ML20137C911
| ML20137C911 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/14/1997 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137C897 | List: |
| References | |
| NUDOCS 9703250169 | |
| Download: ML20137C911 (68) | |
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ATTACHMENT A-1
.i I
Beaver Valley Power Station, Unit No. 1 l
Proposed Technical Specification Change No. 236 The following is a list of the affected pages:
Affected Pages:
XV i
XVI XVII j
XVIII j
XIX i
XX XXI j
3/4 7-30 6-5 i
j 6-6 6-7 i
6-8 4
6-9
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6-10 l
6-11 j
6-12 2
6-13
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6-21 i
6-22 6-24 1
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9703250169 970314 PDR ADOCK 05000334 P
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INDEX DESIGN FEATURES SECTION PAGE 5.5 EMERGENCY CORE COOLING SYSTEMS 5-5 5.6 FUEL STORAGE 5.6.1 Criticality.
5-5 5.6.2 Drainage.
5-5 5.6.3 Capacity.
5-6 5.7 SEISMIC CLASSIFICATION 5-6
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5.6 METEOROLOGICAL TOWER LOCATION.
5-6 ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY 6-1 J
6.2 ORGANIZATION 4
6.2.1 Onsite and Offsite Organizations.
6-1 4
6.2.2 Unit Staff.
6-2 i
6.3 FACILITY STAFF OUALIFICATIONS.
6-5 6.4 TRAINING 6-5 D ELE.TED 6.5 OC"!E5? AND AUDIT 5.5.1 OMCITE CATCT'l CO!O'ITTCC '000)
-0.5.1.1 Function.
0 5 5.5.1.2 C: ;::itic.-
0 5 0.5.1.0 Alternet;;.
0-0 0.5.1.4 M;; ting Tr;quency.
00 BEAVER VALLEY - UNIT 1 XV Amendment No. +GB-Of A
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1 SPR-65 e
-p INDEX ADMINISTRATIVE CONTROLS y
l SECTION PAGE i
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6.6 REPORTABLE EVENT ACTION.
6-H 5
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l 6-H 6.7 SAFETY LIMIT VIOLATION L
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6-i+
6.8 PROCEDURES 6
6-16 6.9 REPORTING REOUIREMENTS 6-16 6.9.1 Routine Reports.
6-16 6.9.1.1,2,3 Startup Reports.
6-17 6'9.1.4,5 Annual Reports.
6.9.1.6 Monthly Operating Report.
6-18 BEAVER VALLEY - UNIT 1 XVI Amendment No. -Me.
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. 0"!!!!CTPl.TI"E CO::TROLC
- Move. +o Pac 3e XVI 6.9.1.10 Annual Radiological Environmental Operating i
Report..................................
6-18 6.9.1.11 Annual Radioactive Effluent Release j
Report..................................
6-19 6.9.1.12 Core Operating Limits Report............
6-19 4
l 6.9.2 SPECIAL REPORTS.........................
6-20
(~(DELETED 6.10 R ECORO D ET E!!T I O!'................................
0 21 6.11 RADIATION PROTECTION PROGRAM....................
6 -N-21 6.12 HIGH RADIATION AREA.............................
6-23 l
1 6.13 PROCESS CONTROL PROGRAM (PCP) 6-24 4
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-24 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS...............................
6-25 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM........
6-25 BEAVER VALLEY - UNIT 1 XVII Amendment No. 6 (Peoposed W'ocenc3)
J DPR-66 NOVC ko Pac 3e XVIIA '" '*
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 i
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 Feature Response Times 3/4 3-25 4.3-2 Engineered Safety Feature Actuation System 3/4 3-29 Instrumentation Surveillance Requirements i
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 Monitorir.g Instrumentation 3/4 3-39 4.3-4 Seismic Monitoring Instrumentation 3/4 3-40 Surveillance Requirements 1
3.3-8 Meteorological Monitoring Instrumentation 3/4 3-42 4.3-5 Meteorological Monitoring Instrumentation 3/4 3-43 Surveillance Requirements f
3.3-9 Remote Shutdown Panel Monitoring 3/4 3-45 Instrumentation l
4.3-6 Remote Shutdown Monitoring Instrumentation 3/4 3-46 Surveillance Requirements
_ _j (Proposed W'ordino) Amendment No.
XVIII BEAVER VALLEY - UNIT 1
i
-Mo\\/e +o Pac 3e XVIII-
fT Index (cont.)
l 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 i
4.4-1 Minimum Number of Steam Generators to be 3/4 4-10e Inspected During Inservice Inspection i
4.4-2 Steam Generator Tube 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 Chemistry Limits 3/4 4-17 i
Surveillance Requirements i
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 Minimum Burnup 3/4 9-15 vs. Initial U235 Enrichment For Storage in Region 2 Spent Fuel Racks BEAVER VALLEY - UNIT 1 XIX Amendment No. 103
{ ffO O$Lb YOU' bin
b e Index (cont.)
h0V6 fo Page. XIX=
TABLE TITLE PAGE B 3/4.4-1 Reactor Vessel Toughness Data (unirradiated)
B 3/4 4-7 l
6.2-1 Minimum Shift Crew Composition 6-4 i
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2 i
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I BEAVER VALLEY - UNIT 1 XX Amendment No. 444 (Proposed Wordincp
,D""-55 Ficure Index Move to Pqe XX FIGURE 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 i
Rated Thermal Power with the Primary Coolant Specific Activity > 1.0 Ci/ 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 i
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>l Mev) as a B 3/4 4-6a i
Function of Full Power Service Life B 3/4.4-2 Effect of Fluence, Copper Content, and B 3/4 4-6b f0E Phosphorus Content on ARTNDT Reactor Vessel Steels Per Reg. Guide 1.99 B 3/4.4-3 Isolated Loop Pressure-Temperature Limit B 3/4 4-10a Curve 1
5.1-1 Site Boundary for Gaseous and Liquid 5-lb Effluents for the Beaver Valley Power Station 5.1-3 Exclusion Area - Beaver Valley Power Station 5-id 5.1-4 Low Population Zone - Beaver Valley Power 5-le Station I
5.1-5 Gaseous Release Points - Beaver Valley Pcwer 5-2 Station 5.1-6 Liquid Release Points - Beaver Valley Power 5-3 Station urm,
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i PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l I
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Snubber Service Life Monitorina*
l The service life of hydraulic and mechanical snubbers shall be i
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 parts shall be i
determined and established based on engineering information and may be extended or shortened based on monitored test results and i
failure history.
Critical parts shall be replaced so that the I
maximum service life will not be exceeded during a period when l
the snubber is required to be OPERABLE.
The parts replacements shall be documented and the documentation shall be retained in i
accordance with 0;::ificetier 5.1^.2.
I I
bthe opp icable record refenhon provision l
Oh the UCLl0f ASSUYC\\nct froc)rorrt CbescriftOn referenced in the U clafed Rnal Safe +v P
Ano. lysis Reporty l
l For purposes of establishing a baseline for the determination of service life monitoring, this program will be implemented over 3 successive refueling periods.
1 BEAVER VALLEY - UNIT 1 3/4 7-30 Amendment No. -t+%
i (Proposed Wordiny
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.DPR-66 ADMINISTRATIVE CONTROLS 4
6.3 TACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility and Radiation Protection staff l
shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable 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 l
accidents.
J l
6.4 TRAINING i
l 1
6.4.1 A
retraining and replacement training program for the l
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 vf6ELETED3 6.5 7E"Irf
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ADMINISTRATIVE CONTROLS 4
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PR-66
/
ADMINISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The chairman and all members of the ORC shall be appointed by the Senior Vice President, Nuclear Power Division.
The membership shall consist of a
minimum of five individuals who collectively possess a
broad based level of experience. and competence enabling the committee to review and audit those 4
activities designated in 6.5.2.1 above and to recognize when it is necessary to obtain technical advice and counsel.
An individual i
may possess expertise in more than one specialty area.
The collective competence of the committee will be maintained as changes to the membership are made.
ALTERNATES l
6.5.2.3 All alternate members shall be appointed in writing by the l
ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC.
k 6-8 Amendment No Q AVER VALLEY - UNIT 1 (Proposed wordinop
De.bte. Erstire Pqe.
opR-66 ADMINISTRATIVE CONTROLS MEETING FREOUENCY a
6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.
I OUORUM l
consist of the Chairman or his least4membersincludingalternates.[
6.5.2.6 A
quorum of ORC shall designated alternate and at No more than a minority of the quorum shall have line responsibility for operation of the facility.
REVIEW i
6.5.2.7 The ORC shall review:
a.
The safety evaluations for lj changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify tha.t such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
t c.
Proposed testsorexperimentswhichinvolveanunreviewedsafety}
question as defined in Section 50.59, 10 CFR.
d.
Proposed changes in Technical Specifications or licenses.
e.
Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operation abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g.
All REPORTABLE EVENTS h.
All recognized indications of an unanticipated deficiency in aspect of design or operation of safety-related structures, i
some systems, or components.
i.
Reports and meeting minutes of the OSC.
j.
The results of the Radiological Environmental Monitoring Program annual report provided in accordance with Specification 6.9.1.10, prior to submittal.
BEAVER VALLEY - UNIT 1 6-9 Amendment No. 110
(.Peoposed Worchg)
~
4 i>R-66
~
ADMINISTRATIVE CONTROLS l
AUDITS j
6.5.2.8 Audits of facility activities shall be performed under the cognizance of the ORC.
These audits shall encompass:
a.
The conformance of faci.'.ity operation to provisions contained within the Technical Specifications and f
applicable license conditions.
b..
The performance, training and qualifications of the entire j
facility staff.
c.
The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or 3
methods of operation that affect nuclear safety.
d.
The performance of activities required by the Quality Assurance Program to meet the criteria of. Appendix "B",
)
e.
Not used.
l f.
Not used.
l g.
Any other area of facility operation considered appropriate l
by the ORC or Senior Vice President, Nuclear Power i
Division.
h.
The Facility Fire Protection Program and implementing j
procedures at least once per 24 months.
f 1.
An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified off-site licensee j
personnel or an outside fire protection firm.
I j.
An inspection and audit of the fire protection and loss i
prevention program shall be performed by a
qualified j
outside fire consultant at least once per 36 months.
i k.
The OFFSITE DOSE CALCULATION MANUAL and implementing procedures.
1.
The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive waste.
j BEAVER VALLEY - UNIT 1 6-10 Amendment No.191 (Proposed Wordirip
i l
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~ Move to Pa3e 6 -
6.6 REPORTABLE EVENT ACTION i
6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
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 10CFR Part 50, and b.
Each REPORTABLE EVENT shall be reviewed by the OSC, and J
results of this review shall be submitted to the ORC.
6.7 SAFETY LIMIT VIOLATION 1
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 />.
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-- Move. to Pac 3e 6-5
-7 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 I
circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, k_
and (3) corrective action taken to prevent recurrence.
Move to P%e 6-6 d.
The Safety LimitViolationReportshallbesubmittedtothe)
Commission, the ORC and the General
- Manager, Nuclear y 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.
b.
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.
&_:..tItitC] dure p =:
2nd ad=ini tr:tiv: p;licy of 0.0.1 abov: and
6.8.2 change
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- rtry ch nger te preendure Of 5.".1 ch
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c.
The change ir decurented, revier:d by th: 000 :nd ;ppr;.;d i
by th:
Cener:1 M:nt;::
Muel :: Op;;;ti:n:, pr:d::i;n:t:d l
- lterrete er a
preferignited M:nt;cr te "her the cener:1 M
- rr Murleir Operstiene har erri;ned in "riti ;
the r;;; ::ibility f::
- vi:r
- nd
- ppr:v:1 cf p::ifi cuhj::te, *?ithin l'. dry: ef inpl::::t ti:n.
i 6.8.4 A
Post-Accident monitoring program shall be established, l
implemented, and maintained:
j A program which. will provide the capability to obtain and analyze 4
reactor. coolant, radioactive iodines and particulates in plant gaseous affluents, and containment atmosphere samples following an i
accident.
The program shall include the following:
i j
(i)
Training of personnel, (ii)
Procedures for sampling and analysis, and samplingandanalysis)
(iii)
Provisions for maintenance of equipment.
Move to Page 6-7 6.8.5 A
program for monitoring of secondary water chemistry t 1
inhibit steam generator tube degradation shall be implemented.
This shall be described in the station chemistry manual and shall j
program i
include:
a.
Identification of a
sampling schedule for the critical parameters and control points for these parameters; i
j b.
Identification of the procedures used to measure the values i
of the critical parameters; identification for process sampling points; c.
d.
Procedures for the recording and management of data; j
e.
Procedures defining corrective actions for off control point chemistry conditions; and f.
A procedure identifying:
1) the authority responsible for the interpretation of 2
the data, and i
2) the sequence and timing of administrative events required to initiate corrective action.
i CEAVER 7ALLE'l L'UIT 1 C 13 A :nd;;nt N. 150 (Proposed Wordinep
- - - ~ _. -- -
{
1
?-
/ ADMINISTRATIVE CONTROLS 1
4 SPECIAL REPORTS (Continued) h.
Steam Generator Tube Inservice Inspection Results Report, i
Specification 4.4.5.5.
1 i.
Liquid Hold Up Tanks, Specific. ition 3.11.1. 4.
j.
Gas Storage Tanks, Specification 3.11.2.5.
i k.
Explosive Gas Monitoring Instrumentation, Specification 3.3.3.11.
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BEAVER VALLEY - UNIT 1 6-21 Amendment No. MO-(. Proposed k/o,-dincy
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PROGRAM.
- Move to Pa.Se 02 l -
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 30 CFR Part 20 and shall be
- approved, maintained and adhered Lc for all operations involving personnel radiation exposure.
m___
._..m-._...
t,..
.....nn
.,vm (Peoposed Woch)
- .DPu-6G
' ADMINISTRATIVE CONTROLS
.e 6.13 P'ROCESS CONTROL PROGRAM ( PCP ).
Changes to the PCP:
a.
Shall be documented and records of reviews performed shall be retained r:quir:d by Specific ti:r 5.10.2.r.
This documentation shall contain:
AL t et A 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 applicable 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 2:
required by Sp0cific:ti:r 5.10.2.
This documentation shall contain:
dt.1A5erf/s 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.
Shall become effective after review and acceptance by the OSC and the approval of the General Manager Nuclear predesignated Operations, predesignated alternate or a
Manager to whom ',he General Manager Nuclear Operations hat-.
assigned ir riting the responsibility for review and approval of specific subjects.
BEAVER VALLEY - UNIT 1 6-24 Amendment No. 444 (Peoposed W'ordiny
e
)
,=.
/ Attachment A-1 Insert A in accordance with the applicable record retention provision of the quality assurance program description referenced in the Updated Final Safety Analysis Report.
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i 4
i 4
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=
l ATTACHMENT A-2
}
Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 110 a
1 The following is a list of the affected pages:
Affected Pages:
XIV XV i
XVI l
3/4 7-27 i
6-2 l
6-3 6-6 6-7 6-8 i
6-9 l
6-10 6-11 i
6-12
{
6-13 1
6-21 6-22 l
6-23 J
6-24 l
6-25 i
i b.
I i
i I
I i
l NPF-73 l
INDEX l
DESIGN FEATURES SECTION PAGE 5.3 REACTOR CORE 5.3.1 Fuel Assemblies.
5-6 i
l 5.3.2 Control Rod Assemblies.
5-6 5.4 REACTOR COOLANT SYSTEM i
5.4.1 Design Pressure and Temperature.
5-6 5.4.2 Volume.
5-7 5.5 EMERGENCY CORE COOLING SYSTEMS 5-7 5.6 FUEL STORAGE 5.6.1 Criticality.
5-7 5.6.2 Drainage.
5-7 5.6.3 Capacity.
5-7 5.7 SEISMIC CLASSIFICATION 5-7 5.8 METEOROLOGICAL TOWER LOCATION.
5-7 ADMINISTRATIVE CONTROLS SECTION PAGE 1
6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 1
6.2.1 ONSITE AND OFFSITE ORGANIZATIONS.
6-1 6.2.2 UNIT STAFF.
6-1 0.2.2
!!?OE"E!!OE!!T S.'.FET'l E'J.'.L".".TIO!'
C".0""
'!SEC).
C --- 2 l
6.3 FACILITY STAFF OUALIFICATIONS.
6-6 BEAVER VALLEY - UNIT 2 XIV Amendment No. M-(ht"Of056d Nord r\\
mn. g:c:g
,~;c w
- . ; ;.y
.,oNPF-73 i
INDEX 1
ADMINISTRATIVE CONTROLS SECTION PAGE i
i 1
6.4 TRAINING 6-6 1
DELETED 6.5
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6.6 REPORTABLE EVENT ACTION.
6-M 6
BEAVER VALLEY - UNIT 2 XV Amendment No. (Proposed Worclim3)
' NPF-73 INDEX ADMINISTRATIVE CONTROLS PAGE SECTION f
g 6.7 SAFETY LIMIT VIOLATION..............................
6-M 6
6.8 PROCEDURES..........................................
6-M 7
6.9 REPORTING REOUIREMENTS 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 f
6-19 Report.......................................
6.9.1.11 Annual Radioactive Effluent Release Report...
6-19 l
6.9.1.12 Core Operating Limits Report.................
6-19 6.9.2 SPECIAL REPORTS..............................
6-20 DELETED)
C 21 6.10 0 2 2 " " n ""' C.. ! C !'...................................
6.11 RADIATION PROTECTION PROGRAM.......................
6-M El 6-M 6.12 HIGH RADIATION AREA................................
21 6-24 6.13 PROCESS CONTROL PROGRAM (PCP)......................
6-25 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM).............
6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT 6-25 SYSTEMS (Licuid. Gaseous and Solid) 6-25 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM...........
BEAVER VALLEY - UNIT 2 XVI Amendment No..gg.
(Propased W'ordinc)
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _... _ _ _. _ _ _ _ _ _ _ _. _ _ _ _. - _ _ _ _.. _. _. _ _ _ _ _ _ _. ~. _.
1 wn ?
j E,1 6
j NPF-73
' \\
- 1 i
l PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l i
e.
Hydraulic Snubbers Functional Test Accentance Criteria i
The hydraulic snubber functional test shall verify that:
1.
Activation (restraining action) is achieved within the i
specified range of velocity or acceleration in both tension and compression.
2.
Snubber bleed, or release rate, where required, is within l
the specified range in compression or tension. For snubbers specifically required to not displace under continuous load, i
the ability of the snubber to withstand load without
{
displacement shall be verified.
i l
f.
Mechanical Snubbers Functional Test Accentance Critaria The mechanical snubber functional test shall verify that:
i i
1.
The force that initiates free movement of the snubber rod in j
either tension or compression is less than the specified maximum drag force.
2.
Activation (restraining action) is achieved within the j
specified range of velocity or acceleration in both tension and compression.
j 3.
Snubber release
- rate, where
- required, is within the j
specified range in compression or tension.
For snubbers i
specifically required not to displace under continuous load, i
the ability of the snubber, to withstand load without displacement shall be verified.
j g.
service Life wonitorina i
The service life of hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between j
surveillance inspections.
The maximum expected service life for various seals,
- springs, and other critical parts shall be i
determined and established based on engineering information and may be extended or shortened based on monitored test results and l
fmilure history.
Critical parts shall be replaced so that the j
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 shall be retained in i
accordance with !; :ifi::tien 5.10 '.
Service life will be-j definedtocommenceatplantstartup{subsequenttoinitialfuel load
- the a licAble record re,tenhon provision of Ne gunkth R$5Urance prograrn de5Cripbn f
in We Updoked h Safa+y Analysis Report l
t i
BEAVER VALLEY - UNIT 2 3/4 7-27 Amendment No.~ 49-l (Proposeci Word Q
'll 2
(
APF-73 I
ADMINISTRATIVE CONTROLS l,
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 l
be onsite when fuel is in the reactor.
i 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 j
other concurrent responsibilities during tJ11s 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 extended 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 to the working hour guidelines shall be documented and available for NRC review.
l knee 4 Pese is h5)
BEAVER VALLEY - UNIT 2 6-2 Amendment No.-M
{ Pfcf Oseb Yorb in
y (iiPr-23 S M c M ive A go ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY EVALUATION GROUP (ISEG)
FUNCTION l
6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event
- Reports, and other sources of unit design and operating experience information, including units of similar design, which may indicate areas for improving unit safety.
The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to corporate management.
If not otherwise implemented, all recommendations shall then be made to the Senior Vice President, Nuclear Power Division.
COMPOSITION j
l J
6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers located on site.
Each shall have either:
(1)
A bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1
year' experience shall be in the nuclear of which field, or (2)
At least 5 years of nuclear experience and hold or have held a Senior Reactor Operator license, or (3)
At least 10 years of professional level experience in his
- field, at least 5 years of which experience shall be in the nuclear field.
A minimum of 50%
of these personnel shall have the qualifications specified in (1) above.
RESPONSIBILITIES beresponsibleformaintainingsurygfilance 6.2.3.3 The ISEG shall of unit activities to provide independent verificationt that these activities are performed correctly and that human errors are reduced as much as practical.
RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be
- prepared, maintained, and a
summary report shall be forwarded each calendar month to the Senior Vice President, Nuclear Power Division.
(1)
Not responsible for sign-off function.
l L BEAVER VALLEY - UNIT 2 6-3 (next page is 6-5)
Amendment No.74 (Peoposed WorclinQ
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nes
~'
.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 qualifications of ANSI N18.1-1971 for comparable 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 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.
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(Peoposed Worcling) l
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4 BEAVER VALLEY - UNIT 2 6-7 Amendment No. 7. :
(Peopose.d w'oedMe;)
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Deleke. En+ ire. Pqe -
ADMINISTRATIVE CONTROLS AUTHORITY 6.5.1.7 The OSC shall:
a.
Recommend to the General Manager, Nuclear Operations written approval or disapproval of items considered under 6.5.1.6.a through d above.
b.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6.a through e above constitutes an unreviewed safety question.
c.
Provide wrikten notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Senior Vice President, Nuclear Power Division and the Offsite Review Committee of disagreement between the OSC and the General
- Manager, Nuclear Operations; however, the General
- Manager, Nuclear Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
)
RECORDS 6.5.1.8 The OSC shall maintain written minutes of each meeting and copies shall be provided to the General Manager, Nuclear Operations and Chairman of the Offsite Review Committee.
6.5.2 OFFSITE REVIEW COMMITTEE (ORC)
FUNCTION 6.5.2.1 The ORC shall function to provide independent review and audit of designated activities in the areas of:
nuclear power plant operations a.
b.
nuclear engineering l
c.
chemistry and radiochemistry d.
metallurgy e.
instrumentation and control f.
radiological safety mechanical and electrica'l engineering g.
h.
quality assurance practices BEAVER VALLEY - UNIT 2 6-8 Amendment No.74 (Proposed W'ording')
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ADMINISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The chairman and all members of the ORC shall be appointed by the Senior Vice President, Nuclear Power Division.
The t
l membership shall consist of a
minimum of five individuals who l
l collectively possess a
broad based level of experience and competence enabling the committee to review and audit those activities designated in 6.5.2.1 above and to recognize when it is necessary to. obtain technical advice and counsel.
An individual may possess expertise in more than one specialty area.
The collec'tive competence of the committee will be maintained as changes to the membership are made.
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC.
MEETING FREOUENCY 6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.
OUORUM 6.5.2.6 A
quorum of ORC shall consist of the Chairman or his designated alternate and at least four members including alternates.
No more than a minority of the quorum shall have line responsibility for operation of the facility.
REVIEW 6.5.2.7 The ORC shall review:
a.
The safety evaluations for 1) changes to procedures, equipment, or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
l l
l BEAVER VALLEY - UNIT 2 6-9 Amendment No.74 hf0f0Seb NOrbin
3 bhPF-73 OL % -
ADMINISTRATIVE CONTROLS REVIEW (Continued) b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR. 2 c.
Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
d.
Proposed changes in Technical Specifications or licenses.
e.
Violations of applicable
- statutes, codes, regulations,
- orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g.
All REPORTABLE EVENTS.
h.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components, i.
Reports and meeting minutes of the OSC.
j.
The results of the Radiological Environmental Monitoring Program prior to submittal of the annual report provided in accordance with Specification 6.9.1.10.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the ORC.
These audits shall encompass:
a.
The
'conformance of facility _ operations to provisions contained within the Technical Specifications and applicable license conditions.
b.
The performance, training, and qualifications of the entire
{
facility staff.
c.
The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operation that affect nuclear safety.
d.
The performance of activities. required by the Quality f
Assurance Program to meet the criteria of Appendix "B",
10 I
CFR 50.
BEAVER VALLEY - UNIT 2 6-10 Amendment No. 74 J
$f0f 0$eb YOrdint
Dele +e Enyce Pac;e-n,m%
r.NPF-73 P
3 ADMINISTRATIVE CONTROLS AUDITS (Continued)
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e.
Not used.
S f.
Not used, l
1 g.
Any other area of facility operation considered appropriate j
by the ORC or the Senior Vice President, Nuclear Power I
Division.
h.
The Facility Fire Protection Program and implementing procedures at least once per 24 months.
i.
An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified off-site licensee personnel or an outside fire protection firm.
j.
An inspection and audit of the fire. protection and loss prevention program shall be performed by a
qualified outside fire consultant at least once per 36 months.
k.
The OFFSITE DOSE CALCULATION MANUAL and implementing procedures.
)
1.
The PROCESS CONTROL PROGRAM and implementing procedures for T
l processing and packaging of radioactive waste.
AUTHORITY 6.5.2.9 The ORC shall report to and advise the Senior Vice President, Nuclear Power Division on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Records of ORC activities shall be prepared, approved and distributed as indicated by the following:
4 a.
Minutes of each ORC meeting shall be prepared for and approved by the ORC Chairman or Vice Chairman within 14 days following each meeting.
l b.
Reports of reviews encompassed by Section 6.5.2.7 above, shall be documented in the ORC meeting minutes.
c.
Audit reports encompassed by Section 6.5.2.8 above, shall
~
be forwarded to the Senior Vice President, Nuclear Power Division and to the management positions responsible for the areas audited within 30 days after. completion of the I
audit.
BEAVER VALLEY - UNIT 2 6-11 Amendment No. y (Proposed W'ording)
APF-72
?.0MI:::0Trf.TI';; 00 'Tn Lc
~ Move to Pahe 6-6 6
REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
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 I
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.
The Safety Limit violation shall be reported to the General
- Manager, Nuclear Operations and to the ORC wi' thin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
j c.
A Safety Limit Violation Report shall be prepared.
The i
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.
f B4ove to. Pane 6 d.
The Safety Limit Violation teport 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.
b.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Not used.
e.
Not used.
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PROCESS CONTROL PROGRAM implementation.
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OFFSITE DOS 5 CALCULATION MANUAL implementation.
i Datated 6.8.2
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Post'-Accident monitoring program shall be established, j
implemented, and maintained. The program will provide the capability l
to obtain and analyze reactor
- coolant, radioactive iodines and l
particulates in plant gaseous effluents, and containment atmosphere 1
samples following an accident.
The program shall include the j
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(i)
Training of personnel, I
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Procedures for sampling and analysis, and (iii) Provisions for maintenance of sampling and analysis j
equipment.
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NPF-73 f
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SPECIAL REPORTS (Continued) 1 c.
Inoperable Meteorological Monitoring Instrumentation, i
specification 3.3.3.4.
J.
l d.
Seismic event analysis, Specification 4.3.3.3.2.
4 e.
Sealed source leakage in excess of limits, Specification l
i 4.7.9.1.3.
4 4
f.
Miscellaneous reporting requirements specified in the ACTION Statements for Appendix C of the ODCM.
l g.
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Steam generator tube inservice inspection, Specification 4.4.5.5.
1 l
i.
Inoperable accident monitoring, Specification 3.3.3.8.
I
-j.
Liquid Hold-Up Tanks, Specification 3.11.1.4.
i k.
Gas Storage Tanks, Specification 3.11.2.5.
p i
1.
Explosive Gas Monitoring Instrumentation, Specification 4
3.3.3.11.
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n;;:rd ef the 2:rfire li/:: Of 211 h dr uli: :nd
- hnni;;l j
- nubber including the date et hich th
- r.i :
lif:
2
- n
2nd 2:20 izted inct:112 tic.-
2nd m:intentn :
- rdr-i i
22: rd:
Of 2n lycer required by the Redi:1:;i;;l e_.
- u__:..
l n____2-4_.._
___,____2
-u.____
.2.
._u_
mee-,me OOC CALC"LAT!OF MA""Ai :nd th:.n00 ESC CO"TEOL PCOCRAM.
Move. to Page. 6-2.1--
j l
6.11 RADIATION PROTECTION PROGRAM i
l Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be j
- approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA l
6.12.1 In lieu of the " control device" or " alarm signal" required by j
paragraph 20.1601 of 10 CFR 20, each high radiation area in which the i
intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high andentrancetheretoshallgycontrolledbyrequirinal L radiation J
area WorX Permit or Radiological Access Radiological issuance.or a
l Control Permit.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied i
by one or more of the following:
I a.
A radiati.on monitoring device which continuously indicates the radiation dose rate in the area.
i 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 l
monitoring device may be made after the dose rate level in the area nos been established and personnel have been made' i
j knowledgeable of them.
1 (1)
Health physics personnel, or personnel escorted by health 4
i 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 1
procedures for entry into high radiation areas.
M ove +o Pqe. 6-2 2
l l
LEAVZK VALLC" - "JNIT 2 0 22 A ; ;..l ; ; 7.2.'!;.
"^
W r o p o $ t b Y o r Y if 1
h* PF-73.
/
- ADMINISTRATIVE CONTROLS d
cu nmnymmrnu anrn tr--*i-n us c.
An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
i This individual shall be responsible for providing positive i
control over the activities within the area and shall perform l
periodic radiation surveillance at the frequency specified by a
facility health physics supervisor in the Radiological Work Permit'or Radiological Access Control Permit.
6.12.2 The requirements of 6.12.1, above, also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.
In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained i
under the administrative control of the Shift Supervisor on duty L_and/or a facility health physics supervisor.
)
6.13 PROCESS CONTROL PROGRAM (PCP)
Changes to the PCP:
)
- be documented and records of reviews performed shall be a.
Shall' retained r quired by Sp::ific ti:n 5.10.2.n.
This
/(-((f\\5erk k documentation shall contain:
1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and j
2)
A determination that the change will maintain the overall conformance of the solidified waste product to
)
existing requirements of
- Federal, State, or other applicable 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.
BEAVER VALLEY - UNIT 2 6-24 Amendment No. 74-(Proposed Wordine)
a
- NPF-73 ADMINISTRATIVE CONTROLS 6.14.OFFSITE DOSE CALCULATION MANUAL (ODCM) 4 Changes to the ODCM:
a.
Shall be documented and records of reviews performed shall be retained :: required by S;::ifi::ti:n c.12. 2...
This documentation shall contain:
Q gg 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 j
1 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.
Shall become effective after review and acceptance by the OSC and the approval of the General Manager Nuclear Operations, predesignated alternate or a predesignated j
Manager to whom the General Manager Nuclear Operations has i
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 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.
4 6,16 Moved to the PROCESS CONTROL PROGRAM.
6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR 50, pendix J,
Option B,
as modified by approved exemptions (
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
(1) Exemptions to Appendix J of 10 CFR 50, as stated in the operating license.
BEAVER VALLEY - UNIT 2 6-25 Amendment No.49-(P<oposed Wardn%Q
.. ~ -.
i i
,' Attachment A-2
.-1 Insert A in accordance with the applicable record retention provision of the quality assurance program description included in the Updated Final i
Safety Analysis Report.
i i
a 5
I J
l i
i 1
)
l i
d l
i
i ATTACHMENT B i
Beaver Valley Power Station, Unit Nos. 1 and 2 Proposed Technical Specification Change Nos. 236 and 110 RELOCATION OF ADMINISTRATIVE CONTROLS RELATED TO QUALITY ASSURANCE A.
DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would relocate certain technical specification administrative controls to the quality assurance program description included or referenced in the Updated Final Safety Analysis Report (UFSAR).
Unit No. 2 UFSAR Section 17.2 presents the quality assurance program description for both Unit No. 1 and Unit No.
2.
Unit No. 1 UFSAR Section A.2.2, Operations Quality Assurance Program, was revised in 1988 to reference Section 17.2 of the Unit No. 2 FSAR.
Technical specifications to be relocated include the following:
1 TECHNICAL RELOCATED BVPS-2 UFSAR UNIT SPECIFICATION TO >
SECTION 4
INDEPENDENT SAFETY EVALUATION GROUP 2
6.2.3, Independent Safety 17.2.1 Evaluation Group Organization 4
REVIEW AND AUDIT i
1&2 6.5.1, Onsite Safety Committee 17.2.1 (OSC)
Organization
\\
1&2 6.5.2, Offsite Review Committee 17.2.1 (ORC)
Organization J
]
PROCEDURES l
1&2 6.8.2, Describing Procedure 17.2.5, Review \\ Approval Instructions, Procedures and Drawings 4
1&2 6.8.3, Describing Temporary 17.2.5 Procedure Changes RECORD RETENTION i
j 1&2 6.10.1, Identifying Records to 17.2.17 Quality j
be Retained for 5 Years Assurance Records 1&2 6.10.2, Identifying Records to 17.2.17 be Retained for the Duration of the Operating License B-1 l
_ - -..._.~--.
i i ;
- 4-
. ATTACHMENT B, continued j
Propo22d Technical Specification Change Nos. 236 and 110 i
Page 2 i
The following paragraphs present a description of changes to the UFSAR and Technical Specifications.
UFSAR CHANGES The Onsite Safety Committee.(OSC) and the Offsite Review i
Committee (ORC) descriptions found in Unit No. 1 UFSAR Section 12.4 will be updated as appropriate (in accordance with 10 CFR l
50.59) to reflect the revised quality assurance program description following NRC approval of this amendment request.
l Unit No.
l' UFSAR Section A.2.2, Operations Quality Assurance Program, will be updated to reference the Unit No. 2 "UFSAR" l
instead of:the "FSAR."
f The descriptions of the
- OSC, ORC, and Independent Safety Evaluation Group (ISEG) presented in Unit No. 2 UFSAR Sections 13.4.1, 13.4.2, and 13.4.4 will also be updated as appropriate i
(in accordance with 10 CFR 50.59) to reflect the revised' quality assurance program description following NRC approval of this amendment request.
The quality assurance program description included in Unit No. 2 UFSAR Section 17.2 would be revised to incorporate the text of technical specifications referenced in the table above.
Specific discussion of the proposed wording is presented under the heading TECHNICAL SPECIFICATION CHANGES below. '
The text of Specifications 6.5.1, Onsite Safety Committee, and 6.5.2, Offsite Review Committee, would replace the description of the OSC and ORC found in Unit No. 2-UFSAR Section 17.2.
The text to be deleted from the UFSAR is consistent with provisions in the technical specifications noted above; therefore, there is no reduction in commitments.
A spelling correction to the word " quantitative" in the second paragraph of Unit No. 2 UFSAR Section 17.2.5 would also be made.
This editorial change.would not reduce commitments.
TECHNICAL SPECIFICATION CHANGES In addition to the relocation of technical specifications the following editorial changes will be made:
1.
References to Specification 6.10 record retention requirements (found in Specifications 4.7.12.g, 6.13.a, and 6.14.a) would be revised to reference the quality assurance program description included (Unit 2) or referenced (Unit 1) in the UFSAR.
2.
The technical specification index would be updated to reflect the amended text and to reference appropriate page numbers.
)
B-2 l
l
- " ATTACHMENT'B, ccntinutd Proposed Technical Specification Change Nos. 236 and 110 Page 3 i
3.
Grammar and punctuation changes would be made that do not affect-the intent of provisions relocated to the UFSAR.
4.
Technical specification numbers would be omitted or replaced in the text' relocated to the UFSAR by appropriate references and UFSAR paragraph numbering conventions.
5.
The title "Onsite Safety Committee Supervisor" in Specification 6.5.1.2 would be revised to read "Onsite Safety Committee Coordinator."
The technical specification provisions listed in the table above will be relocated intact to the UFSAR quality assurance program description in Section 17 with the editorial changes noted above.
Specific changes and exceptions are presented below.
INDEPENDENT SAFETY EVALUATION GROUP Specification 6.2.3, Independent Safety Evaluation Group, would be moved intact to Unit No. 2 UFSAR Section 17.2.1 except that:
1.
The words "50%" in Specification 6.2.3.2 would be revised to read "50 percent."
In addition, UFSAR text numbering and formatting conventions would be used.
2.
The note associated with Specification 6.2.3.3 would be incorporated as a parenthetical statement.
3.
The first sentence 'of Specification 6.2.3.2 would be omitted, and the second sentence would be ' changed from:
"Each shall have either:
" to read:
"ISEG personnel-shall have-either:
The first sentence specifies the minimum' number of dedicated, full-time ISEG engineers
' located on site.
REVIEWS AND AUDITS Specification 6.5.1, onsite safety committee, would be moved intact to Unit No. 2 UFSAR Section 17.2.1 with the following exceptions.
1.
Specification 6.5.1 would be revised to incorporate UFSAR text numbering and formatting conventions.
{
)
2.
The title "Onsite Safety Committee Supervisor" located in the first sentence of Specification 6.5.1.2 would be
'l revised to read "Onsite Safety Committee Coordinator."
Reference to Specification 6.5.1.1 would be replaced by the word "above."
3.
Item "1)"
in Specification 6.5.1. 6.a would be revised by changing " required by Specification 6.8" to " required by Technical Specification 6.8."
B-3
_m
_-__._.y
- ATTACHMENT B continued Proposed Technical Specification Change Nos. 236 and 110 Page 4 j
)
l 4.
The words "of the type described in 10' CFR 50.73" would be i
added to the end of Specification 6.5.1.6.f.
This change reflects the definition of reportable events provided in the technical specifications.
4 J
5.
Specification 6.5.1.7.a.
would be revised by changing j
"6.5.1.6.a through d" to "OSC responsibility 1 through 4."
6.
Specification 6.5.1.7.b would be revised by changing "6.5.1.6.a through e" to "OSC responsibility 1 through 5."
^
7.
Specification 6.5.1.7.c would be revised by changing the words
" pursuant to 6.1.1 above" to read
" pursuant to Technical Specification 6.1.1."
Specification 6.5.2, Offsite Revlev Committee, would be moved intact to Unit No.
2 UFSAR Section 17.2.1 with the following exceptions.
j i
i 1.
Specification 6.5.2 would be revised to incorporate UFSAR
{
text numbering and formatting conventions.
2.-
The second sentence of-Specification 6.5.2.2 would be 4
revised by deleting the words "in 6.5.2.1."
- f i
3.
The words "of the type described in 10 CFR 50.73" would be added to the end of Specification 6. 5.2.7.g.
.This change reflects the definition of reportable events provided in the technical specifications.
1 4.
Unit No. 1 Technical Specification 6.5.2.7.j wording would i
be relocated to the UFSAR.
Unit No.
2 Technical Specification 6.5.2.7.j is worded differently but has the same intent.
Reference to " Specification 6.9.1.10" would be revised to read " Technical Specification 6.9.1.10."
i 5.
Items identified as "Not used" in specifications 6.5.2.8.e i
and f would be omitted.
I 6.
Specification 6.5.2.9 would be revised by changing the words "in Section 6.5.2.7 and 6.5.2.8" to read "in the j
paragraphs above that list ORC review and audit topics."
7.
Specification 6.5.2.10.b -would be revised by changing the words "Section 6.5.2.7 above" to read "the ORC review topics listed above."
8.
Specification 6.5.2.10.c would be revised by changing the words "Section 6.5.2.8 above" to read "the ORC audit topics listed above."
B-4
. hTTACHMENT B, continued Proposed Technical Specification Change Nos. 236 and 110 l
Page 5 PROCEDURE REVIEW PROCESS Specification 6.8.2, Procedure Review / Approval, would be moved intact to Unit No. 2 UFSAR Section 17.2.5 except that:
1.
This specification would be revised to incorporate UFSAR text numbering and formatting conventions.
2.
The words "of 6.8.1 above" in the first and second sentence would be changed to read "of Technical Specification 6.8.1."
. Specification 6.8.3, Temporary Procedure Changes, would be moved intact to Unit No. 2 UFSAR Section 17.2.5 except that:
1.
This specification would be revised to incorporate UFSAR text numbering and formatting conventions.
2.
The words "of 6.8.1 above" in the first sentence would be changed to read "of Technical Specification 6.B.1."
RECORDS AND RECORD RETENTION Specification 6.10.1, Record Retention - 5 Years, would be moved intact to Unit No. 2 UFSAR Section 17.2.17 with the following exceptions.
1.
This specification would be revised to incorporate UFSAR text numbering and formatting conventions.
2.
The words
" covering time interval" in Specification 6.10.1.a would be revised to read
" covering the time interval."
3.
The words "of the type described in 10 CFR 50.73" would be added to the end of Specification 6.10.1.c.
This change reflects the definition of reportable events provided in the technical specifications.
4.
The words "these Technical Specifications" in Specification 6.10.1.d would be revised to read "the Technical Specifications."
Specification 6.10.2, Record Retention - Life of License, would be mcVed intact to Unit No.
2 UFSAR Section 17.2.17 with the following exceptions.
1.
Tnis specification would be revised to incorporate UFSAR text numbering and formatting conventions.
2.
The words "these Technical Specifications" in Specification 6.10.2.h would be revised to read "the Technical Specifications."
B-5
- . ATTACHMENT B, continued Proposed Technical Specification Change Nos. 236 and 110 Page 6 B.
BACKGROUND Technical specification administrative controls are the 1
provisions relating to organization and management, procedures, i
record keeping, review and audit, and reporting.
This amendment request involves relocation of certain of these administrative controls related to quality assurance requirements and editorial changes.
The administrative controls will be relocated to the 1
NRC approved quality assurance program description for Beaver l
Valley Power Station.
The quality assurance program description is included in Unit No.
2 UFSAR Section 17.2.
Unit 1 UFSAR Section A.2.2 will be clarified to reference Unit 2 UFSAR Section 17.2 for the quality assurance program description.
NRC Administrative Letter 95-06 discusses relocation of certain technical specifications that are located in the " administrative controls section" and are related to quality assurance i
requirements.
This submittal was developed based on the discussion in the administrative letter.
C.
JUSTIFICATION The Technical Specifications administrative controls to be relocated are related to quality assurance.
There are no limiting conditions for operation or surveillance requirements associated with these administrative controls.
When relocated, these administrative controls will be included in the NRC approved quality assurance program description and will be subject to the established quality assurance change control process in 10 CFR 50.54(a).
Location of these administrative controls in the UFSAR is appropriate since they describe implementation of licensee commitments to industry quality assurance standards contained in the UFSAR.
In addition, Technical Specifications will be revised to remove excessive detail, thereby gaining flexibility in making changes to these administrative controls without the need for a license amendment.
Relocation of technical specification requirements in cases where adequate controls are provided by other methods can reduce the resources spent by licensees and the U.
S.
Nuclear Regulatory Commission staff ~ in preparing and reviewing license amendment requests.
NUREG-0737,Section I.B.1.2 instructs all applicants for operating licenses to implement an Independent Safety Evaluation Group (ISEG).
Thus Beaver Valley Unit 2,
with the issuance of its operating license in 1987, established an ISEG.
As per NUREG-0737, the principal function of the ISEG is to examine plant operating characteristics, NRC issuances, industry advisories and other appropriate sources of plant design and operating experience information that may indicate areas for B-6
T
. ATTACHMENT B, c ntinu:d i
Proposed Technical Specification Change Nos. 236 and 110 Page 7 improving plant safety.
The ISEG is also to perform independent i
review of plant activities including maintenance, modifications, operational problems. and operational analysis.
Where useful improvements can be achieved, it is expected that this group will i
develop and present detailed recommendations to corporate j
management.
NUREG-0737 specifies the staffing levels for an ISEG.
The I
. staffing requirement for five dedicated engineers is burdensome l
l to a utility because it restricts the capability to utilize i
resources to their maximum advantage and does not result in an increase in the protection afforded to the health and safety of the public.
Therefore, the staffing requirement will be omitted l
from the relocated provisions.
j This proposed change would provide flexibility to accomplish the ISEG review function.
The purpose, scope, and thoroughness of j
independent technical reviews will not be affected.
A sufficient number of personnel will continue to be utilized to accomplish j
the ISEG function.
+
4 D.
SAFETY ANALYSIS
[
This proposed relocation of technical specification i
administrative controls to the quality assurance program description is consistent with the discussion provided in NRC Administrative Letter 95-06.
The proposed amendment does not l
. reduce commitments or change implementation of the affected i
i administrative controls except as provided in the paragraph below.
The method of controlling revisions to the administrative j
controls will be changed.
Instead of requiring a
license amendment, changes to the administrative controls would be made in accordance with the established quality assurance program change process in 10 CFR 50.54(a).
Since the affected administrative controls will continue to be subject to adequate change
- controls, continued safe operation of the plant is assured.
The provision of Technical Specification 6.2.3.2 which states i
that: "The ISEG shall be composed of at least five, dedicated, i
full-time engineers located on site," would be omitted from the i
provisions relocated to the quality assurance program description.
Since no system, component or operational procedure j
changes are involved, and the ISEG function will continue to be implemented, the change can have no effect on safe operation of
}
the plant.
i 1
B-7
1
- ATTACHMENT B, c:ntinurd Proposed Technical Specification Change Nos. 236 and 110 Page 8 E.
NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the 4
proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a
facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed i
amendment would not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or i
(2)
Create the possibility of a new or different kind of accident from any accident previously evaluated; or t
(3)
Involve a significant reduction in a margin of safety.
The following evaluation is provided for the no significant i
hazards consideration standards.
j i
1.
Does the change involve a
significant increase in the probability or consequences of an accident previously j
evaluated?
This proposed change would relocate technical specification administrative controls to the quality assurance program description.
Adequate controls are provided by the established quality assurance program change process in 10 CFR 50.54(a).
The provision of Technical Specification 6.2.3.2 which states that:
"The ISEG shall be composed of at least
- five, dedicated, full-time engineers located on site," would be omitted from the provisions relocated to the quality assurance program description.
Since no system, component or operational procedure changes are involved, and the ISEG i
function will continue to be implemented, the change can have no effect on safe operation of the plant.
The likelihood that an accident will occur is not increased by this proposed technical specification change which involves administrative controls.
No systems, equipment, or components are affected by the proposed change.
Thus, the consequences of a malfunction of equipment important to safety previously evaluated in the Updated Final Safety Analysis Report (UFSAR) are not increased by this change.
B-8
. ATTACHMENT B, continu!d i
Proposed Technical Specification Change Nos. 236 and 110 Page 9 Relocation of technical specification provisions and related changes do not affect possible initiating events for accidents previously evaluated or any system functional requirement.
The proposed changes have no impact on accident initiators or plant equipment, and do not affect the probabilities or consequences of an accident.
Therefore, the proposed changes will not involve a
significant increase in the probability or consequences of a i
previously evaluated accident.
2.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
1 The proposed relocation of technical specification provisions i
to the quality assurance program description and related changes do not involve changes to the physical plant or operations.
Since the proposed changes to administrative controls do not affect equipment or its operation, they cannot contribute to accident initiation and cannot produce a new accident scenario or a new type of equipment malfunction.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Does the change involve a significant reduction in a margin of safety?
The proposed changes are administrative in nature and do not directly affect plant equipment or operation.
Safety limits and limiting safety system settings are not affected by this proposed change.
The proposed changes do not affect the UFSAR design
- bases, accident assumptions, or technical specification bases.
In addition, the proposed changes do not affect release limits, monitoring equipment or practices.
Therefore, the proposed changes would not involve a
significant reduction in the margin of safety.
F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request satisfy the no significant hazards consideration standards of 10 CFR 50.92(c)
- and, accordingly, a
no significant hazards consideration finding is justified.
3 I
B-9
i
. ATTACHMENT B, continuId Proposed Technical Specification Change Nos. 236 and 110 Page 10 j
G.
- changes, including quality assurance program description changes, are provided in Attachments C-1 and C-2.
I 1
B-10
I 1
l,*.
ATTACHMENT C-1 i
l Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 236 i
i
)
Applicable UFSAR Changes i
Affected Page:
A.2-9 i
l i
i i
1 i
i 4
i l
l
BVPS-1-UPDATED FSAR Rev. 6 (1/88) i A
file system has been established to assure retrievability of such records.
In the case where records i-locations for storage are are_not stored at the site, reviewed and/or approved by the Duquesne
- i Light Company.
Vendors retaining records for certain periods of time i.
are_ instructed that at the conclusion of his (the vendor)such records must be submitted to the applicant storage responsibility.
i A.2.1.18 Audits s
i The Duquesne Light Company retains the ultimate responsibility for Quality Assurance for -BVPS-1.
This
~ responsibility is exercised mainly through an audit program.
These audits are performed against i
the S&W and Westinghouse commitments.
These include quality i
assurance / quality control procedures, specifications,
- drawings,
}
instructions, and similar information.
Internal audits of the i
Duquesne Light Company are also documented and maintained in the j
file.
i As agents for the Duquesne Light Company, S&W also performs audits at i
the vendor shops through their Procurement Quality Control Division and audits at the site through the Field Quality Control Division.
Both
' divisions are audited periodically Eby the Duquesne Light j
j.
Company.
Audit reports are utilized to assess the adequacy of the Duquesne Light Company Quality Assurance Program.
A record of all audits is maintained and periodically reviewed to assure necessary follow-up action.
Audits are reported at each project Management Committee
' Meeting to assure that the necessary level of management is involved and has an awareness of the audit program.
All audits are performed utilizing a
preplanned checklist.
The checklist will include address to the specific criteria of Appendix B of
- 10CFR50, results of previous
- audits, inspection
- reports, nonconformance
- reports, adherence to specifications, and other items as identified in the respective organization quality assurance manuals.
A.2.2 Operations Quality Assurance Program Deleted by Revision
,6,
January 1988.
Refer to Section 17.2 of the BVPS-2{SAR.
j U
A.2-9
4 i
ATTACHMENT C-2 4
4 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 110 l
4 l
Applicable UFSAR Changes l
Affected Pages:
l l
17.2-3 j
17.2-4 l
N:':
i i
l 1
i
1.
BVPS-2 UFSAR R v. 8 e
The General Manager Nuclear Human Resources Unit maintains detailed job d.escriptions of principal Nuclear Power Division personnel.
Customer ooerations k
. Customer Operations is the Vice I
Reporting to the Sr. Vice President, i
President Customer Services.
The Vice President Customer Services is responsible for the customer services Group.
I The customer Services Group provides support services as requested li by the Nuclear Power Division.
These services are performed
[
primarily by its departments in its General Services Unit, its l.
Western Power Delivery Division, and its Customer Operations Unit.
L "he General Services Unit provides support to the Nuclear Power Division for meter and relay calibration.
The Western Powstr Delivery Division provides support services to the Nuclear Power Division that includes 1.
Performing field testing, maintenance, and/or calibrations on specified types of relays in accordance with written appreved procedures, i
j-2.
Protective relay application and coordination down to and including 480 VAC electrical systems.
3.
Engineering services as requested in the areas of electrical, mechanical, civil, and telecommunications.
+
l 4.
Providing cupport in the station for specified electrical maintenance and telecommunications systems.
The Customer operations Unit provides engineering, technical, and field operations support to the Nuclear Power Division.
1 j
]
All nuclear quality-related activities of customer operations are documented by policies, directives, procedures, instructions, etc.,
I of a type appropriate to the activity.
Qasite Safety h ittee fosci s
The onsite Safety committee, a.s required by Technical special committee aomposed of experienced and 4' specifications, is a highly skilled Nuclear Power Division pe:sonnel.
The Onsite Safety Connoittee meets monthly (and when called by the Chairman) and j
advises the General Manager Nuclear Operations Unit on all matters j
related to nuclear safety.
i a
-+Imert A i
(
e i
17.2-3 I
m
-_e
Y IM2ERT A ONSITE SAFETY COMMITTEE The. OSC shall function to advise the General Manager, Nuclear operations on all matters related to. nuclear safety and shall provide j
review capability in the areas of:
i 1.
Nuclear power plant operations.
2.
Radiological. safety..
3.
Maintenance.
4.
Nuclear engineering.
5.
Nuclear power plant testing.
j.
6.
Technical advisory engineering.
l 7.
Chemistry.
8.
Quality control.
9.
Instrumentation and control.
[
The Onsite Safety Committee Coordinator is the OSC Chairman and shall i
appoint all members of the OSC.
The membership shall consist of a
]
minimum of one individual from each of the areas designated above.
j OSC members and alternates shall meet or exceed the minimum i
qualifications of ANSI N18.1-1971 Section 4.4 for comparable l
positions.
The nuclear power plant operations individual shall meet the qualifications of Section 4.2.2 and the maintenance individual j
shall meet the qualifications of Section 4.2.3.
i appointed in writing by the OSC 1.
All alternate members shall be j
Chairman 1to serve on a temporary basis; however, no more than two alternates shall participate as voting members in OSC activities at any one time.
The OSC shall meet at least once per calendar month and as convened by the OSC Chairman or his designated alternate._ A quorum of the OSC shall consist of the - Chairman or his designated alternate and at l
least one half of the members including alternatec.
The OSC shall be responsible for:
1.
Review of a) all procedures required by Technical Specification 6.8 and changes of intent thereto, b) any other i
proposed procedures or changes thereto as determined by the l
General Manager, Nuclear Operations to affect nuclear safety.
2.
Review of all proposed tests and experiments that - affect nuclear safety.
3.
Review of all proposed changes to the Technical Specifications.
4.
Review of all proposed changes or modifications to ' plant systems or equipment that affect nuclear safety.
5.
Investigation of all violations of the Technical Specifications including the preparation and forwarding of
~
reports covering evaluation and recommendations to prevent recurrence to the General Manager, Nuclear Operations and to the Chairman of the Offsite Review Committee.
Page 1 of 2
y INTERT A (C:ntinued)
ONSITE SAFETY COMMITTEE 6.
Review of all reportable events of the type described in 10 CFR 50.73.
7.
Review of facility operations to detect potential safety
)
hazards.
8.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Offsite Review Committee.
J The OSC shall:
l 1.
Recommend to the General Manager, Nuclear Operations written
~
approval or disapproval of items considered under OSC responsibility 1 through 4 above.
2.
Render determinations in writing with regard to whether or not each item considered under OSC responsibility 1 through 5 above constitutes an unreviewed safety question.
3.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Senior Vice President, Nuclear Power Di' vision and the Offsite Review Committee of disagreement between the OSC and the General
- Manager, Nuclear Operations;
- however, the General Manager, Nuclear Operations shall have responsibility for resolution of such disagreements pursuant to Technical Specification 6.1.1.
\\
The OSC shall maintain written minutes of each meeting and copies shall be provided to the General Manger, Nuclear Operations and Chairman of the Offsite Review Committee.
l l
Page 2 of 2
.. _ _ _ _. ~. _ -. _ _.
... _ _.. _ _. _ _ _. _ ~ _. _ _
m i
1 BVPS-2 UFSAR Rev. 8 Offaite Revlaw Committee fORC) f The Offsite Review Committee, as required by Technical Specifications, functions as an independent safety review and audit group to advise the Sr. Vice President and Chief Nuclear Officer of the Nuclear Power Division on all matters concerning 4 safe performance and operation of the nuclear power station.
It 1
has the authority to perform periodic nuclear safety audits of plant operations.
The Offsite Review Committee is comporad i
primarily of personnel outside of the Operations Department, ameh with emnerience in a field Dertinent to nuclear safety.
A Insert Banca C.
Contractors contractors who perform activities affecting quality shall be required to establish and implement a QA program that is consistent with the pertinent requirements of 10 CFR 50, Appendix B and reviewed by the DLC Quality services Unit or implement the applicable portions of the Duquesne Light company Operations Quality Assurance Program. Tne applicable QA Program shall be in effect when the safety-related work is performed.
Duquesne Light Company retains responsibility for assuring that the requirements of 10 CFR 50, Appendix B are satisfied regardless of specific responsibilities assigned to contractors or vendors.
Vendors Vendors supplying safety-related items shall document and j
' implement a Quality. Assurance Program that addresses safety-l related activities performed by the vendor.
The Duquesne Light I
company Quality services Unit shall review and concur with the vendor's Quality Assurance Program.
The vendor's Quality Assurance Program shall be in effect when the safety-related work is performed.
17.2.2 Quality Assurance Program The operations Quality Assurance Program is established and managed by the Quality services Manager who reports to the Senior Vice President.and Chief Nuclear Officer of the Nuclear Power Division.
The Quality Services Manager has the authority to report quality matters to any level necessary within the Duquesne Light Company, including: the Chairman of the Board and Chief Executive Officer; tha President and Chief Operating Officer; and the Senior Vice l
President Customer Operations in order to establish effective corrective action.
The Duquesne Light Company Chairman of the Board j
and Chief Executive Officer; the Senior Vice President and Chief Nuclear Officer of the Nuclear Power Division; the President and chief Operating Officer, i
17.2-4
IN!BRT S OFFSITE REVIEW COMMITTEE The ORC shall tunction to provide independent review and audit of designated activities in the areas of:
1.
Nuclear power plant operations.
2.
Nuclear engineering.
3.
Chemistry and radiochemistry.
4.
Metallurgy 5.
Instrumentation and control.
6.
Radiological safety.
7.
Mechanical and electrical engineering.
8.
Quality assurance practices.
The Chairman and all members of the ORC shall be appointed by the Senior Vice President, Nuclear Power Division.
The membership shall consist of a minimum of five individuals who collectively possess a
. broad based. level of experience and competence enabling the committee to review and audit those activities designated above and to recognize when it is necessary to obtain technical advice and counsel.
An individual may possess expertise in more than one specialty area.
The. collective competence of the committee will be maintained as changes to the membership are made.
All alternate members shall be appointed in writing by. the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities at any one time.
Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC.
The ORC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.
A quorum of the ORC shall consist of the Chairman or his designated alternate and at least four members including alternates.
No more.than a minority of the quorum shall have line responsibility for operation of the facility.
The ORC shall review:
1.
The safety evaluations for a) changes to procedures, equipment or systems and b) tests or experiments completed under the j
provisions of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
2.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section
)
50.59, 10 CFR. 3.
Proposed tests or experiments which involve an unreviewed i
safety question as defined in Section 50.59, 10 CFR.
i 4.
Proposed changes in Technical Specifications or licenses.
5.
Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
Page 1 of 3
m.
i INSERT S (Cantinu'26)
OFF8ITE REVIEW COMMITTEE l
6.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
]
7.
All reportable events of the type described in 10 CFR 50.73.
8.
All recognized indications of an unanticipated deficiency in i
some aspect of design or operation of safety-related structures, systems, or components.
9.
Reports and meeting minutes of the OSC.
j 10.'The results of the Radiological Environmental Monitoring Program annual report provided in accordance with Technical j
Specification 6.9.1.10, prior to submittal.
t j
Audits of the facility activities shall be performed under the j
cognizance of the ORC.
These audits shall encompass:
1 1.
The conformance of facility operations to provisions contained within the Technical Specifications and applicable license conditions.
2.
The performance, training, and qualifications of the entire facility staff.
3.
The results of actions taken to correct deficiencies occurring in facility equipment, structures,
- systems, or methods of operation that affect nuclear safety.
4.
The performance of activities required by the Quality Assurance Program to meet the criteria of Appendix "B",
j 10 CFR 50.
5.
Any other area of facility operation considered appropriate by the ORC or the Senior Vice President, Nuclear Power Division.
6.
The Facility _ Fire Protection Program and implementing procedures at least once per 24 months.
7.
An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 l
months utilizing either qualified off-site licensee personnel i
or an outside fire protection firm.
8.
An inspection and audit of the fire protection and loss i
prevention program shall be performed by a qualified outside I
fire consultant at least once per 36 months.
9.
The Offsite Dose Calculation Manual and implementing 4
procedures.
j
- 10. The Process Control Program and implementing procedures for i
processing and packaging of radioactive wasta.
]
The ORC shall report to and advise the Senior Vice President, Nuclear i
Power Division on those areas of responsibility specified in the 9
paragraphs above that list ORC review and audit topics.
l i
Page 2 of 3 l
. ~..
INSERT 3 (C:ntinuGd) l' OFFSITE REVIEW COMMITTEE Records of ORC activities shall be prepared, approved and distributed as indicated by the following:
i 1.
Minutes of each ORC meeting shall be prepared for and approved 3
by the ORC Chairman or Vice Chairman within 14 days following each meeting.
2.
' Reports of reviews encompassed by the ORC review topics listed above, shall be documented in the ORC meeting minutes.
3.
Audit reports encompassed by the ORC audit topics listed above,.shall be forwarded to the Senior Vice President, Nuclear Power Division and to the management positions responsible for the areas audited within 30 days after 3
completion of the audit.
f-4 L
i l
k i
1 1
i i
i i
1 4
4 i
i Page 3 of 3
)
i u
- l IN3ERT C INDEPENDENT SAFETY EVALUATION GROUP l
BVPS-2 Indeoendent Safety Evaluation GrouD (ISEG)
The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience information, including units of similar design, which may indicate areas for improving unit safety.
The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to corporate management.
If not otherwise implemented, all recommendations shall then be made to the Senior Vice President, j
Nuclear Power Division.
ISEG personnel shall have either:
1.
A bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field, or 2.
At least 5 years of nuclear experience and hold or have held a Senior Reactor Operator License, or 3.
At least 10 years of professional level experience in his field, at least 5 years of which experience shall be in the nuclear field.
A minimum of 50 percent of these personnel shall have the qualifications specified in 1 above.
The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification (not responsible for sign-off function) that these activities are performed correctly and that human errors are reduced as much as practical.
Records of activities performed by the ISEG shall be prepared, maintained, and a summary report shall be forwarded each calendar month to the Senior Vice President, Nuclear Power Division.
Page 1 of 1
0*
R2v. 2 BVPS-2 UFSAR 17.2.5 Instructions, Procedures, and Drawings The Operations QA Program requires that activ. ties affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and will be accompliched in accordance with these instructions, procedures, or drawings.
4 These instructions, procedures, or drawings include, as appropriate, the requiremer.i.s for special tools, test equipment, processes, controls, or skiUs, in order to attain the required level of i
quality. The instructions, procedures, or drawings will include appropriate quant tive or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished, b
The Beaver Valley Power Station Manual of Operating Procedures includes instructions and procedures covering the requirements of NRC Regulatory Guide 1.33, " Quality Assurance Requirements (Operations)",
Appendix A, as they apply to Pressurized Water Reactors. The Manual of Operating Procedures is implemented, enforced, and maintained by the General Manager Nuclear Operations, and his staff.
These procedures and/or instructions include step-by-step procedures for operating and securing the various systems, actions to be taken in the event of abnormal or emergency conditions and precautions to preclude exceeding system or equipment design.
The applicable requirements of ANSI N18.7, " Administrative Control for Nuclear Power Plants," were used as guidance in the development of startup, operating, emergency, maintenance, and testing procedures.
Maintenance, repair, modifications, testing, and refueling activities which affect the quality or safety of Category I items are prescribed by documented instructions, procedures, or drawings.
These instructions, procedures, or drawings include, as appropriate, the requirements for special tools, test equipment, processes, controls, or skills in order to attain the required level of quality.
--> I nset, D 17.2.6 Document Control l
The Operations QA Program establishes measures to control the issuance of documents such as instructions, procedures, and drawings, affecting the quality of safety-related structures, systems, and components.
The Operations QA Program includes provisions for assuring that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel, and are distributed to and used j
at the location where the prescribed activity is performed, prior to the onset of work.
Control measures are applied to assure that 17.2-12
INSERT D PROCEDURES Each procedure and administrative policy of Technical Specification 6.8.1 and changes of intent thereto, shall be reviewed by the OSC and
)
approved by the General Manager, Nuclear Operations, predesignated alternate or a predesignated Manager to whom the General Manger, Nuclear Operations has assigned in writing the responsibility for review and approval of specific subjects considered by the committee, as applicable.
Changes to procedures and administrative policies of Technical Specification 6.8.1 that do not receive OSC review, such as correcting typographical errors, reformatting procedures and other changes not affecting the purpose for which the procedure is performed shall receive an independent review by a
qualified i
individual and approved by a designated manager or director.
Temporary changes to procedures of Technical Specification 6.8.1 may be made provided:
1.
The intent of the original procedure is not altered.
2.
The change is approved by two (2) members of the plant management staff, at least one (1) of whom holds a Senior I
Reactor Operator's License on the unit affected.
3.
The change is documented, reviewed by the OSC and approved by 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, within 14 days of implementation.
1 Page 1 of 1
9 1
o.
e*
BVPS-2 UFSAR The DLC Quality Services Manager has the authority to prepare written notices to the appropriate level of management requesting changes and/or revisions to any program or procedure which may have resulted in the generation of repeated nonconformances.
In addition, the Quality Services Manager and designated staff may direct the stopping of all work pending corrective action.
17.2.17 Quality Assurance Records The Operations QA Program requires that sufficient records be maintained to furnish evidence of activities affecting quality. The records will include at least the followings operating logs and results of reviews, drawings, inspections, tests, audits, monitoring of work performance, and material analyses. The records will also l
include closely related data much as qualifications of personnel, procedures, and equipment.
Inspection and test records will, as a
- minimum, identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted.
l A r3 cords management system has been established to assure that records will be identifiable and retrievable.
Duquesne Light company has established requirements concerning record retention such as duration, location, and assigned responsibility.
Provisions include facilities for perm &nent records retention, including steps taken to assure preservation, protection, and controlled access.
l The Duquesne Light Company Nuclear Services Division is responsible for the generation and maintenance of a document listing the location and type of all pertinent documents which will be stored offsite.
>Lsect E 17.2.18 Audits The Operations QA Program requires that a comprehensive system of planned and periodic audits be carried out to verify compliance with all aspects of the QA Program and to determine its effectiveness.
Audits will be performed in accordance with written procedures or checklists by appropriately trained personnel not having direct responsibilities in the areas being audited. Audit results will be documented and reported to management having direct responsibility in the area audited.
Responsible management will take the necessary action to respond to any deficiencies or nonconformances identified in the audit report.
Follow-up action, including re-audit of deficient areas, will be taken as necessary to assure that all deficiencies or nonconformances noted have been correc.ed.
17.2-20
--.___.______._____.m.._._--___
m_
d IN2ERT E 0 ".
t RECIRD RETENTICN i
I l
The following records shall be retained for at least five (5) years:
i 1.
Records and logs of facility operation covering the time interval at each power level.
2 2.
Records and logs of principal maintenance activities, l
inspections, repair and replacement of principal items of
- l j
equipment related to nuclear safety.
j
{
3.
All reportable events of the type described in 10 CFR 50.73.
4.
Records of surveillance activities, inspections and 4
calibrations required by the Technical Specifications.
i l
'5.
Records of reactor tests and experiments.
j 6.
Records of changes made to operating procedures.
j 7.
Records of radioactive shipments.
l l
8.
Records of sealed source leak tests and results.
l
'9.
Records of annual physical inventory of all sealed source material of record.
i The. following records shall be retained ' for the duration of the Facility Operating License:
1.
. Records and drawing changes reflecting facility design 3
modifications made to systems and equipment described in the Final Safety Analysis Report.
i 2.
Records of new irradiated fuel invectory, fuel transfers and l
assembly burnup histories.
j 3.
Records of facility radiation and contamination surveys.
l 4.
Records of radiation exposure for all individuals entering l
radiation control areas.
5.
Records of gaseous and liquid radioactive material released to i
the environs.
.6.
Records of transient or operational cycles for those facility I
components designed for a limited number of transients or j
cycles.
{
7.
Records of training and qualification for current members of j
the plant staff.
i 8.
Records of in-service inspections performed pursuant to the Technical Specifications.
j 9.
Records of Quality Assurance activities required by the QA Manual.-
- 10. Records of reviews performed for changes made to procedures or equipment or reviews ' of tests and experiments pursuant to 10 CFR 50.59.
- 11. Records of meetings of the OSC and the ORC..
- p. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.
- 13. Records of analyses required by the Radiological Environmental Monitoring Program.
- 14. Records of reviews performed for changes made to the Offsite Dose Calculation Manual and the Process Control Program.
Page 1 of 1
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