ML20206B236
| ML20206B236 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 03/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0853, NUREG-0853-S08, NUREG-853, NUREG-853-S8, NUDOCS 8704090033 | |
| Download: ML20206B236 (47) | |
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NUREG-0853 Supplement No. 8 Safety Evaluation Report related to the operation of
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Clinton Power Station, Unit No.1 Docket No. 50-461 Illinois Power Company, et al.
U.S. Nuclear Regulatory Commission i
Office of Nuclear Reactor Regulation
- March 1987 5
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8704090033 870331 PDR. Abocl( c5cootg,j E PbR
ABSTRACT Supplement No. 8 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 supplemen't reports the status of items that have been resolved by the. staff since Supple-ment No. 7 was issued.
l Clinton SSER 8 iii
TABLE OF CONTENTS Page 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 6 ENGINEERED SAFETY FEATURES...................................... 6-1 6.2 Containment Systems........................................ 6-1 6.2.5 Combustible Gas Contro1............................. 6-1 6.6 Inservice Inspection of Class 2 and 3 Components........... 6-2 0.6.1 Evaluation of Compliance With 10 CFR 50.55a(g)...... 6-2 7 INSTRUMENTATION AND CONTR0LS.................................... 7-1.
7.5 Information Systems Important to Safety.................... 7-1 7.5.3 Resolution of Issues................................ 7-1 7.5.3.1 Conformance to Regulatory Guide 1.97, i Revision 3................................. 7-1 'l 7.5.3.2 Safety Parameter Display System............ 7-1 l
i 9 AUXILIARY SYSTEMS............................................... 9 9.1 Fuel Storage Facility...................................... .9-1 9.1.1 New Fuel Storage.................................... 9-1 12 RADIATION PROTECTION............................................ 12-1 12.3 Radiation Protection Design Features...................... 12-1 12.3.1 Facility Design Features.......................... 12-1 14 INITIAL TEST PR0 GRAM............................................ 14-1 Clinton SSER 8 v
TABLE OF CONTENTS-(Continued)
P_ age 15 SAFETY. ANALYSIS................................................. ~
15-1 15.6 Single Recirculation Loop 0peration....................... 15-1 15.6.1- MCPR Fuel Cladding Integrity Safety Limit......... 15-1 15.6.2 MCPR Operating Limit.............................. 15-2 15.6.3. Stability Analysis............................... 15-3 15.6.4 Loss-of-Coolant Accident Analysis................ 15-3 15.6.5 Containment Analysis............................. 15-3 15.6.6 Miscellaneous Impact Evaluation.................. 15-4 15.6.7 Technical Specification Chan 15-4 15.6.8 Evaluation Findings.........ges..................
..................... 15-4 16 TECHNICAL SPECIFICATIONS....................................... 16-1 APPENDIX A CONTINUATION OF CHRONOLOGY APPENDIX B REFERENCES APPENDIX D NRC STAFF CONTRIBUTORS APPENDIX F ERRATA TO CLINTON POWER STATION SAFETY EVALUATION REPORT APPENDIX Q STAFF SAFETY EVALUATIONS FOR THE TECHNICAL SPECIFICATION CHANGES PROPOSED BY THE LICENSEE AND THE STAFF'FOR THE CLINTON FULL-POWER OPERATING LICENSE'
'l Clinton SSER 8 vi
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 licensee) 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; SSER 6 was issued in July 1986; and SSER 7 was issued in September 1986. The purpose of this eighth supplement (SSER 8) is to further revise the SER by addressing the re-quests and commitments made by the licensee that have to be resolved before the full power license is issued.
Each section and appendix of this supplement is numbered and titled so that it corresponds to the section or appendix of the SER that is relevant to the NRC staff's additional evaluation. Except where specifically noted, the material in this supplement does not replace the material in the corresponding SER sec-tion or appendix. Appendix A is a continuation of the chronology of correspon-dence between NRC and the licensee and makes current the lists in the SER and in SSER 1 through SSER 7. 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 0 is a list of principal contributors to this supple-ment. Appendix F corrects errors in the SER and its supplements. Appendix Q contains the safety evaluations for the proposed changes to the Technical Specifications for the full power operating license. ,
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 1 had been changed to a con-firmatory status. SSER 2 reported that 6 issues had either been resolved or changed to a confirmatory status. SSER 3 reported that 4 issues had been Clinton SSER 8 1-1 l
. resolved. 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.
SSER 7 resolved the:one remaining issue. '
i SER/SSER
. Issue- Status Section(s)
(1) Transportation accidents Resolved in SSER 3 2.2.1 (2) Effects of Unit 2 excavation Resolved in SSER 2 2.4.2.2 (3) Seismic analysis Became confirmatory 2.5.2.4 '
issue 70, resolved in SSER 3 (4) Internally generated missiles Resolved in SSER 1 3.5 (5) Postulated piping failures Resolved in SSER 6 3.6.1, 3.6.2 (6) Steady-state vibration Resolved in SSER 2- 3.9.2. l acceptance criteria for i balance of plant piping (7a) Environmental qualification Resolved in SSER 5 3.11 l.. of electrical and mechanical and SSER 6 equipment 4
j (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 3.10.2 qualification i NUREG-0737 Item II.E.4.2(6) Resolved in SSER 5 3.10.3
! (8a) Preservice (PSI) and inservice PSI program: became 5.2.4, 6.6.1
- inspection (ISI) programs confirmatory _ issue 67 in SSER 2 ISI. program
- became 5.2.4, 6.6.1 license condition 12 in SSER 2 (8b) Preservice and inservice .Became confirmatory 3.9.6 testing of pumps and valves issue 68 in SSER 1 (9a) Pool dynamic loads due to LOCA Resolved in SSER 6 6.2.1.8 (9b) Pool dynamic loads due to SRV Resolved in SSER 5 6.2.1.8 9
Clinton SSER 8 1-2
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SER/SSER
, Issue Status Section(s) >
(10a) Containment purge- Became confirmatory 6.2.4.1 issue 69 in SSER 2 (10b) Containment isolation Resolved in SSER 2 6.2.4 (10c) Containment leakage testing Resolved in SSER 2 6.2.6 (vent and drain lines)
! (10d) Containment leakage testing Resolved in SSER 2 6.2.2 (secondary containment)
~(10e) Containment bypass leakage Resolved in SSER 2 15.3.1-(11) Control room habitability Resolved in SSER 1 6.4
- (12) Engineered safety features Resolved in SSER 2 7.3.3.7 reset controls (IE Bulletin
- 80-06) l (13) Remote shutdown system Resolved in SSER 3 7.4.3.1
- and SSER 6 (14) Capability for safe shutdown Resolved in SSER 2 7.4.3.2 following loss of bus supply-ing power to instruments and
, controls (IE Bulletin 79-27)
(15) Control system failures Resolved in SSER 6 7.7.3.1 resulting from high-energy-line breaks or common power source or sensor malfunctions (16) Separation of the RPS and MSIV Resolved in SSER 1 8.4.7 solenoid circuits and PGCC circuits (17) Organization and staffing Resolved in SSER 5 13 (18a) Onsite emergency plan Resolved in SSER 4 13.3 (18b) Offsite emergency plan Resolved in SSER 6 13.3 I (19) Security Resolved in SSER 1, 13.7 amended security plans -
i reviewed and approved
- in SSER 5 (20) QA program Resolved in SSER 3 17.2, 17.3 i
Clinton SSER 8 1-3 i
d
- SER/SSER Issue Status Section(s)
(21) Fire Protection Evaluation Resolved in SSER 6 9. 5' 1 Report and Safe Shutdown Analysis 1.10 Confirmatory Issues
- In SER Section 1.10, the NRC staff identified 66 confirmatory issuesLfor which additional information and documentation were required to. confirm preliminary I
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.
SER/SSER Issue Status Section(s)
, (1) Emergency preparedness Resolved in SSER 6 2.2.3,
- meteorological program 13.3.2.8 2
(2) Inspection program around the Resolved in SSER 1 2.4.5, 2.6-4 ultimate heat sink (VHS) and the main cooling lake dam (3) Protection of VHS dam abutments Resolved in SSER 1 2. 6 against soil erosion ,
(4) Internally generated missiles - Resolved in SSER 2 E3.5.1.1 fan failures (5) Design adequacy of cable tray Resolved in SSER 1 3.7.3 system j (6) Containment ultimate strength Removed from list in 3.8.1 analysis SSER 4 (7) Structural integrity of safety- Resolved in SSER 2 3.8.3 related masonry walls (8) NSSS pipe break analysis using Resolved in-SSER 1 . 3.6.1 SRP criteria I
(9) -Vibration assessment of RPV Resolved in SSER 4 3.9.2
Clinton SSER 8 1-4 e w-,,--- e--w- - ---gy- w----- - - --- - -- -- -- yre-
SER/SSER Issue Status Section(s)
(11) Use of SRSS for combining Resolved in SSER 1 3.9.3 Mark III dynamic responses for other than LOCA and SSE (12) IE Bulletin 79-02 regarding Resolved in SSER 2 3.9.3.4 support baseplate flexibility (13) Mark III hydrodynamic loads Became part of out- 3.9.3.1' standing issue 9 to avoid duplication in SSER 4 (14) Feedwater check valve analysis Resolved in SSER 2 3.9.3.1 (15) Seismic and LOCA loadings Resolved in SSER 4 4.2.3.4 on fuel assemblies (LRG II Issue 2-CPB)
(16) Scram discharge system Resolved in SSER 1 4.6 evaluation (17) Fracture toughness data Resolved in SSER 1 5.3.1, 5.3.2 5.3.3 (18) Subcompartment pressure Resolved in SSER 5 6.2.1.6 analysis (19) Combustible gas control Resolved in SSER 3 6.2.5 (20) Containment isolation Resolved in SSER 2 6.~2.4 dependability (21) Containment monitoring, Resolved in SSER 4 6.2.7, II.F.1(1) through II.F.1(6) and SSER 5 12.3.4.1 (22) Plant-specific LOCA analysis, Resolved in SSER 3 6.3.4 II.K.3.31 (23) High drywell pressure Resolved in SSER 1 6.3 interlocks (24) ATWS recirculation pump trip Resolved in SSER 6 15.2.1 (25) Response-time testing Resolved in SSER 1 - 7.2.3.2 (26) Analog trip modules and optical Resolved in SSER 2 7.2.3.4 isolators (27) Susceptibility of the NSPS to Resolved in SSER 1 7.2.3.5 electrical noise Clinton SSER 8 1-5
SER/SSER Issue -Status- Section(s)
(28) Modification of ADS logic, Resolved in SSER 4 7.3.3.4 II.K.3.18 (29) Restart of low pressure Resolved in SSER 1 7.3.3.5
< systems, II.K.3.21 (30) Temperature effects on level Resolved in SSER 2 7.5.3.2J measurements (31) Containment atmosphere Resolved in SSER 4 7.6.3.1 monitoring system (32) Verification that testing is Removed from list in 8.2.3 in accordance with BTP PSB-1 SSER l' (33) Electrical drawing review Removed from list in 7, 8 SSER 1 (34) Verification of diesel Resolved in SSER 4 8.3.1 generator testing
! (35) Class A supervision and power Resolved in SSER 3 9.5.1.4 supply for fire detection system (36) Circulating water system Resolved in SSER 2 10.6 (37) Initial test program Resolved in SER 14 (38) Human engineering aspects of Resolved in SSER 5 18 control room design, I.D.1 (39) Common reference for reactor Resolved in SSER 2 7.5.3.5 vessel level instruments, II.K.3.27 (40) Shielding design review, Resolved in SSER 1 12.3.2 II.B.2 (41) Short-term accident and- Resolved in SSER 4 13.6.3 procedures review, I . C .1,- and SSER 5 I.C.7, I.C.8 l (42)- Training during-low power Resolved in-SSER 5 . 14 testing, I.G.1 (43) Review ESF values, II.K.1.5 Resolved in SSER 1 6.3 (44) Operability status, II.K.1.10 Resolved in SSER 1 13.5
, Clinton SSER 8 , 1-6
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SER/SSER
. Issue ' Status Section(s)
(45): HPCI and RCIC initiation Resolved in SSER 4 6.3.2.3
.l evel s , II . K. 3.'13 -
4 (46) Iso'lation of HPCI and RCIC,- Resolved in SSER 4 '7.3.3.3 II.K.3.15 (47) Qualification of ADS Resolved in SSER 5 6.3.2.2 accumulators, II.K.3.28 (48) Plant-specific analysis, Resolved in SSER 3 6.3.4 II.K.3.30
- (49) ODYN analysis for River Bend Resolved in SSER 1 4.4.1 as applied to Clinton 4.4.1
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(50) Conformance evaluation report Resolved in-SSER 3
.for loose parts monitoring i system (51)L Requirements of NUREG-0313 Resolved in SSER 1 5.2.3 (52) Control room habitability - Resolved in SSER 1 6.4 chlorine gas (53) Debris screen design Resolved in SSER 2: 6.2.4.1' (54) Verification ~of' adequacy of- Removed from list in fire protection systems SSER 1
-I (55) Floca proof door Resolved in SSER 2 10.6 (56) Valves in fire protection Resolved in SSER l' 9.5.1.1-water supply system (57) Break in water supply piping Resolved in SSER l' 9.5.1.1 (58) Test data on fire-ratings Resolved in SSER 3- 9.5.2.1 (59) Three-hour-fire-rated Resolved in SSER 3 9.5.2.1-penetration seals (60) -Install fire protection Resolved in SSER 3 9.5.3.
equipment (emergency
_ lighting) -
(61) Fire-protection administrative Resolved in'SSER 1 9.5.6, controls and training (62) Technical Specification on Resolved in-SSER 1. 9.5~7.
fire protection i
IClinton SSER 8 1-7
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SER/SSER Issue Status Section(s)
(63) Periodic leak testing of Resolved in SSER 1 3.9.6 pressure isolation values (64) Sedimentation in UHS Resolved in SSER 1 2.4 (65) Protection against postulated Resolved in SSER 1 3.6.1 piping failures
-(66) Steam bypass of the suppres- Resolved in SSER 5- 6.2.1.7 sion pool (LRG II Issue 3-CSB)
(67) Preservice inspection program Resolved in SSER 5 5.2.4, 5.2.4.1, 6.6.1 (68) Inservice testing of pumps Resolved in SSER 6 3.9.6 and valves (69a) Containment low-volume purge Resolved in SSER 5 6.2.4.1 system (69b) Low-volume purge valve Resolved in SSER 5 6.2.4.1 operability ,
(70) Seismic analysis Resolved in SSER 3 " 2.5.2.4 (71) Humphrey concerns Resolved in SSER 6 6.2 1.11 License Conditions In SER Section 1.11, the NRC staff identified nine potential _ license conditions that may be required as part of the operating license for Clinton Unit 1 to ensure that NRC requirements are met during plant operations. The status of potential license conditions in the SER supplements 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 on one potential license ce.dition (6); SSER 3 identified one potential license condition (14); SSER 4 rm oi<ed one potential license condition,(8); SSER 5 identified six potentW Me nse conditions (15-20) and resolvid ten potential license cc.1ditions J L E , 9-11, and 13); SSER 6 identified three potential license conditions p -23; A d resolved four potential license conditions (5, 14, 18, and 19); and SSER 7 added one potential license condition (24) and modified one license condition (16). The nine remaining license conditions in SSER 7 (12, 15-17, and 20-24) were made license conditions on the Clinton low power license. SSER 8 resolves three license conditions (12,- 21, and 24).
Six of the nine license conditions that appeared in SSER 7 were not closed in SSER 8 (15, 16, 17, 20, 22, and 23); these have been made license conditions in the full power license.
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j Clinton SSER 8 1-8 l
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-SER/SSER l Issue Status Section(s) j (1) Staffing DeWitt pumping station Resolved in SSER S 2.2 !
SER/SSER ;
(2) New stability analysis before Resolved in SSER 5 4.4.1 i second cycle of-operation (3) Postaccident monitoring Resolved in SSER 5 7.5.3.1 (4) Vacuum relief valva position Resolved in SSER 5- 7.3.2.3 indication.
(5) Hydrogen management Resolved in SSER 6 6.2.7 (6) Postaccident sampling, II.B.3 Resolved in SSER 5 9.3.5 (7) . Diesel generator reliability Resolved in SSER 5 9.6.3.1 (8) Kuosheng-1 test program Resolved in SSER 4- '3.9.2, 6.2.1.8 (9) Visual. examination of Resolved in SSER 5 4.2.3.9 discharged fuel (10) Measurement of groundwater Resolved in SSER 5 2.4.6 level (11) Security Resolved in SSER 5 13.7 (12) Inservice inspection Resolved in SSER 8 5.2.4, (low power license 5.2.4.1, condition 2.C(4))* 6.6.1 q(13) Control of heavy loads Resolved in SSER 5 9.1. 5 (14) Transportation accidents Resolved in SSER 6 2.' 2.1 (15) Fuel zone level channels Addressed in SSER 5, 7.5.3.1 modified in SSER 8 (low power license condition 2.C(11)a)*
(16) Partial feedwater heating Addressed in SSER 5, 15.1 modified in SSER 7 (low power license condition 2.C(12))*
(17) Plant operator experience Addressed in SSER 5 13.1.2.1 (low power license
. condition 2.C(8))*
- Low power license condition refers to the Clinton low power facility operating license No. NPF-55.
-Clinton SSER 8 1-9 r .
(18) Emergency facilities and Resolved in SSER 6 13.3.2.8 equipment (19) Environmental protection plan Resolved in SSER 6 2.6.1.1 l for Unit 2 site (20) Detailed Control Room Design Addressed in SSER 5 18 Review '(low power license condition 2.C(11)b)*
l 7.5.3.2
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(21)- Safety parameter display system Resolved in SSER 8 l (low power license condition 2.C(11)c)*
(22) Fire protection Addressed in SSER 6 9.5 (low power license condition 2.C(7))*
(23) Control system failures Addressed in SSER 6 7.7.3.1-(low power license condition 2.C(5))*
(24) New-fuel storage Resolved in SSER 8 9.1.1 (low power license
- condition 2.C(6))*
- Low power license condition refers to the Clinton low power facility operating license No. NPF-55.
Clinton SSER 8 1-10
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 Vendor 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 l 'inton SSER 8 1-11
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' Table'1.2.(Continued);
Action Action' No. -SSER Section
. 44 Improvements in Maintenance and Test. N/A.
Procedures for B&W Plants
- 4.5' System Functional 4
- 4.5.1' . Test Diverse Trip Features- 15~.2.'2.4 (SSER 5)
- . 4.5.2 l Justify Not Modifying the' Reactor Trip *
- ' System to Permit Periodic On-Line. Testing
' 4. 5. 3 ~ Review Existing. Intervals for On-Line ~*
g_ Functional Testing Required.byLTechnical i
Specifications 1
i Notes: N/A = not applicable'to Clinton, i
- Staff's review is ongoing (SSER 6; -Section 15.2.2).
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! Clinton SSER 7 1-12
6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.5 Combustible Gas Control As discussed in the SER, the staff noted that a backup purge system is provided in accordance with Regulatory Guide (RG) 1.7. The purpose of the system is-to aid in atmospheric cleanup of the postaccident containment environment. As noted in RG 1.7, the purge / vent system may be a separate system or part of an existing system.
By letter dated March 2, 1987, the licensee indicated that the backup purge system was incapable of meeting the self-imposed design function requirements.
However, the licensee does not intend to pursue design changes on this system to achieve these objectives. Rather, as discussed in the submittal, the licensee cites the recent amendment to 10 CFR 50.44 and its compliance to the new hydro-gen rule as sufficient to fulfill the intent of the regulatory guide; i.e. ,
the hydrogen igniter system (HIS) installed at the Clinton Power Station as discussed in SSER 6, Section 6.2.7.
As another means of backup, the licensee indicated that hydrogen control through purging will be accomplished through the standby gas treatment system (SGTS) post-loss-of-coolant accident (LOCA) purge mode of operation. Also, as part of the severe accident response for hydrogen control, the licensee is in the process of developing emergency procedures guidelines (EPGs) to identify appropriate operator actions to support this purging operation. This effort will incitde the assessment of additional purge / vent paths that can be used for hydrogen control.
The staff has concluded that the HIS is not functionally equivalent to the-backup purge system; however, the purge operation in accordance with the EPGs during post-LOCA conditions does satisfy the RG 1.7 backup purge requirement.
Therefore, the staff finds that the intent of RG 1.7 has been met. Furthermore, the staff's review of the hydrogen control emergency procedures' guidelines is on going and, as part of the staff's evaluation, the vent paths selected for postaccident containment venting along with the consequences of venting hydrogen will be pursued generically in connection with the BWR Hydrogen Control Owners Group. Accordingly, upon completion of the staff's review, a plant-specific '
evaluation will be provided for Clinton. In the interim, the staff agrees that the SGTS vent path may be used for postaccident purging, provided the licensee's emergency procedures preclude the use of the system at pressures approaching the design capability of the SGTS and at containment hydrogen concentrations greater than 2%.
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Clinton SSER 8 6 t
6.6 Inservice Inspection of Class 2 and 3 Components 6.6.1 Evaluation of Compliance.With 10 CFR 50.55a(g)-
In Supplement 5 to the SER the staff stated that since the licensee had not submitted the initial inservice inspection program it would be made a license condition (license condition 2.C(4) to the Clinton low power Facility Operating License No. NPF-55). By letter dated November 7, 1986,.the licensee submitted the inservice inspection program plan for the first 10 year interval for piping and components and component supports and, therefore, satisfies the requirement of license condition 2.C(4) contained in the Clinton low power Facility Operat-ing License No. NPF-55.
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Clinton SSER 8 6-2 1 .
7 INSTRUMENTATION AND CONTROLS p 7.5 Information Systems Important to Safety 7.5.3 Resolution of Issues 7.5.3.1 Conformance-to Regulatory Guide 1.~97, Revision 3 In this section of Supplement 5 to the SER.a discussion was provided relative to the fuel zone range level instrumentation and the extent to which it con-formed to the guidance contained in RG 1.97. The purpose of this' supplement is to provide clarification.of-the information contained in Supplement 5 with regard to this issue.
The power sources provided to operate the'two redundant channels ~of.the fuel zone range level instrumentation at Clinton presently do not conform to the-guidance contained in RG 1.97 since both channels are powered from the same 120-V ac non-Class 1E power sources. By letter dated December 11, 1984, the licensee committed to modify this water level instrumentation to conform to the guidance provided in RG 1.97 before startup following the first fueling out-age. On the basis of the licensee's commitment to provide separate Class 1E power sources to operate each fuel zone range level instrumentation, the staff, based on its review of the Technical Evaluation Report provided by EG&G Idaho (SSER 5, Appendix G), finds the use of the existing instrumentation acceptable on an interim basis.
Since the fuel zone instrumentation currently does not conform.to RG 1.97 guide--
lines, the following full power license condition will replace license condi-tion 2.C(11)a contained in the Clinton low power Facility Operating License-No. NPF-55: '
The licensee, in accordance with the commitment contained in a letter dated December 11, 1987, shall provide separate power sources for each of the fuel zone level channels as recommended in the guidance ,
contained in RG 1.97 before startup after the first refueling outage. '
7.5.3.2 Safety Parameter Display System In Supplement 6 to the SER the staff evaluated and approved the modifications to the safety parameter display system (SPDS), related to the display of con-tainment isolation valve group status, proposed by the licensee in a letter dated June 11, 1986. A license condition (Facility Operating License No. NPF-55, paragraph 2.C(11)c) was imposed on the facility license that required the pro-posed modifications to be completed on the schedule contained in the licensee's June 11, 1986 letter. By letter dated January 28, 1987,-the licensee stated' that these modifications have been completed and the requirement of paragraph 2.C(11)c of the Clinton low power Facility Operating License No. NPF-55 has been fulfilled. On the basis of this letter, the staff has determined this-license condition is no longer required. Region III shall, as previously :
stated in Supplement 6 to 'the SER, verify the completion of these modifica-tions described in the licensee's June 11, 1986 letter.
Clinton SSER 8- 7-1
9 AUXILIARY SYSTEMS 9.1 Fuel Storage Facility 9.1.1 New Fuel Storage Containment Fuel Storage Pool SER Supplement 7 to the SER, the staff approved the dry storage of the fuel assemblies for the initial core in the containment storage pool and incorporated as part of license condition 2.C(6) in Clinton low power operating license NPF-55,-specific restrictions related related to the dry storage of these fuel assemblies (SSER 7, license condition 24). Since the initial fuel assemblies have been loaded into the core of the reactor vessel, this license condition is no longer required.
l j
1 J
4 Clinton SSER 8 9-1
12 RADIATION PROTECTION 12.3 Radiation Protection Design Features 12.3.1 Facility Design Features-
, In Supplement 5 to the SER the staff stated that the licensee, in response to TMI Action Plan Item III.D.1.1, has committed to a program to reduce leakage to as-low-as practical levels from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident. The staff determined the licensee's program was acceptable provided the initial leak rate tests results are submitted within 120 days after fuel load.
By letter dated January 27, 1987, the licensee stated that because reactor pres-surization will occur after the expiration of the 120-day period, the schedule for submittal will not be met because some of the systems to be inspected require reactor pressurization to provide meaningful leakage measurement results. The licensee, in the January 27, 1987 letter, requested that the submittal date for the test results be extended.
The staff has reviewed the licensee's submittal and determined that the defer-ral of submittal to provide the test results requested 's necessary and that the extension of the submittal date to 30 days after reaching 5% of rated power or 30 days after obtaining a full power license, whichever comes later, does not represent a safety concern and is therefore acceptable.
l \
I l
CLINTON SSER 8 12-1 !
l
l l
4 14 INITIAL TEST PROGRAM ,
Preoperational Test Deferrals In Supplement 6 to the SER the staff approved deferral of various preopera-tional tests with different completion schedules after issuance of the low-power license. Those preoperational tests that required schedular exemptions from Appendix A to 10 CFR Part 50 were discussed in Appendix N to Supplement 6 to the SER and appeared as schedular exemptions from Appendix A in the lower-power license No. NPF-55 (Section 2D, items a-i).
By letter dated March 3, 1987, the licensee stated that all the schedular pre-operational test deferrals through heatup granted from Appendix A to 10 CFR 50 in Section 20 of the low power license No. NPF-55 have been completed in accor-dance with completion schedules specified in Section 2D of the low power license.
Since the licensee has completed these preoperational tests in accordance with the schedules contained in the low power license, these exemptions identified in Table 14.1 are no longer required.
By letter dated March 12, 1986 the licensee requested a deferral of fuel handling (FH) system preoperational testing related to the transfer of fuel bundles under wet loading conditions until before exceeding 5% of rated reactor power. By letter dated March 27, 1986, the licensee requested an exemption to 10 CFR 50, i
Appendix A, General Design Criterion (GDC) 61 for the operation of the FH system under wet conditions. GDC 61 requires, in part, that the FH system be designed to prevent significant reduction in fuel storage coolant inventory under acci-dent conditions.
The initial fuel loading has been performed under dry conditions, and operation of the FH system under wet conditions was not required. Consequently, the staff granted a schedular exemption from the requirements of GDC 61, and per-mitted deferral of completion of preoperational testing of the FH system under wet loading conditions until before exceeding 5% of rated reactor power, or before removal of the reactor pressure vessel head after initial criticality.
By letter dated October 24, 1986, the licensee requested a revision to the exemption to 10 CFR 50, Appendix A, GDC 61, for preoperational testing of the FH system under wet loading conditions until before off-loading irradiated fuel (prior to removal of the reactor pressure vessel head after initial criticality).
I The licensee's current request for revision to the exemption previously granted above is to remove the restriction on test completion before exceeding 5% of rated reactor power and to allow continued power operation up to full power and until prior to off-loading irradiated fuel or before removing the reactor pressure vessel head after initial criticality.
The licensee stated that performance of this test before exceeding 5% of rated reactor power would delay completion of the startup test and power ascension programs.
Clinton SSER 8 14-1
The staff has determined that potential events which could require emergency off-loading of irradiated fuel are possible, such as repairing a reactor vessel internal component; however, there are no events which require off-loading of fuel in a time frame that would preclude performance of this preoperational test on the FH system first.
-Transfer of irradiated fuel is not required until either refueling, or for maintenance purposes that require the vessel head to be removed. Therefore, it is acceptable to defer testing until such operations are necessary.
On the basis of its review, the staff concludes that the schedular exemption to 10 CFR 50, Appendix A, GDC 61, requested by the licensee is acceptable provided that the preoperational testing of the FH system under wet loading conditions is completed before the removal of the reactor pressure vessel head after initial criticality. This schedular exemption will be included in the full-power license.
Table 14.1 Completion of deferred preoperational testing of systems that required schedular exemptions Low power license NPF-55 Plant system exemption /SSER 6 section Completion milestone Turbine electrohydraulic 2.D(a)/ Appendix N, S2.1 Heatup control Offgas 2.D(c)/ Appendix N, 62.3 Heatup Containment monitoring 2.D(d)/ Appendix N, $2.4 Initial criticality Leakage detection 2.D(e)/ Appendix N, 62.5 Initial criticality In place filter testing of 2.D(h)/ Appendix N, 62.8 Initial criticality control room heating, ventilation, and air conditioning Heating, ventilation, and 2.D(i)/ Appendix N, S2.9 Heatup air conditioning Clinton SSER 8 14-2
.a . .w. a w w.; .- . ~.u=. :. .- - -
i 15 SAFETY ANALYSIS
- l. 15.6 Single Recirculation Loop Operation I l
By letter dated October 7, 1986, the licensee submitted a single recirculation ;
loop operation (SLO) analysis and also requested Technical Specification changes necessary to implement SLO for the operation of the Clinton Power Station. The;.
, staff has reviewed the licensee's evaluation of accidents and abnormal opera- j
, tional transients with only one pump operative. This evaluation was performed ;
, for a P8X8R fueled core on an equilibrium cycle basis up to a maximum power of '
l approximately 70% of rated power. The analysis is applicable to both the ;
- initial fuel cycle and reload cycles.
4
! The licensee provided a General Electric (GE) report entitled, "Clinton Single
{ issues contained in this report and the proposed Technical Specification changes pertaining to Clinton Power Station to justify extended operation with one loop:
out of service follows. ;
- 15.6.1 MCPR Fuel Cladding Integrity Safety Limit The net effect of increased uncertainties in the core total flow and traversing i in-core probe (TIP) readings for the single-loop operation is a 0.01 incremen-a tal increase in the minimum critical power ratio (MCPR) fuel cladding integrity
safety limit. Operating with one recirculation loop results in a maximum power; i output which is about 30% below that which is attainable for two pump opera-i tion. Therefore, consequences of abnormal operation transients from one-loop i- operation will be less severe.than those from a full power two-loop operational mode as provided in the FSAR.
l The transient peak value results and critical power ratio (CPR) results for
} load rejection with bypass failure (LR8PF) and feedwater controller failure l (FWCF) with maximum demand are summarized below. i i
SIM1ARY OF TRANSIENT PEAK VALUE AND CPR RESULTS LRBPF FWCF 4
Initial power / flow (% rated) 70.2/53.6 70.2/53.6 Peak neutron flux (% NBR) 70.3 83.6
! Peak heat flux (% initial) 100.3 106.5
. Peak dome pressure (psig) 1167 1051 l
Peak vessel bottom pressure (psig) 1181 1065 Required two-loop initial MCPR ,
operating limit at SLO condition 1.39 1.39 ACPR 0.05 0.12
L Clinton SSER 8 15-1
I l
'l For the two limiting transient events analyzed, the MCPRs are all above the single-loop operation safety limit value of 1.07 so that'no fuel failure will occur because of boiling transition. The peak vessel pressures are all below the.ASME Code value of 1375 psig. Therefore, since the pressure barrier integ-rity is maint'ained under single-loop operation conditions, the staff has deter-mined this is acceptable.
15.6.2 MCPR Operating Limit (1) Accidents (Other Than LOCA) and Transients Affected by One Recirculation Loop Out of Service One Pump Seizure Accident Plant-specific analysis was not performed for this event. However previous analysis for the Monticello plant has shown that this event results in an MCPR value significantly above the SLO safety limit MCPR (Musolf, 1986).
This was also stated in the Clinton FSAR, Section 15.3.3, " Recirculation Pump Seizure" which was approved by the staff, and further documented in a transmittal by the licensee on December 15, 1986 (Riley).
(2) Abnormal Operating Transients
. Although the increased uncertainties in core total flow and TIP readings
! resulted in a 0.01 increase in MCPR fuel cladding integrity safety limit during single-loop operation, the limiting-transients analyzed in the GE report (GE, 1986) indicate that there is more than enough MCPR margin dur-ing single-loop operation to compensate for this increase in safety limit.
For single-loop operation at off-rated conditions, the steady-st'te a operat-ing MCPR limit is established by the power-dependent and flow-dependent MCPR curves. For the most limiting transient events analyzed, the GE report also shows that the present power-dependent MCPR limits are bounding for single-loop operation. Further, the present flow-dependent MCPR limits are also bounding for single-loop operation since the maximum core flow runout during single-loop operation is only about 54% of rated core flow runout. Since the transient consequence from one-loop operation is bounded by previously submitted full power analyses, the staff has determined this is acceptable.
(3) Rod Withdrawal Error The rod withdrawal error at rated power is given in the FSAR for the ini-tial core and in cycle-dependent reload supplemental submittals. These analyses were performed to demonstrate that, even if the operator had a ignored all instrument indications and alarms during the course of the
- transient, the rod block system would stop rod withdrawal at_an MCPR which is higher than the fuel cladding integrity safety limit. The GE report also shows that correction of the rod block equation for single-loop operation ensures that the MCPR safety limit is not. violated.
One pump operation results in backflow through 10 of the 20 jet pumps while i
flow is being supplied to the lower plenum from the 10 active jet pumps.
Because of this backflow through the inactive jet pumps, the present rod block equation and APRM settings were modified for use during one pump Clinton SSER 8 15-2 i
E operation. The staff has reviewed these modifications to_the rod block equation and APRM settings and found them acceptable.
The staff has determined that one-loop transients and accidents other.than
- loss-of-coolant accident (LOCA), which is discussed below, are bounded by the two-loop operation analyses and are, therefore, acceptable.
15.6.3 Stability Analysis With one recirculation loop not in service, the primary contributing factors to the stability performance are the power / flow ratio and the recirculation loop characteristics. At forced circulation with one recirculation loop not.in operation, the reactor core stability is influenced'by the inactive recircula-tion loop. Staff evaluations'have considered whether increased noise in SLO.
was being caused by reduced stability margin as SLO core flow was increased.
, Results of analyses and tests-indicate that the SLO stability characteristics are not significantly different from two-loop operation.
^
At low core flows, SLO may be slightly less stable than two-loop operation, but as core flow is increased and reverse flow is established, the stability performance is similar.
At' higher core flows with substantial-reverse flow in the inactive recircula-tion loop, the effect of cross-flow on the flow noise results in an increase in system noise (jet pump, core flow, and neutron flux noise), but core thermal-hydraulic stability margin is very high, similar to two-loop operation. GE has f
developed a Service Information Letter (SIL-380), (GE, 1984) informing plant operators how to recognize and suppress unanticipated oscillations when encountered during plant operation. The NRC staff has approved the GE generic stability analysis for application to single-loop operation (Thomas, 1985) provided that the recommendations of-SIL-380 have been incorporated into the plant Technical Specifications. Since the licensee has incorporated the recom-mendations of SIL-380 into the Clinton Technical Specifications, the staff has determined this is acceptable.
15.6.4 Loss-of-Coolant Accident Analysis SAFE /REFLOOD calculations were performed for a full spectrum of large break sizes for the recirculation suction line breaks for the single-loop operation.
The small differences in uncovered time and reflood time for the limiting break size, i.e., 169 seconds for the single-loop vs. 168 seconds for the two-loop operation, would result in a small change in the calculated peak cladding temperature. The maximum average planar linear heat generation rate (MAPLHGR) reduction factors for the most limiting single-loop operation for P8X8R fuel is 0.85, which is conservative.
In the event of a small-break LOCA, the slight increase (s50 F)-in peak clad-ding temperature (PCT) is offset by the effect of the decreased MAPLHGR (equi-
< valent to 300 F to 500 F PCT) for the single-. loop operation. -The calculated PCT values for small breaks will therefore be well below the 1496*F PCT value previously analyzed for small breaks which is acceptable to the staff.
15.6.5 Containment Analysis
! The GE analysis (GE, 1986) indicates that under SLO conditions limiting-case ,
i accidents would result in peak containment pressures, containment temperatures, I and suppression pool temperatures which are less severe than'those estimated i Clinton SSER 8 15-3
for design-basis accidents under two-loop operation. GE also evaluated the chugging and pool swell load under SLO conditions and stated that these loads are less severe than those estimated for accidents postulated during two-loop operation. The staff has reviewed the GE analysis and determined it to be acceptable.
15.6.6 Miscellaneous Impact Evaluation (1) Anticipated Transient Without Scram Since the SLO initial power / flow condition is less than the rated condition used for the two-loop anticipated transient without scram (ATWS) analysis, GE found the transient response less severe and therefore bounded by the FSAR analyses. The staff has determined this is acceptable.
(2) Fuel Mechanical Performance As a result of the substantial reverse flow established during SLO, both the average power range monitor (APRM) noise and core plate differential pressure noise are increased slightly. GE has stated that the APRM fluctuation should not exceed a flux amplitude of 115% of rated amplitude and the core plate dif-ferential pressure fluctuation should not exceed 3.2 psi peak to peak if opera-tion is to be consistent with the fuel rod and assembly design bases. The staff has determined this is acceptable.
(3) Vessel Internal Vibration GE imposed a recirculation pump drive flow limit for single-loop operation, which is about 33,000 gpm for rated reactor water temperature and pressure.
This is based on measured prototypical value from the Kuosheng Unit 1 plant which has been accepted by the staff as the valid prototype for Clinton Unit 1 in SSER 4'(NUREG-0853). With maximum flow thus limited, vibration levels of the reactor internal components will be within acceptance limits during SLO at Clinton Power Station. The staff has determined this is acceptable.
. 15.6.7 Technical Specification Changes The staff has reviewed all the Technical Specification (TS) changes proposed by the licensee and found them to be consistent with results of the GE analysis and also with TS changes approved for other BWR/6 plants for single-loop opera-tion. On the basis of this review, the staff has determined the proposed TS changes are acceptable.
I 15.6.8 Evaluation Findings The licensee has submitted SLO analysis and proposed TS changes which would incorporate requirements for extended operation with one recirculation loop out of service. The staff has evaluated the licensee submittal and determined that the analysis provided by the licensee and proposed TS changes for extended SLO are acceptable.
Clinton SSER 8 15-4
4 16 TECHNICAL SPECIFICATIONS The Technical Specifications in a license define certain features, characteris-tics, and conditions governing the operation of the facility that cannot be changed without prior approval of the NRC staff. The Clinton Technical Speci-fications are included as Appendix A to the Clinton license. Included in the Technical Specifications are sections covering definitions, safety limits, limiting safety system settings,-limiting conditions for operation, surveillance requirements, design features, and administrative controls.
In letters dated January 8, February 4, and March 3 and 20,.1987, the licensee requested certain changes to the Technical Specifications issued as part of the Clinton low power license. These changes would be incorporated into the Tech-nical Specifications for the full power license and would be effective upon issuance of that license. The licensee requested these changes as a result of experience to date, to clarify and enhance the Clinton Technical Specifica-tions, and to provide consistency between the FSAR, the SER, and the as-built facility. In addition, the staff audited the final Clinton Technical Specifi-
, cations in accordance with NRR* Office Letter No. 51, dated July 31, 1986.
Several of the findings of this audit identified in a November 20, 1986 memo-randum resulted in staff changes to the Clinton Technical Specifications. The licensee's requested changes, their status, and the staff changes are contained in Table 16.1. The new Technical Specification requests contained in the March 20, 1987 submittal will be addressed in a future safety evaluation report.
- NRC's Office of Nuclear Reactor Regulation.
Clinton SSER 8 16-1
i Table 16.1 Technical Specification changes O
F Date of IP letter request
{ and TS section Licensee change request description
- Staff evaluation **
y January 8, 1987 TS 3/4.3.1, Table 4.3.1.1-1, Change frequency of surveillance require- Unacceptable. See Appendix Q, pp. 3/4 3-8, -9, -10 ments for some of the reactor protection TS 3/4.3.1, pp. 3/4 3-8, -9, system instrumentation. -10.
TS 3/4.3.2, Table 3.3.2-1, Add a note to identify operability relief Acceptable. See Appendix Q,
- p. 3/4 3-13 and Table 3.3.2-1, for the containment high pressure trip TS 3/4.3.2, pp. 3/4 3-13, -18.,
- p. 3/4 3-18 function for containment isolation when associated valves are sealed closed.
TS 3/4.3.2, Table 3.3.2-1 Modify Action statements 21, 25, and 29 Acceptable. See Appendix Q,
- p. 3/4 3-19 to provide clarification. TS 3/4.3.2, p. 3/4 3-19.
- TS 3/4.3.7, Table 3.3.7.1-1, Permit a channel of the main control room Acceptable. See Appendix Q, i M pp. 3/4 3-71, -72 air intake radiation monitor to be placed TS 3/4.3.7, pp. 3/4 3-71, -72 4 in inoperable status during required sur-veillance without placing the tripped sys-
' tem in the tripped condition, provided there is at least one other operable channel.
TS 3/4.6.4, p. 3/4 6-29 [ Action statement for primary containment The BWR licensing staff agrees isolation valves should address valves with the audit findings and a which have dual functions and the effects footnote has been added to this of their inoperability on their other specification identifying the safety function.]t problem.
TS 3/4.6.4, Table 3.6.4-1, Update the TS to satisfy the commitment of Acceptable. See Appendix Q, 4
- p. 3/4 6-38 '
TMI Item II.E.4.2 as described in Appen- TS 3/4.6.4, p. 3/4 6-38.
dix D of the FSAR.
See footnotes at end of table.
l
- - . . - - .- - ... - - - - . - --- .~ .. . - .- .- -. .-.
b I
a Table 16.1 (Continued) e
- E Date of IP letter request
- y and TS section Licensee change request description
- Staff evaluation **
h January 8,1987 (Continued) 1
.] TS 3/4.6.4, Table 3.6.4-1, Delete. valve IE12-F351 from the table. Acceptable. See Appendix Q, j p. 3/4 6-44 TS 3/4.6.4, p. 3/4 6-44.
4 t l TS 3/4.6.4, Table 3.6.4-1, Delete valves IB21-F098A, B, C, D from the Acceptable. See Appendix.Q, pp. 3/4 6-51, -52 table. TS 3/4.6.4, pp. 3/4 6-51, -52.
i TS 3/4.6.4, Table 3.6.4-1, Delete valves IIA 128A, B from the table. Acceptable. See Appendix'Q, i pp. 3/4 6-56, -60 TS 3/4.6.4, pp. 3/4 6-56, -60.
j TS 3/4.6.4, Table 3.6.4-1, Add a note (c) to valve ICM053 to be Acceptable. See Appendix Q, j p. 3/4 6-59 consistent with FSAR. TS.3/4.6.4, p. 3/4 6-59.
TS 3/4.6.4, Table 3.6.4-1, Delete words in note (g) and insert revised Acceptable. See Appendix Q, 5 p. 3/4 6-61 note wording to satisfy TMI Item II.E.4.2 TS 3/4.6.4, p. 3/4 6-61.
j O requirements.
4 TS Bases B 3/4.6.4, Revise the Bases to clarify that TMI Item Acceptable. See Appendix Q,
! p. B 3/4 6-7 II.E.4.2 is satisfied for containment TS Bases B 3/4.6.4, p..B 3/4 6-7.
l integrity.
! TS 3/4.7.6, p. 3/4 7-17 [An inconsistency exists between the The staff agrees with the 3 applicability and Action statement.]t audit findings. See .
Appendix Q, TS 3/4.7.6,
- p. 3/4 7-17.
TS 3/4.8.4, Table 3.8.4.1-1, Delete from table columns entitled: "Cir- Acceptable. These columns are-pp. 3/4 8-27 through 43 cuit Breaker Trip," " Penetration Cable purely. informational and do not-I Size," and " Cable. Number." contain any safety requirements i
related to the TS.
See footnotes at end of table.
3 i a 1
Table 16.1 (Continued)
'3 (g Date of IP letter request and TS section Licensee change request description
- Staff evaluation **
h January 8, 1987 (Continued)
[ TS 3/4.8.4, Table 3.8.4.2-1, Add valves IWO551A, B and 1WO552A, B to the Acceptable. This is in accordance
- p. 3/4 8-52 table of motor-operated valves that have with Position C.1 of RG 1.106 thermal overload protection devices. " Thermal Overload Protection for Electric Motors on MOVs."
TS 3/4.9.12, p.3/4 9-19, -20 Extend the surveillance times to determine See staff evaluation under March 3, the operability of the inclined fuel 1987 submittal, TS change 3/4.9.12, transfer system (IFTS) after system is pp. 3/4 9-19, -20.
determined operable.
TS 6.1.2, p. 6-1 Change the appropriate level of management Acceptable. See Appendix Q, to the Vice President-Nuclear for issu- TS 6.1.2, p. 6-1.
ance of the endorsement letter describing y the control room command function.
TS 6.2.1, Figure 6.2.1-1, Change the organizational chart to reflect Acceptable. See Appendix Q,
- p. 6-3 current organization. TS 6.2.1, p. 6-3.
TS 6.4.1, p. 6-7 Revise training requirements for the unit Unacceptable. See Appendix Q, staff. TS 6.4.1, p. 6-7.
TS 6.5.1.2, p. 6-7 Change the membership of Facility Review Acceptable. See Appendix Q, ;
Group. TS 6.5.1.2, p. 6-7. '
February 4, 1987 TS 1.28, Table 1.2, Delete word "recoupled" and insert word Acceptable. See Appendix Q,
- p. 1-11 " moved." Additional justification provided TS 1.28, p 1-11.
in March 3, 1987 submittal.
TS 3/4.2.3, Figure 3.2.3-2, Replace current figure, which was incor- Acceptable. This is an editorial
- p. 3/4 2-9 rectly drafted, with new figure. change and has been incorporated into the TS.
See footnotes at end of table.
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. - - - . - - _ . . . _ - - . _ _ _ . - _ - _ - _ - _ - . - - _ ~ _ - - - . _ . - - - . - . - . . . .. -
l 1
n Table 16.1 (Continued) i g Date of IP letter _ request g and TS section Licensee change request description
- Staff evaluation **
k{ February 4, 1987 (Continued) m oo TS 3/4.3.6, Table 4.3.6-1, Change the surveillance frequencies for Unacceptable. See Appendix Q, t
- p. 3/4 3-68 channel functional' tests for some control TS 3/4.3.6, p. 3/4 3-68. '
i rod block instrumentation.
'i TS 3/4.3.7, Substitute Action statement 121 for_126 Acceptable. See Appendix Q, Table 3.3.7.12-1, for items 2.a and 2.b. Delete Action 126. TS 3/4.3.7, pp. 3/4 3-102, -104.
pp. 3/4 3-102, -104 TS 3/4.3.9,-Table 3.3.9-2, Change the trip setpoint and allowable. Acceptable. See Appendix Q, i
i p. 3/4 3-113 value for the high containment pressure TS 3/4.3.9,-p. 3/4 3-113.
trip setpoints.
1 1 y TS 3/4.3.10, p. 3/4 3-115 Change the note to permit self-test system Acceptable. See Appendix Q, T to be taken out of automatic mode of opera- TS 3/4.3.10, p. 3/4 3-115.
4 u'
tion for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to perform. surveil-lance testing, preventive or corrective maintenance.
TS 3/4.4.1, SR 4.4.1.1,tt -Correct a typographical error. Acceptable. Change is editorial
- p. 3/4 4-2 and has been incorporated into the TS.
l TS 3/4.4.4,-SR 4.4.4.c,tt Delete the number of days the continuous Acceptable. See Appendix Q, 4 -p. 3/4 4-15 recording conductivity monitor may be in- TS 3/4.4.4, p. 3/4 4-15.
operable before obtaining in-line conduc-tivity measurements.
4 TS 3/4.6.2, SR 4'6.2.2,tt.
. Change to permit the drywell bypass leak- Acceptable. See March 3, 1987
- p. 3/4 6-15 ' age rate test to be performed during each submittal, TS 3/4.6.2, p. 3/4 6-15.
refueling outage.
t See footnotes at end of table.
^
Table 16.1 (Continued)
O F Date of IP letter request
{ and TS section Licensee change request description
- Staff evaluation **
y February 4, 1987 (Continued)
TS 3/4.6.3, SR 4.6.3.1.c.3,tt Correct a typographical error. Acceptable. Change is editorial
- p. 3/4 6-25 and has been incorporated into the TS.
TS 3/4.6.4, Table 3.6.4-4, Add a note related to B and R isolation Acceptable. See Appendix Q, pp. 3/4 6-32, -33 signals for valves IE51-F031 and IE51-F064. TS 3/4.6.4, pp. 3/4 6-32, -33.
TS 3/4.6.4, Table 3.6.4-1, Correct a typographical error. Acceptable. Change is editorial
- p. 3/4 6-51 and has been incorporated into the TS.
TS 3/4.8.4, Table 3.8.4.2-1, Add valve IE51-C002E to table of motor- Acceptable. See Appendix Q,
- p. 3/4 8-49 operated valves with thermal overload TS 3/4.8.4, p. 3/4 8-49.
M protection.
E TS 3/4.9.12, pp. 3/4 9-19, -20 Change the appropriate note to state that See staff evaluation under two components of the inclined fuel trans- March 3, 1987 submittal.
fer system need not be operable until prior TS 3/4.9.12, pp. 3/4 9-19, -20.
to off-loading irradiated fuel.
TS 3/4.4.3, p. B 3/4 4-3 Withdrawn TS 5.1.1 and 5.1.3, Change unrestricted area boundary as shown Acceptable. See Appendix Q, Figures 5.1.1-1 and 5.1.3-1, in figures. TS 5.1.1 and 5.1.3, pp. 5-2, -4.
pp. 5-2, -4 TS 6.2.2, Change the " Director" Plant Operations to Acceptable. See Appendix Q, Figure 6.2.1-1,' p. 6-4 " Assistant Manager" Plant Operations and TS 6.2.2, p. 6-4.
correct typographical error.
See footnotes at end of table.
Q Table 16.1 (Continued)
'3 ET Date of IP letter request is and TS section Licensee change request description
- Staff evaluation **
y February 4, 1987 (Continued) .
" TS 6.2.3.4, p. 6-6 Acceptable. . Change is editorial Correct a typographical error.
- and has been. incorporated into the TS.
TS 6.7.1.d, p. 6-14 Require safety limit violation report to Acceptable. See Appendix Q, be provided to all parties within 30 days TS 6.7.1.d, p.'6-14.
of violation.
TS 6.12, p. 6-23 Incorporated portions of TS 6.12.2 and TS Acceptable. See Appendix Q, 6.12.3 inadvertently left out upon issu- TS 6.12, p. 6-23.
ance of low power license into these TS sections.
- g. March 3, 1987
' TS 3/4.3.2, Table 3.3.2-2, Withdrawn.
- p. 3/4.3-23 TS 3/4.3.3, Table 3.3.3-2, Change the trip setpoint and allowable Acceptable. See Appendix Q,-
pp. 3/4 3-39 -40 value for the low pressure systems injec- TS 3/4.3.3, pp. 3/4 3-39, -40.
tion valve permissives. Additional justi-fication provided in March 20, 1987 ,
submittal.
TS 3/4.3.7, LC0 3.3.7.12,tt Add a note to table applicable to minimum Acceptable. See Appendix Q, Table 3.3.7.12-1, number of operable channels to enable TS 3/4.3.7, pp. 3/4 3-102, -103, pp. 3/4 3-102, -103, -104 channels to be placed in an inoperable
-104.
status for up to I hour to perform
-surveillance of this TS and TS 3/4.11.2.1.-
TS 3/4.4.3, SR 4.4.3.2.1.a,tt Delete surveillance requirement related to Unacceptable. Change to TS
- p. 3/4 4-11 monitoring drywell atmospheric particulate proposed is inadequate to and gaseous radioactivity once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. support requested change.-
See footnotes at end of table.
Table 16.1 (Continued)
O F Date of IP letter request g and TS section Licensee change request description
- Staff evaluation **
- s N March 3, 1987 (Continued) '
E TS 3/4.6.1, SR 4.6.1.4.c.2,tt Change the surveillance requirements of.the Acceptable. See Appendix Q,
- p. 3/4 6-7 MSIV leakage control. system blowers. TS 3/4.6.1, p. 3/4 6-7..
TS 3/4.6.2, SR 4.6.2.2,tt Change to permit the drywell bypass leakage Acceptable. See Appendix Q,
- p. 3/4 6-15 rate test to be performed during each re TS 3/4.6.2, p. 3/4 6-15.
fueling outage. Amends the February 4, 1987 submittal.
TS 3/4.9.12, Additional changes related to inclined fuel Acceptable. See Appendix Q, pp. 3/4 9-19,-20 transfer system (IFTS) TS were provided TS 3/4.9.12, pp. 3/4 9-19, -20.
to supplement January 8, 1987 submittal.
These changes ensure that sufficient pre-cautions are taken to prevent personnel E from entering areas adjacent to where
& irradiated fuel is being handled by the IFTS.
TS Bases 3/4.4.3, Update Bases to be consistent with a revi- Unacceptable. Change to TS
- p. 8 3/4 4-3 sion to the FSAR. (Related to revisian on proposed is inadequate to TS 3/4.4.3, SR 4.4.3.2.1.a p. 3/4 4-11 of- support requested change.
March 3, 1987 submittal.)
March 20, 1987 TS 3/4.3.3, Table 3.3.3-2 Clarification of same TS change request See TS 3/4.3.3, pp. 3/4 3-39 -40, pp. 3/4 3-39, -40 contained in the March 3, 1987 submittal.- under March 3, 1987 submittal.
TS 3/4.3.7, p.- 3/4.3-92 and Modify appropriate TS.to account for an TS Bases 3/4.3.7.8, alternative method for use and storage of
- p. B 3/4 3-7 and TS 3/4.7.2, chlorine gas on site in individual con-
- p. 3/4 7-5 tainers having an inventory capacity of
< 150 pounds.
See footnotes at end of table.
A- _ _ _ _
Table 16.1 (Continued)
O I Date of IP letter request y and TS section Licensee change request description
- Staff evaluation **
h March 20, 1987 (Continued)
[ TS 3/4.6.1., SR 4.6.1.7,tt Add a note regarding the instrument loca-
- p. 3/4 6-11 tions and the number of instruments to be used to determine primary containment average air temperature.
TS 3/4.6.2, SR 4.6.2.6,tt Add notes regarding the instrument loca-
- p. 3/4 6-20 tions and the number of instruments to be used to determine drywell average air temperature.
TS 3/4.8.2, Revise the load profile for the Division SR 4.8.2.1.d.2.att I, II, and III batteries due to installa-l g pp. 3/4 8-13, -14 tion of a prelubrication system on the i
9 diesel engines.
l e
! *For a more complete description, see the referenced letter.
- Appendix Q is in this supplement.
tBracketed material [] is a staff audit recommendation.
i ttSR = Surveillance Requirement; LCO = Limiting Condition for Operation.
I
APPENDIX Q STAFF SAFETY EVALUATIONS FOR THE TECHNICAL SPECIFICATION CHANGES PROPOSED BY THE LICENSEE AND THE STAFF FOR THE CLINTON FULL-POWER OPERATING LICENSE By letters dated January 8, February 4, and March 3, 1987, the licensee pro-posed changes to the Clinton Technical Specifications (TS). The changes were requested for issuance with the Clinton full power operating license. The staff evaluations of the proposed changes are given below. In addition, the staff evaluation of a staff audit recommendation is also included.
TS 1.28; page 1-11 During operational conditions of hot or cold shutdown, the present TS allows placement of the reactor mode switch (RMS) in the refuel position while a single control rod is being moved solely for recoupling, provided that the one rod-out interlock is operable (TS Table 1.2). The licensee has proposed to allow for movement of a single control rod for reasons other than recoupling. The li-censee, in a submittal dated March 3, 1987, stated that the design basis for the one-rod-out interlock is such that the single, highest worth control rod can be removed from the reactor core at any time while preserving the required shutdown margin. The change will not adversely affect the safe shutdown of the reactor and is similar to changes approved for other facilities; therefore, the staff finds the change acceptable.
TS 3/4.3.1; pages 3/4 3-8, -9, -10 The licensee proposed to change, in TS Table 4.3.1.1-1, the frequency of the channel functional test for certain reactor protection system instrumentation to quarterly from the present weekly or monthly. Justification for this change is based on a technical analysis provided in GE's Topical Report NEDC-30851P, submitted by the BWR Owners Group for generic applicability. However, the staff's generic review of NEDC-30851 P has not been completed. Until this review is completed and the applicability of the report to the Clinton Power Station is evaluated, the proposed change is unacceptable.
TS 3/4.3.2; pages 3/4 3-13, -18 This requested change would add a note to Table 3.3.2-1 providing operability relief for the containment pressure monitor when the two associated automatic isolation valves are sealed closed per TS 3/4.6.4. This change would bring the TS into agreement with the FSAR commitments. Since the valves are required to be closed to satisfy TMI Action Item II.E.4.2, the need for the operability of instrument signal to the valves is nullified and the change is acceptable.
TS 3/4.3.2; page 3/4 3-19 The requested change would modify Actions 21, 25, and 29 of Table 3.3.2-1 to delineate more clearly the actions required relative to instrumentation of the Clinton SSER 8 1 Appendix Q
containment and reactor vessel isolation control system (CRVICS) under specific operational conditions. The present TS is based on the Grand Gulf TS model; the proposed revision represents the Clinton specific design. Since the revised Actions are consistent with the intent of the boiling water reactor Standard Technical Specifications (BWR STS) (Draft BWR/6), they are acceptable.
TS 3/4.3.3; pages 3/4 3-39, -40 The staff has reviewed the proposed TS changes to Table 3.3.3-2, " Emergency Core Cooling System Actuation Instrument Setpoints," for the low pressure core injection (LPCI) and low pressure core spray (LPCS) injection valve permissives.
The proposed TS changes for items A.1.C and B.1.C, the trip functions for
" reactor vessel pressure - low," are as follows:
Trip setpoint: 472 psig Allowable value: 1 452 psig and 2 478 psig The signal to actuate the LPCS/LPCI injection valve interlocks originates from Rosemount pressure transmitter 11538. The licensee stated (Memorandum from B. Siegel, NRC, to W. Butler, NRC, dated March 26, 1987) that based'upon a revised GE methodology a more conservative basis is used for determining the setpoint of the pressure transmitters. Since a complete release of fission products from an equilibrium core at time zero post-LOCA is assumed, the ac-curacy of the existing pressure transmitters would be significantly affected by the radiation associated with a loss-of-coolant accident (LOCA). Therefore, replacement pressure transmitters (Rosemount 1153R) having higher radiation tolerance and higher accuracy are now being used.
The trip setpoint and allowable values for the LPCI/LPCS valve permissives are determined from analytic values to ensure the valve opening: (1) will not over-pressurize the low pressure portions of the LPCI or LPCS systems, and (2) will initiate in sufficient time to prevent core damage.
The staff expressed concern about changes to the analytical limit and the poten-tial effects on the emergency core cooling system (ECCS) analysis. By letter dated March 20, 1987, the licensee advised that a new ECCS analysis had been performed for Clinton. The impact of three changes to input parameters for i 1
the Clinton LOCA design-basis accident were evaluated. This included (1) up- '
dated analytical limits for the LPCI/LPCS injection valve permissive, (2) up-dated analytical limits for level setpoints LL1, LL2, and LL3, and (3) motor-operated valve (MOV) closure time relaxation. The resulting peak cladding temperature (PCT) increased to 2105*F, which is below the 2200*F limit. Most of the PCT increase was due to the relaxation in MOV closure time. The in-crease in cladding oxidation fraction was not significant. The staff finds the proposed Technical Specification change acceptable since the licensee has met the requirements of 10 CFR 50.46.
TS 3/4.3.6; page 3/4 3-68 The proposed change increases the channel functional test surveillance interval from at least once every 30 days to at least once every 90 days. The licensee justified this change based on the BWR Owners Group report (BWROG-8623) en-titled " Technical Specification Analysis for Control Rod Block Instrumentation."
The staff's generic review of this report is scheduled for completion in the Clinton SSER 8 2 Appendix Q
third quarter of 1987. Until the review is completed and the applicability of
~
the report to the Clinton Power Station is evaluated, the staff finds these proposed changes unacceptable.
TS 3/4.3.7; pages 3/4 3-71, -72 The requested change would add a note to the Table'3.3.7.1-1 table notations that would modify the surveillance Actions for. area monitors and pretreatment offgas monitors, providing at least one other operable channel in the same trip system that is monitoring that parameter. The staff does not consider that channels under functional test.and/or calibration modes are in " inoperable status," which requires the appropriate Actions specified in the' Limiting Condi-tion for Operation (LCO). Thus, the staff finds that the licensee's request to-add the proposed note "a"1to (1) item 1 on page 3/4 3-71 and (2) page 3/4 3-72, as a-table notation, is acceptable.
TS 3/4.3.7; pages 3/4 3-102,-103, -104 The requested changes would add a note (#)-to Table 3.3.7.12-1, table notations, on page 3/4 3-104, and to the Minimum Channels Operable column of Table 3.3.7.12-1, on pages 3/4 3-102 and 103. The note would allow-the monitor channels to be in an inoperable. status for up to one hour for the purpose of monitor surveillance.
This surveillance is required in Specifications 3/4.11.2.1 and 3/4.3.7.12. The basis for the staff's acceptance is that a one-hour allowance for. surveillance of the monitors is well within the Action limitations of Spectfications 3/4.11.2.1 and 3/4.3.7.12.
TS 3/4.3.7; pages 3/4 3-102, -104 This requested change would correct an error in the Actions required by Table 3.3.7.12-1 of the present Clinton TS. The' change would substitute Action.
121 for Action 126 on page 3-102, and delete Action 126 on page 3-104. The staff finds that 'the licensee's proposed change (Action 121 rather than Action 126) is acceptable. The bases for acceptance are that Action 121 meets the requirements in General Design Criterion 64 of Appendix A to 10 CFR 50 and Sec-tion 50.36a of 10 CFR 50, and the change'is consistent wit the model Radiologi-cal Technical Specifications, as well as with Technical Specifications for recently licensed BWRs.
s TS 3/4.3.9; page 3/4 3-113
-J The requested Technical Specification change would modify the trip setpoint and allowable value for trip function 1.b., " Containment Pressure - High," to 22.3 psia and 22.4 psia, respectively.
During the latter part of calendar year 1982, the NRC staff requested that several operating license applicants with General Electric nuclear steam supply systems document the methodology used to establish the protective system actua- <
tion instrumentation setpoints in plant Technical Specifications.
The licensee participated in the instrument setpoint methodology ~(ISM) program sponsored by the Boiling Water Reactors Owners Group (BWROG). The BWROG author-ized General Electric Co. to perform a validation calculation of the Technical Specification operating limits; Based on the analysis performed, calculations Clinton SSER 8 3 Appendix Q
(ISM) for the residual heat removal (RHR) containment spray initiation ir.stru-mentation indicate that the loop accuracy is significantly affected by high environmental temperatures. The inaccuracy of tht' loop requires lowering the setpoint of the affected instruments, IE12-N062 A-D and IE12-N662 A-D.
The staff has found that reducing the trip setpoint and allowable value for trip function 1.b. , " Containment Pressure - High," is a conservative change.
The acceptance of this change should not affect the safety of the plant.
On this basis, the staff has concluded that the proposed changes to the Clinton Power Station Technical Specifications are consistent with the SRP Section 7.2 acceptance criteria and, therefore, the proposed changes are acceptable.
TS 3/4.3.10; page 3/4 3-115 The licensee requested that Clinton Power Station Technical Specifica-tion 3/4.3.10 footnote be amended to read:
The STS may be periodically taken out of the fully automatic mode of operation for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing sur-veillance testing and preventive or corrective maintenance to satisfy technical specification requirements for those components the STS is designed to monitor.
The licensee's request to change, from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the amount of time that the self-test system (STS) may be taken out of the fully automatic mode of operation is for the purpose of providing greater flexibility to perform de-tailed " troubleshooting" and preventive or corrective maintenance and surveil-lance testing before returning the STS to automatic operation.
The Clinton Power Station's STS is an automatic testing system that injects
' short-duration pulses into the solid-state nuclear system protection system (NSPS) circuits and verifies proper response to various input combinations.
The STS is designed to maintain surveillance over NSPS cabinet circuitry essen-tial to reactor protection, emergency core cooling, and safe shutdown of the reactor. The STS is an overlay testing and surveillance subsystem that provides the capability to continuously and automatically perform testing of circuitry within the NSPS panels. The primary purpose of the STS is to improve the availability of the NSPS by optimizing the time to detect and determine the location of a failure in the system.
Each of the four Class 1E powered NSPS channels is tested once per 15 minutes.
Thus, all four NSPS divisions are tested in roughly 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
STS surveillance testing to satisfy Technical Specification requirements in-cludes logic system functional tests and response time tecting. In addition, the STS will be used to augment conventional testing methods to perform channel checks, channel functional tests, and channel calibrations. The STS will test.
the functional operability and response time of the NSPS logic from the input of the analog trip modules to the output of the actuated device load driver.
Upon replacement of a faulted printed circuit (PC) card, the STS is used as the primary piece of maintenance and test equipment that has the unique capability Clinton SSER 8 4 Appendix Q
l l
to test the replace / repaired PC card once installed in the NSPS panels. In summary, the STS is used for functions other than surveillance testing.
The staff has found that changing, from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the amount of time that the STS may be taken out of the fully automatic mode of operation should allow the time needed to perform detailed troubleshooting, preventive mainte-nance, corrective maintenance, and surveillance testing before returning the STS to the fully automatic mode of operation. The acceptance of this change should not affect the safety of the plant and is, therefore, acceptable.
TS 3/4.4.4; page 3/4 4-15 This requested change would correct a typographical error introduced in the published Clinton TS by removing the phrase "for up to 31 days" from the cited Surveillance Requirements. On publication of the Clinton TS, the phrase "for up to 31 days" was included erroneously in Surveillance Requirement 4.4.4.c.
This phrase has not been included in the TS of other recently licensed boiling-water reactors (BWRs) and is not in the STS for BWRs. Therefore, deletion of the typographical error is acceptable.
TS 3/4.6.-1; page 3/4 6-7 This requested change would allow the two blowers in the outboard MSIV leakage control system to operate in the design mode of the syst(.m's normal operation and, hence, permit them to be tested simultaneously. This change would also eliminate the need to replace one of the blowers with a blind flange, which currently is required to permit individual testing of each blower. Since the new test procedures are simplified and the new test reproduces the operating conditions more accurately, the staff finds that these changes are acceptable.
TS 3/4.6.2; page 3/4 6-15 The requested change would allow the 18-month test interval to be adjusted to coincide with the refueling outages. The BWR Mark III dryvell is not governed by Appendix J. This change would bring this TS into agreement with the provi-sions of Specificatiun 4.0.2. Granting flexibility for surveillance in accord-ance with plant outage is consistent with Specification 4.0.2 and has been accepted for other plants. This change is, therefore, acceptable.
TS 3/4.6.4; pages 3/4 6-32, -33 This proposed editorial change provides clarification and consistency between table notations on Table 3.3.2-1 earlier in the TS and Table 3.6.4-1. The change adds a note to isolation signals B and R to indicate that they are not independent signals. Specifically, the note indicates that signal R isolates appropriate values, only, following signal B. The note is added for clarifica-tion and this makes the description of signal R consistent with the one in Table 3.3.2-1 for containment and reactor vessel isolation control system (CRVICS) instrumentation. Thus the change is acceptable.
T_S 3/4.6.4; page 3/4 6-38 The requested change would modify TS operating conditions for four valves so that they are sealed closed for all operating conditions except refueling in Clinton SSER 8 5 Appendix Q
accordance with 10 CFR 50.36 and the comm'tu nt to TMI Action Item II.E.4.2 as described in the Clinton FSAR Appendix 0. Two of the valves are the containment-vent and purge air supply outboard and inbcard isolation valves. The other two are the containment vent system supply outboard and inboard bypass valves. This change is needed to satisfy the commitment to TMI Action Item II.E.4.2 to in-stall containment isolation signals for radiation. These valves do not have radiation isolation capability, so the licensee closed them off. The licensee made.the commitment in the Clinton FSAR, Appendix D, page D-49. The' change is, therefore, acceptable.
TS 3/4.6.4; page 3/4 6-44 The requested change would delete a valve from the containment isolation valve Table 3.6.4-1 because it does not serve the requirements of containment isolation identified in GDC 54, 55, 56, and 57. The valve involved in the requested change does not serve to satisfy requirements of containment isolation. The valve is used for containment isolation of the RHR/LPCI test line. Since the penetration with which this valve is associated is fully isolated by the outboard isolation valve, 1E12F042C, and a closed loop that returns to the containment which meets GDC 57, this valve, IE12-F351, is not necessary for containment isolation.
Hence the change is acceptable.
TS 3/4.6.4; pages 3/4 6-51, -52 The requested change would delete four valves from the containment isolation valve Table 3.6.4-1 because they are shutoff valves that do not satisfy the requirements for containment isolation. Since these valves are shutoff valves located downstream of the two main steam isolation valves (MSIVs) on each main steamline, and since they are not necessary for containment isolation, this change is acceptable.
TS 3/4.6.4; pages 3/4 6-56, -60 The requested change would delete two instrument air line relief valves from the containment and drywell isolation valve Table 3.6.4-1 because, through a design change to relocate the valving to the auxiliary building, they no longer serve as containment isolation valves. By a design change, these two valves have been replaced with blind flanges and relocated outside the containment isolation boundary. Since these valves no longer serve as containment isolation valves, the change is acceptable.
TS 3/4.6.4; page 3/4 6-59 The requested change would add a notation to one valve of Table 3.6.4-1, to be consistent with the Clinton FSAR. This notation was omitted in the Clinton TS through a typographical error. Note 6 of the Clinton FSAR Table 6.2-47 was inadvertently omitted from the Clinton TS Table 3.6.4-1 at the is' suance of the Clinton TS. Since addition of the note makes the Clinton TS consistent with ,
l the Clinton FSAR, the change is acceptable.
l TS 3/4.6.4; page 3/4 6-61 i
The requested change would modify the wording of table notation for Table 3.6.4-1 containment isolation valve notations to satisfy staff positions for l Clinton SSER 8 6 Appendix Q
implementation of TMI Action Item II.E.4.2 as described in the Clinton FSAR, Appendix D. In order to satisfy TMI Action Item II.E.4.2, the Clinton FSAR Appendix 0, page 49, was changed to specify that valves should be " sealed" closed by utilizing mechanical devices to seal or lock the valve closed, or to prevent power from being supplied to the valve operator. Since the proposed change corrects the wording of the Clinton TS table notation to correspond to that of the FSAR, it is acceptable.
TS Bases B 3/4.6.4; page B 3/4 6-7 The requested change would modify the Bases for TS 3/4.6.4 containment isolation valves to ensure that it is understood that the operability of the containment isolation valves is consistent with TMI requirements as described in the Clinton FSAR, Appendix D. The purpose of this clarification is to ensure that the full Bases are understood and that the requirements for containment integrity are amended by the TMI requirements of NUREG-0660 as clarified in NUREG-0737. The change is acceptable.
TS 3/4.7.6; page 3/4 7-17 The applicability statement presently requires the turbine bypass system to be operable from 0 to 100 percent of rated thermal power, while the Action statement allows the system to be inoperable below 25 percent of rated thermal power.
Since the objective of this Technical Specification is to provide the maximum plant operational flexibility without incurring unnecessary scrams, the appli-cability statement should be changed to require the turbine bypass system to be operable from 25 to 100 percent of rated thermal power. Therefore, the Clinton Technical Specifications are being amended to reflect this change. At 100 per-cent rated thermal power, a 45 percent total load shedding capability'(35% by the turbine bypass system and 10% by the reactor) is sufficient to contend with the expected operating transients. At a potential load of 25 percent rated thermal power level or less, at least 10 percent of the total load shedding capability that is provided by the reactor is sufficient to contend with expected operating transients. Permitting the turbine bypass system to be inoperable below 25 percent of rated thermal power will not alter the scram likelihood and is, therefore, acceptable.
TS 3/4.8.4; page 3/4 8-49 The licensee requested that Clinton Power Station TS Table 3.8.4.2-1 be modified to add valve IE51-C002E.
Valve 1E51-C002E contains a thermal overload bypass device and is safety re-lated. The thermal overload is continuously bypassed in both the open and closed directions. Therefore valve 1E51-C002E should be included in Table 3.8.4.2-1. The acceptance of this change does not impact the safety of the plant. '
TS 3/4.9.12, page 3/4 9-19, -20 In a submittal dated January 8, 1987, the licensee requested, in part, to revise l the Clinton Plant Technical Specification 3/4.9.12, " Inclined Fuel Transfer l System" (IFTS), in order to conform to the Technical Specifications which were specified for River Bend and Perry. The specific request deals with relief on Clinton SSER 8 7 Appendix Q l
l l
the specifications regarding the frequency of determining the operability of IFTS interlocks for the access doors, the hydraulic power unit blocking valve, and the access on control-transfer system keylock switches. The current require-ment is for these interlocks to be verified operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before the operation of the IFTS and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. The licensee h- proposed to reorganize this Technical Specification, so that the surveillance frey ancy would be at least once per seven days after the initial surveillance, in lieu of the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The River Bend Technical Specification 3/4.9.12 was evaluated against the pro-posed Standard Technical Specifications for a BWR/6. The purpose of this speci-fication is to provide adequate assurance that the possibility of accidental exposure of personnel to high radiation fields, as the result of the operation of the IFTS, is minimized. The deviation, therefore, was determined to be acceptable on the basis of adequate compensatory measures. .There are three compensatory measures which provided the basis for accepting the proposed River Bend Technical Specification. First, River Bend has warning lights outside of each access door. By Technical Specification, these warning lights are required to be operable, and their operability is to be verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after their initial verification. To remove one of the access doors (located behind a floor plug) requires the use of a crane. The need for a crane to gain access to this area would preclude accidental entry to the area by personnel, hence minimizing personnel exposures. The second compensatory measure is the verifi-cation, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, that the floor plug is installed and, after the floor plug has been removed, the verification, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before re-installation of the floor plug, of the operability of the access interlocks and palm switches for the fuel building and shield building annulus IFTS support rooms. The third
, measure is for the plant procedures to require that a guard be stationed outside each personnel access door to prevent any accidental entry into these potentially high radiation areas. On the basis of these three compensatory measures, the staff concluded that a 7-day surveillance frequency for the access interlocks, palm switches, blocking valve, and keylock switch was acceptable.
The licensee's request for a 7-day surveillance frequency is not based on any compensatory measure and the licensee has stated that there are no warning lights or palm switches at Clinton. There are only two " doors" which are covered by this Technical Specification. Although the licensee proposed no compensatory measures, the two doors have significant design features which, in themselves and in conjunction with administrative controls, provide the compensatory measures. The first door is a motorized vertical shield hatch and the second door is a floor plug. The lifting of the floor plug would require a crane, and access to the plug requires personnel to obtain and use a ladder. This area is fully visible to personnel operating the IFTS. The hatch is located in a high radiation area which is normally roped off. Furthermore, access to this area would require a work permit and a special confined space permit; thus access to the hatch is administratively controlled. The motorized hatch and the floor plug pose a sufficiently difficult barrier to' prevent per-sonnel from accidentally entering the protected high radiation areas, adequately minimizing potential personnel exposures.
In a submittal dated March 3, 1987, the licensee proposed a revision to the January 8, 1987 submittal to incorporate changes to identify that the " doors" include the removable shields and to commit to verify at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during operation of the IFTS that no personnel are in areas immediately Clinton SSER 8 8 Appendix Q
adjacent to the IFTS snd that all access doors to rooms through which-the IFTS penetrates are closed and locked.
On the basis that no personnel doors are covered by Technical Specification
~3/4.9.12 and with consideration being given both for the physical attributes of the " doors" and the administrative controls which would control their use, the staff concludes that proposed modifications to Technical Specification 3/4.9.12 as submitted on January 8 and amended on March 3, 1987, are acceptable.
TS 5.1.1 and 5.1.3; pages 5-2, -4 The requested change would make small changes in the documentation of unre-stricted areas recorded in these TS figures to make them consistent with the Clinton Offsite Dose Calculation Manual (00CM). The definition of unrestricted area, as it is used in the ODCM for calculation of doses to members of the public, is within the purview of the licensee. Since the proposed change follows the guidance of the model Radiological Effluent Technical Specifica-tions (RETS), as described in NUREG-0133, it is acceptable.
TS 6.1.2; page 6-1 The licensee proposed to designate the Vice President-Nuclear, rather than the Executive Vice President, as the appropriate level of management for issuance of the management directive describing the control room command function. This authority is consistent with FSAR Section 13.1.1.2, which states that the Vice President has the authority to change the organization as needed, and with the acceptance criteria of Section 13.1.1 of the Standard Review Plan (NUREG-0800),
and is, therefore, acceptable.
TS 6.2.1; page 6-3 The licensee proposed to change the title of one position shown on TS Fig-ure 6.2.1-1 and to delete certain positions from that figure. The proposed position title is consistent with the FSAR and the staff's Supplement 6 to the SER and corrects an error of omission and is, therefore, acceptable. The pro-posed position deletions are acceptable because they relate to plant construc-tion, which is essentially complete, or to the plant Startup Department, whose functions are now the responsibility of the plant operations staff. These changes are normal and are consistent with the staff's evaluation given in SER Supplement 6.
TS 6.2.2; page 6-4 The licensee proposed to change the title of the first level of management for l plant operations from Director-Plant Operation to Assistant Manager-Plant Opera- I tions. ThisproposalisconsistentwithSERSupplement6, Figure ~13.2," Plant Operating Organization for Clinton Power Station' and, therefore, the staff finds the change acceptable. The licensee also proposes to delete the asterisk associated with the position designation of Manager / Assistant Manager Clinton ;
Power Station that denotes that the person (s) in the position is required to be '
qualified as a Senior Reactor Operator (SRO). Since there is no requirement for this level of manager to be qualified as an SRO, the staff finds that a correction to an error and therefore, the proposed change is acceptable.
Clinton SSER 8 9 Appendix Q
TS 6.4.1; page 6-7 The licensee proposed to delete reference to the ANSI /ANS Code, Appendix A of 10 CFR Part 55, and the March 28, 1980 NRC letter to all licensees, leaving in "10 CFR Part 55," in anticipation of proposed changes to 10 CFR Part 55. The licensee's proposal is not acceptable at this time because the proposed changes to the rule have not yet been accomplished.
TS 6.5.1.2; page 6-7 The licensee proposed to change the membership of the Facility Review Group.
The Directors of Plant Maintenance and Plant Operations would be replaced by
, the Assistant Managers of Plant Maintenance and Plant Operations. This would be a position title change only. However, the organization chart, TS Fig-ure 6.2.2-1, still shows a Director of Plant Operations. By letter dated February 4, 1987, the licensee proposed to change the title Director Plant Operations to Assistant Manager Plant Operations on TS Figure 6.2.2-1. This
' change is acceptable because it makes this figure consistent with SER Supple-ment 6 in which the plant organization and position titles were found to be acceptable by the staff. Therefore, the proposed changes to the membership of the Facility Review Group are acceptable because the position levels of the membership have not been lowered and are consistent with other relevant Clinton Technical Specifications.
TS 6.7.1.d; page 6-14 The proposed change involves changing the time period for making safety limit violation reports available to the Nuclear Review and Audit Group (NRAG) and to the corporate Vice President. Reports of safety limit violations are required, in accordance with 10 CFR 50.73, to be reported to the Commission within 30 days of their occurrence. The licensee proposes to change the existing 14-day safety limit violation reporting reouirement to 30 days for reports made to the Vice President and to the NRAG. The change will not affect'the technical speci-fication requirement that the Vice President and the NRAG be notified of safety
{
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limit violations within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of their occurrence. The change is consistent with the reporting requirements of 10 CFR 50.73 for safety limit violations and it retains a reasonable time period for issuing a report of such violations to the appropriate corporate personnel; therefore, the staff finds this change
, acceptable.
TS 6.12; page 6-23 This requested change would correct a significant editorial omission in the published Clinton TS by adding the remainder of paragraph 6.12.2 and the two paragraphs of TS 6.12.3 to the Clinton TS. On publication of the Clinton TS, ten lines of TS 6.12.2 and all of TS 6.12.3 (plus footnotes) were inadvertently omitted. These added lines are consistent with the BWR Standard Technical Specifications, NUREG-0123 Revision 4 (Draft BWR/6) and have been included in the TS of other recently licensed BWRs. Therefore, the requested change is acceptable.
Clinton SSER 8 10 Appendix Q l
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