ML20148G070
| ML20148G070 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 01/13/1988 |
| From: | Perkins K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148G074 | List: |
| References | |
| NUDOCS 8801260399 | |
| Download: ML20148G070 (20) | |
Text
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION g
j WASHINGTON, D. C. 20555
\\...../
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET N0. STN 50-483 AMENDMENT 70 FACILITY OPERATING LICENSE Amendment No.30 License No. NPF-30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by Union Electric Company (the licensee) dated February 19, 1987, as supplemented by letter dated October 30, 1987, complies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will 03erate in confonnity with the application, the provisions of tie Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; i
D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Par ^ 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, Facility Operating License No. NPF-30 is amended as follows:
(A) Change the Technical Specifications as indicated in the attachment to this license amendment, and amend paragraph 2.C.(2) to read as follows:
8801260399 880113 PDR ADOCK 050004B3 P
. (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 30, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license.
UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(B) Amend paragraphs 2.C.(5)(c) and (d) to read as follows:
(c) The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the SNUPPS Final Safety Analysis Report for the facility through Revision 15, the Callaway site addendum through Revision 8, and as approved in the SER through Supplement 4, subject to provision d below.
(d) The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely j.
affect the ability to achieve and maintain safe shutdown in the event of a fire.
(C) Deleteparagraph2.C.(5)(e).
3.
This license amendment is effective upon issuance and shall be implemented upon the licensee's completion of the necessary procedural changes. The licensee will notify the Commission in writing when the necessary procedural changes have been completed.
FOR THE NUCLEAR REGULATORY COMMISSION Kenneth E.
e ins, D e Project Directorate III-3 Division of Reactor Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: January 13, 1988
ATTACHMENT TO LICENSE AMENDMENT NO. 30 OPERATING LICENSE N0. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the encloseo pages. The revised pages are identified.by the captioned amendment number and contain marginal lines indicating the area of change. Corresponding overleaf pages are provided to maintain document completeness.
REMOVE INSERT VI VI X
X XVI XVI 3/4 3-57 3/4 3-58 3/4 3-59 F
3/4 3-60 3/4 3-61 3/4 3-57 through 3/4 3-61 3/4 7-27 3/4 7-28 3/4 7-29 3/4 7-30 t
3/4 7-31 3/4 7-32 3/4 7-33 3/4 7-34 3/4 7-35 3/4 7-36 3/4 7-27 through 3/4 7-36 B 3/4 3-5 B 3/4 3-5 B 3/4 7-7 B 3/4 7-7 B 3/4 7-8 8 3/4 7-8 I
6-9 6-9 6-15 6-15 l
l l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.3 MONITORING INSTRUMENTATION Radiation Moni toring for Plant Operations.................
3/4 3-38 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PL AN T 0 P E RAT I O N S.....................................
3/ 4 3 - 3 9 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS...........
3/4 3-41 Movabl e Incore Detec tors..................................
3/4 3-42 Seismi c Ins trumen ta ti on...................................
3/4 3-4 3 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.....................
3/4 3-44 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................
3/4 3-45 Meteorological Instrumentation............................
3/4 3-46 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION..............
3/4 3-47 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................
3/4 3-48 Remote Shutdown Ins trumentation...........................
3/4 3-49 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION.............
3/4 3-50 g
- )
J TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION j
SURVEILLANCE REQUIREMENTS............................
3/4 3-51 Accident Moni toring Instrumentation.......................
3/4 3-52 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION...................
3/4 3-53 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-55 l
Loose-Part Detection System..............................
3/4 3-62 Radioactive Liquid Effluent Monitoring Instrumentation....
3/4 3-63
\\
l l
l CALLAWAY - UNIT 1 VI Amendment No. 30 l
l l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.....................................
3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RAT E D T HE RMAL P0 W E R.................................
3/ 4 2 - 3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fg(Z)......................
3/4 2-4 l
FIGURE 3.2-2 K(Z)_ NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT....
3/4 2-5 9
N 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F3............
3/4 2-8 l
3/4.2.4 QUADRANT POWER TILT RATI0.................................
3/4 2-10 3/4.2.5 DNB PARAMETERS............................................
3/4 2-13 i
TABLE 3.2-1 DNB PARAMETERS.........................................
3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.......................
3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION....................
3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.....
3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................................
3/4 3-9 V
f 3/4.3.2
' ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.........................................
3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION......................................
3/4 3-14 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS.......................
3/4 3-22 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES..............
3/4 3-29 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............
3/4 3-33 CALLAWAY - UNIT 1 V
Amendment No. 28
INDEX LIMITING; CONDITIONS FOR'0PERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued)
TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P......................
3/4 7-3 Auxi l i a ry Feedwa ter Sys tem................................
3/4 7-4 Condensa te Storage Tank...................................
3/4 7-6 S p e c i fi c Ac ti v i ty.........................................
3/ 4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................
3/4 7-8 Main Steam Line Isola tion Valves..........................
3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........
3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................
3/4 7-11 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM............................
3/4 7-12 3/4.7.5 ULT IMATE H EAT S I N K........................................
3/ 4 7-13 1
3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.................
3/4 7-14 3/4.7.7 EMERGENCY EXHAUST SYSTEM..................................
3/4 7-17 3/4.7.8 S N U B B E RS..................................................
3/ 4 7-1 9 ii FIGURE 4.7-1 SAMPLING PLAN 2) FOR SNUBBER FUNCTIONAL TEST..........
3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION...............................
3/4 7-25 3/4.7.10 Deleted 3/4.7.11 Deleted 3/4.7.12 AREA TEMPERATURE MONIT0 RING...............................
3/4 7-37 TABLE 3.7-4 AREA TEMPERATURE M0NITORING............................
3/4 7-38 CALLAWAY - UNIT 1 X
Amendment No. 30
INDEX LIMllING CONDIIIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE EMERGENCY CORE COOLING SYSTEMS (Continued) 3/4.5.5 REFUELING WATER STORAGE TANK.............................
3/4 5-10 3/4.6 ' CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity...................
3/4 6-1 Containment Leakage.............................
3/4 6-2 Containment Air Locks....................................
3/4 6-4 Internal Pressure........................................
3/4 6-6 A i r T e mp e ra t u re..........................................
3/4 6-7 Containment Vessel Structural Integrity..................
3/4 6-8 Containment Ventilation System...........................
3/4 6-11 3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.................................
3/4 6-13 Spray Additive System....................................
3/4 6-14 Containment Cooling System...............
3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES....................
3/4 6-16 TABLE 3.6-1 CONTAINMENT ISOLATION VALVES..........................
3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.....................................
3/4 6-31 Ilydrogen Control Systems.................................
3/4 6-32
,3/4. / PLANI SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves.........................................
3/4 7-1 TABLL 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION...................
3/4 7-2 Call AWAY - UNIT 1 IX
INDEX BASES SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 STEAM GENERATORS.........................................
B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE...........................
B 3/4 4-4 3/4.4.7 CHEMISTRY................................................
B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY........................................
B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..............................
B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS..........................
B 3/4 4-10 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................
B 3/4 4-12 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RTNOT FOR REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550 F....
B 3/4 4-13 3/4.4.10 STRUCTURAL INTEGRITY.....................................
B 3/4 4-17 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................
B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................
B 3/4 5-1 3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS.....................
B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK.............................
B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT......................................
B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.....................
B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................
B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L..................................
B 3/4 6-4 CALLAWAY - UNIT 1 XV l
j.
INDEX
(
BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................
B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........
B 3/4 7-3 3/e.7.3 COMPONENT COOLING WATER SYSTEM...........................
B 3/4 7-3 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM...........................
B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK.......................................
B 3/4 7-3 j
3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................
B 3/4 7-4 3/4.7.7 EMERGENCY EXHAUST SYSTEM.................................
B 3/4 7-4 3/4.7.8 SNUBBERS.................................................
B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION..............................
B 3/4 7-6 3/4.7.10 Deleted I
3/4.7.11 Deleted I
3/4.7.12 AREA TEMPERATURE MONITORING..............................
B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and j
ONSITE POWER DISTRIBUTION...............................
B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES..................
B 3/4 8-3 il 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION..........................................
B 3/4 9-1 l
3/4.9.3 DECAY TIME....
B 3/4 9-1 l
l 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................
B 3/4 9-1 I
3/4.9.5 C O MM U N I C AT I O N S...........................................
B 3/4 9-1 CALLAWAY - UNIT 1 XVI Amendment No. 30
. - = _ _ _ _ _
+
INTENTIONALLY BLANK I
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CALLAWAY - UNIT 1 3/4 3-57 through 3/4 3-61 Amendment No. 30
INSTRUMENTATION LOOSE-PART DETEC110N SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.3 The Loose-Part Detection System shall be OPERABLE.
APPLICABILITY:
MODES 1 and 2.
ACTION:
With one or more Loose-Part Detection System channels inoperablo for a.
more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status, b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.8 Each channel of the Loose-Part Detection System shall be demonstrated OPERABLE by performance of:
A CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a.
b.
An ANALOG CHANNEL OPERATIONAL TEST except for verification of Setpoint at least once per 31 days, and c.
A CHANNEL CALIBRATION at least once per 10 months.
CALLAWAY - UNIT 1 3/4 3-62 1
I
- [?
INTENTI0llALLY BLANK h-i CALLAWAY - UNIT 1 3/4 7-27 through 3/4 7-36 Amendment No. 30 L
)
7-e INSTRUMENTATION BASES 3/4.3.3.8 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components.
The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, "Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors,"
May 1981.
3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION I
The radioactive liquid effluent instrumentation is provided to monitor i
and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with.the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
i i
I CALLAWAY - UNIT 1 B 3/4 3-5 Amendment No. 30
INSTRUMENTATION
. BASES 3/4.3.3.10 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm /
Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitor used to show compliance with the gaseous effluent release requirements of Specifica-tion 3.11.2.2 shall be such that concentrations as low as 1 x 10 6 pCi/cc are measurable.
3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Although the orientation of the turbine is such that the number of potentially damaging missiles which could impact and damage safety-related components, equipment, or structures is minimal, protection from excessive turbine overspeed is required.
i l
CALLAWAY - UNIT 1 B 3/4 3-6 l
PLANT SYSTEMS BASES SEALED' SOURCE CONTAMINATION (Continued)
Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not.
Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3/4.7.10 Deleted i
l 3/4.7.11 Deleted 2
s CALLAWAY - UNIT 1 B 3/4 7-7 Amendment No. 30
-m
l l
PLANT SYSTEMS BASES 1
3/4.7.12 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of +3'F.
t i
L i
d CALLAWAY - UNIT 1 B 3/4 7-8 Amendment No. 30
ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) m.
Review of Unit operations to detect potential hazards to nuclear safety; n.
Investigations or analysis of special subjects as requested by the Chairman of the NSRB; I
o.
Review of Unit Turbine Overspeed Protection Reliability Program and revisions thereto; and I
p.
Review of the Fire Protection Program and submitting recommended changes to the NSRB.
6.5.1.7 The ORC shall:
a.
Recommend in writing to the Manager, Callaway Plant approval or disapproval of items considered under Specifications 6.5.1.6a.
through e.,
i., J.,
k.,
l.,
o., and p. above:
g b.
Render determinations in writing with regard to whether or not each item considered under Specifications 6.5.1.6b. through e., and m.,
above, constitutes an unreviewed safety question; and c.
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 Vice President-Nuclear and the Nuclear Safety Review Board of disagreement between the ORC and the Manager, Callaway Plant; however, the Manager, Callaway Plant shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.
RECORDS 6.5.1.8 The ORC shall maintain written minutes of each ORC meeting that, at a minimum, document the results of all ORC activities performed under the responsibility provisions of these Technical Specifications.
Copies shall be provided to the Vice President-Nuclear and the Nuclear Safety Review Board.
6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)
V' FUNCTION i
6.5.2.1 The NSRB shall function to provide independent review and audit of designated activities in the areas of:
a.
Nuclear power plant operations, b.
Nuclear engineering, c.
Chemistry and radiochemistry; d.
Metallurgy, e.
Instrumentation and control, f.
Radiological safety, 9
Mechanical and electrical engineering, and h.
Quality assurance practices.
The NSRB shall report to and advise the Vice President-Nuclear on those areas of responsibility stated in Specifications 6.5.2.8 and 6.5.2.9.
CALLAWAY - UNIT 1 6-9 Amendment No. 30
ADMINISTRATXVE CONTROLS COMPOSITION 6.5.2.2 The NSP.B shall be composed of at least the following members:
Chairman:
General Manager, Engineering (Nuclear)
Member:
Manager, Nuclear Engineering Member:
Manager, Nuclear Safety and Emergency Preparedness Member Manager, Quality Assurance Member:
General Manager, Nuclear Operations l
Member:
Supervising Engineer, Nuclear Fuels Additional members and Vice Chairmen may be appointed by the Chairman.
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NSRB Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSRB activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NSRB Chairman to provide expert advice to the NSRB.
-MEETING FREQUENCY 6.5.2.5 The NSRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter.
QUALIFICATIONS 6.5.2.6 The NSRB members shall hold a Bachelor's degree in an engineering or physical science field, or equivalent experience, and a minimum of 5 years of technical experience of which a minimum of 3 years shall be in one or more of the disciplines of Specification 6.5.2.1.
QUORUM 6.5.2.7 The quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least two-thirds of the HSRB members including alternates.
No more than a minority of the auorum shall have line responsibility for operation of the unit.
For the purpose of a quorum, those considered to have line responsibility will include the General Manager, Nuclear Operations, and personnel reporting to the General Manager, Nuclear Operations.
CALLAWAY - UNIT 1 6-10 Amendment No. 16 a
ADMINISTRATIVE CONTROLS SAFETYLIMITVIOLATION(Continued) c.
The Safety Limit Violation Report shall be submitted to the Commission, the NSRB and the Vice President-Nuclear within 14 days of the violation; and d.
Critical operation of the unit shall not be resumed until authorized by the Commission.
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
f a.
The applicable procedures recommended tri Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; b.
The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7.1 of Generic Letter No. 82-33; c.
Plant Security Plan implementation; d.
Radiological Emergency Response Plan implementation; e.
PROCESS CONTROL PROGRAM implementation; f.
0FFSITE DOSE CALCULATION MANUAL implementation; g.
Quality Assurance Program implementation for effluent and environ-mental monitoring; h.
Turbine Overspeed Protection Reliability Program; and i.
Fire Protection Program implementation.
l 6.8.2 Each procedure and administrative policy of Specification 6.8.1 above, and changes thereto, including temporary changes shall be reviewed prior to implementation as set forth in Specification 6.5 above.
6.8.3 The plant Administrative Procedures and changes thereto shall be reviewed in accordance with Specification 6.5.1.6 and approved in accordance with C
Specification 6.5.3.1.
The associated implementing procedures and changes thereto shall be reviewed and approved in accordance with Specification 6.5.3.1.
6.8.4 The following programs shall be established, inplemented and maintained:
a.
Reactor Coolant Sources Outside Containment A' program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation portion of the Containment Spray System, Safety Injection System, Chemical and Volume Control System, and RHR System. The program shall include the following:
1)
Preventive maintenance and periodic visual inspection requirements, and CALLAWAY - UNIT 1 6-15 Amendment No.
30
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 2)
Integrated leak test requirements for each system at refueling cycle intervals or less, b.
In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
- 1) Training of personnel,
- 2) Procedures for monitoring, and
- 3) Provisions fon maintenance of sampling and analysis equipment, c.
Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation.
This program shall include:
1)
Identification of a sampling schedule for the critical variables and control points for these variables, 2) 7dentification of the proceJures used to measure the values of the critical variables, 3)
Identification of process sampling points, which shall include monitoring the. discharge of the condensate pumps for evidence of condenser in-leakage, j
- 4) Procedures for the recording and management of data, g
- 5) Procedures defining corrective action for all off-control point chemistry conditions, and
- 6) A procedure identifying:
(a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
d.
Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.
The program shall include the following:
- 1) Training of personnel,
- 2) Procedures for sampling and analysis, and 3)
Provisions for maintenance of sampling and analysis equipment.
CALLAWAY - UNIT 1 6-16
-