By these actions, air will Waterford SSER #1 7-2
be recirculated and the communication with the outside atmosphere prevented.
SIAS reset will cause the fan bypass dampers to change state from open to
. closed.
The areas included in this ventilation complex will continue to be isolated from the outside environment until the operator (manually) restarts the fans. When conditions return to normal and the operator is able to reset SIAS, after doing so, he has only to start the fans in order to restore the normal ventilation of the areas within this scheme.
He can start these fans under the condition that both bypasses are closed.
- However, if any isolation signal.is present, this will by design prevent the reset, so the dampers could not be closed.- Therefore, the present design for SIAS reset feature of the fan dampers simplifies the operator's steps in restoring normal ventilation and does not result in these areas coming in contact with external environments.
On this basis, the design is acceptable.
(3) Evaluation of Control Room, Conference Room and Kitchen Exhaust Fan -
Bypass Damper D-19 (SA and SB).
The staff's evaluation and conclusion as well as the applicant's justifica-tion for Control Room, Conference Room and Kitchen Exhaust Fan B mass Damper D-19 (SA and SB) is similar to Item (2) above. The associated fan is E-42.
7.5 Safety-Related Display Instrumentation TMI Action Plan Item II.D.3:
Direct Indication of Relief and Safety Valves Positions Section 7.5.1.1.0, of the FSAR, Amendt.;ent 17, specifies the indicators provided, their function, qualification, indication / annunciation and backup methods.
Waterford 3 does not use PORVs on the primary loop.
Safety valve position indication is provided utilizing the loose parts monitoring
'mtrumentation.
The existence of fluid flow on the discharge side of the s u ty valves is sensed by the ultrasonic vibration monitoring devices on the uilpiece of each safety valve. The devices are to be seismically qualified.
An alarm has been provided in the control room. The system is powered by a static uninterruptible power supply, from a vital bus.
Appropriate filtering and gain setting of the flow pickup signals reduce noise, impact effects, cross-talk disturbances, and provide an unambiguous valve position indication.
Backup indication of valve position is derived from independent on-line measurement of quench tank level and temperature.
On the basis of the above, the staff finds that TMI action plan item II.D.3 has been satisfied.
Waterford SSER #1 7-3
\\
7.7 Control Systems Not Required for Safety-
-Single Failure of Control Systems St'udy The staff requested that the applicant address the following concern:
Common Electrical Power Sources or Sensor Malfunctions May Cause Multiple Control System failures With regard to the effects.of control system failures or malfunctions, the analyses reported in Chapter 15 of the Final Safety Analysis Report are intended to demonstrate the adequacy of safety systems in mitigating anticipated opera-tional occurrences and accidents, including those related to control systems.
Based on the conservative assumptions made in defining these " design bases" events and on our detailed review of the analyses, it is likely that these analyses adequately bound the consequences of single control system-failures.
To provide assurance that the Chapter 15 analyses adequately bound events initiated by a single credible failure.or malfunction, the staff requires that a review be conducted to identify any power sources or sensors which provide power or signals to two or more control systems, and to demonstrate that failures or malfunctions of these power sources or sensors will not result in consequences outside the bcunds of the Chapter 15 analyses or beyond the capability of operations of safety systems.
The staff requires that the applicant resolve these concerns before startup following the first refueling.
Accordingly, the operating license will be conditioned to reflect this requirement.
Waterford SSER #1 7-4
13 CONDUCT OF OPERATIONS 13.3 Emergency Preparedness Evaluation 13.3.1 Introduction The staff's evaluation of the applicant's emergency plans is provided in Section 13.3 of the Safety Analysis Report (SER) dsted July 9, 1981 (NUREG-0787).
The Waterford 3 Emu gency Plan (Plan) as amended (Amendment 17, April 1981),
was reviewed against the requiren ents of 10 CFR Section 50.47(b), Appendix E to 10 CFR Part 50, and the criteria of the 16 Planninp Standards in Part II of the " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654/ FEMA-FEP-1, Rev. 1, dated November 1980.
In the SER, the staff specifically identified-five items requiring resolution for which additional information and commitments were to be provided by the applicant.
Since the issuance of the SER, the applicant has provided the staff with additional information and commitments in response to the open items. This information was provided in Amendment 19 to the fSAR and submittals on September 29 and October 7, 1981.
The applicant's responses to five identified items have been evaluated and are discussed in this supplement.
With regard to offsite emergency preparedness, the State of Louisiana Peacetime Radiological Response Plan was formally submitted, on September 1, 1981, to the Federal Emergency Management Agency (FEMA) Region VI for review by the Regional Assistance Committee (RAC VI).
The local parish plans are to be submitted for RAC VI review in October 1981.
FEMA will review these plans in accordance with its Memorandum of Understanding with the NRC and FEMA's proposed emergency planning rule, 44CFR350.
A joint emergency response exercise, designed to determine the onsite and offsite emergency response capabilities, is scheduled for June 1982.
The final NRC approval of the state of emergency preparedness for the Waterford 3 sito will at made following implementation of the emergency plans to include development of procedures, training and qualification of personnel, installation of equipment and facilities, and a joint exercise involving participation by the response organizations (site, State, and local).
13.3.2 Evaluation of Applicant's Emergency Plan The applicant's responses to the five items previously identified by the staff as requiring additional information and commitments have been evaluated and are discussed below.
(The order of presentation corresponds to the listing of unresolved items that appears in Section 13.3 of the SER.)
Waterford SSER #1 13-1
1.
Existing agreement letters must be updated with regard to access control of the-10-mile Emergency Planning Zone (EPZ).
Discussion and Conclusion The applicant, in a letter dated October 7, 1981, committed to clarify access control of the Mississippi River within the 10 mile EPZ, by.
identifying an agreement letter between the U.S. Coast Guard and the Louisiana Department of Public Safety, Office of Emergency Preparedness.
The letter, dated October 17, 1980, specifically addresses control of the marine traffic on that portion of the Mississippi River which traverses the Waterford 3 10-mile EPZ. -The applicant, in a letter dated October 7, 1981, committed to update the agreement letter with the Missouri Pacific Railroad Company and to provide written agreements established between the State and the other railway systems in St. Charles and St. John the Baptist Parishes, within the 10 mile EPZ,.with regard to access control of the railway within the 10-mile EPZ.
LP&L will make appropriate changes to the Plan by amending the FSAR in December 1981.
Based on our review of their Plan and submittal as outlined and discussed above, we find that the applicant has provided an acceptable response to this item.
2.
The jurisdiction of the Lafourche Basin Levee District over that portion of the levee that lies within the plume EPZ must be clarified and written arrangements for access control thereof must be estabiished.
Discussion and Conclusion The applicant's response to the staff's question Q-810.1-3 (Amendment 17 to the FSAR) stated that the Lafourche Basin Levee District does not pro-
'~
vide access control on any levees in which it has jurisdiction.
Discus-sions with the applicant and a tour of portions offthe roadway network in the 10-mile EPZ during an on-site visit on July 21,'1981, verified that access to the levees via land can effectively be controlled by access, control of the river roads - Highway 18 on the west bank, and Highway 628 on th'e
..The State Plan specifies east bank, both of which run parallel to the levees.
that traffic control within the parishes of St. Charles and St. John ths Baptist is the responsibility of the respective Sherif f s' Office and Police Departments.
Access to the levee via the Mississippi River will be controlled by the U.S. Coast Guard as discussed in item 1 above.
Based on our review of their Plan and submittal as outlined and discussed above, we find that the applicant has provideb an acceptable response to this item.
v' Waterford SSER #1 13-2
____.______~_,a a_
3.
Describe the specific conditions under which utility officials above the
)osition of Plan _t Manager, Nuclear would succeed to the position of Emergency Coordinator.
DiscussionandConclusion Amendment 19 to the Plan states that the responsibilities of the Emergency:
Coordinctor will not.be assumed by any management personnel more senior The Plan identifies the line of succes-than-the Plant Manager - Nuclear. ~
sion for the position of Emergency Caordinator, up to and including the Plant Manager - Nuclear.
BasedbnourreviewoftheirPlanasdiscussedabove,wefind'thatthe applicant has prceided an acceptable response to this item.
4.
Make minor clarifications to the classification and EAL section of the Plan.
Discussion and Conclusion Amendmer.t 19 to the Plan incorporated, in the classification and EAL section, initiating conditions.for security emergencies for the Unusual Event, Alert and Site Area Emergency Classifications, and initiating conditions for entry of uncontrolled flammable or toxic gases into vital areas of the plant for the Site Area Emergency Classification.
Based on our review cf their Plan as discussed above, we find that the applicant has provided an acceptable response to this item.
5.
Evacuation time estimates, currently under review by the staff, must satisfy the criteria of Appendix 4 to NUREG-0654.
Discussion and Conclusion The evacuation time estimate submitted by LP&L for the Waterford 3 site-a on June 25, 1980, was reviewed against the guidance given in Appendix 4 to NUREG-0654/ FEMA-REP-1, Revision 1, " Criteria for'the Preparation and Evaluation of Radiological Emergency Response Plans in Support of Nuclear Power Plants." The review found that the applicant's evacuation study was deficient in many areas.
On August 14, 1981, the NRC requested LP&L to provide additional information in the areas identified by the review.
We are in receipt of the additional material submitted on September 29, 1981.
An initial review indicates,that the response addresses all of the elements in the August 14 request.
However, the results of the staff's review will be reported following the full review of. all information submitted by the applicant.
J Waterford SSER #1 13-3
13.5 Plant Procedures 13.5.2 Operating and Maintenance Procedures The applicant is responsibt, for describing a plan for the implementation of operating and maintenance procedures in the Final Safety Analys s Report (FSAR).
The first part of the $ n should deal Nith proceduro that are performed by licensed operators in tne control ro.m.
Each such operating procedure should be identified by title and included in a classification system.
The general format and content for each class should be described.
The following categories should be included:
(1) System procedures.
(2) Ger.eral plant procedures.
(3) Off-normal operating procedures.
(4) Emergency procederes.
(5) Alarm response procedures.
(6) Temporary procedures.
In category 5, individual alarm response procedures need not be listed.
- However, the system employed to classify or subclassify alarm responses, and the methods to be employed by operators to retrieve or refer to alarm response procedures should be described.
Immediate action procedt ns required to be memorized should be identified.
The second part of the plan should describe how other operating and maintenance procedures are classified, what group or groups within the operating organiza-tion have the responsibility for following each class of procedures, and the general objectives and character of each class and subclass.
If gene.yl objec-tives and characters of the procedures are described elsewhere in the FSAR or application, these may be described by specific reference thereto.
The following categories of procedures should be included:
(1) Plant radiation protection (2) Emergency preparedness (3) Instrument calibration tests (4) Chemical and radiochemical control (5) Radioactive waste management (6) Maintenance and modification (7) Materials control Waterford SSER #1 13-4
~-
(8) Plant security (9) Fire protection Subsection 13.5 of the Final Safety Analysis Report was reviewed in accordance with Seciton 13.5.2 (Rev. 1) of the Standard Review Plan, NUREG 75/087.
Staff review confirmed that the applicant's procedures plan includes the necessary classes of opetu'.".9 and maintenance procedures and a statement that the procedures will be completed approximately six months prior to initial fuel loading.
The applicant's plan appears to be consistent with Regulatory Guide 1.33, Revision 2, February 1978, " Quality Assurance Program Requirements (Operation)" ana it meets the requirements of 10 CFR 50.34.
Therefore, the staff concludes that the applicant's plan for operating and maintenance procedures is acceptable.
1 Waterford SSER #1 13-5
15 ACCIDENT ANALYSIS 15.3 Limiting Accidents 15.3.1 Steam Line Breaks Tha staff's SER questioned the ability of the CESEC-I computer code to properly account for steam formation within the reactor vesse', during depressurizing events.
The applicant responded to the staff's question by submitting a reanalysis of the limiting steam line break (SLB) event.
The purpose of this new analysis was to demonstrate that the FSAR conclusions would not be altered by reanalyzing the limiting depressurization event utilizing a computer program which properly accounts for primary system voiding.
The new analysis incorpo-rated automatic initiation of the auxiliary feedwater system (AFWS), as required by NUREG-0737.
Since the FSAR analyses did not account for the automatic initia-tion of the AFWS, the new analysis (utilizing the CESEC-III computer program) cannot be directly compared to the FSAR analysis (analvzed with CESEC-I).
However, the new analysis demonstrated that the FSAR conclusions derived from the CESEC-I program would not be altered by utilizing a code (CESEC-III) which was specifically designed to model primary system voiding.
Comparisons between the CESEC-I and the CESEC-III codes indicate that the system response differed appreciably.
However, since the CESEC-I program underpredicted the system pres-sure, then the DNBR was also conservatively evaluated.
The staff, therefore, finds the FSAR analysis acceptable.
The staff requested the applicant to address the consequences of manually tripping the reactor coolant pumps by the operators upon low pressure ECCS l
initiation, per the present guidance of IE Bulletin 79-068.
The applicant has addressed this request by performing a paraiaetric study of losing offsite power at various times during a large SLB. These analyses were performed utilizing.
the CESEC-III computer program.
The analyses demonstrated that the limiting time for tripping the reactor coolant pumps #as coincident with the initiation of the large MSLB.
Large steam line breaks are reviewed as post-reactor trip events since the reactor rapidly trips on low steam generator pressure.
The applicant has not analyzed s'aall and intermediate MSLBs.
During small steam line breaks, the system does not depressurize to the low pressure ECCS actuation setpoint, and therefore the reactor will trip upon an ECCS actuation signal and the resulting event will therefore not be limiting.
The applicant has not addressed the consequences of a small steam line break concurrent with loss of offsita power.
This analysis is required by the Standard Review plan Section 15.1.5, in -der tu assure that a spectrum of break sizes has been reviewed and assessed.
Although the applicant has stated that the large steam line break is limiting, we require a confirmatory analysis be pro-vided of the small s ham line break to support the applicant's conclusions.
Waterford SSER #1 15-1
3 In analyzing the above depressurization events, the applicant utilized two I
versions of the CESEC computer program.
The CESEC program consists of mathemati-cal computer models which are presently under review by the staff.
The review at this time indicates reasonable assurance that the applicant's submittal will not be appreciably altered by the completion of our review.
The applicant will be required to implement the results of any changes resulting from these reviews.
15.3.2 Feedwater System Pipe Breaks The staff's safety evaluation report requested additional information regarding i
the analytical methodology utilized in assessing the consequences of a main feedwater line break (FWLB) and the system performance during small and inter-mediate sized breaks.
The following is our evaluation of the applicant's response.
4 The applicant's methodology for analyzing FWLBs for Waterford differed from other applicants' methodology with CE-designed NSSS.
First, the applicant assumed credit for two phase flow exiting the break.
The treatment of break flow quality has typically been a major conservatism in feedwater line break j
analyses.
Other applicants with CE-designed plants assumed single phase liquid exiting the break.
Since FWLBs result in primary system pressurization, this assumption minimizes the primary system cooling throughout the transient and is therefore conservative.
During the initial phase of the blowdown, the applicant assumed the break flow to be that of saturated liquid.
As the steam generator liquid mass inventory decreased, the quality of the break flow was permitted to increase by assuming that all of the steam generator liquid mass is contained within the downcomer region, and that it forms a homogeneous two-phase mixture.
This modeling technique results in a higher energy removal rate by the break flow than if the flow was restricted to single phase liquid, but still results in a lower evaluated quality than would be expected during the actual course of events.
i j
Another conservatism, typically utilized in previous CE plant analyses, is the treatment of the broken steam generator heat transfer coefficient.
Through i
sensitivity studies, documented in Appendix 15B of the CESSAR FSAR, it was i
demonstrated that the peak reactor coolant pressure is a sensitive function of l
the degradation of the broken steam generator heat transfer rate.
The limiting system pressure is obtained by assuming a step decrease in primary to secondary system heat transfer (from the initial steady-state value to a value of zero).
This maximizes the heat imbalance between the primary and secondary systems.
As a result, the system pressurization rate is maximized.
Rather than assuming a step decrease in the broken steam generator heat transfer coefficient, the applicant assumed a heat transfer coefficient to be linear ramp function of the remaining steam generator liquid inventory, varying it from the 100% steady-state value to zero.
To demonstrate the conservative nature of this assumption, the applicant provided a curve of the expected heat transfer coefficient as a l
function of steam generator liquid inventory.
However, the applicant has not provided any basis (analytical or experimental) for the expected behavior.
A conservatism utilized by other applicants is the time assumed for reactor trip.
Reactor trip is typically assumed on high pressurizer pressure.
This is conservative, since a steam generator low level trip was not credited during the entire duration of liquid inventory blowdown from the broken secondary Waterford SSER #1 15-2
system. The Waterford analysis deviated from this practice by assuming a low-level trip at 9000 lbm of remaining liquid inventory in the secondary system.
The 9000 lbm of liquid inventory was derived by utilizing the same assumptions used in developing the break flow quality (discussed above).
The staff agrees that although the applicant's assumption is less conservative than the assumption made on previous applications, they are still conservative with respect to expected behavior.
However,~no detailed analysis or experimental verification was provided to justify the assumption made, nor has the applicant verified the net conservatism of the analyses.
In addition to the large FWLB analysis, the applicant has provided the peak reactor coolant system pressure as a function of break size.
The results showed the peak reactor coolant system pressure to exceed 110% of the design pressure and to be insensitive to break size. That is, the peak pressure for small and large feedwater line breaks are similar.
This is considered to be a result of the analytical technique and assumptions made in the analyses.
It is the staff position that small feedwater lines breaks, in the absence of single failures, should not exceed 100% of the design pressure, in accordance with the SRP.
The staff will permit the pressure to reach levels associated with the ASME Pressure Vessel Service Level C Limits for a feedwater line break in conjunction with a single active failure.
Therefore, the applicant is required to provide confirmatory analyses which demonstrate that small feedwater line breaks in the absence of single failures will not result in a primary system pressure in excess of 110% of design pressure. However, the staff has reasonable assurance that utilizing a justifiable analytical model, the calculated peak primary system pressure for small feedwater line breaks without single failures will be below 110% of the design pressure.
Therefore, this item will remain open but confirmatory in nature.
The staff intends to continue its evaluation of the modeling assumptions made in the applicant's analysis of a feedwater line break.
Should the results of this evaluation indicate the assumptions made to be inappropriate, we will require the applicant to modify the analysis model accordingly and reanalyze the event.
Appropriate modifications will be required to assure the SRP criteria are met.
15.3.4 Steam Generator Tube Rupture The applicant's analyses of the steam generator tube rupture events were performed using the CESEC-I computer program.
CESEC,I does not account for formation of voids in the reactor coolant system when the pressurizer empties or when the primary system saturates.
For this event some void formation is expected and the CESEC-I results might not be applicable.
The staff required the applicant to reanalyze the steam generator tube rupture (SGTR) with a suitable model or provide justification why the CESEC-I analyses properly account for steam for. nation in the primary system.
The applicant has provided an analysis of the SGTR event with a loss of offsite power, using the
.CESEC-III computer program.
CESEC-III explicitly models steam formation in the reactor vessel upper head region.
This event is the most limiting with respect to the duration of voids in the reactor coolant system.
Waterford SSER #1 15-3
l ihe modeling of the upper head region results in a predicted maximum void volume of 535 cubic feet.
The amount of voids predicted is not large enough to expand the steam bubble beyond the upper head region and to the elevation i
of the hot legs.
Natural circulation cooldown of the reactor coolant system will not be impaired. The plant can be maintained in a stable condition by the collapse of the upper head voids through manual control of the cooldown rate.
Shutdown cooling entry conditions can be achieved by cooling down the reactor coolant system at a prescribed cooldown rate using the intact steam generator.
The CESEC-III analysis shows the minimum DNBR decreases from 1.28 (CESEC-1) to 1.21.
The offsite dose to the thyroid, at the exclusion area boundary, is estimated to increase from 66.5 rem to 73.0 rem, which is acceptable because it is still less than 10 CFR Part 100 criteria.
The staff finds the applicant's analysis to be acceptable and adequately addresses the concerns associated with the formation of voids in the reactor coolant system.
The applicant has demonstrated compliance with GDC 10, 15 and 26 for this event.
15.3.6 Anticipated Transients Without Scram In the SER, the staff required the applicant to develop emergency procedures for the postulated ATWS.
In response to NRC requirements, the applicant has submitted an ATWS procedure which has been reviewed by the staff and personnel from Battelle Pacific Northwest Laboratories.
The review disclosed a need for several minor revisions to the procedure.
The applicant has committed to revise the procedure to include our comments.
The staff will verify that these revi-sions are completed prior to issuance of an operating license.
The revised procedure will provide an acceptable basis for licensing and interim operation of Waterford 3 pending the outcome of the Commission rulemaking on ATWS in accordance with General Design Criteria 10, 15, 26, 27, and 29 of 10 CFR 50, Appendix A.
The Commission will, by rulemaking, determine any future modi-fications necessary to resolve ATWS concerns and the required schedule for implementation of such modifications.
Waterford SSER #1 15-4
18 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS On July 9,1981, the staff issued a Safety Evaluation Report on Waterford 3.
During its 256th meeting on August 6-8, 1981, the ACRS reviewed all aspects.'
the Waterford 3 application for an 7perating license.
A copy of the Committee's report, dated August 11, 1981 is included in Appendix C to this report.
The Cominittee was concerned with Louisana Power and Light's management and staffing for Waterford 3.
Their letter asks for another meeting on this subject when progress is made toward iraprovement.
The staff's review of this issue has not yet been completed, and will be discussed in a future supplement to the SER.
Waterford SSER #1 18-1
20 FINANCIAL QUALIFICATIONS The Nuclear Regulatory Commission's regulations relating to the determination of an applicant's fina mial qualifications for a facility construction permit are 10 CFR 50.33(f) ar.; Appendix C to 10 CFR Part 50.
At our request, Louisiana Power and Light Company submitted information detailing its proposed consolida-tion with another utility and the effect such consolidatien would have upon the ability to finance the operation of the Waterford facility.
The following analysis summarizes our review of this submittal and evaluates the financial qualifications of the surviving corporation resulting from the consolidation -
Louisiana Power and Light Company - to obtain the necessary funds to continue the construction of the Waterford Steam Electric Station, Unit No. 3.
Furthermore, this evaluation updates the staff's analysis of the applicant's financial qualifications as stated in Section 20 of the Waterford Safety Evaluation Report.
20.1 Business of Applicant (1) Description of Applicant and Party to Consolidation Louisiana Power and Light Company (LP&L) is an electric utility operating in two retail regulatory jurisdictions.
Approximately 97 percent of LP&L's business is under the jurisdiction of the Louisiana Public Service Commission with the remainder subject to the jurisdiction of the City Council of the City of New Orleans.
LP&L is a wholly owned subsidiary of Middle South Utilities, Inc.,
(MSU) an investor-owned holding company.
New Orleans Public Service Company, Inc. (N0 PSI) is an electric utility subject to regulation by the City Council of the City of New Orleans.
N0 PSI is also a wholly owned subsidiary of MSU.
On November 28, 1981, the voters of Orleans Parish will vote on changing the jurisdiction of both LP&L and N0 PSI from the City Council to the Public Service Commissica.
In all retail rate cases to date the City Council has granted LP&L the same rates for the customers under its jurisdiction as were granted by the Louisiana Public Service Commission.
(2) Proposed Consolidation LP&L and NOPSI have begun development of a plan to consolidate their operations which could be effective by early 1982.
The new company to emerge from the consolidation will be called Louisiana Power and Light Company (LP&L) and the area now served electricity by N0 PSI will become a fourth operating division of LP&L.
The new LP&L company will be a wholly owned subsidiary of MSU.
20.4 Reasonable Assurance of Funds 20.4.1 General Like the present LP&L company, the consolidated corporation will remain an electric utility providing service to the combined service area.
The present LP&L would represent about 80 percent of the combined company in customer Waterford SSER #1 20-1
i electric energy use and peak requirements and N0 PSI would constitute approxi-mately 20 percent.
Tables 20-1 and 20-2 show the respective Pro-Forma Consolidated Statements of Capitalization and Income for the emerging LP&L company.
As will be noted from review of Tables 20-1 and 20-2, the emerging LP&L company j
will be an expanded entity for financial purposes.
Accordingly, it is reason-able that the expansion of LP&L's financial resources as a result of the con-solidation will enhance its financial qualifications to operate the Waterford facility.
As stated in the Waterford Safety Evaluation Report for the present LP&L company, the emerging company will likewise recover both operating and decommissioning costs for Waterford through revenues derived from rates charged for electric service.
i 20.5 Conclusion Accordingly, the staff has determined that LP&L has reasonable assurance under 10 CFR 50.33(f) of obtaining the necessary funds to cover the estimated opera-ting costs for the facility.
In this respect, the applicant has demonstrated that it has available resources sufficient to cover estimat'ed costs for each of the first five years of operations plus the estimated costs of permanent shut down and maintenance of the facility in a safe condition, as required by 10 CFR Part 50, Appendix C(I) (B).
As a consequence of this, the staff finds that the consolidated LP&L will be financially qualified to operate and safely decommission Waterford Steam Electric Station, Unit No. 3.
In summary, this conclusion is based upon the company's status as a public utility, the increased size of its operations, the demonstrated ability to achieve revenues sufficient i
to cover the operating and capital costs of both the present LP&L company and NOPSI, their successful history of obtaining capital in amounts both internally generated and in the external markets, and the reasonable expectation that the consolidated LP&L company will continue to do so.
l l
l l
Waterford SSER #1 20-2
Table 20-1 Louisiana Power & Light Company (LP&L) and New Orleans Public Service Inc.
(N0 PSI) Pro Forma Consolidated Statement of Capitalization June 30, 1981 (Unaudited)
(In Thousands)
Consolidated LP&L N0 PSI Pro-Forma Long-Term Debt
$ 903,523
$126,508
$1,030,031 Preferred Stock with Sinking Fund 121,381 14,582 135,963 Preferred Stock without Sinking Fund 145,882 20,117 165,999 Common Stock 498,900 59,359 558,259 Retained Earnings 65,900 9,797 75,697 Total Capitalization 1,735,586 230,363 1,965,949 Notes Payable due within one year 63,192 4,000 67,192 Current Maturing Long-Term Debt 52,224 52,224 Total Capitalization Including Short-Term Debt
$1,851.002
$234,363
$2,085,365 Waterford SSER #1 20-3
Table 20-2 Louisiana Power & Light Company (LP&L) and New Orleans Public Service Inc.
(N0 PSI) Pro Forma Consolidated Statement of Income For the Twelve Months Ended June 30, 198.1 (Unaudited)
(In Thousands)
Consolidated LP&L NOPSI Pro-Fo rma Operating Revenues
$977,324
$431,798
$1,409,122 Operating Expenses 849,211 417,735 1,266,946 Operating Income 128,113 14,063 142,176 4
i Other Income 60,420 2,441 62,861 Interest and Other Charges 70,684 10,277 80,961 Net Income 117,849 6,277 124,076 Preferred Dividend Requirements 26,783 3,281 30,064 i
j '
Balance for Common Sto:k
$ 91,066
$ 2,946 94,012 i
Note: No attempt has been made to eliminate intercompany transactions, the only significant items being sales for resale and purchased power transactions through the Middle South Power Pool.
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i Waterford SSER #1 20-4 I
22 TMI-2 REQUIREMENTS 22.2 Discussion of Requirements I.C.1 - Guidance for the evaluation and development of procedures for transients and accidents Position In letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guide-lines, upgrade emergency procedures, and to conduct operator retraining (see also item I.A.2.1).
Emergency procedures are required to be consistent with the l
actions necessary to cope with the transients and accidents analyzed.
Analyses
{
of transients and accidents were to be completed in early 1980 and implement-l ation of procedures and retraining were to be completed 3 months after emergency procedure guidelines were established; however, some difficulty in completing j
these requirements has been experienced.
Clarification of the scope of the task and appropriate schedule revisions were included in NUREG-0737, item I.C.1.
Pending staff approval of the revised analysis and guidelines, the staff will l-continue the pilot monitoring of emergency procedures described in Task Action Plan item I.C.8 (NUREG-0660).
For PWRs, this will involve review of loss-of-coolant, steam generator-tube rupture, loss of main feeoYater, and inadequate core cooling procedures. The adequacy of each PWR vendor's guidelines will be identified for each near term operating license (NTOL) during the emergency procedures review.
I Discussion and Conclusion The Combustion Engineering (CE) Owners' Group revised analysis and guidelines required by Task Action Plan item I.C.1 (3), as clarified in NUREG-3737, were submitted for staff review on June 30, 1981.
Their review by the rtaff has not l
been completed. Therefore, interim CE Owners' Group guidelines that have been l
reviewed by the staff were used in the evaluation of the selected emergency l
procedures. The CE Owners' Group interim guidelines were submitted in CEN-117 (Inadequate Core Cooling, October 1979) and CEN-128 (Transients and Accidents).
Review of the CE Owners' Group interim guidelines has been performed in conjunc-tion with the review for I.C.8 - Pilot Monitoring of Selected Emergency Proce-dures for NT0Ls.
The interim guidelines are adequate except for the CE Owners' Group inadequate core cooling guidelines.
The guidelines for the long-term reanalysis of transients and accidents, including inadequate core cooling, were under development when this deficiency was identified.
Therefore, CE did not expend resources to address staff concerns on the interim guidelines.
Since additional work on the interim guidelines was not performad by CE, the staff conducted a review of the San Onofre Units 2 and 3 inadequate core cooling procedure to be used as an interim technical guideline.
The review of this Waterford SSER #1 22-1
. ~. - - - -- -
San Onofre p mcedure to determine if it was technically adequate was performed to the same depth as would have been performed.for a generic guideline.
At a meeting on April 10, 1981, the remaining unresolved staff comments on the inadequate core cooling procedure were discussed, and the applicant for the San Onofre Units 2 and 3 committed to specific changes in the procedures to address the staff's concerns. After incorporation of these concerns in the i
inadequate core cooling procedure, the use of the San Onofre procedure as a guideline was considered acceptable for Waterford 3.
The Waterford emergency operating procedures were received on May 19, 1981, and have been reviewed by l
the staff and personnel from Battelle Pacific Northwest Laboratories in accord-ance with item I.C.8 of the Task Action Plan.
Based on.the staff review of the emergency operating procedures developed from l
the CE Owners' Group Guidelines and the San Onofre inadequate core cooling pro-cedure, and the staff's observation of the procedures being implemented on a simulator and in a walk-through in the control room, the staff concludes that the guidelines have been adequately incorporated.
This fulfills the requirements of Section I.C.1 of NURE6-0694.
i I.C.7 - NSSS vendor review of procedures Position Obtain NSSS veador review of power ascension and emergency operatir, procedures to further verify their adequacy.
l This requirement must be met before issuance of a full power license.
Discussion The NSSS Vendor, Combustion Engineering, will review the emergency operating procedures, the low power test procedures and the power ascension test procedures.
The applicant has committed to ensuring these reviews are complete prior to fuel l
loading.
The staff will review the final revision of the procedures selected for I.C.8 to verify that the vendor review has been implemented acceptably, prior to i
issuance of an operating license.
I.C.8 - Pilot monitoring of selected emergency procedures for NT0L Applicants.
l Position Correct emergency procedures as necessary based on the NRC audit of selected plant emergency operating procedures (e.g., small-break LOCA, loss of feedwater, restart of engineered safety features fol'owing a loss of ac power and steam-
[
line break).
This action will 'oe completed prior to issuance of a full power license.
Discussion Emergency operating procedures for mitigating the consequences of steam gener-ator tube rupture, loss-of-coolant accident, inadequate core cooling, anticip-ated transients without scram, and loss of normal feedwater were received on l
l Waterford SSER #1 22-2
-~
May 19, 1981. The staff and personnel from Battelle Pacific Northwest Laborator-ies reviewed the procedures to ensure that they were consistent with the plant design and incorporated applicable human factors considerations.
A meeting was held with the applicant on June 25, 1981, to discuss the results of these reviews, the Waterford 3 plant characteristics, and the CE Owners' Group Guidelines.
The staff and contractor personnel reviews resulted in several pages of general comments and numerous detailed comments on the procedures.
The general comments included human factors consideration on the use of the procedures under stress, inappropriate operator actions not specifically prohibited by the pro edures, and comments related to consistency among the procedures.
The specific comments concerned a lack of dett il in the procedures based on an assumed level of operator knowicdge, procedurai compatibility with the control room indications and controls, resolution of abiguous wording and logic statements, use of cautions and i;otes, usability of figures and tables, and the interface with normal and abnormal operating procedures.
These con ents p ovided a basis for discussion. During the discussions on June 25, 1981, many of the comments were resolved by consideration of the control room design, olant characteristics, administrat've controls, and operator training.
As a result of these discussions and an independent review of the procedures by the applicant, many changes were made to the procedures.
On July 28 and 29, 1981, procedures revised to incorporate the changes discussed in the June 25, 1981 meeting were employed to direct operator responses to simulated transients and accidents on the Palo Verde simulator.
A team of NRC and contractor personnel observed the simulator exercises and discussed the l
exercises with the operations personnel following each simulation.
The transients l
and accidents simulated included a wide range of simulations from minor transients to major accidents with multiple system failures.
The transients and accidents involving multiple systems failures simulated included:
(1) Feed pump turbine trips at 100% power; (2) Steam generator tube rupture at 100% power; (3) Total loss of feedwater at 100% power followed by a complete loss of AC power; (4)
Loss of all AC power followed by a main steam line rupture; (5) A small break loss of coolant; (6) A spu.ious scram with the failure of all control rods to scram (ATWS). Other transients and accidents simulated included:
(1) Loss of offsite power and failure of one and two diesel generators, (2) Failure of scram breakers to open (ATWS), (3) Failure of individual components of the emergency core cooling systems, and (4) Support system failures such as the computer-driven system displays. The discussions between the first few simula-tions centered on the use of the procedures and the roles of the operators and supervisors.
This included the division of labor, operator to operator communica-tions, and mutual support of operator activities.
The applicant has agreed to l
incorporate relevant information on the technical basis of procedural steps and procedure utilization into the training program to be conducted on the I
revised procedures.
As a result of these exercises and discussions, some additional editorial changes and clarifications for the operator were added to the emergency operating procedures.
On July 29, 1981, the same team of reviewers observed the same team of Waterford 3 control room operators participate in a walk-throuch of the revised procedures in the Waterford 3 control room.
The simulated event was a small break LOCA, taken to the point of inadequate core cooling, recovery of adequate core cooling, and plant shutdown.
The scenario included multiple failures and several failures beyond the design basis, which exercised a majority of the Waterford SSER #1 22-3
emergency procedures. The procedures were discussed with the operationc personnel during and after the event.
The manner in which the procedures were executed indicated that the emergency operating procedures were generally clear, properly sequenced, and compatible with the control room equipment and arrangements.
The applicant has committed to revise the remaining emergency operating proce-dures to incorporate revisions similar to those made to the procedures that were reviewed in detail t;y the staff under I.C.8.
With incorporation of these comments we conclude that the emergency operating procedures will be acceptable for operation to 100% rated power.
The staff will vcrify that the appropriate revisions have been incorporated prior to issuance of an operating license.
Future actions required by additional staff review of the submittal from the CE Owners' Group and staff p.sitions developed to implement Task Action Plan item I.C.9, Long-Term Program for Upgrading of Procedures, may require future revisions to the emergency operating procedures.
Waterford SSER #1 22-4
l i
APPENDIX A l
CONTINUATION OF CHRONOLOGY OF l
RADIOLOGICAL REVIEW l
i l
June 15, 1981 Letter to applicant regarding schedule change for l
submittal and evaluation of upgradad emergency plan June 16, 1981 Letter to applicant transmitting request for additional information on radiation emergency plan June 17, 1981 Letter from applicant concerning net shutdown group worth / stuck CEA test June 17, 1981 Letter to applicant transmitting request for additional information June 19, 1981 Letter from applicant transmitting "Evaluath n Report on Control of Heavy Loads," Part I, June 1981 L
June 19, 1981 Letter from applicant transmitting information requested by Effluent Treatment Systems Branch June 24, 1981 Letter to applicant concerning fire protection review June 24, 1981 Letter from applicant forwarding information to address SER Open Items 48 and 49, " Reanalysis of Category I Structures" and "Re evaluate Foundation Mat for Changes in the Value of the Subgrade Modulus" June 24, 1981 Letter from applicant advising that security plan transmitted June 11 should be withheld from public disclosure June 29, 1981 Letter to applicant transmitting request for additional ~
information June 29, 1981 Letter from applicant transmitting Revision 6 of Physical Security Plan i
June 30, 1981 Letter from applicant forwarding training program for mitigating accidents involving core damage July 1, 1981 Letter to applicant transmitting "AE0P Observations and Recommendations Concerning the Problem of Steam Generator Overfill and Combined Primary and Secondary Side Blowdown" July 2, 1981 Meeting with applicant to discuss safe shutdown analysis Waterford SSER #1 A-1
4 APPENDIX A (continued)
July 6, 1981 Letter to applicant advising that INPO reports will be handled through Regional offices July 8, 1981 Letter from applicant transmitting schedule for responding to June 29 letter July 8, 1981 Letter from applicant discussing open items of Effluent Treatment Systems Branch July 8, 1981 Letter from Ebasco transmittirq information to support response to questions concerning containment system compliance with Gener&l Design Criterion 51 July 8, 1981 Letter from applicant providing revised response regarding its ability to initiate shutdown cooling.from control room July 9, 1981 Letter to applicant concerning privacy and proprietary material in emergency plans July 9, 1981 Issuance of Safety Evaluation Report July 14, 1981 Letter from applicant transmitting Amendment No. 20, consisting of additional information to address SER open items, additional TMI information, and revised-responses to certain questions July 17, 1981 Letter from applicant forwarding information on post 4
accident sampling caoability July 17, 1981 Meeting with applicant to discuss turbine missiles and tour plant July 17, 1981 Letter to applicant regarding capability to provide prompt notification in the event of an emergency July 21, 1981 Letter from applicant confirming commitment to adopt methods given in Regulatory Guides 1.58 and 1.146 July 24, 1981 Letter from applicant transmitting physical security plan in connection with its application for Special Nuclear Material License July 28 31, 1981 Meeting wit.h applicant to conduct emergency operating procedures walk-through July 30, 1981 Meeting with applicant to hear applicant's presentation on the post-accident sampling system July 31, 1981 Letter to applicant re-transmitting letter of July 1, 1981 j
Waterford SSER #1 A-2
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APPENDIX A (continued)
August 3, 1981 Letter from applicant concerning effect of potential accidental releases of hazardous chemicals from nearby industrial and transportation facilities August 5, 1981 ACRS Subcommittee meeting with staff and applicant August 6, 1981 ACRS meeting with staff and applicant August 7, 1981 Meeting with applicant to discuss applicant's recent announcement of plans to consolidate with New Orleans Public Service, Inc.
August 11, 1981 Interim Report from ACRS August 14, 1981 Letter from applicant transmitting Amendment No. 21, consisting of additional information to address SER open items, including site hazards sad the Q list, and miscellaneous changes August 19, 1981 Letter to applicant transmitting request for additional information August 19, 1981 Letter from applicant forwarding containment purge valve operability study August 20, 1981 Letter from applicant transmitting revised physical security plan August 26, 1981 Meeting with applicant to discuss preservice insp'ection program August 31, 1981 Meeting with applicant to discuss safe shutdown analysis and other fire protection issues August 31, 1981 Letter from applicant forwarding updated financial information, information on organizational structure, and need for power in connection with proposal to consolidate with New Orleans Public Service, Inc.
September 4, 1981 Letter from applicant forsarding information regarding steam voiding in the reactor vessel analysis, loss of offsite power, and clarification of transient analyses with potential for fuel damage i
September 9-11, 1981 Meeting with applicant to discuss fire protection September 14, 1981 Letter from applicant reporting its activities relating to recruiting and upgrading total corporate capability September 15-18, 1981 Seismic qualification review team audit Waterford SSER #1 A-3
APPENDIX A (continued)
September 21, 1981 Letter from applicant transmitting Part II of report on requirements for overhead handling systems (with regard to control of heavy loads)
September 22, 1981 Letter from applicant transmitting revised response to question on validity of safety analyses for future fuel cores September 22, 1981 Letter to aFplicant transmitting request for additional information on quality assurance September 28, 1981 Letter from applicant transmitting Amendment No. 22 consisting of revision to Chapter 17 of FSAR and other revised information September 29, 1981 Meeting with applicant to discuss staffing and management October 1, 1981 Meeting with applicant to discuss fire dampers October 2, 1981 Meeting with applicant to discuss Waterford and CESSAR licensing and scheduling impacts of CPC software modifications required to incorporate reactor power cutback system Waterford SSER #1 A-4
APPENDIX B Continuation of Bibliography
- Letter dated September 29, 1981, from L. V. Maurin to R. L. Tedesco submitting additional emergency planning information.
Letter dated October 7, 1981, from L. V. Maurin to R. L. Tedesco submitting additional emergency planning information.
USNRC Regulatory Guide 1.33, Revision 2, Rebruary 1978, " Quality Assurance Program Requirements (0peration)."
She11 tech Report 80-2, " Evaluation of Reinforced Openings in large Steel Pressure Vessels."
USNRC, " Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No.
3," NUREG-0787, July 1981.
Electric Power Research Institute Report NP-232, " Fracture Toughness Date for Ferritic Nuclear Pressure Vessel Materials" Department of Defense, Aerospace Structural Materials Handbook.
USNRC Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor vessel Materials."
American National Standards Instituta, Standard N101.2, " Protective Coatings (Paints) for the Nuclear Industry," 1972; N5.12, " Protective Coatings (Paints) for the Nuclear Industry,"1974.
USNRC Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants."
"Available for inspection and copying for a fee in the NRC Public Document Room, 1717 H Street, NW, Washington, D. C.
Waterford SSER #1 B-1
APPENDIX C
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UNITED STATES i
!4 NUCLEAR REGULATORY COMMISSION
{.-
i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS t',
g WASHINGTON, D. C. 20555
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August 11, 1981 The Honorable Nunzio J. Palladino
' Chairman
-U.S. Nuclear Regulatory Commission Washington, D.C.-
20555
SUBJECT:
. INTERIM REPORT ON THE WATERFORD STEAM ELELIRIC STATION UNIT 3
Dear Dr. Palladino:
4 During'its 256th meeting, August 6-8 1981, the. Advisory Committee on Reactor Safeguards reviewed the application of Louisiana Power & -Light Company (Applicant) for a license to operate the Waterford Steam Electric Station Unit 3 (Waterford-3).
This project has been considered at Subcommittee meetings on June 18-19, 1981 in St. Charles Parish,- Louisiana, and on August 5, 1981 in Washington, D.C.
A tour of the facility was made by Subcommittee members on June 18, 1981.
During its review, the : Committee had-the benefit of discussions with representatives of the Applicant and the NRC Staff.
The Committee also had!the benefit of the documents listed.= The Committee commented on the construction permit application for this unit in its report t
dated January 17, 1973.
Waterford-3-is located on the bank of the Mississippi River near Taft, Louisiana in St. Charles Parish.
The city of New Orleans is approximately 25 miles east-southeast from the: plant site and Baton.Rcuge is approximately 50 miles north-northwest. The largest town within 10 -miles of: the site is Reserve, Louisiana, which had a population of approximately.7000 in 1977.
Waterford-3 uses a Combustion Engineering nuclear steam supply system with a rated power level of 3410 MWt.
The architect-engineer is Ebasco Services, Inc.
The containment is-a free. standing. steel pressure vessel enclosed wi thin a ' reinforced concrete shield-building.1 The - containment building,
' auxiliary building, fuel-handling building,' and ultimate heat-sink are located on'a-common base mat, forming a self-contained nuclear island.
~
Louisiana Power & Light (LP&L) is a part of Middle South Utilities (MSU).
Although Waterford-3 is the first nuclear plant-to ~be operated by the Appli-cant; the MSU system has two operating nucleer plants, Arkansas-Nuclear One Units 1 and 2, which are being operated Ly Arkansas Power and Light Company.
Two additional plants in the MSU system, Grand. Gulf Nuclear Station-Units 1 and 2, are under construction by Mississippi Power and Light.. MSU provides some technical services to support the nuclear units in its system.
Waterford SSER#1 C-1
Honorable Nunzio J. Palladino August 11, 1981 The Applicant described the management, the operating organization, and the status of staffing.
The NRC Staff has not completed its review of these matters, but reported its conclusion that the management and staffing at Waterford-3 is less well established than at other nuclear plants at a similar time during their construction and startup schedule.
The LP&L management has not yet been successful in putting together the team of experienced and qualified personnel which we believe will be necessary to successfully operate the plant. Of particular concern is the lack of nuclear experience throughout the organization and the apparent lack of appreciation by high-level management of the magnitude of the project it is undertaking.
We believe that an extraordinary effort will be required to prepare the LP&L management and staff for operation of the Waterford-3 plant. We also believe that a more concerted effort is needed to build an integrated organization of LP&L and contractor personnel for startup and operation of Waterford-3.
We recomeend that the adequacy of management and staffing be established prior to fuel loading.
We will continue to review this matter with the Applicant and the NRC Staff.
The Applicant described -the three safety review committees which will be a permanent part of the Waterford-3 organization.
We believe that better use could be made of experts from sources other than the Applicant's organization and its contractors to provide professional experience in areas such as training, human factors engineering, and reactor safety.
We recommend that the Applicant make a greater effort to include recognized experts, especially on its Safety Review Committee.
[
Although a sincere effort has been made to establish a comprehensive training program at Waterford-3, it has suffered from a lack of professional direction.
We believe the Applicant should move as soon as possible to employ a highly qualified professional for the key position of training director and provide him with the resourr.s needed to build an effective program.
Waterford-3 is located in a highly industrialized area with an unusually large concentration of sources of hazardous substances from nearby industries and transportation r%ces.
We believe.the Applicant has done a comme.ndable job in analyzing these hazards and providing for protection of the plant by both equipment design and administrative procedures.
The NRC Staff has not completed its review of this matter, but we believe it can be resolved sati sfactorily.
The Waterford-3 control room makes extensive use of a computer system for monitoring and control of the plant, and for evaluating plant perfonnance.
We commend the initiative the Applicant has shown in this area and the continuing effort to integrate the control room equipment with opera + ' '
procedures and human factors considerations.
Watirford SSER#1 C-2 e
Honorable Nunzio J. Palladiro August 11, 1981 Waterford-3 has a unique ultimate heat sink design.
It is contained within the nuclear island and is protected from extreme environmental effects.
It consists of two trains of wet and dry cooling towers.
Sufficient water is stored on the nuclear island to meet the needs for shutdown decay heat removal. We believe the design is acceptable.
The Applicant has performed an analysis of total loss of AC power.
The DC power supply is capable of supplying essential loads for at least two hours and the condensate supply is sufficient for a longer period.
We recommend that the Applicant expand this analysis to consider the effect of loss of space cooling on essential electrical equipment and to also consider the effect of coolant leakage from the primary system.
Evaluation of these matters is a generic issue.
Studies for this plant need not be completed prior to startup.
We note that a number of items have been identified as Outstanding issues in the NRC Staff Safety Evaluation Report dated July 1981.
These include some TMI-2 Action Plan requirements. We believe these issues can be resolved in a manner satisfactory to the NRC Staff, subject to the concerns on instrumenta-tion for detection of inadequate core cooling expressed in the ACRS letter to the Executive Director for Operations dated June 9,- 1981.
The Committee believes that, contingent on the Applicant's attainment of an adequate level of management and staffing, if due consideration is given to the recommendations above, and subject to satisfactory compl etion of construction and preoperational testing, there is reasonable assurance that Waterford Steam Electric Station Unit 3 can be operated at power levels up to 3410 MWt without undue risk to the health and safety of the public.
We expect to report further on the adequacy of the staffing and management as progress is made toward improvement.
Sincerely, J. Carson Mark Chairman
References:
1.
Louisiana Power & Light Company, "Waterford Steam Electric Station, Unit 3 Final Safety Analysis Report," with Amendments 1 through 20.
2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit 3," Docket No.
50-382, USNRC Report NUREG-0787, July 1981.
Waterford SSER#1 C-3
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- 1. REPORT NUMBE R Lamped by CDCI WREWW BIBLIOGRAPHIC DATA SHEET Supplement No. 1 4 Ti1LE AND SU8TsTLE (Add Volume No.. d worconeNi 7 fleave Diankt Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3 3 RECIPIENT 3 ACCESSION No.
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- 5. DATE REPORT CoMPLE TED MONTH lvtAn October 1981 9 PERFORMING ORGANIZATION NAVE AND MAltlNG AcoRESS Onclude l@ Codel DATE REPORT ISSUED uoNTH lvtAm U.S. Ihclear Regulatory CoCmission October 1981 Office of Nuclear Reactor Regulation s tre e u.a*J Washington, D.C.
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information submitted since the Safety Evaluation Report was issued, (2) the' staff's evaluation of matters under review when the Safety Evaluation Report was l
issued, and (3) the staff's responses to comments made by the Advisory Committee en Reactor Safeguards in its report dated August 11, 1981.
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