ML20197J967
| ML20197J967 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/10/1997 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duquesne Light Co, Ohio Edison Co, Pennsylvania Power Co |
| Shared Package | |
| ML20197J974 | List: |
| References | |
| DPR-66-A-209, NPF-73-A-087 NUDOCS 9801050156 | |
| Download: ML20197J967 (31) | |
Text
- _ _ _ _.
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UNITE) STATES p"
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. enmaa mang o
DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BIAVER VALLEY POWER STATION. UNIT NO. l MSDMENT TO FAr,ILITY OPERATING LICENSE Amendment No. 209 License No. DPR-66 1.
The Nuclear Regulatory Conaission (the Comission) has found that:
A.
The application for amendment by Duquesne Light Company, et al. (the licensee) dated March 14, 1997, as supplemented July 29, 1997, and August 13, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will oaerate in conformity with the application, the provisions of tie Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
9001050156 971210 PDR ADOCK 05000334 e
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-66 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. Y09, are hereby incorporated in the license.
~
The licensee shall operate the fa'cility in accordance with the Technical Specifications.
In addition, the' license is amended by changes to paragraph 2.C.(10) to the Facility Operating License No. DPR-66 as follows:
(10)- Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 209, are hereby incorporated into this license. Duqu:sne Light Company shall operate the facility in accordt a with the Additional Conditions.
3.
This license amendment is effective as of the date of its issuance, to be implemented within 60 days.
Implementation of this amendment shall include the relocation of these technical specification requirements to the appropriate documents, as described in the licensee's application dated March 14, 1997, as supplemented July 29 and August 13, 1997, and evaluated in the staff's safety evaluation attached to this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION o i F. Stolz, Dirq9 or ject Directoratf I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachments:
1.
Page I to Appendix C of License
- DPR-66 2.
Changes to the Technical Specifications Date of Issuance: December 10, 1997
- Peqe 1 of Appendix C is attached, for convenience, for the composite license to reflect this change.
b ATTACHMENT TO LICENSE AMENDMENT NO.209 FACILITY GPERATING LICENSE NO. DPR-66 QQCKET NO. 50-334 1.
Revise Appendix C of the License as follows:
Remove Paae Insert Pace 1
1 2.-
Replace the following pages of Appendix A Technical Specifications, with the enclosed pages as indicated.- The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove inigrl XIV XIV XV XV XVI XVI XVII XVII XVIII XVIII XIX XIX XX 3/4 7-30 3/4 7-30 6-5 6-5 6-6 6-6 6-7 6-7 6-8 6-9 6-10 6-11 6-12 6-13 6-21 6-21 6-22 6-24 6-24 6-25 6-25
l
(
I l
APPENDIX C ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR-66 Duquesne Light Company Ohio Edison Company, and Pennsylvania Power Company shall comply with the following conditions on the schedules noted below:
Amendment Additional Condition Implementation Number Date 202 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.
Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from April 14, to the appropriate documents, as described in 1997 the licensee's application dated September 9, 1996, and evaluated in the staff's safety evaluation attached to this amendment.
208 The licensee commits to perform the post weld The amendment heat treatment of sleeve welds and the shall be NRC-recom..anded inspections fot.epaired tubes implemented as described in the licensee's application within 60 days dated March 10, 1997, as supplemented July 28 from and September 17, 1997, and evaluated in the November 25, staff's safety enluation attached to this 1997 amendment.
209 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.
Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from to the appropriate documents, as described in December 10, the licensee's application dated March 14, 1997 1997, as supplemented July 29 and August 13, 1997, and evaluated in the staff's safety evaluation attached to this amendment.
1 Amendment No. G02, B98,209
' DPR-66 INDEX BASES SECTION PAGE 3/4.11.2-GASEOUS EFFLUENTS 3/4.11.2.5 Gas Storage Tanks.......................
B 3/4 11-1 3/4.11.2.6 Explosive Gas Mixture...................
B 3/4 11-1 DESIGN FEATURES SECTION PAGE 5.1 SITE LOCATION....................................
5-1 5.2 REACTOR CORE.....................................
5-1 5.3 FUEL STORAGE.....................................
5-1
-ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY...................................
6-1 6.2 ORGANIZATION 6.2.1 Onsite and Offsite Organizations........
6-1 6.2.2 Unit Staff..............................
6-2 6.3 FACILITY STAFF OUALIFICATIONS....................
6-5
,6. 4 TRAINING 6-5 6.5 DELETED 6.6 REPORTABLE EVENT ACTION..........................
5-5 l
6.7 SAFETY LIMIT VIOLATION...........................
6-5 l
BEAVER VALLEY - UNIT 1 XIV Amendment No. 209
DPR-66 INDEX
. ADMINISTRATIVE CONTROLS SECTION PAGE 6.8 PROCEDURES.......................................
6-6 l
6.9 REPORTING REQUIREMENTS...........................
6-16 6.9.1 Routine Reports.........................
6-16 6.9.1.1,2,3 Startup Reports.........................
6-16 6.9.1.4,5 Annual Reports..........................
6-17
- 6. 9.1. 6 -
Monthly Operating Report................
6-18 6.9.1.10 Annual Radiological Environmental Operating Report........................
6-18 6.9.1.11 Annual Radioactive Effluent Release Report..................................
6-19 6.9.1.12 Core Operating Limits Report............
6-19 6.9.2 SPECIAL REPORTS.........................
6-20 6.10 DELETED l
6.11 RADIATION PROTECTION PROGRAM....................
6-21 l
6.12 -HIGH RADIATION AREA.............................
6-23 6.13 PROCESS CONTROL PROGRAM (PCP) 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-24 6.16-MAJOR CHANGES TO RADIOACTIVE WASIE TREATMENT. SYSTEMS...............................
6-25 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM........
6-25 BEAVER VALLEY'- UNIT-1 XV Amendment No. 209
DPR-66 TABLE INDEX TABLE TITLE PAGE 2.2-1 Reactor Trip System Instrumentation Trip 2-6 Setpoints 3.1-1 Accident Analyses Requiring Reevaluation 3/4 1-19a in the event of an Inoperable Full or Part Length Rod 3.2-1 DNB Parameters 3/4 2-13 3.3-1 Reactor Trip System Instrumentation 3/4 3-2 3.3-2 Reactor Trip System Instrumentation 3/4 3-9 Response Times 4.3-1 Reactor Trip System Instrumentation 3/4 3-11 Surveillance Requirements 3.3-3 Engineered Safety Features Actuation System 3/4 3-15 Instrumentation 3.3-4 Engineered Safety Features Actuation System 3/4 3-22 Instrumentation Trip Setpoints 3.3-5 Engineered Safety Featurc Response Times 3/4 3-25 4.3-2 Engineered Safety Feature Actuation System 3/4 3-29 Instrumentation Surveillance Requirements 3.3-6 Radiation Monitoring Instrumentation 3/4 3-34 4.3-3 Radiation Monitoring Instrumentation 3/4 3-36 Surveillance Requirements 3.3-7 Seismic Monitoring Instrumentation 3/4 3-39 4.3-4 Seismic Monitoring Instrumentation 3/4 3-40 Surveillance Requirements 3.3-8 Meteorological Monitoring Instrumentation 3/4 3-42 4.3-5 Meteorological Monitoring Instrumentation 3/4 3-43 Surveillance Requirements 3.3-9 Remote Shutdown Panel Monitoring 3/4 3-45 Instrumentation 4.3-6 Remote Shutdown Monitoring Instrumentation 3/4 3-46 Surveillance Requirements BEAVER VALLEY - UNIT 1 XVI Amendment No. 209 i
_m.-___
4 DPR-66.
Table Index (cont TABLE TITLE PAGE 3.3-11 Accident Monitoring Instrumentation 3/4 3-51 4.3-7 Accident Monitoring Instrumentation 3/4 3-52 Surveillance Requirements 3.3-13 Explosive Gas Monitoring Instrumentation 3/4 3-55 4.3-13 Explosive Gas Monitoring Instrumentation 3/4 3-57 Surveillance Requirements 4.5-1 Minimum Number of Steam Generators to be 3/4 4-10e Inspected During Inservice Inspection 4.4-2 Steam Generator Tuoe Inspection 3/4 4-10f 4.4-3 Reactor Coolant System Pressure Isolation 3/4 4-14b Valves 3.4-1 Reactor Coolant System Chemistry Limits 3/4 4-16 4.4-10 Reactor Coolant System Chenistry Limits 3/4 4-17 Surveillance Requirements 4.4-12 Primary Coolant Specific Activity Sample 3/4 4-20 and Analysis Program 3.7-1 Maximum Allowable Power Range Neutron Flux 3/4 7-2 High Setpoint With Inoperable steamline Safety Valves During 3 Loop Operation
- 3.7-2 Maximum Allowable Power Range Neutron Flux 3/4 7-3 High Setpoint with Inoperable Steam Line Safety Valves During 2 Loop Operation 3.7-3 Steam Line Safety Valves Per Loop 3/4 7-4 4.7-1 Snubber Visual Inspection Interval 3/4 7-31 4.7-2 Secondary Coolant System Specific Activity 3/4 7-9 Sample and Analysis Program 3.8-1 Battery Surveillance Requirements 3/4 8-9a 3.9-1 Beaver Valley Fuel Assembly Mir. mum Burnup 3/4 9-15 d
vs. Initial U235 Enrichment For Storage in Region 2 Spent Fuel Racks BEAVER VALLEY - UNIT 1 XVII Amendment No. 209
DPR - Table Index=(cont.)-
TABLE TITLE PAGE B 3/4.4-1 Reactor Vessel' Toughness Data (unirradiated)
B 3/4 4-7 6.2-1 Minimum Shift Crew Composition 6-4 d
' BEAVER VALLEY - UNIT 1 XVIII Amendment No. 209
DPR-66 Fleure Index PIGURE TITLE PAGE 2.1-1 Reactor Core Safety Limit - Three Loop 2-2 operation 3.4-1 Dose Equivalent I-131 Primary Coolant 3/4 4-21 Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.35 pCi/ gram Dose Equivalent I-131 3.4-2 Beaver Valley Unit 1 Reactor Coolant 3/4 4-24 System Heatup Limitations Applicable for the First 16.0 EFPY
-3.4-3 Beaver Valley Unit 1 Reactor Coolant 3/4 4-25 System'Cooldown Limitations Applicable for the First 16.0 EFPY 3.6-1 Maximum Allowable Primary Containment Air 3/4 6-7 Pressure Versus River Water Temperature B 3/4.2-1 Typical Indicated Axial Flux Difference' B 3/4 2-3 Versus Thermal Power at BOL B 3/4.4-1 Fast Neutron Fluence (E>1 Mev) as a B 3/4 4-6a Function of Full Power Service Life B 3/4.4-2 Effect of Fluence, Copper Content, and B 3/4 4-6b Phosphorus Content on ARTNDT for Reacter Vessr.1 Steels per Reg. C_ide 1.99 B 3/4.4-3 isolated Loop Pressure-Temperature Limit B 3/4 4-10a Curve l
l l
BEAVER VALLEY - UNIT 1 XIX Amendment No. 209 l
- DPR-66 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) q.
Snubber service Life Monitoricq*
The service life of hydraulic anc) mechanical _ snubbers sha17. be monitored to ensure that the nervice life is not exceeded between surveillance inspections.
The maximum expected.sertvice life for various seals, springs, and other critical parts shall be determined and established based on engirieering infornation and may be extended or shortened based on monitored test rasults and failure h! story.
Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE.
The. parts replacements shall be documented and the documentation r. hall be retained in accordance with the applicable record retention provision of the quality assurance program der.cription referenced in the Updated Final Safety Analysis Report.
For purposes of establishing a baseline for the determination of service life monitoring, this program will ba implemented over 3 successive refueling perioda.
BEAVER VALLEY - UNIT 1-3/4 7-30 Amendment No. 209 L
DPR-66 ADMINISTRAT.TVE CONTROLS 6.3 - FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility and Radiation Protection staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparcble positions, except for the Health Physics Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the technical advisory engineering representative who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response analysis of the plant for transients and accidents.
6.4 TRAINUiG 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.
l 6.5 DELETED 6.6 REPORTABLE EVENT ACTIQ 1{
6.6.1 The rollowing actions shall be taken for REPORTABLE EVENTS:
a.
The Comnission shall be notified in accordance with 10 CFR 50.72 and/or a
report be submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.
Each REPORTABLE EVENT shall be reviewed by the OSC, and results of this review shall be submitted to the ORC.
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a.
The facility shall be placed in at least HOT STANDBY within one (1) hour, b.
The Safety Limit violation shall be reported to the Commission within one hour and to the General Manager, Nuclear Operations and to the ORC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the On-site Safety Committee (OSC).
This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
BEAVER VALLEY - UNIT 1 6-5 Amendment No.209
DPR-66 ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued).
d.
The Safety Limit Violation Report shall be submitted to the Commission, the ORC and the General
- Manager, Nuclear Operations within 30 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
a.
The applicable procedures recommended in Appendix "A"
of Regulatory Guide 1.33, Revision 2, February 1978.
h.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Not used.
e.
Not used.
f.
Fire Protection Program implementation, g.
PROCESS CONTROL PROGRAM implementation.
h.
OFFSITE DOSE CALCULATION MANUAL implementation.
6.8.2 Deleted 6.8.3 Deleted 6.8.4 A Post-Accident monitoring program shall be established, implemented, and maintained:
A program which will provide the capability to obtain and analyze reactor
- coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples following an accident.
The program shall include the following:
(i)
Training of personnel, (ii)
Procedures for sampling and analysis, and (iii)
Provisions for maintenance of sampling and analysis equipment.
BEAVER VALLEY - UNIT 1 6-6 Amendment No. 209
- lDPR-66 ADMINISTRATIVE CONTROLS PROCEDURES (Continued) 6.8.5-A program for monitoring ' of ' secondary water chemistry to inhibit steam generator tube degradation shall be implemented.- This program shall be described in the station chemistry manual and shall include:
a.
Identification of a sampling schedule for the critical parameters and control points for these parameters; b.
Identification of the procedures used to measure the values of the critical parameters; c.
Identification for process sampling points; d.
Procedures for the recording and management of data; e.
Procedures defining corrective actions for off control point chemistry conditions; and f.
A procedure identifying:
1) the authority responsibic for the interpretation of the data, and-2) the sequence and. timing of administrative events required to initiate corrective action, i
BEAVER VALLEY - UNIT 1 6-7 Amendment No.209 (Next Page is 6-14) l n
e
DPR-66 ADMINISTRATIVE CONTROLS SPECIAL REPORTS (Continued) f.
Miscellaneous reporting requirements specified in the Action Statements for Appendix C of the ODCM.
g.
, DELETED h.
Steam Generator Tube Inservice Inspection Results Report, Specification 4.4.5.5.
i.
Liquid Hold Up Tanks, Specification 3.11.1.4.
4 Gas Storage Tanks, Specification 3.11.2.5.
k.
Explosive Gas Monitoring Instrumentation, Specification 3.3.3.11.
6.10 DELETED l
'6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
BEAVER VALLEY - UNIT 1 6-21 Amendment No. 209 (Next Page is 6-23) l
DPR-66 ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP)
Changes to the PCP:
a.
Shall be documented and records of reviews performed shall be retained in accordance with the applicable record retention provision of the quality assurance program description referenced in the Updated Final Safety Analysis Report.
This documentation shall contain:
1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)
A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of
- Federal, State, or other spplicable regulations, b.
Shall become effective after review and acceptance by the OSC and the approval of the General Manager Nuclear Operations, predesignated alternate or a
predesignated Manager to whom the General Manager Nuclear Operations has assigned in writing the responsibility for review and approval of specific subjects.
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)
Changes to the ODCM:
a.
Shall be documented and records of reviews performed shall be retained in accordance with the applicable record retention provision of the quality assurance program description referenced in the Updated Final Safety Analysis Report.
This documentation shall contain:
1) sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)
A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent,
- dose, or setpoint calculations.
b.
Shal) become effective af ter review and acceptance by the OSC and the approval of the General Manager Nuclear Operations, predesignated alternate or a
predesignated Manager to whom the General Manager Nuclear Operations has BEAVER VALLEY - UNIT 1 6-24 Amendment No.209
. ~,
d DPR-66 ADMINISTRATIVE CONTROLS 3
~
OFFSITE DOSE CALCULATION MANUAL (ODCN) (Continued) assigned ~ in; writing the responsibility,for -review -and' approval of specific subjects._-
c.;
Shall be submitted to the - Commission in the form " of a complete, legible copy of - the entire ODCM as a part of.or concurrent with the Annual Radioactive Effluent Release Report for the period of - the report in which any change to the ODCM - was made.-
Each change shall be identified by marhings in.the margin of' the= affected,pages,. clearly
~ indicating _the area _of the page that was changed, and shall indicate the date (e.g.,
month / year) the change -was implemented.
1 11 Moved to the PROCESS CONTROL PROGRAM.
6.17 Containment Leakaae Rate Testina Proaram A program shall be established to implement the leakage rate testing _of the containment as required by 10 CFR 50.54 (o) and 10 exemptions (qFR
)
50, Appendix - J, Option. B, as modified by approved This program shall-be in accordance with the guidelines contained in Regulatory Guide 1.163,
" Performance-Based Containment Leak-Test L
-Program," dated September--1995.
The. peak calculated containment internal pressure for the design basis loss of coolant accident, P.,
is 40.0 psig.
The maximum allowable containment leakage rate, L.,
at Pa, shall be 0.10% of containment air weight per day.
Leakage Rate acceptance criteria are:
a.
Containment leakage rate _ acceptance criterion is 5
- 1. 0 L.
for the overall Type A, leakage test and < 0. 60 L.
for the Type B and Type C tests on a minimum pathway leakage _ rate (MNPLR) basis.
During the first unit startup following testing in - accordance with this program, the leakage ' rate acceptance criteria
< 0.60 L.
on a maximum pathway Lleakage rate (MXPLR)Igrebasis for Type B and Type C tests I
4
. and < 0.75 La for Type A tests.
~ (1) Exemptions to Appendix.J of 10 CFR 50 dated November.19, 1984, December 5, 1984, and July 26, 1995.
(2) For' penetrations which are isolated by use of a closed valve (s),
blind _ flange (s), or de-activated automatic valve (s), the MXPLR of the. isolated panetration is assumed to be the measured leakage
-through the~ isolation device (s).
BEAVER VALLEY - UNIT 1 6-25 Amendment No. 209
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UNITED STATES a
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NUCLEAn REGULATORY COMMISSION i
WASHINGTON, D.C. 30eeH001
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DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No. NPF-73 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Duquesne Light Company, et al. (the licensee) dated March 14, 1997, as supplemented July 29, 1997, and August 13, 1997, complies with the standards and requirements of the Atomic Energy Act of"1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendi.ent, and paragraph 2.C.(2) of Facility Operating License No. NPF-73 is hereby amended to read as follows:
-(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 87, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. DLC0 shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
In addition, the license is amended by changes to paragraph 2.C.(11) to the Facility Operating License No. NPF-73 as follows:
(11) Additional Conditions The_ Additional Conditions contained in Appendix D, as revised through Amendment No. 87, are hereby incorporated into this license.
Duquesne Light Company shall operate the facility in accordance with the Additional Conditions.
3.
This license amendment is effective as of its date of issuance, to be implemented within 60 days.
Implementation of this amendment shall include the relocation of these technical specification requirements to the appropriate documents, as described in the licensee's application dated March 14, 1997, as supplemented July 29 and August 13, 1997, and evaluated in the staff's safety evaluation attached to this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Jo F. Stolz, Direct P
ect Directorate I-2 Division of Reactor Pro,Jects - I/II Office of Nuclear Reactor Regulation Attachments:
1.
Page I to Appendix D of License
- No. NPF-73 2.
Changes to the Technical Specifications Date of Issuance: December 10, 1997
- Page 1 of Appendix D is attached, for convenience, for the composite license to reflect this change.
ATTACHMENT TO LICENSE AMEN 0 MENT NO. A7 FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 1.
Revise Appendix D of the License as follows:
Remove Paae Insert Pace 1
1 2.
Replace' the following pages of Appendix A Technical Specifications, with the enclosed pages as indicated. The revised pages'are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert XIV XIV XV XV XVI 3/4 7-27 3/4 7-27 6-2 6-2 6-3 6-6 6-6 6-7 6-7 6-8 6-9 6-10 6-11 6-12 6-13 6-21 6-21 6-22 6-22 6-23 6-24 6-24 6-25 6-25
4 APPENDIX D ADDITIONAL CONDITIONS OPERATING LICENSE NO. NPF-73 Duquesne Light Company, Ohio Edison Company, The Cleveland Electric Illuminating Company, and The Toledo Edison Company shall comply with the following conditions on the schedules noted below:
Amendment Additional Condition Implementation Number Date 83 The licensee is authorized tol291Xreloce6etain The amendment Technical Specification requirements to shall be licensee-controlled documents.
Implementation implemented of this amendneent shall include the relocation within 60 days of these technical specification requirements from April 14, to the appropriate documents, as described in 1997 the licensee's application dated September 9, 1996, and evaluated in the staff's safety evaluation attached to this amendment.
87 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.
Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from to the appropriate documents, as described in December 10' the licensee's application dated March 14, 1997 1997, as supplenented July 29 and August 13, 1997, and evaluated in the staff's safety evaluation attached to this amendment.
1 Amendment No. 83, 87
m.
.~.. ___.._. _.___..
_.m_
ypy_73 ~
DII2EK ADMINISTRATIVE CONTROLS-
-6.2.2 UNIT-STAFF................................-
6-1--
-l 6.3 FACILITY STAFF OUALTFICATIONS.......................-
6-6 6.4 TRAINING 6-6 i
p 6.5 DELETED l
l 6.6 REPORTABLE EVENT ACTION.............................
6-6 r
6.7 EAFETY LIMIT VIOLATION..............................
6-6 l
l 6.8 PROCEDURES..........................................
6-7 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS..............................
6-16 6.9.1.1 Startup Reports..............................
6-16
-6.9.1.4 Annual Reports...............................
6-17 6.9.1.6 Monthly Operating Report.....................
6-18 6.9.1.10 Annual Radiological Environmental Operating Report.......................................
6-19 6.9.1.'11 Annual Radioactive Effluent Release Report...
6-19 6.9.1.12 Core Operating Limits Report.................
6-19 6.9.2:
SPECIAL REPORTS..............................
6 !6.10' DELETED _
l 6'.11 RADIATION PROTECTION PROGRAM.....................
6-21 l
' BEAVER 1 VALLEY - UNIT 2 XIV
. Amendment No. 87
=. = -
_ _ _ _ =
a NPF-73~
i INDEK
'l l
ADMINISTRATIVE CONTROLS-6.12 HIGH RADIATION AREA................................
6-21 l
-6.13 - PROCESS CONTROL' PROGRAM (PCP)'.......................
6-24
~
.1,,14 OFFSITE DOSE CALCULATION MANUAL fODCM) 6-25 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Licruid, Gaseous and Solid) 6-25 1
6.17 CONTAINMENT Y.MKAGE RATE TESTING PROGRAM...........
6-25 4
1 T
4 P
BEAVER VALLEY.,- UNIT 2~
- XV-Amendment No.87-
,a
+
c
- .NPF-73 PIANT. SYSTEMS i
SURVEILIANCE REQUIREMENTS (Continued)-
e.
- Hydraull'c Snubbers Functional Test'Accentance Criteria The hydraulic _ snubber functional, test snall verify thati-
-1.
Activation -(restraining action)_ is achieved within' the specified-range of velocity or acceleration in both tension and compression.
1 l
2.
Snubber bleed, or release ~ rate, where required, is within the specified range in compression or _ tension.
For
. snubbers specifically required to not displace-under L
continuous load, the ability of the snubber to withstand load without displacement shall be verified.
j f.
Mechanical Snubbers Functional-Tact Accentance Criteria
{-
The mechanical snubber functional test shall verify that:
1.
The' force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.
2.
Activation (restraining action) is achieved within-the specified range of velocity or acceleration in both. tension and compression.
3.
Snubber release
- rate, where
- required, is within the specified range in compression or tension.
For snubbers spec.ifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.
g.
service Life Monitorina The service life of-hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between. surveillance inspections.-
The maximum expected service-life--for various seals, springs, and other critical partsLshall be determined and established based on-engineering information and may be extended or shortened based on monitored test results-and; f ailure._ history.
Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE.
The parts replacements shall be-documented and the documentation shallibe retained in : accordance with the applicable record ratantion provision of the quality assurance program description _ included.
in_the Updated Final' Safety Analysis: Report.
Service-life will Lbe defined'to_ commence'at plant startup subsequent' to. initial e
fuel load.
_ BEAVER VALLEY - UNIT.2-3/4 7-27 Amendment No. 87-
,.~-
NPF-73 ADMINISTRATIVE CONTROLS UNIT STAFF-(Continued) c.
At least two licensed operators shall be in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
d.
An individual qualified in radiation protection procedures shall be onsite when fuel is in the reactor.
e.
All CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
f.
Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; senior reactor operators, reactor operators, radiation control technicians, auxiliary operators, meter and control repairman, and all personnel actually performing work on safety related equipment.
The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is operating.
- However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary - basis, the following guidelines shall be followed:
a.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time, b.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period, all excluding shift turnover time.
c.
A break of at least eight hours should be allowed between work periods, including shift turnover time, d.
Except during extentad shutdown
- periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the General
- Manager, Nuclear Operations or predesignated alternate, or higher levels of management.
Authorized deviations at the working hour guidelines shall be documented and available for NRC review.
BEAVER VALLEY - UNIT 2 6-2 Amendment No.87 (Next page is 6-5) l 1
NPF-73 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility and Radiation Protection staff shall meet or exceed the minimum qualif f cations of ANSI N18.1-1971 for comparable positions, except for the Health Physics Manager who shall meet or exceed the qualificationis of Regulatory Guide 1.8, September 1975, and the technical advisory engineering representative who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response analysis of the plant for transients and accidents.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.
6.5 DELETED l
6.6 REPORTABLE EVENT ACTICN 6.6.1 The following actions shall be taken for REPORTABLE EVENTG:
a.
The Commission shall be notified in accordance with 10 CFR 50.72 and/or a
report be submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.
Each REPORTABLE EVENT shall be reviewed by the OSC, and the results of this review shall be submitted e the ORC.
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a.
The facility shall be placed in at least HOT STANDBY within one (1) hour.
b.
The Safety Limit violation shall be reported to the Commission within one hour.
The Safety Limit violation shall be reported to the General
- Manager, Nuclear Operations and to the ORC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, c.
A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the On-Site Safety Committee (OSC).
This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
BEAVER VALLEY - UNIT 2 6-6 Amendment No.87
)
HPF-7 3 l
ADMINISTRATIVE CONTROLS I
SAFEM LIMIT VIOLATION (Continued) d.
The Safety Limit Violation Report shall be submitted to the Commission, _the ORC and the General
- Manager, Nuclear Operations within 30 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, icplemented, rnd maintained covering the activities referenced belows a.
The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.
b.
Refueling operations.
c.
Surveillance and test activities of cafety related equipnant.
d.
Not used, e.
Not used.
f.
Fire Protection Program implementation.
f g.
PROCESS CONTROL PROGRAM implementation.
h.
OFFSITE DOSE CALCULATION MANUAL implementation.
6.8.2 Deleted 6.8.3 Deleted 6.8.4 A Post Accident monitoring program shall be established, implemented, and maintained.
The program wi31 provide the capability to obtain and analyzo reactor coolant, Ladioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples following an accident.
The program shall include the followings (i)
Training of personnel,.
(ii)
Procedures for sampling and analysis, and (iii)
Provisions for maintenance of sampling and analysis equipment.
BhAVER VALLEY - UNIT 2 6-7 Amendment No. 87 (Next page is 6-14) l
HPF-73 ADMINISTRATIVE CONTROLS SPECIAL REPORTS (Continued) b.
Inoperable seismic Monitoring Instrumentation, specification 3.3.3.3.
c.
Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
d.
Seismic event analysis, Specification 4.3.3.3.2.
a.
Sealed source leakage in excess of limits.
Specification 4.7.9.1.3.
f.
Miscellaneous reporting requirements specified in the ACTION Statements for Appendix C of the ODCH.
g.
DELETED h.
Steam generator tube inservice inspection, Specification 4.4.5.5.
1.
Inoperable acciosnt monitoring, Specification 3.3.3.0.
j.
Liquid Hold-Up Tanks, Specification 3.11.1.4.
k.
Gas Storage Tanks, Specification 3.11.2.5.
1.
Explosive Gas Monitoring Instrumentation, Specification 3.3.3.11.
6.10 DELETED l
6.11 RADIATION PROTECTION PROGEAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, main \\ained and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control oevice" or " alarm signal" required by paragraph 20.1601 of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring BEAVER VALLEY - UNIT 2 6-21 Amendment No. 87
- __. ~
- ~.
NPF-73 ADMINISTRATIVE CONTROLE HIGH RADIATION AREA (Continued) issuance of a Radiological Work Permit (1) or Radiological Access control Permit.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the followings a.
A radiation monitoring device which continuously indicates the radiation dose rate in the area, b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
c.
An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specifted by a facility health physics supervisor in the Radiological Work Permit or Radiological Access control Permit.
6.12.2 The requ.'
9ts of 6.12.1, above,. also apply to each high radiation area in wa tcn the intensity of radiation is greater than 1000 mres/hr.
In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or a facility health physics supervisor.
(1)
Health physics personnel, or personnel escorted by halth physics personnel in accordance with approved emergency procedures, shall be exempt from the RWP issuance requirement during the performance of their radiation protection
- duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.
BEAVER VALLEY - UNIT 2 6-22 Amendment No. 87 (Next page is 6-24) l
i l
NPF '=
ADMINISTRATIVE CONTRSLS 6.13 PROCESS CONTROL PROGRAM fPCP1 Changes to the PCP j
a.-
Shall be documented and records.of reviews performed shall be retained in accordance with the applicable record retention provision of. the quality assurance program description included' in -the Updated Final Safety Analysis Report.
This documentation shall containt 1) sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and l
2)
A determination that the change will maintain the f
overall conformance of the solidified waste product to existing requirements of
- Federal, State, or other applicable regulations.
b.
Shall-become effective af ter review and acceptance by the osc and the approval of the General Manager Nuclear operations, predesignated alternate 'or a
predesignated t
Manager to whom the General Manager Nuclear ooerations has assigned in writing the responsibility for review end approval of specific subjects.
t R
W f
i i
BEAVER VALLEY - UNIT 2-6-24 Amendment No.87--
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NPF-73 ADMINISTRATIVE CONTROLS 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM1 l
Changes to the ODCM a.
Shall be documented and records of reviews performed shall be retained in accordarace with the applicable record retention provision of the quality assurance program description included in the Updated Final Safety Analysis Report.
This documentation shall contains 1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)
A determination that the change will maintain the level of radioactive affluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of
- effluent, dose, or setpoint calculations.
b.
Shall become effective after review and acceptance by the OSC and the approval of the General Manager Nuclear Operations, predesignated alternate or a
predesignated Manager to whom the General Manager Nuclear Operations has assigned in writing the responsibility for review and approval of specific subjects.
c.
Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM vas made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.,
month / year) the change was implemented.
f,.11 Moved to the PROCESS CONTROL PROGRAM.
6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implemer.t the leakage rate testing of thu containment as required by 10 CFR 50.54 (o) and 10 exemptions (gR 50, Appendix J,
Option B,
as modified by ' approved This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,
" Performance-Based Containment Leak-Test Program," dated Septembkr 1995.
(2) Exemptions to Appendix J of 10 CFR 50, as stated in the operating license.
BEAVER VALLEY - UNIT 2 6-25 Amendment No. 87
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