ML20210M465
| ML20210M465 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 09/30/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0853, NUREG-0853-S07, NUREG-853, NUREG-853-S7, NUDOCS 8610030414 | |
| Download: ML20210M465 (56) | |
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.g i NUREG-0853 Supplement No. 7 Safety Evaluation Report related to the operation of Clinton Power Station, Unit No.1 Docket No. 50-461 lilinois Power Company, et al.
I U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation ,
September 1986 ps >* "'%,
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BBA 288 M 888383u E PDR
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l NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available foc inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices:
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ABSTRACT Supplement No. 7 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 items that have been resolved by the staff since Supple-ment No. 6 was issued.
i Clinton SSER 7 iii
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TABLE OF CONTENTS PaILe ABSTRACT............................................................. iii 1 INTRODUCTION AND GENERAL DISCUSSION............................. 1-1 1.1 Introduction............................................... 1-1 1.9 Outstanding Issues......................................... 1-1 1.10 Confirmatory Issues........................................ 1-4 1.11 License Conditions......................................... 1-8 1.13 Contentions................................................ 1-10 2 SITE CHARACTERISTICS............................................ 2-1 2.2 Nearby Industrial, Transportation, and Military Facilities. 2-1 2.2.2 Nearby facilities .................................. 2-1 2.6 Stability of Subsurface Materials and Foundations ......... 2-1 2.6.1 Site Conditions..................................... 2-1
- 2. 6.1.1 Genera 1.................................... 2-1 2.6.4 Stability of S1 opes................................. 2-2 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS......... 3-1 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment........................ 3-1 3.10.1 Sei smic and Dynamic Quali fication. . . . . . . . . . . . . . . . . . . 3-1 4 REACT 0R......................................................... 4-1 i
4.3 Nuclear Design............................................. 4-1 i
4.3.2 Design Description.................................. 4-1 4.3.2.7 Vessel Irradiation.....................'.... 4-1 4.6 Functional Design of Reactivity Control System............. 4-3 6 ENGINEERED SAFETY FEATURES...................................... 6-1 6.2 Containment Systems........................................ 6-1 i
Clinton SSER 7 v
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TABLE OF CONTENTS (Continued)
.P39e 6.2.2 Secondary Containment............................... 6-1 6.2.5 Combustible Gas Control............................. 6-1 7 INSTRUMENTATION AND CONTR0L..................................... 7-1 7.3. Engineered Safety Features Systems......................... 7-1 7.7.3 Resolution of Issues................................ 7-1 8 ELECTRIC POWER SYSTEM........................................... 8-1 8.2 Offsite Power System....................................... 8-1 8.2.1 General Description................................. 8-1 8.2.2 Load Flow and Grid Stabili ty Analysis. . . . . . . . . . . . . . . 8-1 8.3 Onsite Emergency Power Systems............................. 6-1 8.3.1 AC Power System..................................... C-1 8.3.2 DC Power System..................................... B-3 8.4 Other Electrical Features and Requirements for Safety...... 8-3 8.4.1 Containment Electrical Penetrations................. 8-3 8.4.4 Use of a Load Sequencer With Offsite Power. . . . . . . . . . 8-4 8.4.7 Physical Identification and Independence of Redundant Safety-Related Electrical Systems........ 8-4 9 AUXILIARY SYSTEMS.................................. 9-1 3...........
9.1 Fuel Storage Facility...................................... 9-1 9.1.1 New Fuel Storage.................................... 9-1 l
9.3 Process Auxiliaries........................................ 9-2 9.3.1 Instrument Air System............................... 9-2 9.3.3 Standby Liquid Control System. . . . . . . . . . . . . . . . . . . . . . . 9-3 9.3.5 Postaccident Sampling Capability (NUREG-0737,
- Item II.B.3)........................................ 9-4 9.4 Ventilation Systems........................................ 9-5 l
! 9.4.1 Control Room Area Ventilation System (Control Room
! Heating, Ventilating and Air Conditioning (HVAC)
System)............................................. 9-5 Clinton SSER 7 vi
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TABLE OF CONTENTS (Continued)
P,afte 11 RADI0 ACTIVE WASTE MANAGEMENT.................................... 11-1 11.5 Time-Dependent Correction Factors for Effluent Radiation Monitors................................................... 11-1 13 CONDUCT OF OPERATIONS........................................... 13-1 13.1 Organizational Structure of Applicant...................... 13-1 13.1.2 Operating Organization............................. 13-1 13.2 Training Program........................................... 13-1 13.2.2 Unlicensed Personne1................................ 13-1 13.2.3 Replacement and Retraining.......................... 13-2 14 INITIAL TEST PR0 GRAM............................................ 14-1 15 SAFETY ANALYSIS................................................. 15-1 15.1 Anticipated Operational Occurrences........................ 15-1 17 QUALITY ASSURANCE............................................... 17-1 17.2 Organization............................................... 17-1 17.4 Conclusions................................................ 17-3 APPENDIX A CONTINUATION OF CHRONOLOGY APPENDIX B REFERENCES APPENDIX D NRC STAFF CONTRIBUTORS APPENDIX F ERRATA TO CLINTON POWER STATION SAFETY EVALUATION REPORT APPENDIX P CONTENTIONS l
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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 No. 1, Docket No. 50-461. Supplement No. 1 (SSER 1) to the Clinton SER was issued in July 1982; SSER 2 was issued in May 1983; SSER 3 was issued in May 1984; SSER 4 was issued February 1985; SSER 5 was issued in January 1986; and SSER 6 was 4,
issued in July 1986. The purpose of this seventh supplement (SSER 7) is to further revise the SER by providing results of the NRC staff's review of infor-mation submitted by the applicant to address the remaining unresolved issues listed in Sections 1.9 and 1.11 of the SER and its supplements.
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 i in this supplement does not replace the material in the corresponding SER sec-tion or appendix. Appendix A is a continuation of the chronology of correspon-dence between NRC and the applicant and makes current the lists in the SER and in SSER 1 through SSER 6. Appendix B is a list of references cited in this re-port; the availability of the references is described on the inside front cover of this report. Appendix D is a list of principal contributors to this supple-u ment. Appendix F corrects errors in the SER and its supplements. Appendix P H* provides detailed information about the status of specific issues addressed in a joint stipulation agreement resolving the two remaining contentions related
! to quality assurance and control room design.
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 on the inside front cover.
The NRC Project Manager assigned to the operating license application for Clinton Unit 1 is Byron L. Siegel. Mr. Siegel may be contacted by calling (301) 492-9474 or by writing to Mr. Byron L. Siegel Division of BWR Licensing, Mail Stop P-924 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 1.9 Outstanding Issues 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 4 of those issues had been satisfactorily resolved and I had been changed to a con-firmatory status. SSER 2 reported that 6 issues had either been resolved or Clinton SSER 7 1-1
a changed to a confirmatory status. SSER 3 reported that 4 issues had been re-solved.. SSER 4 partially resolved 1 issue and reopened another. SSER 5 re-ported that 1 issue had been resolved and added an additional outstanding issue.
SSER 6 resolved 6 outstanding issues and partially resolved 1 outstanding issue.
The current status of each of the 21 issues is tabulated below. For the one remaining issue which is resolved in this supplement, the relevant section in this document is indicated.
SSER 7 Issue Status Section(s)
(1) Transportation accidents Resolved in SSER 3 --
(2) Effects of Unit 2 excavation Resolved in SSER 2 --
(3) Seismic analysis Became confirmatory --
issue 70, resolved in SSER 3 (4) Internally generated missiles Resolved in SSER 1 --
(5) Postulated piping failures Resolved in SSER 6 --
(6) Steady-state vibration Resolved in SSER 2 --
acceptance criteria for balance of plant piping (7a) Environmental qualification Resolved in SSER 5 --
of electrical and mechanical and SSER 6 equipment (7b) Seismic and dynamic qualifi- Resolved in SSER 7 3.10.1 cation of mechanical and electrical equipment (7c) Pump and valve operability Resolved in SSER 6 --
qualification NUREG-0737 Item II.E.4.2(6) Resolved in SSER 5 (8a) Preservice (PSI) and inservice PSI program: became --
inspection (ISI) programs confirmatory issue 67 in SSER 1 ISI program: became --
license condition 12 in SSER 2 (8b) Preservice and inservice Became confirmatory --
testing of pumps and valves issue 68 in SSER 1 (9a) Pool dynamic loads due to LOCA Resolved in SSER 6 --
Clinton SSER 7 1-2
i SSER 7 Issue Status Section(s)
(9b) Pool dynamic loads due to SRV Resolved in SSER 5 (10a) Containment purge Became confirmatory --
issue 69 in SSER 2 (10b) Containment isolation Resolved in SSER 2 --
(10c) Containment leakage testing Resolved in SSER 2 --
(vent and drain lines)
(10d) Containment leakage testing Resolved in SSER 2 --
(10e) Containment bypass leakage Resolved in SSER 2 --
(11) Control room habitability Resolved in SSER 1 --
(12) Engineered safety features Resolved in SSER 2 --
reset controls (IE Bulletin 80-06)
(13) Remote shutdown system Resolved in SSER 3 -.
- and SSER 6 (14) Capability for safe shutdown Resolved in SSER 2 --
following loss of bus supply-ing power to instruments and controls (IE Bulletin 79-27)
(15) Control system failures Resolved in SSER 6 --
- resulting from high-energy-line breaks or common power source or sensor malfunctions (16) Separation of the RPS and MSIV Resolved in SSER 1 --
solenoid circuits and PGCC circuits (17) Organization and staffing Resolved in SSER 5 --
(18a) Onsite emergency plan Resolved in SSER 4 --
(18b) Offsite emergency plan Resolved in SSER 6 --
(19) Security Resolved in SSER 1, --
amended security plans reviewed and approved -
in SSER 5 (20) QA program Resolved in SSER 3 --
Clinton SSER 7 1-3 1
SSER 7 Issue Status Section(s)
(21) Fire Protection Evaluation Resolved in SSER 6 --
Report and Safe Shutdown Analysis 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 issues had been satisfactorily resolved. SSER 2 addressed 11 additional issues that had been resolved, as well as certain issues that still required resolution. SSER 3 addressed 9 addi-tional issues that had been resolved. SSER 4 addressed 10 additional issues that had been totally resolved and 2 that had been partially resolved. SSER 5 addressed 9 confirmatory issues that had been totally resolved. SSER 6 resolved the 4 remaining confirmatory issues. 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. Issue 71 was added in SSER 3. The current status of each of the 71 issues is tabulated below.
SSER 7 Issue Status Section(s)
(1) Emergency preparedness Resolved in SSER 6 --
meteorological program (2) Inspection program around the Resolved in SSER 1 --
ultimate heat sink (UHS) and the main cooling lake dam (3) Protection of UHS dam abutments Resolved in SSER 1 --
against soil erosion (4) Internally generated missiles - Resolved in SSER 2 --
fan failures (5) Design adequacy of cable tray Resolved in SSER 1 --
l system (6) Containment ultimate strength Removed from list in --
analysis SSER 4 (7) Structural integrity of safety- Resolved in SSER 2 --
related masonry walls (8) NSSS pipe break analysis using Resolved in SSER 1 --
SRP criteria (9) Vibration assessment of RPV Resolved in SSER 4 --
internals Clinton SSER 7 1-4
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SSER 7 Issue Status Section(s)
(10) Annulus pressurization loads Resolved in SSER 4 --
(LOCA asymmetric loads)
(11) Use of SRSS for combining Resolved in SSER 1 --
Mark III dynamic responses for other than LOCA and SSE (12) IE Bulletin 79-02 regarding Resolved in SSER 2 --
support baseplate flexibility (13) Mark III hydrodynamic loads Became part of out- --
standing issue 9 to avoid duplication in SSER 4 (14) Feedwater check valve analysis Resolved in SSER 2 --
(15) Seismic and LOCA loadings Resolved in SSER 4 --
on fuel assemblies (LRG II Issue 2-CPB)
(16) Scram discharge system Resolved in SSER 1 --
evaluation (17) Fracture toughness data Resolved in SSER 1 --
(18) Subcompartment pressure Resolved in SSER 5 --
analysis (19) Combustible gas control Resolved in SSER 3 --
(20) Containment isolation Resolved in SSER 2 --
dependability 1
(21) Containment monitoring, Resolved in SSER 4 --
II.F.1(1) through II.F.1(6) and SSER 5 (22) Plant-specific LOCA analysis, Resolved in SSER 3 --
. II.K.3.31 (23) High drywell pressure Resolved in SSER 1 --
interlocks l
(24) ATWS recirculation pump trip Resolved in SSER 6 --
(25) Response-time testing Resolved in SSER 1 --
(26) Analog trip modules and optical Resolved in SSER 2 --
isolators Clinton SSER 7 1-5
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(27) Susceptibility of the NSPS to Resolved in SSER 1 --
electrical noise (28) Modification of ADS logic, Resolved in SSER 4 --
II.K.3.18 (29) Restart of low pressure Resolved in SSER 1 --
systems, II.K.3.21 l
(30) Temperature effects on level Resolved in SSER 2 --
measurements (31) Containment atmosphere Resolved in SSER 4 --
monitoring system (32) Verification that testing is Removed from list in --
in accordance with BTP PSB-1 SSER 1 (33) Electrical drawing review Removed from list in --
SSER 1 (34) Verification of diesel Resolved in SSER 4 --
generator testing (35) Class A supervision and power Resolved in SSER 3 --
supply for fire detection system (36) Circulating water system Resolved in SSER 2 --
(37) Initial test program Resolved in SER --
i (38) Human engineering aspects of Resolved in SSER 5 --
control room design, I.D.1 (39) Common reference for reactor Resolved in SSER 2 --
l vessel level instruments, i II.K.3.27 i
i (40) Shielding design review, Resolved in SSER 1 --
II.B.2 l
(41) Short-term accident and Resolved in SSER 4 --
procedures review, I.C.1, and SSER 5 I.C.7, I.C.8 i (42) Training during low power Resolved in SSER 5 --
testing, I.G.1 l (43) Review ESF values, II.K.1.5 Resolved in SSER 1 --
Clinton SSER 7 1-6 4
SSER 7 Issue Status Section(s)
(44) Operability status, II.K.1.10 Resolved in SSER 1 --
(45) HPCI and RCIC initiation Resolved in SSER 4 --
levels, II.K.3.13 (46) Isolation of HPCI and RCIC, Resolved in SSER 4 --
II.K.3.15 (47) Qualification of ADS Resolved in SSER 5 --
accumulators, II.K.3.28 (48) Plant-specific analysis, Resolved in SSER 3 --
II.K.3.30 (49) ODYN analysis for River Bend Resolved in SSER 1 --
as applied to Clinton (50) Conformance evaluation report Resolved in SSER 3 --
for loose parts monitoring system (51) Requirements of NUREG-0313 Resolved in SSER 1 --
(52) Control room habitability - Resolved in SSER 1 --
chlorine gas (53) Debris screen design Resolved in SSER 2 --
(54) Verification of adequacy of Removed from list in --
fire protection systems SSER 1 (55) Flood proof door Resolved in SSER 2 --
(56) Valves in fire protection Resolved in SSER 1 --
water supply system (57) Break in water supply piping Resolved in SSER 1 --
(58) Test data on fire ratings Resolved in SSER 3 --
(59) Three-hour-fire-rated Resolved in SSER 3 --
penetration seals (60) Ir. stall fire protection Resolved in SSER 3 --
equipment (emergency lighting)
(61) Fire protection administrative Resolved in SSER 1 --
controls and training Clinton SSER 7 1-7
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SSER 7 Issue Status Section(s)
(62) Technical Specification on Resolved in SSER 1 -- '
fire protection (63) Periodic leak testing of Resolved in SSER 1 --
pressure isolation values (64) Sedimentation in UHS Resolved in SSER 1 --
- (65) Protection against postulated Resolved in SSER 1 --
piping failures (66) Steam bypass of the suppres- Resolved in SSER 5 --
sion pool (LRG II Issue 3-CSB)
(67) Presarvice inspection program Resolved in SSER S --
(68) Inservice testing of pumps Resolved in SSER 6 --
and valves (69a) Containment low-volume purge Resolved in SSER 5 --
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1 (69b) Low-volume purge valve Resolved in SSER 5 --
operability (70) Seismic analysis Resolved in SSER 3 --
(71) Humphrey concerns Resolved in SSER 6 --
1.11 License Conditions ,
i 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 l ensure that NRC requirements are met during plant operations. The status of l potential license conditions in the SER suppletents is as follows: SSER 1 identified two potential license conditions (10 and 11); SSER 2 also identified two potential license conditions (12 and 13) and imposed additional requirements i on one potential license condition (6); SSER 3 identified one potential license
! condition (14); SSER 4 resolved one potential license condition (8); SSER 5 identified six potential license conditions (15-20) and resolved ten potential
- license conditions (1-4, 6, 7, 9-11, and 13); SSER 6 identified three potential license conditions (21-23) and resolved four potential license conditions (5, 14, 18, and 19); and SSER 7 added one potential license condition (24). The potential licensing conditions remaining in SSER 7 (12, 15, 16, 17, and 20-24) will appear as license conditions in the low power license.
SSER 7 Issue Status Section(s)
(1) Staffing DeWitt pumping station Resolved in SSER 5 --
Clinton SSER 7 1-8
SSER 7 Issue Status Section(s)
(2) New stability analysis before Resolved in SSER 5 --
second cycle of operation (3) Postaccident monitoring Resolved in SSER 5 --
(4) Vacuum relief valve position. Resolved in SSER S --
indication (5) Hydrogen management Resolved in SSER 6 --
(6) Postaccident sampling, II.B.3 Resolved in SSER 5 --
(7) Diesel generator reliability Resolved in SSER 5 --
(8) Kuosheng-1 test program Resolved in SSER 4 --
(9) Visual examination of Resolved in SSER 5 --
discharged fuel (10) Measurement of groundwater Resolved in SSER 5 --
level Resolved in SSER 5 (11) Security --
(12) Inservice inspection Addressed in SSER 5, --
remains a license condition (13) Control of heavy loads Resolved in SSER 5 --
(14) Transportation accidents Resolved in SSER 6 --
(15) Fuel zone level channels Addressed in SSER 5, --
remains a license condition (16) Partial feedwater heating Addressed in SSER 5, 15.1 modified in SSER 7, remains a license condition (17) Plant operator experience Addressed in SSER 5, --
remains a license condition (18) Emergency facilities and Resolved in SSER 6 --
equipment (19) Environmental protection plan Resolved in SSER 6 --
for Unit 2 site Clinton SSER 7 1-9
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SSER 7 Issue Status Section(s)
(20) Detailed Control Room Design Addressed in SSER 5, --
i Review remains a license '
condition (21) Safety parameter display system Addressed in SSER 6, --
remains a license condition
! (22) Fire protection Addressed in SSER 6, --
I remains a license condition (23) Control system failures Addressed in SSER 6, --
remains a license condition (24) New-fuel storage Added in SSER 7, 9.1.1 remains a license
. condition I
1.13 Contentions On January 29, 1985, a joint stipulation agreement resolving the remaining two contentions related to quality assurance and control room design contentions was signed by all parties involved [ Illinois Power Co. (IPC), U.S. Nuclear Regulatory Commission (NRC), Prairie Alliance, and Illinois Attorney General]
and submitted to the Atomic Safety and Licensing Board (ASLB) withdrawing in their entirety these two conter,tions. On February 26, 1985, the ASLB approved the joint stipulation signed by all parties related to these contentions and f terminated the proceeding. As part of the settlement agreement, the Illinois Attorney General was permitted to participate in meetings between the NRC staff and IPC related to quality assurance and control room design issues until the
, two issues were resolved. In a letter dated July 24, 1986, the applicant re-
- l. ported the status of the issues contained in this joint stipulation agreement.
All of the issues requiring completion before fuel load have been or will be completed and the few items remaining beyond fuel load are addressed by license conditions. Appendix P of this supplement provides more detailed information related to the status of the specific issues addressed in the joint stipulation agreement.
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Table 1.2 Actions required at Clinton Power Station, Unit No. 1, based on generic implications of ATWS events at the Salem plant (Generic Letter 83-28)
Action No. Action SSER Section 1 POST-TRIP REVIEW 1.1 Program Description and Procedure 15.2.2.1 (SSER 5) 1.2 Data and Information Capability 15.2.2.1 (SSER 6) 2 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE 2.1 Reactor Trip System Components 2.1.1 Equipment Classification (Part 1) 15.2.2.2 (SSER 6) 2.1.2 Vendor Interface (Part 2) *
- 2. 2 Programs for All Safety-Related Components 2.2.1 Equipment Classification (Part 1)
- 2.2.2 Vendor Interface (Part 2)
- 3 POST-MAINTENANCE TESTING 3.1 Reactor Trip System Components 3.1.1 Results of Review of Test and Maintenance 15.2.2.3 (SSER 5)
Procedures and Technical Specifications 3.1.2 Results of Check of Vendor and Engineering 15.2.2.3 (SSER 5)
Recommendations 3.1.3 Identify Post-Maintenance Test Requirements 15.2.2.3 (SSER 5) in Existing Technical Specifications Which Degrade Safety 3.2 All Other Safety-Related Components 3.2.1 Submit Report Documenting the Extension of 15.2.2.3 (SSER 5)
Test and Maintenance Procecedures and Technical Specifications Review 3.2.2 Submit Results of Check of Venocr and 15.2.2.3 (SSER 5)
Engineering Recomendations 3.2.3 Identify Post-Maintenance Test Requirements 15.2.2.3 (SSER 5) in Existing Technical Specifications 4 REACTOR TRIP SYSTEM RELIABILITY 4.1 Vendor-Related Modifications N/A 4.2 Preventive Maintenance and Surveillance N/A Program for Reactor Trip Breakers 4.3 Automatic Actuation of Shunt-Trip Attachment N/A for Westinghouse and B&W Plants Clinton SSER 7 1-11
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i Table 1.2 (Continued)
Action No. Action SSER Section 4.4 Improvements in Maintenance and Test N/A Procedures for S&'n' Flants 4.5 System Functional 4.5.1 Test Diverse Trip Features 15.2.2.4 (SSER 5)
, 4.5.2 Justify Not Modifying the Reactor Trip
- System to Permit Periodic On-Line Testing 4.5.3 Review Existing Intervals for On-Line
- Functional Testing Required by Technical Specifications Notes: N/A = not applicable to Clinton.
- Staff's review is ongoing (SSER 6, Section 15.2.2).
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2 SITE CHARACTERISTICS t
2.2 ,N,eapnindus,tga,1drysyortation,andMilitaryFacilities 2.2.2 Nearby Facilities The SER states that Revere Copper and Brass, Inc. stores propane gas which is
- transported to the Revere facility by trucks. Since the SER was issued, the
Inc. now uses natural gas instead of propane at its site. The propane is
, transported by pipeline. In addition, the SER states that the Charles Todd
! Industrial Uniform Services stores a maximum of 12,000 gallons of dry cleaning solvent. This has since been changed in the FSAR to 15,000 gallons. Both of l these facilities are more than 5 miles away from the plant.
The staff has determined these changes do not affect the staff's conclusion in
- the SER, and that the types and quantities of materials stored or used at these I locations, although different from those reported previously, pose no undue risk to the plant.
2.6 Stability of Subsurface Materials and Foundations 2.6.1 Site Conditions
- 2. 6.1.1 General In Supplement 5 to the SER, the staff, in response to the applicant's motion to j terminate proceedings for Unit 2 before the Atomic Safety and Licensing Board (ASLB), stated that it was satisfied with the condition of the Unit 2 site sub-ject to two site redress conditions. One of these redress conditions was related l to the construction of a berm around the Unit 2 excavation in accordance with j the applicant's FSAR commitments. By letter dated August,5, 1986, the applicant
- stated that (1) the berm around the Unit 2 excavation is,51n general, about
( 6 inches lower than the elevations shown in FSAR Figure 2.5-484 and (2) one area next to the radwaste building is about 12 inches below the elevations shown in the FSAR figure. The applicant has stated that the berm height rec,uirements l will be eliminated on this figure in the next FSAR amendment because Clinton i
Unit 1 is designed for flooding of the Unit 2 excavation in the event that the berm should fail. The staff reviewed the applicant's flooding analyses in Section 2.4.2.2 of Supplement 2 to the SER and found it acceptable. The staff agrees with the applicant that the berms are not necessary to protect Unit 1 l from flooding or hydrostatic loading during a local maximum precipitation event.
The ASLB order on terminating Unit 2 mentioned the berm and stated it should
- prevent a vehicle from accidentally going over the rim of the excavation.
Although the design purpose of the berm was to divert floodwater, the staff agrecs with the applicant that the berm, as built, is still sufficient to pre-vent a vehicle from accidentally going over the rim of the excavation.
Clinton SSER 7 2-1 6
2.6.4 Stability of Slopes The safety evaluation of the slopes of the ultimate heat sink (UHS) pond was contained in the SER. Sections XX', HH', and YY' are critical sections on the shore of the UHS and were analyzed for stability. Recently the applicant found a more limiting worst-case condition and recalculated the factors of safety.
Amendment 35 to FSAR Section 2.5.5.2.3.8 contains the results of the revised stability analyses. The revised computed factors of safety are presented below.
Factor of safety Conditions Section XX' Section HH' Section YY" End of construction Static 2.60 --
2.42 Pseudostatic 1.24 --
1.21 Normal cooling lake Static 2.33 2.32 2.15 Pseudostatic 1.07 1.02 1.03 Sudden drawdown Static 1.96 2.16 2.09 Section HH' was not analyzed for the end-of-construction condition because the fill was already in place and stable. The staff agrees with the applicant's conclusion that the UHS slopes are stable and possess an adequate factor of safety.
i Clinton SSER 7 2-2
3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment 3.10.1 Seismic and Dynamic Qualification In SSER 5, the staff presented a list of generic as well as equipment-specific concerns related to the %ncant's seismic qualification program, which were identified during the plant site audit conducted on August 26 through August 30, 1985, by the Seismic Qualification Review Team (SQRT). The SQRT conducted a second audit on January 28, 1986, at the office of Sargent & Lundy Engineers (S&L), the applicant's consultant, in Chicago, placing primary consideration on the qualification of active valves. Specifically discussed was the availability of dynamic similarity between the various valve groups as installed at Clinton and the corresponding valves that have been actually tested. This auditing re-vealed that the qualification packages for active valves were deficient in this information. At the end of the second audit, the SQRT asked the applicant to upgrade the active valve qualification packages by incorporating proper docu-mentation of the information on dynamic similarity.
Several additional meetings and conference calls, between the SQRT and the ap-plicant, were held after the second audit to further clarify the issue of seis-mic qualification for active valves, in particular, and the overall seismic qualification program, in general.
The applicant's continuing efforts, as documented in its submittals of April 18 and June 23, 1986, have resolved most of the SQRT concerns, including the issue on active valve qualification. Several concerns remained to be resolved, how-ever; namely, the adequacy of qualification packages of nuclear steam supply system (NSSS) equipment, the fatigue cycling effects on hydraulic control unit (HCU) components, and the maintenance procedure for pcssible loosening of HCU bolts during an operating basis earthquake (OBE), which were detailed in SSER 5.
In the April 4, 1986, submittal, the applicant stated that all the NSSS seismic qualification packages had been reviewed to ensure that they include all perti-nent qualification information. On this premise, the SQRT conducted a sample audit on one NSSS equipment package, the reactor core cooling system control panel 1H13-601. The package was subsequently found to be deficient in that pertinent information regarding dynamic similarity between the Clinton panels l (including devices housed inside the panels) and the panels actually tested was not provided nor documented to the staff's satisfaction. In the meeting i with the SQRT held on June 6, 1986, in Bethesda, Md., the applicant committed l to improve all the NSSS equipment packages to meet the staff's requirement for
! documentation. On July 1, 1986, the applicant provided the SQRT with the re-
, quired similarity evaluation. This, together with a subsequent letter from the l General Electric Co. (GE) to the applicant, dated July 14, 1986, confirming the j configuration of the tested panel, sufficed to resolve the SQRT concern and to i close the issue.
l Clinton SSER 7 3-1 I
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Regarding the concern about the HCU components, the GE Qualification Report on the HCU has been incorporated into the Equipment Qualification Report for the i HCUs (SQ-CL534). As stated in the applicant's letter of August 5, 1986, the
- report now includes a fatigue analysis which demonstrates a qualified life of 6 years for the components. The same letter also indicates that Equipment Qualification Report MEQ-CLO75 has been revised to require the inspection and tightening of HCU bolts following any earthquake which exceeds OBE levels.
This requirement has also been incorporated into the Equipment Qualification Maintenance Manual.
In addition, the applicant has confirmed in the August 5,1986, letter that the similarity evaluations for all the NSSS equipment have been completed and have been incorporated into the appropriate qualification packages.
i The last remaining SQRT concern is the completion of the seismic qualification of all safety related equipment. The applicant stated in its letter dated September 12, 1986, that the only equipment items that will not be seismically qualified before fuel load are the thermal cutout temperature switches in the i;
standby gas treatment system (VG). According to the applicant, a plant modi- 1 fication (VG-09) was performed to use United Electric (UE) temperature switches i (Model F400) that meet ANSI N509 Standards, but are not environmentally or seis-mically qualified. The Technical Specification operability of the VG system is supported by these switches. The applicant states that although these specific
, UE temperature switches have not been seismically qualified, seismic testing of i
a similar UE400 series switch (UE model F402D-5BS) has been completed and is documented in the Clinton Qualification Report SQ-LC346. The non qualified switch is similar to the qualified switch since they are both "F" type switches, have the same mode of operation, and have the same weight, dimensions, materials, j and weight distribution. During a seismic event, these switches may have con-
! tact chatter lasting 15 milliseconds. The effect of the contact chatter is that
! it may trip the heater on and off. The effect of the heater being off for this l short period is negligible and does not have a detrimental effect upon the effi-
- ciency of the standby gas treatment system filter units. The s,taff finds this l
justification for interim operation acceptable.
I
- The staff has reviewed all the information provided by the applicant and con-
- cludes that all the SQRT concerns relating to the equipment seismic qualifica-
- tion program have been satisfactorily resolved.
! On the basis of the satisfactory resolution of the specific findings and generic
- comments from the SQRT site visit, the staff concludes that the seismic and dynamic qualification program for safety-related equipment at Clinton meets the applicable portions of General Design Criteria (GDC) 1, 2, 4, 14, and 30 of
- . Appendix A to 10 CFR 50, Appendix B to 10 CFR 50, and Appendix A to 10 CFR 100.
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I l Clinton SSER 7 3-2 l
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4 REACTOR 4.3 Nuclear Desian 4.3.2 Design Description 4.3.2.7 Vessel Irradiation 4.3.2.7.1 Evaluation Historically, vessel fluence calculations have been subject to large uncertain-ties, and pressure vessel materials sample surveillance programs have been required, in part, to confirm predictions of vessel fluence and lifetime. Accu-rate calculations of vessel fluence are made difficult by several factors. The neutron fluence is attenuated by three to four orders of magnitude between the core and vessel inner wall and is sensitive to nuclear cross-section data and numerical solution techniques. Also, the pressure vessel fluence is sensitive to the core / shroud / vessel configuration geometry, material compositions, and assumptions in modeling the core neutron source. Vessel fluence calculations have improved rapidly in recent years from simplified kernel diffusion tech-niques to multidimensional higher order discrete ordinates schemes,' capable of relatively accurate predictions of vessel fluence.
The Standard Review Plan (NUREG-0800) gives no specific criteria that must be met by the analytical methods or data that are used by the applicant. It states, in general, that analytical methods and data base should be representa-tive of the state of the art and that the experiments used to validate the ana-lytical methods should be adequate and should encompass a sufficient range.
The only specific criteria in NUREG-0800 are that the calculations be performed by higher-order approximation than diffusion theory and that the geometric model-ing be detailed enough to properly estimate the relative flux spectra at vari-ous positions from the reactor core boundary into the pressure vessel wall.
The applicant, on its own initiative, has upgraded the original one-dimensional calculations of neutron fluence at the vessel wall and has also provided addi-tional information on the details of the calculations and the methods used to validate the analytical methods. The calculations were made using the programs DOT IV (from Oak Ridge) and SNID (based on ANSIN).
4.3.2.7.2 Transport Calculations The neutron transport calculation should be carried out with sufficient spatial detail and with a complete representation of the materials present in the neutron field. In the present case, calculations were carried out using a two-dimensional calculation, discrete ordinates, S ntransport code with P3 anisotropic scatter ing. The Ss angular quadrature used has been found to be adequate for pressure vessel flux calculations. The discrete ordinates code was used in a distributed source mode with cylindrical geometry. The geometry described seven regions with the core modeled as two homogenized regions. The coolant water region Clinton SSER 7 4-1
between the core and the shroud contained saturated water at 550*F. Subcooled water at 530 F and 1040 psi was used for the coolant between the shroud and the vessel. The model of the material compositions for the stainless steel shroud and the alloy steel vessel contained the mixtures by weight as specified in the ASME material specifications (ASME Boiler and Pressure Vessel Code, Section 2,
" Materials Specifications") for ASME SA240 (304L) and ASME SAS33 Grade B, respectively.
In the region between the shroud and the vessel, the presence of the jet pumps was ignored. This is justified because the maximum fluence to the vessel occurs in the top half of the reactor core, and the jet pump inlet nozzles are not at the maximum fluence elevation. Only the inlet recirculation risers are at that elevation and because the area blanketed by the risers is small and does not include the peak fluence point, they were not modeled. The major effect of including the jat pumps would be a slight reduction in the calculated fluence, a result of substituting 7.90 g/cc steel for 0.74 g/cc water. The wall thick-ness of.the jet pump is less than inch and, therefore, this could possibly result in a 20% reduction to a highly localized portion of the vessel. The more conservative approach was to ignore the jet pumps.
The cross-section file, with angular scattering approximated by a third order Legendrd expansion, was taken from a General Electric Co. file, GMUG. This file was based on Los Alamos Scientific Laboratory (LASL) library LIB-IV and is, therefore, based on END F/8-IV. The liorary is reduced to fewer groups by using a program that performs a one-dimensional diffusion calculation for'each region.
The code-calculated diffusion spectrum is used to weight the cross-sections.
4.3.7.2.3 Neutron Source Modeling The source in the two-dimension DOT problem was represented in an (R,0) geometry for a 1/4-core model. In this model, the outer four layers of Dundles are modeled by assigning the appropriate power densities to the (R,0) intervals. Selection of the appropriate power densities is difficult because the location of the axial peak and the relative bundle powers change during the cycle. The axial location is controlled by the outer two layers of bundles. The maximum location is identified by running a series of one-dimensional cases for the axial power nodes to locate the maximum. Given the location of the maximum, the two-dimension case is modeled at that node.
The representation of the neutron source should reflect the geometric and operational characteristics of the core during the time the neutron fluence is accumulated. An exact calculation of the true accumulated neutron fluence would require knowledge of the exposure and power history of all fuel assembly loca-tions. It can be seen that the estimation of end-of-life fluence exposure at the beginning of service requires many assumptions. Therefore, the neutron fluence is periodically recalculated from core physics data obtained during service and reevaluated from the dosimetry in the reactor vessel surveillance capsules as they are pulled. For these initial calculations, a distributed source, separated in space and energy, was obtained from the core power shape and a neutron spectrum calculated from cross-sections originally based on LASL data. The integral over space and energy was normalized to the total number of neutrons in the core region. Mid-cycle data for an equilibrium fuel cycle were used as representative of conditions throughout the cycle. As a conserva-l tism, a general 30% increase was applied to the entire analysis.
Clinton SSER 7 4-2
4.3.2.7.4 Validation of Methods The analytical methods were validated by comparison to measurements made in the Browns Ferry reactor. The results of calculation by the codes used for Clinton have been compared to measurements from vessel surveillance capsule flux wires and the Browns Ferry in-vessel flux monitors. The flux monitors were located at the shroud in the center of the downcomer region and at the vessel wall.
These data provide both magnitude and attenuation shape in the water.
Because surveillance sample data comparisons to measured flux at equivalent lo-cations in equivalent plants show spread in the data, a safety factor was ap-plied that forces the calculation to envelope the range of surveillance flux wire measurements. For the Clinton calculations, the flux and fluence values were increased 30% to cover uncertainties. The staff finds this to be accept-able for preservice calculations because, as pointed out above, the calculations will be refined as part of the vessel surveillance program during service life.
4.3.2.7.5 Conclusions The neutron fluence, as determined from the calculated cycle average flux by as-suming that the plant is operated 90% of the time at 90% power level for 40 years, is 7.3 x 101s neutrons /cm2 at the vessel wall and 5.5 x 101s neutrons /cm2 at 1/2 T (thickness of vessel wall) for energies greater than 1 Mev. On the basis of the review as discussed above, the staff considers the analytical methods and data base used to be representative of the state of the art and the methods used to validate the calculations to be adequate and, therefore, acceptable. The staff '
will periodically further review the results of the vessel surveillance program and the adequacy of the pressure and temperature operating limits which comprise a part of the plant Technical Specifications.
4.6 Functional Design of Reactivity Control System In Section 4.6 of the SER, the staff stated that the applicant had committed to conduct preoperational tests to verify the flow rate of the control rod drive (CRD) hydraulic system to determine leakage (return) flow to the reactor vessel, and to verify the proper operation of the CRD system. The staff concluded at that time that the CRD hydraulic system was acceptable, pending satisfactory conclusion of the tests demonstrating adequate return flow to the reactor vessel, as noted in Section 8 of NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking." In a letter dated September 3,1982 (D. L.
- Holtzscher to T. A. Novak), the Licensing Review Group II (LRG-II), of which
! the applicant is a member, recommended that the CRD system makeup test not be performed. In a letter dated September 23, 1982, the applicant incorporated the LRG-II position (identified as Generic Issue 5-ASB) into the Clinton license application, i
j The CRD system makeup test was the result of the staff's evaluation of boiling-
l The results were published in NUREG-0619 and resolve Generic Technical Activity
- A-10. NUREG-0619 included proposed methods for eliminating the nozzle cracking which had occurred in the CRD return line nozzle at the reactor pressure vessel.
All applicants undergoing licensing review for BWRs designed and constructed without the CR0 return line were asked to complete specific flow rate testing.
Clinton SSER 7 4-3 i
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The flow rate tests were to demonstrate that the amount of water from the CR0 system forced to return as leakage around the control rod seals is satisfactory.
This flow test was required as a result of (1) the Browns Ferry fire of March 22, ,
1975, where the CRD flow path provided the only water source to the reactor core for a limited time and (2) the low-level event at Oyster Creek, where this sys-tem once again briefly provided the only water to the core. This flow test was Recommendation (6) of Section 8.1 of NUREG-0619. The criterion to be met (base case) for determining adequate flow was the flow that is required 40 minutes after a reactor scram to keep the reactor levtl above the top of the active fuel, assuming that the level was at its normal value at 40 minutes.
The LRG-II position indicated that the conclusions of an evaluation for LRG-I plants concerning this topic (submitted by letter dated March 26, 1982, P. L.
Power to H. Faulkner) were applicable to the LRG-II plants. The LRG-I evaluation concluded that the present CRD system designs were not capable of meeting this recommendation and that, because of plant modifications, the requirement for a CRD system makeup flow test is no longer necessary. These modifications include (1) plant fire prevention / protection and separation enhancement, (2) development of symptom-oriented emergency procedures guidelines, and (3) post-TMI emergency core cooling system modifications. The staff's review and acceptance of the LRG-I evaluation are documented in the Susquehanna Station's Safety Evaluation Report (NUREG-0776), Supplement 3, Section 4.6.2.
The staff has reviewed the LRG-I evaluation and concludes that the evaluation is applicable to the LRG-II plants (including Clinton). The CRD systems of the LRG-II plants are similar in design and function to system designs used in the LRG-I evaluation. The use of the CRD system as an emergency makeup system was not a design requirement and is beyond the design basis for any of the LRG-II plants. The redundant emergency core cooling systems were designed to meet all applicable staff guidance for emergency vessel makeup functions.
All applicants for operating licenses must now demonstrate the capability to reach cold shutdown following a fire in any area of the plant, including such sensitive areas as the control room and the cable spreading room. This guidance is contained in Standard Review Plan Section 9.5.1, NUREG-0800. The staff spe-cifically reviews plant designs to ensure that no fire is capable of damaging redundant safe shutdown equipment, or where such is not feasible, that alterna-tive or dedicated safe shutdown capability is provided. Thus, fire protection enhancement is similar for both groups of plants.
As a result of the TMI-2 accident, the operator guidelines are now " symptom oriented" and have clear instructions on what raakeup systems are available and in what order they should be used no matter what the initiating event may be.
The CRD system is listed in these guidelines as one of the available high-pressure makeup systems. Additionally, the emergency core cooling system (ECCS),
including high pressure core spray (HPCS) systems and the reactor core isolation cooling (RCIC) system, have been upgraded by such features as automatic reset (on the RCIC) and automatic restart (on the HPCS) in order to make these systems more reliable. Thus, modifications and procedures are similar for both groups of plants.
Thus, on the basis of its previous LRG-I evaluation and its determination that this evaluation is applicable to the LRG-II plants, the staff concludes that
! Clinton SSER 7 4-4
9 the LRG-II position (Generic Issue 5-ASB) is acceptable. The staff, therefore, concludes that the CRD system is acceptable without the flow rate test, provided preoperational testing of the system design function is still performed.
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i Clinton SSER 7 4-5
6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.2 Secondary Containment As discussed in the SER, the results of the applicant's secondary containment drawdown analysis shows that approximately 194 seconds after the onset of a design-basis LOCA, the secondary containment pressure is reduced to the preacci-dent level of -0.25 inch of water gauge. A recent reanalysis by the applicant (FSAR Amendment 36) results in a drawdown time estimate of 188 seconds after the onset of the loss-of-coolant accident (LOCA). The staff conclusions regarding acceptability remain unchanged.
However, the applicant intended to apply a 168-second drawdown time (i.e., with-out diesel electrical delays) as a technical specification limit for testing secondary containment inleakage during normal surveillance periods of at least every 18 months. The staff had discussed with the applicant the inappropriate-ness of such a value as a surveillance requirement. Since the drawdown time was derived with consideration of conditions following the onset of a LOCA which would mechanistically increase the drawdown time (e.g., by heat addition to the secondary containment atmosphere), it would be nonconservative to apply the 168-second value to the normal testing conditions. The applicant has recog-nized this inconsistency, and has committed to analyze the drawdown period under actual test conditions and to provide an appropriate drawdown time estimate at least 60 days before the second fuel cycle begins. This commitment is in the form of a footnote to the relevant surveillance requirement contained in Sec-tion 4.6.6.1.C.1 of the Technical Specification. Because of the detailed nature of the proposed reanalysis, the staff finds the deferment of a final value at this time to be acceptable.
In the interim, the applicant has conducted preoperational testing of the standby gas treatment system (SGTS). The test results demonstrate a nonaccident drawdown time of approximately 35 seconds, thereby adhering to the relevant surveillance requirements. NRC Region III personnel have reviewed the preoperational test and have found it acceptable. Therefore, the staff finds there is a sufficient basis to defer the inclusion of a drawdown value in the Clinton Technical Speci-i fication that reflects only inleakage into the secondary containment until the j first refueling outage, since the staff judges that a nonaccident inleakage
- drawdown time of 35 seconds appears reasonable.
l t
6.2.5 Combustible Gas Control l
! As discussed in the SER, the applicant performed calculations of the post-l accident containment hydrogen concentration in the drywell and in the contain-
! ment to assess the hydrogen recombiner system. As stated in the SER, the cal-culated maximum drywell and containment hydrogen concentration was approximately I 3.5 and 3.0 volume percent respectively.
)
i Clinton SSER 7 6-1
As a result of a recent reanalysis (FSAR Amendment 36), the calculated maximum drywell and containment hydrogen concentrations are approximately 4.0 and 3.7 dry volume percent (i.e. , the effect of steam dilution is neglected), respec-tively. This reanalysis is based primarily on the updated zinc and aluminum inventory inside the Clinton containment. The staff conclusions regarding acceptability remain unchanged.
Clinton SSER 7 6-2
7.. . . -. , - . - . .
7 ItlSTRUMENTATION AND CONTROLS 7.3 Engineered Safety Features Systems 7.3.3 Resolution of Issues 7.3.3.7 Engineered Safety Features Reset Controls NUREG-0737, Item II.E.4.2, requires for control of automatic isolation valves that resetting of the isolation signal will not result in automatic reopening of the isolation valves. In Supplement 6 to the SER, the staff stated the applicant has modified the control circuitry for most of the equipment so that returning safety-related equipment from its emergency mode to its normal operat-ing status requires deliberate operator action after an engineered safety fea-ture (ESF) reset.
By letter dated August 8, 1986, the applicant stated that during startup testing it was discovered that the control logic for the operation of four 36-inch con-tainment isolation valves does not meet the above requirements. The applicant stated in the August 8, 1986, letter that a plant modification is in progress that will provide a seal-in circuit at each individual valve control and, to be consistent with the requirement in Appendix D of the FSAR, each valve control will have to be individually manipulated to open the valve after a high radia-tion isolation signal. The modification will be completed before reactor heatup. Since these valves will be locked closed until the modification is installed and the operability of these valves is not required for fuel loading activities, the staf* finds this modification and the schedule for completing this modification acceptable.
Clinton SSER 7 7-1
0 ELECTRIC POWER SYSTEM 8.2 Offsite Power System 8.2.1 General Description In its SER, the staff described an offsite power system that was designed to provide power for two-unit operation at Clinton. Subsequently, Clinton Unit 2 was cancelled and the offsite power system has been modified for single-unit operation.
The Clinton offsite power system still consists of a 345-kV switchyard and a separate 138-kV three-terminal transmission line which terminates directly (through a circuit switcher) at the emergency reserve auxiliary transformer.
Three transmission lines terminate at the 345-kV switchyard which also serves as the connection point for the unit's main generator and the reserve auxiliary transformer. The second reserve auxiliary transformer has been eliminated.
The 345-kV switchyard is in a ring bus configuration that has intervening circuit breakers between all the connections. The two sources of offsite power to the engineered safety feature (ESF) buses are through the emergency reserve auxil-fary transformer and the reserve auxiliary transformer. The single point failure resulting from a line crossover (previously discussed in the staff's SER) that could result in a total loss of offsite power to the station for 13 seconds, has now been eli.ninated.
The reserve auxiliary transformer is sized to carry the auxiliary load required for Clinton in addition to the total coincidental auxiliary load required if a loss-of-coolant accident (LOCA) occurred. The emergency reserve auxiliary transformer is designed to start and carry the auxiliary load required if a LOCA occurred. The staff has reviewed the current offsite power system utilized for single-unit operation and finds it acceptable.
8.2.2 Load Flow and Grid Stability Analysis The applicant has revised its load flow and grid stability analysis to reflect the single-unit configuration at Clinton. For steady-state stability limit, the applicant has determined that a substantial margin exists between system requirements and the unit stability limit. The transient stability analysis performed, which included sudden loss of Clinton Unit 1 in 1987, as well as various three phase line faults and double-line-to ground faults, concluded that the offsite system remains stable for all the cases. The staff finds the results of these analyses acceptable.
8.3 Onsite Emergency Power Systems 8.3.1 AC Power System I
Independence of the Electrical Protection Assembly In the Clinton SER, the staff described the nuclear systems protection system (NSPS) power supplies to the reactor protection system (RPS) scram solenoids.
Cifnton SSER 7 8-1
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The supplies consist of two Class 1E qualified inverters each with its own power monitor. The power monitor, which is also called an electrical protec-tion assembly (EPA), must be independent of the inverter in order to prevent an overvoltage, undervoltage, or underfrequency condition on the output of the inverter from creating a single failure of the scram system. i During its site visit, the staff questioned the independence of the EPA from the
- inverters. The EPA consisted of a printed circuit card which was located in the i same card cage assembly together with the inverter logic cards. The staff's position was that, in accordance with Section 5.6.2 of IEEE Std. 384-1974, the
- separation distance between the inverter and EPA components should be 6 inches, i
or a barrier should be provided, or the existing separation should be demon-t strated to be adequate by analysis based on tests,
. In a letter dated June 17, 1986, the applicant informed the staff that the RPS NSPS power supplies had been modified to satisfy the staff's concerns on inde-l pendence. The applicant stated that the EPA has been mounted on a separate mounting plate inside the cabinet, satisfying physical and electrical indepen-l dence, and the interface wiring is now separated by more than 6 inches.
i The staff finds these modifications acceptable.
In the Clinton SER, the staff stated that each of the Division I and II diesel generators is capable of attaining rated voltage and frequency within 10 seconds
{ after receiving a starting signal. In a recent amendment to Section 9.5.6.1.1
- of the FSAR, the applicant stated that within 12 seconds after receipt of the i start signal the Division I, II, and III diesel generators are operating at I
rated speed, voltage, and frequency. The 12-second start has been incorporated into the operational sequence of emergency core cooling systems for design-basis accident (Table 6.3-1 of the FSAR). The 12-second start time has also been
- factored into the Clinton Technical Specifications which the staff has reviewed and approved. This change is, therefore, acceptable.
Motor-Operated Valves i
Motor-operated valves (MOVs) at Clinton have minimum specified start?.7 voltages l of 90%. The staff, therefore, asked the applicant if the; motor operated valves l
i would function properly while being supplied from the diesel generators, since there are large voltage transients during load sequencing on the diesel genera-i tors which will be below the 90% minimum voltage requirement specified for the i
MOVs. In a letter dated August 28, 1986, the applicant, using information obtained from the preoperational test for loss of offsite power / emergency core l cooling systems (PTP-LE-01), demonstrated that the diesel generator voltage would recover in enough time to satisfactorily start the MOVs in enough I,
time to meet their safety requirements. The staff has reviewed the applicant's j response to its question and finds it acceptable.
HPCS Diesel Generator (Division 3) Alarms in the Main Control Room i
In the Clinton SER, the staff stated that the conditions which render the high pressure core spray (HPCS) diesel incapable of responding to an emergency start signal are alarmed in the main control room window: "HPCS Not Ready for l
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[
- Clinton SSER 7 8-2 r
1- - -
Auto Start Breaker in Lower Position". In subsequent revisions to the FSAR, the staff identified a condition, control power failure, which made the HPCS diesel incapable of responding to an emergency start condition but is alarmed in the window " Diesel Generator Trouble". The staff was concerned that the
" trouble" alarm is used for many diesel generator events which may or may not place the HPCS diesel out of service, and the labeling is also inconsistent with the Division 1 and 2 diesel generator alarms.
In a letter dated September 4, 1986, the applicant stated that Clinton Power Station Operations are required to locally verify the cause of either a
" trouble" or "out of service" alarm and are trained to recognize these condi-tions that place the diesel generators out of service. The applicant also committed to perform an engineering evaluation to determine the feasibility of adding an "out of service" alarm to the Division 3 diesel generator alarms in the main control room and to report the results of this evaluation and complete the necessary corrective action (i.e., procedural change or a modification to the Division 3 diesel generator alarms) before startup from the first refueling outage. The staff finds this commitment acceptable.
8.3.2 DC Power System In the Clinton SER, the staff stated that the Division III (high pressure core spray) battery charger can recharge its fully discharged battery to a fully charged state in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The response to staff question 430.128 in the FSAR, however, states that the Division III battery can be fully charged from the design minimum discharge level of 1.75 volts per cell in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or less, and the maximum equalizing voltage will be 137.4 volts. Twelve hours is a reasonable timo to recharge the Division III battery. The staff, therefore, finds this acceptable.
During its site visit (October 29-30, 1985), the staff questioned the seismic qualification of the eyewash station and its associated piping in the Division III battery room. The applicant determined that the eyewash station and piping were not adequately supported. In a letter dated March 17, 1986, the applicant provided information that the supports for the eyewash station and associated piping had been replaced and that calculations have been performed which confirm that the modified condition will provide adequate: structural support for the eyewash station and piping during a postulated seismic event. The staff finds this acceptable.
8.4 Other Electrical Features and Requirements for Saf(ty 8.4.1 Containment Electrical Penetrations f In the SER, the staff stated that the 600-V, 15-Hz, low-frequency motor genera-tor (LFMG) set feed to the reactor recirculation pumps (RRP) is protected by l redundant Class 1E molded case circuit breakers. In a letter dated July 8, l 1986, and in FSAR Amendment 38 in response to staff question 40.4, the applicant t stated that the LFMG set maximum output under fault condition is less than the penetration capability and there is a single high quality circuit breaker on the output of the LFMG. The applicant further stated that, if a fault occurs on the containment side of the penetration, the high quality output breaker, whose 12T (current squared x time) let-through is below both the LFMG set and the penetration, will open to protect both the penetration and the LFMG set.
Clinton SSER 7 8-3
l If the breaker fails to open, the I2T of the LFMG set up to and including
, failure of the set is below the 12T of the penetration. The applicant has also i
provided (1) 12T curves which support these statements and (2) a revised technical specification for ten, ting the circuit breakers. l The LFMG sets are used to power the reactor recirculation pumps at low power i
levels primarily during startup and shutdown. The majority of the time the pumps are fed by the 6.9-kV system which has redundant circuit breakers to protect the penetration. The staff, therefore, finds acceptable the provisions j taken by the applicant to protect the reactor recirculation pump (RRP) penetra-tion during the short periods of operation on the LFMG.
l 8.4.4 Use of a Load Sequencer With Offsite Power l In the SER, the staff discussed the use of load sequencing on the offsite power
! system at Clinton. Subsequently, in FSAR Amendment 36, the applicant provided a revised response to Staff Question 430.106 which indicated that loss-of-coolant accident (LOCA) loads are not sequenced to the offsite power sources (reserve
- auxiliary transtormer and emergency reserve auxiliary transformer). The Class j 1E 4160-V engineered safety feature (ESF) loads are sequenced only for operation from their divisional diesel generators. The applicant had performed block start (no load sequencing) calculations on the offsite power sources during the
, Independent Design Review. These calculations concluded that the safety systems would perform their safety functions under block start conditions at the minimum expected offsite system voltage as determined by the Illinois Power System Planning Department. The minimum expected offsite system voltage used, however, was higher than the setpoint of the second level undervoltage protec-tion relays. The staff, therefore, asked whether the safety systems would perform their safety functions under block start conditions if the voltage at the 4160-V safety-related buses is below the minimum expected value, but above the second-level undervoltage setpoint.
. In its letter dated August 28, 1986, the applicant, using information obtained l from the preoperational test for loss of offsite power / emergency core cooling systems (PTP-LE-01), demonstrated that for voltages just above the second-level undervoltage relay setpoint the safety loads would either transfer to the diesel generators because of the reset band of the relays or, if.the voltage was higher than the reset point, the safety loads would satisfactorily start on the offsite power system. Motor-operated valves which are specified for a starting voltage of 90% would have a delayed start but would start in sufficient time to meet their safety requirements. The staff has reviewed the applicant's response to its questions and finds it acceptable.
8.4.7 Physical Identification and Independence of Redundant Safety-Related Electrical Systems Amendment 37 to the Clinton FSAR stated that cable splicing in trays is not permitted and that an analysis will be performed and sent to the NRC staff whenever cable splices are used in trays. In a letter dated March 17, 1986, the applicant provided such an analysis. The analysis is identified as calculation number 19-BD-25 (Rev. 0) dated February 20, 1986. The purpose of the analysis is to justify splicing of Class 1E cables in conduit. The applicant provided this analysis because Position C.9 of RG 1.75 states that cable splices in raceways should be prohibited, although it goes on to say that splices are not,
- Clinton SSER 7 8-4
by themselves, unacceptable. If they exist, the resulting design should be justified by analysis.
The analysis concluded that (1) splices in power cables are seldom made, (2) splices are made in conduit and not in cable tray, (3) the splice material as well as the cable material are flame retardant, and (4) the splices that are used are qualified to the same environment required by the cable on which they are being used.
On the basis of these conclusions, the staff finds that the applicant's use of cable splicing in conduit is unlikely to degrade the Class 1E system and is, therefore, acceptable.
In Amendment 36 to Section 8.3.1.4.5.1 of the FSAR it is stated that, with the exception of a small percentage of special vendor supplied non-Class IE cables, all non-Class 1E cables in open raceways are flame retardant. In SER Supple-ment 6 (Section 8.4.7), the staff described special electrical separation tests conducted by the applicant which supported the use of closer separation dis-tances for circuits than those distances nomally specified by IEEE Std. 384-1974.
. All the cables used in these tests, however, were flame retardant. The staff was, therefore, concerned that the non-flame retardant non-Class 1E cable refer-enced in FSAR Section 8.3.1.4.5.1 may not have sufficient separation distances to Class 1E circuits. Because the non-flame retardant (non-IEEE 383-1974 qual-ified) cables were not a part of the electrical separation tests, the minimum separation distances specified in IEEE Std. 384-1974 should be maintained be-tween these non-Class 1E circuits and any Class IE circuits near them. In a letter dated July 18, 1986, the applicant reviewed all the non-IEEE 383 quali-fied cables for electrical separation. Non-IEEE 383 qualified cables routed entirely in conduit were not specifically addressed in this review because the staff agreed the non-flame retardant characteristic of these non-Class 1E cables provided little additional threat to Class 1E cables nearby when the non-flame retardant cable is entirely enclosed within the conduit. The applicant did determine in its review that 91 cables are partially routed in trays. Some of these cables have been replaced with IEEE 383 qualified cables that have passed the flame test requirements of the standard. Two czbles are fiber optic cables and the rest meet the IEEE 384 separation requirements. Because the results of the review cemonstrate that the non-IEEE 383 qualified cable installation meets the IEEE 384 separation requirements without relying on the reduced separation criteria, the staff finds it acceptable.
Section 5.1.2 of IEEE Std. 384-1974 requires that cables installed in exposed raceways be marked in a manner of sufficient durability and at a sufficient number of points to facilitate initial verification that the cable has been installed in conformance with the separation criteria. It further requires that these markings be applied to the cable before or during installation.
Amendment 36 to the Clinton FSAR, Section 8.1.6.1.14, states that if color-coded l Mylar wraps are used to achieve divisional identification, the marking may be delayed until completion of the cable pull to avoid unnecessary marking of long l
lengths in the duct runs and conduits. In such cases, cables are marked with i
the divisional color code every five feet or less in the exposed part of the raceway and at every manhole, junction box, and pull box.
Clinton SSER 7 8-5 l
Because the cables referenced in FSAR Section 8.1.6.1.14 do not have their color coding applied until after their installation, they do not meet the requirements of IEEE Std. 384-1974. The staff has reviewed the cable installation procedure used by the applicant's contractor (Baldwin Associates Procedure BAP 3.3.2) and the installation procedure used by the plant staff (CPS No. 8491.01). The Baldwin Associates Procedure contains requirements that the quality control (QC) inspector verify correct cable separation and segregation. There is, however, nothing comparable to this in the plant staff procedure. In a letter dated July 18, 1986, the applicant stated that the staff procedure is being revised to include a QC holdpoint to verify that proper cable routing and separation are checked. The staff concludes that these provisions provide a check on the correct installation of cables that is reasonably comparable to that provided by color coding of the cables before installation.
l Clinton SSER 7 8-6
a 9 AUXILIARY SYSTEMS 9.1 Fuel Storage Facility 9.1.1 New Fuel Storage Containment Fuel Storage Pool J
In a letter dated May 10, 1986, the applicant requested that Materials License '
No. SNM-1886 be amended authorizing the dry storage of the fuel assemblies for the initial core in the containment fuel storage pool. Additional information was provided in a letter dated June 27, 1986. By letter dated July 23, 1986, the staff amended SNM-1886 (Amendment 2) authorizing the dry storage of the fuel assemblies for the initial core in the containment fuel storage pool.
The safety evaluation accompanying Amendment 2 to the SNM license stated the following:
The containment fuel storage pool racks have a design and configura-tion similar to the new fuel storage vault racks (i.e., both racks have a center-to-center spacing of 7 inches within a rack and a center-to-center spacing of 12.25 inches between racks). Although authoriza-tion was given to store the fuel assemblies dry in the new fuel storage vault, only underwater storage of the fuel assemblies in the contain-ment fuel storage pool was authorized. The administrative controls
! limiting the water mist from sprinklers that could enter the storage arrays in the event of a fire provided the nuclear criticality safety margin that allowed the dry storage of fuel assemblies in the new fuel storage vault. No such controls were provided in the containment fuel storage pool. The applicant has now provided controls to limit the water mist that could enter the fuel assembly storage array in the containment fuel storage pool.
However, to emphasize water mist control in fighting fires in the containment fuel storage pool area and to prevent the inadvertent addition of water from other sources into the storage pool, license condition 25 was added to the SNM-1886 license. The staff has determined that this condition equally applies to the 10 CFR 50 operating license until the fresh fuel assemblies for the initial core are no longer dry stored in the containment fuel storage pool.
Accordingly, SNM-1886 (Amendment 2) license condition 25 should be incorporated into the operating license as license condition 24 to read as follows:
Fuel assemblies for the initial core, when stored dry in the contain-ment fuel storage pool, shall be stored so that: (a) no more than 12 rows of fuel assemblies shall remain uncovered during the loading or unloading of fuel assemblies; (b) fire-retardant covers meeting the requirements of National Fire Protection Association (NFPA)
Standard 701 shall cover all other rows containing fuel assemblies during loading and unloading of fuel assemblies; (c) when fuel assem-blies are not being loaded or unloaded, fire-retardant covers meeting Clinton SSER 7 9-1
, the requirements of NFPA Std. 701 shall cover all rows of fuel assem-blies; (d) increased surveillance of the containment fuel storage pool vicinity will be maintained to ensure that the area is kept free of extraneous combustible material; (e) the fire protection hose sta-tions servicing the containment fuel storage pool area shall be equipped with solid stream nozzles before the dry storage of fresh fuel assemblies in the containment fuel storage pool takes place; (f) the containment spray system will be tagged out before the dry storage of fresh fuel assemblies in the containment fuel storage pool occurs in a status that will prevent initiation of a spray or mist; and (g) all other sources of water to the containment fuel area shall either be equipped with solid stream nozzles or shall be locked out and shall be tagged shut with a safety tag.
New Fuel Storage Vault The safety evaluation accompanying the SNM license stated that since it has been estimated by the staff that an array of as few as five rows of assemblies fully flooded with wat r mist at optimum density for maximum k,ff may become critical, the applicant has installed fire hose protection stations equipped with solid stream nozzles to fight fires that may, but are unlikely to, occur in the vault. The applicant has committed to establish station administrative procedures specifying actions to be taken for contrrl of combustibles and con-trol of ignition sources, and to control action to ce taken in the svent of a fire.
Accordingly, a license condition was incorporated with the SNM license and identified as license condition 18. The staff has determined that this con-dition equally applies to the 10 CFR 50 operating license and, accordingly, SNM-1886 license condition 18 should be incorporated into the operating license as a license condition, to read as follows:
All fire hoses servicing the new fuel storage vault shall be equipped with solid stream nozzles.
The license conditions specified in this section of SER Supplement 6 related to the storage of new fuel assemblies are still applicable for other than those fuel assemblies for the initial core that are stored in the containment fuel storage pool.
9.3 Process Auxiliaries 9.3.1 Instrument Air System In the course of an inspection of the Clinton Power Station during the periods of February 24-28 and November 3-7, 1985, Region III staff reported a deviation relating to the instrument air system (IA). This deviation involved a lack of acceptance criteria for the size of particulates in the instrument air and a lack of periodic and postmodification/ repair testing of the instrument air quality.
In a submittal dated June 13, 1986, the applicant responded to these omissions as follows:
Clinton SSER 7 9-2
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(1) The cpplicant, in consultation with air-operated equipment vendors, has developed particulate contamination criteria for air-operated, safety-related components. In view of the particularly critical nature of the valves in the control rod drive system, the applicant assured their con-tinued operability by providing additional 5-micron filters upstream of the valves in order to prevent particulates from entering the valve operator.
(2) The applicant agreed to test the air quality at the system filter discharge, at least yearly, for moisture and particulate content using the criteria of minus 40 F for dewpoint and 3 microns for particulate size.
(3) The applicant will test the air quality near selected safety-related com-ponents for compliance with particulate size criteria permissible for them as recommended by the vendor. Those portions of the IA which do not comply with the appropriate particulate size criteria will be blown clean and retested.
(4) The applicant has a procedure to test instrument air quality after system maintenance, modifications, or repairs.
(5) Component filters will be changed periodically in accordance with vendor recommendations.
(6) Periodic blowdowns of the'IA will be performed. Initially, this will be done no more often than at each refueling. However, the frequency may be increased or decreased as justified by experience.
(7) The applicant will inspect air-operated safety-related components for par-ticulate accumulation when replacing elastomeric parts as dictated by pe-riodic preventive maintenance. The periodicity of sucn replacements will vary from 18 months to several years, depending on the component and its location.
(8) The applicant will not test for oil or other contaminants in the instrument air. The applicant notes that the IA compressors are oil-free. In addition, the plant is located in an agricultural area which is isolated from large industrial petrochemical or chemical plants. Intakes for the IA are lo-cated on the roof of the radwaste building and there are no sources nearby which could release hydrocarbons, gases, or other contaminants likely to affect instrument air quality.
In view of the foregoing, the staff finds that the applicant's plan for periodic testing and maintaining of the IA quality is in accordance with the criteria of SRP Section 9.3.1 and is, therefore, acceptable.
9.3.3 Standby Liquid Control System In a Sys, Will Be Revised|letter dated October 16, 1985]], the applicant indicated that the features for mitigating anticipated transient without scram (ATWS) events were con-sistent with the requirements of the ATWS rule (10 CFR 50.62). These features include the standby liquid control system (SLCS) with an injection capacity of 86 gpm of 9.8 weight percent sodium pentaborate solution.
Clinton SSER 7 9-3
In letter dated May 23, 1986, the applicant discussed a plan to increase the sodium pentaborate concentration in order to reduce the required SLCS ficw rate.
The applicant provided additional clarification in a Record of Coordination dated July 16, 1986, which is docketed in an attachment to a memorandum from B. Siegel to W. Butler, dated July 27, 1986. The weight percent concentration would be increased from 9.8 to 10.2 and the SLCS flow rate would be decreased from 86 gpm to 82.4 gpm. The technical specifications will be modified to re-flect these changes.
The changes proposed by the applicant have been reviewed by the staff against the requirements of the ATWS rule (10 CFR 50.62) and Generic Letter 85-03, "Clarificatiois of Equivalent Control Capacity for Standby Liquid Control System.," dated January 28, 1985. The applicant's proposed increase in sodium pentaborate concentration to 10.2 weight percent in conjunction with a flow l rate of 82.4 gpm for Clinton which has a 218-inch vessel will provide a boron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate for the reference plant which has a 251-inch vessel. This is in compliance with 10 CFR 50.62 and is, therefore, acceptable.
The staff has, therefore, concluded that the applicant's proposed design for the SLCS is acceptable because it will provide a boron content equivalent to 86 gpm of 13 weight percent sodium pentaborate as required by 10 CFR 50.62.
9.3.5 Postaccident Sampling Capability (NUREG-0737, Item II.B.3)
By letter dated August 11, 1986, the applicant requested an exception to the requirements of Regulatory Guide (RG) 1.97, Revision 3, relative to the post-accident sampling system (PASS) design capability to sample the Clinton Power Station containment and drywell equipment / floor drain sumps. P,0cause of signif-icant sump pump problems, the pressure head needed to meet the PA35 sump sample flow requirements is not available. The applicant's intent is to delete the PASS sump sampling capability since the objectives of RG 1.97 can be met by alternative means.
l In Section 9.3.5 of SSER 6, the staff concluded that the Clinton Power Station PASS meets the requirements of Item II.B.3 of NUREG-0737. The requested dele-l tion of the PASS sump sample capabilities will not impact: PASS system accept-I ability, since the intent of the PASS is to obtain samples from the reactor cool-ant system, the suppression pool, and the containment atmosphere. The samples i
are intended to determine the extent of core damage. Release of radioactive
( coolant into the drywell floor drain sump will result in sump isolation and j
overflow to the drywell flow, up over the weirwall and into the suppression pool. Therefore, a suppression pool sample can be used as an alternative to a drywell containment equipment drain sump, and the containment floor drain sump can be sampled similarly.
The need for sump samples is indicated in RG 1.97, Revision 3, Table 2, Note 19.
This note states: "An installed capability should be provided for obtaining containment sump, ECCS pump room sumps, and other similar auxiliary building l sump liquid samples." The intent for sump sampling in various primary and sec-l ondary containment areas is for detection and assessment of reactor coolant l leakage in these areas. As discussed above, the primary containment sumps can Clinton SSER 7 9-4 l
l be sampled by the PASS suppression pool sample. Alternate methods for leak de-tection in secondary containment areas include alarms on sump level, room tem-perature and continuous airborne monitors (Emergency Operating Procedures CPS 4406.0, Secondary Containment / Radioactive Release Control Emergency). The ;
intent of RG 1.97, Revision 3, Table 2, Note 19 is, therefore, met. l On the basis of the above evaluation, the staff concludes that the deletion of the PASS sump sample capabilities meets the requirements of Iterj II.B.3 of NUREG-0737 since it will not affect a postaccident core damage assessment.
Furthermore, the intent of RG 1.97, Revision 3, Table 2, Note 19, is met by alternate methods of detection and assessment of reactor coolant leakage to primary and secondary containment sump areas. The staff, therefore, finds ac-ceptable the applicant's exception to the requirements of RG 1.97, Revision 3, relative to the PASS sump sampling capability.
9.4 Ventilation Systems 9.4.1 Control Room Area Ventilation System (Control Room Heating, Ventilating, and Air Conditioning (HVAC) System)
By letter dated September 10, 1986, the applicant requested a change to the Clinton Technical Specifications (CPS-TS 3.3.7.8) that suspends the operability requirement for the chlorine detection system until the initial fuel load is completed and before placing the reactor mode switch in startup for the initial criticality.
The chlorine detection system is used to provide a signal to the control room ventilation system (VC) in the event of an accidental chlorine release. The VC, upon chlorine detection, is automatically placed in the chlorine mode of operation to provide the required protection for the control room operators.
Since the VC is not required to be operable until the initial fuel load is com-pleted and before the reactor mode switch is placed in startup for the initial criticality (per CPS-TS 3/4.7.2), the operability requirement for the chlorine detection system should be the same as that for the VC.
, Because the chlorine detection system is installed and functional, the staff l
concludes that the applicant's proposed revision to the Clinton Technical Spec-ifications as discussed above is acceptable.
l l Clinton SSER 7 9-5
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i 11 RADI0 ACTIVE WASTE MANAGEMENT 11.5 Process and Effluent Radiological Monitoring and Sampling Systems (1) Time-Dependent Correction' Factors for Effluent Radiation Monitors SSER 5 states that the noble gas effluent monitoring systems installed in the Clinton standby gas treatment system (SGTS) and heating, ventilation, and air conditioning (HVAC) stack exhausts are acceptable, contingent upon the appli-cant providing the time-dependent correction factor curve, along with the basis and the assumptions used before the full power operating license is issued.
This factor is used for converting effluent monitor readings to release rates, on the basis of the exhaust air flow rate and radionuclide spectrum variation in the s' ample lines as a function of time after an accident.
By a letter dated June 23, 1986, the applicant provided the methodology, assump-tions, and the time-dependent postaccident correction factor curves for the Clinton low, middle, and high range noble gas effluent monitors installed in the SGTS and HVAC system. In the event of accidents within the design basis at Clinton, all potential radicactivity release paths would be isolated, with the exception of the SGTS and HVAC exhaust stacks.
The correction factor curves are based on noble gas radionuclide decay and moni-tor response characteristics during and following such accidents, as a function of time (10 minutes to 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />). The applicant has demonstrated reasonable assurance for being able to provide an overall monitoring system accuracy which is within a factor of 2 with respect to noble gas releases, as delineated in Regulatory Guide (RG) 1.97, Revision 3. In view of the above, the staff finds i
that the proposed time-dependent correction factor curves developed and provided by the applicant are acceptable and that the applicant's commitment identified in SSER 5 to provide the correction factor curves before the full power operating license is issued has been fulfilled.
l (2) Standby Gas Treatment System Sample Cooler Capacity SSER 5 states that the heat tracing on the SGTS stack effluent sample lines is acceptable provided that the SGTS sample cooler has adequate capacity and has a provision for collecting and draining the condensed water. The sample cooler is used to reduce the sample temperature below the maximum temperature limit (120*F) of the noble gas pallet (sampler and detector) prior to a noble gas activity measurement.
By a letter dated June 23, 1986, the applicant provided the heat transfer and thermal analysis of the SGTS sample line. The applicant's calculation indi-cates that the maximum sample cooler outlet temperature will not exceed 120*F.
The calculation is based on using a maximum expected sample inlet temperature of 180*F at 5 to 6 liters per minute, and a maximum expected cooling water temperature of 105*F at 20 gpm. Furthermore, the applicant stated that the vertical sample tubing section (approximately 12 feet long) will provide addi-tional cooling of the sample before reaching the noble gas pallet.
Clinton SSER 7 11-1
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To provide adequate capacity to collect the condensed water from the sample line, the applicant has added a manually controllable drain, with a capacity of 178 cubic centimeters, immediately downstream of the sample cooler. The appli-cant also has incorporated into the Clinton Plant Operating Procedures the requirement of removing the drain 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after an accident, so as to prevent the potential of interrupting the system gas sample flow.
The staff finds that the applicant's calculation and addition of a manually controllable drain is acceptable. This acceptance is based on (a) the appli-cant's heat transfer and thermal analysis, (b) the current as-built system design for the sample line condensate drain, and (c) the Clinton Plant Operat-ing Procedures. Therefore, the staff concludes that the Clinton SGTS sample cooler has an adequate cooling capacity and the system has adequate provisions to collect and drain gas sample line condensation. This fulfills the appli-cant's requirement to complete these actions before issuance of the full-power license.
(3) Calibration Methods for Noble Gas Effluent Monitors By a letter dated March 10, 1986, the applicant submitted proposed calibration methods for normal range noble gas effluent monitors. The applicant proposes to calibrate the detector assemblies in these monitors using a National Bureau of Standards (NBS) traceable solid source transfer standard and vendor-supplied NBS traceable noble gas calibration data before reactor criticality. The appli-cant proposes a subsequent validation test using the plant process off gas or an NBS xenon-133 calibration gas before exceeding S% of rated reactor power. The staff finds that the applicant's proposed normal range noble gas effluent moni-tor calibration methods are acceptable.
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Clinton SSER 7 11-2
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i 13 CONDUCT OF OPERATIONS 13.1 Organizational Structure of Applicant 13.1.2 Operating Organization 13.1.2.1 Organization Since Supplement 6 to the SER (SSER 6) was issued, the applicant (by letter dated September 10, 1986) modified the organization from that previously re-viewed by the staff. The changes consist of the addition of an Assistant Man-ager, Plant Maintenance and an Assistant Manager, Startup. The individuals who fill these two positions report directly to the Assistant Manager of Clinton Power Station. The modified organizational chart is shown as Figure 13.2 and will be contained in the next amendment to the FSAR.
The staff reviewed the risumss of the two men who fill these new positions and determined they meet the qualification requirements of ANSI /ANS 3.1-1978.
Furthermore, since the addition of these two positions to the onsite organiza-tion will provide higher level management attention to these functions, the staff finds these changes acceptable.
13.2 Training Program The applicant's training program was found acceptable in the Clinton SER. Since the SER was issued, the training program was upgraded in revisions to FSAR Sec-tion 13.2 through Amendment 34, and in commitments provided by the applicant in a letter dated October 29, 1985. These modifications were found-acceptable in Supplement 5 (SSER 5) dated January 1986. The applicant has since modified FSAR Section 13.2 through Amendment 38, and by letter dated July 8,1986. The staff's evaluation of these subsequent modifications follows.
13.2.2 Unlicensed Personnel i Shift Technical Advisor Training (1) Appendix D,Section I.A.1.1(c) of the FSAR discusses on-the-job training for shift technical advisors (STAS). By letter dated July 8, 1986, the applicant provided information that current STAS will acquire "in training" experience while rotating with the shift during preoperational testing and after the plant obtains a low power license. This training will be provided before the plant exceeds 5% of rated power. Every future STA will serve as an STA in training for a minimum of 3 months. The staff finds that this change conforms to guidance contained in Section 5.2 of Appendix C of NUREG-0737. (Appendix C is the Institute of Nuclear Power Operations (INPO) document, " Nuclear Power Plant Shift Technical Advisor, Recommendations for Position Description, Qualifications, Education and Training," dated April 30, 1980.)
Clinton SSER 7 13-1
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(2) The SER compared educational requirements for the STA position at Clinton Power Station with criteria contained in NUREG-0737, Appendix C, Sec-tion 6.1. The applicant has deleted this comparison in Appendix D, Item I.A.1.1, of the FSAR because it has committed to provide a dedicated STA on shift who has a 4 year degree from an accredited institution in an area of engineering, engineering technology, or the physical sciences.
.Therefore, it is not necessary to provide a comparison with the education resiirements contained NUREG-0737, Appendix C, Section 6.1. STAS at Clinton Power Station will receive education in reactor theory. The staff finds that this change conforms to the guidance provided in the Policy Statement on Engineering Expertise on Shift (Generic Letter 86-04, 50 FR 43621, Octnber 28, 1985).
13.2.3 Replacement and Retraining SSER 5 stated that each licensed operator would maintain proficiency by perform-ing one .8-hour shift per month. As an alternative, the licensed operator could spend 4 consecutive days on shift and in this case would be qualified for the next 90 days. On an average, each licensed operator would serve one 8-hour shift per month.
If an individual has not satisfied the proficiency requirement through either one 8-hour shift per month or 4 consecutive days (maintaining proficiency for the next 90 days), a provision has been made in FSAR Amendment 38 to require a licensee to stand one 8-hour shift under the direction of a licensed SRO.
This information was provided by the applicant to clarify its position for maintaining proficiency on shift for a period of less than 4 months. However, if an individual is off shift for 4 months or more, the applicant will certify or demonstrate to the Commission that the licensed operator's knowledge and understanding of facility operation and administration are satisfactory. The staff finds that this change is in conformance with requirements contained in 10 CFR 55.31(e).
On the basis of its review of the revisions to the FSAR in Amendments 35 through 38, and commitments made by the applicant in the July 8, 1986, letter, the staff finds acceptable the applicant's revisions to FSAR Section 13.2 with respect to the training program.
Clinton SSER 7 13-2
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MANAGER.
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(n TECHNICAL l TT1 2 ADVISOR I
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i 14 INITIAL TEST PROGRAM Preoperational Test Deferrals By letters dated August 13, 1986, and September 10, 1986, the applicant re-quested authorization to defer the completion of four additional preoperational tests. The systems for which testing would be deferred are (1) the main steam line isolation valve-leakage control system (MSIV-LCS), (2) the emergency core cooling system (ECCS) equipment cooling, heating, ventilation, and air condi-tioning (HVAC) system (VY), (3) the essential switchgear heat removal system (VX), and (4) the screenhouse and makeup water pump house HVAC system (VH).
In the latter three systems, the only portion of the preoperational testing that is.to be deferred is the area temperature survey.
The deferral of preoperational phase testing of the solid radwaste system beyond the fuel load date but before the primary system heatup, proposed and described in the applicant's submittal of March 12, 1986, was considered and found accept-able in the Clinton SSER 6 Appendix N. Due to an ongaing modification to this system, the applicant, by letter dated September 17, 1986, has requested a re-scheduling of the original deferral date for the test completion milestone.
! The applicant, by letters dated March 12 and May 29, 1986, requested preopera-tional test deferrals for the offgas system and supporting subsystems. In Sec-tion 2.3 of Appendix N of SSER 6, the staff approved the preoperational test deferral for the offgas system but did not explicitly defer the loading / testing for the inplace charcoal filters for the offgas system and the preoperational testing and final air balancing for the offgas vault HVAC system.
These new test deferrals, test deferral extension, and clarification of a test deferral are discussed below.
, MSIV-LCS
. The applicant is requesting deferral of the completion of testing of the MSIV-LCS until after fuel loading with completion of the testing before initial criticality. The reason for this deferral request is that on the basis of the preoperational testing performed on this system to date, it has been determined that the pressure trip units (pressure transmitters which actuate the system valves and blowers) could not perform their function in'the current configura-tion because of the inability to maintain full water legs in the instrument lines. In addition, the system contained a defective flow element and trans-mitter. The design change and any procurement and installation are projected 1 to be completed after fuel loading.
i The MSIV-LCS is a system to limit the release of radioactivity after a loss-of-coolant accident (LOCA). Without significant power operation, there would be insignificant amounts of radioactivity available for release should a LOCA occur. Therefore, the staff finds acceptable the deferral of preoperational
! testing of the MSIV-LCS until after initial fuel loading with test completion i before initial criticality and that this deferral does not require any excep-tions to the plant Technical Specifications or General Design Criteria.
l Clinton SSER 7 14-1 I
Area Temperature Surveys for the ECCS Equipment Cooling HVAC System Essential Switchgear Heat Removal System, and Screenhouse and Makeup Water Pump House HVAC System The applicant requests deferring the area temperature surveys for these three preoperational tests until such times in the startup program as there are significant heat loads so that any flow balancing can be performed for final design verification. These surveys and balancing would be completed before the startup program is completed.
Temperature, heat load, and flow balancing cannot be verified until loads near design values are obtained. Therefore, the staff finds acceptable deferral of the temperature surveys for the above systems until sufficient heat loads are developed during the startup program.
The staff concludes that these preoperational test deferrals requested by the applicant are acceptable with the completion milestones noted. The staff also concludes that these deferrals are acceptable without requiring any exceptions to the plant Technical Specifications or General Design Criteria.
Solid Radwaste System (Section 2.10 of Appendix N to SSER 6)
The applicant is requesting a rescheduling of the deferral for completion of the solid radwaste system preoperational phase test previously granted in Sec-tion 2.10 of Appendix N to SSER 6 to beyond the fuel load date, but before com-mercial operation. GDC 60 requires that a licensed plant have a solid radwaste system. The preoperational test includes a demonstration of trash compacting and handling equipment, as well as an integrated solid radwaste system perform-ance test.
The staff found the use of the Associated Technology Incorporated (ATI) trans-portable system instead of the permanently installed solid radwaste system at Clinton to be acceptable (see SSER 6). The ATI system satisfies GDC 60. The applicant states that the ATI transportable (mobile) solidification system will be in place at Clinton and functionally operable prior to reactor initial crit-icality. The applicant further states that an interim modification of one solid radwaste collection and transfer subsystem was recently completed for transferring radioactive waste slurries to the ATI system. In June 1986, the ATI system was demonstrated to be capable of solidifying radwastes at Clinton.
Therefore, the staff finds the applicant's request for rescheduling the deferral of the preoperational test of the solid radwaste system until commercial plant operation is achieved to be acceptable and no exemption is required.
Offgas System (Section 2.3 of Appendix N to SSER 6)
The applicant requested, by letters dated March 12 and May 29, 1986, authoriza-tion to defer completion of the loading / testing for the inplace charcoal filter for the offgas system and to defer the preoperational test and final air balanc-ing for the offgas vault HVAC system beyond the fuel load date. However, this work will be completed before the installation of the reactor vessel head and tensioning of the head bolts. These two systems support the operation of the condenser offgas system, which is one of the systems that controls the release of radioactive gaseous effluents, and thus GDC 60 is applicable. In Section 2.3 Clinton SSER 7 14-2
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of Appendix N to Supplement 6 to the SER, the staff approved the deferral of completion of the preoperational testing of the offgas system beyond the fuel load date, but before the installation of the reactor vessel head and tensioning of the head bolts. Although not explicitly identified in Supplement 6 to the SER, the two supporting subsystems identified above were considered in the staff's review of the offgas system. The purpose of this evaluation is to make clear the fact that these two supporting subsystems were considered in the staff's prior review of the offgas system and that the schedular exemption from the requirements of GDC 60, granted in Supplement 6 to the SER, was intended to be applicable to these supporting systems as well, since they also are not needed before the reactor vessel head is installed and the head bolts are tensioned.
Clinton SSER 7 14-3
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l l 15 SAFETY ANALYSIS 15.1 Anticipated Operational Occurrences Operation of the Clinton Station with reduced feedwater temperature (RFT) opera-tion is possible during normal operation. With decreasing power level while all feedwater heaters are in service, feedwater temperature normally drops and i'
could be further reduced during maintenance on the feedwater system. Lower feedwater temperature changes the relative values of the reactivity components and possibly the nuclear and thermal hydraulic response during a transient.
Operation with RFT would be outside the bounds of the existing analyses.
The staff has reviewed analyses performed for a BWR/6 using staff-approved models to justify steady-state operation with RFTs. Temperatures ranging from 420*F to 250*F were considered. The anticipated operating transients were reevaluated to determine the required operating MCPR limits. The operating minimum critical power ratio (MCPR) values were found to increase by 0.01 for operation between feedwater temperatures of 370*F and 320*F, and by 0.03 for feedwater tempera-
- tures between 320*F and 250*F.
Since plant-specific RFT analyses were not performed for Clinton, the operating MCPR in the Technical Specifications will be increased by 0.03 for a feedwater j temperature between 400*F and 320*F and by 0.06 between 320*F and 250*F. The additional margin is provided to account for possible differences in transient response between Clinton and the analyzed BWR/6. In addition, before startup
! following the first refueling outage, the staff requires plant-specific analy-i ses to determine the operating limit MCPR from RFT operation if the applicant wishes to maintain the option for RFT operation after the first operating cycle.
l The Clinton operating license will be conditioned as follows to reflect this requirement (license condition 16):
i The facility shall not be operated with reduced feedwater temperature for the purpose of extending the normal fuel cycle. .After the first operating cycle, the facility shall not be operated with a feedwater heating capacity which would result in a rated thermal power feed-water temperature less than 420*F unless analyses supporting such operation are submitted by the applicant and approved by the staff.
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l l Clinton SSER 7 15-1
17 QUALITY ASSURANCE In SSER 3, the NRC staff found that the quality assurance (QA) program descrip-tion contained in Section 17.2 of the FSAR was acceptable. However, since SSER 3 was issued, Illinois Power Company (IP) (the applicant) has amended the FSAR through Amendment 38 with changes affecting the QA organization, the items covered by the QA program, and typographical corrections. In addition, by let-l ter dated August 26, 1986, the organization chart in the FSAR was corrected.
The corrected chart will be included in the next amendment to the FSAR. The e
NRC staff has reviewed this information and the following information super-sedes that described in Sections 17.2 and 17.4 of SSER 3.
17.2 Organization
- The structure of the organization responsible for the operation of the Clinton Power Station (CPS) and for the establishment and execution of the operatior.s phase QA program is shown in Figure 17.1. The Chairman and President of IP has the overall responsibility for the engineering, design, procurement, construc-tion, modification, testing, operation, and quality assurance at Clinton.
- Execution of these responsibilities is delegated to two Executive Vice Presi-l dents. .
l One of the Executive Vice Presidents is responsible for the overall effective-
, ness of the QA program and is responsible for establishing tha QA policies, goals, and objectives. The responsibilities to establish and maintain the applicant's Nuclear Power Quality Assurance Program as well as testing, startup, operations, nuclear support, and engineering are delegated to a Vice President.
This Executive Vice President is also responsible for assuring that annual management reviews are conducted and documented on the status, adequacy, and effectiveness of the overall QA program. This Executive Vice President retains
- the responsibility for assuring that the authority and independence of QA per-sonnel are such that they can effectively assure the conformann to quality requirements and are independent of undue influences and respoalibilities for schedule and cost.
The other Executive Vice President is responsible for quality-related activities associated with environmental affairs. The responsibility to establish and I
maintain procedures which support the applicant's Nuclear Power Program on safety-related matters has been delegated to a Vice President.
A Vice President has been designated as the corporate officer responsible for the overall direction of the applicant's Nuclear Power Quality Assurance Pro-gram, the testing, startup, nuclear support activities (including purchasing),
engineering, and commercial operation of Clinton. This Vice President reports j to an Executive Vice President and has direct access to the President (see Figure 17.1).
The Manager - CPS reports to the Vice President and is responsible for the safe,
- reliable, and efficient operation of the Clinton plant in accordance with the j operating license. This includes ensuring that the applicant's Nuclear Power Clinton SSER 7 17-1
Quality Assurance Program is incorporated in plant procedures and implemented by the Clinton organization.
The Manager - Quality Assurance (Q/A) reports to a Vice President and has di-rect access to an Executive Vice President (see Figure 17.1). The Manager --
Quality Assurance is responsible for the applicant's overall QA program defini-tion, direction, evaluation, and approval, including the Nuclear Power Quality Assurance Program. The Q/A Manager directs the QA departmental activities re-lated to the design, procurement, construction, and operation of the Clinton Power Station. The Q/A Manager's designated alternate has been delegated the responsibility and authority to stop unsatisfactory work.
Four sections reporting to the QA Manager assist the manager in carrying out
- the QA responsibilities. These are Quality Systems and Audits, Quality Engi-neering and Verification, Operations Monitoring Programs, and Procurement j Quality Assurance.
l The Director - Quality Systems and Audits is responsible for providing direction and administration of the Quality Systems and Audits staff in defining, estab-
- lishing, and verifying compliance with the Nuclear Power Quality Assurance i
Program through internal and external audits.
The Director - Quality Engineering and Verification is responsible for direction and management of the Quality Operations and Maintenance, Quality Control, and
- Quality Technical Support staffs for defining, establishing, and verifying com-pliance with the Nuclear Power Quality Assurance Program. The Supervisor - ,
i Quality Operations and Maintenance Section is responsible for the planning and performance of surveillances including assuring timely and responsiv'e corrective action to surveillance findings, and for advising management about the effec-tiveness of QA program implementation for those specific functions under sur-veillance. The Supervisor - Quality Control is responsible for testing, main-tenance, modification, nuclear fuel, and plant support activities which affect quality. This supervisor also initiates reports of nonconforming items or con-ditions discovered during inspections. The Supervisor - Quality Technical Sup-port is responsible for assuring documents involving regulations, standards,
- codes, startup, engineering, radiation, and chemistry conform to the applicable QA program requirements. ;
j The Director - Operations Monitoring Programs is responsible for providing i assistance to the Manager - Quality Assurance in the development and imple-mentation of Quality Assurance Operations Monitoring Programs that meet regu-latory requirements, commitments, and support the safe and reliable operation of Clinton.
The Procurement Quality Assurance Section is responsible for assuring procure-ment documents contain the QA requirements of the FSAR, evaluating supplier QA programs for meeting the FSAR commitments, performing surveys at suppliers' facilities, processing procurement-related nonconformances, performing source i
surveillance, performing receipt inspections, and reviewing and approving ven-
- dor working procedures.
QA personnel do not perform non quality-assurance duties and give their full i
attention to ensuring effective implementation of the QA program.
I Clinton SSER 7 17-2
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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 supplier QA program requirements, (4) per forming reviews and evaluations of procurement sources to determine their capability to meet QA requirements, (5) performing surveillance and evaluations at supplier 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 supplier facili-ties. Personnel performing quality-related activities have the authority to (1) identify problems relating to quality, which include stop-work recommenda-tions by the QA organization; (2) initiate, recommend, or provide solutions to problems relating to quality; (3) verify implementation of solutions; and (4) control further processing, delivery, installation, or utilization of non-conforming items until proper disposition has occurred.
17.4 Conclusions On the basis of its detailed review and evaluation of the QA program descrip-tion contained in Section 17.2 of the FSAR, the additional QA information pro-vided through Amendment 38 and by letter dated August 26, 1986, the staff con-cludes that:
(1) 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 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 con-trol of the QA program.
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Clinton SSER 7 17-3 f
O Et is 3 PRESIDENT w
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4 EXECUTIVE EXECUTIVE VICE "" " "5 VICE PRESIDENT PRESIDENT I
, 1 OFFSITE I ONSITE l VICE PRESIDENT . VICE I PRESIDENT I I 8 MANAGER ENVI ON E TAL l
MANAGER MANAGER CLINTON WER QUAL -- d .
STAFF e,
I 1 I .
DIRECTOR DIRECTOR DIRECTOR OUALITY OPERATIONS QUALITY ENGINEERING & MONITORING SYSTEMS &
VERIFICATION PROGRAM AUDITS j I i I I I I SUPERVISOR SUPERVISOR-SUPERVISOR SUPERVISOR SUPERVISOR OUALITY OUALITY QUALITY QUALITY SUPERVISOR OPERATIONS & TECHNICAL PROCUREMENT AUDITS MAINTENANCE CONTROL
' M SYSTEMS SUPPORT 1
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Figure 17.1 Illinois Power Company'S ClintOn Power Station quality assurance Organization i
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APPENDIX A CONTINUATION OF CHRONOLOGY July 16, 1986 Letter to applicant forwarding Federal Emergency Management Agency's April 4, 1986, interim finding on plant offsite emergency plans. Health and safety of public can be pro-tected with reasonable assurance in event of incident.
July 17, 1986 Letter from applicant forwarding response to NRC question on revised FSAR Chapter 2 concerning hydrology.
Figures 2.4-27, 2.4-28, and 2.4-29 used in analysis of ultimate heat sink have been revised using more conservative Corps of Engineers procedure.
July 18, 1986 Letter from applicant forwarding responses to questions about amendments to FSAR Chapter 8, " Electric Power." FSAR will be updated in future amendments. Response closes out SSER 6 issues.
July 23, 1986 Letter from applicant submitting updated listing of piping subsystems which had arbitrary intermediate breaks elimi-nated, per Chapter 3 of FSAR and April 16, 1985, letter.
July 23,1986 Letter to applicant issuing Amendment 2 to materials license No. SNM-1886.
July 24, 1986 Letter from applicant describing implementation and comple-tion of programs and plans addressed in Atomic Safety and Licen;ing Board's January 28, 1985, joint stipulation, in-cluding fuel zone instrumentation, sensing lines vertical drop, installation of isolators, and review of Regulatory l Guide 1.97.
July 24, 1986 Summary issued of July 15, 1986, telephone conversation with utility about simulator exercise during Office of Nuclear Reactor Regulation's management assessment July 9, 1986, meeting. Observation was made on difficulty in determining actions for event resulting in loss of high pressure injec-tion spray.
July 25, 1986 Letter from applicant identifying an additional exception to Technical Specifications per utility's March 12, 1986, re-quest and justification for deferrals of some preoperational tests. Marked-up Technical Specification 3/4.9.12 on in-clined fuel transfer system is enclosed.
Clinton SSER 7 Appendix A 1
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July 30, 1986 Letter from applicant requesting change to Technical Speci-fication 3/4.10.6, "Special Instrumentation - Initial Core Loading," based on information from General Electric Co.
indicating that signal-to-noise ratio for reduced count rate situation should be greater than or equal to 20.
August 5, 1986 Letter from applicant confirming that General Electric Co.
qualification report on hydraulic control units purchased and incorporated into equipment qualification report, per SER Outstanding Issue 711. Hydraulic control unit bolts are to be tightened following earthquake.
August 5, 1986 Letter from applicant forwarding clarification of utility's position on Regulatory Guide 1.139 about guidance ca resid-ual heat removal, as revised in Amendment 38 to FSAR, per NRC questions.
August 5, 1986 Letter from applicant providing justification for deleting berm height requirement to divert floodwater runoff from entering excavation, per FSAR Section 2.5.4.14.4 and SSER 2 (NUREG-0853). As-built berm is satisfactory and will be included in next FSAR revision.
August 6, 1986 Letter to applicant forwarding questions and comments on pump and valve inservice testing program. Working meeting should be held in Bethesda, Md., 30 to 45 days efter re-ceipt of letter to discuss utility's responses.
August 8, 1986 Letter from applicant advising that control logic for operation of four 36-inch isolation valves does not meet NUREG-0737, Item II.E.4.2, requirement that reopening of containment isolation valves requires deliberate operator action. Plant modifications are pending.
August 8, 1986 Letter from applicant clarifying compliance with TMI Action Plan Item II.K.1.10 on independent verification of emer-l gency core cooling system (ECCS) lineups. FSAR Subsec-tion 6.3.2.8 will be revised to reflect application of in-dependent verification only for ECCS major flow paths.
August 11, 1986 Letter from applicant requesting relief from requirements l of Regulatory Guide 1.97, Revision 3, on postaccident i
sampling system capability to sample containment and drywell equipment / floor drain sumps. Description and justification are enclosed. FSAR will be revised.
i August 11, 1986 Letter from applicant submitting quarterly update on con-l struction schedule, per NRC's May 29, 1981, request.
Current estimate for fuel load is late August 1986.
August 12, 1986 Letter from applicant forwarding additional information for NRC to complete review of Technical Specifications per July 21-25, 1986, meetings. Table 1 summarizes changes Clinton SSER 7 2 Appendix A
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to Technical Specifications and provides technical discus-sion for each change.
August 12, 1986 Letter to applicant forwarding Supplement 6 to NUREG-0853, "SER Regarding Operation of Clinton Power Station, Unit 1."
August 13, 1986 Letter from applicant forwarding technical justification to March 12, 1986, request to defer preoperational test re-garding main steam isolation valve leakage control system until after fuel load.
August 13, 1986 Letter from applicant forwarding request to change Technical Specification Table 3.6.4-1, " Containment Isolation Valves,"
by adding four manual isolation valves to main steam isola-tion valve leakage control system and one manual isolation valve to equipment hatch penetration seals.
4 August 13, 1986 Letter from applicant forwarding application to amend CPPR-137 extending construction completion date to October 1, 1987.
i August 19, 1986 Letter from applicant advising staff that FSAR Subsec-tion 7.2.1.1.6.1.2 will be revised in Amendment 39 to re-flect increase in sequence-of-events monitoring system's response time, per Generic Letter 83-28.
August 20, 1986 Generic Letter 86-14 issued to all power reactor licensees and applicants for operating license regarding operator licensing exams.
August 21, 1986 Letter to applicant forwarding page changes to June 18, 1986, submittal of final draft Technical Specifications for review.
Certification that Technical Specifications are consistent with FSAR and SER is requested 2 weeks before date facility will be ready for licensing.
August 21, 1986 Letter from applicant informing staff.of revision to FSAR Section 13.2.2.1.5 regarding accelerated requalification of licensed operators. Revision ad:is condition that operators be removed from licensed duties until they have been re-trained and reexamined.
August 22, 1986 Letter to applicant forwarding papers delivered at Nuclear l Energy Agency's April 14-18, 1986, symposium on reduction of scram frequency and NUREG-1212, " Status of Maintenance in U.S. Nuclear Power Industry."
August 26, 1986. Letter from applicant informing staff that FSAR Figure 17.2-2 l on quality assurance (QA) organizational chart was not updated in FSAR Amendment 38. Figure will be updated in j next FSAR amendment to correspond to enclosed Figure 1-3 of opera 61onal QA manual.
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Clinton SSER 7 3 Appendix A
August 28, 1986 Letter from applicant forwarding response to NRC questions on FSAR Chapter 8, " Electric Power," regarding effect of block start voltage on equipment and motor-operated valves.
Calculations indicate the system is adequate under listed conditions.
September 4, 1986 Letter from applicant forwarding response to NRC comments on FSAR Chapter 8, " Electric Power," regarding Division 3 diesel generator alarms in main control room.
September 5, 1986 Letter from applicant requesting waiver to 10 CFR 50 Appendix E requirements on annual emergency preparedness exercise scheduled for October 28, 1986. Exercise should be deferred until January 13, 1987, for listed reasons.
September 10, 1986 Letter from applicant rcquesting to defer some preoperational tests until after fuel load. Technical justification for deferrals is included in submittal.
September 10, 1986 Letter from applicant requesting three Technical Specifica-tion changes related to limiting condition for operation for the chlorine detection system, a single addition to the table listing motor-operated valves with thermal overload protection, and an addition to the CPS Onsite Organizational Chart (Fipure 6.2.2-1).
September 12, 1986 Letter from applicant regarding the temporary use of equip-ment and components which are not fully qualified to the re-quired environmental standards before fuel load.
September 17, 1986 Letter from applicant providing technical justification for deferral of preoperational test for the solid radwaste system until prior to commercial operation.
September 18, 1986 Letter from applicant providing certification of final draft Technical Specifications.
September 18, 1986 Letter from applicant providing additional information on source range monitor signal to noise ratio.
September 22, 1986 Letter from applicant providing additional information on the requested deferral for the solid radwaste system.
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i Clinton SSER 7 4 Appendix A
APPENDIX B REFERENCES U.S. Nuclear Regulatory Commission, Generic Letter 85-03, " Clarification of Equivalent Control Capacity for Standby Liquid Control Systems," January 28, 1985.
Generic Letter 86-04, " Policy Statement on Engineering Expertise on Shift,"
February 13, 1986.
-- , NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," Noverber 1980.
-- , NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
-- , NUREG-0776, Supplement 3, " Safety Evaluation Report Related to the Operation of Susquehanna Steam Electric Station Units 1 and 2," July 1982.
-- , NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," LWR Edition, July 1981.
Industrial Standards American National Standards Institute, ANSI N509
-- , ANSI /ANS 3.1-1978 Institute of Electrical and Electronics Engineers, IEEE 383-1974, "IEEE Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations."
-- , IEEE 384-1974, "IEE Trial-Use Standard Criteria for Separation of Class 1A Equipment and Circuits."
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Clinton SSER 7 1 Appendix B
,_ .. . . . . . - . . . . . . . . ~ . . . . . . . . . . . . . , . - . . . - ....a.. . ....s, APPENDIX D NRC STAFF CONTRIBUTORS This SER supplement is a product of the NRC staff. The persons listed below were principal contributors to this report.
Name Branch (Division)*
R. Becker Facility Operations (BWR Lic.)
W. L. Belke Quality Assurance, Office of Inspection and Enforcement A. Chu Plant Systems (BWR Lic.)
H. Conrad Engineering (BWR Lic.)
J. Lazevnick Electrical, Instrumentation & Control System (PWR Lic. A)
A. Lee Engineering (BWR Lic.)
W. Meinke Plant Systems (BWR Lic.)
A. Notafrancesco Plant Systems (BWR Lic.)
M. L. Roe Maintenance Training (Human Factors Technology)
I. Schoenfeld Facility Operations (BWR Lic.)
B. Siegel BWR Project Directorate #4 (BWR Lic.)
P. Sobel Engineering ',8WR Lic.)
T. M. Su Reactor Systems (BWR Lic.)
G. Thomas Reactor Systems (BWR Lic.)
N. Wagner Plant, Electrical, Instrumentation & Control Systems (PWR Lic. B)
F. Witt Plant Systems (BWR Lic.)
- Except where otherwise noted, the contributor works for NRC's Office of Nuclear Reactor Regulation.
Clinton SSER 7 1 Appendix D
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APPENDIX F ERRATA TO CLINTON POWER STATION SAFETY EVALUATION REPORT SER Page 9-23, Section 9.4.3, Delete first sentence.
lines 1 & 2 Page 9-23, Section 9.4.3, Delete " room". After " system" add, " essential line 6 switchgear heat removal, and laboratory system".
SSER 2 Page 6-6, Section 6.2.6, Delete " service water and component cooling line line 10' water valves".
SSER 6 Page viii, Table of Contents, Delete "9.1.2 Spent Fuel Storage... 9-5" line 1 Page 6-12, Section 6.2.7, Change "-61 psig" to "-61 psid".
line 37 Page 7-7, Section 7.4.3.1, Delete lines 26 and 27.
lines 26 & 27 Page 7-7, Section 7.4.3.1, After line 43 add the following line 43 " essential equipment cubicle HVAC systems fan status".
Page 8-1, Section 8.2.3, Delete 69% setting and add "by a 60% setting for line 11 divisions 1 and 2 and a 69% setting for division 3".
Page 8-2, Section 8.3.2, Change "137.4" to "136.5 i d.5%".
line 8 Page 9-5, Section 9.1.1, Delete sentence beginning "The staff recommends".
lines 21-22 Page 9-5, Section 9.1.2, Delete "9.1.2 Spent Fuel Storage" and first lines 23-31 paragraph.
Page 9-6, Section 9.1.2 Change "10 CFR 70.14" to "10 CFR 70.24". line 28 l Page 11-1, Section 11.4.1, Delete " outdoors,".
line 27 Page 11-2s Section 11.4.1, Delete " outdoors,".
I line 14 Page 14-11, Table 14.1 After title add "(spproximate)"
l Clinton SSER 7 1 Appendix F
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Page 22-5, Section 22, Delete'" generator... program." and replace with line 16 " generators was analyzed through basic analysis techniques. The results were demonstrated to be conservative through a computer analysis of a critical piping section."
Appendix A, page 12 Add another May 29, 1986, submittal, " Letter from applicant notifying staff that FSAR Amendment 38 is filed".
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- Clinton SSER 7 2 Appendix F i
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jr APPENDIX P CONTENTIONS As noted in Section 1.13 of this supplement to the SER, this appendix addresses the status of those issues that required resolution as identified in the Janu-ary 29, 1985, joint stipulation agreement between Illinois Power Co. , U.S.
Nuclear Regulatory Commission, Prairie Alliance, and the Illinois Attorney General and reports where resolution of these issues can be found.
Quality Control Contention II, Part 7, of Joint Stipulation Agreement Issue Status 7a. Status of special programs and By letter dated August 11, 1986, from actions identified in applicant's the applicant to Mark Jason, Assistant Quality Improvements and Confirma- Attorney General (IAG) for the State tory Actions Summary Report of Illinois, the status of these pro-grams and actions was provided (IAG request - no NRC review required).
7b. Independent Design Review By letter dated January 18, 1985, (IDR) Bechtel Power Corporation submitted the Clinton IDR. The staff evaluation was contained in SER Supplement 5. By letters dated November 26, 1985, and January 5,1986, the applicant pro-vided additional:information. The staff's final evaluation and accep-tance of the IDR is contained in SER Supplement 6.
7c. Overinspection Program By letter dated April 3, 1986, the applicant informed the staff that the overinspection program was concluded as agreed to in NRC letters dated July 2,1985, and November 21, 1985.
Also, by letter dated January 22, 1986, the applicant terminated the records verification program. By letter dated March 17, 1986, the staff concurred.
Clinton SSER 7 1 Appendix P
Quality Control (Continued)
Contention II, Part 7, of Joint Stipulation Agreement Issue Status 7d. Construction Appraisal Team In accordance with the agreement, the (CAT) staff performed a CAT inspection at Clinton on May 20-31 and June 20-21, 1985 (I&E Report No. 50-461/85-30).
Three technical issues were identified for which resolution is contained in SER Supplement 6. Six findings re-sulting from the inspection were issued by Region III staff in a letter dated August 30, 1985, and responded to by the applicant in a letter dated September 30, 1985. These findings from Report 50-461/85-30 have been
- resolved by Region III in the follow-ing inspection reports
Finding Resolved in 01 50-461/86009 02 50-461/86001 03 50-461/85063 04 50-461/86014 05a 50-461/86014 05b 50-461/86039 05c 50-461/86039 06a 50-461/85062 06b 50-461/85062 i
i Clinton SSER 7 2 Appendix P
Control Room Design Contention III, Part 8, of Joint Stipulation Agreement Issue Status 8.a.2 Fuel zone instrumentation Resolved in Section 7.5.3 of SER Sup-piement 5. Applicant has until before startup from first refueling outage to install fully qualified '
instrumentation.
8.b.2 Water level monitoring system Resolved in Section 4.4.2 of SER Supplement 4.
8.b.3(1) Sensing lines vertical Resolved in Section 4.4.2 of SER drop Supplement 4.
8.d.1 Installation of isolators Resolved in Section 7.5.3 of SER Supplement 5.
8.d.2 Safety parameter display Resolved in Section 7.5.3 of SER system Supplement 5.
8.e.2 Detailed control room design Resolved in Section 18 of SER review Supplement 5.
Identification of RG 1.97 Resolved in Section 18.2 of SER instruments Supplement 5.
8.e.3 Document review of RG 1.97 Resolved in Section 7.5.3 of SER Supplement 5.
Clinton SSER 7 3 Appendix P