ML20236X463
| ML20236X463 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/31/1998 |
| From: | Thomas C NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236X465 | List: |
| References | |
| GL-89-01, GL-89-1, NUDOCS 9808100030 | |
| Download: ML20236X463 (43) | |
Text
_____-_- - -___ - -
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3 UNITED STATES g
,g NUCLEAR REGULATORY COMMISSION
[
WASHINGTON, D.C. 20066 4001 i
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BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.177 License No. DPR-35 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Boston Edison Company (the licensee) dated September 19,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: li) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated j
in the attachment to this license amendment, and paragraph 3.B of Facility Operating J
License No. DPR-35 is hereby amended to read as follows:
3.B. Technical Specificatiqng The Technical Specifications contained in Appendix A, as revised through Amendment No.177, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
In addition, the license is amended to add paregraph 3.J to Facility Operating License No.
DPR-35 as follows:
3.J. Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 177, are hereby incorporated into this license. Boston Edison Company shall operate the facility in accordance with the Additional Conditions.
9008100030 980731 PDR ADOCK 05000293 P
o 2
3.
This license amendment is effective as of TS date of issuance and shall be implemented within 30 days. Implementation of the amendment shallinclude relocation of various sections of the technical specifications as described in the licensee's application dated September 19,1997, and evaluated in the staff's safety evaluation attached to this amendment. Implementation also includes fulfillment of the specified conditions set forth in paragraph 2, except that the licensee shall have until the next scheduled Updated Final Safety Analyses Report (UFSAR) update to incorporate the UFSAR relocations.
FOR THE NUCLEAR REGULATORY COMMISSION Cecil O. Thomas, Director Project Directorate 1-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
- 1. Pages 3*,4*, and 5* of License No. DPR-35
- 2. Changes to The Technical Specifications l
- 3. Appendix B Date of issuance: July 31,1998 l
I
- Pages 3,4, and 5 of the license and Appendix B are attached, for convenience, for the composite license to reflect this change.
L--_-----_--_---_-------__-----------------------.-----------
ATTACHMENT TO LICENSE AMENDMENT NO.177 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 1.
Revise Ucense as follows:
Remove Paaes insert Paoes 3
3 4
4 5
5 2.
Insert page 1 of Appendix B of the License 3.
Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert iii lii 4
1-1 1-1 1-2 1-2 1-3 1-3 1-4 1-4 1-5 1-5 1-6 1-6 i
3/4 8-1 3/4 8-1 3/4 8-2 3/4 8-2 B3/4 8-1 B3/4 8-1
{
B3/4 8-2 B3/4 8-2 B3/4 8-3 B3/4 8-3 B3/4 8-4 B3/4 8-4 B3/4 8-5 B3/4 8-5 B3/4 8-6 B3/4 8-6 B3/4 8-7 B3/4 8-7 B3/4 8-8 B3/4 8-8 B3/4 8-9 B3/4 8-9 4.0-1 4.0-1 4.0-2 4.0-2 5.0-1 5.0-1 5.0-2 5.0-2 5.0-3 5.0-3 5.0-4 5.0-4 5.0-5 5.0-5 5.0-6 5.0-6 5.0-7 5.0-7 5.0-8 5.0-8 5.0-9 5.0-9 5.0-10 5.0-10 5.0-11 5.0-11 5.0-12 5.0-12 5.0-13 5.0-13 5.0-14 5.0-14
a B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 177, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
Records Boston Edison shall keep facility operating records in accordance with the requirements of the Technical Specifmations.
D.
Eaualizer Valve Restriction - DELETED E.
Recirculation Looo Inocerable The reactor shall not be operated with one recirculation loop out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a hot shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner retumed to service.
F.
Fire Protection Boston Edison shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated December 21,1978 as t
supplemented subject to the following provision:
Boston Edison many make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
G.
Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and c; qualification, and safeguards contingency plans including amendments made pursuant to provisions I
of the Miscellaneous Amendments and Search Requirements revisions to l
10CFR73.55 (51FR27817 and 27822) and to the authority of 10CFR50.90 and 10CFR50.54(p). The plans, which contain Safeguards information protected
(
under 10CFR73.21, are entitled:
- Pilgrim Nuclear Power Station Physical Security Plan," with revisions submitted through September 18,1987; " Pilgrim Nuclear l
Power Station Guard Training and Qualification Plan," with revisions submitted through September 24,1904; and " Pilgrim Nuclear Power Station Safeguards Contingency Plan," with revisions submitted through February 15,1984. Changes made in accordance with 10CFR73.55 shall be implemented in accordance with the schedule set forth therein.
Revision 177 Amendment No. 453,177 i
4 H.
Lono Term Proaram (1)
The " Plan for the Long Term Program for Pilgrim Nuclear Power Station" (the Plan) submitted on May 7,1984, is approved.
a)
The Plan shall be followed by the licensee from and after the effective date of this amendment.
b)
Changes to dates for completion of items identified in Schedule B of the Plan do not require a license amendment. Dates specified in Schedule A shall be changed only in accordance with app!icable NRC procedure.
l.
Post-Accident Samofina System. NUREG-0737. Item II.B.3. and Containment Atmospheric Monitorino System. NUREG-0737. Item ll F.1(6)
The licensee shall complete the installation of a post-accident sampling system and a containment atmospheric monitoring system as soon as practicable, but no later than June 30,1985.
J.
Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 177, are hereby incorporated into this license. Boston Edison Company shall operate the facility in accordance with the Additional Conditions.
4.
This license is subject to the following condition for the protection of the environment:
Boston Edison shall continue, for a period of five years after initial power operation of the facility, an environmental monitoring program similar to that presently existing with the Commonwealth of Massachusetts (and described generally in Section C-Ill of Boston Edison's Environmental Report, Operating License Stage dated September,1970) as a basis for determining the extent of station influence on raarine resources and shall mitigate adverse effects, if any, on marine resources.
5.
Boston Edison has not completed as yet construction of the Rad Waste Solidification System and the Augmented Off-Gas System. Limiting conditions conceming these systems are set forth in the Technical Specifications.
6.
Pursuant to Section 105c(8) of the Act, the Commission has consulted with the Attomey General regarding the issuance of this operating license. After said consultation, the Commission has determined that the issuance of this license, subject to the conditions i
l set forth in this subparagraph 6., in advance of consideration of and findings with respect to matters covered in Section 105c of the Act, is necessary in the public interest to avoid unnecessary delay in the operation of the facility. At ihe time this operating license is being issued an antitrust proceeding has not been roticed. The Commission, accordingly, has made no determination with respect to matters covered in Section 105c of the Act, including conditions, if any, which may be appropriate as a result of the Revision 177 '
Amendment No. 75, 85,443, 177
. outcome of any antitrust proceeding. On the basis ofits findings made as a result of an antitrust proceeding, the Commission may continue this license as issued, rescind this license or amend this license to include such conditions as the Commission deems appropriate. Boston Edison and others who may be affected hereby are accordingly on notice that the granting of this license is v.ithout prejudice to any subsequent licensing action, including the imposition of appropriate conditions, which may be taken by the commission as a result of the outcome of any antitrust proceeding. In the course ofits planning and other activities, Boston Edison will be expected to conduct itself accordingly.
7.
This license is effective as of the date ofissuance and shall expire June 8,2012.
FOR THE ATOMIC ENERGY COMMISSION Original Signed by A. Giambusso A. Giambusso, Deputy Diructor for Reactor Projects Directorate of Licensing Attachments:
Appendix A - Technical Specifications (Radiological)
Date of issuance: September 15,1972 l
i t
Revision 177 Amendment 464,177 l
APPENDIX D 1
ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR 35 Boston Edison Company shall comply with the following conditions on the schedules noted below:
Amendment implementation Number Additional Conditions Rait 177 The licensee is authorized to relocate certain The amendment shall be Technical Specifications requirements to implemented within 30 licensee-controlled documents.
days from July 31, 1998, implementation of this amendment shall except that the licensee include relocation of various sections of the shall have until the next technical specifications to the appropriate scheduled Updated Final documents as described in the licensee's Safety Analysis Report application dated September 19,1997, and (UFSAR) update to in the staff's safety evaluation attached to incorporate the UFSAR this amendment.
relocations.
l I Amendment No.
177 I
' TABLE Or CONTENTS 1.0 DEFINITIONS 1-1 2.0 SAFETY LIMITS 2-1 2.1 Safety Limits 2-1 2.2 -
Safety Limit Violation 2-1 BASES B2-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 3/4.1-1 BASES B3/4.1-1 3.2 PROTECTIVE INSTRUMENTATION 4.2 3/4.2-1 A.
Primary Containment isolation Functions A
3/4.2-1 8.
Core and Containment Cooling Systems B
3/4.2-1 C.
Control Rod Block Actuation C
3/4.2-2 D.
Radiation Monitoring Systems D
3/4.2-2 E.
Drywell Leak Detection E
3/4.2-3 F.
Surveillance Information Readouts F
3/4.2-3 G.
Recirculation Pump Trip / Altemate Rod G
Insertion 3/4.2-4 H.
Drywell Temperature H
3/4.2-5 BASES B3/4.2-1 3.3 REACTIVITY CONTROL 4.3 3/4.3-1 A.
Reactivity Limitations A
3/4.3-1 B.
3/4.3-2 C.
Scram insertion Times C
3/4.3-4 D.
3/4.3-5 E.
Reactivity Anomalies E
3/4.3-6 F.
Alternate Requirements 3/4.3-6 G.
3/4.3-6 BASES B3/4.3-1 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 3/4.4-1 BASES B3/4.4-1 3.5 CORE AND CONTAINMENT COOLING 4.5 3/4.5-1 SYSTEMS A.
Core Spray and LPCI Systems A
3/4.5-1 B.
Containment Cooling System B
3/4.5-3 C.
HPCI System C
3/4.5-4 D.
Reactor Core Isolation Cooling (RCIC) System D
3/4.5-5 E.
Automatic Depressurization System (ADS)
E 3/4.5-6 F.
Minimum Low Pressure Cooling and Diesel F
Generator Availability 3/4.5-7 G.
(Deleted)
G 3/4.5-8 H.
Maintenance of Filled Discharge Pipe H
3/4.5-8 BASES B3/4.5-1 PNPS i
Amendment No.177 l
TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 3/4.6-1 A.
Thermal and Pressurization Limitations A
3/4.6-1 B.
Coolant Chemistry B
3/4.6-3 C.
Coolant Leakage C
3/4.6-4 D.
Safety and Relief Valves D
3/4.6-6 E.
Jet Pumps E
3/4.6-7
]
F.
Jet Pump Flow Mismatch F
3/4.6-8
{
G.
Structural Integrity G
3/4.6-8 H.
Deleted H
3/4.6-8 1.
Shock Suppressors (Snubbers) l 3/4.6-9 BASES B3/4.6-1 3.7 CONTAINMENT SYSTEMS 4.7 3/4.7-1 A.
3/4.7-1 B.
Standby Gas Treatment System and Control B
Room High Efficiency Air Filtration System 3/4.7-11 C.
3/4.7-16 BASES B3/4.7-1 3.8 PLANT SYSTEMS 4.8 3/4.8-1 3.8.1 Main Condenser Offgas 4.8.1 3/4.8-1 3.8.2 Mechanical Vacuum Pump 4.8.2 3/4.8-2 BASES B3/4.8-1 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 3/4.9-1 A.
Auxiliary Electrical Equipment A
3/4.9-1 B.
Operation with Inoperable Equipment B
3/4.9-4 BASES B3/4.9-1 3.10 CORE ALTERATIONS 4.10 3/4.10-1 A.
Refueling Interiocks A
3/4.10-1 8.
Core Monitoring B
3/4.10-1 C.
Spent Fuel Pool Water Level C
3/4.10-2 D.
Multiple Control Rod Removal D
3/4.10-2 BASES B3/4.10-1 i
3.11 REACTOR FUEL ASSEMBLY 4.11 3/4.11-1 A.
Average Planar Linear Heat Generation Rate A
(APLHGR) 3/4.11-1 l
B.
Linear Heat Generation Rate (LHGR)
B 3/4.11-2 C.
Minimum Critical Power Ratio (MCPR)
C 3/4.11-2 D.
Power / Flow Relationship During Power D
Operation 3/4.11-4 BASES B3/4.11-1 PNPS ii Amendment No.177 l
TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.12 FIRE PROTECTION 4.12 3/4.12-1 Alternate Shutdown Panels 3/4.12-1 BASES B3/4.12-1 3.13 INSF_RVICE CODE TESTING 4.13 3/4.13-1 A.
Inservice Code Testing of Pumps and Valves 3/4.13-1 BASES B3/4.13-1 4.0 DESIGN FEATURES 4.0-1 4.1 Site Location 4.0 1 4.2 Reactor Core 4.0-1 4.3 Fuel Storage 4.0-1 4.3.1 Criticality 4.0 1 4.3.2 Drainage 4.0-2 4.3.3 Capacity 4.0-2 4.3.4 Heavy Loads 4.0-2 5.0 ADMINISTRATIVE CONTROLS 5.0-1 5.1 Responsibility 5.0-1 5.2 Organization 5.0 2 5.3 Unit Staff Qualifications 5.0-4 5.4 Procedures 5.0-5 5.5 Programs and Manuals 5.0-6 5.6 Reporting Requirements 5.0-10 5.7 High Radiation Area 5.0-13 PNPS iii Amendment No.177 l
1.0 DEFINITIONS The succeeding frequently used terms are explicitly dc9ned so that a uniform interpretation of the specifications may be achieved.
ACTidN ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.
AUTOMATIC PRIMARY Are primary containment isolation valves which receive an CONTAINMENT automatic primary containment group isolation signal.
ISOLATION VALVES COLD CONDITION Reactor coolant temperature equal to or less than 212*F.
I CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, eurces, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following l
exceptions are not considered to be CORE ALTERATIONS:
l
- a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and
- b. Control rod movement, provided there are no fuel assemblies 1
in the associated core cell.
f i
Suspension of CORE ALTERATIONS shall not preclude l
completion of movement of a component to a safe position.
CORE OPERATING The COLR is a reload-cycle specific document that provides core l
LIMITS REPORT (COLR) operating limits for the current operating reload cycle. These cycle specific core operating limits shall be determined for each l
reload cycle in accordance with Specification 5.6.5. Plant operation within these operating limits is addressed in individual specifications.
DESIGN POWER DESIGN POWER means a steady state power level of 1998 thermal megawatts.
FIRE SUPPRESSION A FIRE SUPPRESSION WATER SYSTEM shall consist of: a i
WATER SYSTEM water source (s); gravity tank (s) or pump (s); and distribution piping with associated sectionalizing control or isolation valves. Such valves shall include hydrant post indicator valves and the first valve ahead of the water flow alarm device on each sprinkler, l
hose standpipe or spray system riser.
l HOT STANDBY HOT STANDBY CONDITION means operation with coolant CONDITION temperature greater than 212*F, system pressure less than 600 i
psig, the main steam isolation valves closed and the mode switch in startup.
IMMEDIATE IMMEDIATE means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
l 1
PNPS 1-1 Amendment No.177 l
I 1.0 DEFINITIONS (Cont)
INSTRUMENT An INSTRUMENT CALIBRATION means the adjustment of an CALIBRATION instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including actuation, alarm or trip.
INSTRUMENT CHANNEL An INSTRUMENT CHANNEL means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.
INSTRUMENT CHECK An INSTRUMENT CHECK is a determination of acceptable operability by observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.
INSTRUMENT An INSTRUMENT FUNCTIONAL TEST means the injection of a FUNCTIONAL TEST simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action.
LEAKAGE a.
Identified LEAKAGE:
1.
Reactor coolant LEAKAGE into drywell collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or
- 2. Reactor coolant LEAKAGE into the drywell atmosphere from sources which are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to to Pressure Boundary Leakage.
b.
Unidentified LEAKAGE:
Unidentified LEAKAGE shall be all reactor coolant leakage which is not identified Leakage.
c.
l Pressure Boundary LEAKAGE shall be leakage through a 3
l non-isolable fault in a reactor coolant system component l
body, pipewall or vessel wall.
LIMITING CONDITIONS The LIMITING CONDITIONS FOR OPERATION specify the j
FOR OPERATION (LCO) minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these
{
conditions are met, the plant can be operated safely and j
abnormal situations can be safely controlled.
i PNPS 1-2 Amendment No.177 l
1.0 DEFINITIONS LIMITING SAFETY The LIMITING SAFETY SYSTEM SETTINGS are settings on SYSTF.M SETTING instrumentation which initiate the automatic protective action at a (LSSS) level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.
LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST means a test of all 1
FUNCTIONAL TEST relays and contacts of a logic circuit from sensor to activated
{
device to insure components are operable per design intent.
J l
Where practicable, action will go to completion (i.e., pumps will 1
l be started and valves opened).
MINIMUM CRITICAL The value of critical power ratio associated with the most limiting l
l POWER RATIO (MCPR) assembly in the reactor core. Critical Power Ratio (CPR) is the 1
l ratio of that power in a fuel assembly, which is calculated to I
cause some point in the assembly to experience boiling transition, to the actual assembly operating power.
MODE The reactor MODE is that which is established by the mode-selector-switch. The MODES include:
Startuo MODE j
in this MODE the reactor protection scram trip, initiated by main steam line isolation valve closure, is bypassed when reactor pressure is less than 600 psig, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with IRM neutron monitoring system trips and control rod withdrawal interlocks in service.
Run MODE in this MODE the reactor system pressure is at or above 785 psig and the reactor protection system is energized with APRM protection and RBM interlocks in service.
Shutdown MODE The reactor is in the shutdown MODE when the reactor mode switch is in the shutdown mode position and no core alterations are being performed.
- a. Hot Shutdown means conditions as above with reactor coolant temperature greater than 212*F.
- b. Cold Shutdown means conditions as above with reactor coolant temperature eoual to or less than 212*F.
Refuel MODE The reactor is in the refuel MODE when the mode switch is in the refuel mode position. When the mode switch is in the refuel l
position, the refueling interlocks are in service.
PNPS 1-3 Amendment No.177 l l
'=
-
- 1.0 DEFINITIONS (Cont)
OPERABLE-A system, subsystem, division, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other l
auxiliary equipment that are required for the system, subsystem, j
division, component, or device to perform its specified safety l
function (s) are also capable of performing their related support function (s).
OPERATING OPERATING means that a system or component is performing its intended functions in its required manner.
i OPERATING CYCLE Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY means that the drywell INTEGRITY and pressure suppression chamber are intact and all of the following conditions are satisfied:
1.
All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.
2.
At least one door in each airiock is closed and sealed 3.
All blind flanges and manways are closed.
4.
All automatic primary containment isolation valves and all l
instrument line check valves are operable or at least one containment isolation valve in each line having an inoperable valve shall be deactivated in the isolated condition.
5.
A11 containment isolation check valves are operable or at least one containment valve in each line having an inoperable valve is secured in the isolated position.
PRCTECTIVE ACTION An action initiated by the protection system when a limit is reached. A PROTECTIVE ACTION can be at a channel or 3
system level.
PROTECTIVE FUNCTION A system PROTECTIVE ACTION which results from the i
PROTECTIVE ACTION of the channels monitoring a particular 1
plant condition.
i REACTOR POWER REACTOR POWER OPERATION is any operation with the mode OPERATION switch in the "Startup" or "Run" position with the reactor critical and above 1% design power.
REACTOR VESSEL Unless otherwise indicated, REACTOR VESSEL PRESSURES PRESSURE listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.
PNPS 1-4 Amendment No. I77 l
s 1.0 DEFINITIONS (continued) 4 REFUELING INTERVAL REFUELING INTERVAL applies only to ASME Code,Section XI f
lWP and IWV surveillance tests. For the purpose of designating frequency of these code tests, a REFUELING INTERVAL shall mun at least once every 24 months.
REFUELING OUTAGE REFUELING OUTAGE is the period of time between the shutdown of the unit prior to a refueling and the startup of the l
plant after that refueling. For (ne purpose of designating frequency of testing and surveillance, a REFUELING OUTAGE i
shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous REFUELING OUTAGE, the required surveillance testing need not be performed until the next regularly scheduled outage.
SAFETY LIMIT The SAFETY LIMITS are limits below which the reasonable maintenance of the cladding and primary systems are assured.
i Exceeding such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences, but it indicates an operational l
deficiency subject to regulatory review.
SECONDARY SECONDARY CONTAINMENT INTEGRITY means that the CONTAINMENT reactor building is intact and the following conditions are met:
INTEGRITY
- 1. At least one door in each access opening is closed.
- 2. The standby gat treatment system is operable.
- 3. All automatic ventilation system isolation valves are operable or secured in the isolated position.
SIMULATED AUTOMATIC SIMULATED AUTOMATIC ACTUATION means applying a ACTUATION simulated signal to the sensor to actuate the circuit in quest;on.
l SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
STAGGERED TEST A STAGGERED TEST BASIS shall consist of: (a) a test BASIS schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into a equal subintervals; (b) the testing of one system, subsyste.m, train or other designated components at the beginning of each subinterval.
l SURVEILLANCE Each Surveillance Requirement shall be performed within the FREQUENCY specified SURVEILI.ANCE INTERVAL with a maximum allowable extension not to exceed 25 percent of the specified SURVEILLANCE INTERVAL.
The SURVEILLANCE FREQUENCY establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and PNPS 1-5 Amendment No.177 l
I t
1.0 DEFINITIONS (Cent)
SURVEILLANCE consideration of plant operating conditions that may not be FREQUENCY (continued) suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for L
surveillance that are not performed during refueling outages.
This limitatien of this definition is based on engineering judgment and the recognition that the most probable result of any particular surv6tilance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.
SURVEILt.ANCE The SURVEILLANCE INTERVAL is the calendar time between INTERVAL surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the instrument, component, or system is not required to be operable, but the instrument, component, or system shall be tested prior to being declared operable. The operating cycle intervalis 24 months and the 25% tolerance of the definition of " SURVEILLANCE I
FREQUENCY"is applicable. The refueling intervalis 24 months I
and the 25% tolerance specified in the definition of "SURVEILt.ANCE FREQUENCY"is applicable.
TOTAL PEAKING The ratio of the fuel rod surface heat flux to the heat flux of an FACTOR average rod in an identical geometry fuel assembly operating at the core average bundle power.
l TRANSITION BOILING TRANSITION BOILING means the boiling regime between l.
nucleate and film boiling. TRANSITION BOILING is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.
L TRIP SYSTEM A TRIP SYSTEM means an arrarsment of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protecLve trip function. A TRIP SYSTEM may require one or more instrument channel trip signais related to one or more plant parameters in order to initiate trip system action.
initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.
{
PNPS 1-6 Amendment No.177 l l
1
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 Pl. ANT SYSTEMS 4.8 PLANT SYSTEMS 1.
, Main Condenser Offaas 1.
Main Condenser Offaas LCO 3.8.1 1.
NOTE Not Required to be performed The gross gamma activity rate of noble until 31 days after any SJAE in gases measured at a main condenser operation.
retreatment monitor station shall be limited to 500,000 p Ci/second.
APPLICABILITY:
Verify the gross gamma activity rate of the noble gases is At all times when steam is available to s 500,000 Ci/second:
the air ejectors.
a.
At least once per 31 days.
AND ACT M b.
Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a 50% increase in the A.
With the gross gamma activity rate of the noble gases not within limits; n minal steady state fission gas release after factoring ut increase due to
- 1. Restore the gross gamma activity rate of the noble gases changes in THERMAL to within the limit within 72 POWER level or hydrogen hours.
- I B.
Required Action and associated Completion Time not met.
1.
Isolate SJAE within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
98 2.1 Be in HOT SHUTCOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
bHQ 2.2 Be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
PNPS 3/4.8-1 Amendment No.177 l j
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS o
3.8 PLANT SYSTEMS (CONT) 4.8 PLANT SYSTEMS (CONT) 2.
Mechanical Vacuum Pumo isolation 2.
Mechanical Vacuum Pumo isolation instrumentation Instrumentation LCO 3.8.2 NOTE When a channelis placed in an Four channels of the Main Steam Line inoperable status solely for the Radiation Monitoring System Radiation -
performance of required High function for the mechanical vacuum Surveillance, entry into associated pump shall be OPERABLE Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> APPLICABILITY:
provided the associated Function maintains mechanical vacuum Whenever any main steam isolation pump isolation capacity.
valve is open with steam flowing..
ACTIONS:
NOTE 1.
Perform a CHANNEL CHECK Separate Condition Entry is allowed for every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
{
each channel.
2.
Calibrate the trip units every A.
One or more required channels 92 days.
In Pe W e.
3.
Perform a CHANNEL
- 1. Restore channel to OPERABLE CALIBRATION every 24 i
status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, months. The allowable trip g
value shall be s 5.5 x normal background.
2.
NOTE Not Applicable if inoperable 4.
Perform a LOGIC SYSTEM channelis the result of an FUNCTIONAL TEST including inoperable isolation valve.
isolation valve actuation every 24 months.
l Place channel or associated trip system in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Required Action and associated Completion Time of Condition A not met.
Mechanical vacuum pump isolation capability not maintained.
- 1. Isolate mechanical vacuum within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 2. Isolate Main Steam Lines within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 3. Be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, s
I PNPS 3/4.8-2 Amendment No.177 l
B 3/4.8 PLANT SYSTEMS B 3/4.8.1 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the condenser. Air and noncondensible gases are cc'lected in the condenser, then exhausted through the steam jet air ejectors (SJAEs) to the augmented offgas system. The offgas from the main condenser normally includes radioactive gases.
)
Restricting the gross gamma activity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to a member of the public at and beyond the site boundary will not exceed a small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.
The augmented offgas system has been incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radiolytically dissociated l
hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and condensables are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is normeily measured downstream of the moisture separator prior to entering the holdup line, In the event that the augmented offgas system is not in service, the activity rate of noble gases can be measured at the steam jet air ejector offgas sampling system.
APPLICABLE The main condenser offgas gross gamma activity rate is an initial SAFETY ANALYSES condition of the main condenser offgas system fa!!ure event, discussed in the FSAR, Section 14.5.6 "Radwaste System Accidents",
I (Ref.1). This section analyzes the effects, which would result from the failure of the offgas system piping. The gross gamma activity rate is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits of 10 CFR 100 (Ref. 2).
The main condenser offgas limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
(continued)
PNPS B3/4.8-1 Amendment No.177 l
Main Cond2ns2r Offgas 3/4.8-1 BASES LCO.
To ensure compliance with the assumptions of the main condenser offgas system failure event (Ref.1), the fission product release rate -
prior to entering the treatment, adsorption, and delay systems should provide reasonable assurance that the potential total body accident dose to an individual at the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR 100. The limit for gross gamma activity rate of noble gases was derived using the guidance provided in NUREG 0133 (Ref. 3) and documented in reference 4.
APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensables are being processed via the main condenser offgas system.
I I
ACTIONS
.A_d if the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore the gross gamma activity rate to within the limit. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins associated with permissible dose and exposure limits, and the low probability of a main condenser offgas system rupture.
B.1. B.2.1. and B.2.2 If the gross gamma activity rate is not restored to within the limits in the associated Completion Time, the SJAE must be isolated. This isolates the main condenser offgas system from the source of the radioactive steam. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderiy manner and without challenging unit systems.
An attemative to Required Actions B.1 is to place the unit in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in l
COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
1 (continued)
PNPS B3/4.8-2 Amendment No.177 l
MIin Cond:ns:r Offgts 3/4.8-1 BASES SURVEILLANCE SR 4.8.1.1 REQWREMENTS This SR, on a 31 day Frequency, requires an isotopic analysis of an offgas sample to ensure that the required limits are satisfied. The noble gases to be sampled are Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, and Kr-88. These are the isotopes important for monitoring fuelintegrity as recommended by the fuel supplier. If the measured rate of radioactivity increases significantly (by 2 50% after correcting i
for expected increases due to changes in THERMAL POWER or hydrogen injection), an isotopic analysis is also performed within 4 1
hours after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the fission gas release rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on operating experience.
This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after the SJAE is in operation. Only in this
.. condition can radioactive fission gases be in the main condenser offgas system at significant rates.
REFERENCES 1.
FSAR, Section 14.5.6 2.
10 CFR Pad 100 3.
NUREG 0133 4.
Calculation # PNPS-1-ERHS-XII.D-8 PNPS B3/4.8-3 Amendment No.177
i Mechtnical VEcuum Pump Isol: tion Instrum2ntation 3/4.8-2 BASES B 3/4.8 PLANT SYSTEMS l
B 3/4.8.2 Mechanical Vacuum Pump Isolation Instrumentation BASES l
['
BACKGROUND The mechanichi vacuum pump isolation instrumentation initiates a trip of the mechanbl vacuum pump and isolation of the associated isolation valvt following events in which main steam radiation exceeds precstermined values. Tripping and isolating the mechanical vacuum pump limits the offsite doses in the event of a control rod drop accident (CRDA).
I The mechanical vacuum pump isolation instrumentation (Ref.1) includes sensors, relays, and switches that are necessary to cause initiation of a mechanical vacuum pump isolation. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the l
channel output relay actuates, which then outputs an isolation signal to the mechanical vacuum pump isolation logic.
The isolation logic consists of two independent trip systems, with two channels of Main Steam Tunnel Radiation-High in each trip system.
Each trip system is a one-out-of-two logic for this Function. Thus, either channel of Main Steam Tunnel Radiation-High in each trip system are needed to trip a trip system. The outputs of the channels in a trip system are combined in a logic so that both trip systems must trip to result in an isolation signal, There is one isolation valve associated with this function.
t l
APPLICABLE The mechanical vacuum pump isolation is assumed in the safety SAFETY ANALYSES analysis for the CRDA. The mechanical vacuum pump isolation instrumentation initiates an isolation of the mechanical vacuum pump to limit offsite doses resulting from fuel cladding failure in a CRDA (Ref. 2).
The mechanical vacuum pump isolation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
l (continued)
PNPS B3/4.8-4 Amendment No.177 l
Mrchinical Vacuum Pump Isolation Instrum2nti. tion 3/4.8-2 BASES LCO-The OPERABILITY of the mechanical vacuum pump isolation is dependent on the OPERABILITY of the individual Main Steam Line Radiation-High instrumentation channels, which must have a required j
l number of OPERABLE channels in each trip system, with their j
setpoints within the specified Allowable Value of SR 4.8.2.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated isolation valve.
Allowable Values are specified for the mechanical vacuum pump isolation Function specified in the LCO. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between Channel CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place.
The setpoints are compared to the actual process parameter (i.e.,
Main Steam Line Radiation-High), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
APPLICABILITY The mechanical vacuum pump isolation is required to be OPERABLE only if a main steam line is unisolated with steam flowing. If the main steam lines are isolated, the monitors could not detect high radiation due to the monitors location downstream of the main steam isolation valves.
ACTIONS A Note has been provided to modify the ACTIONS related to mechanical vacuum pump isolation instrumentation channels. The Required Actions for inoperable mechanical vacuum pump isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable mechanical vacuum pump isolation instrumentation channel.
(continued)
PNPS B3/4.8-5 Amendment No.177 l
4 Machinical Vacuum Pump Isolttion Instrum ntation BASES 3/4.8-2 ACTIONS A.1 and A.2 (continued).
With one or more channels inoperable, but with mechanical vacuum pump isolation capability maintained (refer to Required Actions B.1, 8.2, and B.3 Bases), the mechanical vacuum pump isolation instrumentation is capable of performing the intended function.
However, the reliability and redundancy of the mechanical i
vacuumpump isolation instrumentation is reduced such that a single failure in one of the remaining channels could result in the inability of the mechanical vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore i
the inoperable channels to OPERABLE status. Because of the low probability of extensive numbers of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of mechanical vacuum pump isolation,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has been shown to be j
{
acceptable to permit restoration of any inoperable channel to OPERABLE status. (Required Action A.1). Altemately, the inoperable channel or associated trip system may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable isolation valve, since this may not adequately compensate for the inoperable valve (e.g., the valve may be inoperable such that it will not isolaf ' If it is not desired to place the channel in trip (e.g., as in the case wi.are placing the inoperable channel would result in loss of mechanical vacuum), or if the inoperable channel is the result of an inoperable valve, Condition B must be entered and its Required Actions taken.
B.1. B.2. and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this
)
status, the plant must be brought to at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.3). Attemately, the associated mechanical vacuum pump may be removed from service since this performs the intended function of the instrumentation (Required Action j
B.1). An additional option is provided to isolate the main steam lines j
(Required Action B.2), which may allow operation to continue,
{
l Isolating the main steam lines effectively provides an equivalent level
{
of protection by precluding fission product transport to the condenser.
{
s Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the Function not maintaining mechanical vacuum pump isolation capability. The i
Function is considered to be maintaining mechanical vacuum pump I
isolation capability when sufficient channels are OPERABLE or in trip such that the mechanical vacuum pump isolation instruments will (continued)
PNPS B3/4.8-6 Amendment No.177 l
M:chinical Vecuum Pump IsoIItion InstrumInt: tion 3/4.8-2 BASES ACTIONS B.1. B.2. and B.3 (continued) generate a trip signal from a valid Main Steam Line-High signal, and the isolation valve will close. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the mechanical J
vacuum pump isolation valve to be OPERABLE.
The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach HOT SHUTDOWN from full power j
conditions, or to remove the condenser pump from service, or to isolate the main steam lines in an orderly manner and without challenging plant systems.
SURVEILLANCE The Surveillance are modified by a Note to indicate that when a REQUIREMENTS channel is placed in an inoperable status solely for performance of required Surveillance, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, provided the associated Function maintains mechanical vacuum pump isolation trip l
capability. Upon completion of the Surveillance or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be retumed to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform a channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does not significantly reduce the probability that the mechanical vacuum pumps willisolate when necessary.
SR 4.8.2.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive f
instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
l l
Agreement criteria are determined by the plant staff based on a combination of the channelinstrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
l The Frequency is based upon operating experience that demonstrates i
channel failure is rare. The CHANNEL CHECK supplements less (continued)
PNPS B3/4.8-7 Amendment No.177 l t
Mschinical VEcuum Pump Isolition InstrumIntition 3/4.8-2 BASES l
[
SURVEILLANCE SR 4.8.2.1 (continued)
REQUIREMENTS formal, but more frequent checks of channels during normal operational use of the displays associated with the required channels of this LCO.
SR 4.8.2.2 Calibration of trip units provides a check of the actual trip setpoints.
The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.8.2.3. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
The Frequency of 92 days is based on the reliabilit/ analysis of Reference 3.
A CHANNEL CALIBRATION is a complete check of the instrument I
loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations, consistent with the plant specific setpoint methodology.
The Frequency is based upon the assumption of a 24 month calibration intervalin the determination of the magnitude of equipment drift in the setpoint analysis.
SR 4.8.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overiaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function.
Therefore, if a breaker is incapable of operating, the associated l
instrument channel (s) would be inoperable.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
(continued) l PNPS B3/4.8-8 Amendment No.177 l
Mechrnical VEcuum Pump IsolLtion InstrumIntatian 3/4.8-2 BASES REFERENCES 1.
FSAR Section 7.12.1 2.
FSAR Section 14.5.1 3.
BECo Calculation I-N1-104, Review Setpoint Calculation for RM-1705-2A,2B,2C, and 2D - Main Steam Line High Radiation".
l 1
I l
i (continued)
PNPS B3/4.8-9 Amendment No.177 l
.~
Design Fe'_turzs 4.0 4.0. DESIGN FEATURES 4.1 Site Location Pilgrim Nuclear Power Station is located on the westem shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts and contains approximately 517 acres owned by Boston Edison Company as shown on FSAR Figures 2.2-1 and 2.2-2. The site boundary is posted and a perimeter security fence provides a distinct i
security boundary for the protected area of the station.
l The reactor (center line) is located approximately 1800 feet from the nearest property boundary.
i 4.2 Reactor Core The reactor vessel core design shall be as described in the CORE OPERATING LIMITS REPORT and shall be limited to those fuel assemblies which have been analyzed with NRC approved codes and methods and approved by the NRC in its acceptance of Amendment 22 of GESTAR 11.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum k-infinity of 1.32 for standard core geometry, calculated at the bumup of maximum bundle reactivity, and an average U 235 enrichment of 4.6 %
averaged over the axial planar zone of highest average enrichment; and b.
K., s 0.95 if fully flooded with unborated water, tyhich includes an allowance for uncertainties as described in Section 10.3.5 of the FSAR.
(continued)
PNPS 4.0-1 Ainendment No.177 l
l
Design Fetures 4.0 4.0 bESIGN FEATURES l
4.3 FuelStorage (continued) l 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
l a.
K, s0.95 if fully flooded with water, which includes an l
allowanco for uncertainties as described in Section 10.2.5 of
(
the FSAR; b.
K, s0.90 when dry, which includes an allowance for uncertainties as described in Section 10.2.5 of the FSAR; and c.
A nominal 6.60 inch center to center distance between fuel assemblies placed in storage racks.
4.3.2 Droinaae The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 115 ft.
4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3859 fuel assemblies.
4.3.4 Heavy Loads i
a.
Loads in excess of 2000 lb. shall be prohibited from travel over fuel assemblies in the spent fuel storage pool.
i b.
No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the energy absorbing pad.
L l
PNPS TS 4.0-2 Amendment No. I77 l
i Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The Station Director shall be responsible i:. ove'all unit operation and shall delegate in writing the succession to this respe.tibility during his absence.
The Station Director or his designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that l
affect nuclear safety.
j 5.1.2 The Nuclear Operations Supervisor (NOS) shall be responsible for the control f
roc,=1 command function. During any absence of the NOS from the control j
room while the unit is in an operational mode other than Cold Shutdown or i
Refueling, an individual with an active Senior Reactor Operator (SRO) license
)
shall be designated to assume the control room command function. During j
any absence of the NOS from the control room while the unit is in Cold Shutdown or Refueling, an individual with an active SRO license or Reactor Operator (RO) license shall be designated to assume the control room l
command function.
l PNPS 5.0-1 Amendment No.177 l
Organization -
5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall ine!ade the positions for activities affecting safety of the nuclear power plant.
a.
Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be
]
documented in the Pilgrim Station Final Safety Analysis Report 1
(FSAR);
l b,
The Station Director shall be responsible for overa'l safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant; c.
The Senior Vice President-Nuclear shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and d.
The individuals who train the operating staff, carry out health physics, or perform quality assi'rance functions may report to the appropriate i
onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures 3
1 5.2.2 Unit Staff The unit staff organization shall include the following:
l l
a.
A non-licensed operator shall be on site when fuel is in the reactor and an additional non-licensed operator shall be assigned when the reactor is in an operational mode other than Cold Shutdown or Refueling.
(continued)
PNPS 5.0-2 Amendment No.177 l
Org:niz tion 5.2 5.2 Organization -
5.2.2 Unit Staff (continued) b.
At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor, in addition, while the unit is in an operational mode other than Cold Shutdown or Refueling, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.
At least two licensed ROs shall be present in the control room during c.
reactor startup, scheduled reactor shutdown and during recovery from reactor trips.
d, Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.i for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
l Higher grade licensed operators may take the place of lower grade e.
- 1. censed or unlicensed personnel.
f.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position, g.
The amount of overtime worked by unit staff members performing safety-relatea functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12).
h.
. Tne Operations Department Manager, Nuclear Watch Engineers, and Nuclear Operations Supervisors shall hold a Senior Reactor Operator License. The Nuclear Plant Operators sha!! hold a Reactor Operatcr License.
i.
The Shift Control Room Engineer (SCRE) shall provide advisory technical support to the Nucbar Operations Supervisor (NOS) in the areas of engineering and accident astesament. In addition, the SCRE shall meet the qualifications specified m' the Commission Policy i
Statemer' J,n Engineering Expertise on Shift. A Shift Control Room l
Engineer with a Senior Reactor Operator license may simultaneously serve as SCRE and SRO.
PNPS 5.0-3 Amendment No.177 l
e.
Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications as described in the American National Standards institute N18.1-1971,
" Selection and Training of Personnel for Nuclear Power Plants." In addition, the individual performing the function of Radiation Protection Manager shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, l
PNPS 5.0-4 Amendment No.177 l
a Procedurss 5.4 5.0 ADMINISTRATIVE CONTROLS i
5.4 Procedures l
l 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:
a.
The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; b.
The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; c.
Quality assurance for effluent and environmental monitoring; d.
Fire Protection Program implementation; and e.
All programs specified in Specification 5.5.
I PNPS 5.0-5 Amendment No.177 l
Progrcms cnd Minu:Is 5.5 5.0 A' ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented and maintained.
1 5.5.1 Offsite Dose Calculation Manual (ODCM) a.
The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and b.
The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluer,t Release, reports l
required by Specification 5.6.2 and Specification 5.6.3.
Licensee initiated changes to the ODCM:
a.
Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
1.
sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying the change (s), and I
2.
a determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix 1, and not adverc91y impact the accuracy or reliability of effluent, dose, or setpoint calculations; 1
b.
Shall become effective after review and acceptance by the Operations l
Review Committee and the approval of the Chemistry and Radiological l
Department Managers; and Shall be submitted to the NRC in the form of a complete, legible copy c.
of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any l
change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the j
area of the page that was changed, and shallindicate the date (i.e.,
month and year) the change was implemented.
(continued)
PNPS 5.0-6 Amendment No.177 l
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Programs End M:nuals 5.5 5.5 Programs and Manuals 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include the Core Spray (CS), High Pressure Coolant injection (HPCI), Residual Heat Removal (RHR), Reactor Core isolation Cooling (RCIC), Reactor Water Cleanup (RWCU)[Let Down portion), Radwaste Collection System from Reactor Building, sampling system (From Recirc and RWCU), and Standby Gas Treatment (SGTS). The program shallinclude the following:
a.
Preventive maintenance; and b.
Periodic visual inspection to identify and estimate leakage.
5.5.3 Post Accident Samolina This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulate in plant gaseous eifluents and containment atmosphere samples under accident conditions. The program shall include the following:
a.
Training of personnel; b.
Procedures for sampling and analysis; and Provisions for maintenance of sampling and analysis equipment.
c.
5.5.4 Radioactive Effluent Controls Prooram This program conforms to 10 CFR 50.36a for the control of radioactive affluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shallinclude the following elements:
a.
Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; (continued)
PNPS 5.0-7 Amendment No.177 l
e Programs End Minurls 5.5 5.5 Programs and Manuals 5.5A Radioactive Effluent Controls Proaram (continued) b.
Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 CFR 20, Appendix B, Table 2, Column 2; c.
Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM; d.
Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents releasad to unrestricted areas, conforming to 10 CFR 50, Appendix I; e.
Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM st least every 31 days; f.
Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce roleates of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix 1; g.
Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the following:
1.
For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and 2.
For lodine-131, lodine-133, Tritium, and all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.
h.
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the site boundary, conforming to 10 CFR 50, Appendix 1; (continued)
PNPS 5.0-8 Amendment No.177 l t
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ProgrIms cnd Minu Is 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Proaram (continued) 1.
Limitations on the annual and quarterly doses to a member of the public from lodine-131, lodine-133, Tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and j.
Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
I 5.5.5 Component Cvelic or Transient Limit This program provides controls to track the FSAR Section C.3.4.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.
1 5.5.6 Technical Specifications (TS) Bases Control Procram This program provides a means for processing changes to the Bases of these Technical Specifications, Changes to the Bases of the TS shall be made under appropriate a.
administrative controls and reviews.
I b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
1.
a change in the TS incorporated in the license; or 2.
a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
d.
Proposed changes that meet the criteria of Specification 5.5.6b above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall l
be provided to the NRC on a frequency consistent with
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PNPS 5.0.9 Amendment No. I 77 l a
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Reporting Requirements 5.6 5.0 dDMINI'STRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
5.6.1 Occupational Radiation Exposure Report A tabulation on an annual basis of the number of stetion, utility, and other personnel (including contractors) receiving exposures > 100 mrem /yr and their associated man rem exposure according to work and job functions (e.g.,
reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (including description), waste processing, and refueling).
This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter, thennoluminescent dosimeter (TLD), or film badge measurements.
Small exposures totaling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. The report shall be submitted by April 30 of each year.
5.6.2 Annual Radiological Environmental Operatino Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shallinclude summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM),
and in 10 CFR 50, Appendix 1, Sections IV.B.2, IV.B.3, and IV.C.
l The Annual Radiological Environmental Operating Report shall include a I
summary of the results of analyses of all radiological environmental samples i
and of all environmental radiation measurements taken during the period i
pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report sha9 be submitted noting and explaining the reasons for the missing results.
j The missing data shall be submitted in a supplementary report as coon as
)
possible.
t (continued) l PNPS 5.0-10 Amendment No.177 l
Riporting RIquirim nts 5.6 5.6 Reporting Requirements 5.6.3 Radioactive Effluent Release Reoort The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a by May 15th of each year. The report shallinclude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and process control procedures and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix 1,Section IV.B.1.
5.6.4 Monthly Operatina Reports Routine reports of operating statistics and shutdown experience shall be i
submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
5.6.5 Core Operatino Limits Report (COLR)
Core operating limits shall be established prior to each reload cycle, or
(
a.
prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1.
Table 3.1.1 - APRM High Flux trip level setting
' 2.
Table 3.2.C -APRM Upscale trip level setting 3.
3.11.A - Average Planar Linear Heat Generation Rate (APLHGR) 4.
3.11.8 - Linear Heat Generation Rate (LHGR) 5.
3.11.C -Minimum Critical Power Ratio (MCPR) 6.
3.11.D - Power / Flow Relationship During Power Operation 7.
4.2 - Reactor Core b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
NEDE 24011 P-A," General Electric Standard Application for Reactor Fuel,"(the approved version at the time the reload analyses are performed shall be identified in the COLR).
1 (continued)
PNPS 5.0-11 Amendment No.177 l l
Rrporting R:quircmsnts 5.6
,5.6 Reporting Requirements 5.6.5 (continued) 2.
NEDC-31852P, " Pilgrim Nuclear Power Statinn SAFER /GESTR-LOCA Loss of Coolant Accident Analysis", dated Geptember, 1990 (the approved version at the time the reload analyses are performed shall be identified in the COLR), and 3.
NEDC-31312-P," ARTS Improvement Program Analyses for Pilgrim Nuclear Power Station", dated September 1987, (the approved version at the time the reload analyses are performed shall be identified in the COLR).
The core operating limits shall be determined such that all applicable c.
limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
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l PNPS 5.0-12 Amendment No.177 l L
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High Radiation Area 5.7 5.0 ADMINt'STRATIVE CONTROLS 5.7 High Radiation Area 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem /hr but < 1000 mrem /hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics personnel) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates 5; 1000 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
J Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a.
A radiation monitoring device that continuously indicates the radiation dose rate in the area.
b.
A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is j
received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them, l
c.
An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform l
periodic radiation surveillance at the frequency specified by the
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Radiation Protection Manager in the RWP.
5.7.2 in addition to the requirements of Specification 5.7.1, areas with radiation levels 21000 mrem /hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Nuclear Watch Engineer on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an appioved RWP that shall specify the dose rate levels in the I
immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by
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(continued)
PNPS 5.0-13 Amendment No.177 l
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5.7 5.7 Nigh Radiation Area 5.7.2 (continued) personnel qualified in radiation protedion procedures to provide positive exposure control over the activities being performed within the area.
5.7.3 For individual high radiation areas with radiation levels of > 1000 mrem /hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.
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PNPS 5.0-14 Amendment No.177 l