ML20140C120
| ML20140C120 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 05/31/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0853, NUREG-0853-S03, NUREG-853, NUREG-853-S3, NUDOCS 8406190045 | |
| Download: ML20140C120 (40) | |
Text
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NUREG-0853 Supplement No. 3 Safety Evaluation Report related to the operation of Clinton Power Station, Unit No.1 Docket No. 50-461 Illinois- Power Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation May 1984 y no v, a i wa ;
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%1 I NOTICE Availability'of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be availebte from one of the following sources: l
- 1. The NRC Public Document Room,1717 H Street, N.W.
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l GPO Printed copy price: '$4.00 0
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NUREGM53 Supplement No. 3 Safety Evakation Report related to the operation of Clinton Power Station, Unit No.1 Docket No. 50461 lilinois Power Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation May 1984 l pa m.,,,
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ABSTRACT
-Supplement No. 3 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and 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 supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplements No. 1 and No. 2.
Clinton SSER 3 iii
l l
TABLE OF CONTENTS P_ age ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 1 INTR 9 DUCTION AND GENERAL DISCUSSION . . . . . . . . . . . . . . 1-1 1.1 Introduction. . . . . . . .. . . . . . . . . . . . . . 1-1 1.9 Outstanding Issues. . . . ... . . . . . . . . . . . . 1-1 1.10 Confirmatory Issues . . . . . . . . . . . . . . . . . . 1-3 1.11 License Conditions. . . . . . . . . . . . . . . . . . . 1-7 2 SITE CHARACTERISTICS. . . . . . . . . . . . . . . . . . . . . . 2-1 2.2 Nearby Industrial, Transportation, and Military Facilities . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2.1 Transportation Routes . . . . . . . . . . . . . . 2-1 2.3 Meteorology. . .... . . . . . . . . . . . . . . . . 2-2 2.3.3 Onsite Meteorological Measurements Program. . . . 2-2 2.5 Geology and Seismology . . . . . . . . . . . . . . . . . 2-3 2.5.2 Seismology. . . . . . . . . . . . . . . . . . . . 2-3 2.5.2.4 Safe Shutdown Earthquake (SSE) . . . . . 2-3 2.5.2.6 Summary. ... . . . . . . . . . . . . . 2-5 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS'. . . . 3-1 3.7 Seismic Design . . . . . . . . . . . . . . . . . . . . . 3-1 3.7.1 Seismic Input and . .. . . . . . . . . . . . . . 3-1 3.7.2 Seismic System Analysis . . . . . . . . . . . . . 3-1 4 REACTOR . .... ...... . . . . . . . . . . . . . . . . . 4-1 4.4 Thermal and Hydraulic Design . . . . . . . . . . . . . . 4-1.
' 4.4.1 Evaluation. . . . . .. . . . . . . . . . . . . . 4-1 6 ENGINEERED SAFETY FEATURES. . . . . . . . . . . . . . . . . . . 6-1 6.2 Containment Systems. . . . . . . . . . . . . . . . . . . '6 6.2.5 Combustible Gas Control . . . . . . . . . . . . . 1
.Clinton SSER 3' y
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'f, 6. 3 Emergency Core Cooling Syst,em. . . . . . . . . . . . . . 6-1
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6.3.4' Performance Evaluation. . . . . . . . . . . ... 6-1 1 c
\ 9-1 9 AUXILIARY,iSYSTEM,S -
' 9.5 , Fire Protection Systems. . . . . . . . . . . . . . . . . 9-1 o
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' 9!5.1 De'scription and Evaluation. . . . . . . . . ...
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1 9.-5.1.4 ' Fire'Det'ection Systems'. . . .'. . . . . 9-1
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. 9.5.2 '.'OthdOtems Related to ,Fity Protection. . . ... 9-1 V.915:2.1 Fire Barriers and Penetrations . . . . . 9 i 9.5.3 Emergency Lighting. . . . . . .......... 9-2 2 9.5.5 Fire Protection of Safe Shutdown Capabilitj . . . 9 9.5.9 Conclusion. . . . . . . . . . .......... 9-4 12 RADIATI,0N FROTECTION . ....,i.'.,.......c. . ... 12-1
- , a 4 i r 12.5 12-1 Operptional. Radiation, Protectiot
- Program. > .e . . . . ...
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. . , ..,....... 12-1 17 QUALITY ASSURANCE. .f.'.
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?......... 17-1 17.2 Organization. . . . . . . . . . , , , . ......... 17 17.4 Conclusi'oh ' . . . . . . . . . . b . q ......... 17-2 1
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/ APPENDIX A . CONT'INUATION OF CHRONOLOGY ,
, , ' APPENDIX B . REFERENCES- .
, r APPENDIX D NRCISTAFF CONTRIBUTORS AND CONSUL
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1 INTRODUCTION AND GENERAL DESCRIPTION 1.1 Introduction The Nuclear Regulatory Commission staff (referred to as the NRC staff or staff) issued its Safety Evaluation Report (SER) (NUREG-0853) in February 1982 regard-ing the application by Illinois Power Company, et al. (hereinafter referred to as the applicant) for a license to operate the Clinton Power Station, Unit 1, Docket No. 50-461. Supplements No. 1 and No. 2 to the Safety Evaluation Report (SSER 1 and SSER 2) were issued in July 1982 and May 1983, respectively. The purpose of this supplement, No. 3 (SSER 3), is to further update the SER by providing results of the NRC staff's review of information submitted by the applicant to address some of the unresolved issues listed in Sections 1.9 and 1.10 of the SER.
Each section and appendix of this supplement is numbered and titled so that it corresponds to the section or appendix of the SER that is relevant to the NRC staff's additional evaluation. Except where specifically noted, the material in this supplement does not replace the corresponding SER section or appendix.
Appendix A is a continuation of the chronology of correspondence between NRC and the applicant and updates the list in the SER, SSER 1, and SSER 2. Appen-dix B is a list of references cited in this report.* Appendix D is a list of principal staff contributors and consultants to this supplement.
Copies of this SER supplement are available for inspection at the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C. and at the Warner Vespasian Library, Clinton, Illinois. Copies are also available for purchase from the sources indicated an the ' iide front cover.
The NRC Project Manager ,o the operating license application for Clinton is Gregory A ,. . , P . E. Mr. Harrison may be contacted by calling (301) 492-8344 or a Mr. Gregory A .ison, P.E.
Division of Licensing,l tail Stop 144 U.S. Nuclear Regulatory Commission Washington, D.C. 20555
- 1. 9 Outstanding Issues i
In SER Section 1.9, the NRC staff identified 20 outstanding issues that had not been resolved at the time the document was issued. SSER 1 reported that four of those items had been satisfactorily resolved and one had been changed to a
- The availability of the material cited is described on the inside front cover of this report.
Clinton SSER 3 1-1
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s confirmatory status.* SSER 2 reported that six items have either been resolved or changed to a confirmatory status. Therefore, there were nine outstanding issues that had not yet been resolved after the' issuance of SSER 2.
This supplement (SSER 3) closes..four outstanding issues,- as well as nine con-firmatcry issues. The current tatus of-each of the 20 original issues is tabulated below. For those items discussed in'this supplement, the relevant sections of this document are indicated, Resolution of issues that are, to date, unresolved will be reported in futdrp, supplements.
Issue Sk.atus, ,
Section(s)
A .
I -(2) Transportation accidents Resolved this SSER 2.2.1 sc ,
(2) Effects of Unit 2 excavation Resolved SSER 2 -
,3 (3) Seismicanalysisl ,
, Became confirmatory 2.5.2.4, t 4 ,
issue (70), resolved 3.7 I
( this SSER (4) Internally generated missiles Resolved SSER 1- -
(5) Postulated piping failures Under review -
(6) Steady-state vibra'tion' ,
Resolved SSER 2 --
acceptance criteria fnr .
' balance of plant piping; (7) Environmental and seismic Under review --
qualification test programs (8) Preservice (PSI) and inservicel PSI program: became --
inspection (ISI) programs confirmatory issue (67)
< SSER 1 ISI program: became license condition (12)
- SSER 2 .
i (8a) Preservice and inservice Became confirmatory --
testingofpumpsandvalqs issae (68) SSER 1 (9) Pool dynamic loads ,
Under' review --
~i0a)Containmentpurge
( Became confirmatory --
issue (69) SSER 2 (10b) Conta nNent isolation Resolved SSER 2 --
, (10c) Contair.mant leakage testing Resolved SSER 2 --
(vent and drain lines) h, ,
- SSER 2 stated that only three items had been closed and it was silent regarding confirmatory status.
Clinton SSER 3 1-2
Issue Status Section(s)
(10d) Containment leakage testing Resolved SSER 2 --
(10e) Containment bypass leakage Resolved SSER 2 --
(11) Control room habitability Resolved SSER 1 --
(12) Engineered safety features Resolved SSER 2 --
reset controls (IE Bulletin 80-06)
(13) Remote shutdown system Resolved this SSER 9.5 (14) Capability for safe shutdown Resolved SSER 2 --
following loss of bus supply-ing power to instruments and controls (IE Bulletin 79-27)
(15) Control system failures Under review --
resulting from high energy-line breaks or common power source or sensor malfunctions (16) Separation of the RPS and MSIV Resolved SSER 1 --
solenoid circuits and PGCC circuits (17) Organization and staffing Under review --
(18) Emergency Plan Under review --
(19) Security Resolved SSER 1 --
(20) QA Program Resolved this SSER 17.2, 17.4 1.10 Confirmatory Issues In SER Section 1.10, the NRC staff identified 66 confirmatory issues for which additional information and documentation were required to confirm preliminary conclusions. SSER 1 reported that 28 of those items had been satisfactorily resolved. SSER 2 addressed 11 additional issues that have been resolved, as well as.certain issues that still require resolution. The current status of each of-the 66 original issues is tabulated below. Four issues (67, 68, 69, !
and 70) that previously had been outstanding issues in SSER 1 were added to the '
confirmatory list in SSER 2. Resolution of confirmatory issues that are, to date, unresolved will be reported in future supplements.
Issue ' Status Section(s)
(1) Emergency preparedness Under review - 2.3.3 meteorological program Section 2.3.3 updated ,
I Clinton SSER 3 1-3
- Issue Status Section(s)
(2) Inspection program around the Resolved SSER 1 --
ultimate heat sink (VHS) and the main cooling lake dam (3) Protection of UHS dam abutments Resolved SSER 1 --
against soil erosion (4) Internally generated missiles - Resolved SSER 2 fan failures (5) Design adequacy of cable tray Resolved SSER 1 --
system (6) Containment ultimate strength Under review --
analysis (7) Structural integrity of safety- , Resolved SSER 2 --
related masonry walls (8) NSSS pipe break analysis using Resolved SSER 1 --
SRP criteria (9) Vibration assessment of RPV Under review --
internals (10) Annulus pressurization loads Under review --
(LOCA asymmetric loads)
(11) Use of SRSS for combining Resolved SSER 1 --
Mark III dynamic responses for other than LOCA and SSE (12) IE Bulletin 79-02 regarding Resolved SSER 2 --
support baseplate flexibility (13) Mark III hydrodynamic loads Under review- --
(14) Feedwater check valve analysis Resolved SSER 2 --
(15) Seismic and LOCA loadings on Awaiting information --
fuel assemblies (16) Scram discharge system Resolved SSER 1 --
evaluation (17) Fracture. toughness data Resolved SSER 1 --
(18) Subcompartment pressure Under review --
analysis
!- (19) Combustible gas control Resolved this SSER 6.2.5 Clinton SSER 3 1-4
Issue States Section(s)
(20) Containment isolation Resolved SSER 2 --
dependability (21) Containment monitoring, Awaiting information --
II.F.1 (22) Plant-specific LOCA analysis, Resolved this SSER 6.3.4 II.K.3.31 (23) High drywell pressure Resolved SSER 1 --
interlocks (24) ATWS recirculation pump trip Awaiting information --
(25) Response-time testing Resolved SSER 1 --
(26) Analog trip modules and optical Resolved SSER 2 --
isolators (27) Susceptibility of the NSPS to Resolved SSER 1 --
electrical noise (28) Modification of ADS logic, Under review --
II.K.3.18 (29) Restart of low pressure Resolved SSER 1 --
systems, II.K.3.21 (30) Temperature effects on level Resolved SSER 2 --
measurements (31) Containment atmosphere Under review- --
monitoring system (32) Verification that testing is Removed from list SSER 1 --
in accordance with BTP PSB-1 (33) Electrical drawing review Removed from list SSER 1 --
(34) Verification of diesel Position statement --
generator testing sent to applicant-(35) Class A supervision and power Resolved this SSER 9.5.1.4 supply for fire detection system (36) Circulating water system Resolved SSER 2 --
(37) Initial test program 'Resolvad SER --
(38) Human engineering aspects of Under review- --
control room design, I.D.1 l 1
Clinton-SSER 3 1-5
Issue Status Section(s) I l
(39) Common reference for reactor Resolved SSER 2 --
vessel level instruments, II.K.3.27 (40) Shielding design review, Resolved SSER 1 --
11.8.2 (41) Short-term accident and Awaiting information --
procedures review, I.C.1, I.C.7, I.C.8 (42) Training during low power Awaiting information --
testing, I.G.1 (43) Review ESF values, II.K.1.5 Resolved SSER 1 --
(44) Operability status, II.K.1.10 Resolved SSER 1 --
(45) HPCI and RCIC initiation Under review --
levels, II.K.3.13 (46) Isolation of HPCI and RCIC, Under review --
II.K.3.15 i
(47) Qualification of ADS Under review --
accumulators, II.K.3.28 (48) Plant-specific analysis, Resolved this SSER 6.3.4 II.K.3.30 (49) ODYN analysis for River Bend Resolved SSER 1 --
as applied to Clinton (50) Conformance evaluation report Resolved this SSER 4.4.1 l for loose parts monitoring l system I
(51) Requirements of NUREG-0313 Resolved SSER 1- --
(52) Control room habitability - Resolved SSER 1 --
chlorine gas (53) Debris screen design Resolved SSER 2 --
l (54) Verification of adequacy of Removed from list SSER 1 --
! fire protection systems (55) Flood proof door Resolveo SSER 2 --
(56) Valves in fire protection Resolved SSER 1 --
water supply system Clinton SSER 3 1-6
-Issue- Status Section(s)
(57) Break in water supply piping Resolved SSER 1 --
(58) Test data on fire ratings Resolved this SSER 9.5.2.1 (59) Three-hour-fire-rated Resolved this SSER 9.5.2.1 penetration seals (60) Install fire protectior Resolved this SSER 9.5.3 equipment (emergency lighting)
(61) Fire protection administrative Resolved SSER 1 --
controls and training (62) Technical Specification on Resolved SSER 1 --
fire protection (63) Periodic leak testing of Resolved SSER 1 --
pressure isolation values (64) Sedimentation in UHS Resolved SSER 1 --
(65) Protection against postulated Resolved SSER 1 --
piping failures (66) Steam bypass of the Under rev'aw --
suppression pool (67) Preservice inspection program Under. review --
(68) Preservice testing of pumps Under review --
and valves (69) Containment purge Under review --
(70) Seismic analysis Resolved this SSER 2.5.2.4, 3.7 (71) Humphrey concerns Under review --
j 1.11 License Conditions In SER Section 1.11, the NRC staff identified nine potential license conditions that may be required as part of the operating license for Clinton, Unit 1, to ensure that NRC requirements are met during plant operations. Two additional potential license conditions (10 and 11) were identified in SSER 1, and SSER 2 identified two additional conditions (12 and 13), as well as one (6) for which additional requirements were imposed. One condition (14) has been added in this supplement. These are tabulated below, with the relevant section(s) of this report noted.
1 Clinton SSER 3 1-7
Issue- Section(sj (1) Staffing DeWitt pumping station --
(2) New stability analysis before second cycle of operation --
-(3) Postaccident monitoring --
(4) Vacuum relief valve position indication (5) Hydrogen management --
1 (6) Post-accident sampling, II.B.3
! (7)- Diesel generator reliability (8) Kuosheng 1 test program --
(9) Visual examination of discharged fuel (10) Measurement of groundwater level --
(11) Security --
(12) Inservice inspection --
(13) Control of heavy loads --
l 2.2.1 (14) Transportation accidents J
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l Clinton SSER 3 : 1-8
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2 - SITE CHARACTERISTICS i
2.2 Nearby Industrial, Transportation, and Military Facilities 4
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2.2.1 _ Transportation Routes In the SER, the review of the Clinton Power Station Final Safety Analysis Report identified two hazard categories associated with postulated accidents on the Illinois Central Gulf Railroad near the Clinton Power Station: (1) ship-ments of flammable compressed gases and (*) shipments of toxic materials. The latter impacts the review and acceptability of the control room habitability systems. Since the issuance of the SER, the applicant has submitted additional information by letters dated March 4, and August 26, 1983.
I The Illinois Central Gulf Railroad runs parallel to State Route 54 and traverses
, the site approximately 0.75 mile north of the station. Originally, the appli-cant had conducted a survey of hazardous materials transported by the railroad.
1 This survey provided the applicant a basis for determining that the railroad did not pose a significant threat to the plant.
4 i Because of the lapse of time since the first survey, the applicant, at,the
- staff's request, conducted a second survey and determined that a large number 4
of railroad cars per year traveled over the Illinois Central Gulf Railroad line, i The revised traffic level is considerably larger than what was determined during the first survey in terms of numbers and types of hazardous materials. The applicant has provided a conservative analysis that concludes that the-annual l probability of affecting plant safety significantly is about 10 6 for toxic
- materials and about 10 7 for explosive materials.
The staff agrees with this finding. however, in light of the potential for-
- significant changes in the hazardous material shipping patterns for this rail-road, the Staff recommends a license condition that would require a periodic survey (every 3 years) of all shipments of hazardous material and a reassess-ment of the hazards associated with toxic material with respect to the rail-road. With regard to toxic materials, the applicant shall (within 6 months of
' a negative finding) provide sensors that will isolate the control room heating, ventilation, and air conditioning system should identified chemicals present a toxic environment. The initial report shall be submitted to the NRC not later than July 1,1986. The shipping data shall be based on actual records of ship-i ments made during the calendar year 1985. This report shall be updated every 3 years.
9 With respect to explosive materials, only flammable liquids-(such as liquefied petroleum gas) have any potential for being an overpressure hazard with respect
! to the plant. Localized _ explosions (e.g., solid explosives) on the rail line do not present a hazard to the plant because of the separation distance between
-the plant and the' railroad. However, accidents involving flammable liquids can lead to spillage and evaporation,'and, if ignition is delayed, a cloud-of
- flammable vapor can be postulated to drift toward the plant. In the analysis of the hazards _ associated with flammable liquid spills, the applicant used a-
^
Clinton SSER'3 2-1
number of significant conservatisms: an overpressure threshold of 1 psi, although the plant is designed for at least 3 psi (because of design criteria with respect to tornado effects); a railroad tankcar lading loss rate idealized to maximize the extent of flammable concentrations downwind of the spill; and a conservative meteorology that does not include the thermal buoyancy effects of the cooling lake.
The present shipping frequency of flammable material is such that an increase in traffic level of an order of magnitude or more would be required before a more detailed analysis of the hazard would be needed. In view of the above considerations and conservatisms, the staff believes that the risk from explo-sives is acceptably low and is not expected to become significant as a result of annual rail traffic variations.
With the above provisions the staff believes that the hazards at the Clinton Power Station site have been addressed adequately and considers this open item resolved.
2.3 Meteorology 2.3.3 Onsite Meteorological Measurements Program In SER Section 2.3.3.5, the NRC staff stated that a detailed review of the emergency preparedness meteorological program would be included in a supplement to the SER.
Since the issuance of SSER 2, the applicant has submitted additional information by letter dated September 6, 1983. The NRC staff has reviewed the applicant's analysis and concludes that it appears to be adequate, but it will perform a final review during the scheduled onsite emergency preparedness appraisal.
The emergency preparedness meteorological program produces measurements from the primary meteorology tower of wind speed and direction at the 10-meter level and an estimator of atmospheric stability in accordance with the guidance in Regulatory 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 along with the meteorology data can be remotely determined from the emergency operations facility and will be available to offsite agencies through remote interrogation.
If the onsite meteorology data are unavailable, data from National Weather Service offices at Springfield and Peoria, Illinois, will be used to charac-terize 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.
Clinton SSER 3 2-2
... .. . - = . ,,. -- - . . - - -- -
'2.5' Geology and Seismology e 2;5.2 Seismology 2.5.2.4 Safe Shutdown Earthquake (SSE)
In Section 2.5.2.4 of SSER 2, the NRC staff concluded that the most recent ver-
- sion'of the Clinton soil amplification analysis (letter from applicant dated 5
April 18 1983),'which used the upper-bound soil properties published in the
-Final Safety Analysis Report (FSAR) together with a suite of 14 earthquake time histories, provided an adequate site-specific response spectrum to evaluate the seismic design and to demonstrate the adequacy of the earthquake design spec-
!.. trum. This issue was identified in SSER 2 as confirmatory issue 0 0). .Since
! the issuance of SSER 2, the NRC staff has received additional information, met '
- with the applicant, and reviewed certain statistical analyses studies pertaining
! to the applicant's seismic design criteria. The NRC staff's detailed review and resolution of the soil amplification issue now follows.
t The use of site-specific spectra to define the seismic design criteria for j Category I structures has become an accepted. practice in the review of. Safety
- Analysis Reports for nuclear plants (Sequoyah, NUREG-00ll, 1979; Fermi, i NUREG-0798, 1981; Midland, NUREG-0793, 1982; Wolf Creek, NUREG-0881, 1982; Perry, NUREG-0887, 1982). Site-specific spectra are derived from earthquake strong
,t- motion' records. These data can be classified by the magnitudes and distance of I the earthquakes recorded. In addition, the recording stations may be classified
- by the geophysical properties of the underlying rock / soil columns.
1 . :
1 This information allows for the selection of-records of earthquakes of appro-priate energy content (as opposed to having to rely on the more obscure inten-2 sity data) and for optimal representation of actual ground motions generated by .
- the postulated site event by matching the geophysical properties of the record-j ing stations to those of the-site. A disadvantage of this procedure is that'the data base of strong motion records is relatively small.because both the strong
~
motion recording instrumentation'and.the necessary data processing technic.ues
- are fairly 'recent developments in the field of seismology.
I
! As a result, the site-specific spectra that can be.obtained from these records are limited in scope'with' respect-to geologic l site conditions and' magnitude j values. Thus, in cases where the nuclear plant site in question has a geologic profile that cannot'be' matched satisfactorily by records from like recording stations, as is the case with the'Clinton site, spectra have to be augmented.to
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- compensate for this defici.ency. At the Clinton site it was a contrast in shear velocities of a rock / soil interface at'a depth of 200 feet, which.could not be
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matched satisfactorily because=similar geologic conditions could not be found' ..
I among the recording stations from which recordings were available. l 1
For this reason, the' applicant's'ubmitted a study?in a' letter dated May 24,
. 1982, which contained a comparison between the Clinton design-basis; spectra and the site-specific' spectra obtained from 21 strong motion records that best:
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'repre'sented the desired site conditions. The. submittal alsolincluded'a theo-retical soll amplification study assessing the influence of .the 200-foot-depth
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-shear. wave velocity contrast;;however, the method used to calculate,the ampli-fication factors and:the weighting procedure to arrive at an average amplifi-ication were;not adequately discussed. Therefore, fat.the NRC staff's request,
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-Clinton SSER'3 -
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the applicant submitted a report in a letter dated August 26, 1982, in which a detailed presentation was made that also involved comparative results from a computer analysis program (SHAKE). The NRC staff concluded that the results of the two soil analysis methods were sufficiently similar so that the applicant's original analysis could be considered acceptable; however, no documentation was .
presented to justify the weighting procedure, which consisted of assigning a )
weight of 0.25 to the amplification factors obtained from analyses of both the I upper-bound and lower-bound soil properties and a weight of 0.50 to the ampli- )
-fication factors of the mean soil properties. The final amplification factor was determined by calculating the arithmetic average of the three (weighted) sets of amplification factors.
An evaluation of the weighting procedure, conducted by the NRC staff, showed that this method tended to suppress the peak amplification beyond a level which the staf f considered acceptable without further substantiation. As a result of the staff's concerns expressed in a letter dated January 12, 1983, the appli-cant undertook a statistical analysis. The applicant submitted this analysis to the NRC staff by letter dated March 10, 1983. This analysis took into con-sideration random variations in the soil shear moduli and earthquake-induced excitation; however, this analysis in its existing form was generally inade-quate because the randomly chosen shear moduli were not representative of the soil parameters acceptable to the NRC staff and the statistical analysis did not justify the weighting procedure, nor was any other justification presented.
At the staff's request, the applicant, by letter dated April 18, 1983, submitted modified versions of the statistical analysis. The site-specific spectrum that was accepted by the staff (Figure 2.1) was obtained by evaluating the amount of necessary conservatism in each of the parameters influencing the seismic design.
Significant features of the analysis are:
(1) The soil amplification factors were obtained by using the SHAKE soil analysis computer program for layered media. The input parameters to the SHAKE computer program were the soil parameters for each layer of soil and recorded earthquake time histories.
(2) The soil parameters used were the upper-bound soil parameters listed in Table I in the applicant's letter dated April 5, 1983. These (upper-bound) soil properties were considered acceptable for evaluating the amplification potential because they are representative of the typical values for this type of soil (glacial till).
To obtain the amplification factors for the Clinton site, the ratios of the in-put motion (assumed to be at rock interface) to the output motion (assumed to be at plant foundation level) were determined statistically by using 14 earth-quake motion time histories, listed in Table II in the applicant's letter dated April 5, 1983. The staff concluded that the input motions used are sufficiently similar to site-specific SSE rock motions so that the soil amplification analy-sis can be considered acceptable.
To obtain a soil site-specific spectrum for the Clinton site, the applicant computed an average rock site-specific spectrum from two rock spectra, one
! developed by Lawrence Livermore National Laboratory (LLNL) (NUREG/CR-1582) and I one developed by the Tennessee Valley Authority (TVA) (August 1978). The rock l spectra and the amplification factors were combined to obtain an 84th percentile l-soil site-specific spectrum. In evaluating the resulting 84th percentile l
Clinton SSER 3 2-4
^
i amplified site-specific spectrum, shown in Figure 2.1, the NRC staff noted that the amplified site-specific spectrum is enveloped by the Clinton design-basis spectrum except for a portion in the 1.1- to 1.7-hertz range. However, the applicant lerified that no critical structural frequencies fell within this frequency range (see appropriate discussion in Section 3.7).
2.5.2.6 Summary The applicant generated a site-specific spectrum from strong motion recordings that best prese.nted the anticipated ground motion at the site from an earthquake of magnitude 5.8 (mb ) and demonstrated that the Clinton seismic design spectrum enveloped the site-specific spectrum for the Clinton SSE. After reviewing the applicant's material, the NRC staff requested the applicant to evaluate the potential soil amplification that could occur at the site as a result of a geo-logic feature that was not represented in the site-specific spectrum analysis.
The applicant performed a statistical analysis to evaluate the soil amplifica-tion at the site. The statistical analysis was accepted by the NRC staff after several modifications that led to improvement in the site-specific spectrum.
The significant features of the applicant's study that led to a favorable com-parison of the Clinton seismic design spectrum and the postulated (amplified) site specific spectrum are as follows:
(1) The 84th percentile spectrum, obtained by averaging the LLNL rock spectrum and the TVA rock spectrum, was considered conservative.
(2) The methods used to combine the rock spectrum and the amplification fac-tors to obtain the statistical 84th percentile site-specific spectrum were acceptable.
(3) The Clinton seismic design-basis spectrum envelopes the (amplified) site -
specific spectrum except in the small interval between the frequencies of 1.1 hertz and 1.7 hertz. (See also discussion in Section 3.7.)
In view of the above analysis and the evidence submitted by the applicant, the NRC staff concludes that the Clinton seismic design criteria are acceptable.
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Clinton SSER 3 2-5
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Spectrum for Upper bound Soil Properties Considering _
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10 2 4 8 10' 10 10 PERIOD (sec) l Figure 2.1 Comparison of design-basis time history spectrum to i site-specific spectra with and without the effects of a l
soil shear velocity contrast at 200-foot depth Source: Letter No. U-0633 from G. E. Wuller (Illinois Power Company) to A. Schwencer (NRC) dated April 18, 1983.
Clinton SSER 3 2-6
3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.7 Seismic Design 3.7.1 Seismic Input and 3.7.2 Seismic System Analysis In Section 3.7.1 of SSER 1, the NRC staff indicated that the applicant was making use of site-specific spectra to resolve two issue:: (1) the adequacy of the seismic input motion in the free field at the foundation level and .
(2) compliance with the NRC staff's position on the soil-structure interaction analysis. The staff further indicated in SSER 1 that the site-specific spectra were roughly equivalent to a seismic event of 0.20 g anchored to the Regulatory Guide (RG) 1.60 spectra at the foundation level in the free field. The appli-cant performed a seismic reevaluation of plant-complex structures and the cir-culating water screenhouse using the synthetic time history whose spectra enve-lope the RG 1.60 spectra anchored at 0.2 g. In SSER 1, the NRC staff concluded that both of the above issues t -re resolved contingent on the acceptance of the site-specific spectra. Since then, as discussed in detail in Section 2.5.2 of this supplement to the SER, the NRC staff has completed its review of the site-specific spectra.
The major conclusion reached is that the site-specific spectra accepted by the staff are essentially enveloped by the spectra of the synthetic time history used by the applicant in his reevaluation of plant structures; however, the accept, o site-specific spectra do exceed the synthetic time-history spectra in the narrow frequency range of 1.1 hertz to 1.7 hertz.
By a letter dated April 18, 1983, the applicant has confirmed that the plant structures do not have natural frequencies between the range of 1.1 hertz to 1.7 hertz. In fact, the fundamental frequencies of the plant structures are greater than 1.7 hertz. This being the case, the above discussed exceedance in the narrow frequency range of 1.1 hertz to 1.7 hertz will have no impact on the plant structural design.
Therefore, the staff confirms its findings'of SSER 1 that (1) the stresses in the plant structures are within design-basis allowable limits when the allow-able limits are based on the actual mean yield strength derived from field test reports and (2) the applicant has complied with the staff's position on soil-structure interaction analysis. The staff considers both issues fully resolved.
Clinton SSER 3 3-1
l
' 4 REACTOR 4.4 Thermal and Hydraulic Design 4.4.1 Evaluation In the SER, the NRC staff indicated that the applicant will have a loose parts monitoring system (LPMS) operational at the time of initial reactor startup testing. The applicant also had committed to evaluate the system's conformance with Regulatory Guide (RG) 1.133, Revision 1 (May 1981). The staff found the LPMS to be acceptable subject to the review of the LPMS against RG 1.133. This issue was identified as confirmatory issue (50). Since the issuance of SSER 2, the applicant has submitted additional information regarding the LPMS and met with the staff, which has now concluded its review.
4 At the staff's request, the applicant, by Amendment 26 (July 1983) and Amend-ment 27 (September 1983), submitted additional information regarding the LPMS and program. The NRC staff has reviewed these submittals related to the LPMS and has met with the applicant. The applicant's evaluation results indicated that the LPMS complied with RG 1.133 except for Paragraph C.1.g " Operability for Seismic and Environmental Conditions." Paragraph C.1.g requires that com-ponents of the LPMS within containment should be designed and installed to per-form their function following seismic events that do not require plant shutdown.
However, the staff. determined that the applicant's submittal also indicated that all of the LPMS equipment inside containment is seismically qualified. The staff, therefore, has concluded that the applicant's LPMS is in compliance with the guidance stated in Paragraph C.1.g. On the basis of its evaluation of the LPMS against RG 1.133, the staff has concluded that the Clinton LPMS is accept-able. The Clinton Technical Specifications will include appropriate limiting conditions for operation and surveillance requirements to demonstrate the operability of LPMS channels.
f Clinton SSER 3 4-1
. - . . - . ~ . - - - . - - . _ ~ .. -- .- - - -
6 ENGINEERED SAFETY FEATURES 6.2 : Containment Systems
}- 6.2.5 Combustible Gas Control 1
) In the SER, the NRC staff presented its review of the applicant's combustible .
j gas control systems. It found that the applicant complied with appropriate
{ design criteria, except that the selection of hydrogen sample point locations ,
! was not complete. This issue became confirmatory issue (19). Since the issu-ance of the SER, the applicant has completed this selection of the hydrogen 4
sample point locations. The hydrogen sampling points are located in both the drywell and containment to allow detection of any nonuniform hydrogen concen-l trations. The drywell and containment are divided into five zones with two j hydrogen sample points in each zone. The staff has reviewed the applicant's i design and finds that the proposed hydrogen sample point locations are 4 acceptable.
.l
! 6.3 Emergency Core Cooling System 6.3.4 Performance Evaluation j In the SER, the NRC staff noted that the applicant's FSAR loss-of-coolant-
- accident (LOCA) calculations had been performed for a lead plant similar to j Clinton. The applicant committed to supply plant-specific LOCA analyses and i did so in Ar.
- endment 23. The staff reviewed this amendment and concluded that I
the LOCA analyses were acceptable. However, regarding.small-break LOCA methods (TMI Action Plan Item II.K.3.30), the applicant, at the staff's request, pro-j vided information to the NRC at a meeting held on June 18, 1981, which was sub-sequently docketed in a letter dated June 26, 1981. .The NRC staff has now i
evaluated this information relative to confirmatory issues (22) and (48).
l The staff's evaluation and review concluded that the test data comparisons and Other information acceptably demonstrate that the existing General Electric-
! small-break model is in compliance with Part 50 of Title 10 of the Code of
! Federal Regulations (10 CFR 50), Appendix K, and that, therefore, no model i changes are required. In addition, because of the acceptable resolution of TMI I Action Plan Item II.K.3.30 (confirmatory issue (22)), additional small-break I
LOCA analyses are not required for TMI Action Plan Item II.K.3.31 (confirmatory
! issue (48)). On the basis of its review, the staff concludes that the appli-l cant's submittals regarding plant-specific LOCA analyses (TMI Action Plan Items II.K.3.30 and II.K.3.31) have been fully resolved and,-.therefore, are
- acceptable.
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V Clinton SSER 3 6-l' 4
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9 AUXILIARY SYSTEMS 9.5 Fire Protection Systems 9.5.1 Description and. Evaluation 9.5.1.4 Fire Detection Systems In the SER, the NRC staff was concerned that the circuits running between each local supervisory control panel and the fire alarm indicator in the main con-trol room were not supervised and that adequate primary and secondary power supplies had not be provided for the detection systems. This issue was identi-fied as confirmatory issue (35). Since the issuance of the SER, the applicant, by letter dated April 29, 1982, has committed to install redundant fire alarm circuits for all circuits running between local supervisory fire alarm control panels and the main fire alarm indicator in buildings that contain safety-related equipment. One circuit will be supervised so that a single break or ground fault condition on the circuit will be indicated at the main fire alarm indicator in the control room. The redundant circuit will provide alarm indi-cation at the main fire alarm indicator if the supervised circuit becomes dis-abled.
The staff finds that this commitmen't provides the equivalent of a Class A supervised circuit running between all local control par.els in buildings con-taining safety-related equipment and the main fire alarm ii.dicator located in the control room as required by National Fire Protection Association (NFPA) 720.
The staff finds this acceptable.
In Amendment 14, the applicant committed to provide both primary and secondary power supplies for the fire detection systems in accordance with staff guide-lines. The staff find this acceptable.
On the basis of the applicant's commitments, the staff concludes that the fire detection systems meet the guidelines of Branch Technical Postion (BTP)
CMEB 9.5-1 Section C.6.a, and are, therefore, acceptable. This resolves con-firmatory issue (35).
9.5.2 Other Items Related to Fire Protection
- 9. 5. 2.1 Fire Barriers and Penetrations In the SER, the NRC staff indicated that the applicant had verbally committed to provide verification to substantiate the fire rating of both fire barriers and penetration seals. By letters dated February 3, March 25, and April 4, 1983, the applicant provided the information. The information included the manufacturer's verification that the fire rated barriers and penetration seals have been tested in accordance with American Society for Testing _and Materials (ASTM) E-119 for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The staff has reviewed the test results and finds that the fire-rated barriers and penetration seals used are qualified for a Clinton SSER 3 9-1
fire resistance of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. On the basis of this information, the staff con-cludes that fire-rated barriers and fire-rated penetration seals meet the guidelines of BTP CMEB 9.5-1, Section C.5.a, and therefore, are acceptable.
This resc'ves confirmatory issue (58).
9.5.3 Emergency Lighting In the Fire Hazards Analysis and FSAR dated April 1978, the applicant committed to provide 8-hour battery pack emergency lighting units in all areas of the plant needed for operation of safe shutdown equipment and in access and egress routes thereto. In the SER, the staff accepted this commitment as meeting its guidelines. However, by letter clated December 16, 1982, the applicant indicated that the SER evaluation was incorrect and that emergency lighting was not in accordance with Section C.S.g of BTP CMEB 9.5-1, which requires that emergency lighting be provirted for shutdown following a fire. The applicant did not pro-vide a justification for this deviation from staff guidelines. Subsequently, the staff stated that it will require the applicant to meet the commitment made in the Fire Hazards Analysis and FSAR.
By letter dated September 8, 1983, the applicant stated that emergency lighting for safe shutdown would be provided in accordance with the commitments made in the Fire Hazards Analysis and FSAR.
On the basis of this commitment, the staff finds that the emergency lighting meets the guidelines of BTP CMEB 9.5-1, Section C.S.g, and is, therefore, acceptable. This resolves confirmatory issue (60).
9.5.5 Fire Protection of Safe Shutdown Capability In the SER, the NRC staff indicated that the applicant had not provided an analysis of his safe shutdown capability. The staff also indicated that the applicant would be required to meet the fire protection technical requirements for safe shutdown contained in Sections III.G and III.L of Appendix R to 10 CFR 50. Since the issuance of the SER, the applicant, by letter dated December 16, 1982, has provided the " Safe Shutdown Analysis for Fire Protection" for Clinton Unit 1. Additional information concerning the safe shutdown analy-sis was also provided by a letter dated April 5, 1983.
The applicant's safe shutdown analysis indentified the redundant systems avail-able for achieving hot shutdown and cold shutdown. For hot shutdown, either the reactor core isolation cooling system, safety / relief valves, and the sup-pression pool cooling mode of the residual heat removal (RHR) system or the low pressure coolant injection mode of the RHR system, pressure relief system (automatic depressurization system), and the suppression pool cooling mode of the RHR system would be available. For cold shutdown, redundant trains of the shutdown cooling mode of the RHR system would be available. The safe shutdown analysis considered components, cabling, and support equipment for systems identified above that are needed to achieve shutdown. The support equipment includes the shutdown service water system, the diesel generators, and the heating, ventilation, and air conditioning (HVAC) systems for the emergency core cooling system equipment rooms, the shutdown service water pump rooms, the diesel generator rooms, and the essential switchgear rooms. The applicant's analysis indicated that one of the redundant systems needed for safe shutdown would be free of fire damage because of separation, fire barriers, and/or Clinton SSER 3 9-2
1 alternative shutdown capability, with the exception of one fire area. For the auxiliary building division 1 switchgear area, the applicant will use repair 1 procedures for the division 2 RHR valve cables to achieve cold shutdown. The I applicant Ns committed to develop procedures for the repairs, and any materials 1 needed for these repairs will be stored on site.
The applicant's safe shutdown analysis used cable installation drawings. Safe shutdown equipment cabling was identified and traced through each fire area from the components to the power sources. Additionally, cables whose fire-induced spurious operation would affect. shutdown were identified and traced through each fire area. All power and control circuits are protected by coordinated circuit breckers and fuses. During power operation, the power to one of the redundant shutoown supply valves will be locked open. The analysis identified where re-dundant trains (systems) of safe shutdown systems are located within the same fire area. Corrective measures are being taken as necessary to ensure proper separation and/or barriers. The staff has reviewed the applicant's method of analysis and has audited several arrangement drawings to verify correct appli-cation of the methodology. The staff concludes that the applicant has provided an acceptable means of demonstrating that separation and/or barrier (s) exist between redundant safe shutdown system trains.
The applicant's analysis indicated that alternative shutdown measures were re-quired for the control room, the control room HVAC equipment room, and some adjoining areas to the control room and its HVAC equipment room to ensure the availability of the safe shutdown systems. If a fire should disable the equip-ment in these areas, the remote shutdown panel located in the division 1 auxil-iary equipment room in the auxiliary building would provide an alternative to fire protection separation (see Section 7.4.1 of the SER). The control func-tions and indications provided at the local panel that are necessary for safe shutdown are electrically isolated or otherwise separate and independent froi the control room.
In the analysis, the applicant requested four deviations from Section III.G.2 of Appendix R to 10 CFR 50 (Section C.5.b of BTP CMEB 9.5-1). The applicant has requested deviations from providing automatic fire suppression systems in the following areas:
(1) fire zone A.2.1, auxiliary building, el 737 feet (2) fire zone D.2.1, diesel generator building, el 737 feet (3) fire zone 0.3.2, diesel generator building, el 762 feet (4) fire zone F.3.1, fuel building, el 855 feet Each of the above areas contains early warning fire detection. Redundant safe shutdown cable and equipment are separated by more than 20 feet. In addition, the fuel load in each of the above areas is low to moderate. Because the in situ fuel loads are low, the redundant systems have adequate separation, and early warning fire detection is provided, the staff has reasonable assurance that, after a postulated fire in any of these areas, one train of safe shutdown systems will be free of fire damage.
On the basis of its evaluation, the staff concludes that the installation of automatic fire suppression systems in fire zones A.2.1, 0.2.1, D.3.2, and F.3.1 would not significantly increase the level of fire safety. The staff, there-fore, finds that the deletion of the automatic fire suppression systems is an Clinton SSER 3 9-3 L
1 acceptable deviation from Section III.G.2 of Appendix R to 10 CFR 50 (Sec-tion C.5.b of BTP CMEB 9.5.1). Therefore, the fire protection provided for these areas is' acceptable.
On the basis of the above considerations, the systems to be used for safe shut-down during a fire are acceptable, and the methodology used to ensure adequate protection of safe shutdown systems in accordance with Section III.G of Appen-dix R to 10 CFR 50 (Section C.5.b of BTP CMEB 9.5-1) is acceptable. Further, the staff concludes that, with approved deviations, the Clinton Power Station complies with the requirements of Section III.G of Appendix R to 10 CFR 50 (Section C.S.b of BTP CMEB 9.5-1).
FSAR Section 7.4.1.4 describes the remote shutdown control panel's design and capability. The design objective of the remote shutdown control panel is to achieve and maintain cold shutdown in the event of an evacuation of the control room. The design of the remote shutdown control panel provides the capability to electrically isolate the control functions and indications for the shutdown system from the control room. The reactor core isolation cooling (RCIC) system, three safety / relief valves, and one train of the suppression pool cooling mode and shutdown cooling mode of the residual heat removal (RHR) can be controlled from the remote shutdown control panel to achieve cold shutdown should a fire result in evacuation of the control room. Additionally, the shutdown service water and essential HVAC systems can be controlled at the remote shutdown panel.
The design of the remote shutdown control panel was reviewed to determine com-pliance with the performance goals outlined in the requirements of Section III.L of Appendix R to 10 CFR 50 (Section C.5.c of BTP CMEB 9.5-1). Reactivity con-trol will be accomplished by a manual scram before the operator leaves the con-trol room. The capability to manually scram is also available outside the con-trol room. The RCIC system will provide reactor coolant makeup, and the RHR system and the safety / relief valves will be used for reactor decay heat removal.
Reactor vessel water level, reactor vessel pressure, suppression pool water.
level and temperature, RCIC pump turbine speed, and RHR system flow are among the instrumentation available at the remote shutdown control panel to provide direct readings of process variables. The remote shutdown control panel also will include instrumentatien and control of support functions needed for the shutdown equipment. On the basis of this review and evaluation, the staff con-cludes that the remote shutdown control panel complies with the requirements of Section III.L of Appendix R to 10 CFR 50 (Section C.5.c of BTP CMEB 9.5-1) and is, therefore, acceptable.
9.5.9 Conclusion On the basis of its review and evaluation of the applicant's Fire Protection Program, the NRC staff now concludes that the Fire Protection Program, with the approved deviations, meets the guidelines of BTP.CMEB 9.5-1, Appendix R to 10 CFR 50, and General Design Criterion 3 (10 CFR 50, Appendix A) and is, therefore, acceptable.
I i
Clinton SSER 3 9-4
A 12 RADIATION PROTECTION 12.5 Operational Radiation Protection Program 12.5.1 Organization In the SER, the staff found the applicant's radiation protection organization acceptable because it met the intent of RGs 1.8, 8.2, 8.8, and 8.10, and NUREG-0731. By letter dated December 20, 1983, the applicant submitted Amendment 28 of Clinton's FSAR, wherein the applicant revised the radiation protection organization to completely separate the radiation protection func4 tions from the chemistry functions. In addition, the Supervisor - Radiation Protection (RPM) now reports directly to the plant manager instead of through an assistant plant manager. The staff has reviewed the applicant's proposed changes to the radiation protection organization against the criteria of NUREG-0800 and finds the changes acceptable.
1 4
t Clinton SSER 3 12-1
17 QUALITY ASSURANCE In the FER, the NRC staff found that the quality assurance (QA) program de-scription contained in Section 17.2 of the FSAR was acceptable. However, the staff stated that the applicant had to provide additional information and clar-ification with regard to the identification of specific items covered by the QA program and the overall authority and responsibility of the QA organization and the Onsite Compliance Department. Since the issuance of the SER, the applicant has provided the requested additional information in Amendments 25 and 27 to the FSAR. The NRC staff has reviewed this information and prepared the following review, which supersedes that described in Sections 17.2 and 17.4 of the SER.
17.2 Organization The structure of the organization responsible for the operation of Clinton and for the establishment and execution of the operations phase QA program is shown in Figure 17.1. The Executive Vice President has overall responsibility for the program for Illinois Power Company (IPC) during the design, construction, and operation of Clinton. The Vice President of the Clinton Power Station (CPS) reports directly to the Executive Vice President and is responsible for the engineering, quality assurance, testing, startup, and commercial operation of the station.
The Power Plant Manager and the Quality Assurance Manager both report directly to the Vice President of CPS. The Power Plant Manager is responsible for the safe, reliable, and efficient operation of CPS in accordance with the operating license. This includes ensuring that the IPC Nuclear Power QA Program is in-corporated in plant procedures and implemented by the CPS organization. The Quality Assurance Manager is responsible for IPC's overall QA program defini-tion, direction, evaluation, and approval. He directs the QA departmental activities related to the design, procurement, construction, and operation of CPS. This includes the review of program procedures and the inspection and audit of quality-related activities to ensure conformance with FSAR commitments.
The Vice President of CPS, the Power Plant Manager, and the Quality Assurance Manager, including their staff, are located on site.
Five sections reporting to the Quality Assurance Manager assist the manager in carrying out the QA responsibilities. These sections are Quality Engineering, Program and Procedures, Audit and Surveillance, Quality Control, and Admini-stration. The projected overall size of the Quality Assurance Department staff is 50, of which 10 are in the Quality Engineering Section, 10 in the Programs and Procedures Section, 15 in the Audit and Surveillance Section, 10 in the Quality Control Section, and 5 in the Administration Section.
The Quality Engineering Section reviews and approves procurement documents and nonconformance reports; establishes quality control inspection points in test procedures and work documents; specifies QA program requirements in design, modification, and procurement documents; and participates in Facility Review Group activities.
Clinton SSER 3 17-1
l The Programs and Procedures Section is responsible for periodically assessing departmental effectiveness in implementing the IPC Nuclear Power QA Program, trending of conditions adverse to quality, and coordinating QA reviews and approval of QA program requirements associated with the QA Manual and Manage-ment Guides.
The Audit and Surveillance Section is responsible for the planning of internal and external IPC audits and surveillances. This section ensures timely and responsive corrective action to IPC audit and surveillance findings and advises management as to the effectiveness of QA program implementation for those specifi, functions audited and under surveillance.
The Quality Control Section is responsible for conducting and reporting inspec-tions of activites performed by either IPC or by. contractors of IPC that affect the quality of manufacturing, construction, installation, or testing. This section is responsible for the qualification of personnel and procedures and the performance of nondestructive examinations of IPC work. This section is also responsible for the control, issuance, and distribution of documents asso-ciated with the IPC Nuclear Power QA Program, such as the QA Manual, Management Guides, and departmental procedures and instructions.
QA personnel do not perform non quality assurance duties but provide full attention to ensuring effective implementation of the QA program.
The QA organization is responsible for (1) reviewing and approving the Opera-tional QA Manual and reviewing and concurring with operations administrative procedures that detail activities within the scope of the program, (2) review-ing procurement documents and concurring that appropriate QA requirements are specified, (3) reviewing and approving suppliers' QA program requirements, (4) performing reviews and evaluations of procurement sources to determine their capability to meet QA requirements, (5) performing surveillance and eval-uations at suppliers' facilities to verify continued compliance with the QA requirements of the procurement documents, and (6) performing surveillance and audits of all quality-related activities performed on site, off site, and at suppliers' facilities. Personnel performing quality-related activities have the authority to (1) identify problems relating to quality, which include stop-work recommendations by the QA organization; (2) initiate, recommend, or provide solutions to problems relating to quality; (3) verify implementation of solu-tions; and (4) control further processing, delivery, installation, or.utiliza-tion of nonconforming items until proper disposition has occurred.
On'the basis of the review and evaluation of the QA program description con-tained in Chapter 17 of the FSAR (through Amendment 25) for CPS, Unit 1, the staff concludes that the organizations and persons performing QA functions have the required independence and authority to effectively carry out the QA program without undue influence from those directly responsible for cost and schedule.
17.4 Conclusion On the basis of its detailed review and evaluation of the'QA program descrip-tion contained in Section 17.2 of the FSAR and the additional QA information provided in Amendments 25 and 27, the staff concludes that:
Clinton SSER 3 17-2
(1) The organizations and persons performing QA functions have the required independence and authority to effectively carry out the QA program with-out undue influence from those directly responsible for costs and schedules.
(2) The QA program describes requirements, procedures, and controls that, when properly implemented, comply with the requirements of Appendix B to 10 CFR 50.
(3) The FSAR identifies an acceptable listing of items that are under the control of the QA program.
4 l
I Clinton SSER 3 17-3
n I
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8 M EXECUTIVE E VICE PRESIDENT w
' f
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LOCATED ON PRESIDENT t 1
MANAGER MANAGER y QUALITY POWER A
ASSURANCE PLANT l
SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR SUPERWSOR QUALITY PROGRAMS AND AUDITS AND ADMINISTRATION QUALITY SURVEILLANCES ENGINEERING CONTROL PROCEDURES Figure 17.1 Illinois Power Company Clinton Power Station QA organization
j r APPENDIX A
+
CONTINUATION OF CHRONOLOGY L
~
May 10, 1983 Letter from applicant forwarding documentation concerning SER outstanding issue (3),_in response r to request made at April 28, 1983, meeting with j Sargent & Lundy in Bethesda, Maryland.
May 18, 1983 Letter from applicant forwarding information to !
update September 24, 1983, response to April 26, i
1982,-request concerning fast scram loads on control rod drive system.
.May 20, 1983 Letter from applicant forwarding response-to
- review comments on FSAR Section 3.11, " Environ-mental Qualification _of Electrical Equipment."
! May 24, 1983 Letter from applicant advising that fuel load i i date has been changed from January 1984 to
- January 1986.
, May 24, 1983 Letter from applicant presenting design-basis j spectra and site-specific spectra.'
May 31, 1983 Letter from applicant forwarding FSAR Amendment 25 containing revision to Chapter 17 to resolve SER i outstanding issue (20)~and SER confirmatory-issue (6).
May 31, 1983 Letter to applicant forwarding SSER 2. ,
., June 1, 1983 Letter to~ applicant responding to letter dated i
December 3, 1982, and FSAR Amendment 16 concerning emergency operations facility (EOF) and backup
~
- EOF.
i -i
- June 6, 1983 Letter from applicant forwarding schedule ~,
j implementation plan for emergency response j capability and list of'submittals required by NUREG-0737, Supplement l'.
i I June 16, 1983 '
Letter from' applicant forwarding " Emergency-i . Response Facilities Design Report,"'in response-
! ; June'17,1983 ~ Letter from applicant. forwarding Action Plans 1, 2, 3, 4, and 33 in response to HumphreyLconcerns.
,l IClintonSSER3 l' Appendix A-i:.
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.( .
g y .s Lettertroepplicantconcerni$gTMITaskAction 4
,\v. June 29(1983
\, Plan djem'I.G.1, "Special Low Power Testing and 9 '/ Trai Ary; and Recenmendations 7 for BWRs." (Generic Le{te{ (M-24 bJuly,j,1983l 3
[), 'feheT[to applicant concerning clarification of
' ' / syveulance) requirements for fiesel fuel
'g , f ,\
impupty 4 (Generic Letter 83[26) 1
, , July 5,19hN s Let fro applicant forwa,'eding revised emergency 5 ^ responsec capability implememtation plan schedule and subm'ittals to ref. lect riew 'sabmittal dates of 3'
i}I ~l- January 1,1985, for control room review program t, ,
'p. '
plag ,and, July 1,1985, for sunary report in
, p (' , ,t, accordance with NUREG-0737, Supplement 1.
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Ju y 6. 1Q83 , Letty to applicant concernins surveillance intWals in; Standard Technica'i Specifications.
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(( (Gen,erteLauer'p-27)
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' July,8, 1 W 4 Letter to applica'nt concerning required actions
- [/ ,'g hesed dn generic implications of Salem anticipated
- 9 *\ !
/
transientSithou'. scram (ATWS) events. (Generic
}j ( /' f Letter 83' 2h -
?
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July 8,1983 [ y , hetter frc$ appli, cant forwarding listing of f
( , s ' hdchted soudi v ef documentation and comparisen I )i Y I ./ of plant witp 4uo@eng, Grand Gulf, and Perry to
[, &1 , resolve SER opt:tanding issue (9).
. . ?
f ,i ' ' '
July 8, 1983 , q.7hterfrpmapplicantforwardingsupplemental-
- 4 psponse to Gederic Letter 83-01 concerning licensed operatar examinations.
s / '/,
July 14, 1983 #
/./ L4tter fryd,,tilcant clarifying information on
't / fire zones .3.2 and D.3.6 in response to December 16(,1982.,N Atter concerning update of
., safe shu%own analysis for fire protection and rd.jue,stf t.r exempt.fons from 10 CFR 50, Appendix R.
, July 20, 1983 Letter to applican't forwarding evaluation of i
i October 22, 1982/,fsubmittal concerning opera-
biUty pf qwtainment purge and vent valves kI under ptWnis.j loss-of-coolant accident / seismic
., , 3 ioadings.%y g
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Jy1( 22.* D83 etter,fromppoplicantprovidingquarterly
. , j N' const$b etiott schedule updates in response to
(
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\ ' May}9,198), request.
A (gdoly26,1983 Letter from ad licant forwarding summary update
\ j to informalMn{ppviously submitted by December 15,
" ,geB2,,lettdrconcerningelectricalandmechanical
.' N( ' sdsmic aid 'ynun,ic
- a equipment qualification.
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p< f n,
Clinton SSER 3 * . Appendix A
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July 27, 1983 Letter from applicant requesting approval for use of ASME Code, Case N-315, " Repair of Bellows,"
for repairs on guard pipe expansion bellows.
July 28, 1983 Letter from applicant forwarding supplemental information addressing Guidelines 1, 4, and 6 of NUREG-0612, " Control of Heavy Loads."
July 29, 1983 Letter from applicant forwarding FSAR Amendment 26.
August 1, 1983 Letter from applicant forwarding equipment qualification status matrix and annotated FSAR Tables 3.11-3 and 3.11-4.
August 26, 1983 Letter from applicant discussing transportation routes.
August 30, 1983 Letter from applicant forwarding draft FSAR Page 3.2-2Sb, reflecting changes in Table 3.2-1 concerning containment /drywell purge systems to close out SER open item (20).
August 31, 1983 Letter from applicant requesting delay in responding to Generic Letter 83-28 concerning generic implications of Salem ATWS events.
September 6, 1983 Letter from applicant providing proposed responses to questions concerning SER confirmatory issue (1).
September 8, 1983 Letter from applicant advising that emergency lighting requirements for post-fire conditions will be met.
September 9, 1983 Letter from applicant forwarding Revision 3 to complianch report on Regulatory Guide 1.97.
September 12, 1983 Letter from applicant forwarding draft FSAR page changes on THI-2, Issue II.K.3.18, concerning modification to automatic depressurization system logic.
September 16, 1983 Letter from applicant forwarding additional information on conservatisms inherent in March 4, 1983, probabilistic risk assessment thus resolving outstanding issue (1). l l
September 16, 1983 Letter from applicant forwarding proprietary version of NEDE-22146, "Kuosheng 1 Reactor Internals vibration Measurements." SER confirma-tory issue (9) closed.
September 22, 1983 Letter from applicant forwarding information to resolve SER outstanding issue (13).
Clinton SSER 3 3 Appendix A L-
b 1
September 27, 1983 Letter to applicant informing ;iim of conditional acceptance of use of ASME Code, Case N-315,
" Repair of Bellows,Section III, Division 1,"
to repair guard pipe expansion bellows.
September 30, 1983 Letter from applicant informing staff that R. Nelson was appointed Director of Nuclear Licensing and Configuration.
September 30, 1983 Letter from applicant informing staff that utiiity has incorporated three Licensing Review Group II (LRG-II) position papers into operating license application. Position Paper 3-CSB,
" Periodic Low Pressure Leakage Testing of Drywell,"
constitutes response to SER confirmatory issue (66).
October 6, 1983 Letter from applicant forwarding draft FSAR page changes on SER confirmatory issues (45) and (46) concerning TMI Action Plan Items II.K.3.13, "HPCI and RCIC Initiation Levels," and II.K.3.15,
" Isolation of HPCI and RCIC."
October 11, 1983 Letter from applicant forwarding FSAR Amendment 27.
October 14, 1983 Letter from applicant informing staff that high pressure core spray diesel generator preopera-tional testing program complying with Regulatory Guides 1.108 and 1.9 and IEEE Std 387-1977 will be implemented.
October 14, 1983 Letter from applicant forwarding public version of draft FSAR Amendment 28, Appendix 13.B, consisting of Revision 2 to Emergency Plan thus resolving SER outstanding issue (18).
October 17, 1983 Summary of Caseload Forecast Panel meetings with applicant on August 10-12, 1983, in Clinton, Illinois, concerning construction status of facility.
October 27, 1983 Letter from applicant responding to May 29, 1981, request for construction schedule quarterly update.
October 28, 1983 Letter from applicant forwarding safety parameter display system preimplementation package documents.
October 31, 1983 Letter to applicant concerning NUREG-0965, "NRC Inventory of Dams." (Generic Letter 83-38)
November 2, 1983 Letter to applicant concerning clarification of TMI Action Plan Item II.K.3.31. (Generic Letter 83-35)
Clinton SSER 3 4 Appendix A
November 2, 1983 Letter to applicant granting August 31, 1983, request for extension of 120-day response to Generic Letter 83-28.
November 3, 1983 Letter from applicant clarifying analysis requirements to resolve SER confirmatory issue (6).
November 14, 1983 Letter to applicant forwarding request for additional information on containment system and mechanical engineering.
November 15, 1983 Letter to applicant forwarding request for additional information on commitment to Regula-tory Guides 1.58 and 1.46 in accordance with Generic Letter 81-01.
November 17, 1983 Letter to applicant informing him of changes in methodology regarding containment vent and purge valve operability (SER confirmatory issue (69)).
December 2, 1983 Letter to applicant concerning NRC recommenda-tions on operator action for reactor trip and ATWS. (Generic Letter 83-32)
December 8, 1983 Letter from applicant forwarding information on November 3, 1983, commitment to perform local hydrogen detonation analysis thus resolving SER confirmatory issue (6).
December 16, 1983 Letter from applicant forwarding commitment to install emergency lighting.
December 16, 1983 Letter to applicant concerning fast cold starts of diesel generators. (Generic Letter 83-41)
December 19, 1983 Letter to applicant on clarification to Generic Letter 81-07 concerning response to NUREG-0612,
" Control of Heavy Loads at Nuclear Power Plants."
December 19, 1983 Letter to applicant concerning reporting require-ments of 10 CFR 50.72 and 50.73 and Standard Technical Specifications. (Generic Letter 82-43)
December 20, 1983 Letter from applicant forwarding FSAR Amendment 28 consisting of revisions to Appendix 13.B, " Emergency Plan," and Chapter 12, " Radiation Protection."
December 20, 1983 Summary of September 23, 1983, meeting with applicant and Sargent & Lundy in Bethesda, Maryland, concerning SER outstanding issue (1).
Clinton SSER 3 5 Appendix A
December 21, 1983 Letter from applicant forwarding facility staffing survey questionnaire in response to December 6, 1983, request.
December 21, 1983 Letter to applicant forwarding revised positions in response to November 15, 1983, request for additional information on commitments to Regula-tory Guides 1.58 and 1.146 in regard to quality assurance program.
December 21, 1983 Letter from applicant forwarding response to November 7,1983, questions concerning July 28, 1983, submittal of Phase I of NUREG-0612, " Control of Heavy Loads."
January 5, 1984 Letter to applicant concerning NRC use of terms, "important to safety" and safety-related."
January 6, 1984 tetter to applicant concerning notice of meeting on facility staffing. (Generic Letter 84-02)
January 13, 1984 Letter to applicant concerning availability of NUREG-0933, "Prioritization of Generic Safety issues." (Generic Letter 84-03)
January 25, 1984 Letter from applicant advising staff that D. I. Herborn has assumed position of Director, Nuclear Licensing and Configuration, effective January 23, 1984.
January 26, 1984 Letter from applicant clarifying September 25, 1981, letter concerning NUREG-0612, " Control of Heavy Loads."
Janua ry 26, 1984 Letter from applicant providing quarterly update on construction schedule.
January 27, 1984 Letter to applicant forwarding request for additional information on facility's heating, ventilation, and air conditioning (HVAC) structural design.
January 31, 1984 Letter from applicant forwarding presentation material used at September 23, 1983, meeting to close SER outstanding issue (1).
February 6, 1984 Letter to applicant advising him that October 14, 1053, letter was unresponsive to concerns in NRC April 1, 1983, letter regarding high pressure core spray diesel generator qualification test requirements.
6 Appendix A Clinton SSER 3 i
February 8,'1984 Letter from applicant forwarding presentation material used at September 23, 1983, meeting concerning SER outstanding issue (1) that was inadvertently omitted in January 31, 1984, letter.
February 9, 1984 Letter from applicant forwarding revised diagrams reflecting implementation of NUREG-0737, TMI Action Plan Items II.K.3.13, II.K.3.15, and II.K.3.18, thus closing out SER confirmatory issue (28), (45), and (46).
February 9, 1984 Letter to applicant advising him that NRC will not review emergency plans to delete privacy or proprietary information unless specifically requested to do so in accordance with Generic Letter 81-27.
February 10, 1984 Letter from applicant forwarding functional description of safety parameter display system.
Results of review of verification and validation plan recommendations also provided.
February 10, 1984 Letter from applicant responding to Generic Letter 83-42, which indicated that guidelines of NUREG-0612 concerning single-failure proof cranes may be deficient. Review of diagrams indicates no single failure could result in uncontrolled motion of crane.
February 17, 1984 Letter from applicant forwarding response to November 14, 1983, request for additional information on SER issues concerning suppression pool dynamics.
4 February 17, 1984 Letter from applicant responding to January 27, 1984, letter requesting schedule for providing additional information on HVAC design criteria and ductwork.
February 21, 1984 Letter from applicant informing staff that responsibilities for loss prevention / fire protection engineering activities were assumed by E. M. Haagar.
March 1, 1984 Letter to-Franklin Institute / Franklin Research Center forwarding listed reports concerning NRC-03-81.130, " Control of Heavy Loads at Nuclear Power Plants."
~Clinton SSER 3 7 Appendix'A
APPENDIX B REFERENCES Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, Washington, D.C. ' includes General Design Criteria).
Illinois Power Company et al., " Final Safety Analysis 2eport, Clinton Power Station Units 1 and 2," Apr. 1978.
Tennessee Valley Authority, " Justification of the Seismic Design Criteria Used for the Sequoyah, Watts Bar, and Bellefonte Nuclear Power Plants - Phase II,"
Aug. 1978.
U.S. Nuclear Regulatory Commision, NUREG-0011, " Safety Evaluation Report Related to the Operation of Sequoyah Nuclear Plant, Units 1 and 2," Mar. 1979.
-- , NUREG-0731, " Guidelines for Utility Management - Structure and Technical Resources," Sept. 1980.
-- , NUREG-0793, " Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2," May 1982.
-- , NUREG-0798, " Safety Evaluation Report Related to the Operation of Enrico Fermi Atomic Power Plant," July 1981.
NUREG-0800 (formerly NUREG-75/087), " Standard Review Plan for Review of 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," Feb. 1982; Supplement No. 1, July 1982; Supplement No. 2, May 1983.
-- , NUREG-0881, " Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No.1," Apr. 1982.
-- , NUREG-0887, " Safety Evaluation Report Related to the Operation of Perry Nuclear Power Plant, Units 1 and 2," May 1982.
-- , NUREG/CR-1582, " Seismic Hazard Analysis," Vols. 4 and 5, Oct. 1981.
Clinton SSER 3 1 Appendix B
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APPENDIX D . , ,[
NRC STAFF CONTRIBUTORS AND CONSULTANTS -c-
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This supplement to the Clinton Safety Evaluation Report is a product of the ..'
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NRC staff. The NRC staff members listed below were principal contributors to ?' . . . ^"
this report. , .
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Name Title Branch ,7' N. Chokshi Structural Enginear Structural Engineering '
1, N. Fioravante Auxiliary Systems Engineer Auxiliary Systems ?
G. Giese-Koch Geophysicist Geosciences ? 7. . . ..
J. Gilray Quality Assurance Engineer Quality Assurance : ,.
B. Hardin Reactor Engineer Reactor Systems '[( . ' -.' '
M. Lamastra Sr. Radiation Engineer Radiological Assessment : .:
J. Levine Meteorologist Accident Evaluation . _ .' ... .. yC L. Ruth Sr. Containment Systems Engineer Containment Systems ... f . ~
Site Analyst A. Sinisgalli Siting Analysis ..:
J. Spraul Quality Assurance Engineer Quality Assurance ' J;;' ' ,
Fire Protection Engineer J. Stang Chemical Engineering ;l S. Sun Nuclear Engineer Core Performance . . .z; ~ .
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BIBLIOGRAPHIC DATA SHEET N -0853 -
set IN$,.uCriONS ON Ywe alvtast Sucolement No. 3 2 . , L t ANo s. ,i, L t 3 Leave st.Nn Safety luation Reoort related to the operation of __
Clinton P er Station, Unit No.1 "
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Division of Licens 9 -
Office of Nuclear ctor Regulation ... .N,%...
p U. S. Nuclear Regula ry Commission Washington, D. C. 20 to S UNSo.*NG ORGANil A,40N Nawt ANo W A,L a 0084 55s tesi,e. l. Co s 11 ,v,kO.Rt.oh, Safety Evaluation Report _
Same as 7 above .. ioocov.a o ~,,.. .,
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Pertains to Docket No. 50-461 ; A i .....o.m, ,
Supplement No. 3 to the Safety valuati n Report on the application filed 5 by Illinois Power Company, Soyl Po Coooerative, Inc., and Western Illinois Power Cooperative, Inc., s plicants and owners, for a license ._
to operate the Clinton Power Stati Unit No. 1 has been preoared by the Office of Nuclear Reactor Regulatio of the U. S. Nuclear Regulatory -
Commission. The facility is locat n Haro Township, DeWitt County, -
Illinois. This supplement report t status of certain items that had not been resolved at the time of 'ubl' tion of the Safety Evaluation :
Reoort and Supplements No.1 an No. 2. -
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