ML20062H245

Safety Evaluation Report Related to the Operation of Clinton Power Station,Unit No. 1.Docket No. 50-461.(Illinois Power Company,Et Al.)

ML20062H245
Person / Time
Site:	Clinton
Issue date:	07/31/1982
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0853, NUREG-0853-S01, NUREG-853, NUREG-853-S1, NUDOCS 8208130470
Download: ML20062H245 (84)

Text

r NUREG-0853 Supplement No.1 Safety Evaluation Report related to the operation of Clinton Power Station, Unit No.1 Docket No. 50-461 l

Illinois Power Company, et al.

l U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l

July 1982 i t

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ADOtk OCO 1 E PDR

t NOTICE Availability of Reference Materials Cited in NRC Publications - '

Most documents cited in NRC publications will be available from one of the following sources: j

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Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include N RC correspondence and internal NRC memoranda: NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; i Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and 4 licensee documents and correspondence.

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l NUREG-0853 Supplement No.1 i Safety Evaluation Report I related to the operation of l Clinton Power Station, Unit No.1 Docket No. 50-461 Illinois Power Company, et al.

i U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1982 p* ** %

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ABSTRACT Supplement No. 1 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative Inc., end Western Illinois Power Cooperative Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supple-ment reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.

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TABLE OF CONTENTS Page ABSTRACT . ............ .............. ...................... .... iii 1 INTRODUCTION AND GENERAL DISCUSSION .................... .. ..... 1-1 1.1 Introduction ..... .. ..... ........... .................. 1-1

1. 9 Outstanding Issues ......... ...... .......... ... ........ 1-2 1.10 Confirmatory Issues . .. ..... .... ........ ... .......... 1-3 1.11 License Conditions ......... ......... . ... .... .. .. .. 1-7 2 SITE CHARACTERISTICS ..... ... . .. .. ... . ... ............... 2-1 2.3 Meteorology .. .... . . .. ............................... 2-1 2.3.3 Onsite Meteorological Measurements Program . ........ 2-1 2.4 Hydrology . ...................... .. . ...... . .......... 2-1 2.4.4 Hydrologic Evaluation of the Ultimate Heat Sink ..... 2-1 2.4.5 Erosion Protection for Cooling Water Canals and Reservoirs..................... ......... ........... 2-2 2.4.6 Groundwater ............ ...... . ............ .. ... 2-3 2.4.8 Conclusions ............ . ... ....... ....... ..... 2-3 2.6 Stability of Subsurface Materials and Foundations .... ... . 2-3 2.6.1 Site Conditions ... .... ....... ........... ..... . 2-3 l 2.6.1.2 Site Investigation ... . ........... ...... 2-3 1

i 2.6.3 Foundation Stability .. .. ... ....... . . ........ 2-4 l

2.6.3.3 Settlement .. .. . . ...... . . ......... 2-4 2.6.4 Stability of Slopes ... .... ....... ........... .... 2-5 2.6.4.4 Stability Analysis ....... ....... ... . 2-5 2.6.5 Dams . .... ...... . ... . ... .... .. .......... 2-6 2.6.5.5 Embankment .. . ........ .............. . 2-6 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS ..... ... 3-1 3.5 Missile Protection .... ...... .. ... .... ...... . 3-1 3.5.1 Missiles ... . . . .... . .... . ... . . .. 3-1 Clinton SSER 1 v I

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4 3.5.1.1 Internally Generated Missiles (Outside

Containment)................................ 3-1

3.5.1.2 Internally Generated Missiles (Inside 1

Containment)................................

3.6 Protection Against Dynamic. Effects Associated With the i- Postulated Rupture of Piping- ..... ..... .............. ... 3-1 L 3

3.6.1. Determination of Break Locations and Dynamic Effects Associated With the Postulated Rupture of Piping .... 3-1 j  !

! .3.7 Seismic Design .. .... ........... ......................... 3-2 i 3.7.1 Seismic Input ...... ....................... ...... . 3-2 3 3.7.2 Seismic Subsystem' Analysis ..................... ..... 3-2 3.8 Design of Seismic Category I Structures ........ ........... 3-3 3.8.1 Concrete Containment ....... ............. .......... 3-3 3.8.3 Other Category I Structures ...... ............ . ... 3-4 3.9 Mechanical Systems and Components .......................... 3-5 l 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures ..... .... .... 3-5 3.9.6 Inservice Testing of Pumps and Valves ...... ........ 3-6 l.

! 4 REACTOR . ... .............................................. ..... 4-1 4.4 Thermal and Hydraulic Design ............ ..... ............ 4-1  ;

4.4.1 Evaluation ..................... ......... ......... 4-1 i 4.6 Functional Design of Reactivity Control Systems .... .... . 4-1 t

5 REACTOR COOLANT AND CONNECTED SYSTEMS ............... ...... ... . 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary ..... ....... 5-1 5.2.3 Reactor Coolant Pressure Boundary Materials .. . . ... 5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing ........... .... ... .. ...... 5-1 l

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5.3 Reactor Vessel . . .. .............. ........ . ... ....... 5-3 i

5.3.1 Reactor Vessel Materials .... . ..... . .. . ..... 5-3 ,

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5. 3.1.1 Evaluation of Compliance With Appendix G, 10 CFR 50 .... ........ ....... .......... 5-3 5.3.1.2 Evaluation of Compliance With Appendix H, i 10 CFR 50 ... .. ... ....... ........... ... 5-5 i

! Clinton SSER 1 vi

l TABLE OF CONTENTS (Continued)

Page 5.3.3 Reactor Vessel Integrity ............................ 5-6 6 ENGINEERED SAFETY FEATURES ........................................ .6-1 6.3 Emergency Core Cooling System .............................. 6-1 6.4 Control Room Habitability .................................. 6-1 6.6 Inservice Inspection of Class 2 and 3 Components ........... 6-2 7 INSTRUMENTATION AND CONTROLS ...................................... 7-1 7.2 Reactor Protection System (RPS) ............................ 7-1 7.2.3 Resolution of Issues .................. ............. 7-1 7.2.3.2 Response Time Testing .............'......... 7-1 7.2.3.4 Analog Trip Modules (ATMs) and Optical Isolators (0Is) ............................ 7-1 7.2.3.5 NSPS Susceptibility to Electrical Noise .... 7-1 7.3 Fngineered Safety Features System .......................... 7-2 7.3.3 Resolution of Issues ................................ 7-2 7.3.3.5 Restart of Core Spray and Low-Pressure Coolant Injection Systems (TMI Action Plan Item II.K.3.21) .......................... . 7-2 8 ELECTRIC POWER SYSTEM .................................. ........ 8-1 8.4 Other Electrical Features and Requirements for Safety ...... 8-1 8.4.7 Physical Identification and Independence of Redundant Safety-Related Electrical Systems ................... 8-1 9 AUXILIARY SYSTEMS ................. ... ....... . . ............... 9-1 9.5 Fire Protection Systems ..... . .... ....................... 9-1 9.5.1 Description and Evaluation . . . ... ........ ..... 9-1 9.5.1.1 Water Supply Systems ................. ..... 9-1 9.5.2 Other Items Related to Fire Protection .......... .. 9-1 9.5.2.1 Fire Barriers and Penetrations .......... . 9-1 9.5.6 Administrative Controls and Fire Brigade ...... ..... 9-2 9.5.7 Technical Specifications .............. ........ ... 9-2 Clinton SSER 1 vii

i TABLE OF CONTENTS (Continued) i Page

12 RADIATION PROTECTION ............................................ 12-1 12.3 Radiation Protection Design Features ....................... 12-1 ,

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12.3.2 Shielding ........................................... 12-1  !

13 CONDUCT OF OPERATIONS ................. ......................... 13-1 13.3 Emergency Preparedness Evaluation .......................... 13-1 13.3.1 Introduction ........................................ 13-1 13.3.2 Evaluation ............................... . . . . . . . . . . 13-1 13.3.2.1 Assignment of Responsibility 1 (Organizational Control) .................. 13-1 13.3.2.2 Onsite Emergency Organization ............. 13-2 j 13.3.2.3 Emergency Response Support and Resources .. 13-2 13.3.2.4 Emergency Classifications System .......... 13-3 13.3.2.5 Notification Methods and Procedures ....... 13-3 13.3.2.6 Emergency Communications .................. 13-4 13.3.2.7 Public Education and Information .......... 13-4 13.3.2.8 Emergency Facilities and Equipment ........ 13-5 13.3.2.9 Accident Assessment ....................... 13-5 13.3.2.10 Protective Response ....................... 13-6 13.3.2.11 Radiological Exposure Control ............. 13-6 13.3.2.12 Medical and Public Health Support ......... 13-7 )

13.3.2.13 Recovery and Reentry Planning and Postaccident Operations ................... 13-7 13.3.2.14 Exercises and Drills ...................... 13-8 13.3.2.15 Radiological Emergency Response Training .. 13-8 13.3.2.16 Responsibility for the Planning Effort Development, Periodic Review, and Distribution of Emergency Plans ........... 13-8 13.3.3 Conclusions .................. .... ................. 13-9 13.5 Administrative Procedures .................................. 13-9 13.7 Security and Safeguards ............ ... .................. 13-9 1

16 TECHNICAL SPECIFICATIONS . ...................................... 16-1 '

i 18 CONTROL ROOM DESIGN REVIEW ...................................... 18-1 ,

i 18.2 Discussion .... ................................... . . . . . . . 18-1 1 22 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS .......... 22-1 Clinton SSER 1 viii

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TABLE OF CONTENTS (Continued)

APPENDIX A CONTINUATION OF CHRONOLOGY APPENDIX B REFERENCES APPENDIX D NRC STAFF CONTRIBUTORS AND CONSULTANTS.

APPENDIX E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS REPORT ON CLINTON l POWER STATION UNIT 1 l APPENDIX F ERRATA TO CLINTON POWER STATION SAFETY EVALUATION REPORT 4

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Clinton SSER 1 ix i

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1 INTRODUCTION AND GENERAL DESCRIPTION 1.1 Introduction The Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (NUREG-0853) in February 1982 regarding the application by Illinoir. Power Company et al. (hereinatter referred to as the applicant) for a license to operate the Clinton Power Station, Unit 1, Docket No. 50-461. At that time, the staff identi-fied items that were not yet resolved with the applicant. The purpose of this supplement to the Safety Evaluation Report (SER) is to provide the staff evalua-tion of open items that have been resolved, to report on the status of all open items, and to address those recommendations that are contained in the Advisory Committee on Reactor Safeguards (ACRS) letter of March 9, 1982.

At its 263rd meeting on March 5, 1982, the ACRS completed its review of the application. The Committee in its March 9, 1982 letter from Chairman P. Shew-mon to NRC Chairman N. J. Palladino concluded that if due consideration is given to the items mentioned in its letter, and subject to the satisfactory completion of construction, staffing, and preoperational testing, there is reasonable assurance that Clinton Power Station Unit 1 can be operated at power levels up to 2,894 MWt without undue risk to the health and safety of the public.

Each of the following sections or appendices is numbered the same as the corre-sponding SER section or appendix that is being updated. Appendix A is a con-tinuation of the chronology of the staff's actions related to the processing of the Clinton application. Appendix B is a list of references cited in this report.* Appendix D is a list of NRC staff and consultants contributing to this supplement. Appendix E is a copy of the letter from the Advisory Commit-tee on Reactor Safeguards on Clinton Power Station. Appendix F contains the errata to the SER.

Copies of this SER supplement are available for inspection at the NRC Public Document Room at 1717 H Street, NW., Washington, D.C. and at the Warner Ves-pasian Library, Clinton, Illinois. Copies are also available for purchase from the sources indicated on the inside front cover.

The NRC Project Manager assigned to the operating-license application for Clinton is J. H. Williams. Mr. Williams may be contacted by calling (301) 492-9777 or writing to the following address:

Mr. J. H. Williams Division of Licensing, Mail Stop 340 U.S. Nuclear Regulatory Commission Washington, D.C. 20555

• Availabilty of all material cited is described on the inside front cover of this report.

Clinton SSER 1 1-1

1. 9 Outstanding Issues The staff identified certain outstanding issues in the SER that had not been resolved with the applicant. The status of these issues is listed below and discussed further in the sections of this report as indicated. If the staff review is completed for an issue, the item has the notation " closed." The staff will complete its review of these items before the operating license is issued. Resolution of each of these items will be discussed in a supplement to the SER. I Issue Status Section (1) Transportation accidents Under review (2) Effects of Unit 2 excavation Partially closed 2.4, 2.6 (3) Seismic analysis 3.7, 22 (4) Internally generated missiles Closed 3.5 (5) Postulated piping failures Awaiting information (6) S m. ady-state vibration Awaiting information accotance criteria for B0P piping (7) Environmental and seismic Awaiting information qualification test programs (8) Preservice and inservice 5.2.4, 6.6 inspection programs (8a) Preservice and inservice Awaiting information testing of pumps and valves (9) Pool dynamic loads Awaiting information (10a) Containment purge Under review (10b) Containment isolation Under review (10c) Containment leakage testing Awaiting information (vent and drain lines)

(10d) Containment leakage testing Awaiting information (secondary containment)

(10e) Containment bypass leakage Awaiting information (11) Control room habitability Closed 6.4 (12) Engineered safety features Under review reset controls (IE Bul-letin 80-06)

Clinton SSER 1 1-2

Issue Status Section (13) Remote shutdown system Awaiting information ,

(14) Capability for safe shutdown Awaiting information following loss of bus supply-ing power to instruments and controls (IE Bulletin 79-27)

(15) Control system failures Awaiting information resulting from high energy-i line breaks or common power source or sensor malfunctions (16) Separation of the RPS and MSIV Closed 8.4.7 solenoid circuits and PGCC circuits (17) Organization and staffing Awaiting information (18) Emergency Plan 13.3 (19) Security Closed 13.7 (20) QA Program Awaiting information l

1.10 Confirmatory Issues The staff identified confirmatory issues in its SER that required additional information to confirm preliminary conclusions. The status of these issues is listed below and discussed further in the sections of this report as indicated. If the staff review is completed for an issue, the item has the notation " closed."

I Issue Status Section (1) Emergency prepared'1ess 2.3.3, 13.3.2.8 meteorological program i (2) Inspection program around the Closed 2.4.5, 2.6 UHS and the main cooling lake dam I (3) Protection of UHS dam abutments Closed 2.6 against soil erosion (4) Internally generated missiles - Awaiting information fan failures (5) Design adequacy of cable tray Closed 3. 7. 3 ,

system Clinton SSER 1 1-3

Issue Status Section (6) Containment ultimate strength 3.8.1 analysis (7) Structural integrity of safety- 3.8.3 1 related masonry walls (8) NSSS pipe break analysis using Closed 3.6.1 SRP criteria (9) Vibration assessment of RPV Awaiting information internals (10) Annulus pressurization loads Awaiting information (LOCA asymmetric loads)

(11) Use of SRSS for combining Closed 3.9.3 Mark III dynamic responses for other than LOCA and SSE (12) IE Bulletin 79-02 regarding Under review su;oort baseplate flexibility (13) Mark III hydrodynamic loads Awaiting information 1

(14) Feedwater check valve analysis Awaiting information j (15) Seismic and LOCA loadings on Awaiting information fuel assemblies (16) Scram discharge system Closed 4.6 evaluation (17) Fracture toughness data Closed 5.3.1, 5.3.2, 5.3.3 (18) Subcompartment pressure Awaiting information analysis (19) Combustible gas control Awaiting information (20) Containment isolation Under review )

dependability ,

(21) Containment monitoring, Awaiting information ,

II.F.1 j (22) Plant-specific LOCA analysis, Awaiting information II.K.3.31 (23) High drywell pressure Closed in SER interlocks I

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Issue Status Section (24) ATWS recirculation pump trip Awaiting information (25) Response-time testing Closed 7.2.3.2 l (26) Analog trip modules and optical Partially closed 7.2.3.4 isolators 1 (27) Susceptibility of the NSPS to Closed 7.2.3.5 electrical noise i

. (28) Modification of ADS logic, Awaiting information II.K.3.18 (29) Restart of low pressure Closed 7.3.3.5 systems, II.K.3.21 i (30) Temperature effects on level Awaiting information i measurements (31) Containment atmosphere Awaiting information monitoring system (32) Verification that testing is Removed from list in accordance with BTP PSB-1

! (33) Electrical drawing review Removed from list l (34) Verification of diesel Awaiting information generator testing (35) Class A supervision and power Under review supply for fire detection system

(36) Circulating water system Awaiting information i

j -(37) Initial test program Closed in SER (38) Human engineering aspects of 18 l control room design, I.D.1 (39) Con. mon reference for reactor Awaiting information vessel level instruments, II.K.3.27 i

(40) Shielding design review, Closed 12.3.2 l

II.B.2 (41) Short-term accident and Awaiting information .

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procedures review, I.C.1, I.C.7, I.C.8 l

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Clinton SSER 1 1-5

Issue Status Section

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(42) Training during low power Awaiting information '

testing, I.G.1 (43) Review ESF values, II.K.1.5 Closed 6.3 j (44) Operability status, II.K.1.10 Closed 13.5 (45) HPCI and RCIC initiation Awaiting information levels, II.K.3.13 (46) Isolation of HPCI and RCIC, Awaiting information II.K.3.15 (47) Qualification of ADS Under review accumulators, II.K.3.28 (48) Plant-specific analysis, Awaiting information II.K.3.30 l

(49) ODYN analysis for River Bend Closed 4.4.1 as applied to Clinton (50) Conformance evaluation report Under review I for loose parts monitoring system (51) Requirements of NUREG-0313 Closed 5.2.3 (52) Controm room habitability - Closed 6.4 chlorine gas (53) Debris screen design Awaiting inforraation (54) Vertification of adequacy of Removed from list fire protection systems (55) Flood proof door Awaiting information (56) Valves in fire protection Closed 9.5.1.1 water supply system (57) Break in water supply piping Closed 9.5.1.1 (58) Test data on fire ratings 9.5.2.1 (59) Three-hour-fire-rated 9.5.2.1 penetration seals (60) Install fire protection Awaiting information equipment Clinton SSER 1 1-6

! Issue Status Section (61) Fire protection administrative Closed 9.5.6 controls and training (62) Technical Specification on Closed 9.5.7 fire protection i

(63) Periodic leak testing of Closed 3.9.6 pressure isolation values

) (64) Sedimentation in UHS Closed 2.4 (65) Protection against postulated Closed; see out-piping failures standing issue 5 (66) Steam bypass of the Under review suppression pool 4

1.11 License Conditions i There are several issues for which a condition will be included in the oper-ating license to ensure that NRC requirements are met during plant operations.

Nine licensing conditions were listed in the SER. Additional licensing con-ditions are listed below with appropriate references to sections of this report.

Issue Section

.s (10) Measurement of groundwater level 2.4.6 l (11) Security 13.7 4

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2 SITE CHARACTERISTICS

2. 3 Meteorology 2.3.3 Onsite Meteorological Measurements Program The emergency preparedness meteorological program provides measurements of windspeed and direction at the 10-m level of the primary meteorology tower and an estimator of atmospheric stability in accordance with the guidance in Regu-latory Guide 1.97. These measurements are used in an automated dispersion model to calculate relative concentrations within the emergency planning zone.

The results of the dispersion calculations and the meteorological data can be remotely determined from the emergency operations facility and will be avail-able to offsite agencies.

If the onsite meteorological data are unavailable, data from National Wea-ther Service offices at Springfield and Peoria, Illinois, will be used to char-acterize site area conditions. The meteorological measurements program and the application of these data to emergency preparedness dose assessment appear to be reasonable; however, the program's workability cannot be evaluated until an onsite inspection is performed. Therefore, the final acceptability of this program will be determined during the scheduled onsite emergency preparedness implementation appraisal and will be documented in the appraisal report.

2.4 Hydrology 2.4.4 Hydrologic Evaluation of the Ultimate Heat Sink In the Clinton SER, the staff expressed two major concerns regarding sediment inflow into the ultimate heat sink (UHS) during major flood events:

(1) During the probable maximum flood (PMF), the VHS may fill with sediment and not be capable of supplying emergency cooling water if the main dam fails. The main dam is designed to pass the PMF and is currently under an inspection program required by the State of Illinois. However, the staff would require additional inspections for greater assurance of the dam's ability to survive wave runup associated with a PMF if the question of the capacity of the UHS to accommodate the PMF sediment load could not be resolved.

(2) During a combination of the 25 year flood and the safe shutdown earthquake (SSE), the VHS may fill with sediment at about the same time that the main l dam (which is not designed to survive an SSE) fails. To preclude this event, the staff required the applicant to estimate the quantity of sedi-ment that could be carried by the 25 year flood, and to always maintain at least that volume in addition to the volume of water necessary for 30 days' of emergency cooling (110 acre-ft per unit) plus 220 acre-ft of volume for assumed slope failures, into the VHS following the SSE.

Clinton SSER 1 2-1

1 In response to these staff concerns, the applicant submitted an analysis of sedimentation in the UHS during a PMF on the north fork of Salt Creek. In this analysis it was assumed that all of the sediment inflow from the PMF would deposit in the UHS. This resulted in a total estimated deposition of 262 acre-ft. The staff has reviewed the analysis and concludes that it is j conservative. i The UHS has a total capacity of 1,067 acre-ft, and the applicant has committed to a dredging program when the sediment volume in the VHS reaches 500 acre-ft, leaving 567 acre-ft available at all times for sedimentation, emergency cooling water supply, and fire protection. The staff considers this volume adequate for one or two-unit operation of the Clinton Power Station. Therefore, the staff concludes that the additional inspection program for the main dam dis-cussed in the SER is not necessary.

The applicant did not submit an analysis of sediment inflow caused by a 25 year flood on the north fork of Salt Creek. The staff, however, used the applicant's sedimentation analysis and the 100 year flow hydrograph presented in the appli-cant's Final Safety Analysis Report (FSAR) to calculate the sediment inflow from the 100 year flood and determined it to be 35 acre-ft. The volume of sediment from the 25 year flood would be a little less than that.

The staff concludes that the 567-acre-ft capacity, which will always be avail-able because of the applicant's inspection and dredging program, is adequate to contain the sediment volume from the 25 year flood, the volume loss caused by slope failure around the UHS, and the necessary water supply for fire pro-tection and emergency cooling for two units. i i

The staf f concludes that the UHS meets the requirements of General Design Cri-teria (G0C) 2 and 44 with respect to sedimentation.

2.4.5 Erosion Protection for Cooling Water Canals and Reservoirs In the SER the staff concluded that the UHS can meet the requirements of GDC 2 and 44 with respect to erosion induced by flood or dam failure, contingent on staff review and approval of the inspection program to which the applicant had j previously committed.

In FSAR Amendment 14 (Section 2.5.6.8.1, " UHS Monitoring Program"), the appli-cant has submitted a description of the inspection program. This program is to be performed on an annual basis in the fall. Additional inspections will be performed if a major flood, drought, or earthquake occurs. Major floods and droughts are defined as having a recurrence interval of 100 years or more.

A major earthquake is defined by the applicant as one having a horizontal peak ground acceleration of 0.10 g or more at the site.

The UHS shoreline will be observed from both land and by boat for scour and

erosion. Scour and erosion areas shall be defined as those where the surface j vegetation has been disturbed in an area greater than 100 ft2 . When applica- ,

ble, photographic and detailed sketches of any questionable areas shall be '

documented in the inspection report for comparison of conditions from past and j future inspections.

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i The abutments of the UHS submerged dam at or above the water line shall be inspected from a boat for scour and erosion beneath and around the soil cement slope protection of the dam. Photographs and detailed sketches, when made, shall be documented in the inspection report.

The staf f concludes that the applicant's inspection program meets the require-ments of Regulatory Guide 1.127 with respect to scour and erosion.

2.4.6 Groundwater In the SER the staff expressed concern that the design-basis groundwater level of 730 ft mean sea level may be adequate only if the open excavation for Unit 2 is in place because the open excavation tends to dewater the upper portion of the water table and keeps groundwater levels well below those that were observed in the plant area before construction.

In response to this staff concern, the applicant amended the FSAR to state that piezometers will be installed in the vicinity of the power block before either the construction of Unit 2 continues or the backfilling of the excavation of Unit 2 commences. These piezameters will be used to verify the groundwater level used for the main plant hydrostatic loading.

The staff is satisfied that monitoring groundwater levels and verifying the level stated in the FSAR will ensure that the plant will continue to meet GDC 2 with respect to hydrostatic loading on structures. A condition will be placed on the operating license to ensure that these commitments are carried out to the staff's satisfaction before the existing topography is modified significantly.

2.4.8 Conclusions The staff is satisfied with the applicant's responses in regard to sedimenta-tion in the VHS, inspection of the UHS for score and erosion, and design-basis groundwater level.

2.6 Stability of Subsurface Materials and Foundations 2.6.1 Site Conditions 2.6.1.2 Site Investigation In Section 3.7.1 of the SER, the staff stated that the seismic design presented in the FSAR was not acceptable because the deconvolution method was used to establish the design response spectrum at the plant's foundation. Hence, the applicant has developed site-specific response spectra (SSRS) at the plant foundation level (Weston Geophysical Corporation, 1982). In developing the SSRS, the applicant reevaluated the geophysical field test data and used the revisions in the seismic analysis. The following paragraphs describe the results.

Section 2.6.1.2 of the SER presents the geophysical explorations conducted at the project site. Compressional wave velocity of the several soil layers was measured in the geophysical field tests. The shear wave velocity for each of these layers was computed using the analytical relationship between shear and compressional wave velocities using assumed values of the Poisson's ratio.

Clinton SSER 1 2-3

4 During the reevaluation phase, the applicant used data from other soil sites of ,

similar geological origin, soil stratigraphy, and compactness (standard pene- l tration test blow count) where both compressional and shear wave velocities l were measured in field tests (cross-hole survey, up-hole survey). From these l test data, the value of the Poisson's ratio was computed. Using the Poisson's j

i ratio computed from shear and compressional wave velocities measured from field  !

tests at several sites, the applicant estimated the revised value of the Pois-son's ratio for the soils at this site and computed the shear wave velocity using the compressional wave velocity measured at the project site. Revised FSAR Figures 2.5-369 through 2.5-371 present the geological profiles with their geophysical properties. The revisions are changes in the values of the Pois-son's ratio and the corresponding change in the shear wave velocity. Table 2.1 presents the geophysical properties at the station site; this table replaces Table 2.7 in the SER.

The shear modulus parameter used in the dynamic analyses is strain dependent.

The shear modulus-shear strain relationship plot is developed using data from both field and laboratory tests. The shear wave velocity from the field test l is used to compute the shear modulus, and the corresponding strain is in the 10 4% range, or the low-strain end of the plot. The data from laboratory tests are in the 10 3 to 10 2% strain range. All the strain degradation plots were revised to reflect the change in the shear modulus value in the low-strain range as a result of change in the shear wave velocity. FSAR Table 2.5-48

( presents the dynamic properties of the several strata at the project site.

< These revised values have been used in the dynamic analyses. The analyses ,

cover a Nr.ge of shear modulus values (150% of recommended value) to account l for nominal errors and limitations associated with the determination of the ,

! dynamic properties of soil. l 2.6.3 Foundation Stability 2.6.3.3 Settlement

! In Sections 2.6.3.3 and 2.6.3.7 of the SER, the staff stated that the construc-tion of Unit 2 was stopped just below its mat foundation. This has resulted in l

an excavation approximately 250 ft wide by 400 ft long by 25 ft deep adjoining

! the Unit 1 building. The open excavation results in a lack of lateral confine-ment for the foundation on one side of the Unit 1 building and a lack of sur-charge load on the subsoil beneath the foundation.

The staff questioned the effect of this excavation on the assumptions made in the design of Unit 1 and on the additional settlement that would be experienced by Unit 1 as a result of load from Unit 2, if Unit 2 is constructed at a later i j date.

1 In a letter from J. D. Geier of Illinois Power Company to J. R. Miller of NRC,  !

, dated December 3, 1981, the applicant stated that the analysis performed for '

Unit 1 envelops two possible conditions: (1) the building of only Unit 1 or I (2) the building of both Units 1 and 2 as planned. The applicant stated that  !

the absence of Unit 2 was considered in the design and analysis of Unit 1 and l that the lack of lateral confinement was considered in the design. The effect of the lack of surcharge on the subsoil properties was covered in the paramet-ric study performed by the applicant for the Unit 1 building. These factors are discussed below.

I Clinton SSER 1 2-4

The evaluation of the stability of the foundation by a bearing-capacity cri-terion was presented in the SER. The stability of the foundation if sliding should occur, in the absence of an effect from Unit 2 and the contiguous backfill, was evaluated by the applicant. This evaluation showed that the factor of safety against sliding is 1.37 for extreme environmental loading conditions, including an SSE.

The backfill behind the walls of Unit 1 in the vicinity of the excavation is subject to e osion by rain; however, the applicant has committed to protecting this backfill with a revetment composed of a grout intrusion blanket similar to "fabriform." The applicant's commitments for keeping the excavation dry are addressed in Section 2.4.2.2 of the SER.

If Unit 2 were to be constructed at a later date, the additional load imposed on the soil beneath Unit 1 would result in additional settlement of the Unit 1 building. The compacted structural fill and Illinoian till beneath the founda-tion mat would compress under the additional load and contribute to the addi-tional settlement of Unit 1. This additional settlement of Unit 1 is estimated to be a maximum of 0.3 in., which results in a maximum differential settlement of 0.5 in, for the Unit 1 building. The maximum additional settlement to be experienced by the seismic Category I buried piping in the immediate vicinity of Unit 1 is estimated to be 0.5 in. The applicant concludes, and the staff agrees, that these additional settlements would be minimal and would not be harmful to the structure or piping.

The applicant has demonstrated that the effect of Unit 2 excavation on Unit 1 was considered in the Unit 1 design and that the additional settlement experi-enced by Unit 1, if Unit 2 is constructed, would be minimal and would cause no detrimental effects.

The staff has reviewed the results of the applicant's investigations, labora-tory and field tests, analyses, criteria for design and construction, and re-sponses to staff questions. The staff concludes that the foundations of the power block structures (with only Unit 1 built) and the seismic Category I buried piping and electrical duct banks are adequate to support these struc-tures safely and enable them to perform their safety-related functions reli-ably. In addition, they meet the requirements of Appendix A to 10 CFR 100 and GDC 2.

2.6.4 Stability of Slopes 2.6.4.4 Stability Analysis The dynamic stability analyses by the finite-element method (reported in the SER) were repeated using the revised soil parameters. Other than the revisions in the soil parameter, all elements of the analyses (earthquake input motion and method of analyses) are identical to those reported in Sections 2.6.4.4 and 2.6.5.6 of the SER. Table 2.2 presents the results of the dynamic analyses.

l The minimum factor of safety is higher than the acceptable minimum of 1.1 and hence stable under the dynamic loading conditions.

The staff has reviewed the backup data used by the applicant in revising the dynamic properties of the soil and concurs with the revisions in the values of the shear wave velocity and the shear modulus used in the dynamic analyses.

Clinton SSER 1 2-5

I The staff has also reviewed the dynamic analyses of the UHS dam and slopes using the revised soil properties. The staff concludes that the UHS dam and slopes will perform their intended safety-related functions under dynamic loading conditions and that they meet the requirements of 10 CFR 50, Appendix A.

2.6.5 Dams 2.6.5.5 Embankment Ultimate Heat Sink Dam and Baffle Dike A 2-f t-thick layer of a soil-cement mixture is used to protect the slopes of the UHS dam against erosion when the water level in Lake Clinton drops below elevation 675.0 ft. This protective layer is over the entire top surface of the dam and extends, in the abutment area, up to elevation 690.0 ft, which is the operating level of the cooling lake. There has been some erosion of soil where the soil-cement cover meets the natural soils of the UHS shore. This erosion was limited to the vicinity of the junction of soil-cement cover with the natural soil and, even there, it was only at the water level. This indi-cates that the erosion is wave induced. Section 2.6.5.5 of the SER stated the staff's concern over this erosion and the applicant was required to docket remedial measures for this problem. According to a letter from H. Perkins of Illinois Power Company to B. Jagannath of NRC, dated March 9, 1982, the appli-cant has filled the eroded zone with concrete grout and the grout curtain extends a minimum of 1 ft into the abutment. The applicant also committed to ,

inspect the UHS dam, its abutment, and UHS shore area for erosion or local l slope instability and to maintain them in a way to provide sufficient water I for a safe shutdown of the plant. This commitment to monitor and maintain the !

UHS shoreline will be included in the plant Technical Specification. The details of this Technical Specification are under review by NRC.

The staff has reviewed (1) the remedial work performed by the applicant and (2) the commitment by the applicant to monitor and maintain the UHS facility, and concludes that the minor erosion problem has been satisfactorily resolved.

On the basis of these factors and the evaluations presented in the SER, the staff concludes that the UHS dam satisfies the seismic Category I structure criterion and will be stable under both static and dynamic (SSE) conditions.

In addition, it meets the requirements of 10 CFR 50, Appendix A. The appli-cant's investigation, analyses, and monitoring commitments meet the geotech-nical engineering aspects of Regulatory Guide 1.27.

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Clinton SSER 1 2-6

i Table 2.1 Geophysical properties, station site l

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! Compressional Shear. wave velocity, fps wave velocity, Unit description fps Revised Old values Loess, weathered;

overlain by organic top soil 2,000 900 900 Wisconsinan glacial till, I,

weathered to brown 2,000 900 '900 Wisconsinan glacial till,

! unweathered 5,700 1,100 1,900 Interglacial zone, l local alluvial deposits 5,700 1,100 1,900 Illinoian glacial till, local outwash and lacustrine deposits 7,500 2,100 3,600 Lacustrine deposits 7,500 2,100 3,600 i

Pre-Illinoian glacial till, l locally underlain by pre-Illinoian lacustrine deposit 7,500 2,100 3,600 l

Bedrock 10,500 5,700 5,700 Note: Refer to SER Figure 2.7 for thickness and typical stratigraphy at the i station site.

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Table 2.2 Dynamic analysgs of ultimate heat sink (UHS) dam and slopes Minimum factor of safety Distant event, Local event, New Madrid-type Structure SSE event b c UHS dam 1.15 1.81 d

Natural slopes (VHS) 1.27 1.51' See SER Sections 2.6.4 and 2.6.5 for details on the UHS b

dam and slopes.

c See FSAR Table 2.5-56.

d See FSAR Table 2.5-58.

See FSAR Table 2.5-57.

'See FSAR Table 2.5-59.

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s 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS l

3. 5 Missile Protection 3.5.1 Missiles 3.5.1.1 Internally Generated Missiles (Outside Containment)

See Section 3.5.1.2.

3.5.1.2 Internally Generated Missiles (Inside Containment)

In its SER, the staff indicated that the applicant concluded that missiles resulting from failures of pressurized systems would not have sufficient energy to damage or fail safety-related equipment, components, or structures. By FSAR Amendment 13, the applicant has provided information to support this conclusion.

The applicant provided a list of those areas of the plant where high-energy lines in one division, or nondivisional high-energy lines, were located within the same compartment as safety-related equipment of the redundant division. In these areas of the plant, failures of one division or nondivisional equipment could result in missiles that could potentially impact safety-related equipment

, of the redundant division. For these areas, the potential missile sources were reviewed to verify separation from safety-related equipment. Safety-related equipment within 5 ft of potential missiles are protected by at least a 6-in.-

thick concrete barrier. The staff concludes that the applicant's design to maintain the capability for a safe plant shutdown in the event of internally generated missiles is acceptable.

3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping 3.6.1 Determination of Break Locations and Dynamic Effects Associated With the Postulated Rupture of Piping In the SER, the staff identified a confirmatory issue regarding the applicant's method of postulating high-energy pipe breaks in the nuclear steam supply system (NSSS) piping. Standard Review Plan (SRP) Section 3.6.2 (NUREG-0800) states that breaks in ASME Code Section III Class 1 piping not in the contain-ment penetration area be postulated at those locations where the primary and secondary stress intensity range (including the zero load set) as calculated by Equation (10) and either Equation (12) or (13) in Paragraph NB-3653 of ASME Code,Section III exceeds 2.4 S,for normal and upset conditions including the operating-basis earthquake (OBE). In the original evaluation, the applicant postulated a break in the NSSS ASME Code Class 1 piping only when Equation (10) exceeded 3.0 S,and either Equation (12) or (13) exceeded 2.4 S, for normal and upset conditions including the OBE. The applicant committed to reevaluate the NSSS piping using the NUREG-0800 criteria to determine if any additional postu-lated break locations would result.

Clinton SSER 1 3-1

In a letter from G. E. Wuller to J. R. Miller dated April 30, h62, the appli-cant confirmed that in reevaluating the NSSS piping using the d AEG-0800 criteria, no additional breaks were required to be postulated. The staff, therefore, considers the confirmatory issue to be closed.

3.7 Seismic Design I 3.7.1 Seismic Input In the SER, the staff indicated that-the applicant's two-step procedure -in defining the design time history in the free field at the foundation level is not acceptable. It was also indicated that the applicant has undertaken a study to define site-specific spectra to resolve this issue. Since then, the applicant has completed this study and has provided site-specific spectra for the staff's review. The staff is currently reviewing this information and the following evaluation is contingent on the assumption that this information will be acceptable to the staff.

The site-specific horizontal spectra are higher than FSAR design-basis spectra in some frequency ranges. These site-specific spectra are roughly equivalent to a seismic event of 0.20 g anchored to the Regulatory Guide 1.60 spectra at the foundation level in the free field. The applicant has developed a synthetic time-history motion whose spectra envelop the site-specific spectra at all fre-quencies of interest. The vertical site-specific spectra used are also equiva-lent to a seismic event of 0.20 g anchored to the Regulatory Guide 1.60 spectra.

The applicant has evaluated plant-complex structures and the circulating water screenhouse using the synthetic time history. In these analyses, the applicant used the soil-spring approach to account for soil-structure interaction effects i (see Section 3.7.2). The results of the analyses indicate that for plant-complex structures, the increase in the forces in the structural components ranges between 0% and 16% of the design-basis forces, with an average increase of approximately 10%. For the circulating water screenhouse, the new forces are lower than the design-basis forces in the majority of cases. For walls where the soil-spring forces are larger, the increase is less than 10%.

The field material test reports indicate that the actual strength for material used at the Clinton plant is approximately 17% higher than the minimum specified values. This provides an additional 17% load-carrying capacity, fully compen-sating for the increase in forces resulting from the new seismic input. Thus, in general, all stresses are within design-basis allowables when the allowables are based on the actual mean yield strength derived from field test reports.

Based on the above and contingent on the staff's acceptance of the site-specific spectra, this matter is considered fully resolved.

3.7.2 Seismic Subsystem Analysis The staff's position on the soil-structure interaction analysis was described in the SER. This position requires that the soil-structure interaction analysis should include both elastic half-space and finite element approaches for all Category I structures founded on the soil. The applicant has used only the finite element method to perform soil-structure interaction analysis.

Clinton SSER 1 3-2

To comply with the above position of the staff, the applicant conducted the soil-structure interaction analysis using the soil-spring approach and the site-specific seismic input ducussed in Section 3.7.1 of this report. The soil-spring method used by the applicant consists of representation of the e soil media by a visio-elastic layered half space. The soil-spring and dashpot constants are obtained in terms of frequency-dependent impendance functions.

The applicant also has considered appropriate variation in soil properties used in this analysis to account for uncertainties. The staff finds that the approach used by the applicant is acceptable and is similar to those used in the evaluation of some of the recent operating-license applications.

The results of this analysis are disctssed in Section 3.7.1. On the basis of these results, it can be concluded that the applicant has complied with the staff's position on soil-structure interaction analysis. Contingent on the staff's acceptance of site-specific spectra, this matter is considered fully resolved.

The design adequacy of cable trays and support systems for the OBE event was not clearly demonstrated at the time the SER was issued. To demonstrate the design adequacy for the OBE event, the applicant has performed additional analyses and presented the results of these analyses for the staff's review.

The applicant selected the worst-tray system from each building for a detailed dynamic analysis. In selecting these systems, the following factors were con-sidered. (1) the floor with the lowest ratio of peak acceleration of safe shutdown earthquake (SSE) + safety / relief valve (SRV) + loss-of coolant acci-dent (LOCA) spectra to the peak acceleration of SRV + OBE spectra and (2) t'e n tray system with the longest unsupported span. In the dynamic analysis the cable tray was modeled as a lumped mass beam system with the supports modeled by springs. The response-spectrum method was used to compute the dynamic response using the floor response spectrum as the input. The results clearly demonstrate that the cable trays satisfy the staff's acceptance criteria for the OBE loading combinations.

To examine the design adequacy of cable tray supports, the applicant selected 109 hangers for the dynamic analysis. The results indicate that 103 out of 109 hangers comply fully with the staff's acceptance criteria with respect to the OBE design. The other six, although not fully complying with the staff's criteria, possess adequate safety margin against yielding under OBE loading combination; thus, they are judged as meeting the intent of the staff's criteria.

On the basis of the above, the staff concludes that the Clinton cable trays and support systems have adequate design margins for the OBE event and this matter is considered fully resolved.

3.8 Design of Seismic Category I Structures t 3.8.1 Concrete Containment The Clinton Unit 1 containment structure is built of reinforced concrete and consists of a vertical cylinder capped with a hemispherical dome and founded on a base mat. The cylinder has an inside diameter of 124 ft, a height of 150.5 ft, and a thickness of 36 in. The dome and the foundation mat are 2.5 ft and 9 ft 8 in. thick, respectively. The leaktightness of the contain-ment is attained through the use of a 1/4-in.-thick steel liner anchored to Clinton SSER 1 3-3

the inside face of the containment steel and foundation mat. The containment is designed for a pressure of 15 psig.

A study to determine the ultimate capacity of the Clinton Unit 1 containment was made by Sargent & Lundy. The details of the study are contained in Appen-dix A to a report entitled " Degraded-Core Hydrogen Cortrol," dated October 16, 1981. In the study, various probable containment failure modes were examined and two finite element analyses of the containment as an axisymmetric structure were performed, one with the liner considered as a load-resisting element and the other in which the liner strength is neglected. From the results of the analyses, it was found that the most critical section was in the hoop direction around the midheight of the containment where yielding of the hoop steel gov-erns the containment capacity. Using 120 (o = standard deviation) variation of the steel reinforcing and/or steel liner material properties, the upper- and lower-bound ccntainment ultimate capacities are determined as 95 i 7.5 psig with liner as a strength element and 75 1 5.5 psig with liner strength neglected.

An investigation of the ultimate capacity of the penetrations also was made and it was found that the ultimate capacity of the major containment penetrations is controlled by the equipment hatch. The static pressure capacity determined on the basis of the buckling of the equipment hatch spherical head is 76 psig, which was established by using twice the basic allowable buckling stresses.

The 76 psig thus determined is regarded as the critical buckling pressure load for the equipment hatch by the applicant.

However, because there is some uncertainty in establishing this critical buck-ling pressure load as evidenced by the application of a factor of safety in normal design, in the staff's judgment a reduction of this load by a factor of 1.2 should be applied, and the pressure load of 63 psig is determined to be the ultimate capacity of the Clinton containment. In using 63 psig as the ultimate

! capacity, there will be no concern as to the leakage through seals around pene-trations because they have been independently tested to a pressure of 69 psig i as indicated by the applicant. The applicant will confirm that the tested i pressure was 69 psig.

The applicant has evaluated the effects of localized detonations on the con-tainment structure and its penetrations. The applicant contended in a letter dated April 15, 1982 that localized pocketing of hydrogen is minimized by the design of the hydrogen igniter system and, therefore, there is no need to con-sider such effects. The contention is being reviewed by the staff. Until the contention is resolved, this issue remains outstanding.

3.8.3 Other Category I Structures The staff met with the applicant and his consultants on March 8, 1982 to dis-cuss the resolution of the masonry wall issue. In this meeting the applicant i provided the results of analytical studies to address the staff's concerns on I the uncertanties of the parameters used in in the applicant's analysis of the masonry walls. These results indicate that the Clinton masonry walls experi-ence a fairly low level of stress under various applicable loading combinations.

The staff further requested the applicant to assess the impact of using lower damping values and neglecting the contribution of the truss bar reinforcing on the analytical results. The applicant has provided the results of such analysis

! by a letter dated April 23, 1982. Those results indicate that the effect of Clinton SSER 1 3-4 I

neglecting the truss bar reinforcing and using lower damping values lead to higher stresses (increase of 80% in a few cases) when compared with the stresses obtained from the analysis which accounts for truss bar reinforcing and uses higher damping values. However, even with these increases, the stresses are within allowable limits. In addition, considering the fact that these results are based on a conservative parametric variation, the staff believes that the Clinton masonry walls have been shown to be adequate on an analytical basis.

However, because of the uniqueness of this type of unreinforced masonry con-struction, the assumption of uncracked masonry, the traditional uncertainty regarding the overall masonry behavior under intense earthquake shaking, and, particularly, the large scatter associated with the modulus of rupture values for masonry, the staff, in the March 8, 1982 meeting, requested that the appli-cant conduct representative static tests on Clinton-specific masonry construc-tion to verify the analytical assumptions and the parameters used in the analysis.

These tests will consist of either coring or constructing several representative wall samples and subjecting them to the static loads to establish the modulus of rupture and out-of plane wall-bending capacities. The applicant has agreed to conduct such tests and provide the results for the staff's review. In addi-tion, the applicant has provided the staff the scope of the test program and testing procedures to be used for the staff's review by letter dated April 8, 1982. The staff has reviewed this program and finds it acceptable. There will be a minimum of three panels tested to provide an adequate data basis.

On the basis of the results of the above-described wall analysis, the para-metric variation studies, and the fact that the applicant indicated that a quality assurance program consistent with the requirement of Appendix B to 10 CFR 50 was implemented for the design and construction of the walls, the staff concludes that in the event of the design-basis accidents and earthquakes combined with other pertinent loads, there is a reasonable assurance that the walls will maintain their structural integrity without loss of safety functions.

The staff's conclusion is contingent on the following conditions:

(1) The applicant shall provide the results of a confirmatory test program previously approved by the staff to demonstrate the design adequacy and conservatism inherent in the Clinton masonry walls 4 months before the scheduled issuance of the operating license for the staff's review and acceptance.

(2) The applicant shall complete the necessary fixes of the masonry walls before the fuel load date if the review of the information in (1) dictates such fixes.

3.9 Mechanical Systems and Components 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures The applicant has used the square-root-of-the-sum-of-the-squares (SRSS) method for combining all dynamic responses. The use of the SRSS method for combining peak dynamic responses resulting from the loss-of-coolant accident (LOCA) and the safe shutdown earthquake (SSE) has been accepted by the staff in NUREG-0484, Revision 1. Several BWR Mark III owners in response to the Licensing Review Clinton SSER 1 3-5

Group II issue identified as 1-MEB have recently completed a study entitled

" Study To Demonstrate the Generic applicability of SRSS Combination of Dynamic Responses for Mark III Nuclear Steam Supply. System and Balance-of-Plant Piping and Components," dated November 1981, to justify using the SRSS technique for dynamic-response combinations other than the LOCA and the SSE. The report presented 167 cases of response combinations at various locations of mechanical components and equipment selected uniformly from five BWR Mark III plants. In each case, the nonexceedance probability (NEP) of the SRSS method was determined

by creating a cumulative distribution function of the response peak resulting

from combining response-time functions with random-time lags. An independent assessment of all the cases was performed by the Brookhaven National Laboratory (BNL), the NRC consultant. The BNL investigation concludes that the NEP levels at the locations of the mechanical components and equipment are high and meet l the NRC acceptance criteria as delineated in NUREG-0484, Revision 1. The staff

has evaluated the owners' report and the BNL report (NUREG/CR-2686). It concurs with the BNL findings and concludes that the use of the SRSS method for combining the peaks of response-time functions is acceptable for mechanical j components in the Clinton plant.

3.9.6 Inservice Testing of Pumps and Valves In the SER, the staff identified an issue regarding the periodic leak testing i of pressures isolation valves. The staff has completed its review of the

, applicant's response to the issue.

There are several safety systems connected to the reactor coolant pressure boundary that have design pressures below the rated reactor coolant system

, (RCS) pressure. There also are some systems that are rated at full reactor pressure on the discharge side of pumps but that have pump suction below RCS pressure. To protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems. The leaktight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pres-sure of the low pressure systems thus causing an intersystem LOCA.

The pressure isolation valves to be leak tested will be listed in the Tech-

. nical Specifications. The applicant will be required to categorize the

! pressure isolation valves as Category A or AC and to meet the valve leak-rate test requirements of IWV-3420 of Section XI of the ASME Code, except as dis-cussed below. Correct categorization of these valves will be verified when the applicant submits the inservice testing program for valves.

The allowable leakage rate shall not exceed 1 gal per minute (gpm) for each valve as will be stated in the Technical Specifications. The applicant has i

committed to test all pressure isolation valves to the 1 gpm leak rate criteria.

The applicant has proposed to leak test at each periodic test interval, and not each time the valve is disturbed, those systems that are rated at a lower pres-

, sure than the RCS. The staff finds this acceptable because (1) full closure of these valves is verified in the control room by direct-monitoring position indicators and (2) gross intersystem leakages into the low pressure core spray, residual heat removal / low pressure coolant injection, and residual heat removal /

shutdown cooling return and suction lines would be detected by high pressure alarms and increases in the suppression pool level.

i Clinton SSER 1 3-6

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The staff concludes that the applicant's commitments to periodically leak test pressure isolation valves between the RCS and the low pressure systems along with the Technical Specification requirements will provide reasonable assurance that the design pressure of the low pressure systems will not be exceeded thus reducing the probability of an occurrence of an intersystem loss-of-coolant accident. GDC 55 partially considers this matter.

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4 REACTOR 4.4 Thermal and Hydraulic Design 4.4.1 Evaluation In the SER the staff discussed the applicant's use of the ODYN code for River Bend as applied to Clinton Power Station. The ODYN code was used to analyze the following transients: (1) feedwater controller failure - maximum demand, (2) generator load rejection, and (3) turbine trip with and without bypass.

At that time the applicant had not included the report of the analysis in the FSAR. In FSAR Amendment 14, the applicant provided this information. On the basis of its review, the staff considers this confirmatory item closed.

! 4.6 Functional Design of Reactivity Control Systems In the SER the staff indicated that the applicant had committed to modify the scram discharge system to meet the criteria enumerated in the NRC generic study, "BWR Scram Discharge System Safety Evaluation," dated December 1, 1981.

By letter dated December 3,1981, the applicant provided a point-by point description of the scram discharge system against the criteria of the above-referenced NRC generic study.

The design provides two separate scram discharge volumes, with an integral l

instrumented volume at the end of each header, thus providing close hydraulic coupling. Each instrumented volume has redundant and diverse level instrumen-tation (float sensing and pressure sensing) for the scram function attached directly to it. Vent and drain lines are completely separated and contain redundant vent and drain valves equipped with redundant solenoid pilot valves.

High point venting is provided. The staff concludes that the design of the scram discharge volumes fully meets the requirements of the above-referenced NRC generic study and is, therefore, acceptable.

I Clinton SSER 1 4-1

5 REACTOR COOLANT AND CONNECTED SYSTEMS 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.3 Reactor Coolant Pressure Boundary Materials In the SER, the staff indicated that the FSAR would have to be amended to clearly address the requirements of NUREG-0313. The applicant provided this information in FSAR Amendment 10.

The materials used for construction of components of the reactor coolant pres-sure boundary (RCPB) have been identified by specification and found to be in conformance with the requirements of Section III of the ASME Code and in con-formance with the requirements of NUREG-0313, Revision 1. Compliance with the above code provisions for material specifications satisfies the quality stand-ards requirements of GDC 1, GDC 30, and 10 CFR 50.55a.

5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing The July 1981 Edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP, NUREG-0800) includes Sec-tion 5.2.4, " Reactor Coolant Pressure Boundary Inservice Inspection and Test-ing." The Clinton Power Station Unit 1 review is continuing because the applicant has not completed his preservice examinations. The staff's review was conducted in accordance with SRP Section 5.2.4 expect as discussed below.

Paragraph 11.4, " Acceptance Criteria, Inspection Intervals," has not been reviewed because this area applies only to inservice inspection (ISI), not to the preservice inspection (PSI). This subject will be addressed during review of the ISI program after licensing.

Paragraph 11.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed and the applicant has incorporated ASME Code,Section XI, Article IWB-3000, " Standards for Examination Evaluation," into this PSI program. How-ever, ongoing NRC generic activities and research projects indicate that the currently specified ASME Code procedures may not always be capable of detecting the maximum acceptable size flaws specified in the IWB-3000 acceptance stand-ards. For example, ASME Code procedures specified for volumetric examination of reactor vessels, bolts and studs, and piping have not proven to be capable of detecting the maximum acceptable size flaws in all cases.

The staff will continue to evaluate development of improved procedures and will require that these improved procedures be made a part of the inservice examina-tion requirements. The staff has not reviewed the applicant's repair procedures based on ASME Code,Section XI, Article IWB-4000, " Repair Procedures." Repairs are not generally necessary in the PSI program. This subject will be addressed during the staff's review of the ISI program.

Clinton SSER 1 5-1

Paragraph II.8, " Acceptance Criteria, Relief Requests," has not been completed because the applicant has not identified all limitations to examination. Spe-cific areas where ASME Code examination requirements cannot be met will be iden-tified as performance of the PSI progresses.

The staff's complete evaluation of the PSI program will be presented in a later )

supplement to the SER after the applicant submits the required examination in-formation and identifies all plant-specific areas where ASME Code,Section XI, requirements cannot be met and provides supporting technical justification.

GDC 32, " Inspection of Reactor Coolant Pressure Boundary," requires, in part, that components that are part of the reactor coolant pressure boundary be de-signed to permit periodic inspection and testing of important areas and fea-tures to assess their structural and leaktight integrity. To ensure that no deleterious defects develop during service, selected welds and weld heat-affected zones (HAZs) will be inspected periodically at Clinton.

The design of the ASME Code, Class 1 and 2 components of the reactor coolant pressure boundary incorporates provisions for access for inservice inspections as required by Subarticle IWA-1500 of Section XI of the ASME Code. 10 CFR 55.55a(g) defines the detailed requirements for the PSI and ISI programs for light-water-cooled nuclear power facility components. On the basis of the con-struction permit date of February 24, 1976, this section of the regulations requires that a PSI program be developed and implemented using at least the Edition and Addenda of Section XI of the ASME Code that apply to the construc-tion of the particular components. It is the intent of the applicant to comply l with the PSI requirements of the 1974 Edition of the Code with Addenda through ,

Summer 1975. The initial ISI program must comply with the requirements of the I latest Edition and Addenda of Section XI of the ASME Code in effect 12 months before the data of issuance of the operating license, subject to the limita-tions and modifications listed in 10 CFR 50.55a(b).

The staff has reviewed the information in the FSAR and PSI program submitted with a letter dated February 23, 1982. The extent of examination has been determined by the requirements of ASME Code,Section XI, 1974 Edition with Addenda through Summer 1975. The scope of the examinations, procedures, and acceptance criteria meets the requirements of ASME Code,Section XI, 1977 Edition with Addenda through Summer 1979.

At a meeting on April 21, 1982, in Bethesda, Maryland, the applicant presented a detailed technical discussion of the PSI program, the intended piping system examinations, a summary of the automated ultrasonic examination of the reactor pressure vessel, and the nondestructive examination techniques being used. The applicant started the PSI examinations in June 1982. Track systems for the reactor pressure vessel examinations are in place and the vessel examination is scheduled for July or August 1982. A schedule for completing all PSI exami-nations was given as 6 months after start of the examinations. Examination procedures which were identified in the PSI plan submittal as "in course of pre-paration" have been completed and will be submitted to the staff for review.

l The specific areas where the Code requirements cannot be met will be identified after the examinations are performed. The applicant has committed to identify all plant-specific areas where the Code requirements cannot be met and provide supporting technical justification.

Clinton SSER 1 5-2

1 1

On the basis of the staf f's review of the FSAR, the PSI program submitted, and the results of the meeting with the applicant on April 21, 1982, the staff finds the selection of Class 1 and Class 2 welds for examination is accept-able. It considers the review of the PSI program to be a confirmatory issue contingent on the applicant's provisions of the following information:

(1) All procedures identified in the PSI program as "in course of preparation."

(2) All relief requests with supporting technical justification. The staff will complete its evaluation of the Clinton PSI program in a supplement to the SER after the applicant provides an acceptable response.

The initial inservice inspection program has not been submitted by the appli-cant. The staff will evaluate the program after the applicable ASME Code Edition and Addenda can be determined based on 10 CFR 50.55a(b), but before the first refueling outage when inservice inspection commences.

To ensure that no deleterious defects develop during service, selected welds and weld heat-affected zones will be inspected before plant startup and period-ically throughout the life of the plant. The applicant has stated that his inservice inspection program will comply with the rules published in 10 CFR 50.55a.

The design of the reactor coolant system incorporates provisions for access for inservice inspection in accordance with Section XI of the ASME Code (1974 Edition, including Addenda through Summer 1975). Suitable equipment has been developed and will be installed to facilitate the remote inspection of those areas of the reactor coolant pressure boundary that are not readily accessible to inspection personnel. The conduct of periodic inspections and leakage and hydrostatic testing of pressure-retaining components of the reactor coolant pressure boundary in accordance with the requirements of Section XI of the ASME Code and 10 CFR 50 provides reasonable assurance that evidence of structural degradation or loss of leaktight integrity occurring during service will be de-tected in time to permit corrective action before the safety function of a com-ponent is compromised. Compliance with the inservice inspections required by this Code and 10 CFR 50 constitutes an acceptable basis for satisfying in part the requirements of GDC 32.

5.3 Reactor Vessel 5.3.1 Reactor Vessel Materials The staff has reviewed the fracture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials and the materials surveillance pro-gram for the reactor vessel beltline. The acceptance criteria and references that are the basis for the evaluation are in Paragraph II.3a of SRP Section 5.2.3 and Paragraphs 11.5, II.6, and II.7 (Appendices G and H, 10 CFR 50) of SRP Sec-tion 5.3.1 (NUREG-0800). Welding and fabrication aspects if the reactor pres-sure vessel are covered in Section 5.2.3 of this supplement.

5.3.1.1 Evaluation of Compliance With Appendix G, 10 CFR 50 In the SER it was found that the requirements of Appendix G, 10 CFR 50, were met for Unit 1 except for the specific requirements of Paragraphs III.A, Clinton SSER 1 5-3

i i

l III.R.4, III.B.5, and IV.A.3. In addition, in FSAR Amendment 13 (February j 1982), the applicant requested an exemption to Paragraph IV.A.2.C of Ap-pendix G. The staff's review of each of these items follows.

Paragraph III.A of Appendix G states that to demonstrate minimum fracture tough-ness requirements, ferritic materials of the reactor coolant pressure boundary I (RCPB) must be tested in accordance with NB-2300 of the ASME Code. The mate-l rials must be subject to the Charpy V notch (CVN) impact test and, when required, i the drop-weight test. The applicant has supplied CVN impact test and drop-weight

test data which indicate that the ferritic materials of the RCPB were tested in accordance with NB-2300 of the ASME Code. Therefore, the staff concludes that

, no exemption to the requirements of Paragraph III.A is required.

I j Paragraph III.B.4 of Appendix G requires that the testing personnel shall be qualified by training and experience and shall be able to perform the tests in

) accordance with written procedures.

i The applicant states that the testing personnel were qualified to written pro-i cedures, on-the-job training, and past experience, but that there are insuffi-I J

cient records to document full compliance.

Because these tests are relatively routine in nature and are continually being I

performed in the laboratory that conducted these tests, it is unlikely that the i tests were conducted improperly. Consequently, the staff cnncludes that an exemption for not performing the tests in accordance with written procedures is justified.

i i f

Paragraph III.B.5 of Appendix G states that fracture toughness test results i shall be recorded and shall include a certification by the licensee or persons performing the test for the licensee that (1) The tests have been performed in compliance with the requirements of Appendix G.

1 (2) The test data are correctly reported and identified with the material i intended for a pressure-retaining component.

l (3) The tests have been conducted using machines and instrumentation with available records of periodic calibration.

(4) Records of the qualifications of the individuals performing tests are l available upon request.

In Amendment 13 to the FSAR, the applicant indicated that he complies with all these requirements. The justification was a General Electric letter from Sherwood to Case (NRC) dated October 17, 1977, indicating that all the requirements of Paragraph III.B.5 have been met. The staff has reviewed this letter and has concluded that all the requirements of Paragraph III.B.5 have been met and that no exemption is necessary.

j Paragraph IV.A.2.c of Appendix G requires that whenever the core is critical, the metal temperature of the reactor vessel shall be high enough to provide an adequate margin against fracture, taking into account such factors as the poten-tial for overstress and thermal shock during anticipated operational occurrences j

i Clinton SSER 1 5-4 l

l l

in the control of reactivity. -In no case when the core is critical (other than for the purpose of low-level physics tests) shall the temperature of the reactor vessel be less than the minimun permissible temperature for the in-service system hydrostatic pressure test or less than 40 F above the tempera-ture required by Paragraph IV.A.2.a. The applicant requested that he be allowed to employ procedures other than the limits established by the inservice system hydrostatic pressure test. In FSAR Amendment 13 the applicant indicated that the intent of the proposed alternative method of compliance with Appendix G of 10 CFR 50 is to use operating limitations on pressure and temperature that provide a margin of safety against a nonductile failure that will be equivalent to that that for a vessel built to Summer 1972 Addenda of the ASME Code. General Electric also has proposed that 10 CFR 50 be amended to make this procedure acceptable in the regulation. The proposed modification to 10 CFR 50, Appen-dix G, Paragraph IV.A.2.c, is described in the General Electric Topical Report NEDO-21778-A, " Transient Pressure Rises Affecting Fracture Toughness Require-ments for Boiling Water Reactors." As reported in the November 13, 1978 letter from O. D. Parr to G. G. Sherwood, the staff has reviewed this topical report, found it acceptable, and concurred that the proposed alternative to the criti-cality hydrostatic temperature limit is acceptable. The evaluation of accept-ance of NEDO-21778-A provides sufficient information for the staff to conclude that the proposed alternate method is adequate and that an exemption of Para-graph IV.A.2.c of Appendix G is justified.

Paragraph IV.A.3 of Appendix G requires that materials for piping, pumps, and valves meet the requirements of Paragraph NB-2332 of the ASME Code. The appli-cant has certified that the above components have met the requirements in Para-graph NB-2332 of the ASME Code. Therefore, an exemption to Paragraph IV.A.3 of Appendix G is not required.

5.3.1.2 Evaluation of Compliance With Appendix H, 10 CFR 50 In the SER it was found that the requirements of Appendix H were met for Unit 1, except for those of Paragraph II.B. The staff's review of this item follows.

Paragraph II.B of Appendix H requires, in part, that the surveillance program for the ferritic materials in the reactor vessel beltline comply with American Society for Testing Materials (ASTM) E 185-73, " Standard Recommended Practices for Surveillance Tests for Nuclear Reactor Vessel." ASTM E 185-73 requires that the lead factors of the surveillance capsule specimens be known so that capsule withdrawal times can be calculated. In the SER the staff noted that the applicant has not reported the lead factors for the material surveillance capsules. In FSAR Amendment 13 dated February 1982, the applicant has reported the lead factors. The staff has evaluated them and finds them to be in accord-ance with ASTM E 185-73.

The staff's evaluation has shown that the applicant has submitted sufficient data to fulfill the requirements of Paragraphs III.A, III.B.5, and IV.A.3 of Appendix G, and Paragraph P.B of Appendix H. The staff's evaluation has not identified any practical methods by which the existing Clinton Unit 1 RCPB components can comply with the specific requirements of Paragraphs III.B.4 and IV.A.2.c of Appendix G, 10 CFR 50. However, the alternate methods proposed by the applicant were reviewed and evaluated and have been found to demonstrate that the intent of Appendix G has been achieved.

Clinton SSER 1 5-5

Based on the fr going, pursuant to 10 CFR 50.12, exemptions from the specific requirements < nppendices G and H of 10 CFR 50, as discussed above, are author-ized by law anu can be granted to Clinton Unit 1 without endangering life or property or the cc.nmon defense and security and are otherwise in the public interest. The staff concludes that the public is served by not imposing cer-tain prov ans of Appendices G and H, 10 CFR 50 that have been determined to  !

either be impractical or result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.

5.3.3 Reactor Vessel Integrity The staff has reviewed the fracture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, the pressure temperature limits for operation of the reactor vessels, and the materials surveillance program for the reactor vessel beltline. The acceptance criteria and references, which are the basis for this evaluation, are in Paragraphs 11.2, II.6, and II.7 (Appendices G and H, 10 CFR 50) of SRP Section 5.3.3 (NUREG-0800).

The staff has reviewed the above criteria as they relate to the structural integrity of the reactor vessels and concludes that the Clinton Unit 1, the applicant has complied with Appendices G and H, 10 CFR 50, except for the following items.

Paragraph III.B.4 of Appendix G requires the applicant to conduct impact testing according to written procedures. Although the applicant cannot provide docu-centation to demonstrate that these tests were performed to formal written pro-  ;

cedures for Clinton Unit 1 impact tests, the applicant has supplied sufficient I information to demonstrate that the tests were conducted correctly, and,there-fore, the staff has concluded that an exemption to Paragraph III.B.4 is justified.

Paragraph IV.A.2.c of Appendix G requires, in part, that whenever the core is critical, the temperature of the reactor vessel shall be greater than the mini-mum permissible temperature for the inservice system hydrostatic pressure test.

The applicant has provided in Topical Report NED0-21778-A sufficient information for the staff to conclude that the proposed alternate method is adequate and that an exemption to Paragraph IV.A.2.c of Appendix G is justified.

The staff has reviewed all factors contributing to the structural integrity of the reactor vessel and concludes there are no special considerations making it necessary to consider potential reactor vessel failure for Clinton Unit 1.

Clinton SSER 1 5-6

f

( 6 ENGINEERED SAFETY FEATURES

! 6.3 Emergency Core Cooling System The applicant was asked to review all valve positions, positioning requirements, positive controls, and related test and maintenance procedures to ensure the proper functioning of the engineered safety features (ESF) (II.K.1.5). In the SER, the staff reported that the applicant will develop written procedures which include provisions to ensure proper valve alignments for ESF functioning on re-turning systems to service. The applicant has indicated that these procedures and reviews have been completed. The staff will inspect the records and proce-dures on site and will confirm that the requirements of NUREG-0737 have been met before the operating license is issued.

6.4 Control Room Habitability The requirements for the protection of the control room personnel under accident conditions are specified in GDC 19. The applicant has proposed to meet these requirements by incorporating shielding and emergency ventilation systems in the control room design and by having an adequate supply of self-contained breathing apparatus in the control room for the emergency team. The applicant has stated in FSAR Section 6.4 that the control room ventilation system is designed to meet

[

l single-failure and seismic Category I criteria.

i The operation of the emergency ventilation system can be controlled by redundant radiation monitors and local chlorine detectors, located in the supply air sys-tem ducts. The emergency pressurization units utilize 99% efficient, engineered safety feature grade atmospheric cleanup units. In the event of high-radiation detection, an alarm in the control room will be activated and the heating, ven-tilating, and air conditioning (HVAC) system will switch to the emergency mode of operation. The control room is equipped with dual intakes and the operator can operate handswitches to choose the intake with the lower radioactivity. On detection of chlorine, the outside air intake dampers will automatically close and the HVAC system will operate with 100% recirculation.

In addition to the primary ventilation system, the control room HVAC recircu-lation system would be in operation during radiological emergencies to remove radioactive iodine from the control room atmosphere. This system is not in strict conformance with Regulatory Guide 1.52; however, an iodine decontamina-tion efficiency of 70% (independent of chemical form) is appropriate provided that (1) the filter trains will be leak tested and (2) the iodine removal effi-ciency of the activated charcoal will be determined by laboratory tests in accordance with Sections 5 and 6 of Regulatory Guide 1.52, Revision 2. These requirements will be incorporated into the Technical Specifications.

The staff has evaluated the control room doses following a postulated loss-of-coolant accident in accordance with SRP Section 6.4 (NUREG-0800). The calcu-lated thyroid and whole-body doses are within the guidelines of GOC 19.

The staff has also evaluated the habitability of the control room with respect to toxic gases in accordance w:th SRP Section 6.4 and Regulatory Guides 1.78 Clinton SSER 1 6-1

l and 1.95. As indicated in Section 2.2 of the FSAR 32 tons of chlorine in two

tank cars are proposed to be stored on site within 300 m of the control room. I j

In view of this potential chlorine hazard, the staff will require a Technical 1 Specification to the operating license which will recuire periodic testing to i

limit the control room leak rate to (4,000 cfm at > 1/4-in. water gauge. This j Technical Specification value is based on an analysis provided by the applicant in January 1982 (Amendment 12). The applicant has demonstrated in the analysis I that the control room ventilating system will adequately protect the control room operators against chlorine release in accordance with SRP Section 6.4 and i Regulatory Guides 1. 78 and 1.95, and the staf f's review confirms the applicant's finding.

l In meeting the guidance and position of the Regulatory Guides and the SRP, the applicant has demonstrated that the control room ventilating system will ade-quately protect the control room operators and will remain habitable in accord- a

ance with NUREG-0737, Item III.D.3.4.

6.6 Inservice Inspection of Class 2 and 3 Components 1 The July 1981 Edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP, NUREG-0800) includes Sec-tion 6.6, " Inservice Inspection of Class 2 and 3 Components." The Clinton Power Station Unit 1 review is continuing because the applicant has not com-pleted his preservice examinations. The staff's review was conducted in accordance with SRP Section 6.6 except as discussed below.

Paragraph II.4, " Acceptance Criteria, Inspection Intervals," has not been reviewed because this area applies only to inservice inspection (ISI), not to preservice inspection (PSI). This subject will be addressed during review of l the ISI program after licensing.

Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed and the applicant has incorporated ASME Code, Articles IWC-3000

) and IWD-3000, " Standards for Examination Evaluation," into his PSI program.

f However, ongoing NRC generic activities and research projects indicate that the currently specified ASME Code procedures may not always be capable of detecting the maximum acceptable size flaws specified in these standards. For example, ASME Code procedures specified for volumetric examination of vessels, bolts and studs, and piping have not proven to be capable of detecting maximum acceptable size flaws in all cases.

The staff will continue to evaluate development of improved procedures and will require that these improved procedures be made a part of the inservice examina-tion requirements. The staff has not reviewed the applicant's repair procedures based on ASME Code, Articles IWC-4000 and IWD-4000, " Repair Procedures." Repairs are not generally necessary in the PSI program. This subject will be addressed during the review of the ISI program.

Paragraph II.7, " Acceptance Criteria, Augmented ISI To Protect Against Postu-lated Piping Failures," has not been completed because this subject has not yet been addressed in the applicant's PSI program. The staff will review the I

applicant's augmented ISI program af ter it is submitted.

Paragraph II.9, " Acceptance Criteria, Relief Requests," has not been completed because the applicant has not identified the limitations to examination.

Clinton SSER 1 6-2

l Specific areas where ASME Code examination requirements cannot be met will be identified as performance of the PSI progresses. The complete evaluation of the PSI program will be presented in a future supplement to the SER after the applicant submits the required examination information and identifies all plant-specific areas where ASME Code,Section XI, requirements cannot be met and pro-vides a supporting technical justification for relief.

l GDC 36, 39, 42, and 45 require, in pai that the Class 2 and 3 components be designed.to permit appropriate periodic inspection of important components to ensure system integrity and capability 10 CFR 50.55a(g) defines the detailed requirements for the PSI programs for light-water-cooled nuclear power facility components. Based on the construction permit date of February 24, 1976, thic section of the regulations requires that a PSI program for Class 2 and 3 compo-nents be developed and implemented using at least the Edition and Addenda of Section XI of the ASME Code applied to the construction of the particular com-ponents. It is the intent of the applicant to comply with th PSI requirements of the 1974 Edition of the Code with Addenda through Summer 1975. The initial l ISI program must comply with the requirements of the latest Edition and Addenda of Section XI of the ASME Code in effect 12 months before the date of issuance of the operating license, subject to the limitations and modifications listed in 10 CFR 50.55a(b).

The staff has reviewed the information in the FSAR and the PSI program submitted with a letter dated February 23, 1982. The extent of examination has been deter-mined by the requirements of ASME Code,Section XI, 1974 Edition with Addenda through Summer 1975. The scope of the examinations, procedures, and acceptance

! criteria meets the requirements of ASME Ccde Section XI, 1977 Edition with l Addenda through Summer 1979.

l At a meeting on April 21, 1982, in Bethesda, Maryland, the applicant presented a detailed technical discussion of the PSI program, the intended piping system examinations, and the nondestructive examination techniques being used. The baseline examinations on all Class 2 loop multiple stream systems will not have specific streams exempted from PSI. At the April meeting the applicant agreed to evaluate an augmented program to substitute volumetric examination of selected Class 2 piping welds with less than 1/2-in. wall thickness for the Code-required surface examinations and report the results to the staff. The applicant started the PSI examinations in June 1982. A schedule for completing all PSI examinations was given as 6 months after start of the examinations.

Examination procedures that were identified in the PSI plan submittal as "in course of preparation" have been completed and will be submitted to the staff for review. The specific areas where the Code requirements cannot be met will be identified after the examinations are performed. The applicant has committed to identify all plant-specific areas where the Code requirements cannot be met and provide supporting technical justification.

Based on a review of the FSAR, the PSI program submitted, and the results of the meeting with the applicant on April 21, 1982, the staff finds the selection of Class 2 and 3 welds for examination is acceptable. The staff considers the review of the PSI program to be a confirmatory issue contingent on the applicant providing the following information:

(1) all procedures identified in the PSI program as "in the course of preparation" Clinton SSER 1 6-3

Y (2) an evaluation of the impact of doing augmented volumetric examinations of i Class 2 piping welds with less than 1/2-in. wall thickness (3) all relief requests with supporting technical justification The staff will complete its evaluation of the Clinton Unit 1 PSI program in a future supplement to the SER after the applicant provides an acceptable response.

The applicant has not submitted the initial inservice inspection program.

The staff will evaluate the program af ter the applicable ASME Code Edition and Addenda can be determined based on 10 CFR 50.55a(b), but before the first refueling outage when inservice inspection commences.

To ensure that no deleterious defects develop during service in ASME Code Class 2 system components, selected welds and weld-heat-affected zones are inspected before plant startup and periodically throughout the life of the plant. In addition, ASME Code Class 2 and 3 systems receive visual inspec-tions while the systems are pressurized, in order to detect leakage, signs of mechanical or structural distress, and corrosion.

The applicant has stated that his ISI program will comply with the rules pub-lished in 10 CFR 50.55a, and the applicable editions of ASME Code,Section XI.

Compliance with the preservice and inservice inspections required by the ASME Code and 10 CFR 50 constitutes an acceptable basis for satisfying the appli-cable requirements of GDC 36, 39, 42, and 45.

Clinton SSER 1 6-4

(

7 INSTRUMENTATION AND CONTROLS l

7.2 Reactor Protection System (RPS) 7.2.3 Resolution of Issues 1

7.2.3.2 Response Time Testing As stated in the SER, the applicant did not originally plan to perform response time testing for the solid-state safety system [ nuclear system protection sys-tem (NSPS)] logic. The staff informed the applicant that response time testing of protection system channels described in IEEE Std 338-1977 as supplemented by Regulatory Guide 1.118 would be required, and requested him to formally docu-ment his commitment to response time testing of the solid-state NSPS logic.

! By letter dated December 1, 1981 (U-357), the applicant stated that response time testing of the solid state logic of the NSPS (from the initiating parameter to the actuated device) will be included in the overall response time testing as required by the Clinton Unit 1 Technical Specifications. The Technical Specifi-cation information to be provided will include the channels and systems to be tested, the frequency of the testing, and the required response times.

On the basis of this commitment from the applicant, the staff concludes that this item is satisfactorily resolved contingent on acceptance of the Clinton Unit 1 Technical Specifications.

! 7.2.3.4 Analog Trip Modules (ATMs) and Optical Isolators (0Is)

As stated in the SER, ATMs are used extensively as a bistable trip unit in the NSPS. Each ATM compares a conditioned analog signal, proportional to a moni-i tored process variable, with a trip setpoint and provides a change of state of the output if the setpoint is exceeded. The applicant was required to provide a detailed design description of the ATMs.

l The applicant provided a description of the ATMs in FSAR Amendments 11 and 12.

The ATMs are part of the NSPS design concept and are designed to meet IEEE Stds 323-1974 and 344-1975 and the protection system testing requirements of IEEE Std 338-1977 as supplemented by Regulatory Guides 1.22 and 1.118. On the basis of its review of the ATM design description provided by the applicant, j the staff finds this design acceptable.

l The applicant has not, however, provided required confirmation that the Ols l used at Clinton are identical to those approved for Grand Gulf. This remains l a confirmatory item pending submittal of the required confirmation.

7.2.3.5 NSPS Susceptibility To Electrical Noise As stated in the SER, the staff was concerned that the low-voltage complemen-l tary metal oxide semiconductor (CMOS) devices used in the NSPS could be sensi-tive to electrical noise and other types of interference. The applicant was Clinton SSER 1 7-1

l required to provide information describing the shielding techniques or other l precautions taken to prevent system maloperation from such effects.

The applicant provided this information in FSAR Amendment 12, which describes the tolerance of the CMOS logic to noise spikes, the isolation and buffering of all inputs and outputs to the NSPS cabinets, the routing of internal NSPS wiring, and the use of switching power supplies and filtering to prevent con-

',I duction of interference through the power lines. On the basis of its review of the applicant's submittal, the staff concludes that the NSPS logic is ade- l

! quately protected from the effects of electrical noise and other types of 1 interference and, therefore, is acceptable.

7.3 Engineered Safety Features System 7.3.3 Resolutica of Issues 7.3.3.5 Restart of Core Spray and Low-Pressure Coolant Injection Systems (TMI Action Plan Item II.K.3.21) i

. TMI Action Plan Item II.K.3.21 states that the core spray and low pressure coolant injection (LPCI) system logic sPould be modified so that these systems I

will restart, if required, to ensure adequate core cooling. As stated in the SER, the applicant has taken the BWR Owners Group position that the current

low pressure emergency core cooling system (ECCS) logic design (low pressure core spray system and LPCI) is adequate and that overall boiling-water-reactor (BWR) safety would not be enhanced by modifying these systems to provide automatic restart capability. No modifications were planned for these systems.

This position is acceptable to the staff.

i A modification was planned for the high pressure core spray (HPCS) system, how-j ever, to provide automatic reset of the autoinitiation signal for low water i level and block the continuing autoinitiation signal for high drywell pressure to allow autorestart of HPCS on low water level after the operator stopped the

, HPCS pump. Decrease in drywell pressure below the trip level returns the HPCS logic to its original status.

However, the current NRC position, identified in a letter from J. R. Miller to D. L. Holtzscher, dated February 26, 1982, is that the automatic restart of HPCS after manual termination is optional and not necessarily required. The justi-fication for not modifying the HPCS logic follows.

Immediately following a loss-of-coolant accident that produces either high drywell pressure or low reactor water level, the HPCS automatically starts.

Injection of emergency cooling water into the reactor occurs. Flow from the HPCS system is automatically terminated when the reactor water level reaches '

its high level trip point (level 8). This control feature prevents flooding of the reactor vessel and steamlines. Termination of HPCS injection can occur either automatically or by operator action. In the event of the former, the HPCS system will restart automatically if and when reactor water level decreases from the high level trip point to the low level initiation setpoint. For the i latter event, a manual action is required to restart HPCS. It was the staf f's l concern about reliance on the operator to restart the HPCS after manual termina-j tion that prompted the original proposed design modification. Such a modifica-l tion is not necessary for the following reasons:

i Clinton SSER 1 7-2

l (1) The ECCS logic design that permits operator intervention is based on a legitimate assumption that the operators are not likely to terminate ECCS flow prematurely and thereby jeopardize the core cooling process. In actual practice, one of the activities with highest priority for an operator during an accident is to ensure that emergency systems have started correctly and r are effectively maintaining core coverage. This guidance is provided to the operator through the plant's emergency operating procedures.

If the, operator should terminate HPCS system flow, such termination would be based on event-specific conditions such as (a) availability of adequate coolant flow from other systems (feedwater and reactor core isolation cooling (RCIC))

(b) HPCS system equipment problems (gross seal leakage, pipe breaks, and equipment flooding)

(c) required vessel coolant makeup rate much less than HPCS system capa-bility (5,010 gpm) and well within RCIC system capability (600 gpm)

(2) For the long-term core cooling situation, the plant operators manually set up the auxiliary systems to support eventual termination of the incident.

Consequently, adequate core cooling depends on correct operator actions.

Such actions are not constrained by strict time requirements. This aspect of ECCS design is considered fully acceptable because of the time available between attainment of level 1 and the occurrence of high fuel clad temperatures.

(3) A key incentive of vessel water level control is to keep the core covered but also to prevent water level from reaching level 8, where, in addition to HPCS, the RCIC and feedwater systems (if operating) would be tripped off.

(4) Automatic vessel water level control will be available from the RCIC system.

This system will be capable of automatic restart on level 2 after automatic termination at level 8, as provided for in response to TMI Action Plan Item II.K.3.13.

(5) Inadequate core cooling as a result of operator failure to reinstate the HPCS system would not occur because the automatic depressurization system initiation level would be reached eventually. This would result in reactor blowdown and core flooding by the low pressure ECCS.

The manual override option is deliberate and is considered to be an impor-tant safety feature of the BWR ECCS network. This feature provides the plant operators with flexibility in dealing with unforeseen but credible conditions requiring the shutdown of a particular system. This option, complemented by the other means available to automatically maintain ade-quate core cooling, provides adequate justification for not implementing the HPCS system automatic restart after manual termination modification.

Based on the above, the applicant has decided not to modify the HPCS system as stated in the SER. The staff has concluded that this is acceptable and that this issue is resolved.

Clinton SSER 1 7-3

8 ELECTRIC POWER SYSTEM 8.4 Other Electrical Features and Requirements for Safety 8.4./ Physical Identification and Independence of Redundant Safety-Related Electrical Systems As stated in the SER, Section 8.4.7, the Clinton power generation control com-plex (PGCC) design allows non-Class 1E fire detection, intercom, and utility.

services wiring in flexible conduit to be routed through and between Class IE divisional ducts. The wiring from the reactor protection system (RPS) inver-ters to the RPS and main steam isolation valve (MSIV) solenoid circuits is also run in conduit and routed through Class 1E ducts and raceways. This does not satisfy the physical separation requirements identified in Institute of Elec-trical Electronics Engineers (IEEE) Std 384-1974 and Regulatory Guide 1.75;

, therefore, in accordance with IEEE Std 384, an analysis based on tests must be provided to verify that the Class 1E wiring will not be compromised. Tests have been performed by General Electric Company (GE) to demonstrate that a failure of the wiring in the flexible conduit would not fail any wiring exter-nal to it. However, neither the test nor the analysis included the effects of a line-to ground (conduit) fault or accounted for the temperature rise during

faulted conditions of a flexible conduit that is buried in cables. In discus-l sions with GE the staff indicated that additional tests and/or analysis cover-ing these conditions would have to be provided. The staff also indicated that

, the tests and analysis could include the use of protective devices to interrupt the fault provided the devices were of proven quality and testable and met the single-failure criterion (requires two protective devices in series).

GE presented the additional analysis (Smith, 1982) in a meeting with the staff on February 9, 1982. The analysis showed that given a line-to ground fault of a utility cable internal to the conduit with the ground return path through the conduit, the heat rise on the outer surface of the flexible conduit resulting from the current flow would be insufficient to fail any cable in contact with it before a fuse or circuit breaker interrupted the fault. It is noted that GE did not provide an analysis for the heat rise at the point where the cable was shorted to the conduit. The temperature there could be greater than that previously analyzed; however, it would be a localized heat rise that would affect only the cables in one division. Therefore, it is not necessary that this analysis be provided.

GE has also verified that faults in the power wiring adjacent to the RPS and MSIV cables in the PGCC do not pose a threat to the RPS and MSIV cables in the flexible conduit. GE further indicated they would provide redundant fault pro-tection for those 120-V ac fire protection circuits and 120-V ac communication circuits which run in PGCC Class 1E ducts. The low-voltage, low energy por-tions of those circuits do not require any additional protection.

In the February 9th meeting, Sargent & Lundy acting for Illinois Power Company indicated that the applicant would run the RPS and MSIV circuits located in the plant through a flexible conduit and route them in Class 1E trays. Before they Clinton SSER 1 8-1

i enter-the. trays, the circuits will be double fused and all RPS circuits lead-ing to the scram solenoid valves of one group of hydraulic control units (HCUs), i will only be run through Class 1E trays of a single division. Sargent & Lundy j indicated they would not ground the flexible conduit in the trays as GE has done in the PGCC. This would not provide the same level of ground fault pro-tection as the circuits in the PGCC, but is acceptable, since any failure of an RPS or MSIV circuit internal to the conduit or Class lE circuit external to the  !

conduit would only affect one group of HCUs or one Class lE division.

In summary, the staf f considers the use of the flexible conduit in the PGCC and in the RPS and MSIV circuits to be acceptable provided redundant protective devices are used in the 120-V circuits which have at least the equivalent fault current-interrupting characteristics as those used in the GE analysis. The devices should be of proven quality, and a representative sample of at least 10% of the devices should be tested at least every 18 months in a manner simi-lar to that specified for penetration circuit breakers and fuses in the Stand-ard Technical Specifications.

The staff, therefore, requires that the above-described testing be incorporated into the Clinton Technical Specifications. With this Technical Specification '

stipulation for Clinton, the staff considers this item resolved.

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l 9 AUXILIARY SYSTEMS 9.5 Fire Protection Systems 9.5.1 Description and Evaluation 9.5.1.1 Water Supply Systems Iri the SER, the staf f reported that valves in the fire protection water supply system are electrically supervised with alarms in'the control room or monitored by administrative procedures. The applicant verbally committed.to key-lock open, with strict key control procedures and monthly verification of valve posi-tion, all valves that are not electrically supervised. This procedure was to be documented in an FSAR amendment. By letter dated November 19, 1981 and FSAR Amendment 14, the applicant provided the required documentation. Therefore, the supervision of valves meets the guidelines of Appendix A, Section E-3.6 of Branch Technical Position (BTP) ASB 9.5-1 and is acceptable.

The staff reported in the SER that the applicant must document that a single break in the water supply piping will not eliminate both the primary and secon-

dary water suppression in any fire zone. The applicant was to provide this i documentation in an FSAR Amendment. By letter dated November 19, 1981 and FSAR Amendment 14, the applicant provided the required documentation. This is acceptable to the staff.

9.5.2 Other Items Related to Fire Protection 9.5.2.1 Fire Barriers and Penetrations Walls that separate buildings and wall and floor / ceiling assemblies used to enclose rooms containing safety-related equipment must be qualified to a 3-hour-fire rating.

The applicant originally had extrapolated test data in determining that the fire ratings of walls and floor / ceiling assemblies were qualified.

In the SER, the staff reported that the applicant had verbally indicated that all fire-rated wall and floor / ceiling assemblies had been qualified in accord-ance with American Society for Testing Materials (ASTM) E-119, " Fire Test of Building Construction and Materials," for a minimum of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The applicant had agreed to provide test data in the FSAR to substantiate such fire ratings.

The applicant also had verbally committed to provide 3-hour-fire-rated penetra-tion seals for conduit, cable tray, or piping penetrations qualified for a 3-hour-fire rating in accordance with ASTM E-119. Test data to substantiate this committment were to be included in an FSAR amendment.

By letter dated November 19, 1981 and FSAR Amendment 14, the applicant provided information on test data. However, the submitted information was not sufficient Clinton SSER 1 9-1

1 to confirm the verbal commitments to verify the design of 3-hour-fire-rated bar-riers and penetration seals. The staff will require the applicant to provide design information, such as Underwriters Laboratory (UL) or Factory Mutual (FM) fire resistance rating design numbers, test report numbers, or actual fire test data to verify that the fire barriers and penetration seals are rated for a fire resistance of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in accordance with the guidelines of BTP CMEB 9.5-1, Sec- '

! tion C.S.a. The staff will report on these issues in a future supplement to  !

the SER.

I 9.5.6 Administrative Controls and Fire Brigade

In the SER, the staff reported that the applicant had verbally committed to implement the fire protection program contained in the staf f supplement guid-ance, " Nuclear Plant Fire Protection Functional Responsibilities, Administra-tive Controls, and Quality Assurance," dated July 1977. This program includes (1) fire brigade training, (2) control of combustibles, (3) control of igni-tion sources, (4) fire fighting procedures, and (5) quality assurance.

The applicant was to document this committment in an FSAR amendment. By letter dated November 19, 1981 and FSAR Amendment 14, the applicant provided the l required documentation. The applicant will implement the plant administrative controls and procedures before fuel loading. This is acceptable to the staff.

1 9.5.7 Technical Specifications In the SER, the staff reported that the applicant had verbally committed to follow the staff's Standard Technical Specifications with respect to fire pro-tection. The applicant was to document the committment in an FSAR amendment.

By letter dated November 19, 1981 and FSAR Amendment 14, the applicant provided the required information. The staff finds this acceptable.

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l 12 RADIATION PROTECTION L

12.3 Radiation Protection Design Features 12.3.2 Shielding In Section 12.3.2 cf the SER, the staff found the applicant's design review of the plant shielding allowing access to plant areas after an accident acceptable, subject to receipt of work dose data and a list of required plant modifications.

This information has been provided.

The applicant's shielding design review demonstrated that the criterion of less than 15 mr/ hour is met by Clinton for vital areas requiring extended or contin-uous occupancy (the control rcom and technical support center). Additionally, GDC 19 is met for those areas requiring only infrequent access.

The applicant has identified two plant modifications required as a result of postaccident radiation and shielding design review: (1) additional shielding around the postaccident sampling panel and (2) routing the exhaust from the main steam isolation valve leakage control system into the standby gas treat-ment system.

The staff concludes that postaccident access for Clinton has met the criteria of NUREG-0737, Item II.B.2, and, therefore, is acceptable.

Clinton SSER 1 12-1

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l 13 CONDUCT OF OPERATIONS 13.3 Emergency Preparedness Evaluation 13.3.1 Introduction As required by the regulations, Illinois Power Company filed with the NRC an emergency plan for Clinton Power Station. This plan (Appendix 13.B of the FSAR) was revised as FSAR Amendment 7 in September 1981 and as Amendment 16 in May 1982.

i The plan (Amendments 7 and 16) was reviewed against t'le 16 planning standards of 10 CFR 50.47(b), the requirements of Appendix E to 10 CFR 50, and the spe-l cific criteria of NUREG-0654/ FEMA-REP-1, Revision 1, as required by SRP Sec-

! tion 13.3 (NUREG-0800).

The Federal Emergency Management Agency (FEMA) review of the offsite emergency i

plans has not been done because the plans are not complete. This review and l subsequent submittal of findings to the NRC by FEMA must be completed before a license for operation above 5% of rated power is issued.

This report follows the format of Part II of NUREG-0654 in that each of the i

planning standards is listed separately followed by a summary of applicable portions of the plan that relate principally to that specific standard. The conclusions of the staff review are provided in Section 13.3.3.

13.3.2 Evaluation 13.3.2.1 Assignment of Responsibility (Organizational Control)

The Federal, State, and local, organizations that are intended to be part of the I

overall response organization for emergency planning zones are identified. The l

role of the State of Illinois is fully described, with reference to the State l Radiological Emergency Plan.

The concept of operations and its relationship to the total effort is described, and a block diagram showing the interfaces between and among the principal response organizations is provided.

l The Emergency Coordinator is identified as the individual who will be in charge l of emergency response on site. The Emergency Coordinator is the Power Plant Manager, but the position is filled by the Shift Supervisor until the Power Plant Manager arrives on site to relieve him. The Emergency Coordinator reports to the i Recovery Manager and is responsible for keeping the proper authorities informed of the status of the emergency. The position of Recovery Manager is filled by the Vice President, Nuclear Station Engineering.

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The applicant provides a description of the capabilities of the onsite and offsite organizations for continuous (24-hour) operation for a protracted j period.

l Written agreements are included to verify assistance arrangements between the '

plant and other support organizations. j 13.3.2.2 Onsite Emergency Organization In an emergency situation, the Shift Supervisor assumes the function of Emer-gency Coordinator and, as such, has the authority and responsibility to imple-ment the plan and initiate any necessary emergency actions, including notifi-cation of and recommendation of protective actions to local authorities. The line of succession for the Emergency Coordinator is the Assistant Shift Super-visor, Control Room Supervisor, and Senior Licensed Reactor Operator. The

! Plant Manager or designee assumes the position of Emergency Coordinator upon j arrival on site and after becoming thoroughly cognizant of the situation.

Police, ambulance, medical, hospital, and fire-fighting support that can be provided by local agencies is identified.

The functional responsibilities of the Emergency Coordinator are established in Guide 4-1. This guide clearly specifies that the Emergency Coordinator may not delegate the responsibility to notify and to make protective action recommenda-tions to offsite authorities.

Most station staff emergency assignments have been made and the relationships between the emergency organizations and normal staff complement are speci- i fied in the plan. Positions and/or titles of shift and plant staff personnel,  !

both onsite and offsite, assigned emergency functional duties are listed in Guides 4-1 through 4-14. Minimum shift manning requirements and provisions for timely shift augmentation are provided in Table 4-1. However, no Director of the emergency operations facility (EOF) is specified, although Clinton Power Station commits to having the EOF fully functional within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

13.3.2.3 Emergency Response Support and Resources The EOF will be activated for a site area or general emergency. It will be the central location for collecting and providing information and making recommen-dations for offsite protective actions. Provisions are made to accommodate representatives from Federal, State, and local government organizations and contractor and other support groups. Additionally, provisions are made for the dispatch of the applicant's representatives to the offsite governmental emer-gency operation centers (E0Cs).

The plan identifies the radiological laboratories and their availability.

Table 5-1 lists the available onsite equipment and specifications. Additional mobile nuclear laboratory facilities can be provided by the Illinois Department of Nuclear Safety on request. Its Radiological Assessment Field Team mobile lab has the capability of field analysis and radiological assessment. Esti-mated response time for the mobile lab is 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less. Other laboratory assistance is available from other reactor sites in Illinois and from Radiation Management Corporation.

Clinton SSER 1 13-2

Technical assistance may be requested from the Clinton architect-engineer, Sargent & Lundy, and from the nuclear steam supplier, General Electric, as well as from suppliers of various equipment used on site. The Institute of Nuclear l

Power Operations is also available as is the U.S. Department of Energy at Argonne, Illinois.

13.3.2.4 Emergency Classification System The applicaht defines five emergency classes:

(1) transportation accident (2) unusual event (3) alert (4) site emergency (5) general emergency Emergency action levels (EALs) are established based on onsite and offsite radiation monitoring information and on readings from various reactor sensors. <

The applicant has established action level criteria, which are essentially a generic classification for conditions that could initiate one of the five emergency classes.

Clinton has identified plant system and effluent parameters and annunciators characteristic of a spectrum of offnormal conditions and accidents. These parameters are divided into 10 broad sections:

(1) emergency core cooling systems (2) radiation monitoring (3) abnormal temperatures (4) failure of safety system (5) control power (6) fires (7) security (8) natural phenomenon (9) Technical Specifications (10) other hazardous conditions Each of these sections is then further divided into specific areas that fall under the section. Listed under the specific areas are EALs indicative of off-normal plant conditions with information on how the EAL can be identified by control room personnel.

13.3.2.5 Notification Methods and Procedures Provisions for notification of response organizations and response personnel are described, including a means for verifying the authenticity of the notifi-cation. Onshift personnei will be notified by the plant public address system, and offduty and corporate personnel can be notified by either telephone or the companywide radio and pocket paging systems.

Clinton has committed to provide a prompt public notification system that will meet the design objectives of Appendix 3 to NUREG-0654. A general description Clinton SSER 1 13-3

of the system is provided in the plan but the exact design has not been com-pleted. In general, the system will consist of (1) tone-activated radio receivers in areas of low population density and in special facilities and (2) outdoor warning devices in areas of high population density anc in recreation areas. It will be designed so that the intended recipient will be alerted by an acoustic signal without any special action required of him.

13.3.2.6 Emergency Communications Communications between the nuclear facility, State, and local E0Cs and radio-logical monitoring teams are provided.

The station communication system is designed to provide reliable, redundant, and diverse communications to all essential onsite and offsite locations during normal operations and under accident conditions. Station systems consist of a public address system, pocket pager system, diai-telephone system (PBX), microwave sys-tem, sound powered telephone, and intraplant two-way radio system. In addition to regular, inplant communications systems, there exist several specialized telephone and radio systems that have been developed for use in an emergency.

These are (1) nuclear accident reporting system (2) NRC/Clinton Power Station emergency notification system (3) health physics network (4) emergency automatic ringdown Communications between the technical support center and the E0F consist of several telephone systems and a radio system.

There will be a radio base station remote control console in the EOF to communi-cate with mobile monitoring teams. Communications with the State and local governments will be tested monthly. Joint tests with State, local, and Federal agencies will be conducted quarterly, and tests between Clinton, State, and local emergency operation centers and field assessment teams are tested annually.

13.3.2.7 Public Education and Information The applicant provides for periodic dissemination of information to the public in the 10-mi emergency planning zone. This will be done by various means and may include periodic information notices in utility bills, posting of informa-tion in public areas, and distribution of publications on an annual basis. The information will include instructions to the public on how they will be notified of an emergency and what their initial actions should be.

In emergency situations, the Public Relations Manager will disseminate initial and followup information through the news media by means of periodic press releases. This individual will also be responsible for responding to rumors.

On an annual basis the manager will contact members of news agencies to acquaint them with the emergency plans, information concerning radiation, and the points of contact for the release of public information in an emergency.

Clinton SSER 1 13-4

13.3.2.8 Emergency Facilities and Equipment l'

Emergency. facilities needed to support an emergency respons'e are provided.

They include a technical support center (TSC), emergency operations facility (E0F), and operations support center (OSC). The TSC is adjacent to the control room and was originally designated as the Unit 2 control room. This allows easy and timely access to the control room and inherently meets the habita-bility criteria for a TSC. The OSC is adjacent to the TSC and essentially consists of the balance of the area originally designated as the Unit 2 control room. Both the TSC and the OSC are activated for an alert or higher emergency classification.

The primary EOF is located in the Clinton Power Station construction office building outside the plant security fence about 1,000 ft east of the control room. It will be used to evaluate and coordinate emergency reentry / recovery operations on a continuing basis by the applicant and Federal and State offi-cials. It also will be the center for coordination of field-monitoring infor-mation. The applicant provides for an alternate EOF at Illinois Power Company's Electric and Gas Dispatch Center. This location places the backup E0F about 21.7 mi from the main control room and technical support center.

Onsite monitoring systems including geophysical phenomena monitors (meteoro-logical, hydrological, and seismic), radiological and process monitors, and fire and combustion product detectors are established. Arrangements have been made to acquire data from or have emergency access to offsite monitoring and analysis equipment.

The applicant will maintain a number of emergency kits which contain protective )

equipment, communications equipment, radiological monitoring supplies, and other emergency supplies. Provisions are made to inventory, inspect, and check these kits at least once each quarter and after each use. There are enough  :

reserves to replace equipment removed from kits for maintenance or calibration.

The applicant commits to maintaining meteorological instrumentation to ensure that sufficient meteorological data are available. However, the description presented must be improved to demonstrate that the equipment will satisfy the criteria of NUREG-0654, Appendix 2.

13.3.2.9 Accident Assessment The plan identifies the instruments and equipment used to identify the spectrum of offnormal conditions in Table 11-1 and relates the spectrum to accident classifications.

The applicant will have a radiation monitoring system (RMS), which includes the area radiation monitoring system, the process radiation monitoring system, and centralized digital processing, annunciating, and control equipment. Using informaticn from this system, the applicant can determine the source term and magnitude of releases of radioactive materials by use of the RMS computer. The system is also capable of calculating the relationship between effluent monitor readings and onsite and offsite exposure and contamination for various meteoro-logical conditions. A postaccident sampling system also will be installed to Clinton SSER 1 13-5

I allow sampling of the primary coolant, suppression pool water, drywell and ,

containment sumps, drywell and containment atmospheres, and effluent from the I t

reactor water cleanup system while limiting radiation exposure to operating l

personnel.

As stated in the previous section the applicant will have the capability and resources to acquire and evaluate meteorological data that meet Appendix 2, j NUREG-0654 criteria. These data will be available on the RMS cathode-ray tube terminal in the main control room, the technical support center, and the emergency operations facility. No provisions are made for access of this information by an offsite NRC center, or for interconnections to the State if it desires to independently verify the calculations.

The plan describes the capabilities arid resources for field monitoring within the plume emergency planning zone (EPZ), including notification of personnel and the equipment and expertise needed for making rapid assessment of actual or potential radiological hazards. The Plan also states that the applicant has the capability to detect and measure radioiodine levels as low as 10 7 pCi/cc under field conditions. The means for relating measured field contamination levels to dose rates will be provided in the implementing procedures. l Dose information may be used in conjunction with dose projections generated by l the RMS computer to develop a picture of actual plume travel, integrated dose, and dose estimates.

13.3.2.10 Protective Response The applicant has established onsite protective measures for employees, contractor personnel, and visitors who may be on site at the time of an emergency. These measures consist of warning and notification, relocation and accountability, and protective actions. Primary, secondary, and alternate assembly areas are designated to which nonessential personnel would be relocated in a site area or general emergency. Initial accountability would then be completed within about 30 minutes. Radiological monitoring and decontamination will be performed at the offsite reassembly areas, if necessary. For individuals remaining on site, Clinton will provide protective clothing, respiratory equipment, and radio-protective drugs.

The plan provides for the prompt notification and recommendations of protective actions to State and local authorities for the population-at-risk in the plume exposure pathway EPZ. The specific recommendation may be sheltering or evacu-ation depending on the magnitude of the projected dose, meteorological condi-tions, the nature of the release, and the predetermined evacuation time esti-mates for the sector (s) affected. Evacuation time estimates are provided in addition to the decision tree to aid Clinton personnel in making this recom-mendation.

Evacuation routes and distribution of permanent and transient population are also provided.

13.3.2.11 Radiological Exposure Control The applicant has established a radiation protection program for controlling radiological exposures in an emergency. The applicant will use onsite exposure Clinton SSER 1 13-6

guidelines consistent with the Environmental P.rotection Agency's Emergency i Worker and Lifesaving Activity Protective Action Guides for (1) removing injured persons

( (2) undertaking corrective actions (3) performing assessment actions (4) providing first aid (5) performing personnel decontamination (6) providi.ng ambulance service (7) providing medical treatment services The Emergency Coordinator or designee can authorize emergency workers to receive doses greater . hat 10 CFR 20 limits in the course of carrying out lifesaving activities.

Onsite contamination control measures for personnel, equipment, and access control are provided. The criteria for decontamination of personnel and equipment are specified in emergency procedures, and guidance criteria are given for returning previously restricted areas to normal use.

The plan provides for 24-hour per-day determination of doses received by onsite emergency workers and offsite response personnel and for appropriate record-keeping. Personnel monitoring devices will be read at appropriate frequencies to ensure accurate dose measurement and records.

13.3.2.12 Medical and Public Health Support Arrangements have been made for local and backup hospital and medical sources with a capability for treating contaminated patients. The local hospital is the John Warner Hospital in Clinton and the backup hospital is Northwestern Memorial Hospital in Chicago. . Agreements have been made with the Clinton Ambulance Service to provide emergency transportation of the injured personnel.

Local medical practitioners have also been trained in treating radiation victims.

Decontamination and first aid facilities exist on site in the service building.

Additional decontamination facilities exist in the control building and the radwaste building.

13.3.2.13 Recovery and Reentry Planning and Postaccident Operations The plan describes the applicant's general plans for recovery and reentry. It describes a recovery organization that is activated on declaration of a site area emergency or a gene r?l emergency. It Consists of Corporate management, administrative, and technical support personnel and is the official interface between Illinois Power Company, Clinton Power Station, and all outside organiza-tions. It is headed by a Recovery Manager who is the Vice President in charge of Nuclear Station Engineering. The Plan also describes the responsibilities and duties of other managers in the recovery organization.

A method for periodically estimating total population exposure is established using the radiation monitoring system computer.

Clinton SSER 1 13-7

13.3.2.14 Exercises and Drills An emergency preparedness exercise will be conducted before initial fuel load and on an annual basis. Provisions are made to start an exercise between  ;

6:00 p.m. and midnight, and another between midnight and 6:00 a.m. once every <

6 years. Some exercises will be unannounced. In addition to the annual exer- l cises, drills will also be conducted. The plan includes a description and schedule for the following drills: -

(1) communications (2) fire (3) medical emergency (4) radiological monitoring (5) radiation protection Critiques will be conducted after all drills and exercises with formal evalua-tions resulting from them. The Emergency Coordinator is responsible for 1 coordinating these critiques, for reviewing reports associated with the exer- i cises, and for initiating action to correct any deficiencies.

13.3.2.15 Radiological Emergency Response Training j The applicant will provide training and annual retraining for those personnel who have emergency response responsibilities. This training will include j participation in practical drills. The Supervisor-Training is responsible for providing this activity.

The applicant will also provide training for those offsite organizations such j as fire, police, medical support, and rescue personnel, whose services may be required in an emergency. The training will be consistent with the organiza-tion's emergency functions and will include procedures for notification and i basic radiation protection. For those local services personnel who will enter the site, training also will include site access procedures.

13.3.2.16 Responsibility for the Planning Effort Development, Periodic Review, and Distribution of Emergency Plans The Power Plant Manager has the authority and responsibility for the appli-cant's emergency response planning and will designate a member of the Clinton staff as the station Emergency Planning Coordinator to assist him. This t coordinator will be responsible for (1) ensuring coordination of the Clinton plan with other plans, (2) ensuring that procedures are consistent with the plan, (3) coordinating the review and update of the plan and procedures, and (4) maintaining an up-to-date listing of telephone numbers.

The plan, as well as any changes thereto, is provided to the organizations and l individuals having a responsibility for its implementation. It will be con- i trolled in the same manner as other station procedures are controlled. Provisions exist for an annual review of the plan and for incorporation of necessary revi- i

< sions. An independent review of the overall emergency preparedness program will I be conducted at least annually by the Corporate Quality Assurance Department.

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13.3.3 Conclusions On the basis of its review against the criteria in NUREG-0654, Revision 1, the staff concludes that providing the items identified are accomplished, the Clinton Power Station Emergency Plan will provide an adequate planning basis for an acceptable state of emergency preparedness and will meet the require-ments of 10 CFR 50 and Appendix E.

The following actions must be taken:

(1) The applicant must specif, by title the person who will function as the EOF Director and shall revise Table 4-1 to include the position of EOF Director.

(2) The public prompt notification system design shall be completed and the system installed.

(3) The description of the meteorological instrumentation (as discussed in Section 13.3.2.8) must be improved to demonstrate that the instrumentation will meet the criteria of Appendix 2 to NUREG-0654, including the capa-bility for remote interrogation.

13.5 Administrative Procedures The applicant was asked to review and modify procedures for removing safety-related systems from service (and restoring to service) to ensure that the operability status is known (II.K.1.10). In the SER, the staff reported that the applicant indicated that the procedures will ensure that independent verification of system lineups is applied to (1) valve anc electrical lineups for all equipment important to safety, (2) surveillance procedures, and (3) restoration to service following maintenance. The applicant has indicated that these procedures and review have been completed. The staff will inspect the records and procedures on site and will confirm that the requirements of NUREG-0737 have been met before the operating license is issued.

13.7 Security and Safeguards The applicant has submitted security plans entitled "Clinton Power Station Physical Security Plan," "Clinton Power Station Security Force Training and Qualification Plan," and "Clinton Power Station Safeguards Contingency Plan,"

for protection against radiological sabotage. The plans were reviewed in accordance with SRP Section 13.6, " Physical Security" (NUREG-0800).

As a result of the staff's evaluation, certain portions of the Clinton Power Station Physical Security Plan were identified as requiring additional informa-tion or upgrading to satisfy the requirements of 10 CFR 73.55. The applicant has been informed of the areas requiring revision and has made the necessary changes shown in the attachment to an Illinois Power Company letter dated March 23, 1982. Based on these revisions, the plan has been determined to meet the requirements of 10 CFR 73.55 and is, therefore, acceptable.

The Clinton Power Station Security Force Training and Qualification Plan and the Clinton Power Station Safeguards Contingency Plan have also been determined Clinton SSER 1 13-9

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to meet the requirements of Appendices B and C to 10 CFR 73 and are, therefore, acceptable.

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] An ongoing review of the progress of the implementation of the approved plans 4

will be performed by the staff to ensure conformance with the performance  ;

requirements of 10 CFR 73. '

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' The identification of vital areas and measures used to control access to these ar.eas, as described in the plan, may be subject to amendments in the future.

i The staff has determined that the above-referenced plans contain safeguards j information that must be protected against unauthorized disclosure in accord-ance with 10 CFR 73.21.

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l l 16 TECHNICAL SPECIFICATIONS The staff identified nine issues in the SER that were to be included in the Technical Specifications. Four additional issues are identified in this supplement. These issues are listed below and discussed further in the sections of this report as indicated by parentheses.

(10) Monitor and Maintain VHS Shoreline (2.6.5.5)

(11) Periodic Leak Testing of Pressure Isolation Valves (3.9.6)

(12) Leak Testing of Filter Trains (6.4)

(13) Testing of Protective Devices (8.4.7)

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l 18 CONTROL ROOM DESIGN REVIEW l 18.2 Discussion As described in the SER, the staff conducted a control room design review at the Clinton Power Station on November 3 through November 5, 1981. The human engineering discrepancies (HEDs) identified by the review team are documented in the staff's control room design review audit (CRDR/A) report dated Decem-ber 3, 1981, which was transmitted to the applicant in December 1981.

The staff met with the applicant in Bethesda, Maryland, on January 5,1982.

Identified HEDs were discussed, measures for correcting the HEDs were resolved, and a schedule for correcting them.was established. The applicant's proposed i

resolutions and implementation schedules for HEDs listed in the CRDR/A report l are documen+ed in letters of January 13 and February 9,1982, and are accept-able to the staff.

, As stated in the SER, a large number of systems and items were not available l for review'during the site visit. Using the guidance provided in Section 6 of l NUREG-0700, the staff requires that the applicant perform an evaluation of these systems and items and submit the findings, proposed corrective actions, and schedule for implementing the actions. The evaluation of these items shall be submitted for staff review and approval not later than 120 days before the scheduled issuance of the operating license.

The applicant's proposed resolutions and implementation schedules for HEDs listed l in the CRDR/A report are acceptable to the staff. Actions implemented to correct discrepancies will be audited by the NRC. The applicant has committed to perform an evaluation of certain items to resolve them before licensing. For these items, the staff requires the applicant to report the findings, proposed corrective actions, and schedule for implementing the actions for staff review not later than 120 days before the scheduled issuance of the operating license.

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l Clinton SSER 1 18-1

'22 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS A Subcommitee of the Advisory Committee on Reactor Safeguards considered the application for an operating license for Clinton Unit 1 at a meeting in Decatur, Illinois, on February 25-26, 1982. The Subcommittee visited the site and toured the facility on February 25, 1982. The full Committee reviewed the application at its 263rd meeting on March 5, 1982. A copy of the Committee's report to the chairman of the Nuclear Regulatory Commission dated March 9,1982 is included in this supplement as Appendix E.

The Committee's review included an evaluation of the management organization, the operational staff, and the training program. It was noted that staffing for plant startup and operation was not yet complete and the staff will ccm-h plete its review before fuel loading. The staff's review will be reported in a future supplement to the SER.

The Committee noted that the applicant is currently restructuring the construc-tion and operational quality assurance and quality control organization in response to staff concerns. The staff has met with the applicant on several occasions over the last 3 months to discuss the operational quality assurance (QA) program. The applicant has made major changes to the QA operational program.

The staff will complete its review when the applicant submits the revised QA program information. This review will be reported in a future supplement to the SER.

The Committee noted that the staff had not completed its review of the suppres-sion pool dynamic loads and hydrogen control systems for Mark III containments.

These items will be considered on a generic basis by the Committee.

The Committee reviewed the applicant's site-specific spectra and reanalysis of certain structures and components. The Committee recommended that specific attention be given to the seismic capability of the emergency ac power supplies, the dc power supplies, and the small components such as actuators and instrument lines that are part of the decay heat removal system. By letter dated May 19, 1982, the applicant submitted a proposed program to address the Committee's recommendations. The staff will report on this program in a future supplement to the SER.

The Committee noted that the SER dated February 1982 identified a number of unresolved safety issues, outstanding issues, confirmatory issues, and license conditions that the staff must review further.

Clinton SSER 1 22-1

APPENDIX A CONTINUATION OF CHRONOLOGY December 3, 1981 Letter from LRG-II forwarding Vol. II, " Licensing Review Group II Position Papers."

December 4, 1981 Letter from applicant responding to violations noted in Inspection Report 50-461/81-24. Corrective actions:

requirements beyond ANSI B96.1-73 deleted by Amendments 7 and 11 to FSAR.

December 10, 1981 Letter to applicant forwarding Federal Emergency Manage-ment Agency (FEMA) Guidance Memorandum 1-18 regarding transfer of offsite emergency preparedness to FEMA.

December 11, 1981 Letter to applicant forwarding request for information l to support the seismic qualification review team site visit.

December 14, 1981 Letter from applicant forwarding additional informa-tion on fire detection system and power generation control complex (PGCC) fire detection for the facility.

l December 14, 1981 Letter from applicant forwarding Inspection Report

50-461/81-31. No compliance was noted.

December 16, 1981 Letter from applicant responding to Inspection Report 50-461/81-24. Corrective actions: initiated first interaction surveillance and set conservative interim interaction separation criteria.

! December 18, 1981 Letter to applicant with Draft Environmental Statement (DES) for proposed operation of facility.

1 l December 18, 1981 Letter to applicant requesting additional information.

December 21, 1981 Letter to applicant indicating that use of Perry simu-l lator for training cold-license candidates is accept-l able. Program modifications will depend on results of first sets of examinations.

December 28, 1981 Letter from applicant with evacuation time estilaates.

l December 28, 1981 Letter from applicant responding to violations noted in Inspection Report 50-461/81-27. Corrective ac-tions: spreading room drawings will be reviewed, revised, and reissued to include necessary fire bar-rier provisions for all ladder rack chutes.

Clinton SSER 1 A-1

December 30, 1981 Letter from applicant forwarding Amendment 11 to FSAR.

December 31, 1981 Letter from applicant advising of dynamic soil proper-ties review to define curves of soil shear modules values for seismic soil structure interaction analysis.

Estimated values of Poisson ratio in FSAR considered too low. Revised tables enclosed.

December 31, 1981 Letter to applicant requesting additional information on preservice inspection program.

January 6, 1982 Letter from applicant forwarding list of estimated quantities of organic materials in containment.

January 7, 1982 Letter to applicant requesting additional information on emergency plan submitted in Amendment 7 dated September 1981. Plan should be revised to address enclosed comments per provisions of revised 10 CFR 50 and NUREG-0654, Rev. 1.

January 8, 1982 Letter to applicant informing of acceptance of facil-ity contingency plan and security force qualification training program.

January 8, 1982 Letter from applicant responding to violations noted in Inspection Report 50-461/81-25. Corrective ac-tions: construction materials and debris on and around cabinets were removed and protective coverings

• restored.

January 11, 1982 Letter from applicant responding to December 31, 1981 request for schedule of response to questions on pre-service inspection program abstract. Response to Ques-tions 250.1 and 250.2 will be submitted by April 1, 1982 or before start of inspection program.

January 12, 1982 Letter to applicant on new license application survey.

January 12, 1982 Letter to applicant advising of schedule change for information to be submitted before emergency prepared-ness exercise.

January 13, 1982 Letter from applicant forwarding draft FSAR Chapter 14 pages and answers to questions.

January 13, 1982 Letter from applicant forwarding " Demonstration of Compliance to NRC Regulations" per request of September 29, 1980.

January 13, 1982 Letter from applicant forwarding update to containment l fracture toughness information submitted on August 7, 1981.

Clinton SSER 1 A-2

i January 13, 1982 Letter from applicant responding to letter of l December 11, 1981 concerning the Control Room Design Review / Audit Report.

January 15, 1982 Letter from " Hydrogen Control" owners group forwarding

" Hydrogen Control Program Document."

January 15, 1982 Letter to applicant forwarding Inspection Report 50-461/81-30. No noncompliance noted.

January 15, 1982 Letter from applicant forwarding additional informa-tion supporting special nuclear material application.

January 19, 1982 Letter from applicant forwarding site-specific spectra and hazard analysis information.

January 20, 1982 Letter from applicant forwarding response to request for additional information on containment ultimate l

capacity.

January 22, 1982 Letter to applicant confirming January 29, 1982 meet-ing with utility in Glen Ellyn, Illinois, to discuss

! concerns on proposed organizational arrangement of operational quality assurance (QA) program.

l January 22, 1982 Letter to applicant forwarding Inspection Report 50-461/81-32 and notice of violation.

January 25, 1982 Letter from LRG-II forwarding " Licensing Review Group II Position Papers," Vol. 3 consisting of technical uis-cussions and resolutions of 12 issues.

January 25, 1982 Letter to applicant forwarding list of confirmatory issues discussed with utility. Requests schedule for activities associated with closing each item.

January 27, 1982 Letter to applicant submitting confirmation of action letter on problems identified during ongoing inspec-j tion concerning safety-related electrical work.

l Stop-work order will be effective until improvements

approved by Region III.

f January 27, 1982 Letter from applicant forwarding rationale and ref-erences used to estimate shear wave velecity for l

glacial tills at site.

January 27, 1982 Letter from applicant forwarding proposed schedule for providing information to close draft SER confirmatory l

issues.

January 27, 1982 Letter from applicant incorporating into operating license (OL) application 12 position papers provided by LRG-II.

Clinton SSER 1 A-3

January 28, 1982 Letter from applicant forwarding Information Fotice 82-01, " Auxiliary Feedwater Pump Lockout Resulting from Westinghouse W-2 Switch Circuit Mod."

i January 29, 1982 Letter from applicant forwarding Amendment 12 to FSAR. )

January 29, 1982 Letter from applicant forwarding response to request for information on cable tray hangar design for OBE and safety / relief valve load combinations with revised writeup of expected maximum stresses in burned pipes and ducts from seismic shear wave.

February 1, 1982 Letter from applicant forwarding schedule for provid-ing information to close SER confirmatory issues.

February 4,1982 Letter to applicant forwarding SER (NUREG-0853).

February 8, 1982 Letter to applicant forwarding policy on factors caus-ing fatigue of operating personnel at nuclear reactors.

February 8, 1982 Letter to applicant forwarding pages 3-40 and 7-16 of SER, which were omitted from February 4, 1982 submittal.

February 9, 1982 Letter from aplicant forwarding Central Files version of responses to questions on emergency plan.

February 9,1982 Letter from applicant with changes to response to

" Central Room Design Review / Audit-Report," Appendix C. )

February 10, 1982 Letter to applicant advising that physical security plan needs further revision.

February 15, 1982 Letter from applicant clarifying onsite staffing pro-jections.

February 16, 1982 Letter from applicant forwarding " Site-Specific Response 1 Spectra" report.

February 16, 1982 Letter to applicant forwarding SER (NUREG-0853) and Federal Register notice.

l February 17, 1982 Letter from applicant forwarding additional informa- i I

tion on shear wave velocity revisions. '

February 19, 1982 Letter from applicant forwarding comments on NUREG-0854, ,

DES.  !

l February 19, 1982 Letter to applicant advising that purpose of February 23, 1982 meeting is to discuss corrective actions being taken to resolve QA organization and policies discussed during January 29, 1982 meeting.

February 22, 1982 Letter from applicant responding to violations noted in Inspection Report 50-461/81-32. Corrective action:

Clinton SSER 1 A-4

l cleaning activities on steam separator stopped and procedures revised.

February 23, 1982 Letter from applicant forwarding report on bypass leak-age of secondary containment used in design-basis analy-sis of radiological consequences of loss-of-coolant accident (LOCA).

February 23, 1982 Letter from applicant forwarding preservice inspection program consisting of nondestructive examination pro-gram ASME Code Class 1, 2, and 3 components and pre-service testing program for valve.

February 23, 1982 Letter from applicant forwarding list of actions taken since Jaruary 15, 1982 increasing utility involvement in contractor construction management and quality pro-grams and strengthening QA program.

February 26, 1982 Letter from applicant forwarding Amendment 13 to FSAR, consisting primarily of complete revision of Chapter 14.

February 26, 1982 Letter to applicant requesting additional information on automatic restart of high pressure core spray after manual termination.

February 26, 1982 Letter from applicant forwarding information on pre-operational test program.

March 2, 1982 Letter from applicant indicating no plans for new application or request for reviews.

March 2, 1982 Letter from applicant forwarding revised response to i FSAR Question 220.15 on shear wave velocities. '

March 3, 1982 Letter from applicant forwarding draft FSAR response to TMI Action Plan Item II.B.2.

March 4, 1982 Letter to applicant forwarding Inspection and Enforce-ment (IE) Management Meeting Report 50-461/82-03.

March 5, 1982 Letter to applicant forwarding report on antitrust review.

March 9, 1982 Letter to applicant on use of Institute of Nuclear Power Operations SEE-IN program.

March 10, 1982 Letter to applicant forwarding Information Notices 82-04, " Potential Deficiency of Certain AGASTAT E-7000 Series Time-Delay Relays," and 82-05, " Increasing Frequency of Drug Related Incidents."

March 10, 1982 Letter from applicant forwarding report on Mark III containment ventilation / purge.

Clinton SSER 1 A-5

March 11, 1982 Letter from applicant forwarding revised responses to questions on containment ultimate capacity.

March 11, 1982 Letter from applicant on static test program to deter-mine modules of rupture of concrete block masonry walls. j l

March 11, 1982 Letter to applicant forwarding Inspection Report 50-461/82-01. No noncompliance noted.

March 12, -1982 Letter from LRG-II forwarding Vol. IV of " Licensing Review Group II Position Papers" covering 11 of 56 i issues.

March 15, 1982 Letter from applicant forwarding background documents on thermal effluent limitations.

March 15, 1982 Letter to applicant indicating that staff fails to

. find acceptable basis for changing bypass leakage from 4% to 11%.

March 16, 1982 Letter to applicant forwarding Information Notice 82-07, " Inadequate Security Screening Programs."

March 16, 1982 Letter to applicant forwarding letter from Advisory Committee on Reactor Safeguards (ACRS).

I March 16, 1982 Letter to applicant forwarding draft acceptance cri-teria on Mark III LOCA-related pool dyrjamic loads. 1 i

March 16, 1982 Letter from applicant on violations noted in Inspec-tion Report 50-461/81-24. Additional corrective action: Sargent & Lundy (S&L) Procedure PI-CP-034 will be revised to reflect requirement for maintaining program controls.

March 17, 1982 Letter to applicant requesting detailed description of

actions taken to determine quality of installed com-pleted and working electrical items.

March 19, 1982 Letter from applicant providing status of SER issues.

March 22, 1982 Letter from applicant forwarding Endorsements 12, 13, l and 14 to American Nuclear Insurers' Policy NF-261. l 1

March 23, 1982 Letter from applicant forwarding soil-structure inter-action analysis of circulating water screenhouse and additional information on effect of far-field earth- l quake on buried pipe design.

March 23, 1982 Letter from applicant forwarding Revision 3 to physi-cal security plan.

I

, Clinton SSER 1 A-6

March 24, 1982 Letter from applicant with calculations to support ultimate heat sink sedimentation of 262 acre-f t during probable maximum flood.

1 March 26, 1982 Letter from applicant forwarding 218 oversize drawings associated with the preservice inspection program.

March 29, 1982 Summary of February 1982 meeting on separation of reactor protection system and main steam isolation valve solenoid circuits and PGCC circuits.

March 29, 1982 Letter to applicant requesting additional information on fast scram hydrodynamic loads on control rod drive system.

March 30, 1982 Letter from applicant advising of incorporation of position papers on LRG-II issues.

March 31, 1982 Letter from applicant forwarding Amendment 14 to FSAR.

April 1, 1982 Summary of March 30, 1982 meeting on Instrumentation and Controls Systems Branch position on remote shut-down panel.

April 2, 1982 Letter from applicant advising that plant design will comply with acceptance criteria for Mark III LOCA-i related pool dynamic loads.

• - April 2, 1982 Letter from applicant updating construction schedule.

April 5, 1982 Letter from applicant forwarding Supplement 3 to environmental report - OL stage.

April 6, 1982 Letter to applicant confirming April 8, 1982 meeting to discuss response to nonconformance and discrep-ancies and enforcement actions.

April 7, 1982 Letter from applicant forwarding revised procedures on as low as reasonably achievable program.

April 7, 1982 Letter from applicant advising that response to hydro-dynamic loads in control rod drive system will be pro-vided by September 30, 1982.

April 7, 1982 Letter to applicant forwarding Inspection Report 50-461/82-04. No noncompliance noted.

April 8, 1982 Letter from applicant forwarding transparencies pre-sented at March 8, 1982 meeting.

April 8, 1982 Letter from applicant addressing certain issues raised in review of remote shutdown panel.

Clinton SSER 1 A-7

April 12, 1982 Letter from applicant forwarding information on control room doses.

April 13, 1982 Letter from applicant forwarding response to Question 6 3 on degraded core hydrogen control.

April 14, 1982 Letter from applicant forwarding Illinois Power Company Annual Financial Report 1981; :oyland Power Cooperative, Inc. Annual Financial Report 1980; and Western Illinois Power Cooperative, Inc. Annual Financial Report 1981.

April 15, 1982 Letter to applicant transmitting NUREG-0909 on Ginna tube rupture.

April 20, 1982 Letter to applicant on environmental qualification of safety-related electrical equipment. NRC position on qualification requirements enclosed.

April 20, 1982 Letter from applicant forwarding computerized listing of status of outstanding and confirmatory issues.

April 20, 1982 Letter from applicant forwarding GE paper, " Analysis of Flexible Conduit Used as Electrical Fault Separa-tion Barrier Within PGCC," and description of S&L sketch of protection cable segregation, presented at February 9, 1982 meeting. l April 21 1982 Letter from applicant informing that current schedule i for construction completion and fuel loading is e January 1984. Commercial operation projected for August 1984.

April 23, 1982 Letter from applicant with data assessing impact on masonry walls of using lower damping values and ignor-ing joint reinforcement.

April 2?, 1982 Letter from applicant requesting that schedule for construction completion date be extended.

April 29, 1982 Letter from applicant forwarding response to request for additional information on fire detection system at facility.

April 30, 1982 Letter from applicant forwarding FSAR Amendment 15.

April 30, 1982 Letter from applicant dealing with nuclear steam sup-ply system (NSSS) pipe break analysis using SRP cri- 3 teria. j May 5, 1982 Letter to applicant informing that American Telephone

& Telegraph has revised time intervals for normal ser-vice requests for emergency notification system.

Clinton SSER 1 A-8

May 10, 1982 Letter from applicant forwarding approved draft ANSI /

ASME OM-3 on steady-state piping vibration acceptance criteria.

May 13, 1982 Letter to applicant forwarding draft technical evalu-ation, " Control of Heavy Loads at Nuclear Power Plants."

May 14, 1982 Letter to applicant informing of telephone number change for NRC Operations Center. Numbers should be used as backup to emergency notification system.

May 17, 1982 Letter from applicant forwarding results of review of NSSS engineered safety features system in response to IE Bulletin 80-06. Update incorporates results of balance of plant.

May 17, 1982 Letter from applicant forwarding " Licensing Review Group II Position Papers," Vol. V.

May 19, 1982 Letter from applicant forwarding equipment seismic assessment program in response to ACRS request.

May 20, 1982 Letter from applicant confirming new information for Final Environmental Statement on revised schedule for fuel load.

Y _

Clinton SSER 1 A-9

l i

APPENDIX B l-REFERENCES BWR Mark III Owners Group, " Study to Demonstrate the General Applicability of SRSS Combination of Dynamic Responses for Mark III Nuclear Steam Supply System and Balance-of-Plant Piping and Components," November 1981.

Code of Federal Regulations, Title 10, " Energy" (includes General Design Cri-teria).

General Electric Topical Reports, NEDO-21778-A, " Transient Pressure Rises l Affecting Fracture Toughness Requirements for Boiling Water Reactors,"

! December 1978.

-- , NED0-24708, " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," September 16, 1980.

! Illinois Power Company et al., " Final Safety Analysis Report, Clinton Power Station Units 1 and 2."

Sargent & Lundy, " Degraded-Core Hydrogen Control," Chicago, IL, October 16, l 1981.

Y l Smith, E. D., " Analysis of Flexible Conduit Used as an Electrical Fault Separa-I tion Barrier Within PGCC," General Electric Company, San Jose, CA, Febru-ary 5,1982.

U.S. Nuclear Regulatory Commission, "BWR Scram Discharge System Safety Evalua-tion," December 1, 1981.

-- , " Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls, and Quality Assurance," July 1977.

-- , NUREG-0313, " Technical Report on Material Selection and Processing Guide-lines for BWR Coolant Pressure Boundary Piping" Rev.1, July 1977.

l l -- , NUREG-0484, " Methodology for Combining Dynamic Responses," Rev. 1, May l 1980.

l -- , NUREG-0654/ FEMA-REP-1, " Criteria for Preparation and Evaluation of Radio-l logical Emergency Response Plans and Preparedness in Support of Nuclear l

Power Plants," Rev. 1, November 1980.

i

-- , NUREG-0700, " Guidelines for Control Room Design Reviews," September 1981.

-- , NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

Clinton SSER 1 B-1

-- , NUREG-0799, " Draft Criteria for Preparation of Emergency Operating Proce- 1 dures," June 1981. j

-- , NUREG-0800 (formerly NUREG-75/087), " Standard Review Plan for Review of l Safety Analysis Reports for Nuclear Power Plants--LWR Edition," July 1981 (includes Branch Technical Positions).

-- , NUREG-0853, " Safety Evaluation Report Related to the Operation of Clinton Power Station, Unit No. 1," February 1982.

-- , NUREG/CR-2686, " Review of Load Combinations for NSSS and B0P Piping and Equipment of Mark III Plants," Brookhaven National Laboratory, May 1982.

-- , Regulatory Guide 1.22, " Periodic Testing of Protection System Actuation Functions."

-- , Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants," l Rev. 2.

-- , Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," Rev. 2.

-- , Regulatory Guide 1.60, " Design Response Spectra for Seismic Design of Nuclear Power Plants," Rev. 1. l

-- , Regulatory Guide 1.75, " Physical Independence of Electric Systems," Rev. 2.

T

-- , Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release."

-- , Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," Rev. 1.

-- , Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident, Rev. 2.

-- , Regulatory Guide 1.118, " Periodic Testing of Electric Power and Protection

Systems," Rev. 2.

-- , Regulatory Guide 1.127, " Inspection of Water-Control Structures Associated With Nuclear Power Plants," Rev. 1.

-- , SECY 81-582, "TMI Action Plan II.F.2 (NUREG-0737) Additional Instrumenta-tion for Detection of Inadequate Core Cooling," October 7, 1981.

l U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement (I&E)

Bulletin 79-02, " Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts," March 2, 1970.

Clinton SSER 1 B-2

-- , Bulletin 79-27, " Loss of Non-Class 1E Instrumentation and Control Power System Bus During Operation," November 30, 1979.

-- , Bulletin 80-06, " Engineered Safety Features (ESF) Reset Controls,"

March 12, 1980.

Weston Geophysical Corporation, " Site Specific Response Spectra for Clinton Power Station - Unit 1 of Illinois Power Company," Boston, MA, May 1982.

Industry Codes and Standards American Society of Mechanical Engineers (ASME), " Boiler and Pressure Vessel Code,"Section III.

-- ,Section III, NB-2300, " Fracture Toughness Requirements for Materials."

l l

-- ,Section III, NB-2332, " Material for Piping, Pumps, and Valves, Excluding Bolting Material."

l -- ,Section III, NB-3653, " Consideration of Level A Service Limits."

-- ,Section XI, IWA-1500, " Accessibility."

-- ,Section XI, IWB-3000, " Standards for Examination Evaluation."

-- ,Section XI, IWB-4000, " Repair Procedures."

-- ,Section XI, IWC-3000, " Standards for Examination Evaluation."

! -- ,Section XI, IWC-4000, " Repair Procedures."

-- ,Section XI, IWD-3000, " Standards for Examination Evaluation."

-- ,Section XI, IWD-4000, Repair Procedures."

-- ,Section XI, IWV-3420, " Valve Leak Rate Test."

l -- ,Section XI, 1974 Edition with Addenda through Summer 1979.

-- , Summer 1972 Addenda.

, American Society for Testing Materials (ASTM), E-119, " Fire Test of Building l

Construction and Materials."

-- , E 185-73, " Standard Recommended Practices for Surveillance Tests for l Nuclear Reactor Vessels."

i Institute of Electrical and Electronics Engineers (IEEE), Standard 323-1974, i "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generat-ing Stations."

-- , Standard 338-1977, "IEEE Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems."

Clinton SSER 1 B-3

-- , Standard 344-1975, " Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations."

-- , Standard 384-1974, "IEEE Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits."

(

4 1

Clinton SSER 1 B-4

l APPENDIX D NRC STAFF CONTRIBUTORS AND CONSULTANTS This supplehent to the Safety Evaluation Report is a product of the NRC staff and its consultants. The NRC staff members listed below were principal con-tributors to this report. A list of consultants follows the list of the staff members.

NRC Staff Name Title Branch J. Lazevnich Reactor Systems Engineer Power Systems N. Fioravante Auxiliary Systems Engineer Auxiliary Systems B. Hardin Reactor Engineer Reactor Systems R. Kendall Reactor Engineer Instrumentation & Control Systems

! J. Read Nuclear Engineer-Chemist Accident Evaluation br K. Dempsey Nuclear Engineer A.cident Evaluation M. Lamastra Sr. Radiation Engineer Radiological Assessment A. Ramey-Smith Engineering Psychologist Human Factors Engineering D. Terao Mechanical Engineer Mechanical Engineering N. Chokshi Structural Engineer Structural Engineering C. Tan Sr. Structural Engineer Structural Engineering G. Giese-Koch Geophysicist Geosciences R. Wescott Hydraulic Engineer Hydrologic & Geotechnical Engineering B. Jaggannath Geotechnical Engineer Hydrologic & Geotechnical Engineering B. Elliott Materials Engineer Materials Engineering D. Smith Sr. Materials Engineer Materials Engineering M. Hum Sr. Materials Engineer Materials Engineering J. Stang Fire Protection Engineer Chemical Engineering Clinton SSER 1 D-1

Name Title Branch R. Skelton Plant Protection Analyst Physical Security Licensing ,

1 J. Mathis Reactor Safety Engineer Emergency' Preparedness Licensing B. DeFayette Emergency Preparedness Emergency Preparedness Licensing Analyst S. Sun Nuclear Engineer Core Performance Consultants Name Company R. Skaggs Pacific Northwest Laboratory B. Brown Idaho Nuclear Engineering Laboratory P. Nagata Idaho Nuclear Engineering Laboratory i

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I Clinton SSER 1 D-2

T APPENDIX E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS REPORT l ON CLINTON POWER STATION UNIT 1 I

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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WA$mNG TON, D. C. 20555

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• • • • • March 9, 1982 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

REPORT ON CLINTON POWER STATION UNIT 1

Dear Dr. Palladino:

During its 263rd meeting, March 4-6, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of the Illinois Power Company, the Soyland Power Cooperative, Inc., and the Western Illinois Power Cooperative, Inc. (Applicant) for a license to operate the Clinton l Power Station Unit 1. The plant is to be operated by the Illinois Power l Company. A tour of the facility was made by members of the Subcommittee on the morning of February 25, 1982 and a Subcommittee meeting was held I in Decatur, Illinois on February 25-25, 1982 to consider this applica-

- tion. During its review the Committee had the benefit of discussion with representatives of the Applicant and the NRC Staff. The Committee also had the benefit of the documents listed. The Committee commented on the applicatica for a permit to construct this Station in its report dated April 8,1975.

The Clinton Power Station is located in DeWitt County in east-central Illinois about 6 miles east of the city of Clinton and 22 miles north-northeast of Decatur. Unit i uses a General Electric BWR-6 nuclear steam supply system with a rated power level of 2894 MWt and a Mark III pres-sure suppression containment system with a design pressure of 15 psig.

Construction of Unit 1 is about 85% complete and Unit 2 is about 35 ccmpl ete . Construction of Unit 2 has been deferred indefinitaly, and the Applicant's motion to sever the Unit 2 proceedings from Unit i licensing proceedings has been granted. Ccnsequently, both the Committee and the NRC Staff have limited this rev'ew to Unit 1.

The Ccmmittee's review included an evaluation of the management organi-zati on , the operational staff, and the training program. The Clinton Power Station is the Applicant's first nuclear station and staffing for plant startup and operation is not yet comolete. The Applicant, however, has made considerable progress and has a well-established training pro-gram. The MRC Staff will continue to monitor the Appl fcant's progress and expects to complete its review before fuel loading.

The Applicant is currently restructuring the construction and operational quality assurance and quality control o ganization in response to NRC Staff concerns. The revised organization will be reviewed and audited by the NRC Staff. The Ccmmittee wishes to be kept informed on this matter.

Clinton SSER 1 E-1

Honorable Nunzio J. Palladino March 9, 1982 The Mark III suppression pool dynamic loads have been identified as an Outstanding Issue in the NRC Staff'r review. The NRC Staff has provided the Applican't with a proposal for the appropriate design basis loads, and it appears that the Clinton design will be able to accommodate these loads. The Committee will continue to discuss, on a generic basis, the Mark III suppression pool dynamic loads with the NRC Staff.

Hydrogen control systems for Mark III containments are being developed by the Mark III Owners Group. Efforts by this Nners Group are being directed toward the development of a hydrogen ignition system which makes use of distributed ignition sources. The NRC Staff has indicated that they will be able to meet with the Committee on this matter in the near future. The Committee expects to review this system on a generic basis.

Acceptability of this system is a License Condition.

The Applicant, in response to NRC Staff requirements, has reevaluated certain safety-related systems of the Clinton design using the ground motion parameters that describe the site-specific spectra equivalent to a design basis earthquake of Mb equal to 5.8. The Applicant has reana-lyzed what he believes to be the limiting structures and components using this new response spectruni and has concluded that all Seismic Category 1 structures will withstand the design basis earthquake. Work by tha Applicant is continuing. The Committee believes that specific ,

attention should be given to the seismic capability of the emergency AC power supplies, the DC power supplies, and small ccaponents such as actuators and instrument lines that are part of the decay heat removal system. This matter should be resolved in a manner satisfactory to the NRC Staff. The Committee wishes to be kept informed.

1 In its Safety Evaluation Report dated February 1982, the NRC Staff has identi fied a number of Unresolved Safety Issues as being applicable to Clinton as well as a number of Outstanding Issues, Confimatory Issues, and License Conditions. We believe that if due consideration is given to these matters and to our recommendations above, and subject to satis-factory completion of construction, staffing, and preoperational testing, there is reasonable assurance that the Clinton Power Station Unit I can be operated at power levels up to 2894 MWt without undue risk to the health and safety of the public.

Sincerely,

\.

P. Shewmon Chairman References

1. Illinois Power Company, et al., " Final Safety Analysis Report, Clinton Power Station Units 1 and 2" with Amendments 1-12.

2. U.S. Nuclear Regulatory Commission, " Safety Analysis Report Related to the Operation of Clinton Power Station Unit 1," NUREG-0853, dated February 1982.

Clinton SSER 1 E-2

APPENDIX F ERRATA TO CLINTON POWER STATION SAFETY EVALUATION REPORT Pm Line/ Item Change 1-4 Item II.B.3 Change SER Section from "9.3.4" to "9.3.5".

1-4 Item II.B.4 Change SER Section from "13.2.2" to "13.2.1".

1-5 Item II.K.3.16 Change SER Section from "5.2.3" to "5.2.2" 1-5 Item II.K.3.45 Change SER Section from "5.2.3" to "5.2.2".

1-9 Design Features Change " maximum critical pc.ecr ratio" to

" minimum critical power ratio".

1-11 uutstanding Issue (12) Change "(7.3.3.5)" to "(7.3.3.7)".

1-12 Confirmatory Issue (9) Change "(3.9.3)" to "(3.9.2)".

(

1-12 Confirnatory Issue (10) Change "(3.9.2)" to "(3.9.3)"

k 1-12 Confirmatory Issue (11) Change "(3.9.2)" to "(3.9.3)".

1-12 Confirmatory Issue (12) Change "(3.9.2)" to "(3.9.3.4)".

1-12 Confirmatory Issue (21) Omit second reference to Section 6.2.7.

1-12 Confirmatory Issue (44) Change "(13.6.1)" to "(13.5)".

1-12 Confirmatory Issue (46) Change "II.K.2.15" to "II.K.3.15".

1-13 Confirmatory Issue (60) Change "(9.5.4.5)" to "(9.5.4.6)".

1-13 Confirmatory Issue (65) Omit Confirmatory Issue (65).

3-41 32 Change "May 20" to "May 27" 7-21 11 Change the sentence " Staff results were in substantial agreement...GE value of 10 6 per year." To read " Staff results show that the probability of a second valve to open as a result of a single failure is 6.6 x 10-3 per year as compared with the GE value of 1.3 x 10 6 per year."

7-21 18 Omit the sentence "In either case, the.

is very low."

Clinton SSER 1 F-1

(

Page Line/ Item Change 7-21 20 Change sentence "In addition,. . acceptable limits." to read "The probability value has been found to be acceptably low by the staff in view of the fact that the staff has evalu- l ated the consequences of simultaneous open-ing of two relief valves on second and sub-sequent valve pops and has found that the loading to the containment is within accept-able limits."

10-5 15 Af ter " Class 2" omit "and 3".

11-3 19 Insert the words " Appendix A" before Table 3.2. Also, change " Table 3.2" to

" Table 3.3" 12-1 39 After " modify" add the words "as appropriate."

13-1 22 Change "in November. " to "on November. ".

13-3 38 Change "yr" to " years" i

15-8 14 Change "13.5.2" to "13.6.3" <

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l Clinton SSER 1 F-2

F oRu 335 f,R , u s NUCLE AR REGUL ATORY COMMISSION BIBLIOGRAPHIC DATA SHEET NUREG-0853 Supplement 1

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L 4 Tt TLE AND SUBTlTLE (Add Volume No., et wormeran) 2. (Leave blanal i

Safety Evaluation Report related to the operation of Clinton Power Station, Unit No.1 3. RECIPIENT'S ACCESSION NO.

I ,

e l / AUTHOHISI 5. D ATE REPORT COMPLE TED MONTH l YEAR

.1o n n 1982 9 PE HF OHMING ORGANIZATION N AME AND M AILING ADDRESS (/nclude 2,p Codel DATE REPORT ISSUED MONTH l YEAR U. S. Nuclear Regulatory Commission July 1982 Office of Nuclear Reactor Regulation s (t e,ve b,a,,

• i l Washington, D. C.

20555 8 (Leave blann) 12 SPONSOHING OHGAN12 ATION N AME AND M AILING ADDRESS (/nclude 2,p Codel p

Same as 9, above

11. CONT R ACT NO l 1.4 iYPE OF REPORT PE RIOD COV E RE D (Inclusive dal*s) lSafetyEvaluationReport 15 57PPL EMENTARY NOTES 14 (Leave osana s Docket No. 50-461 p su aosT H AcT (200 anads or ie,so Supplement No.1 to the Safety Evaluation Report on the application filed by Illinois I i Power Company, Soyland Pcwer Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit f No. I has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission. The facility is located in Harp Township, Dewitt County, Illinois.

This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.

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