ML18038A232
| ML18038A232 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/30/1986 |
| From: | Mangan C NIAGARA MOHAWK POWER CORP. |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| NMP1L-0122, NMP1L-122, NUDOCS 8701050171 | |
| Download: ML18038A232 (839) | |
Text
REGULATORY 1 ORMAT ION DISTRIBUTION SYS (R IDS >
ACCESSION NBR: 870i050i7i DOC. DATE: 86/i2/30 NOTARIZED: NO DOCKET 0 FACIL: 50-220 Nine Mile Point Nuclear Station> Unit i> Niagara Poue 05000220 AUTH. NAl lE AUTHOR AFFILIATION MANQAN> C. V. Niagav a Mohawk Power Corp.
RECIP. NAME RECIPIENT AFFILIATION ZWOLINSKI> J. A. BWR Pv object Div ectorate i
SUBJECT:
Fovwards corv ected Tech Spece> consisting of typo 5 page numbev'orv ections. Retyped version v evieeed against Amend 89 approved on 86iii7. Redv a ing of f igures unnecessav g.
DISTRIBUTION COD : 'iD COPIE ECEIVED: LTR ENCL SIZE:
TITLE: QR Submi tal: General Distribution NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID 'CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL BWR EB BWR EICSB BWR FQB BWR PDi LA BWR PDi PD Oi KEl LY> J BWR PSB g 8WR RSB INTERNAL: ADl1/LFMB ELD/HDS3 NRR/DHFT/TSCB 04 i ji+ NRR/GRAS RQNi EXTERNAL'Q8(G BRUSKE> S LPDR 03 NRC PDR 02 i + NSIC 05 yZ"~'
TOTAL NUMBER OF COPIES REQUIRED: LTTR 23 ENCL
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T NIAGARA N ~ MOHAWK NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511 December 30, 1986 NMPlL 0122 Director of Nuclear Reactor Regulation Attention: Mr. John A. Zwolinski, Project Director BWR Project Directorate Number 1 Division of BWR Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Nine Mile Point Unit 1 Docket No. 50-220 DPR-63
Dear Mr. Zwolinski:
As requested in your letter dated November 4, 1986, enclosed please find corrections to the retyped Nine Mile Point Unit Technical Specifications. 1 These corrections consist of minor typographical errors and page number corrections. In addition, the retyped version has been reviewed against Amendment No. 89, approved November 17, 1986. We have not included a schedule for redrawing any figures. Niagara Mohawk believes that the existing figures are sufficiently legible for use in the retyped Technical Specifications. If future proposed amendments require the revision of any figures, we will consider redrawing them at that time. Sincerely, NIAGARA MOHAWK POWER CORPORATION C. V. Manga Senior Vice President KBT/pns 2365G Enclosure 8701050171 861230 PDR ADQCK 05000220
RADIOLOGICAL TECHNICAL SPECIFICATIONS APPENDIX A TO FACILITY OPERATING LICENSE NO. DPR-63 FOR THE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION UNIT 1 DOCKET NO. 50-220 DECEMBER 26, 1974
FOREWORD These revised specifications supersede in their entirety the previous technical specifications and are issued as Appendix A to Full-Term Operating License DPR-63 issued to Niagara Mohawk Power Corporation by the Atomic Energy Commission. The Environmental Technical Specifications are issued as Appendix B to License DPR-63. Nine Mile Point - Unit 1
NINE MILE POINT NUCLEAR STATION UNIT 1 - TECHNICAL SPECIFICATIONS CONTENTS SECTION DESCRIPTION PAGE 1.0 Definitions
- 2. 0 Safety Limits and Limiting Safety System Setting Safet Limits Limitin Safet Settin
- 2. 1.
2.2. 1,'el C Reactor adding Integrity 2.1.2 Fuel Cladding Integrity S stem 1 Coolant System 2.2.2 Reactor Coolant System 2-7 g.o S9sL~$ 3'~ g 82-l 3.0 Limiting Condition for Operation and Surveillance Requi rements 3/4 0-1
- 3. 1.0 Fidel Cladding 3/4 1-0 Limitin Condition for 0 eration Surveillance Re uirements 3.1.1 Control Rod System 4.1.1 Control Rod System 3/4 1-1 3.1.2 Liquid Poison System 4.1.2 Liquid Poison System 3/4 1-8 3.1.3 Emergency Cooling System 4.1.3 Emergency Cooling System 3/4 1-13 3.1.4 Core Spray System 4. 1.4 Core Spray System 3/4 1-15 3.1.5 Solenoid-Actuated Pressure 4.1.5 Solenoid-Actuated Pressure Relief Valve Relief Valve 3/4 1-20 3.1.6 Control Rod Drive Coolant Injection 4.1. 6 Control Rod Drive Coolant Injection 3/4 1-22 3.1.7 Fuel Rods 4.1. 7 Fuel Rods 3/4 1-24 3.1.8 High Pressure Coolant Injection 4.1.8 High Pressure Coolant Injection 3/4 1 3. 2. 0 Reactor Coolant System 3/4 2-.1-Limitin Condition for 0 eration Surveillance Re uirements 3.2.1 Reactor Vessel Heatup and Cooldown Rates 3/4 2-2 3.2.2 Minimum Reactor Vessel Temperature 4.2.2 Minimum Reactor Vessel Temperature for Pressurization for Pressurization 3/4 2-3 3.2.3 3.2.4 Coolant Chemistry Coolant Activity
- 4. 2. 3 Coolant Chemistry 3/4.2~ (/
- 4. 2.4 Coolant Activity 3/4 2-% ]~
t4o. 6 Nine Mile Point - Unit 1 1
SECTION DESCRIPTION PAGE 3.2.5 Reactor Coolant System Leakage 4.2.5 Reactor Coolant System Leakage 3/4 2-16-~4 3.2.6 In-.Service Inspection and Testing 4.2.6 In-Service Inspection and Testing 3/4 2-MIG 3.2.7 Isolation Valves 4.2.7 Isolation Valves 3/4 2-MI8
- 3. 2.8 Pressure Relief Systems-Safety Valves 4.2.8 Pressure Relief Systems-Safety Valves 3/4 2-&qQ 3.2.9 Pressure Relief Systems - Solenoid- 4.2.9 Pressure Relief Systems - Solenoid-Actuated Pressure Relief Valves Actuated Pressure Relief Valves 3/4 2-~ZS 3.3.0 Primary Containment 3/4 3-1 Limstin Condition for 0 eration Survei llance Re uirements 3.3. 1 Oxygen Concentration 4.3.1 Oxygen Concentrati on 3/4 3-2
- 3. 3. 2 Pressure and Suppression Chamber 4.3.2 Pressure and Suppression Chamber Water Water Temperature and Level Temperature and Level 3/4 3-4 3.3.3 Leakage Rate 4.3.3 Leakage Rate 3/4 3-3~
3.3.4 Isolation Valves 4.3.4 Isolation Valves 3/4 3-ZS-i)
- 3. 3. 5 Access Control 4.3.5 Access Control 3/4 3-M~
3.3.6 Vacuum Relief 4.3. 6 Vacuum Relief 3/4 3-M>I 3.3.7 Containment Spray 4.3.7 Containment Spray 3/4 3-26qg 3.4.0 Secondary Containment 3/4 4-1 Limitin Condition for 0 eration Survei llance Re uirements 3.4.1 Leakage Rate 4.4.1 Leakage Rate 3/4 4-2 3.4.2 .Isolation Valves 4.4. 2 Isolation Valves 3/4 4-4 3.4.3 Access Control 4.4.3 Access Control 3/4 4-5 3.4.4 Emergency Ventilation 4 4 Emergency Venti lation 3/4 4-6 3.4.5 Control Room Ventilation 4.4.5 Control Room Ventilation 3/4 4-9 3.5.0 Shutdown and Refueling 3/4 5-1 Limitin Condition for 0 eration Surveillance Re uirements 3.5. 1 Source Range Monitoring 4.5.1 Source Range Monitoring 3/4 5-2 3.5.2 Refueling Platform Inter lock 4. 5. 2 Refueling Platform Interlock 3/4 5-4 3.5.3 Extended Core and ControlCl~oa d5ocJ 4.5.3 Extended Core and Control Road Drive Maintenance Drive Maintenance 3/4 5-6 Amendment No. '76 Nine Mile Point - Unit 1
SECTION DESCRIPTION PAGE 3.6.0 General Reactor Plant 3/4 6-1 Limitin Condition for 0 eration Surveillance Re uirements 3.6.1 Station Process Effluents 4.6.1 Station Process Effluents 3/4 6-2 g.v,~ 3.6.2 3.6.3 Protective Instrumentation Emergency Power'Sources ~
>~~"<<"f><g
~< 4.6.2 s) 4. 6. 3 Protective Instrumentation Emergency Power Sources 3/4 3/4 6-3 6-%,P 3.6.5 Radioactive Haterial Sources . ~4. 6. 5 Radioactive Material Sources 3/4 59~3)Q C lI 3.6.6 3.6.7 Fire Detection g4.g 4.6.6 Fire Detection 3/4 6-79c,y 4,
Fire Suppression 4.6.7 Fire Suppression 3/4 6-+ qg 3.6.8 Carbon Dioxide Suppression System 4.6.8 Carbon Dioxide Suppression System 3/4 6-84'~ 3.6.9 Fire Hose Stations 4.6.9 Fire Hose Stations 3/4 6-869s
- 3. 6. 10 Fire Barrier Penetration Fire Seals 4.6.10 Fire Barrier Penetration Fire Seals 3/4 6-99 %
- 3. 6. 11 Accident Monitoring Instrumentation 4.6.11 Accident Monitoring Instrumentation 3/4 6-M 9g 3.6. 12 Reactor Protection System Motor 4. 6. 12 Reactor Protection System Motor Generator Set Monitoring Generator Set Monitoring 3/4 6-M2 <oi 3.6. 13 Remote Shutdown Panels Remote Shutdown Panels 3/4 6-~to&
3.6. 14 Radioactive Effluent Instrumentation Radioactive Effluent Instrumentation 3/4 6-188 f o7 3.6. 15 Radioactive Effluents Radioactive Effluents 3/4 6
- 3. 6. 16 Radioactive Effluent Treatment Radioactive Effluent Treatment Systems Systems 3/4 6-134 3.6. 17 Explosive Gas Mixture Explosive Gas Mixture 3/4 6-TR I DC 3.6. 18 Mark I Containment Hark I Containment 3/4 6~lb 7 3.6. 19 Liquid Waste Holdup Tanks Liquid Waste Holdup Tanks 3/4 6-~ (8 P 3.6. 20 Radiological Environmental Monitoring Radiological Environmental Monitoring
- 3. 6 21 Program Interlaboratory Comparison Program 3/4 6-~ I>+
~ Program 4. 6. 21 Interlaboratory Comparison Program 3/4 6-150 3.6. 22 Land Use .Census 4.6. 22 Land Use Census 3/4 6-~iS I 5gsis Scc IF~ s 3 g'-'1=
5.0 Design Features 5-1 5.1 Site 5-1 5.2 Reactor 5-1 5.3 Reactor Vessel 5-1 5.4 Containment 5-4 5.5 Storage of Unirradiated and Spent. Fuels 5-5
- 5. 6 Seismic Design 5-5 Amendment No. )1 Nine Mile Point Unit 1
SECTION DESCRIPTION PAGE
- 6. 0 Administrative Controls 6-1 6.1 Responsibility 6-1 6.2 Organization 6-1
- 6. 3 Facility Staff qualifications 6-7 6.4 Training 6-7
- 6. 5 Review and Audit
- 6. 6 Reportable Event Action
- 6. 7 Safety Limit Violation 6.8 Procedures 6-16
- 6. 9 Reporting Requirements 6-17
- 6. 10 Record Retention 6-25
- 6. 11 Radiation Protection Program 6-26 6.12 High-Radiation Area 6-26
- 6. 13 Fire Protection Inspection 6-28
- 6. 14 Systems Integrity 6-28
- 6. 15 Iodine Monitoring 6-28 APPENDIX B APPENDIX B Amendment No. 66 Nine Mile Point - Unit 1 iv
r Document Name: NMP-I TS SEC I Requestor's ID: NORMA Author's Name: JAMERSON C. Document Comments: PH-364 Rev. 9/22/86 KEEP THIS SHEET WITH DOCUMENT
1.0 OEFINITIONS 1.1 Reactor 0 eratin Conditions The various reactor operating conditions are defined below. Individual technical specifications amplify these definitions when appropriate.
- a. Shutdown Condition - Cold (1) The reactor mode switch is in the shutdown position or refuel position.
(2) No core alterations leading to an addition of reactivity are being performed. (3) Reactor coolant temperature is less than or equal to 212F.
- b. Shutdown Condition - Hot (1) The reactor mode switch is in the shutdown position.
(2) No core alterat'ions leading to an addition of reactivity are being performed. (3) Reactor coolant temperature is greater than 212F.
- c. Refuelin Condition (1) The reactor mode switch is in the refuel position.
(2) The reactor coolant temperature is less than 212F. (3) Fuel may be loaded or unloaded. (4) No more than one operable control rod may be withdrawn.
- d. Power 0 eratin Condition (1) Reactor mode switch is in startup or run position.
(2) Reactor is critical or criticality is possible due to control rod withdrawal. Ma or Maintenance Condition (1) No fuel is in the reactor. Nine Nile Point Unit 1
0 The ratio of the fuel rod heat flux to the heat flux of an average rod in an identical geometry bundle operating at the average core power.
- g. Total Peakin Factor The Total Peaking Factor (TPF) is the highest product of radial, axial, and local peaking factors simultaneously operative at any segment of fuel rod.
- h. Critical Power That assembly power which causes some point in the assembly to experience transition boiling.
Critical Power Ratio (CPR) The ratio of critical power to the bundle power at the reactor condition of interest.
- j. Hinimum Critical Power Ratio (HCPR)
The minimum in-core critical power ratio. Ama4nee? go. 6 Nine Hile Point - Unit 1 1-2
- 1. 2 ~0enab le A system, subsystem, train, component or device shall be operable when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, except as noted in 3. 0, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).
- 1. 3 ~0eratin Operating means that a system or component is performing its required functions in its required manner.
1.4 Protective Instrumentation Lo ic Definitions
- a. Instrument ontrol 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.
A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals 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. 1.5 Sensor Check A sensor check is a qualitative determination of acceptable operability by observation of sensor behavior during operation. This determination shall include, where possible, comparison of the sensor with other independent sensors measuring the same variable.
- 1. 6 Instrument Channel Test Instrument channel test means injection of a simulated signal into the channel to verify its proper response including, where applicable, alarm and/or trip initiating action.
Amendment No. 55 Nine Nile Point - Unit 1 1-3
1.7 Instrument Channel Calibration Instrument channel calibration means adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip.. 1.8 Ma'or Refuelin Outa e For the purpose of designating frequency of testing and surveillance, a major refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the end of the previous refueling outage, the test or surveillance need not be performed until the next regularly scheduled outage. An operating cycle is that portion of Station operation between reactor startups following each major refueling outage.
- 1. 10 Test Intervals The test intervals specified are only valid during periods of power operation and do not apply in the event of extended Station shutdown. cf l.ll Primar Containment Inte rit Primary containment integrity means that the drywell and absorption chamber are closed and all of the following conditions are satisfied:
- a. All non-automatic primary containment isolation valves which are not required to be open for plant operation are closed.
- b. At least one door in the airlock is closed and sealed.
- c. All automatic containment isolation valves are operable or are secured in the closed position.
- d. All blind flanges and manways are closed.
Nine Mile Point - Unit 1 1-4
1.12 Reactor Buildin Inte rit Reactor building integrity means that the reactor building is closed and the following conditions are met:
- a. At least one door at each access opening is closed..
- b. The standby gas treatment system is operable.
- c. All reactor building ventilation system automatic isolation valves are operable or are secured in the closed position.
- 1. 13 Core Alteration A core alteration is the addition, removal, relocation, or other manual movement of fuel or controls in the reactor core. Control rod movement with the control rod drive hydraulic system is not considered to be a core alteration.
- l. 14 Rated flux Rated flux is the neutron flux that corresponds to a steady-state power level of 1850 thermal megawatts. The use of the term 100 percent also refers to the 1850 thermal megawatt power level.
- 1. 15 Surveillance Surveillance means that process whereby systems and components which are essential to plant nuclear safety during,all modes of operation or which are necessary to prevent or mitigate the consequences of incidents are checked, tested, calibrated and/or inspected, as warranted, to verify performance and availabi lity at optimum intervals. Unless otherwise specified, those intervals specified in calendar or clock time may be adjusted plus or minus 25K to accommodate normal operating and test schedules. The total maximum combined interval time for any 3 consecutive surveillance intervals shall not exceed 3. 25 times the specified interval.
Amendment No. 6 Nine Nile Point - Unit 1 1-5
- l. 16 Fire Su ression Water S stem A Fire Suppression Water System shall consist of: a water supply system, fixed extinguishing systems of both automatic sprinklers and sprays, and manual fire fighting equipment consisting of standpipe risers with hose connections and hose reels.
- 1. 17 Fire Watch Patrol At least each hour an area with inoperable Fire Protection Equipment shall be inspected for abnormal conditions.
1.18 Gaseous Radwaste Treatment S stem A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting main condenser offgas and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
- 1. 19 Member s of the Public Member(s) of the public shall include persons who are not occupationally associated with the Nine Mile Point Nuclear Station. This category .does not include employees of Niagara Mohawk Power Corporation, the New York State Power Authority, its contractors or vendors who are occupationally associated with Nine Mile Point Unit l. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational; or other purposes not associated with Nine Mile Point Unit l.
- 1. 20 Milk Sam lin Location A milk sampling location is that location where 10 or more head of milk animals are available for the collection of milk samples.
I.21 Offsite Oose Calculation Manual OOCH) The Offsite Oose Calculational Manual shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the cal-culation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program. Amendment No. Z2, S8, 66 Nine Mile Point - Unit 1
1.22 Process Control Pro ram (PCP) The process control program shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of radioactive wastes, based on demonstrated processing of actual or simulated wet or liquid wastes, will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71, and Federal and State regulations and other requirements governing the transport and disposal of radioactive waste. Purge or purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. The purge is completed when the oxygen concentration exceeds 19. 5 percent. The site boundary shall be that line around the Nine Mile Point Nuclear Station beyond which the land is neither owned, leased, nor otherwise controlled by Niagara Mohawk Power Corporation or the New Yor k Power Authority.
- 1. 25 Solidification Solidification shall be the conversion of wet or liquid waste into a form that meets shipping and burial ground requirements.
1.26 Source Check A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
- 1. 27 Unrestricted Area The unrestricted area shall be any area at or beyond the site boundary access that is not controlled by Niagara Mohawk Power Corporation or the New York Power Author ity for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. That area outside the restricted area (10 CFR 20.3(a)(14)) but within the site boundary will be controlled by the owner as required.
Amendment No. 66 Nine Mile Point - Unit 1 1-7
1.28 Ventilation Exhaust Treatment S stem A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing venti lation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be venti lation exhaust treatment system components. 1.29 .~Ventin Venting is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during venting. Vent, used in system names, does not imply a venting process. 1.30 Reactor Coolant Leaka e
- a. Identified Leaka e (1) Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered and conducted to a sump or collecting tank, or (2) Leakage into the primary containment atmosphere from sources that are both specifically located and known not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.
- b. Unidentified Leaka e All other leakage of reactor coolant into the primary containment area.
Correction Letter of 3-27-85 Amendment No. 88, 70 Nine Nile Point - Unit 1 1-8
Document Name: NMP-1 TS SEC 2 Requestor's ID:, NORMA Author's Name: JAMERSON C Document Comments: PH-364 Rev. 9/8/86 KEEP THIS SHEET WITH DOCUMENT
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.1 FUEL CLADDING INTEGRITY 2.1.2 FUEL CLADDING INTEGRITY Applies to the interrelated variables asso- Applies to trip settings on automatic protec-ciated with fuel thermal behavior. tive devices related to variables on which the fuel loading safety limits have been placed.
~0b'ective: ~bb ective:
To establish limits on the important To provide automatic corrective action to thermal-hydraulic variables to assure the prevent exceeding the fuel cladding safety integrity of the fuel cladding. limits. Fuel cladding limiting safety system settings shall be as follows:
- a. When the reactor pressure is greater than a. The flow biased APRM scram trip settings 800 psia and the core flow is greater shall be less than or equal to that than lOX, the existence of a Minimum Cri- shown in Figure 2.1.1.
tical Power Ratio (MCPR) less than the Safety Limit Critical Power Ratio (SLCPR) (Reference 12) shall constitute viola-tion of the fuel cladding integrity safety limit.
- b. When the reactor pressure is less than b. The IRM scram trip setting shall not or equal to 800 psia or core flow is exceed 12K of rated neutron flux.
less than lOX of rated, the core power shall not exceed 25K of rated thermal power. Amendment No. 8X, 41 Nine Mile Point - Unit 1 2-1
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- c. The neutron flux shall not exceed its c. The reactor high pressure scram trip scram setting for longer than 1.5 seconds setting shall be < 1080 psig.
as indicated by the process computer. When the process computer is out of service, a safety limit violation shall be assumed if the neutron flux exceeds the scram setting and control rod scram does not occur. To ensure that the Safety Limit estab-lished in Specifications 2. l. la and
- 2. l. 1b is not exceeded, each required scram shall be initiated by its expected scram signal. The Safety Limit shall be assumed to be exceeded when scram is accomplished by a means other than the expected scram signal.
- d. Whenever the reactor is in the shutdown d. The reactor water low level scram trip condition with irradiated fuel in the setting shall be no lower than -12 inches reactor vessel, the water level shall not (53 inches indicator scale) relative to be more than 6 feet, 3 inches (-10 inches the minimum normal water level (302'9").
indicator scale) below minimum normal water level (Elevation 302'9") except as specified in "e" below.
- e. For the purpose of performing major e. The reactor water low-low level setting maintenance (not to exceed 12 weeks in for core spray initiation shall be no less duration) on the reactor vessel; the than -5 feet (5 inches indicator scale) reactor water level may be lowered relative to the minimum normal water level the minimum normal water level 9'elow (Elevation 302'9")
(Elevation 302'9"). Whenever the reactor water level is to be lowered below the low-low-low level setpoint redundant instrumentation will be provided to monitor the reactor water level. l4, Amendment'4 Nine Nile Point - Unit 1 2-2
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING Mritten procedures will be developed and followed whenever the reactor water level is lowered below the low-low level set point. (5 feet below minimum normal water level) The procedures will define the valves that will be used to lower the vessel water level. All other valves that have the potential of lowering the vessel water level will be identified by valve number in the procedures and these valves will be red tagged to preclude their operation during the major maintenance with the water level below the low-low level set point. In addition to the Facility Staff require-ments given in Specification 6. 2. 2. b, there shall be another control room operator present in the control room with no other duties than to monitor the reactor vessel water level.
- f. The flow biased APRM rod block trip settings shall be less than or equal to that shown in Figure 2.1.1.
No. Amendment'X4, 32 Nine Mile Point - Unit 1 2-3
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- g. The reactor low pressure setting for main-steam-line isolation valve closure shall be > 850 psig when the reactor mode switch is in the run position.
- h. The main-steam-line isolation valve closure scram setting shall be < 10 percent of
-valve closure (stem position) from full open.
The generator load rejection scram shall be initiated by the signal for turbine control valve fast closure due to a loss of oil pressure to the acceleration relay any time the turbine first stage steam pressure is above a value corresponding to 833 Mwt, i.e., 45 percent of 1850 Mwt.
- j. The turbine stop valve closure scram shall be initiated at < 10 percent of valve closure setting /Stem position) from full open whenever the turbine first stage steam pressure is above a value corresponding to 833 Mwt, i.e., 45 percent of 1850 Mwt.
Nine Mile Point - Unit 1 2-4
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~. CO>if Pn(SS>>nf >S ~ COO p>I ~
>CHAN
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W ROO BLt. 100 0 z UI ~J 4l 00 J El 0 60 Z SA -- ~so'PF-TPF 3.00 for all Gx8 fuel W>> fl>e> SA 1>'f tlfvlSC>lnl >n>lo >>U>> >>>.OCK uTPF - CA>.cu>.nl oo >>A:I>>>uv ToTA>. PEAK>t>o FACToll 20 sc>>na>> >oo ococK s>>own ncovE 0 '10 CO 50 60 lO 60 110 nfc)I>cu>.ATlott F>.O~(.rfncf>IT( F OfsiGn Figure 2.1.1. Flow Biased Scram and APRN Rod Block Amendment No. 8Z, 41 Nine Mile Point - Unit 1 2-5
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.2. 1 REACTOR COOLANT SYSTEM 2.2.2 REACTOR COOLANT SYSTEM
- a. The settings on the safety valves of the pressure vessel shall be as shown Applies to the limit on reactor coolant below. The allowable initial set point system pressure. error on each setting will be + 1 percent.
~0b 'ective:
Set Point Number of To define those values of process vari- (P51(P Saf et Val ves ables which shall assure the integrity of the reactor coolant system to prevent 1218 an uncontrolled release of radioactivity. 1227 1236 1245 1254 The reactor vessel or reactor coolant sys-tem pressure shall not exceed 1375 psig at b. The reactor high-pressure scram trip any time with fuel in the vessel. setting shall be < 1080 psig.
- c. The flow biased APRM scram trip settings shall be shown in Figure 2.1.1.
Nine Mile Point - Unit 1 2-6
Document Name: NMP-1 TS SEC B2 Requestor's ID: EVA Author's Name: JAMERSON C. Document Comments: PH-364 Rev. 9/10/86 KEEP THIS SHEET WITH DOCUMENT
BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not directly observable, a step-back approach is used to establish a safety limit such that the Minimum Critical Power Ratio (MCPR) is no less than the Safety Limit Critical Power Ratio (SLCPR) (Reference 12). The SLCPR represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one'f the physical barriers which separate radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding safety limit is defined with margin to the conditions which would produce onset of transition boiling, (MCPR of 1. 0). These conditions represent a significant departure from the condition intended by design for planned operation. Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boiling transition, is riot a directly observable parameter in an operating reactor. Therefore, at reactor pressure > 800 psia and core flow > 10X of the rated margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the Critical Power Ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the Minimum Critical Power Ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective set points via the instrumented variables, by the nominal expected flow control line. The SLCPR has sufficient conservatism to assure that in the event of an abnormal operational transient initiated from a normal operating condition more than 99.9X of the fuel rods in the core are expected to avoid boiling transition. The margin between MCPR of 1. 0 (onset of transition boiling) and the SLCPR is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as des-cribed in References 1 and 12. Amendment No. 8, 8X, 41 Nine Mile Point - Unit 1 B 2-1
BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of the SLCPR would not produce boiling transition. Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity. However, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increase to approximately 1100'F which is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Test Reactor (GETR) where similar fuel operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation. If reactor pressure should ever exceed 1400 psia during normal power operating (the limit of applicability of the boiling transition correlation) been violated. it would be assumed that the fuel cladding integrity safety limit has In addition to the boiling transition limit SLCPR operation is constrained to a maximum LHGR of 13.4 kW/ft for Sx8 fuel and 13.4 kW/ft for SxSR fuel. At 100K power this limit is reached with a Maximum Total Peaking Factor (HTPF) of 3.02 for SxS fuel and F 00 for SxSR fuel. For the case of the MTPF exceeding these values, operation is permitted only at less than lOOX of rated thermal power and only with reduced APRH scram set-tings as required by Specification 2. 1.2.a. (In cases where for a short period the total peaking factor was above 3.02 for SxS fuel and 3.00 for SxSR fuel the equation in Figure 2. 1.1 will be used to adjust the flow biased scram and APRN rod block set points.) At pressure equal to or below.800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi. At low power and all core flows, this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and all flows will always be greater than 4.56 psi. Analyses show that with a bundle flow of 28x103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Therefore, due to the 4.56 psi driving head, the bundle flow will be grea-ter than 28xlOs lb/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.-7 psia to 800 psia indicate that the fuel assembly critical power at 28xl03 lb/hr is approximately 3. 35 MWt. With the design peaking factor, this corresponds to a core thermal power of more than 50K Thus, a core thermal power limit of 25K for reactor pressures below 800 psia or core flow less than lOX is conservative. Amendment No. 9, 8g, 41 Nine Mile Point.- Unit 1 B 2-2
BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT During transient operation the heat flux (thermal power-to-water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel which is 8 to 9 seconds. Also, the limiting safety system scram settings are at values which will not allow the reactor to be operated above the safety limit during normal operation or during other plant operating situations which have been analyzed in detail. (3,4) In addition, control rod scrams are such that for normal operating transients the neutron flux transient is-terminated before a significant increase in surface heat flux occurs. Scram times of each control rod are checked periodically to assume adequate insertion times. Exceeding a neutron flux scram setting and a fail-ure of the control rods to reduce flux to less than the'scram setting within 1.5 seconds does not necessarily imply that fuel is damaged; however, for this specification a safety limit violation will be assumed any time a neutron flux scram setting is exceeded for longer than l. 5 seconds. If the scram occurs such that the neutron flux dwell time above the limiting safety system setting is less-than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected. These analyses show that even if the bypass system fails to operate, the design limit of the SLCPR is not exceeded. Thus, use of a 1.5-second limit provides additional margin. The process computer has a sequence annunication program which will indicate the sequence in which scrams occur such as neutron flux, pressure, etc. This program also indicates when the scram set point is cleared. This will provide information on how long a scram condition exists and thus provide some measure of -the energy added during a transient. Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 2.1. l.c will be relied on to determine if a safety limit has been violated. Amendment No. 5, 8l, 41 Nine Mile Point - Unit 1 B 2-3
BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height. The lowest point at which the reactor water level can 'normally be monitored is approximately 7 feet 11 inches below minimum normal water level or 4 feet 8 inches above the top of the active fuel. This is the location of the reactor vessel tap for the low-low-low water level instrumentation. The actual low-low-low water leveg trip point is 6 feet 3 inches (-10 inches indicator scale) below minimum normal water level (Elevation 302~"). The 20 inch difference resulted from an evaluation of the recommendations contained in General Electric Service Information Letter 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instru-mentation." The low-low-low water level trip point was raised 20 inches to conservatively account for possi-ble differences in actual to indicated water level due to potentially high drywell temperatures. The safety limit has been established here to provide a point which can be monitored and also can provide adequate mar-gin. However, for performing major maintenance as specified in Specification 2.l.l.e, redundant instrumenta-tion will be provided for monitoring reactor water level below the low-low-low water level set point. (For example, by installing temporary instrument lines and reference points to redundant level transmitters so that the reactor water level may be monitored over the required range.) In addition, written procedures, which identify all the valves which have the potential of lowering the water level inadvertently, are esta-blished to prevent their operation during the major maintenance which requires the water level to be below the low-low-.level set point. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e. g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a safety limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis. Amendment No. 64 Nine Mile Point - Unit 1 B 2-4
0 BASES FOR 2.1'.2 j=UEL CLAODING LS The abnormal operational transients applicable to operation of the plant have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 1850 MMt. The analyses were based upon plant operation in accordance with the operating map given in Reference 11. In addition, 1850 MMt is the licensed maximum power level, and represents the maximum steady-state power which shall not knowingly be exceeded; Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance. Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model. The comparisons and results are summarized in Reference 2. The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25K greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 80K of the total scram worth of the control rods. The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is as-sured by the time requirements for 5X and 20K insertion. By the time the rods are 60K inserted, approxi-mately four dollars of negative reactivity have been inserted which strongly turns the transient, and accom-plishes the desired effect. The times for 50K and 90K insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shut-down steady-state condition. This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.
- a. The Average Power Range Monitoring (APRH) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated thermal power. Because fission chambers Nine Nile Point - Unit 1 B 2-5
BASES FOR 2.1.2 FUEL CLADDING - LS provide the basic input signals, the APRH system responds directly to average neutron flux. During. transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal opera-tional transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting: ,Analyses (5,6,8,9,10,11,13) demonstrate that with a 120K scram trip setting, none of the abnormal operational transients analyzed violate the fuel safety limit and there is a substantial margin from fuel damage.- However, in response .'to expressed beliefs (7) that variation of APRH flux scram with recirculation flow is a prudent measure to assure safe plant operation during the design confirmation phase of plant opera-tion, the scram setting will be varied with recirculation flow. An increase in the APRH scram trip setting would decrease the margin present before the fuel cladding integrity safety limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRH scram trip setting was selected because it provides adequate margin for the fuel cladding integrity safety limit yet allows operating margin that reduces the possibility of unnecessary scrams. The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of HTPF and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Figure 2.1.1 when the maximum total peaking factor is greater than the limiting total peaking factor.
- b. Normal operation of the automatic recirculation pump control will be in excess of 30K rated flow; therefore, little operation below 30K flow is anticipated. For operation in the start-up mode while the reactor is at low pressure, the IRH scram setting is 12K of rated neutron flux. Although the operator will set the IRH scram trip at 12K of rated neutron flux or less, the actual scram setting can be as much as 2.5X of rated neutron flux greater. This includes the margins discussed above. This provides adequate margin between the setpoint and the safety limit at 25K of rated power. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. There are a few possible sources of rapid reactivity input to the system in the low power flow condition. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during Amendment No. 36 Nine Mile Point - Unit 1 B 2-6
BASES FOR 2.1.2 FUEL CLADDING - LS startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5X of rated per minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit. Procedural controls will assure that the IRM scram is maintained up to 20K flow. This is accomplished by keeping the reactor mode switch in the startup position until 20K flow is exceeded and the APRM's are on scale. Then the reactor mode switch may be switched to the run mode, thereby switching scram protection from the IRM to the APRM system. j In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcriti-cal and the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. -The results of this analysis show that the reactor is scrammed and peak power limited to 1X of rated power, thus maintaining a limit above the SLCPR. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM. As demonstrated in Appendix E-I" and the Technical Supplement to Petition to Increase Power Level, the reactor high pressure scram is a backup to the neutron flux scram, turbine stop valve closure scram, generator load rejection scram, and main steam isolation valve closure scram, for various reactor isolation incidents. However, rapid isolation at lower power levels generally results in high pressure scram preceding other scrams because the transients are slower and those trips associated with the turbine generator are bypassed. The operator will set the trip setting at 1080 psig or lower. However, the actual set point can be as much as 15.8 psi about the 1080 psig indicated set point due to the deviations discussed above. Amendment No. 9, 81, 41 Nine Mile Point - Unit 1 B 2-7
BASES FOR 2.1.2 FUEL CLADOING - LS A reactor water low level scram trip setting -12 inches (53 inches indicator scale) relative to the minimum normal water level (Elevation 302'") will assure that power production will be terminated with adequate coolant remaining in the core. The analysis of the feedwater pump loss in the Technical Supplement to Petition to Increase Power Level, dated April 1970, has demonstrated that approximately 4 feet of water remains above the core following the low level'scram. The operator will set the low level trip setting no lower than -12 inches relative to the lowest normal operating level. However, the actual set point can be as much as 2.6 inches lower due to the deviations discussed above.
- e. A reactor water low-low level signal -5 feet (5 inches indicator scale) relative to the minimum normal water level (Elevation 302'") will assure that core cooling will continue even if level is dropping.
Core spray cooling will adequately cool the core, as discussed in LCO 3. 1.4. The operator will set the low-low level core spray initiation point at no less than -5 feet (5 inches indicator scale) relative to the minimum normal water level (Elevation 302'"). However, the actual set point can be as much as 2. 6 inches lower due to the deviations discussed above. Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a MCPR less than the SLCPR. This rod block trip setting, which is automatically varied with recirculation flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 105K of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRM rod block peaking factor exceeds the design peaking trip setting is adjusted downward if the maximum total factor, thus preserving the APRM rod block safety margin. Amendment No. 9, 8Z, 41 Nine Mile Point - Unit 1 B 2-8
BASES FOR 2.1.2 FUEL CLADDING - LS g-h. The low pressure isolation of the main steam lines at 850 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus, the combina-tion of main steam line isolation on reactor low pressure and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel clad-ding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. With the scrams set at < 10K valve closure, there is no increase in neutron flux and peak pressure in the vessel dome is limited to 1141 psig. (8,9,10) The operator will set the pressure trip at greater than or equal to 850 psig and the isolation valve
, stem position scram setting at less than or equal to 10K of valve stem position from full open. How-ever, the actual pressure set point can be as much as 15.8 psi lower than the indicated 850 psig and the valve position set point can be as much as 2. 5X of stem position greater. These allowable deviations are due to instrument error, operator setting error, and drift with time.
In addition to the above m~ntioned LS , other reactor protection system devices (LCO 3. 6. 2) serve as a secondary backup to the LS chosen. These are as follows:
/
High fission product activity released from the core is sensed in the main steam lines by the high radiation main steam line monitors. These monitors provide a backup scram signal and also close the main steam line isolation valves. The scram dump volume high level scram trip assures that scram capability will not be impaired because of insufficient scram dump volume to accommodate the water discharged from the control rod drive hydraulic system as a result of a reactor scram (Section X-C. 2. 10). * ~FSAR Amendment No. 5 Nine Mile Point - Unit 1 B 2-9
BASES FOR 2.1.2 FUEL CLADDING - LS The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to the worst case transient of a load rejection and subsequent failure of the bypass. In fact, analysis (9,10) shows that heat flux does not increase from its initial value at all because of the fast action of the load rejection scram; thus, no significant change in HCPR occurs. j ~ The scram turbine stop valve closure scram is provided for the same reasons as discussed in i above. Mith setting of < 10K valve closure, the resultant transients are nearly the same as for those de-a scribed in i above; and, thus, adequate margin exists. Amendment No. 5 Nine Nile Point - Unit 1 B 2-10
REFERENCES FOR BASES FOR 2.1.1 AND 2.1.2 FUEL CLADDING (1) General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO-10958 and NEDE-10958. (2) Linford; R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NED0-10801, February 1973. (3) FSAR, Volume II, Appendix E. (4) FSAR, Second Supplement. (5) FSAR, Volume II, Appendix E. (6) FSAR, Second Supplement. (7) Letters, Peter A. Morris, Director of Reactor Licensing, USAEC, to John E. Logan, Vice-President, Jersey Central Power and Light Company, dated November 22, 1967 and January 9, 1968. (8) Technical Supplement to Petition to Increase Power Level, dated April 1970. (9) Letter, T. J. Brosnan, Niagara Mohawk Power Corporation, to Peter A. Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972. (10) Letter, Philip D. Raymond, Niagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973. (ll) Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, NED0-24012, May 1977. (12) Licensing Topical Report General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P-A, August 1978. (13) Nine Mile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6), NED0-24185, April 1979. (14) General Electric SIL 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation." Amendment No. 64 Nine Mile Point - Unit 1 B 2-11
~ . BASES FOR 2.2.1 REACTOR COOLANT SYSTEM SAFETY LIMIT The pressure safety limit of 1375 psig was derived from the design pressures and applicable codes for the reactor pressure vessel and the reactor coolant system piping. (ASME Boiler and Pressure Vessel Code Section I applies to the reactor pressure vessel and ASA Piping Code, Section B31.1 applies to the coolant system piping.) The ASME Code permits pressure transients up to 10 percent over design pressure (llOX x 1250 = 1375 psig) and the ASA Code permits pressure transients up to 15 percent over the design pressure (115K x 1200 = 1380 psig). Oata presented in Volume IV, Section I-B" includes the design analyses which were performed to demonstrate that the reactor pressure vessel would meet the applicable code requirements. As a part of these analyses, both design and non-design events (Tables 7 and 8) were postulated to evaluate their strain effect to the vessel. Among the non-. design events, a postulated over-pressure of 3750 psig was expected to result in vessel destruction. Comparable data concerning the piping system is not available, however, ASA Code (B31.1) indicates a margin of safety factor, code allowable (10,800 psi at 600F) versus yield strength (75,000 psi), of 6.8 for the process piping system while the margin of safety factor (15,000 psi vs. 60,000 psi) for the high pressure feedwater system is 4. Additional data in Supplement 2, Table IV-1" indicates a calculated feedwater valve burst pressure of 13,000 psi based upon a yield strength of 36,000 psi. Based upon the available data and for safety valve sizing calculations, 1375 psig was selected as a safety limit for the reactor coolant system. The maximum pressure of the critical hydro test of the unfueled system was selected as 1800 psig, while the normal system operating pressure will be 1030 psig. "FSAR Nine Nile Point - Unit 1 B 2-12
BASES FOR 2. 2. 2 REACTOR COOLANT SYSTEH - LS
- a. The range of set points for a safety valve actuation is selected in accordance with code requirements.
A safety valve capability3study presented in the Technical Supplement to Petition to Increase Power Level using the stated LS values has demonstrated the maximum pressures occur ring at the bottom of the reactor vessel and the bottom of the recirculation piping are 1303 psig and 1315 psig, respectively, some 72 psig below the 1375 psig safety limit. This analysis has assumed the highly improbable event of reactor isolation occurring without scram, in spite of separate and redundant scram signals such that the power output reached 167 percent of rated (1850 Hwt). In addition to the safety valves, the solenoid-actuated relief valves are used to prevent safety valve during rapid reactor isolation at power coupled with failure of the bypass system. Any five of these valves lift opening at 1090 psig to 1100 psig will keep the maximum vessel pressure below the lowest safety valve setting, as demonstrated in Appendix E-I. 3. 11 (p. E-35). " (The Technical Supplement to Petition to Increase Power Level, and letter from T. J. Brosnan, Niagara Hohawk Power Corporation, to Peter A. Horris, Oivision of Reactor Licensing, USAEC, dated February 28, 1972. ) Subsequently, six valves were provided due to the blowdown requirements, following a small line break. The capacity of a solenoid-actuated relief valve is about the same as a safety valve. Therefore, even without scram any combination of 16 safety valves and solenoid-actuated valves will limit the pressure below the safety limit following the worst isolation situation.
- b. The reactor high pressure scram setting is relied upon to terminate rapid pressure transients if other scrams, which would normally occur first, fail to function. As demonstrated in Appendix E-I of the FSAR and the Technical Supplement to Petition to Increase Power Level, Page II-12, the reactor high pressure scram is a backup to the neutron flux scram, generator load rejection scram, and main steam isolation-valve closure scram for various reactor isolation incidents. However, rapid isolation at lower power levels generally results in high pressure scram preceding other scrams because the transients are slower and those trips associated with the turbine-generator are bypassed.
The operator will set the trip setting at 1080 psig or lower. However, the actual set point can be as much as 15.8 psi above the 1080 psig indicated set point due to the deviations discussed above. "FSAR Nine Hi le Point - Unit 1 B 2-13
BASES FOR 2. 2. 2 REACTOR COOLANT SYSTEH - L
- c. As shown in Appendix E-I. 3. 8 and 3. 11,* rapid Station transients due to isolation valve or turbine trip valve closures result in coincident high-flux and high-pressure transients. Therefore, the APRH trip, although primarily intended for core protection, also serves as backup protection for pressure transients.
Although the operator will set the scram setting at less than or equal to that shown in Figure 2.1.1 the actual neutron flux setting can be as much as 2.7 percent of rated neutron flux above the line. This includes the errors discussed above. The flow bias could vary as much as one percent of rated recirculation flow above or below the indicated point. In addition to the above-m~ntioned 3 LS , other reactor protection system devices (LCO 3. 6. 2) serve as secondary backup to the LS chosen. These are as follows: The primary containment high-pressure scram serves as backup to high reactor pressure scram in the event of lifting of the safety valves. As discussed in Vol. I, VIII, 2.0.c (p. VIII-9)*a pressure in excess of 3. 5 psig due to steam leakage or blowdown to the drywell will trip a scram well before the core is uncovered. A low condenser vacuum situation will result in loss of the main reactor heat sink, causing an increase in reactor pressure. The scram feature provided, therefore, anticipates the reactor high-pressure scram. A loss of main condenser vacuum is analyzed in Appendix E-I.3. 17." The scram dump volume high-level scram trip assures that scram capability will not be impaired because of insufficient scram dump volume to accommodate the water discharge from the control-rod-drive hydraulic system as a result of a reactor scram (Section X-C. 2. 10). " In the event of main-steam-line isolation valve closure, reactor pressure will increase. A reactor scram is, therefore, provided on main-steam-line isolation valve position and anticipates the high reactor pressure scram trip.
- FSAR Nine Hile Point - Unit 1 8 2-14
Document Name: NMP-1 TS SEC 3/4 0 Requestor's ID: NORMA Author's Name: Jamerson, C. Document Comments: 9/8/86 Revisions
3.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS OPERABILITY REQUIREMENTS When a system, subsystem, train, component or device is determined to be inoperable solely because its emer-gency power source is inoperable, or solely because its normal power source is inoperable, it may be considered operable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is operable; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are operable, or likewise satisfy the require-ments of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in a condition stated in the individual specification. In the event a Limiting Condition for Operation and/or associated surveillance requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in a condition consistent with the individual specification unless corrective measures are completed that permit operation under the permissible surveillance requirements for the specified time interval as measured from initial discovery or until the reactor is placed in an operational condition in which the specification is not applicable.
- 3. 1. 0 FUEL CLADDING A) GENERAL APPLICABILITY Applies to the power level regulation, control rod system, liquid poison system, emergency cooling system, and core spray system. LCO's for the minimum allowable circuits corresponding to the LS3 settings are included in the Reactor Protection System LCO (3. 6. 2).
B) GENERAL OBJECTIVE LIMITING CONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of the systems and associated components which will assure the integrity of the fuel cladding as a barrier against the release of radioactivity. SURVEILLANCE REQUIREMENTS - To define the tests or inspections required to assure the functional capability or performance level of the required systems or components. Amendment No. 55 Nine Mile Point - Unit 1 3/4 0-1
Document Name: NMP-1 TS SEC 3/4 1 Requestor's ID: NORHA Author's Name: Jamerson, C Document Comments: PH-364 Rev. 9/22/86 PLEASE RETURN THIS SHEET WITH REVISIONS
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UI REHENT 3.1.1 CONTROL ROD SYSTEH 4.1.1 CONTROL ROD SYSTEH Applies to the operational status of Applies to the periodic testing requirements the control rod system. for the control rod system.
~0b ective: ~0b ective:
To assure the capability of the control To specify the tests or inspecti.ons required rod system to control core reactivity. to assure the capability of the control rod system to control core reactivity. The control rod system surveillance shall be performed as indicated below.
- a. Reactivity Limitations a. Reactivity Limitations (1) Reactivity margin - core loading (1) Reactivity margin - core loading The core loading shall be limited Sufficient control rods shall be to that which can. be made subcri- withdrawn following a refueling outage tical in the most reactive con- when core alterations were performed to dition during the operating cycle demonstrate with a margin of 0.25 percent with the strongest control rod in bk that the core can be made subcritical its full-out position and all at any time in the subsequent fuel cycle other operable rods fully inserted. with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.
Nine Nile Point - Unit 1 3/4 1-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Reactivity margin - stuck control (2) Reactivity margin stuck control rods rods Control rods which cannot be moved Each partially or fully withdrawn with control rod drive pressure control rod shall be exercised at least shall be considered inoperable. once each week. This test shall be Inoperable control rods shall be performed at least once per 24 hours in valved out of service, in such the event power operation is continuing positions that Specification 3. 1. 1 with two or more inoperable control rods a(l) is met. In no case shall the or in the event power operation is con-number of non-fully inserted rods tinuing with one fully or partially valved out of service be greater withdrawn rod which cannot be moved and than six during power operation. for which control rod drive mechanism If this specification is not met, damage has not been ruled out. The the reactor shall be placed in the surveillance need not be completed cold shutdown condition. If a partially or fully withdrawn within 24 hours if the number of in-operable rods has been reduced to less control rod drive cannot be moved with drive or scram pressure the than two and if it has been demonstrated that control rod drive mechanism collet reactor shall be brought to a housing failure is not the cause of an shutdown condition within 48 hours immovable control rod. unless investigation demonstrates that the cause of the failure is not due to a failed control rod drive mechanism collet housing.
- b. Control Rod Withdrawal b. Control Rod Withdrawal (1) The control rod shall be coupled to (1) The coupling integrity shall be verified its drive or completely inserted for each withdrawn control rod by either:
and valved out of service. When removing a control rod drive for (a) Observing the drive does not go to the inspection, this requirement does overtravel position, or not apply as long as the reactor is in a shutdown or refueling (b) A discernible response of the nuclear condition. instrumentation. Amendment No. Nine Mile Point - Unit 1 3/4 1-2
P~ LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) The control rod drive housing (2) The control rod drive housing support support system shall be in place system shall be inspected after reassembly. during power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3. 1.1a(1) is met. (3)(a) Control rod withdrawal sequences (3)(a) To consider the rod worth minimizer shall be established so that max- operable, the following steps must imum reactivity that could be add- be performed: ed by dropout of any increment of any one control blade would not (i) The control rod withdrawal sequence make the core more than 0.013 for the rod worth minimizer computer delta k supercritical. shall be verified as correct. (ii) The rod worth minimizer computer on-line diagnostic test shall be successfully completed. (iii) Proper annunciation of the select error of at least one out-of-sequence control rod in each fully inserted group shall be verified. (iv) The rod block function of the rod worth minimizer shall be verified by attempting to withdraw an out-of-sequence control rod beyond the block point. Nine Mile Point - Unit 1 3/4 1-3
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT (b) Whenever the reactor is in the (b) If the rod worth minimizer is inoperable startup ot run mode below 20K while the reactor is in the startup or I rated thermal power, no control run mode below 20K rated thermal power rods shall be moved unless the and a second independent operator or rod worth minimizer is operable, engineer is being used he shall verify except as noted in that all rod positions are correct prior 4.l.l.b(3)(a)(iv), or a second to commencing withdrawal of each rod independent operator or engineer group. verifies that the operator at the reactor console is following the control rod program. The second operator may be used as a substitute for an inoperable rod worth minimizer during a startup only if the rod worth minimizer fails after withdrawal of at least twelve control rods. (4) Control rods shall not be with-drawn for approach to critical-ity unless at least three source range channels have an observed count rate equal to or greater than three counts per second. Amendment No. 16 Nine Nile Point - Unit 1 3/4 1-4
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. Scram Insertion Times c. Scram Insertion Times (1) The average scram insertion (1) After each major refueling outage time of all operable control and prior to power operation with rods, in the power operation reactor pressure above 800 psig, condition, shall be no greater all operable control rods shall be than: scram time tested from the fully withdrawn position.
X Inserted Average Scram From Fully Insertion Withdrawn Times (sec) 5 0. 375 20 0.90 50 2.00 90 5.00 (2) Except as noted in 3. l. 1.c(3), (2) Following each reactor scram from the maximum insertion scram time, rated pressure, the mean 90K Inser-in the power operation condition, tion time shall be determined for shall be no greater than: eight selected rods. If the mean 90K insertion time of the selected X Inserted Maximum Scram control rod drives does not fall From Fully Insertion within the range of 2.4 to 3. 1 sec-Withdrawn Times (sec) onds or the measured scram time of any one drive for 90K insertion 5 0. 398 does not fall within the range of 20 0.954 1. 9 to 3. 6 seconds, an evaluation 50 2. 12 shall be made to provide reasonable 90 5.30 assurance that proper control rod drive performance is maintained. Nine Mile Point - Unit 1 3/4 1-5
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT (3) Control rods with longer scram (3) Following any outage not initiated insertion time will be per- by a reactor scram, eight rods shall mitted provided that no other be scram tested with reactor pres-control rod in a nine-rod sure above 800 psig. The same cri-square array around this rod teria of 4.l.l.c(2) shall apply. has a: (a) Scram insertion time greater than the maxi-mum allowed, tionn. (b) Malfunctioned accumulator, (c) Valved out of service in a non-fully inserted posi-
- d. Control Rod Accumulators d. Control Rod Accumulators At all reactor operating pressures, Once a shift check the status of the a rod accumulator may be out of accumulator pressure and level alarms service provided that no other con- in the control room.
trol rod in a nine-rod square array around this rod has a: (1) Malfunctioned accumulator, (2) Valved out of service in a non-fully inserted position, (3) Scram insertion greater than maximum permissible insertion time. Nine Nile Point - Unit 1 3/4 1-6
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT If a control rod with a malfunctioned accumulator is inserted "full-in" and valved out of service, it shall not be considered to have a malfunctioned accumulator.
- e. If Specification 3.1.1.a through d, above, are not met, the reactor shall be placed in the hot shutdown condition within ten hours except as noted in 3. 1. 1. a(2).
- f. Reactivity Anomalies f. Reactivity Anomalies The difference between an observed The observed control rod inventory shall and predicted control rod inventory be compared with a normalized computed shall not exceed the equivalent of prediction of the control rod inventory one percent in reactivity. If this during startup, following refueling or limit is exceeded, the reactor shall major core alteration. These comparisons be brought to the cold, shutdown con- will be used as base data for reactivity dition by normal orderly shutdown pro- monitoring during subsequent power opera-cedure. Operation shall not be per- tion throughout the fuel cycle. At mitted until the cause has been eval- specific power operating conditions, the uated and the appropriate corrective actual control rod configuration will be action has been completed. compared with the expected configuration based upon appropriately corrected past data. This comparison will be made every equivalent full power month.
Amendment No. 84 Nine Mile Point - Unit 1 3/4 1-7
LIMITING CONDITION FOR OPERATION SURYEILLANCE REQUIREMENT A licabilit: Applies to the operating status of the Applies to the periodic testing require-liquid poison system. ments for the liquid poison system.
~0b'ective: ~0b ective:
To assure the capability of the liquid To specify the tests required to assure the poison system to function as an indepen- capability of the liquid poison system for dent reactivity control mechanism. controlling core reactivity. S ecification: S ecification: The liquid poison system surveillance shall be performed as indicated below:
- a. During periods when fuel is in the a. Overall S stem Test:
reactor and the reactor is not shut-down by the control rods, the liquid (1) At least once durin each o er-poison system shall be operable ex- at)n c c e-cept as specified in 3. 1.2. b. Manually initiate the system from the control room. Demineralized water shall be pumped to the re-actor vessel to verify minimum flow rates and demonstrate that valves and nozzles are not clogged. Remove the squibs from the valves and verify that no deterioration has occurred by actual field fir-ing of the removed squibs. In ad-dition, field fire one squib from the batch of replacements. Nine Mile Point - Unit 1 3/4 1-8
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT Disassemble and inspect the squib-operated valves to verify that valve deterioration has not occurred. (2) At least once er month-Demineralized water shall be recycled to the test tank. Pump discharge pressure and minimum flow rate shall be verified.
- b. If a redundant component becomes in- b. Boron Solution Checks:
operable, Specification 3. l. 2. a. shall be considered fulfilled, pro- (1) At Least once er month-vided that the component is returned to an operable condition within 7 Boron concentration shall be de-days and the additional surveillance termined. required is performed. (2) At least once er da Solution volume shall be checked. In addi-tion, the boron concentration shall be de-termined any time water or boron are added or if the solution temperature drops below the limits specified by Figure 3.1.2.b. (3) At least once er da The solution temperature shall be checked. Nine Mile Point - Unit 1 3/4 1-9
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- c. The liquid poison tank shall contain c..Surveillance with Ino erable Com onents a boron-bearing solution that satis-fies the volume-concentration re- When a component becomes inoperable its re-quirements of Figure 3. 1.2. a Revised dundant component shall be demonstrated to at all times when the liquid poison be operable immediately and daily thereafter.
system is required to be operable.
- d. The liquid poison solution tempera-ture shall not be less than the tem-perature presented in Figure 3. 1. 2. b.
- e. If Specifications "a" through "d" are not met, initiate normal orderly shutdown within one hour.
Nine Mile Point - Unit 1 3/4 1-10
Figure 3.1.2a Revised VOLUME-CONCENTRATION LIMITS 22 20 18 REGION OF REQUIRED VOLUME - I CON ENTRAT ON 16 14 12 a. 2000 al. 20.4X ga
- c. 4080 gal ., 10.7$
- d. 3800 gal ., 10.7X 1500 2000 2500 3000 3500 4000 4500 VOLUME OF SOLUTION IN TANK (GALLONS)
Figure 3.1.2b HIHIftUtl ALLOWABLE SOLUTION TEMPERATURE 160 140 cr 120 I-Ul 100 C) 80 eo 40
.10 20 30 40 WEIGHT PERCENT SOOIUH PEHTABORATE IN SOLUTIOH
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 1.3 EMERGENCY COOLING SYSTEM 4. 1. 3 EMERGENCY COOLING SYSTEM Applies to the operating status of the Applies to periodic testing requirements emergency cooling system. for the emergency cooling system.
~0b 'ective: ~0b ective:
To assure the capability of the emergency To assure the capability of the cooling system to cool the reactor coolant emergency cooling system for cooling of in the event the normal reactor heat sink the reactor coolant. is not available. e gati The emergency cooling system surveillance shall be performed as indicated below:
- a. During power operating conditions and a. At least once ever five ears-whenever the reactor coolant tempera-ture is greater than 212'F, except for The system heat removal capability hydrostatic testing with the reactor shall be determined.
not critical, both emergency cooling systems shall be operable except as specified in 3. 1. 3. b and c.
- b. During the remainder of Cycle 8 with b. At least once dail one emergency cooling system inoperable, Specification 3.1.3a shall be consider- The shell side water level and ed fulfilled, provided the additional makeup tank water level shall be surveillance required in 4. 1. 3.f is checked.
performed. Amendment No. 79, 82 Nine Mile Point - Unit 1
i LIHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREHENT
- c. During Cycle 9 and subsequent cycles, c. At least once er month-if one emergency cooling system becomes inoperable, Specification 3.1.3.a shall The makeup tank level control valve be considered fulfilled, provided that shall be manually opened and closed.
the inoperable system is returned to an operable condition within 7 days and the additional surveillance required in 4.1.3.f is performed.
- d. Hake up.water shall be available d. At least once each shift from the two gravity feed makeup water tanks. The area temperature shall be checked.
- e. During Power Operating Conditions, e. Durin each ma or refuelin outa e-each emergency cooling system high point vent to torus shall be Automatic actuation and functional operable. system testing shall be performed during each major refueling outage and whenever
- 1. With a vent path for one major repairs are completed on the emergency cooling system system.
inoperable, restore the vent path to an operable condition Each emergency cooling vent path shall within 30 days. be demonstrated operable by cycling each power-operated valve (05-01R, 05-'ll,
- 2. With vent paths for both 05-12, 05-04R, 05-05 and 05-07) in the emergency cooling systems vent path through one complete cycle of inoperable, restore one vent full travel and verifying that all path to an operable condition manual valves are in the open position.
within 14 days and both vent paths within 30 days.
- f. If Specification 3.1.3.a, b, c, d f. Surveillance with an Ino erable S stem or e are not met, a normal orderly shutdown shall be initiated within During Cycle 8 with one of the emergency one hour, and the reactor shall be cooling systems inoperable, the level Amendment No. 75 Nine Hile Point - Unit 1 3/4 1-14
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT in the cold shutdown conditions control valve and motor operated within ten hours. isolation valve in the operable system shall be demonstrated to be operable weekly. During Cycle 9 and subsequent cycles, when one of the emergency cooling systems is inoperable, the level control valve and the motor-operated isolation valve in the operable system shall be demonstrated to be operable immediately and daily thereafter. Amendment No. 75 Nine Mile Point - Unit 1 3/4 1-15
0 LIMITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 1.4 CORE SPRAY SYSTEM 4. 1. 4 CORE SPRAY SYSTEM Applies to the operating status of the Applies to the periodic testing requirements core spray systems. for the core spray systems.
~bb ective: ~0b ective:
To assure the capability of the core spray To verify the operability of the core spray systems to cool reactor fuel in the event systems. of a loss-of-coolant accident. The core spray system surveillance shall be performed as indicated below.
- a. Whenever irradiated fuel is in the a. At each major refueling outage automatic reactor vessel, each of the two core startup of one set of pumps in each core spray systems shall be operable except spray system shall be demonstrated.
as specified in Specifications b, c and d, below.
- b. If a redundant component of a core b. At least once per quarter pump operability spray system becomes inoperable, that shall be checked.
system shall be considered operable provided that the component is returned to an operable condition within 15 days and the additional surveillance required is performed. Nine Nile Point - Unit 1 3/4 1-16
LIHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREHENT If a redundant component in each of the c. At least once per quarter the operability core spray systems becomes inoperable, of power-operated valves required for both systems shall be considered operable proper system operation shall be checked. provided that the component is returned to an operable condition within 7 days and the additional surveillance required is performed. If a core spray system becomes inoperable d. Core spray header bP instrumentation and all the components are operable in the other system, the reactor may remain check Once/day in operation for a period not to exceed calibrate Once/3 months 7 days. test Once/3 months
- e. If Specifications a, b, c and d are not e. Surveillance with Ino erable Com onents met, a normal orderly shutdown shall be initiated within one hour and the Mhen a component or system becomes reactor shall be in the cold shutdown inoperable its redundant component or condition within ten hours. system shall be demonstrated to be operable immediately and daily thereafter.
If both core spray systems become in-operable the reactor shall be in the cold shutdown condition within ten hours and no work (except as specified in "f" and "h" below) shall be performed on the reactor or its connected systems which could result in lowering the reactor water level to more than six feet, three inches below minimum normal water level (-10 inches indicator scale). Amendment No. 64 Nine Hile Point - Unit 1 3/4 1"17
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- f. Work may be performed on control rod f. Surveillance during control rod drive drives at times when water is not in maintenance which is simultaneous with the the suppression chamber and the core suppression chamber unwatered shall include
'spray system shall be considered at least hourly checks that the conditions operable provided that the following listed in 3.1.4f are met.
are met:
- 1. No more than one control rod drive housing or instrument penetration will be opened at any time.
- 2. A blind flange will be installed on the control rod drive housing whenever a control rod drive has been removed for maintenance.
- 3. Work will not be performed in the reactor vessel while a control rod drive housing is open.
- 4. A control rod drive will not be removed if the backseat seal does not function.
- g. During reactor operation, except during g. At least once per month verification that core spray system survei llance testing, the piping system between valves 40-03, 13 core spray isolation valves 40-02 and and 40-01, 09, 10, 11 is filled with water.
40-12 shall be in the open position and the associated valve motor starter circuit breakers for these valves shall be locked in the off position. In addition, redundant valve position indication shall be available in the control room. Amendment No. Q, 83 Nine Mile Point - Unit 1 3/4 1"18
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT h.. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel, the reactor water level may be lowered to 9'elow the. minimum normal water level (elevation 302'9"). Mhenever the re-actor water level is to be lowered below the low-low-low level set point redundant instrumentation will be pro-vided to monitor the reactor water level and written procedures will be developed and followed whenever the reactor water level is lowered below the low-low level set point. The pro-cedures will define the valves that will be used to lower the vessel water level. All other valves that have the potential of lowering the vessel water level will be identified by valve number in the procedures and these valves will be red tagged to preclude their oper-ation during the major maintenance with the water level below the low-low level set point. During the period of major maintenance requiring lowering the water level to more than 6 feet, 3 inches below mini-mum normal water level (-10 inches indicator scale), either both Core
~
Spray Systems must be operable or, one Core Spray System is inoperable if because 'of the -maintenance, all of the redundant components of the other Core Spray System must be operable. Amendment No. 64 Nine Hile Point - Unit 1 3/4 1-19
0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.5 SOLENOID-ACTUATED PRESSURE RELIEF VALVES 4.1.5 SOLENOID-ACTUATED PRESSURE RELIEF VALVES (AUTOMATIC DEPRESSURIZATION SYSTEM) (AUTOMATIC DEPRESSURIZATION SYSTEM) Applies to the operational status of the Applies to the periodic testing requirements solenoid-actuated relief valves. for the solenoid-actuated pressure relief valves.
~0b ective: ~Ob ective:
To assure the capability of the solenoid- To assure the operability of the solenoid-actuated pressure relief valves to provide actuated pressure r el i e f val ves to per form a means of depressurizing the reactor in their intended functions. the event of a small line break to allow full flow of the core spray system. The solenoid-actuated pressure relief valve surveillance shall be performed as indicated bel ow.
- a. During the power operating condition a. At least once during each operating cycle whenever the reactor coolant pressure with the reactor at pressure, each Amendment No. 86 Nine Mile Point Unit 1 3/4 1-20
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT is greater than 110 psig and the valve shall be manually opened until acoustic reactor coolant temperature is monitor s or thermocouples downstream of the greater than saturation temperature, valve indicate that the valve has opened and all six solenoid-actuated pressure steam is flowing from the valve. relief valves shall be operable.
- b. If Specification 3. 1.5a above is not b. At least once during each operating met, the reactor coolant pressure and cycle, automatic initiation shall be the reactor coolant temperature shall demonstrated.
be reduced to 110 psig or less and saturation temperature or less respectively within ten hours. Amendment No. 86 Nine Hile Point - Unit 1 3/4 1-21
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.1.6 CONTROL ROD DRIVE PUMP COOLANT INJECTION 4.1.6 CONTROL ROD DRIVE PUMP COOLANT INJECTION Applies to the operational status of the Applies to the periodic testing requirements control rod drive pump coolant injection for the control rod drive pump coolant system. injection system.
~0b ective: ~0b ective:
To assure the capability of the control rod To assure the capability of the control rod drive pump coolant injection system to: drive pump coolant injection system in performing its intended functions. Provide core cooling in the, event of a small line break, and Provide coolant makeup in the event of reactor coolant leakage (see LCO 3. 2. 5). The control rod drive pump coolant injection system surveillance shall be performed as indicated below.
- a. Mhenever irradiated fuel is in the reactor vessel and the reactor
- a. At least once er o eratin c cle-coolant temperature is greater Automatic starting of each pump shall be than 212F, the control rod drive demonstrated.
pump coolant injection system shall be operable except as specified in "b" below. Nine Mile Point - Unit 1 3/4 1-22
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- b. If a redundant component becomes b. At least once er uarter-inoperable, the control rod drive pump coolant injection system shall Pump flow rate shall be determined.
be considered operable provided that the component is returned to an operable condition within 7 days and the additional surveillance required is performed.
- c. If Specifications "a" or "b" above are c. Surveillance with Ino erable Com onents not met, the reactor coolant tempera-ture shall be reduced to 212F or less Mhen a component becomes inoperable its within ten hours. redundant component shall be demonstrated to be operable immediately and daily thereafter.
Nine Mile Point Unit 1 3/4 1-23
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.7 FUEL RODS 4.1.7 FUEL RODS The Limiting Conditions for Operation asso- The Surveillance Requirements apply to the ciated with the fuel rods apply to those parameters which monitor the fuel rod parameters which monitor the fuel rod operating conditions. operating conditions.
~0b 'ecti ve: ~0b ective:
The objective of the Limiting Conditions The objective of th'e Surveillance Requirements for Operation is to assure the performance is to specify the type and frequency of sur-of the fuel rods. veillance to be applied to the fuel rods.
- a. Avera e Planar Linear Heat Generation a. Avera e Planar Linear Heat Generation ate L Rate A LHG During power operation, the APLHGR for The APLHGR for each type of fuel as a each type of fuel as a function of function of average planar exposure average planar exposure shall not exceed shall be determined daily during the limiting value shown in Figures reactor operation at > 25K rated 3.1.7a, 3.1.7b, 3.1.7c, 3.1.7d, 3.1.7e, thermal power.
and 3.1.7f. If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded at any node in the core, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits within two (2) hours, reactor power re-ductions shall be irotiated at a rate not less than 10'er hour until APLHGR at all nodes is within the prescribed limits. Amendment No. 47, 81 Nine Mile Point - Unit 1 3/4 1-24
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- b. Linear Heat Generation Rate LHGR b. Linear Heat Generation Rate (LHGR)
During power operation, the Linear Heat The LHGR as a function of core height Generation Rate (LHGR) of any rod in any shall be checked daily during reactor fuel assembly at any axial location operation at > 25K rated thermal power. shall not exceed 13.4 KW/FT. If at any time during power operation it is determined by normal survei llance that the limiting value for LHGR is being ex-ceeded at any location, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR at all locations is not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10X per hour until LHGR at all locations is within the prescribed limits. Amendment No. 8, 81, 41 Nine Mile Point - Unit 1 3/4 1-25
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UI REMENT C. Minimum Critical Power Ratio MCPR c. Minimum Critical Power Ratio MCPR During power operation, the HCPR for all MCPR shall be determined daily during 8 x 8 fuel at rated power and flow shall reactor power operation at >25K rated be as shown in the table below: thermal power. LIMITING CONDITION FOR OPERATING MCPR Core Average Incremental Limiting Ex osure HCPR" BOC to EOC minus 2 GWD/ST > l. 40 EOC minus 2 GWD/ST to EOC minus j. GWD/ST > 1.45 EOC minus 1 GWD/ST to EOC > 1.50 If at any time during power operation it is determined by normal surveillance that these limits are no longer met, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If all the operating MCPRs are not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10X per hour until MCPR is within the prescribed limits.
*These limits shall be determined to be applicable each operating cycle by analy-ses performed utilizing the ODYN transient code.
Amendment No. 59 Nine Mile Point - Unit 1 3/4 1-26
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT For core flows other than rated the MCPR limits shall be the limits iden-tified above times Kf where Kf is as shown in Figure 3.1.7-1.
- d. Power Flow Relationshi Durin 0 eration d. Power Flow Relationshi The power/flow relationship shall not Compliance with the power flow relation-exceed the limiting values shown in ship in Section 3. 1.7.d shall be determined Figure 3.1.7.aa. daily during reactor operation.
If at any time during power operation it is determined by normal surveillance that the limiting value for the power/ flow relationship is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the power/ flow relationship is not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than lOX per hour unti 1 the power/ flow relationship is within the prescribed limits.
- e. Partial Loo 0 eration e. Partial Loo 0 eration During power operation, partial loop Under partial loop operation, surveillance.
operation is permitted provided the requirements 4. 1. 7. a,b,c, and d above are following conditions are met. applicable. Nine Mile Point - Unit 1 3/4 1-27
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT When operating with four recirculation loops in operation and the remaining loop unisolated, the reactor may oper-ate at 100 percent of full licensed power level in'ccordance with Figure 3.1.7aa and an APLHGR not to exceed 98 percent of the limiting values shown in Figures 3.1.7a, 3.1.7b, 3.1.7c, 3.1.7d, and 3.1.7e: When operating with four recirculation loops in operation and one loop iso-lated, the reactor may operate at 100 percent of full licensed power in accordance with Figure 3. 1. 7aa and an APLHGR not to exceed 98 percent of the limiting values shown in Figures 3. 1.7a,
- 3. 1. 7b, 3.1. 7c, 3. 1. 7d, and 3. 1. 7e, provided the following conditions are met for the isolated loop.
- 1. Suction valve, discharge valve and discharge bypass valve in the iso-lated loop shall be in the closed position and the associated motor breakers shall be locked in the open position.
- 2. Associated pump motor circuit breaker shall be opened and the breaker removed.
If these conditions are not met, core power shall be restricted to 90.5 per-cent of full licensed power. Amendment No. 89, 47 Nine Mile Point - Unit 1
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT When operating with three recirculation loops in operation and the two remain-ing loops isolated or unisolated, the reactor may operate at 90 percent of full licensed power in accordance with Figure 3.1.7aa and an APLHGR not to exceed 96 percent of the limiting values shown in Figures 3. 1. 7a, 3. 1. 7b, 3.1.7c, 3.1.7d, and 3.1.7e. During 3 loop operation, the limiting HCPR shall be increased by 0.01. Power operation is not permitted with less than three recirculation loops in operation. If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR under one and two isolated loop oper-ation is being exceeded at any node in the core, action shall be initiated within 15 minutes to restore oper-ation to within the prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits for one and two isolated loop operation within two (2) hours, reactor power reduction shall be initiated at a rate not less than 10 percent per hour until APLHGR at all nodes is within the prescribed limits. Amendment No. 47 Nine Nile Point - Unit 1 3/4 1-29
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- f. Recirculation Loo s During all operating conditions with
'irradiated fuel in the reactor vessel, at least two (2) recirculation loop suction valves and their associated discharge valves will be in the full open position except when the reactor vessel is flooded to a level above the main steam nozzles or when the steam separators and dryer are removed.
- g. Re ortin Re uirements If any of the limiting values iden-tified in-Specification 3. 1.7a, b, c, d, and e are exceeded, a Reportable Occurrence Report shall be submitted.
If the corrective action is taken, as described, a thirty-day written report will meet the requirements of this Specification. Amendment No. 39 Nine Mile Point - Unit 1
Nine H)le Point Un)t 1 100 L)m)t)ng Power/Flow Lkm 60 20 0 60 100 f'c.".nt Rated Core Floe Figure 3.1.7aa LIMITING POWER FLOW LINE Amendment No. ZS, 99, 41 Nine Ni 1e Point - Unit 1 3/4 1"31
NINE MlLE POINI UN I I l 10 9 8.73 8.7) 8.64 8.60 8.57 8.67 8.58 8.50 8.49 8.58 8 C7'. ( Lal 7 10 l5 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GHD/ST) Figure 3. 1.7a MAXIMUM ALLOWABLE AVERAGE PLANAR LHGR APPLICABLE TO 8DB250 FUEL AS DESCRIBED IN REFERENCE 8. Amendment No. 8X, gx, 47 Nine Mi.1e Point - Unit 1 3/4 1-32
NINE MILE POINT UNIT 10 5 8.74 8. 70 8.65 8 62- 8.66 8.60 8.56 8.48 8. 48
- 8. 58 7
10 20 25 30 35 40 45 AVERAGE PLANAR EXPOSURE (GWD/ST) Figure 3.1.7b MAXIMUM ALLOWABLE AVERAGE PLANAR LHGR APPLICABLE TO SDB274L AND SDB274H FUEL AS DESCRIBED IN REFERENCE S. Amendment No. 8l, EZ, 47 Nine Mile Point - Unit 1 3/4 1-33
NINE HILE POINT UNIT 10 9.23 9.22 9.20 9.16
.24 9.13 9.12 8.83 8.55 10 15 20 25 30 35 40. 45 AVERAGE PLANAR EXPOSURE (GWD/ST)
Figure 3.1.7c HAXIHUH ALLONBLE AVERAGE PLANAR LHGR APPLICABLE TO 8DNB277 FUEL AS DESCRIBED IN REFERENCE 8. Amendment No. 8X, AX, 47 Nine Hile Point - Unit 1 3/4 1-34
NNE MILE POINT UNIT
- 9. 30 9,23 9.20 9.1i
.24 9. 06 8.90
- 8. 72
- 8. 46 0 10 15 20 25 r 30 35 40 AVERAGE PLANAR EXPOSURE (GWD/ST)
Figure 3. 1. 7d MAXIMUM ALLOWABLE AVERAGE PLANAR LHGR APPLICABLE TO PSDNB277 AND FUTURE RELOAD FUEL AS DESCRIBED IN REFERENCE 8. Amendment No. 87, 4i, 47 Nine Nile Point - Unit 1 3/4 1-35
Hlht. lllLt I'UJHI Unl I I 10 G. 73 8.71 8.57 8.67 8.64 8.60 8.58 8.58 I 0 10 15 20 25 30 35 40 AVEPAGE PLANAR EXPOSURE (G'!0/ST) Figure 3.1.7e MAXIMUM ALLOWABLE AVERAGE PLANAR LHGR APPLICABLE TO 8DB252 FUEL AS DESCRIBED IN REFERENCE 8. Amendment No. 8Z, 41 Nine Mile Point - Unit.l 3/4 1-36
MAPLHGR LIMITS R P8DRB299 10 9.15 . 9.15 9.15 9.13 9.10 LEGEND 8.67 8.61 8.51 MAPLHGR 8.20 b 7.65 lK 2 b C7 0 5 C7 1 E E
>C O
- 0. 5. 1 0. 1 5. 20. 25. 30. 35. 40. 4'5.
AVERAGE PLANAR EXPOSURE (GMD/STI Figure 3. 1.7f MAXIMUM ALLOWABLE AVERAGE PLANAR LHGR APPLICABLE TO PBDRB299 AND FUTURE RELOAD FUEL AS DESCRIBED IN REFERENCE 8. Amendment No. 8I, 81 Nine Mile Point - Unit 1 3/4 1-37
1.4 1.3 12 AUTOMATICFLOW CONTROL MANUALFLOW CONTROL SCOOP.TUBE SETO'OINT CALIBRATION POSITIONED SUCH THAT 1.0 F LOWMAX ~ 102.51I
~ I07.0%
~ 112.01I I 'I 7.0'4 30 a0 50 60 BD 90 100 120 CORE FLOW IIC)
Figure 3.1.7-1 Kf Factor Amendment No. N, 81 Nine Mile Point - Unit 1 3/4 1-38
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UI REHENT 3.1.8 HIGH PRESSURE COOLANT INJECTION 4.1.8 HIGH PRESSURE COOLANT INJECTION Applies to the operational status of the Applies to the periodic testing requirements high pressure coolant injection system. for the high pressure coolant injection system. if'he
~bb'ective: ~0b ective:
To assure the capability of the high To verify the operability of the high pressure pressure coolant injection system to coolant injection system. cool reactor fuel in the event of a loss-of-coolant accident. e high pressure coolaht injection sur-veillance shall be performed as indicated below:
- a. During the power operating condition a. At least once er o eratin c cle-whenever the reactor coolant pressure is greater than 110 psig and the reactor Automatic start-up of the high pressure coolant temperature greater than satu- coolant injection system shall be demon-ration temperature, the high pressure strated.
coolant injection system shall be operable except as specified in Specification "b" below.
- b. If a redundant component of the high b. At least once er uarter-pressure coolant injection system becomes inoperable the high pressure Pump operability shall be determined.
coolant injection shall be considered operable provided that the component is returned to an operable condition within 15 days and the additional sur-veillance required is performed. Nine Hile Point - Unit 1 3/4 1-39
0 LIMITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. If Specification "a" and "b" are not c. Surveillance with Ino erable Com onent met, a normal orderly shutdown shall be initiated within one hour and When a component becomes inoperable its reactor coolant pressure and temper- redundant component shall be demonstrated ature shall be reduced to less than to be operable immediately and daily 110 psig and saturation temperature thereafter.
within 24 hours. Nine Mile Point - Unit 1 3/4 1-40
Document Name: NMP-1 TS SEC 3/4 2 Requestor's ID: NORMA Author's Name: Jameson, C. Document Comments: ETPB Rev. 9/22/86 KEEP THIS SHEET WITH DOCUMENT
- 3. 2. 0 REACTOR COOLANT SYSTEM A) GENERAL APPLICABILITY Applies to the operating conditions of the reactor coolant system and its associated systems and components.
B) GENERAL OBJECTIVE LIMITING CONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of the systems which will assure the integrity of the reactor coolant system as a barrier against the uncontrolled release of radioactivity. SURVEILLANCE RE(UIREMENTS - To define the tests or inspections required to assure the functional capability or performance level of the above. Nine Mile Point - Unit 1 3/4 2-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.1 REACTOR VESSEL HEATUP AND COOLDOWN RATES Applies to the reactor vessel heating or cooling rate.
~0b 'ective:
To assure that thermal stress resulting from reactor heatup and cooldown are within allowable code limits. During the startup and shutdown operations of the reactor, the reactor vessel temperature shall not be increased more than 100 F in any one hour period nor decreased more than 100 F in any one hour period. Nine Mile Point - Unit 1 3/4 2-2
LIMITING CONDITION. FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 2. 2 MINIMUM REACTOR VESSEL TEMPERATURE FOR 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION PRESSURIZATION Applies to the minimum vessel temperature Applies to the required vessel temperature for required for vessel pressurization. press uri zati on.
~0b ective: ~0b 'ective:
To assure that no substantial pressure is To assure that the vessel is not subjected to imposed on the reactor vessel unless its any substantial pressure unless its temperature temperature is considerably above its Nil is greater than its Nil Ductility Transition Ductility Transition Temperature (NDTT). Temperature (NDTT).
- a. During reactor vessel heat-up and cool- a. Reactor vessel temperature and pressure-down when the reactor is not critical shall be monitored and controlled to assure the reactor vessel temperature and pres- that the pressure and temperature limits are sure shall satisfy the requirements of met.
Figure 3.2.2.a.
- b. During reactor vessel heat-up and cool- b. Vessel material surveillance samples located down when the reactor is critical the within the core region to permit periodic reactor vessel temperature and pressure monitoring of exposure and material properties shall satisfy the requirements of Figure shall be inspected on the following schedule:
3.2.2.b, except when performing low power physics testing with the vessel First capsule - one fourth service life head removed at power levels not to Second capsule three fourth service life exceed 5 mw(t). Third capsule - standby In the event the surveillance specimens at one quarter of the vessels service life indicate a shift of reference temperature greater than predicted the schedule shall be revised as follows: Amendment No. S8, 85 Nine Mile Point - Unit 1 3/4 2-3
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT Second capsule - one half service life Third capsule - standby
- c. During hydrostatic testing, the reactor vessel temperature and pressure shall satisfy the requirements of Figure
- 3. 2. 2. c if the core is not critical.
- d. The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head are equal to or greater than 100F.
Amendment No. S8, 85 Nine Hile Point - Unit 1 3/4 2-4
EA I OJ fT7 1400 1400 1200 Limit For Non-Crate'cal Operation Zncluding 1000 Beatup/CooLdown at up to 100 F/HR ~p 0 CL 800 600 400 236 200 X00 200 300
.Minimum Vessel Temperature (F)
Figure 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEATUP OR COOLDOWN (REACTOR NOT CRITICAL) O (HEATING.OR COOLING RATE < 1QQFlHR} FOR UP TO ELEVEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION
TABLE 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING H AT-UP OR COOLOOWN REACTOR NOT CRI CAL HEATING OR COOLING RATE 100F HR
~
FOR UP TO ELEVEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION LIMIT FOR NON-CRITICAL OPERATION INCLUDING HEAT-UP/COOLDOWN AT UP TO 100F/HR PRESSURE si ) TEMPERATURE F) 236 100 300 135 350 154 400 170 450 182 500 192 550 201 600 209 650 217 700 223 750 229 800 234 850 240 900 244 950 248 1000 253 1050 256 1100 260 1150 263 1200 267 1300 273
'400 279 Amendment No. 98, 85 Nine Mile Point - Unit 1 3/4 2-6
I C4 1600 1400 1200 1105 U Limit For
~g
~ el vj 0 Power Operation Cl (Core Cr'tical) Zncludinc 1000 Heatup/Cooldown, at o 4 0 up t,o l40 Z/HR o ~ N c 4 800 C 600 Q o l5 400 Mate Level Nust 200 Be in Normal 194: Operatina Band For Core to he C"'tical at Temperatures
~200 F 0 100 200 300 Hinimum Vessel Tempe'ture (F)
Figure 3.2.2.b MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEATUP OR COOLDOWN (REACTOR CRITICAL) O (HEATING OR COOLING RATE < 100F/HR) FOR UP TO ELEVEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION
TABLE 3.2.2.b MINIMUM TEMPERATURE FOR PRESSURIZATION DURING H A -UP OR COOLDOWN REA OR CR CAL HEATING OR COOLING RA E 100 HR FOR UP TO ELEVEN EFFECTIVE FULL POWER Y ARS CORE OPERATION LIMIT FOR POWER OPERATION (CORE CRITICAL) INCLUDING HEAT-UP/ COOLDOWN AT UP TO 100F/HR PRESSURE si ) TEMPERATURE F) 194 100 250 150 300 176 350 194 400 210 450 222 500 232 550 241 600 249 650 257 700 263 750 269 800 274 850 280 900 284 950 288 1000 293 1050 296 1100 300 1150 303 1200 307 1300 313 1400 319 Amendment No. 98, 85 Nine Mile Point - Unit 1 3/4 2-8
1600 1400 1400 Limit For Inservice Test (Core Not Critical, 1200 Fuel in Vessel} 1000 800 628 600 400 200 100 130 200 300 Hinimuin Vessel Temperature {F) Figure 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING (REACTOR NOT CRITICAL) FOR UP TO ELEVEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION
TABLE 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING REACTOR NOT CRITICAL FOR UP TO ELEVEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION LIMIT FOR IN-SERVICE TEST (CORE NOT CRITICAL, FUEL IN VESSEL) PRESSURE ( si TEMPERATURE F) 360 100-130 628 130 700 152 800 174 900 190 1000 204 1050 210 1100 215 1150 2-20 1200 225 1300 233 1400 241 Amendment No. Sg, 85 Nine Mile Point - Unit 1 3/4 2-10
0 0
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2. 3 COOLANT CHEMISTRY 4.2. 3 COOLANT CHEMISTRY Applies to the reactor coolant system Applies to the periodic testing requirements chemical requirements. of the reactor coolant chemistry.
~0b ective: ~0b ective:
To assure the chemical purity of the To determine the chemical purity of the reactor coolant water. reactor coolant water.
~Siti
- a. The reactor coolant water shall not Samples shall be taken and analyzed for con-exceed the following limits with -
ductivity and chloride ion content at least steaming rates less than 100,000 pounds 3 times per week with a maximum time of 96 hours per hour except as specified in 3.2.3c: between samples. In addition, if the conduc-tivity becomes abnormal (other than short term Conductivity 2 pmho/cm spikes) as indicated by the continuous conduc-Chloride ion 0. 1 ppm tivity monitor, samples shall be taken and analyzed within 8 hours and daily thereafter until conductivity returns to normal levels.
- b. The reactor coolant water shall not exceed the following limits with steam- When the continuous conductivity monitor is ing rates greater than or equal to inoperable, a reactor coolant sample shall be 100,000 pounds per hour except as speci- taken and analyzed for conductivity and chloride fied in 3.2.3c: ion content at least once per 8 hours.
Conductivity 5 ymho/cm Chloride ion 0.2 ppm Amendment No. 9 Nine Mile Point - Unit 1 3/4 2-11
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. The limits specified in 3.2.3a and 3.2.3b may be exceeded for a period of time not to exceed 24 hours. In no case shall (1) the conductvity exceed a maximum limit of 10 pmho/cm, or (2) the chloride ion concentration exceed a maximum limit of 0.5 ppm.
- d. If Specifications 3.2.3.a, b, and c are not met, normal orderly shutdown shall be initiated within one hour and the reactor shall be in the cold shutdown condition within ten hours.
- e. If the continuous conductivity monitor is inoperable for more than 7 days the reactor shall be placed in the cold shutdown condition within 24 hours.
Amendment No. 9 Nine Mile Point - Unit 1 3/4 2-12
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.4 REACTOR COOLANT ACTIVITY 4.2.4 REACTOR COOLANT ACTIVITY Applies to the limits on reactor coolant Applies to the periodic testing requirements activity at all operating conditions. of the reactor coolant activity.
~0b ective: ~0b'ective:
To assure that in the event of a reactor To assure that limits on coolant activity coolant system line break outside the are not exceeded. drywell permissible doses are not exceeded.
- a. The reactor coolant system radioactivity a. Samples shall be taken at least every concentration in water shall not exceed 96 hours and analyzed for gross gamma 25 microcuries of total iodine per gram activity.
of water.
- b. If Specification 3.2.4 a, above, cannot b. Isotopic analyses of samples shall be made be met after a routine surveillance check, at least once per month.
the reactor shall be placed in the cold shutdown condition within ten hours. Nine Mile Point - Unit 1 3/4 2-13
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT
- 3. 2. 5 REACTOR COOLANT SYSTEM LEAKAGE 4.2.5 REACTOR COOLANT SYSTEM LEAKAGAE A 1icabil it Applies to the limits on reactor coolant Applies to the monitoring of reactor coolant system leakage rate and leakage detection system leakage.
! systems.
~0b ective: ~0b 'ective:
To assure that the makeup capability To determine the reactor coolant system provided by the control rod drive pump is leakage rate and assure that the leakage not exceeded. limits are not exceeded.
- a. Any time irradiated fuel is in the a. A check of the reactor coolant leakage reactor vessel and the reactor shall be made very four hours.
temperature is above 2120F, reactor coolant leakage into the primary containment shall be limited to:
- 1. Five gallons per minute unidentified leakage.
- 2. A two gallon per minute increase in unidentified leakage within any period of 24 hours or less.
- 3. Twenty-five gallons per minute total leakage (identified plus unidentified) averaged over any 24 hour period.
Amendment No. 70 Nine Nile Point - Unit 1 3/4 2-14
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- b. Any time irradiated fuel is in the b. The following surveillance shall be performed reactor vessel and reactor coolant on each leakage detection system:
temperature is above 212~F, at least one of the leakage measurement channels l. An instrument calibration once associated with each sump (one for the each refueling outage. drywell floor drain and one for the equipment drain) shall be operable. 2. An instrument functional test once every three months. If conditions a or.b cannot be met, the reactor will be placed .in the cold shutdown condition within 24 hours. Amendment No. 70 Nine Mile Point - Unit 1 3/4 2-15
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.6 INSERVICE INSPECTION AND TESTING 4.2.6 INSERVICE INSPECTION AND TESTING Applies to components which are part of Applies to the periodic inspection and testing the reactor coolant pressure boundary and of components which are part of the reactor their supports and other safety-related coolant pressure boundary and their supports pressure vessels, piping, pumps, and and other safety-related pressure vessels, valves. piping, pumps, and valves.
~0b 'ective: ~0b ective:
To assure the integrity of the reactor To verify the integrity of the reactor coolant coolant pressure boundary and the pressure boundary and the operational readiness operational readiness of safety-related of safety-related pressure vessels, piping, pressure vessels, piping, pumps, and valves. pumps, and valves.
- a. Inservice Ins ection a. Inservice Ins ection
- 1. To be considered operable, Quality 1. Inservice inspection of Quality Group A, Group A, B and C components shall B and C components shall be performed in satisfy the requirements contained accordance with the requirements for ASME in Section XI of the ASME Boiler Code Class 1, 2 and 3 components, re-and Pressure Vessel Code and appli-- spectively, contained in Section XI of cable Addenda for continued service the ASME Boiler and Pressure Vessel Code of ASME Code Class 1, 2 and 3 com- and applicable Addenda as required by ponents, respectively, except where 10 CFR 50, Section 50.55a(g), except relief has been granted by the where relief has been granted by the Commission pursuant to 1(1CFR 50, Commission pursuant to 1%it)FR Part 50 Section 50. 55a(g)(6)(i). Section 50. 55a(g)(6)(i).
Amendment No. 57 Nine Mile Point - Unit 1 3/4 2-16
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 2. An augmented inservice inspection program shall be performed in accordance with the schedules contained in NUREG 0313 Revision 1. The augmented inservice inspection program shall be performed on service sensitive components. The fol-lowing systems contain service sensitive components: core spray, shutdown cooling, emergency condensers, liquid poison, reactor head spray and control rod drive return.
- b. Inservice Testin b. Inservice Testin
- 1. To be considered operable, Quality 1. Inservice testing of Quality Group A, Group A, B and C pumps and valves B and C pumps and valves shall be shall satisfy the requirements con- performed in accordance with the tained in Section XI of the ASME requirements for ASME Code Class 1, 2 Boiler and Pressure Vessel Code and and 3 components contained in Section applicable Addenda for continued XI of the ASME Boiler and Pressure service of ASME Code Class 1, 2 and Vessel Code and applicable Addenda as 3 components, respectively, except required by 10 CFR 50, Section 50.55a(g),
where relief has been granted by the except where relief has been granted Commission pursuant to 1[0)FR 50, by the Commission pursuant to 10C@R Section 50.55a(g)(6)(i). Part 50 Section 50.55a(g)(6)(i). Amendment No. 57 Nine Mile Point - Unit 1 3/4 2-$ 7
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 2. 7 REACTOR COOLANT SYSTEM ISOLATION VALVES 4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Applies to the operating status of the Applies to the periodic testing requirement system of isolation valves on lines con- for the reactor coolant system isolation nected to the reactor coolant system. valves.
~0b ective: ~0b'ective:
To assure the capability of the reactor To assure the capability of the reactor coolant system isolation valves to minimize coolant system isolation valves to minimize reactor coolant loss in the event of a reactor coolant loss in the event of a rupture of a line connected to the nuclear rupture of a line connected to the nuclear steam supply system. steam supply system. l~ifi The reactor coolant system isolation valves surveillance shall be performed as indica-ted below (see Table 3. 2. 7).
- a. During power operating conditions a. At least once er o eratin c cle the whenever the reactor head is on, all opera e automatica y snstiate power-reactor coolant system isolation operated isolation valves shall be valves on lines connected to the reactor tested for automatic initiation and coolant system shall be operable except closure times.
as specified in "b" below.
- b. In the event any isolation valve becomes b. At least once er uarter all normally inoperable the system shall be considered open power-operated iso ation valves operable provided at least one valve in (except the feedwater and main-steam-each line having an inoperable valve is line power-operated isolation valves) in the mode corresponding to the isolated shall be fully closed and reopened.
condition. Nine Mile Point - Unit 1 3/4 2-18
LIMITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. If Specifications 3. 2. 7a and b above c. At least twice er week the feedwater are not met, initiate normal orderly and masn-steam- >ne power-operated shutdown within one-hour and have isolation valves shall be exercised reactor in the cold shutdown condition by partial closure and subsequent within ten hours. reopening.
- d. At least once er uarter the scram
>sc arge system acr operated vent and drain valves shall be fully closed and reopened.
Amendment No. 43 Nine Mile Point - Unit 1 3/4 2-19
LIMITING CONDITIONS FOR OPERATION TABLE 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Haximum Action on Initiating Signal No. of Valves to Primary Normal Oper. Time Initiating (All Valves Have
~tine or S stem ~Each Line Containment Position Motive Power ~sec ~SI nai Remote Hanual Backu Main Steam 1 Inside Open A. I.P.O." 10 Close Reactor water level low-Etwvo me s) 1 Outside Open A. I. P.O. " 10 Close low, or main steam line high radiation, or main steam line high flow, or Hain Steam Warm-u 1 Outside Closed A. I.P.O Close low condenser vacuum, or wo ines high tempelature in the pipe tunnel.
Hain Steam-Emer enc Coolin Vents wO lneS 2 Outside Open A. I.P.O 5 Close Feedwater 1 Outside Open R. P.O." 60 ~omen) 1 Outside Self Act. Ck. Emer enc Coolin Steam Leavin Reactor 1 Outside Open A. I.P.O. 38 Close wo 1 mes 1 Outside Open A.I.P.O. 38 Close Nigh system flow Condenser Return to Reactor 1 Inside Self Act. Ck. wo ines 1 Outside Cl os ed A. I.P.O. 60 Close Reactor Cleanu Water Leavin Reactor 1 Inside Open A. I.P.O. 18 Close Reactor water level ne lne 1 Outside Open A. I.P.0. 18 Close low-low, or high area temperature, liquid Water Return to Reactor 1 Inside Open A. I.P.0. 18 Close poison initiation or ne lne 1 Outside Self Act. Ck. high system pressure, or low system flow, or high system temperature Amendment No. 60 Nine Mile Point - Unit 1 3/4 2-20
LIMITING CONDITIONS FOR OPERATION TABLE 3.2.7 (Cont'd) REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Action on Initiating Signal No. of Valves to Primary Normal Oper. Time Initiating (All Valves Have
~ttne or 5 stem ~tach tine "
Containment Position Shutdown Coolin Water Leavin Reactor Inside Closed A. I. P.O. 40 Close ne ne Outside Closed A. I.P.O. 40 Close Reactor water level low-low, or high area Water Return to Reactor Inside Closed A. I.P.O. 40 Close temperature ne ne Outside Self Act. Ck. Reactor Head S ra Inside Self Act. Ck. ne- ne Outside Closed R.H.P.O. 30 Li uid Poison Inside Self Act. Ck. ne ne Outside Self Act. Ck. Control Rod Drive H draulic Inside Self Act. Ck. ne ne Outside Self Act. Ck. Core S ra Hi h Point Vent Inside Closed A.C. Hotor 30 Close Reactor Water Level wo nes Outside Closed Air/D.C. 30 Close Low-Low or High Solenoid Drywell Pressure "A. I.P.O. - Automatically Initiated Power Operated "R.M.P.O. - Remote Hanual Power Operated Amendment No. 44 Nine Hile Point - Unit 1 3/4 2-21
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.7.1 PRIMARY COOLANT SYSTEM PRESSURE 4.2.7.1 PRIMARY COOLANT SYSTEM PRESSURE L N VAL LA N AL Applies to the operating status of isolation Applies to the periodic testing of primary valves for systems connected to the primary coolant system pressure isolation valves. coolant system.
~0b 'ective: ~0b 'ective:
To increase the reliability of primary cool- To increase the reliability of primary coolant ant system pressure isolation valves thereby system pressure isolation valves thereby reducing reducing the potential of an intersystem loss the potential of an intersystem loss of coolant of coolant accident. accident.
- a. The integrity of all pressure isolation a. Periodic leakage testing on each valve valves listed in Table 3.2.7.1 shall be listed in Table 3.2.7.1 shall be accomplished demonstrated. Valve leakage shall not prior to exceeding 2X power while in the exceed the amounts indicated. power operating condition every time the plant is placed in a cold shutdown condition for refueling, each time the plant is placed in a cold shutdown condition for 72 hours testing has not been accomplished in the if preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is performed.
- b. If Specification a cannot be met, an orderly shutdown shall be initiated within 1 hour and the reactor shall be 7 i f in the cold shutdown condition within measured indirectly (as from the performance of 10 hours. pressure indicators) if accomplished in accor-dance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
Order Dated: April 20, 1981 Nine Mile Point - Unit 1 3/4 2-22
TABLE 3.2.7. 1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum
~Setem Valve No. Allowable Leaka e
- 1. Core Spray System 40-03 <5.0 gpm 40"13 <5 0 gpm
~
- 2. Condensate Supply to 40"20 <5.0 gpm Core Spray 40-21 <5.0 gpm (Keep Fill System) 40-22 <5.0 gpm 40"23 <5.0 gpm Footnote:
(a) 1. Leakage rates less than or equal to 1.0 are considered acceptable. gpm
- 2. Leakage rates greater than l. 0 gpm but less than or equal to 5. 0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50X or greater.
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5. 0 gpm by 50K or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
- 5. Test differential pressure shall not be less than 150 psid.
Order Oated: April 20, 1981 Nine Mile Point Unit 1 3/4 2-23
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.8 PRESSURE RELIEF SYSTEMS-SAFETY VALVES 4.2.8 PRESSURE RELIEF SYSTEMS-SAFETY VALVES Applies to the operational status of the Applies to the periodic testing requirements safety valves. for the safety valves.
~bb ective: ~0b ective:
To assure the capability of the safety To assure the capability of the safety valves valves to limit reactor overpressure to limit reactor overpressure to below the below the safety limit in the event of safety limit. rapid reactor isolation and failure of all pressure relieving devices. e gati
- a. During power operating conditions and At least once during each operating cycle at whenever the reactor coolant pressure least eight of the sixteen safety valves shall is greater than 110 psig and temperature be removed, tested for set point and partial greater than saturation temperature all sixteen of the safety valves shall be lift, and then returned to operation or replaced.
operable.
- b. If Specification 3. 2.8a is not met, the reactor coolant pressure and temperature shall be reduced to 110 psig or less and saturation temperature or less, respec-tively, within ten hours.
Nine Mile Point - Unit 1 3/4 2-24
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.2.9 PRESSURE RELIEF SYSTEMS - SOLENOID- 4.2.9 PRESSURE RELIEF SYSTEMS - SOLENOID-ACTUATED PRESSURE RELIEF VALVES ACTUATED PRESSURE RELIEF VALVES OVERPRESSURIZATION Applies to the operational status of the Applies to the periodic testing requirements solenoid-actuated pressure relief valves. for the solenoid-actuated pressure relief val ves.
~0b ective: ~0b ective:
To assure the capability of the solenoid- To assure the operability of the solenoid-actuated pressure relief valves to limit actuated pressure relief valves to limit reactor overpressure below the lowest safety reactor overpressure in the event of rapid valve setpoint in the event of rapid reactor isolation. reactor isolation. The solenoid-actuated pressure relief valve surveillance shall be performed as indicated below.
- a. During the power operating condition a. The setpoints of the six relief valves and whenever the reactor coolant shall be as follows:
pressure is greater than 110 psig and temperature greater than satura- No. of tion, five of the six solenoid- Valves ~Set oint actuated pressure relief valves shall be operable. 2 <1090 psig 2 <1095 psig 2 <1100 psig Amendment No. 86 Nine Mile Point - Unit 1 3/4 2-25
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 2.9 PRESSURE RELIEF SYSTEMS - SOLENOID- 4.2.9 PRESSURE RELIEF SYSTEMS SOLENOID-ACTUATED PRESSURE RELIEF VALVES ACTUATED PRESSURE RELIEF VALVES OVERPRESSURIZATION
- b. If Specification 3. 2. 9a is not met, b. At least once during each operating cycle the reactor coolant pressure and with the reactor at pressure, each valve temperature shall be reduced to 110 shall be manually opened until acoustic psig or less and saturation temper- monitors or thermocouples downstream of ature or less, respectively, within the valve indicate that the valve has ten hours. opened and steam is flowing from the valve.
- c. At least once during each operating cycle, relief valve setpoints shall be verified.
Amendment No. 86 Nine Mile Point - Unit 1 3/4 2-26
V Document Name: NMP-1 TS SEC 3/4 3 Requestor's ID: CYNTHIA Author's Name: Jamerson, C Document Comments: ETPB REV 9/9/86 Please return this sheet with revisions
- 3. 3. 0 PRIMARY CONTAINMENT APPLICABILITY Applies to the operating status of the primary containment systems.
OBJECTIVE To assure the integrity of the primary containment systems. SPECIFICATION Primary containment, integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 215F and fuel is in the reactor vessel except while per-forming low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 Mwt. Nine Mile Point - Unit 1 3/4 3-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 3.1 OXYGEN CONCENTRATION 4.3.1 OXYGEN CONCENTRATION Applies to the limit on oxygen concentra- Applies to the periodic testing requirement tion within the primary containment system. for the primary containment system oxygen concentration.
~0b'ective: ~0b ective:
To assure that in the event of a loss-of- To assure that the oxygen concentration coolant accident any hydrogen generation within the primary containment system is will not result in a combustible mixture within required limits. within the primary containment system.
- a. After completion of the startup test At least once a week oxygen concentration program and demonstration of plant shall be determined.
electrical output, the primary con-tainment atmosphere shall be reduced to less than four percent oxygen with nitrogen gas whenever the reactor coolant pressure is greater than 110 psig and the reactor is in the power operating condition, except as specified in "b" below. Nine Mile Point - Unit 1 3/4 3-2
LIMITING CONDITION FOR OPERATION ~ SURVEILLANCE REQUIRBIENT
- b. Within the 24-hour period subsequent to the reactor being placed in the run mode for the power operating condition, the containment atmosphere oxygen con-centration shall be reduced to less than four percent by weight, and main-tained in this condition. Deinerting may commence 24 hours prior to a major refueling outage or other scheduled shutdown.
- c. If Specifications "a" or "b" above are not met, the reactor coolant pressure shall be reduced to 110 psig or less within ten hours.
Nine Nile Point - Unit 1
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT 3.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE 4. 3. 2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL TEHPERATURE AND LEVEL Applies to the interrelated parameters of Applies to the periodic testing of the pressure suppression system pressure and pressure suppression system pressure and suppression chamber water temperature and suppression chamber water temperature level. and level.
~0b ective: ~0b ective:
To assure that the peak suppression chamber To assure that the pressure suppression pressure does not exceed design values in system pressure and suppression chamber the event of a loss-of-coolant accident. water temperature and level are within required limits.
- a. The downcomers in the suppression a. At least once per day the suppression chamber shall have a minimum submergence chamber water level and temperature and of three feet and a maximum submergence pressure suppression system pressure of four and one half feet whenever the shall be checked.
reactor coolant system temperature is above 215F.
- b. During normal power operation, the com- b. A visual inspection of the suppression bination of primary containment pressure chamber interior, including water line and suppression chamber bulk pool tem- regions, shall be made at each major perature shall be within the shaded area refueling outage.
of (1) Figure 3. 3. 2a when downcomer submergence is greater than or equal to 4 feet, or (2) Figure 3.3.2b when downcomer submergence is greater than or Amendment No." 76 Nine Nile Point Unit 1 3/4 3-4
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UI REMENT equal to 3 feet but less than 4 feet. If these temperatures are exceeded, pool cooling shall be initiated immediately. C. If Specifications a and b above are not c. Whenever heat from relief valve operation met within 24 hours, the reactor shall is being added to the suppression pool, be shut down using normal shutdown the pool temperature shall be continually procedures. monitored and also observed and logged every 5 minutes until the heat addition is terminated.
- d. During testing of relief valves which d. Whenever operation of a relief valve is add heat to the torus pool, bulk pool indicated and the bulk suppression pool temperature shall not exceed 10F above temperature reaches 160F or above while normal power operation limit specified the reactor primary coolant system pres-in-b above. In connection with such sure is greater than 200 psig, an external testing, the pool temperature must be visual examination of the suppression reduced within 24 hours to below the . chamber shall be made before resuming normal power operation limit specified normal power operation.
in b above.
- e. The reactor shall be scrammed from any e. Whenever there is indication of relief operating condition when the suppression valve operation with the local temperature pool bulk temperature reaches 110F. of the suppression pool reaching 200F or Operation shall not be resumed until the more, an external visual examination of pool temperature is reduced to below the the suppression chamber shall be conducted normal power operation limit specified before resuming normal power operation.
in b above. During reactor isolation conditions, the reactor pressure vessel shall be depres-surized to less than 200 psig at normal cooldown rates if the pool bulk tempera-ture reaches 120F. Amendment No. N, 76 Nine Mile Point Unit 1 3/4 3-5
Figure 3.2.2 a ALLOWABLE PRESSURE SUPPRESSION SYSTEM 4 FOOT DOWNCOMER SUBMERGENCE 5.0 4.0 C) 3.0 C) W I V) 2.0 tA C) Vl M CL 1.0 CA 0 68 72 76 80 84 88 92 96 '00 SUPPRESSION CHAMBER MATER OPERATIONAL TEMPERATURE F Amendment No. N, 76 Nine Mile Point - Unit 1 3/4 3-6
Figure 3.3.2- b ALLOMABLE PRESSURE SUPPRESSION SYSTEM 3 FOOT DOWNCOMER SUBMERGENCE 5.0 CD C/7 CL I CCJ CC lA 4.0 l/l UJ CC A Cl I- 3.0 C) tU I Cll 2.0 ID W CL 0 68 72 76 80 84 88 92 96 100 SUPPRESSION CHAMBER MATER OPERATIONAL TEMPERATURE F Amendment No. 28, 76 Nine Mile Point - Unit 1 3/4 3-7
0 LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT
- 3. 3. 3 LEAKAGE RATE 4. 3. 3 LEAKAGE RATE A licabilit Applies to the allowable leakage rate Applies to the primary containment system of the primary containment system. leakage rate.
~0b ective: ~0b ective:
To assure the capability of the containment To verify that the leakage from the primary in limiting radiation exposure to the public containment system is maintained within from exceeding values specified in 10 CFR 100 specified values. in the event of a loss-of-coolant accident accompanied by significant fuel cladding failure and hydrogen generation from a metal-water reaction. Mhenever the reactor coolant system tem- a. Integrated Primary Containment Leakage perature is above 215 F the primary con- Rate Test tainment leakage rate shall be within the limits of 4.3.3.b. (1) Integrated leak rate tests shall be performed prior to initial Station operation at the test pressure of 35 psig (P ) and the test pressure (P ) of 22 psig to obtain the respective measured leak rates L (35) and L (22). (2) Subsequent leakage rate tests shall be performed without preliminary leak detection surveys or leak repairs immediately prior to or during the test, at an initial pressure of approximately 22 psig. Nine Nile Point - Unit 1 3/4 3-8
0 0
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (3) Leak repairs, if necessary to permit integrated leakage rate testing, shall be preceded by local leakage measure-ments. The leakage rate difference, prior to and after repair when cor-rected to P shall be added to the final integer'ated leakage rate result. (4) Closure of the containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves. (5) After the containment test conditions have stabilized, the test duration shall not be less than eight hours for integrated leak rate measurements. The test shall be extended for sufficient duration, to verify by a supplemental test method the accuracy of the inte-grated leak rate test results. (b) Acce tance Criteria (1) The maximum allowable leakage rate L shall not exceed 1.5 weight percent Bf the contained air per 24 hours at the test pressure of 35 psig (P ). (2) The allowable test leak rate L (22) shall not exceed the value established as follows: Lt (22) = 1.5 L (22)/L (35) Nine Mile Point Unit Amendment No. 52 1 3/4 3-9
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT (3) The allowable operational leak rate, L (22) which shall be met prior to resumption of power operation following a test (either as measured or following repairs and retest) shall not exceed 0'75 Lt 22
- c. Corrective Action If leak repairs are necessary to meet the allowable operational leak rate, the inte-grated leak rate test need not be repeated provided local leakage measurements are conducted, and the leak rate differences prior to and after repairs, when corrected to Pt and deducted from the integrated leak rate measurement, yield a leakage rate value not in excess of the allowable operational leak rate Lt (22).
- d. Ere<rfuenc r Three integrated leak rate tests shall be performed at approximately equal intervals during each 10-year service period with the third test in each ten-year interval cor-responding with the ten-year scheduled in-service inspection shutdown.
- e. Local Leak Rate Tests (1) Primary containment testable penetra-tions and isolation valves shall be tested at a pressure of 35 psig each major refueling outage except bolted Nine Nile Point - Unit 1 3/4 3-10
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT double-gasketed seals shall be tested whenever the seal is closed after being opened, and at least at each refueling outage. (2) Personnel air lock door seals shall be tested once within 24 hours after opening when the reactor is in a power operating condition, at a pressure of 10 psig arid the leak rate extrapolated to 35 psig. Air lock seals shall also be leak rate tested at a pressure of 35 psig at the beginning of each operating 'cycle. An additional 35 psig leak rate test shal.l be performed near the middle of the operating cycle should a shutdown requiring de-inerting arise. If the above shutdown does not occur or is not anticipated, the air lock seals will be tested at 10 psig. In each test the leak rate corrected to 35 psig shall not exceed 5 percent L . a (3) Containment components not included in (1) and (2) which required leak repairs following any integrated leakage rates in order to meet the allowable leakage rate unit L shall be subjected to local leak tests $ t a pressure of 35 psig at each refueling outage. (4) The main steam line isolation valves are to be tested at a pressure of 35 psig during each refueling outage. Nine Nile Point - Unit 1 3/4 3-11
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- f. Corrective Action (1) If the total leakage rates listed below as adjusted to a test pressure of 22 psig are exceeded, repairs and retests shall be performed to correct the condition.
(a) double-gasketed seals lOX Lt 22 (b) (i) testable penetrations and isolation valves 30K Lt (22) (ii) any one penetration or. isolation valve
)
(c) primary containment air purge penetrations and reactor building to torus vacuum relief valves 50~ Lt (22
- g. Continuous Leak Rate Monitor (1) When the primary containment is inerted the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements.
Nine Mile Point Unit 1 3/4 3-12
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as soon as practical.
- h. ~Ine ection The accessible interior surfaces of the drywell shall be visually inspected each operating cycle for evidence of deterioration.
Nine Mile Point - Unit 1 3/4 3-13
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.4 PRIMARY CONTAINMENT ISOLATION VALVES 4.3.4 PRIMARY CONTAINMENT ISOLATION VALVES A 1icabil it: A 1 icabi1 it: Applies to the operating status of the sys- Applies to the periodic testing requirements tem of isolation valves on lines open to the of the primary containment isolation valve free space of the primary containment. system.
~0b ective: ~0b ective:
To assure that potential leakage paths from To assure the operability of the primary con-the primary containment in the event of a tainment isolation valves to limit potential loss-of-coolant accident are minimized. leakage paths from the containment in the event of a loss-of-coolant accident. S ecification: S ecification: The primary containment isolation valves sur-veillance shall be performed as indicated (see Table 3.3.4)
- a. Whenever the reactor coolant system tem- a. At least once per operating cycle the operable perature is greater than 215F, all con- isolation valves that are power operated and tainment isolation valves on lines open automatically initiated shall be tested for to the free space of the primary contain- automatic initiation and closure times.
ment shall be operable except as speci-fied in 3.3.4b below.
- b. In the event any isolation valve becomes b. At least once per quarter all normally open inoperable the system shall be considered power operated isolation valves shall be operable provided at least one valve in fully closed and reopened.
each line having an inoperable valve is in the mode corresponding to the isolated condi ti on. Nine Mile Point - Unit 1 3/4 3-14
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- c. If Specifications 3.3.4 a and b are not c. At least once per operating cycle, each met, the reactor coolant system tempera- instrument-line flow check valve will be ture shall be reduced to a value less tested for operability.
than 215F within ten hours. Nine Nile Point - Unit 1 3/4 3-15
LIHITING CONDITION FOR'OPERATION TABLE 3.3.4 PRIMARY CONTAINMENT ISOLATION VALVES LINE NIE RING H N A NMENT Initiating Signal Location Relative Maximum Action on (All Valves Have No. of Valves to Primary Normal Oper. Time Initiating Remote Manual
~tine or 5 stem ~Each Line Containment Position Hotive Power 0 ell Vent 8 Pur e N2 Connection Outside Closed (a) Air/D.C. Sol. 60 Close
~ne me Outside Closed (a) A.C. Hotor 60 Close Reactor water level low-low or drywell Air Connection Outside Closed (a) Air/D.C. Sol. 60 Close high pressure ne >ne Outside Closed (a) A.C. Hotor 60 Close Su ression Chamber Vent 6 Pur e N2 Connection Outside Closed (a) Air/D.C Sol. 60 Close ne L>ne Outside Closed (a) A.C. 60 Close Reactor water level Hotor'losed low-low or drywell Air Connection Outside (a) Air/D.C. Sol. 60 Close high pressure ne L>ne Outside Closed (a) A.C. Motor 60 Close Drywell N2 Hakeup Reactor water level ne L>ne Outside Closed (b) Air/D.C. Sol. 60 Close low-low or drywell high pressure Su ression Chamber a eup Reactor water level (Ofe Line) Outside Closed (b) Air/D.C. Sol. 60 Close low-low or drywell high pressure Nine Mile Point. - Unit I 3/4 3-16
LIMITING CONDITION FOR OPERATION TABLE 3.3.4 (Cont'd) PRIMARY CONTAINMENT ISOLATION VALVES LIN MENT Initiating Signal Location Relative Haximum Action on (All Valves Have No. of Valves to Primary Normal Oper. Time Initiating Remote Hanual
~tine or S steo ~Each Line Contatuoent Position Motive Power ~sec ~st nal Backu 0 ell E ui ment Drain Line Inside Open A.C. Hotor 60 Close ne ne Outside Open Air/D.C. Sol. 60 Close Reactor water level low-low or drywel 1 high pressure Floor Drain Line Inside Open A.C. Hotor 60 Close ne one Outside Open Air/D.C. Sol. 60 Close Su ression Chamber Hater makeu Outside Closed (b) A.C. Motor 60 Remote manual ne >ne Outside Self Act. Ck.
Negative pressure Vacuum Relief relative to atmos- ~osp ere to Pressure
'Three Lines)
Suppression Systeu Outside Outside Closed A.C. Hotor Self Act. Ck. Open phere Reactor Cleanu S stem Relief Valve Oischar e ne ne to uppressson am er Outside Self Act. Ck. o2 Sampling Orywell Outside Closed (b) O.C. Sol. 60 Close Reactor water level (Three Lines) low-low or high drywell pressure Su ression Chamber ne ine Outside Closed (b) D.C. Sol. 60 Close NOTES: (a) These valves may be open for containment fill with nitrogen. (b) These valves will periodically be opened for sampling or nitrogen makeup. Nine Hile Point - Unit I 3/4 3-17
LIMITING CONDITION FOR OPERATION TABLE 3.3.4 (Cont'd) PRIMARY CONTAINHENT ISOLATION VALVES LINE MENT Initiating Signal Location Relative Maximum Action on (All Valves Mave No. of Valves to Primary Normal Oper. Time Initiating Remote Manual
~ttoe or 5 stem ~Each tree Containment Position Hotive Power
~Core 5 ra (c) Pum our Suction
>nes from Suppression Chamber) Outside 'pen AC Hotor 90 Remote manual Pum ) Disc~har e Reactor water level worst~>nes to Suppression Chamber) 1 Outside Closed AC Motor 90 Close low-low LINES WITH A CLOSED LOOP INSIDE CONTAINMENT VESSELS Recir. Pum Coolin Water Su 1 (c) upp y one Outside Open Self Act. Ck.
Return Line Outside Open DC Hotor 30 Remote manual 0 ell Cooler Water Su 1 (c) upp y >ne 1 - Outside Open Self Act. Ck. Return Line 1 . Outside Open DC Motor 30 Remote manual LINES WITH A CLOSED LOOP OUTSIDE CONTAINMENT VESSELS Containment S ra (c) Reactor level low-e u ression Chamber Common Su 1 low and high dry-our snes Outside Open Air/DC Sol. 60 Open well pressure Nine Hi le Point - Unit 1 3/4 3-18
LIMITING CONDITION FOR OPERATION TABLE 3.3.4 (Cont'd) PRIMARY CONTAINMENT ISOLATION VALVES LINE MENT Initiating Sign'al Location Relative Maximum Action on (All Valves Have No. of Valves to Primary Normal Oper. Time Initiating Remote Manual
~tine or S stem ~Each tine Contairment Position Hotive Power Drywell Branch (Po~ur sees Outside Self Act. Ck.
Su ression Chamber Branch ne rane or ac ystem) Outside Self Act. Ck. P Suction From Su ression Chamber our Ines Outside Open AC Motor 70 Remote manual "One Valve in each separate line and one valve in each common line. (c) These are classified as not"testable valves and penetrations. Nine Mile Point - Unit 1 3/4 3-19
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.5 ACCESS CONTROL 4.3.5 ACCESS CONTROL A licabilit : Applies to the access control to the Applies to the surveillance on primary containment primary containment. access control.
~0b ective: ~0b ective:
To assure the integrity of the primary To assure the operability of the primary contain-containment system. ment access control interlocks. S ecification: S ecification: Mhenever the. reactor coolant system temper- A mechanical. interlock will be maintained to ature is above 215F the following shall be prevent simultaneous opening of two doors. in effect.
- a. Only one door in each of the two double-door drywell access locks will be opened at one time.
- b. The equipment hatch and drywell head and other flanged openings will be secured.
- c. If following a routine surveillance check "a" or "b" is not met, initiate normal orderly shutdown within one hour and have reactor in the cold shutdown condition within ten hours.
Nine Mile Point - Unit 1 3/4 3-20
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.3. 6 VACUUM RELIEF 4.3.6 VACUUM RELIEF A licabilit : A 1icabil it: Applies to the operational status of the Applies to the periodic testing of the vacuum primary containment vacuum relief system. relief system.
~bb
'e cti ve: ~bb'ective:
To assure the capability of the vacuum To assure the operability of the containment relief system in the event of a loss-of- vacuum relief system to perform its intended ~ coolant accident to: functions.
- a. Equalize pressures between the dry-well and suppression chamber, and
- b. Maintain containment pressure above the vacuum design values of the dry-well and suppression chamber.
e~if i
- a. Mhen primary containment is required, a. Periodic 0 erabilit Tests all suppression chamber-drywell vacuum breakers shall be operable except during Once each month and following any release testing and as stated above. Suppression of energy to the suppression chamber, each chamber-drywell vacuum breakers shall be suppression chamber-drywell vacuum breaker considered operable if: shall be exercised. Operability of valves, position switches, and position indicators (1) The valve is demonstrated to open and alarms shall be verified monthly and fully with the applied force at all following any maintenance on the valves and valve positions not exceeding that associated equipment.
equivalent to 0. 5 psi acting on the suppression chamber face of the valve disk. Nine Mile Point - Unit 1 3/4 3-21
0 0
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIRENENT (2) The valve can be closed by gravity, when released after being opened by remote or manual means, to within not greater than, the equivalent of 0.06 inch at the bottom of the di sk. (3) The position alarm system will annunciate in the control room if the valve opening exceeds the equivalent of 0. 06 inch at the bottom of the disk.
- b. Any drywell-suppression chamber vacuum b. Refuelin Outa e Tests breaker may be non-fully closed as in-dicated by the position indication and (1) All suppression chamber-drywell vacuum alarm systems provided that drywell to . breakers shall be tested to deter-suppression chamber differential pres- mine the force required to open each sure decay rate is demonstrated to be valve from fully closed to fully not greater than 25K of the differential open.
pressure decay rate for all vacuum break-ers open the equivalent of 0.06 inch at (2) All suppression chamber-drywell vacuum the bottom of the disk. breaker position indication and alarm systems shall be calibrated and func-tionally tested. (3) Once each operating cycle, each vacuum breaker valve shall be visually in-spected to ensure proper maintenance and operation. (4) A drywell to suppression chamber leak rate test shall demonstrate that with an initial differential pressure of not'less than 1.0 psi, the differen-tial pressure decay rate shall not exceed the equivalent of the Nine Nile Point - Unit 1 3/4 3-22
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT leakage rate through a 1-inch orifice. C. When it is determined that one or more c. Pressure Su ression Chamber-Reactor vacuum breaker valves are not fully Bui 1 ds n Vacuum Breakers closed as indicated by the position indication system at a time when such (1) The pressure suppression chamber-closure is required, the apparently mal- reactor building vacuum breaker functioning vacuum breaker valve shall systems and associated instrumenta-be exercised and pressure tested as tion, including set point, shall be specified in 3.3.6 b immediately and checked for proper operation every every 15 days thereafter until appro- three months. priate repairs have been completed. (2) During each refueling outage, each vacuum breaker shall be tested to determine that the force required to open the vacuum breaker does not exceed the force specified in Speci-fication 3.3.6.f(l) and each vacuum breaker shall be inspected and veri-fied to meet design requirement.
- d. One drywell-suppression chamber vac-uum breaker may be secured in the closed position.
- e. If Specifications 3.3.6 a, b, c, or d cannot be met, the situation shall be corrected within 24 hours or the reactor shall be placed in a cold shut-down condition within 24 hours.
Nine Mile Point - Unit 1 3/4 3-23
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- f. Pressure Su ression Chamber-Reactor u1 sn acuum reakers The three pressure suppression chamber-reactor building vacuum breaker systems shall be operable at all times when the primary containment integrity is required.
The set point of the differential pressure 'instrumentation which actuates the pressure suppression chamber-reactor building air-operated vacuum breakers shall be < 0.5 psid. The self-actuating vacuum breakers shall open fully when subjected to a force equivalent to or less than 0. 5 psid acting on the valve disk. (2) From and after the date that one of the pressure suppression chamber-reactor building vacuum breaker systems is made or found inoperable for any reason, the vacuum breaker shall be locked closed and reactor operation is permissible only during the succeeding seven (7) days unless such vacuum breaker system is sooner made operable, provided that the procedure does not violate contain-ment integrity. Nine Nile Point - Unit 1
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.3.7 CONTAINMENT SPRAY SYSTEM 4.3.7 CONTAINMENT SPRAY SYSTEM Applies to the operating status of the Applies to the testing of the containment containment spray system. spray system.
~0b ective: ~0b ective:
To assure the capability of the containment To verify the operability of the containment spray system to limit containment pressure spray system. and temperature in the event of a loss-of-coolant accident. The containment spray system surveillance shall be performed as indicated below:
- a. During all reactor operating conditions a. Containment Spray Pumps whenever reactor coolant temperature is greater than 215 F and fuel is in the (1) At least once er o eratin c cle, reactor vessel; each of the two contain- automatic startup of the containment ment spray systems and the associated spray pump shall be demonstrated.
raw water cooling systems shall be operable except as specified in 3. 3. 7. b. (2) At least once er uarter, pump operabs sty shall be checked.
- b. If a redundant component of a containment b. Nozzles spray system becomes inoperable, Specifi-cation 3.3.7.a shall be considered ful- At least once per operating cycle, an filled, provided that the component is air test shall be performed on the returned to an operable condition within spray headers and nozzles.
15 days and that the additional survei 1-lance required is performed. Nine Mile Point - Unit 1 3/4 3-25
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. If a redundant component in each of the c. Raw Mater Cooling Pumps containment spray systems or their asso-ciated raw water systems become inoper- At least once per quarter manual startup able both systems shall be considered and operability of the raw water cooling operable provided that the component is pumps shall be demonstrated.
returned to an operable condition within 7 days and that the additional surveil-lance required is performed.
- d. If a containment spray system or its d. Sur veil lance with Inoperable Components associated raw water system becomes in-operable and all the components are Mhen a component or system becomes in-operable in the other systems, the re- operable its redundant component or actor may remain in operation for a system shall be demonstrated to be period not to exceed 7 days. operable immediately and daily thereafter.
- e. If Specifications "a" or "b" are not met, e. Surveillance during control rod drive shutdown shall begin within one hour and maintenance which is simultaneous with the the reactor coolant shall be below 215 F suppression chamber unwatered shall within ten hours. include at least hourly checks that the conditions listed in 3.3.7.f are met.
If both containment spray systems become inoperable the reactor shall be in the cold shutdown condition within ten hours and no work (except as specified in "f" below) shall be performed on the reactor which could result in lowering the reactor water level to more than six feet, three inches (-10 inches indicator scale) below minimum normal water level (Elevation 302'"). Amendment No. 64 Nine Nile Point - Unit 1 3/4 3-26
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- f. Mork may be performed on control rod drives at times when water is not in the suppression chamber and the containment spray system shall be considered oper-able provided the following are met:
A (1) No more than one control rod drive housing or instrument penetration will be opened at any time. (2) A blind flange will be installed on the control rod drive housing whenever a control rod drive has been removed for maintenance. (3) Mork will not be performed in the reactor vessel while a control rod drive housing is open. (4) A control rod drive will not be removed if the backseat seal does not function. (5) A minimum condensate storage volume of 300,000 gallons and a minimum hot well storage volume of 40,000 gallons will be maintained during the period that the torus water level is below that corresponding to minimum NPSH requirement. (6) The control rod drive removal and instrument replacement shall not be concurrent items. Amendment No. 83 Nine Mile Point - Unit 1
Document Name: NMP"1 TS SEC 3/4 4 Requestor's ID: NORMA Author's Name: Jamerson, C. 1 Document Comments: PH-364 Revised 9/8/86 /KEEP THIS SHEET WITH DOCUMENT
3.4. 0 REACTOR BUILDING APPLICABILITY Applies to the operating status of the reactor building. OBJECTIVE To assure the integrity of the reactor building. SPECIFICATION Reactor building integrity must be in effect in the refueling and power operating conditions and also whenever irradiated fuel or the irradiated fuel cask is being handled in the reactor building. Nile Point Nine Unit 1 3/4 4-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.4.1 LEAKAGE RATE 4.4.1 LEAKAGE RATE A 1 icabi1 it Applies to the leakage rate of the secondary Applies to the periodic testing requirements containment. of the secondary containment leakage rate.
~0b ective: ~0b ective:
To specify the requirements necessary to limit To assure the capability of the secondary exfiltration of fission products released to containment to maintain leakage within the secondary containment as a result of an al 1 owabl e 1 imits. accident. Mhenever the reactor is in the refueling or Once dur in each o eratin c cle - isolate power operating condition, the reactor build- the reactor busldsng and start emergency ing leakage rate as determined by Specification ventilation system fan to demonstrate 4.4. 1 shall not exceed 2000 cfm. If this negative pressure in the building relative cannot be met after a routine surveillance to external static pressure. The fan flow check, then the actions listed below shall be rate shall be varied so that the building taken: internal differential pressure is at least as negative as that on Figure 3.4.1 for the
- a. Suspend immediately irradiated fuel wind speed at which the test is conducted.
handling, fuel pool and reactor cavity The fan flow rate represents the reactor activities, and irradiated fuel cask building leakage referenced to zero mph with handling operations in the reactor building internal pressure at least 0.25 building. inch of water less than atmospheric pressure. The test shall be done at wind speeds less
- b. Restore the reactor building leakage rates than 20 miles per hour.
to within specified limits within 4 hours or initiate normal orderly shutdown and be in a cold shutdown condition within 10 hours. Amendment No. 38 Nine Mile Point - Unit 1 3/4 4-2
-0.25 -0.35 8 10 12 '4 16 18 20 MIND SPEED (MPH) 167 Figure 3.4.1 Reactor Building Pressure Nine Nile Point - Unit 1 3/4 4-3
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.4.2 REACTOR BUILDING INTEGRITY - ISOLATION 4.4.2 REACTOR BUILDING INTEGRITY - ISOLATION VALVES VALVES Applies to the operational status of the Applies to the periodic testing requirements reactor building isolation valves. of the reactor building isolation valves.
~00 ective: ~0b'ective:
To assure that fission products released to To assure the operability of the reactor the secondary containment are discharged to building isolation valves. the environment in a controlled manner using the emergency ventilation system.
- a. The normal Ventilation System isolation At least once per operating cycle, automatic valves shall be operable whenever the initiation of valves shall be checked.
reactor is in the refueling or power operating conditions, and whenever irrad-iated fuel or the irradiated fuel cask is being handled in the reactor building.
- b. If Specification 3.4. 2a is not met, the reactor shall be in the cold shutdown condition within ten hours and handling of irradiated fuel cask shall cease.
Nine Mile Point - Unit 1 3/4 4-4
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.3 ACCESS CONTROL 4.4.3 ACCESS CONTROL Applies to the access control to the reactor Applies to the periodic checking of the bui 1 ding. condition of portions of the reactor building.
~0b ective: ~0b ective:
To specify the requirements necessary to To assure that pump compartments are properly assure the integrity of the secondary contain- closed at all times. ment system.
- a. Only one door in each of the double-doored a. The core and containment spray pump com-access ways shall be opened at one time. partments shall be checked once per week
~
and after each entry.
- b. Only one door or closeup of the railroad bay shall be opened at one time.
- c. The core spray and containment spray pump compartments'oors shall be closed at all times except during passage in order to consider the core spray system and the containment spray system operable.
Amendment No. 1 Nine Mile Point - Unit 1 3/4 4-5
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.4. 4 EMERGENCY VENTILATION SYSTEM 4.4. 4 EMERGENCY VENTILATION SYSTEM A 1icabilit A licabilit Applies to the oper ati ng status of the Applies to the testing of the emergency emergency ventilation system. venti lation system.
~0b ective: ~0b ective:
To assure the capability of the emergency To assure the operability of the emergency ventilation system to minimize the release of ventilation system. radioactivity to the environment in the event of an incident within the primary containment or reactor building. Emergency venti lation system survei llance shall be performed as indicated below:
- a. Except as specified in Specification a. At least once per operating cycle, not 3.4.4e below, both circuits of the to exceed 24 months, the following emergency ventilation system and the conditions shall be demonstrated:
diesel generator s required for opera-tion of such circuits shall be operable (1) Pressure drop across the combined at all times when secondary containment HEPA filters and charcoal adsorber integrity is required. banks is less than 6 inches of water at the system rated flow rate (+ 10K). (2) Operability of inlet heater at rated power when tested in accordance with ANSI N.510-1980. Amendment No. 73 Nine Mile Point - Unit 1 3/4 4-6
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- b. The results of the in-place cold DOP and b. The tests and sample analysis of Specifi-halogenated hydrocarbon tests at design cation 3.4.4b, c and d shall be performed flows on HEPA filters and charcoal ad- at least once per operating cycle or once sorber banks shall show >99K DOP removal every 24 months, or after 720 hours of and >99K halogenated hydrocarbon removal system operation, whichever occurs first when tested in accordance with ANSI or following, significant painting, fire or N.510-1980. chemical release in any ventilation zone communicating with the system.
C. The results of laboratory carbon sample c. Cold DOP testing shall be performed analysis shall show >90K radioactive after each complete or partial replace-methyl iodide removaT when tested in ment of the HEPA filter bank or after accordance with ANSI N.510-1980 at 80'C any structural maintenance on the system and 95K R.H. housing. Fans shall be shown to operate within d. Halogenated hydrocarbon testing shall be 210K design flow. performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing. From and after the date that one circuit e. Each circuit shall be operated with the of the emergency ventilation system is inlet heater on at least 10 hours every made or found to be inoperable for any month. reason, reactor operation and fuel handling is permissible only during the succeeding seven days unless such circuit is sooner made operable, provided that during such seven days all active com-ponents of the other emergency venti la-tion circuit shall be operable. If these conditions cannot be met, within f. Test sealing of gaskets for housing 36 hours, the reactor shall be placed in doors downstream of the HEPA filters a condition for which the emergency and charcoal adsorbers shall be performed ventilation system is not required. at and in conformance with each test per-formed for compliance with Specification 4.4.4b and Specification 3.4.4b. Amendment No. 73 Nine Mile Point - Unit 1 3/4 4-7
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREHENT
- g. At least once per operating cycle, not to exceed 24-months, automatic initiation of each branch of the emergency ventila-tion system shall be demonstrated.
- h. At least once per operating cycle, not to exceed 24 months, manual operability of the bypass valve for filter cooling shall be demonstrated.
When one circuit of the emergency venti-lation system becomes inoperable all active components in the other emergency ventilation circuit shall be demonstrated to be operable within 2 hours and daily thereafter. Amendment No. 73 Nine Hile Point Unit 1 3/4 4-8
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.5 CONTROL ROOM AIR TREATMENT SYSTEM - 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM Applies to the operating status of the Applies to the testing of the control room control room air treatment system. air treatment system.
~0b ective: ~0b ective:
To assure the capability of the control room To assure the operability of the control air treatment system to minimize the amount room air treatment system. of radioactivity or other gases entering the control room in the event of an incident.
- a. Except as specified in Specifiction a. At least once per operating cycle, or 3.4. 5e below, the control room air treat- once every 24 months, whichever occurs ment system and the diesel generators first, the pressure drop across the required for operation of this system combined HEPA filters and charcoal shall be operable at all times when con- adsorber banks shall be demonstrated to tainment integrity is required. be less than 6 inches of water at system design flow rate (+lOX).
- b. The results of the in-place cold DOP and b. The tests and sample analysis of halogenated hydrocarbon test design flows Specification 3.4.5b, c and d shall be on HEPA filters and charcoal adsorber performed at least once per operating banks shall show >99K DOP removal and cycle or once every 24 months, or after
>99K halogenated hydrocarbon removal when 720 hours of system operation, whichever tested in accordance with ANSI N.510-1980. occurs first or following significant painting, fire or chemical release in any ventilation zone communicating with the system.
Amendment No. 73 Nine Mile Point - Unit 1 3/4 4-9
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. The results of laboratory carbon sample c. Cold DOP testing shall be performed analysis shall show >90% radioactive methyl after each complete or partial iodide removal when tested in accordance replacement of the HEPA filter bank with ANSI N.510-1980 at 80'C and 95K R.H. or. after any structural maintenance on the system housing.
- d. Fans shall be shown to operate within d. Halogenated hydrocarbon testing shall be flow. +10'esign performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing.
- e. From and after the date that the control e. The system shall be operated at least room air treatment system is made or found 10 hours every month.
to be inoperable for any reason, reactor operation or refueling operations is per-missible only during the succeeding seven days unless that system is sooner made operable.
- f. If these conditions cannot be met, reactor f. At least once per operating cycle, not shutdown shall be initiated and the reactor to exceed 24 months, automatic initia-shall be in cold shutdown within 36 hours tion of the control room air treatment for reactor operations and refueling opera- system shall be demonstrated.
tions shall be terminated within 2 hours. Amendment No. 73 Nine Nile Point - Unit 1 3/4 4-10
Document Name: NMP-1 TS SEC 3/4 5 Requestor's ID: CYNTHIA Author's Name: Jamerson, C. Document Comments: ETPB Rev. 9/ll/86 KEEP THIS SHEET WITH DOCUMENT
- 3. 5.0 SHUTDOWN AND REFUELING A) GENERAL APPLICABILITY Applies to the neutron instrumentation systems required during shutdown and refueling operations.
B) GENERAL OBJECTIVE LIMITING CONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of equipment required during shutdown and refueling operations. SURVEILLANCE RE(UIREMENTS - To define the test or inspections required to assure the functional capability or performance level of the above items. Nine Nile Point Unit 1 3/4 5-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5. 1 SOURCE RANGE MONITORS 4.5.1 SOURCE RANGE MONITORS Applies to the operating status of the source Applies to the periodic testing of the range monitors. source range monitors.
~bb ective: ~0b 'ective:
To assure the capability of the source range To assure the operability of the source monitors to provide neutron flux indication range monitors to monitor low-level required for reactor shutdown and startup neutron flux. and refueling operations. Mhenever the reactor is in the shutdown, The source range monitoring system surveil-refueling or power oper ating conditions lance will be performed as indicated below. (unless the IRH's or APRM's are on scale) or whenever core alterations are being made at Durin each o eratin c cle - check in-core least three SRM channels will be operable to out-of-core signal ratio and minimum except as noted in Specification 3. 5. 3. To count rate. be considered operable, the following condi-tions must be satisfied:
- a. Inserted to normal operating level and available for monitoring the core. Hay be withdrawn as long as a minimum count rate of 100 cps is maintained.
- b. A 3/1 in-core to out-of-core signal ratio and a minimum count rate of 3 cps at a ke ff equi val ent to the initial cl ean core with all rods and poison control curtains inserted.
Amendment No. 27 Nine Mile Point - Unit 1 3/4 5-2
LIMITING CONOITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- c. If following a routine surveillance check "a" or "b" is not met, the reactor shall be in the cold shutdown condition within ten hours.
Nine Mile Point - Unit 1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5. 2 REFUELING PLATFORM INTERLOCK 4. 5. 2 REFUELING PLATFORM INTERLOCK Applies to the refueling platform on inter- Applies to the periodic testing requirements locks. for the refueling platform interlocks.
~0b ective: ~0b ective:
To,assure that a loaded refueling platform To assure the operability of the refueling hoist is never over the core when one or platform interlock. more control rods are withdrawn. e S ecification: if'uring the refueling condition with the mode The refueling platform interlocks shall be switch in the "refuel" position the following tested prior to any fuel handling with the interlocks must be operative: head off the reactor vessel, at weekly intervals thereafter until no longer required
- a. Control rod withdrawal block with a fuel and following any repair work associated assembly on the hoist over the reactor with the interlocks.
core.
- b. With a control rod withdrawn from the core the refuel platform, if loaded with a fuel assembly, is blocked from travel-ling over the core.
- c. If the interlocks for either "a" or "b" or both are not operable, double proce-dural control will be used to ensure that "a" and "b" are met.
Nine Mile Point - Unit 1 3/4 5-4
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.5.3 EXTENDED CORE AND CONTROL ROD DRIVE 4.5.3 EXTENDED CORE AND CONTROL ROD DRIVE Applies to core reactivity limitations during Applies to monitoring during major core major core alterations. alterations'0b
~0b 'ective: ective:
To assure that inadvertent criticality does To assure that inadvertent withdrawal of not result when control rods are being removed an incorrect control rod does not occur. from the core.
~l'ati e' Whenever the reactor is in the refueling 'henever the reactor is in the refuel mode condition, control rods may be withdrawn from and rod block interlocks are being bypassed the reactor core provided the following con- for core unloading, one licensed operator ditions are satisfied: and a member of the reactor analysis staff will verify that all the fuel from the cell
- a. The reactor mode switch shall be locked in has been removed before the corresponding the "Refuel" position. The refueling control rod is withdrawn.
interlock input signal from a withdrawn control rod may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core. All other refueling inter-locks shall be operable, except those necessary to pull the next control rods. Amendment No. 27 Nine Mile Point - Unit 1 3/4 5-6
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- b. During core alterations two SRH's shall be operable, one in and one adjacent to any core quadrant where fuel or control rods are being moved. Operable SRM's shall have a minimum of 3 counts per second except as specified in d and e below.
C. The SRM's shall be inserted to the normal operating level. Use of special movable dunking type detectors during major core alterations is permissible as long as detector is connected into the normal SRH circuit.
- d. Prior to spiral unloading, the SRM's shall have an initial count rate of 3 cps. During sprial unloading, the count rate on the SRH's may drop below 3 cps.
- e. During sprial reload, SRM operability will be verified by using a portable external source every 12 hours until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, two fuel assemblies will be loaded in different cells containing control blades around each SRM to obtain the required 3 cps. Until these two assemblies have been loaded, the 3 cps requirement is not necessary.
Amendment No. 27 Nine Mile Point - Unit 1
Document Name: NMP"1 TS SEC 3/4 6 Requestor's ID: CYNTHIA Author's Name: Jamerson, C. Document Comments: ETPB Rev. 9/11/86 PLEASE RETURN THIS SHEET WITH REVISIONS
- 3. 6. 0 GENERAL REACTOR PLANT A) GENERAL APPLICABILITY Applies to Station process effluents, reactor protection system and emergency power sources.
- 8) GENERAL OBJECTIVE LIMITING CONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of the equipment to assure overall Station safety.
SURVEILLANCE RE(UIREMENTS - To define the test or inspection required to assure the functional capability or performance level of this equipment. Nine Mile Point - Unit 1 3/4 6-1
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- 3. 6.1 STATION'ROCESS EFFLUENTS 4. 6. 1 STATION PROCESS EFFLUENTS
- a. Effluent release limits are described in a. Monitoring the radioactive discharges from Specification 3. 6. 15. Nine Mile Point Unit 1 is described in Specification 4. 6. 15.
- b. The mechanical vacuum pump line shall be b. At least once during each operating cycle capable of automatic isolation by closure (prior to startup), verify automatic of the air-operated valve upstream of the securing and isolation of the mechanical pumps. The signal to initiate isolation vacuum pump.
shall be from high radioactivity (five times normal) in the mainsteam line.~ Mithin 24 hours prior to the planned start of the hydrogen injection test with the reactor power at greater than 20K rated power, the normal full-power radiation background level and associated trip and alarm setpoints may be changed based on a calculated value of the radiation level expected during the test. The background radiation level and associated trip and alarm setpoints may be adjusted during the test program based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip .and alarm setpoints shall be reset within 24 hours of re-establishing normal radiation levels after completion of the hydrogen injection or within 12 hours of establishing reactor power levels below 20X rated power, while these functions are required to be operable. At reactor power levels below 20K rated power, hydro-gen injection shall be terminated and the injection system secured. Amendment No. 88, 87 Nine Mile Point - Unit 1 3/4 6-2
LIMITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 2 PROTECTIVE INSTRUMENTATION 4. 6. 2 PROTECTIVE INSTRUMENTATION Applies to the operability of the plant Applies to the surveillance of the instrumenta-instrumentation that performs a safety tion that performs a safety function. function.
~0b'ective: ~0b ective:
To assure the operability of the instru- To verify the operability of protective instru-mentation required for safe operation. mentation.
- a. The set points, minimum number of trip a. Sensors and instrument channels shall be systems, and minimum number of instru- checked, tested and calibrated at least ment channels that must be operable for as frequently as listed in Tables 4.6.2a each position of the reactor mode switch to 4.6.2m.
shall be as given in Tables 3.6.2a to 3.6.2m. If the requirements of a table are not met, the actions listed below for the respective type of instrumentation shall be taken. (1) Instrumentation that initiates scram - control rods shall be inserted, unless there is no fuel in the reactor vessel. Amendment No. 87, 73 Nine Mile Point - Unit 1 3/4 6-3
0 LIMITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Primary Coolant and Containment Isolation - Isolation valves shall be closed or the valves shall be considered inoperable and Spec-ifications 3.2.7 and 3.3.4 shall be applied. (3) Emergency Cooling Initiation or Isolation - The emergency cooling system shall be considered inop- . erable and Specification 3. 1. 3 shall be applied. (4) Core Spray Initiation - The core spray system shall be considered inoperable and Specification 3.1.4 shall be applied. (5) Containment Spray Initiation - The containment spray system shall be considered inoperable and Spec-ification 3.3.7 shall be applied.
'I (6) Auto Oepressurization Initiation--
The auto depressurization system shall be considered inoperable and Specification 3.1.5 shall be applied. (7) Control Rod Withdrawal Block - No control rods shall be withdrawn. Amendment No. 5 Nine Mile Point Unit 1
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (8) Off-Gas and Vacuum Pump Isolation-The respective system shall be isolated or the instrument channel shall be considered inoperable and Specification 3.6.1 shall be applied. (9) Diesel Generator Initiation - The diesel generator shall be considered inoperable and Specification 3. 6. 3 shall be applied. (10) Emergency Ventilation Initiation - The emergency ventilation system shall be considered inoperable and Specification 3.4.4 shall be applied. (11) High Pressure Coolant Injection Initiation - The high pressure coolant injection system shall be considered inoperable and Specification 3. 1. 8. c shall be applied. (12) Primary Containment Monitoring - The primary containment monitoring instrumentation shall be considered inoperable and Specification 3.3.2. shall be applied. (13) Control Room Ventilation - The control room ventilation system shall be considered inoperable and Specification 3.4. 5 shall be applied.
- b. During operation with a Maximum Total b. Each trip system shall be tested each Peaking Factor (MTPF) greater than the time the respective instrument channel design value, either: is tested.
Amendment No. 78, 76 Nine Mile Point - Unit 1 3/4 6-5
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT (1) The APRM scram and rod block settings shall be reduced to the values= given by the equations in Specification 2.1.2.a; or (2) The power distribution shall be changed such that the MTPF no longer exceeds the design value. C. At least daily during reactor power operation, the core power distribution shall be checked for Maximum Total Peaking Factor (MTPF) and the flow-referenced APRM scram and rod block signals shall be adjusted, if necessary, as specified in Figure 2. 1. 1. Amendment No. 9, 73 Nine Mile Point - Unit 1 3/4 6-6
TABLE 3.6.2a INSTRUMENTATION THAT INITIATES SCRAM Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (1) Manual Scram x x x (2) - High Reactor Pressure 1080 psig x x x (3) High Drywell Pressure 2 3.5 psig x (a) (a) (4) Low Reactor Water Level 2(m) 53 inches x x x (Indicator Scale) (5) High Water Level Scram < 45 gal. (b) x x Oischarge Volume Nine Mile Point Unit Amendment No. 48, 79 1 3/4 6-7
TABLE -3.6.2a (Cont'd) INSTRUMENTATION THAT INITIATES SCRAM Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (6) Hain-Steam-Line 2 4(h) <10 percent (c) (c) x Isolation Valve valve closure Position from full open (7) High Radiation <5 times normal x x x Main-Steam-Line background rated power g) (8) Shutdown Position (k) x x of Reactor Mode Switch (9) Neutron Flux (a) IRM (i) Upscale 3(d) <96 percent of (g) (g) (g) full scale Amendment No. 8T, 87 Nine Mile Point - Unit 1 3/4 6-8
TABLE 3.6.2a (Cont'd) INSTRUMENTATION THAT INITIATES SCRAM Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which .
. Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (ii) Inoperative 3(d) X X (b) APRM (i) Upscale 3(e) Figure 2.1.1 X X X (ii) Inoperative 3(e) X X X (iii) Downscale 3(e) >5 percent of (g) (g) (9) 7ull scale (10) Turbine Stop <lOX val ve Valve Closure closure (11) Generator Load Rejection Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-9
0 TABLE 4.6.2a INSTRUMENTATION THAT INITIATES SCRAM Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) Manual Scram None Once per 3 months None (2) High Reactor None Once per month Once per 3 months Pressure (3) High Drywell None Once per month Once per 3 months Pressure (4) Low Reactor Water Once/day Once per month Once per 3 months Level (5) High Water Level None Once per month Once per 3 months Scram Discharge Volume (6) Main-Steam-Line None Once per 3 months Once per operating Isolation Valve cycl e Position (7) High Radiation Once/shift Once per week Once per 3 months Main-Steam Line Amendment No. 61 Nine Mile Point - Unit 1 3/4 6-10
TABLE 4.6.2a (Cont'd) INSTRUMENTATION THAT INITIATES SCRAM Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (8) Shutdown Position None Once during each None of Reactor Mode major refueling Switch outage (9) Neutron Flux (a) IRM (i) Upscale (ii) Inoperative (b) APRM (i) Upscale None Once/week Once/week (ii) Inoperative None Once/week Once/week (iii) Downscale None Once/week Once/week (10) Turbine Stop None Once per 3 months Once per operating Valve Closure cycle (11) Generator Load None Once per month Once per 3 months Rejection Amendment No. 61 Nine Mile Point - Unit 1 3/4 6-11
NOTES FOR TABLES 3.6.2a and 4.6.2a (a) May be bypassed when necessary for containment inerting. (b) Hay be bypassed in the refuel and shutdown positions of the reactor mode switch with a keylock switch. (c) Hay be bypassed in the refuel and startup positions of the reactor mode, switch when- reactor pressure is less than 600 psi. (d) No more than one of the four IRH inputs to each trip system shall be bypassed. (e) No more than two C or D level LPRH inputs to an APRM shall be bypassed and only four LPRM inputs to an APRM shall be bypassed in order for the APRM to be considered operable. No more than one of the four APRM inputs to each trip system shall be bypassed provided that the APRH in the other instrument channel in the same core quadrant is not bypassed. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRH input if the TIP is positioned in close proximity to the failed LPRM it is replacing. (f) Calibrate prior to starting and normal shutdown and thereafter check once per shift and test once per week until no longer required. (g) IRM's are bypassed when APRM's are onscale. APRM downscale is bypassed when IRM's are onscale. (h) Each of the four isolation valves has two limit switches. Each limit switch provides input to one of two instrument channels in a single trip system. (i) Hay be bypassed when reactor power level is below 45K. (j) Trip upon loss of oil pressure to the acceleration relay. (k) Hay be bypassed when placing the reactor mode switch in the SHUTOOWN position and all control rods are fully inserted. (1) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2a, the primary sensor will be calibrated and tested once per operating cycle. (m) One instrument channel in each trip system may be bypassed in the cold shutdown and refuel conditions during the Spring 1986 refueling outage to perform the emergency condenser piping replacement. Amendment No. 87, 79 Nine Mile Point - Unit 1 3/4 6-12
NOTES FOR TABLES 3.6.2a and 4.6.2a (con't.) (n) Within 24 hours prior to the planned start of the hydrogen injection test with the reactor power at greater than 20K rated power, the normal full-power radiation background level and associated trip and alarm setpoints may be changed based on a calculated value of the radiation level expected during the test. The background radiation level and associated trip and alarm setpoints may be adjusted during the test program based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip and alarm setpoints shall be reset within 24 hours of re-establishing normal radiation levels after completion of the hydrogen injection or within 12 hours of establishing reactor power levels below 20X rated power, while these functions are required to be operable. At reactor power levels below 20X rated power, hydrogen injection shall be terminated and the injection system secured. Amendment No. 87 Nine Nile Point - Unit 1 3/4 6-13
TABLE 3.6.2b INSTRUMENTATION THAT INITIATES PRIMARY LAN M NM N ISOLATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable PRIMARY COOLANT
~masn team, Cleanup, and Shutdown)
(1) Low-Low Reactor Water Level >5 inches X X (Indicator Scale) (2) Manual X X X X MAIN-STEAM-LINE L N (3) High Steam Flow Main-Steam Line <105 psid X X Amendment No. X, f4, 37 Nine Mile Point - Unit 1 3/4 6-14
TABLE 3. 6. 2b (Cont'd) INSTRUMENTATION THAT INITIATES PRIMARY COOLAN SYST M OR N A NM N ISOLATION Limi tin Condi tion for 0 erati on Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stem Set Point 0 erable (4) High Radiation Hain-Steam Line <5 times X X normal back-ground at rated power (5) Low Reactor >850 psig Pressure (6) Low-Low-Low >7 in. mercury (a) x Condenser Vacuum vacuum (7) High Temperature <200F x x Main-Steam-Line Tunnel Amendment No. 97, 87 Nine Mile Point - Unit 1 3/4 6-15
TABLE 3.6.2b (Cont'd) INSTRUMENTATION THAT INITIATES PRIMARY LAN Y A NM N ISOLATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operabl e Instr ument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable CLEANUP SYSTEM (8) High Area Temperature <190 X X X X SHTUDOWN COOLING LA (9) High Area Temperature <170 X X X X CONTAINMENT (10) Low-Low >5 inches (c) X X Reactor Water /Indicator Scale) Amendment No. L, LE, 37 Nine Mile Point - Unit 1 3/4 6-16
TABLE 3.6.2b (Cont'd) INSTRUMENTATION THAT INITIATES PRIMARY OOLANT SYSTEM OR CON A NM N ISOLATION Limitin Condition for 0 eration Minimum No. of . Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (11) High Drywell <3.5 psig (c) (b) (b) Pressure (12) Manual X X X X Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-17
TABLE 4. 6. 2b INSTRUMENTATION THAT INITIATES PRIMARY LAN Y M N A NM N ISOLATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration PRIMARY COOLANT nanna ~a>n team, Cleanup and Shutdown) (1) Low-Low Reactor Once/day Once per month Once per 3 months Water Level (2) Manual Once during each major refueling outage MAIN-STEAM-LINE (3) High Steam Flow Once/day Once per month( ) Once per 3 months Main-Steam Line (4) High Radiation Once/shift Once/week Once per 3 months Main-Steam Line (5) Low Reactor Once/day Once per month Once per 3 months (d) Pressure Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-18
TABLE 4.6.2b (Cont'd) INSTRUMENTATION THAT INITIATES PRIMARY LAN Y M N A NM N ISOLATION Survei llance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (6) Low-Low-Low None Once during each Once during each Condenser Vacuum major refueling major refueling outage outage (7) High Temperature None Once during each Once during each Hain-Steam-Line major refueling major refueling Tunnel outage outage CLEANUP SYSTEM (8) High Area Once/week Once during each Once during each Temperature major refueling major refueling outage outage SHUTDOWN COOLING (9) High Area Once/week Once during each Once during each Temperature major refueling major refueling outage outage Nine Mile Point Unit 1
TABLE 4..6.2b (Cont'd) INSTRUMENTATION THAT INITIATES PRIMARY L N Y M ISOLATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration CONTAINMENT LA N (10) Low-Low Reactor Once/day Once per month Once per 3 months Mater Level (ll) High Drywell Once/day Once per month Once per 3 months Pressure (12) Manual Once during each operating cycle Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-20
NOTES FOR TABLES 3.6.2b AND 4.6.2b (a) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi. (b) May be bypassed when necessary for containment inerting. r (c) May be bypassed. in the shutdown mode whenever the reactor coolant system temperature is less than 215 F (d) Only the trip'circuit will be calibrated and tested at the frequencies specified in Table 4.6.2b, the primary sensor will be calibrated and tested once per operating cycle. (e) Within 24 hours prior to the planned start of the hydrogen injection test with the reactor power at greater than 20K rated power, the normal full-power radiation background level and associated trip and alarm setpoints may be changed based on a calculated value of the radiation level expected during the test. The background radiation level and associated trip and alarm setpoints may be adjusted during the test program based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip and alarm setpoints shall be reset within 24 hours of re-establishing normal- radiation levels after completion of the hydrogen injection or within 12 hours of establishing reactor power levels below 20K rated power, while these functions are required to be operable. At reactor power levels below 20X rated power, hydrogen injection shall be terminated and the injection system secured. Amendment No. Xg, 87, 87 Nine Mile Point - Unit 1 3/4 6-21
TABLE 3.6.2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Limitin Condition for 0 eration
~
Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable EMERGENCY COOLING N A N (1) High-Reactor Pressure <1080 psig (b) X X (2) Low-Low Reactor Water Level >5 inches
/Indicator Scale) (b) X X EMERGENCY COOLING oemm
'Paar eac i of two systems) (3) High Steam Flow Emergency Cooling System 2(a) 19 psid X X Amendment No. 60 Nine Mile Point - Unit 1 3/4 6-22
TABLE 4. 6. 2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Survei llance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration EMERGENCY COOLING N A N (1) High Reactor None Once/per month Once per 3 months Pressure (2) Low-Low Reactor Water Level Once/day Once per month Once per 3 months EMERGENCY COOLING ~for each of two systems) (3) High Steam Flow None Once per 3 months( ) Once per 3 months Emergency Cooling System Amendment No. 60 Nine Mile Point - Unit 1 3/4 6-23
NOTES FOR TABLES 3.6.2c AND 4.6.2c (a) Each of two differential pressure switches provide inputs to one instrument channel in each trip system. (b) Hay be bypassed in the cold shutdown condition. (c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2c, the primary sensor will be calibrated and tested once per operating cycle. Amendment No. XA, 37 Nine Hile Point - Unit 1 3/4 6-24
TABLE 3.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Mhich Operabl e Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable START CORE EPRAY~UM S (1) High Drywel 1 < 3.5 psig (d) x (a) (a) Pressure (2) Low-Low Reactor Mater Level 2(f) > 5 inches (b) x x x
/Indicator Scale)
OPEN CORE SPRAY (3) Reactor Pressure > 365 psig x x x x and either (1) or (2) above. Amendment No. Q, 87, 79 Nine Mile Point - Unit 1 3/4 6-25
TABLE 4.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration START CORE EHRF7 UUUS (1) High Drywel 1 Once/day Once per month Once per 3 months Pressure (2) Low-Low Reactor Once/day Once per month Once per 3 months Water Level OPEN CORE SPRAY (3) Reactor Pressure None Once per month Once per 3 months and either (1) or (2) above Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-26
NOTES FOR TABLES 3. 6. 2d AND 4. 6. 2d (a) Hay be bypassed when necessary for containment inerting. (b) May be bypassed when necessary for performing major maintenance as specified. in Specification 2. 1. 1. e. (c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2d, the primary sensor will be calibrated and tested 'once per operating cycle. (d) Hay be bypassed when:necessary for integrated leak rate testing. (e) The instrumentation that initiates the Core Spray System is not requried to fuel in the reactor vessel. be operable, if there is no (f) One instrument channel in each trip system may be bypassed in the cold shutdown and refuel conditions during the Spring 1986 refueling outage to perform the emergency condenser piping replacement. Amendment No. Xg, 87, 79 Nine Mile Point - Unit 1 3/4 6-27
TABLE 3.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (1) a. High Drywel 1 < 3.5 psig (a) X X Pressure and
- b. Low-Low Reactor Water Level > 5 inches
/Indicator Scale) (a) X X Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-28
S 0 0
TABLE 4.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) a. High Drywell Once/day Once per month( ) Once per 3 months (b) Pressure
- b. Low-Low Reactor Once/day Once per month Once per 3 months Mater Level Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-29
NOTES FOR TABLES 3. 6. 2e AND 4. 6. 2e (a) Hay be bypassed in the shutdown mode whenever the reactor coolant temperature is less than 2150F. (b) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2e, the primary sensor will be calibrated and tested once per operating cycle. Amendment No. 37 Nine Nile Point - Unit 1 3/4 6-30
TABLE 3.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION
,Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable 'perable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable INITIATION (1) a. Low-Low-Low Reactor Water 2 (a) 2 (a) > -10 inches* (b) (b) x Level (Indicator scale) and
- b. High Drywell 2 (a) 2 (a) < 3.5 psig (b) (b) x Pressure
- greater than (>) means less negative Amendment No. 64 Nine Mile Point - Unit 1 3/4 6-31
TABLE 4.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration INITIATION (1) a. Low-Low-Low None Once per month Once per 3 months Reactor Mater and
- b. High Orywel 1 Once/day Once per month Once per 3 months Pressure Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-32
NOTES FOR TABLES 3. 6. 2f AND 4. 6. 2f (a) Both instrument channels in either trip system are required to be energized to initiate auto depressuri-zatlon. One trip system is powered from power board 102 and the other trip system from power board 103. (b) May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature. (c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2f, the primary sensor will be calibrated and tested once per operating cycle. Amendment No. 37 Nine Mile Point - Unit 1 3/4 6-33
TABLE 3. 6. 2 INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operab'le Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (1) SRM
- a. Detector not 2 (a), (e) X X in Startup Position
- b. Inoperative 2 (a) X X
- c. Upscale 2 (a) <10s counts/sec ,eX X (2) IRM
- a. Detector not 3 (b) X X in Startup Position
- b. Inoperative 3 (b) X X Nine Mile Point - Unit 1 3/4 6-34
INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable
- c. Downscale 3 (b) >5 percent of X X full scale for each scale
- d. Upscale 3 (b) <88 percent of X X full scale for each scale (3) APRM
- a. Inoperative 3 (c) X X X
- b. Upscale (Biased 3 (c) Figure 2. 1. 1 x x x by Recirculation Flow)
- c. Downscale 3 (c) >2 percent of (d) (d) x Vull scale Am~~~ng M0. 5 Nine Mile Point - Unit 1 3/4 6-35
0 INSTRUMENTATION THAT INITIATES CONTROL ROO WITHDRAWAL BLOCK Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (4) Recirculation Flow
- a. Comparator X X X Off Normal
- b. Flow Unit 2. X X X Inoperative
- c. Flow Unit Figure 2.1.1 X X X Upscale (5) Refuel Platform 2 (f) and Hoists (6) Mode Switch in Shutdown Amendment No. 5 Nine Mile Point Unit 1 3/4 6-36
0 0
~TA LE .6.2 fC INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (7) Mode Switch in Refuel (Blocks withdrawal of more than 1 rod)
(8) Scram Oump Volume X X Water Level Scram Bypass Nine Mile Point - Unit 1 3/4 6-37
INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Surveillance Re uirement Instrument Instrument - Channel Parameter Sensor Check Channel Test Cal ibrati on (1) SRM
- a. Detector not in Startup Position
- b. Inoperative
- c. Upscale (2) IRM
- a. Detector not in Startup Position
- b. Inoperative
- c. Downscale (g) (g)
- d. Upscale (g) (g)
Nine Mile Point - Unit 1 3/4 6-38
~A.. 2 INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (3) APRM
- a. Inoperative None Once per month Once per 3 months
- b. Upscale (Biased None Once per month Once per 3 months by Recirculation Flow)
- c. Downscale None Once per month Once per 3 months (4) Recirculation Flow
- a. Comparator None Once per month Once per month Off Normal b.. Flow Unit None Once per month Once per month Inoperative
- c. Flow'nit None Once per month Once per month Upscale Nine Mile Point - Unit 1
~TAB E .. tC 'd)
INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (5) Refuel Platform (see 4.5.2) and Hoists (6) Mode Switch in Once during each Shutdown major refueling outage (7) Mode Switch Once during each in Refuel major refueling (Blocks withdrawal) outage or more than 1 rod) (8) Scram Dump Volume Once during each Water Level Scram major refueling Bypass outage Nine Mile Point - Unit 1 3/4 6-40
NOTES FOR TABLES 3.6.2g AND 4.6.2g (a) No more than one of the four SRH inputs to the single trip system shall be bypassed. (b) No more than one of the four IRM inputs to each instrument channel shall be bypassed. These signals may be bypassed when the APRH's are onscale. (c) No more than one of the four APRM inputs to each instrument channel shall be'ypassed provided that the APRH in the other instrument channel in the same core quadrant is not bypassed. No more than two C or 0 level LPRH inputs to an APRM shall be bypassed an'd only four LPRH inputs to only one APRH shall be bypassed in order for the APRM to be considered operable. In the Run mode of operation, bypass of two chambers from one radial core location in any one APRM shall cause that APRM to be considered inoperative. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRH input if the TIP is positioned in close proximity to the failed LPRM it is replacing. If one APRM in a quadrant is bypassed and meets all requirements for operability with the exception of the requirement of at least one operable chamber at each radial location, it may be returned to service and the other APRM in that quadrant may be removed from service for test and/or calibration only test. if no control rod is withdrawn during the calibration and/or (d) Hay be bypassed in the startup and refuel positions of the reactor mode switch when the IRM's are onscale. (e) This function may be bypassed when the count rate is >100 cps. (f) One sensor provides input to each of two instrument channels. Each instrument channel is in a separate trip system. (g) Calibrate prior to startup and normal shutdown and thereafter check once per shift and test once per week until no longer required. Nine Mile Point - Unit 1 3/4 6-,41
TABLE 3.6.2h VACUUM PUMP ISOLATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable VACUUM PUMP High Radiation Main Steam <5 times nppal X X' Line background' Amendment No. 88, 87 Nine Mile Point - Unit 1 3/4 6-42
TABLE 4.6.2h OFFGAS AND VACUUM PUMP ISOLATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration GFFGAS (j.) High Radiation Offgas Line
- a. Upscal e Once/shift Once per week Once per 3 months
- b. Down seal e Once/shift Once per week Once per 3 months VACUUM PUMP (2) High Radiation Main Steam Once/shift Once per week Once per 3 months Line NOTES FOR TABLES 3.6.2h and 4.6.2h
'ine Mile Point - Unit 1 3/4 6-43
I NOTES FOR TABLES 3. 6. 2h and 4. 6. 2h (a) Mithin 24 hours prior to the planned start of the hydrogen injection test with the reactor power at greater than 20K rated power, the normal full-power radiation background level and associated trip and alarm setpoints may be changed based on a calculated value of the radiation level expected during the test. The background radiation level and associated trip and alarm setpoints may be adjusted during the test program based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip and alarm setpoints shall be reset within 24 hours of re-establishing normal radiation levels after completion of the hydrogen injection or within 12 hours of establishing reactor power levels below 20K rated power, while these functions are required to be operable. At reactor power levels below 20X rated power, hydrogen injection shall be terminated and the injection system secured. Amendment No. 87 Nine Nile Point - Unit 1 3/4 6-44
TABLE 3.6.2i DIESEL GENERATOR INITIATION Limitin Condition for 0 eration Reactor Mode Switch Minimum Position in Mhich Total No. Channels Channels Function Must Be Parameter of Channels ~to Tri ~0erahl e 0 erable Loss of Power
- a. 4.16kV PB 102/103 Emergency Bus Undervoltage (Loss of Voltage) 3 per Bus 2 per Bus 2 per Bus X X X X
- b. 4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) 3 per Bus 2 per Bus 2 per Bus X X X X (1) If one out of three channels becomes inoperable, the inoperable channel will be placed in the trip condition.
Amendment No. 67 Nine Mile Point - Unit 1 3/4 6-45
Table 3.6.2i (Cont'd) DIESEL GENERATOR INITIATION Limitin Condition for 0 eration Parameter Set Point (Inverse Time Undervolta e Rela s Loss of Power
- a. 4.16kV PB 102/103 Emergency Bus >3200 volts 0 volts <3.2 seconds Undervolt (Loss of Voltage)
- b. 4.16kV PB 102/103 Emergency Bus >3600 volts 3580 volts 18.5 + 3 seconds Undervol tage (Degraded Voltage)
(a) The operating time indicated in the table is the time required for the relay to operate its contacts when the voltage is suddenly decreased from operating voltage level values to the voltage level listed in the table above. Amendment No. 67 Nine Nile Point - Unit 1 3/4 6-46
TABLE 4. 6. 2i DIESEL GENERATOR INITIATION Survei llance Re uirement~ Instrument Instrument Parameter Sensor Check Channel Test Channel Calibration Loss of Power
- a. 4.16kV PB 102/103 Emergency Bus Undervoltage (Loss of Voltage) NA Once per month Once per refueling cycle
- b. 4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) Once per month Once per refueling cycle (a) The instrument channel test will demonstrate the operability of the instrument channel by simulating an undervoltage condition to verify that the tripping logic functions properly.
(b) The instrument channel calibration will demonstrate the operability of the instrument channel by simulating an undervoltage condition to verify that the tripping logic functions properly. In addition, a sensor calibration will be performed to verify the set points listed in Table 3.6.2.i. Amendment No. 67 Nine Mile Point - Unit 1 3/4 6-47
TABLE 3.6.2 EMERGENCY VENTILATION INITIATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (1) High Radiation <5mr/hr X X X X Reactor Building Ventilation Ouct (2) High Radiation <1000 mr/hr (a) (a) (a) (a) Refueling Platform Nine Mile Point - Unit 1 3/4 6-48
p 4~ l a I ~
TABLE 4.6.2'MERGENCY VENTILATION INITIATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) High Radiation Once/shift Once during Once per quarter Reactor Building each operating Ventilation Duct cycle (2) High Radiation (b) (c) Once per quarter Refueling Platform Nine Mile Point Unit 1
NOTES FOR TABLES 3. 6. 2j AND 4. 6. 2j (a) This function shall be operable any time that irradiated fuel or the irradiated fuel cask is being handled in the reactor building. (b) Once pei shift whenever this function is required to be operable. (c) Immediately prior to when function is required and once per week thereafter until function is no longer required. Nine Mile Point - Unit 1 3/4 6-50
TABLE 3.6.2k HIGH PRESSURE COOLANT INJECTION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable q + (1) Low Reactor Mater Level > 53 inches (a) (a) x (Indicator scale) (2) Automatic Turbine Trip (a) (a) x Amendment No. I, X4, 37 Nine Mile Point - Unit 1 3/4 6-51
TABLE 4.6.2k HIGH PRESSURE COOLANT INJECTION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) Low Reactor Once per day Once per month Once per 3 months Mater Level (2) Automatic None Once during each None Turbine Trip operating cycle Amendment No. 37 Nine Nile Point - Unit 1 3/4 6-52
I~ 0 4
NOTES FOR TABLES 3.6.2k AND 4.6.2k (a) Hay be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature. (b) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2k, the primary sensor will be calibrated and tested once per operating cycle. Amendment No. X8, 37 Nine Nile Point - Unit 1 3/4 6-53
TABLE 3.6.21 PRIMARY CONTAINMENT MONITORING Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (1) Suppression Chamber Water Level Specification 3.3.2 X X Amendment No. 28, 76 Nine Mile Point - Unit 1 3/4 6-54
TABLE 4.6.21 PRIMARY CONTAINMENT MONITORING Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) Suppression Chamber Once/day N/A Once Every Six Months Mater Level Amendment No. N, 76 Nine Mile Point - Unit 1 3/4 6-55
TABLE 3. 6. 2m CONTROL ROOM AIR TREATMENT SYSTEM INITIATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels Per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable (1) High Radiation Ventilation Intake <1000 CPM X X X Amendment No. 73 Nine Mile Point - Unit 1 3/4 6-56
TABLE 4.6.2m CONTROL ROOM AIR TREATMENT SYSTEM INITIATION Survei llance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) High Radiation Once/shift Once per quarter Once each operating Venti lation Intake cycle not to exceed 24 months Amendment No. 73 Nine Mile Point - Unit 1 3/0 6-57
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- 3. 6. 3 EMERGENCY POWER SOURCES 4.6.3 EMERGENCY POWER SOURCES A licabilit :
Applies to the operational status of Applies to the periodic testing requirements the emergency power sources. for the emergency power sources.
~0b'ective: ~0b ective:
To assure the capability of the emer- To assure the operability of the emergency gency power sources to provide the power power sources to provide emergency power required for emergency equipment in the required in the event of a loss-of-coolant event of a loss-of-coolant accident. accident. S ecification:
- a. For all reactor operating conditions The emergency power systems surveillance except cold shutdown, there shall will be performed as indicated below. In normally be available two 115 kv addition,-components on which maintenance external lines, two diesel generator has been performed will be tested.
power systems and two battery sys-tems, except as further specified in a. Durin each ma'or refuelin outa e-
"b," "c," "d," "e," and "h," below. test for automat)c startup and pickup of load required for a loss-of-coolant accident.
- b. One 115 kv external line may be de- b. ~Honthl - manual start and operation energized provided two diesel-generator at rated load shall be performed for a power systems are operable. If a minimum time of one hour. Determine 115 kv external line is de-energized, the specific gravity of each cell.
that line shall be returned to service Determine the battery voltage. within 7 days. Nine Mile Point - Unit 1 3/4 6-58
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- c. One diesel-generator power system c. W~ieekl
- determine the ce)l vnitege end may be inoperable provided two 115 specific gravity of the pilot cells of kv external lines are energized. each battery.
If a diesel-generator power system becomes inoperable, it shall be returned to an operable condition within seven days. In addition, a diesel-generator power system if becomes inoperable coincident with a 115 kv line de-energized, that diesel-generator power system shall
, be returned to an operable condi-tion within 24 hours.
- d. If a reserve power transformer be- d. Surveillance for startu with an in-comes inoperable, it shall be re- o crab 1 e di ese 1- enerator - prior to turned to service within seven days. startup the operable diesel-generator shall be tested for automatic startup and pickup of the load required for a loss-of-coolant accident.
- e. For all reactor operating conditions e. Surveillance for o eration with an except startup and cold shutdown, >no crab e d>ese - enerator - the the following limiting conditions operab e diese -generator shall be shall be in effect: manually started and operated at rated load for a minimum time of one hour immediately and once per week thereafter.
(1) One operable diesel-generator power system and one energized 115 kv external line shall be available. If this condition is not met, normal orderly shutdown will be initiated within one hour and the reactor will be in the cold shutdown condition within ten hours. Nine Mile Point - Unit 1 3/4 6-59
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) If no 115 kv external line is available, both diesel-generator power systems shall be operable with one diesel-generator running. If no 115 kv external line is available after 24 hours, normal orderly shutdown will be initiated within one hour and the reactor will be in the cold shutdown con-dition within ten hours.
- f. For all reactor operating conditions except cold shutdown, there shall be a minimum of two day's fuel supply onsite for one diesel-generator or normal orderly shutdown will be initiated within one hour and the reactor will be in the cold shutdown condition within ten hours.
- g. When operating with only one diesel-generator, all emergency equipment aligned to the operable diesel-generator shall have no inoperable components.
- h. If a battery system becomes inoper-able that system shall be returned to service within 24 hours.
Nine Mile Point - Unit 1 3/4 6-60
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.4 SHOCK SUPPRESSORS SNUBBERS) 4.6.4 SHOCK SUPPRESSORS SNUBBERS) Applies to the operational status of Applies to the periodic testing requirement shock suppressors (snubbers). for shock suppressors (snubbers).
~00'ective: ~0b'ective:
To assure the capability of the To assure the operability of the snubbers to snubbers to: perform their intended functions. Prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, and Allow normal thermal motion during startup and shutdown. Amendment No. Xg, 74 Nine Mile Point - Unit 1 3/4 6-61
LIMITING CONDITION FOR OPERATION'URVEILLANCE REQUIREMENT The following surveillance requirements apply to snubbers. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related system.
- a. During all reactor operating a. Visual Ins ection conditions, except cold shutdown, snubbers shall be operable on (i) Visual Ins ection Fre uenc those systems required to be operable during that particular Snubbers shall be visually inspected operating condition except as in accordance with the following noted in 3.6.4.b, c and d below. schedule:
Snubbers excluded from this in- Number of Snubbers spection program are those in- Found Inoperable stalled on nonsafety-related During Inspection or Next Required systems and then only if their failure or failure of the system During Inspection Inspection Interval Interval on which they are installed, would have no adverse effect on any 0 Refueling period safety-related system. 1 12 months + 25K 2 6 months + 25K 3,4 124 days + 25K 5,6,7 62 days + 25K 8, or more 31 days + 25K The required inspection interval shall not be lengthened more than one step at a time. Amendment No. N, 74 Nine Mile Point - Unit j. 3/4 6-62
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Snubbers may be categorized into two types (mechanical and hydraulic). These may then be classified as "accessible" or "inaccessible" based on accessibility for inspection during operation. These four groups may be inspected independently according to the above schedule. (ii) Visual Ins ection Acce tance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired operability, (2) attachments to the foundation or supporting structure are secure, and (3) in those locations where snubber movement can be manually induced with-out disconnecting the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined operable for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; or (2) the affected snubber is functionally tested in the as found condition and determined operable per Specification 4.6.4.b as applicable. Amendment No. X8, 74 Nine Mile Point - Unit 1 3/4 6-63
LIMITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT
- b. With one or more snubbers inoper- b. Functional Testin able, within 72 hours replace or restore the inoperable snubber(s) (i) Functional Test.Fre uenc to the operable status or perform an engineering evaluation to deter- At least once each refueling cycle, mine that the components supported 10X of the total of each type (mechan-by the snubber(s) were not adversely ical or hydraulic, accessible or in-affected by the inoperability of accessible) of snubber in use in the the snubber(s), i. e. the snubber(s) plant shall be functionally tested is (are) not required for the system either in place or in a bench test.
operability. For each snubber that does not meet the functional test acceptance criteria of 4.6.4b(ii) an additional 10K of that type of snubber shall be functionally tested. / (ii) Functional Test Acce tance Re uirement Hydraulic snubber functional test shall verify that:
- l. Activation (restraining action) is achieved within the specified range of velocity.
- 2. Freedom of movement exists in both tension and compression.
Mechanical snubber functional test shall verify that:
- 1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.
Amendment No. N, 74 Nine Mile Point - Unit 1 3/4 6-64
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE(UI REHENT
- 2. Acti vati on (restraining acti on) is achieved within the specified range of velocity or acceleration in both tension and compression.
- c. If after 72 hours the actions as de-scribed in Section 3 '.4b have not been completed, the supported system shall be declared inoperable and the appropriate action statement for that system will be followed.
- d. If the actions described in 3.6.4.b or c resulted in replacement or restoration to the operable status of the effected snubber(s), perform an engineering evaluation to de-termine if the components supported by the snubber(s) were adversely affected by the inoperabi lity of the snubber.
Amendment No. 78, 74 Nine Hile Point - Unit 1 3/4 6-65
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- 3. 6. 5 RADIOACTIVE MATERIAL SOURCES 4. 6. 5 RADIOACTIVE MATERIAL SOURCES A 1icabilit Appl i es to the 1 imi t on sour ce 1 eakage Applies to the periodic testing requirements for sealed or start-up sources. for source leakage.
~Ob'ective: ~0b 'ective:
To specify the requirements necessary To assure the capability of each source to limit contamination from radioactive material container to limit leakage within source materials. allowable limits. e if'. Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or an agreement State, as follows: The leakage test shall be capable of l. Each sealed source, except start-up detecting the presence of 0.005 micro- sources subject to core flux, containing curie of radioactive material on the radioactive material, other than hydrogen test sample. If the test reveals 3, with a half-life greater than 30 days the presence of 0.005 microcurie or and in any form other than gas shall be more of removable contamination, shall immediately be withdrawn from use, it tested for leakage and/or contamination at intervals not to exceed six months. decontaminated and repaired or be dis-posed of in accordance with Commission regulations. Sealed sources are exempt from such leak tests when the source contains 100 microcuries or less of beta and/or gamma emitting material or 10 microcuries or less of alpha emitting material. Amendment No. 63 Nine Nile Point - Unit 1 3/4 6-66
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT I 3.6.5 RADIOACTIVE MATERIAL SOURCES (Cont'd) 4.6. 5 RADIOACTIVE MATERIAL SOURCES (Cont')
- 2. Results of required leak tests performed 2. The periodic leak test required does on sources, if the tests reveal the not apply to sealed sources that are presence of 0.005 microcurie or more stored and not being used. The sources of removable contamination, shall be excepted from this test shall be tested reported within 90 days. for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a trans-feror indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
- 3. A complete inventory of radioactive 3. Start-up sources shall be leak tested by-product materials, exceeding the within 31 days prior to being subjected limits set forth in 10 CFR 30.71, to core flux and following any repair or in sealed sources in possession shall maintenance.
be maintained current at all times. Amendment No. 63 Nine Mile Point - Unit 1 3/4 6-67
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 6. 6 FIRE DETECTION 4. 6. 6 FIRE DETECTION A 1 icabil it: A 1 icabi lit:
Applies to the operational status of the fire Applies to the periodic surveillance of the detection system. fire detection system.
~0b'ective: ~0b'ective:
To assure the capability of fire detection To assure the operability of the fire instrumentation for each fire detection zone detection instrumentation for each fire shown in Table 3.6.6a to provide fire detection. detection zone shown in Table 3.6.6a to-provide fire detection. S ecification: S ecification:
- a. With the number of detectors OPERABLE less a. Each of the fire detectors shall be than the number required by Table 3.6.6a. demonstrated OPERABLE:
- 1. Within one hour, establish a fire watch 1. By performance of an instrument patrol to inspect the zone with the channel test at least once per six inoperable detector(s); and months for all detection devices.
- 2. Restore the inoperable detector(s) to OPERABLE status within 14 days OR
- 3. Prepare and submit a report in accordance with 6. 9. 2. b.
Amendment No. 22, 53 Nine Mile Point - Unit 1 .3/4 6-68
TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPMENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL 81 LOCATION 8 DETECTORS MINIMUM 8 OPERABLE 1 DA-2031 - Turb. Bldg. 250 North of Cable Spread Room 17 17 2 DA-2041N - Turb. Bldg, 250 Diesel Gen. 102 12 12 3 DA-2041S - Turb. Bldg. 250 Diesel Gen. 103 6 6 4 DA-2051E - Turb. Bldg. 250 South Side East 14 14 5 DA-2051W - Turb. Bldg. 250 South Side West 16 16 6 DA-2081S - Turb. Bldg. 261 East Corridor 41 41 7 DX-2141A - Turb. Bldg. 261 Diesel Gen. 102 3 3 8 DX-2141B - Turb. Bldg. 261 Diesel Gen. 102 3 3 9 D-2151 -,Turb. Bldg. 261 D.G. 103 Cable Tray 3 2 (Note 1) 10 DX-2151A - Turb. Bldg. 261 Diesel Gen. 103 3 3 ll DX-2151B - Turb. Bldg. 261 Diesel Gen. 103 3 3 12 DA-2161E - Turb. Bldg. 261 South Side East 23 23 13 DA-2161M - Turb. Bldg. 261 P.B. 11 8 12 Area 22 22 14 DX-3011A - Turb. Bldg. 250 Cable Spreading Room 6 6 15 DX-3011B - Turb. Bldg. 250 Cable Spreading Room 6 6 16 0-3031PL - Turb. Bldg. 261 Aux. Control Room Panels 99 80 (Note 1) 17 DX-3031A - Turb. Bldg. 261 Aux. Control Room 16 16 Amendment No. 22, 53 Nine Mile Point - Unit 1 3/4 6-69
TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPMENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL ¹1 (Cont'd) LOCATION ¹ DETECTORS MINIMUM ¹ OPERABLE 18 DX-3031B Turb. Bldg. 261 Aux. Control Room 16 16 19 D-8151 - South Yard Foam Room 2 1 20 DA-2141 Turb. Bldg. 261 Diesel Gen. 102 4 21 DA-2151 Turb. Bldg. 261 Diesel Gen. 103 4 Note 1: No two (2) adjacent detectors may be out of service simultaneously. Amendment No. ZZ, 53 Nine Mile Point - Unit 1 3/4 6-70
0 TABLE 3. 6. 6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPMENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL ¹2 LOCATION ¹ DETECTORS MINIMUM ¹ OPERABLE 22 DA-2022N - Turb. Bldg. 250 North Corner 20 20 23 DA-2022S - Turb. Bldg. 250 West Side 22 22 24 DA-2092E - Turb. Bldg. 261-277 Booster Pump Area 40 40 25 DA-2092W - Turb. Bldg. 261-277 Recirc. MG Set Area 36 36 26 DA-2162W - Turb. Bldg. 261 West Side South 33 33 27 DA-2092MG - Turb. Bldg. 261 Recirc. MG Sets 15 15 Amendment No. 53 Nine Mile Point - Unit 1 3/4 6-71
TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPMENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL ¹3 1 LOCATION ¹ DETECTORS MINIMUM ¹ OPERABLE 28 DA-2013S -. Turb. Bldg. 250 North Side East 11 11 29 DA-2013N - Turb. Bldg. 250 Cond. Stor. Tank 16 16 30 DA-2083M - Turb. Bldg. 261 Cool Water Pump Area 44 31 DA-2083N - Turb. Bldg. 261 Cond. Stor. Tank 23 23 32 DX-2113A - Turb. Bldg. 261 Power Board 103 Room 1 1 33 DX-2113B - Turb. Bldg. 261 Power Board 103 Room 1 1 34 DX-2123A - Turb. Bldg. 261 Power Board 102 Room 1 1 35 DX-2123B - Turb. Bldg. 261 Power Board 102 Room 1 1 36 D-5013 - Screen House 250-261 P.B. 176 Area 6 5 37 D-5023 - Screen House 243-256 South Side 17 14 (Note 1) Note 1: No two (2) adjacent detectors may be out of service simultaneously. Amendment No. 53 Nine Mile Point - Unit 1 3/4 6-72
TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPMENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL ¹4 LOCATION ¹ DETECTORS MINIMUM ¹ OPERABLE 38 D-2224 - Turb. Bldg. 277 P.B. 101 Area 22 18 (Note 1) 39 0-2234 - Turb. Bldg. 277 South East Side 27 22 (Note 1) 40 0-3054 - Turb. Bldg. 277 Control Room 26 22 (Note 1) Note 1: No two (2) adjacent detectors may be out of service simultaneously. Amendment No. 53 Nine Mile Point - Unit 1 3/4 6-73
0 TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPMENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL ¹5 LOCATION ¹ DETECTORS MINIMUM ¹ OPERABLE 41 D-2345 - Turb. Bldg. 305 Rx Bldg. Supply Fan Area 13 11 (Note 1) 42 0-2395 - Turb. Bldg. 300 Control Ventilation Area 7 6 Note 1: No two (2) adjacent dete0tors may be out of service simultaneously. Amendment No. 53 Nine Mile Point - Unit 1 3/4 6-74
TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPHENT DETECTION ZONES LOCAL FIRE ALARH CONTROL PANEL ¹6 LOCATION ¹ DETECTORS MINIHUH ¹ OPERABLE 43 D-4016 - Rx Bldg. 198 Northeast Corner 2 1 44 D-4026 - Rx Bldg. 198 Northwest Corner 2 1 45 0-4036 - Rx Bldg. 198 Southwest Corner 2 1 46 D-4046 - Rx Bldg. 198 Southeast Coner 2 1 47 DA-4076E - Rx Bldg. - 237 East Side 16 16 48 DA-4076W - Rx Bldg. - 237 West Side 21 21 49 D-4086 - Rx Bldg. Drywell 9 7 (Note 1) (Note 2) 50 DA-4116E - Rx Bldg. 261 East Side 11 11 51- DA-4116W - Rx Bldg. 261 West Side 21 21 52 D-4156 - Rx Bldg. 281 West Side 16 13 (Note 1) 53 D-4166 - Rx Bldg. 281 East Side 16 13 (Note 1) Note 1: No two (2) adjacent detectors may be out of service simultaneously. Note 2: Detectors in service ~onl when unit is shutdown and drywell is open for major maintenance. Amendment No. 53 Nine Nile Point - Unit 1 3/4 6-75
0 TABLE 3.6.6a FIRE DETECTORS PROTECTING SAFETY-RELATED E UIPHENT DETECTION ZONES LOCAL FIRE ALARM CONTROL PANEL ¹7 LOCATION ¹ DETECTORS MINIMUM ¹ OPERABLE 54 D-4197 - Rx Bldg. 298 North Side 9 7 (Note 1) 55 D-4207 - Rx Bldg. 298 South Side 7 6 56 DA-4237 - Rx Bldg. 318 Storage Area 30 30 57 D-4267 - Rx Bldg. 340, 13 10 (Note 1) 58 DX-4217A - Rx. Bldg. 298 Emerg. Cond. Vlv. Room 4 59 DX-4217B - Rx. Bldg. 299 Emerg. Cond. V'. Room 4 v', Note 1: No two (2) adjoining detectors may be out of service simultaneously. Amendment No. 53 Nine Mile Point - Unit 1 3/4 6=76
A LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 6. 7 FIRE SUPPRESSION 4. 6 7
~ FIRE SUPPRESSION A licabilit :
Applies to the operational status of the fire Applies to the surveillance of the fire suppression system. suppression system.
~0b'ective: ~bb ective:
To assure the capability of the fire sup- To assure the operability of the fire pression system to provide fire suppression suppression system to provide fire in the event of a fire. suppression in the event of a fire. S ecification: S ecification:
- a. The FIRE SUPPRESSION MATER SYSTEM shall be a. The FIRE SUPPRESSION WATER SYSTEM OPERA L with; shal be demonstrated OPERABLE:
- 1. Two high pressure pumps each with a l. At least once per 31 days by capacity of 2500 gal./min. with their starting each pump and operating discharge aligned to the fire it for 30 minutes on recircula-suppression header. tion flow.
- 2. Automatic initiation logic for each 2. At least once per 31 days by fir'e pump. verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position.
- 3. At least once per 12 months by cycling each manually-operable valve through one complete cycle.
- 4. At least once per 6 months by a flush of the hydrants.
Amendment No. 22, 53 Nine Mile Point - Unit 1 3/4 6-77
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.7 FIRE SUPPRESSION (Cont'd) 4.6 ~ 7 FIRE SUPPRESSION (Cont'd)
- 5. At least once per operating cycle.
(a) By performing a system automatic start on low header pressure. (b) By verifying that each pump will develop a flow of at
~
least 2500 gpm at a pump discharge of 115 psig. (c) Cycling each valve in the flow path that is not test-able during plant operation through at least one com-plete cycle of full travel. (d) Verifying that each automatic valve in the flow path actuates to its correct position.
- 6. At least once per 3 years by per-forming a flow test of the system in accordance with Chapter 8, Sec-tion 16 of the Fire Protection Handbook, 15th Edition, published by the National Fire Protection Association.
- b. Mith an inoperable redundant pump or b. The fire pump diesel engine shall be water supply line inoperable, restore demonstrated OPERABLE:
the inoperable equipment to OPERABLE status within 7 days or prepare and 1. Daily by checking the starting air submit a report in accordance with tank pressure 6.9.2.b. Amendment No. gg, 53 Nine Nile Point - Unit 1 3/4 6-78
LIHITING CONOITION FOR OPERATION SURVEILLANCE REPUIREHENT 4.6.7 FIRE SUPPRESSION (Cont'd)
- 2. At least once per 31 days by verifying:
(a) That the fuel day storage tank contains at least 150 gallons of fuel. (b) The fuel storage tank contains at least 1000 gallons of fuel. (c) The fuel transfer pump starts and transfers fuel from the storage tank to the day tank. (d) The diesel starts from ambient conditions and operates for greater than or equal to 30 minutes on recirculation flow. (e) The method of starting the diesel fire engine will alternate between the normal= air start method and the low air pressure start.
- 3. At least once per 92 days by
. verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTH-0270-65, is within the acceptable limits specified in Table 1 of ASTH 0975-74 with respect to viscosity, water control and sediment.
Amendment No. 22, 53 Nine Hile Point Unit 1 3/4 6-79
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.7 FIRE SUPPRESSION (Cont'd) 4.6.7 FIRE SUPPRESSION (Cont'd)
- 4. At least once per six months by using the manual bypass of the-solenoid on the starting air system.
- 5. At least once per 18 months, subjecting the diesel to an in-spection in accordance with procedures prepared in conjunc-tion with its manufacturer's recommendations for the class of service, and verifying the diesel starts from ambient conditions on the auto-start signal and operates for greater than or equal to 30 minutes while loaded with the fire pump.
- c. With no FIRE SUPPRESSION WATER SYSTEM c. The spray systems shall be demonstrated to operable, w>th>n 24 hours: be OPERABLE:
- l. Establish a backup fire suppression 1. At least once per 31 days by verifying system, and that each valve, manual, power operated or automatic, in the flow path is in its correct position.
- 2. Report to the NRC in accordance with 2. At least once per year by cycling each 6.9.2.a. manually operable valve through one complete cycle.
Amendment No. gg, 53 Nine Mile Point - Unit 1 3/4 6-80
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6.7 FIRE SUPPRESSION (Cont'd) 4.6.7 FIRE SUPPRESSION (Cont'd)
- 3. At least once per operating cycle.
(a) By performing a system func-tional test which includes simulated automatic actuation of the system and verifying that the automatic deluge valves in the flow path actuate to their correct positions. (b) By visual inspection of spray headers to verify their integrity. (c) By visual inspection of each nozzle to verify no blockage.
- 4. At least once per 3 years by per-forming an air or water flow test through each open head spray header and verifying each open head spray nozzle is unobstructed.
- d. The spray and sprinkler systems located d. The sprinkler system shall be demon-in the following areas shall be OPERABLE: strated to be OPERABLE:
- l. Automatic water spray systems 1. At least once per operating cycle.
(a) Reserve Transformer 101N (a) By performing a system (b) Reserve Transformer 101S functional test which in-cludes simulated automatic
- 2. Automatic Sprinkler System for the actuation of the system.
Diesel Fire Pump Room in the Screen House. Amendment No. g2, 53 Nine Mi le Point - Unit 1 3/4 6-81
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.7 FIRE SUPPRESSION (Cont'd) 4.6;7 FIRE SUPPRESSION (Cont'd)
- 3. Pre-Action Systems: (b) By visual inspection of sprinkler headers to verify (a) Rx Bldg., El. 237- their integrity.
(b) Rx Bldg., El. 261 (c) Rx Bldg., El. 318 (c) By visual inspection of (d) Turb. Bldg., El. 250 South each nozzle to verify no (e) Turb. Bldg., El. 250 West bl ockage. (f) Turb. Bldg., El. 250 North (g) Turb. Bldg., El. 250 East (h) Diesel Gen., El. 250 (i) Cable Spreading Room (j) Turb. Bldg., El. 261 South (k) Turb. Bldg., El. 261 North (1) Turb. Bldg., El. 261 East (m) Turb. Bldg., El. 277 East (n) Turb. Bldg., El. 300 Storage Area
- e. With a spray or sprinkler system in-operable, establish a fire watch patrol with backup fire suppression equipment for the unprotected area within one hour.
- f. With a pre-action system inoperable, trip system wet or establish a fire watch patrol with backup fire suppression equip-ment 'for the unprotected area within one hour.
- g. Restore the system to OPERABLE status within 14 days or prepare and submit a report in accordance with 6. 9. 2.b.
Amendment No. g2, 53 Nine Mile Point - Unit 1 3/4 6-82
0 0 0
LIHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.8 CARBON DIOXIDE SUPPRESSION SYSTEH 4.6.8 CARBON DIOXIDE SUPPRESSION SYSTEM Applies to the operational status of the to the periodic surveillance carbon dioxide suppression system. requirements of the carbon dioxide suppression system.
~0b 'ective: ~0b'ective:
To assure the capability of the carbon To verify the operability of the carbon dioxide suppression system to provide dioxide suppression system. fire suppression in the event of a fire. e tank.'pplies S ecification: if'. The C02 system, which supplies the a. The C02 system shall be demonstrated Recirculation Pumps Hotor-Generator Sets, operable. Power Boards 102 and 103, Diesel Generators 102 and 103, Cable Room fire 1. At least once per 7 days by hazards, shall be OPERABLE with a minimum verifying the C02 storage tank level of 40X of tank and a minimum level and pressure. pressure of 250 psig in the storage
- 2. At least once per 31 days by verifying that each valve, manual, power operated or automatic, in the flow path is in its correct position.
- 3. At least once every six months by verifying the system valves and associated ventilation dampers actuate automatically to a simu-lated actuation signal. A brief flow test shall be made to verify flow from each nozzle (" Puff Test" ).
Amendment No. ZZ, 53 Nine Hile Point - Unit 1 3/4 6-83
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6.8 CARBON DIOXIDE SUPPRESSION SYSTEM (Cont'd)
- b. With one or more of the above required COq systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged.
- c. The Auxiliary Control Room CO> system shall be operated as a manual backup for the Halon System.
- d. Restore the system to OPERABLE status within 14 days or prepare and submit a report in accordance with 6.9.2.b.
Amendment No. gg, 53 Nine Mile Point -- Unit 1 3/4 6-84
0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.9 FIRE HOSE STATIONS 4.6.9 FIRE HOSE STATIONS Applies to the operational status of the Applies to the periodic surveillance of fire hose stations. the fire hose stations.
~0b'ective: ~bb ective:
To assure the capability of the fire hose To assure the operability of the fire hose stations to provide fire suppression in the station to provide fire suppression in the event of a fire. event of a fire. e if'. The fire hose stations in the locations a. Each fire hose station shown in shown in Table 3.6.9a shall be operable. Table 3.6.9a shall be verified to be OPERABLE:
- 1. At least once per 31 days by visual inspection of the fire hose stations accessible during plant operation to assure all required equipment is at the station.
- b. Mith one or more of the fire hose b. At least once per operating cycle by:
stations shown in Table 3. 6. 9a inoperable, route an additional equivalent capacity 1. Visual inspection of the fire hose fire hose to the unprotected area(s) from stations not accessible during an operable hose station within 1 hour if plant operation to assure all the inoperable fire hose is the primary required equipment is at the hose means of fire suppression, otherwise station. route the additional hose within 24 hours.
- 2. Removing the hose for inspection and re-racking.
Amendment No. gg, 53 Nine Nile Point - Unit 1 3/4 6-85
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.9 FIRE HOSE STATIONS (Cont'd) 4.6.9 FIRE HOSE STATIONS (Cont'd)
- 3. Inspecting all gaskets and re-placing any degraded gaskets in the couplings.
- c. Restore the inoperable fire hose c. At least once per 3 years by:
station(s) to operable status within 14 days or prepare and submit a report 1. Partially opening each hose in accordance with 6.9.2.b. station valve to verify valve operability and no flow blockage.
- 2. Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any hose station.
Nine Mile Point Unit Amendment No. gg, 53 1 3/4 6-86
TABLE 3.6.9a FIRE HOSE STATIONS
~Bui1 din Elevation feet Column Station Number
- 1. TURBINE 267
- 2. TURBINE Aa"7 FS-144 267 C -3
- 3. TURBINE 267 FS-132 TURBINE G -2 FS-128 267 H -9
- 5. REACTOR 346 FS-123
- 6. REACTOR L "12 FS-112 346 P "4 7.
8. REACTOR REACTOR 324 324 K -ll
-5 FS-106 FS-ill
- 9. P FS-105 REACTOR 309 "11 10.
11. REACTOR REACTOR
'04 K P -5 FS-110 FS-104 287 K -11
- 12. REACTOR 287 FS-109 P -5
13. 14. REACTOR REACTOR 267. 267 K -ll FS-103 FS-108
- 15. REACTOR P -5 FS-102 243 K -11
- 16. REACTOR 243 FS-107
- 17. WASTE P -5 FS-101 267 H -19
- 18. WASTE 267 FS-301
- 19. WASTE NB-19 FS-300 267 HB-16
- 20. TURBINE 375 FS"116
- 21. TURBINE H -8 FS-126 357 H -10
- 22. TURBINE 311 FS-121
- 23. TURBINE H -9 FS-125 311 G -2
- 24. TURBINE FS" 130 311 C -3
- 25. TURBINE 297 FS-134
- 26. TURBINE H -9 FS-124 297 G -2
- 27. TURBINE 267 FS-129
- 28. DIESEL F -15 FS" 117 267 C -18
- 29. DIESEL FS-164 267 Ba-17 FS-166 Amendment No. 22, 53 Nine Nile Point - Unit 1 3/4 6-87
TABLE 3. 6. 9a (Cont') FIRE HOSE STATIONS
~Bnil din Elevation feet) Column Station Number
- 30. TURBINE 256 Aa-13 FS-152
- 31. DIESEL 256 Aa-17 FS-163
- 32. DIESEL 256 C -17 FS-165
- 33. TURBINE 256 H -9 FS-122
- 34. TURBINE 267 Aa-14 FS-156
- 35. TURBINE 267 B -2 FS-139
- 36. TURBINE 267 P -14 FS-114
- 37. TURBINE 297 C -3 FS-133
- 38. TURBINE 283 B -2 FS-140
- 39. TURBINE 283 Aa-7 FS-145
- 40. TURBINE 283 Aa-13 FS-153
- 41. TURBINE 283 F -15 FS-118
- 42. TURBINE 256 Aa-7 FS-143
- 43. TURBINE 256 B -2 FS-138
- 44. TURBINE 256 C "3 FS-131
- 45. TURBINE 256 G -2 FS-127
- 46. TURBINE 256 H -13 FS" 115
- 47. SCREEN 256 UV-16 FS-408
- 48. TURBINE 277 H -9 FS-405
- 49. TURBINE 256 F -16 FS-406
- 50. REACTOR 267 Track Bay FS-401
- 51. REACTOR 267 M -12 FS-422
- 52. REACTOR 243 P -9 FS-402
- 53. REACTOR 243 Drywell Entrance . FS-403
- 54. REACTOR 243 P-Q FS-404
- 55. SCREEN 262 R-14 FS-113
- 56. TURBINE 306 Aa-13 FS-154 Amendment No. 22, 53 Nine Mile Point - Unit 1 3/4 6-88
0 LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- 3. 6. 10 ADDITIONAL FIRE E UIPMENT 4. 6. 10 ADDITIONAL FIRE E UIPMENT 3.6.10.1 FIRE BARRIER PENETRATIONS 4.6.10.1 FIRE BARRIER PENETRATIONS A 1icabilit Applies to the condition of the fire Applies to the periodic surveillance barrier penetrations, including cable requirements for the fire barrier penetration barriers, fire doors and penetrations.
fire dampers.
~0b ective: ~0b ective:
To assure the capabi lity of the fire To verify the condition of the fire barrier barrier penetrations to perform their penetrations. intended function.
- a. All fire barrier penetrations a. Fire barrier penetrations shall be protecting safety related areas verified to be functional by:
shall be intact.
- l. A visual inspection at least once per operating cycle.
- 2. A visual inspection of.a fire barrier penetration after repair or maintenance, prior to restoring the fire barrier penetration to functional status.
- b. With one or more of the above re-quired fire barrier penetrations non-functional, within one hour establish a continuous fire watch on one side of the affected penetration, or Amendment No. gg, 53 Nine Nile Point - Unit 1 3/4 6-89
0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.10 ADDITIONAL FIRE E UIPHENT (Cont'd
- c. Verify the operability of fire detectors on one side of the non-functional fire barrier and establish a fire watch patrol.
- d. Restore the non-functional fire barrier penetrations to functional status within 14 days or prepare and submit a report in accordance with 6.9.2. b.
Amendment No. 22, 53 Nine Mile Point - Unit 1
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- 3. 6. 10. 2 HALON SUPPRESSION SYSTEM 4.6. 10. 2 HALON SUPPRESSION SYSTEM Applies to the operational status of Applies to the periodic surveillance require-the Halon suppression system. ment of the Halon suppression system.
~0b ective: ~0b'ective:
To assure the capability of the Halon To verify the operability of the Halon suppression system to provide fire suppression system. suppression in the event of a fire.
~limni
- a. The Halon systems which supply the Each of the required Halon systems Auxiliary Control and Emergency shall be demonstrated operable:
Condenser I.V. Rooms shall be oper-able. with the storage tanks having 1. At least once per 31 days by at least 95K of full charge weight verifying that each valve, manual, (level) and 90K of full charge power operated or automatic, in the pressure. flow path is in its correct position.
- 2. At least once per 6 months by verifying Halon storage tank weight (level) and pressure.
- 3. At least once per 18 months by:
(a) Verifying the system and associated ventilation dampers and fire door release mechan-isms actuate manually and automatically. Amendment.No. 53 Nine Nile Point - Unit 1 3/4 6-91
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.10.2 HALON SUPPRESSION SYSTEM (Cont'd) 4.6.10.2 HALON SUPPRESSION SYSTEM (Cont'd) (b) Performance of a flow test thr ough headers and nozzl es to assure no blockage.
- b. With a Halon system inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment.
- c. Restore the system to operable status within 14 days or prepare and submit a report in accordance with 6. 9. 2. b.
Amendment No. 53 Nine Mile Point - Unit 1 3/4 6-92
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.10.3 YARD FIRE HYDRANTS AND HYDRANT 4.6.10.3 YARD FIRE HYDRANTS AND HYDRANT H H U H H U
~0b To fire if'pplies Applies to the operational status of the Yard Fire Hydrants and Hose Houses.
ective: assure the capabi lity of hydrant to provide fire the yard suppression To assure to the periodic surveillance requirement of the yard fire hydrants and associated hose houses.
~0b'ective:
the operability of the yard fire hydrant to provide fire suppression in the in the event of a fire. event of a fire. 4 e
- a. The yard fire hydrants shown in a. Each of the yard fire hydrants and Table 3.6.10.3a shall be operable. associated hose houses shown in Table 3.6.10.3a shall be demonstrated operable:
- 1. At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.
- 2. At least once per 6 months during March, April, May and during September, October and November by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.
Amendment No. 53 Nine Mile Point Unit 1 3/4 6-93
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.10.3 YARD FIRE HYDRANTS AND HYDRANT 4.6.10.3 YARD FIRE HYDRANTS AND HYDRANT H HOUS S ont d HO HOU S Cont d
- 3. At least once per 12 months by:
(a) Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any yard fire hydrant. (b) Replacement of all degraded gaskets in couplings.
- b. Mith one or more of the yard fire hydrants or associated hydrant houses shown in Table 3.6.10.3a inoperable, route sufficient addi-tional lengths of 2-1/2 inch diam-eter hose located in an adjacent operable hydrant hose house to pro-vide service to the unprotected area(s) within one hour, if the in-operable fire hydrant is the pri-mary means of fire suppression, otherwise, route an additional hose within 24 hours.
- c. Restore the inoperable hydrant(s) and/or hose house(s) to operable status within 14 days or prepare and submit a report in accordance with
- 6. 9. 2. b.
Amendment No. 53 Nine Nile Point - Unit 1 3/4 6-94
TABLE 3.6.10.3a YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES H drant Number Location El. 261' West El. 261' Southwest Amendment No. 53 Nine Mile Point - Unit 1 3/4 6-95
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 11 ACCIDENT MONITORING INSTRUMENTATION 4.6. 11 ACCIDENT MONITORING INSTRUMENTATION
~0b if'pplies Applies to the operability of the plant instrumentation that performs an accident monitoring function.
ective: To assure high reliability of monitoring instrumentation. the accident
~0b To to the survei llance of the instrumentation that performs an accident monitoring function.
'ective:
verify the operability of accident monitor ing instr umentation. e
- a. During the power operating condition, Instrument channels shall be tested and the accident monitoring instrumentation calibrated at least as frequently as channels shown -in Table 3.6.11-1 shall listed in Table 4.6.11.
be operable except as specified in Table 3. 6.11-2. Amendment No. 72 Nine Mile Point - Unit 1 3/4 6-96
TABLE 3.6.11-1 ACCIDENT MONITORING INSTRUMENTATION Minimum Number Total Number of Operable Action Parameters of Channels Sensors or Channels See Table 3.6.11-2
- 1) Relief Valve Position Indication 2/Valve 1/Valve
- 2) Safety Valve Position Indication 2/Valve 1/Valve
- 3) Reactor Vessel Mater Level
- 4) Drywell Pressure Monitor
- 5) Suppression Chamber Mater Level
- 6) Containment Hydrogen Monitor
- 7) Containment High Range Radiation Monitor
- 8) Suppression Chamber Mater Temperature 2 Amendment No. 72, 76 Nine Mile Point - Unit 1 3/4 6-97
0 TABLE 3.6.11-2 ACCIDENT MONITORING INSTRUMENTATION A ONSA MNS ACTION - 1
- a. With the number of OPERABLE accident monitoring instrumentation channels 1 less than the total number shown in Table 3.6.11-1, restore to an OPERABLE status during the next cold shutdown when there is access to the drywell.
- b. Mith the number of OPERABLE accident monitoring instrumentation channels less than the minimum number shown in Table 3.6.11-1, restore the inoperable channel to an OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours.
- c. The total number of channels shown in Table 3.6.11-1 will be OPERABLE prior to the begining of each cycle.
ACTION - 2
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the total Number of Channels shown in Table 3.6. 11-1, restore the inoperable channel(s) to OPERABLE status within seven days or be in at least HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the minimum Channels OPERABLE requirements of Table 3. 6. 11-1, restore the inoperable channel(s) to OPERABLE status within 48 hours or-be in at least HOT SHUTDOWN within the next 12 hours.
ACTION - 3 34.
- a. Mith the number of OPERABLE channels less than the total Number of Channels shown in Table - . 11-1, prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
Amendment No. 72 Nine Nile Point - Unit 1 3/4 6-98
I TABLE 3.6.11-2 (Cont'd) ACCIDENT MONITORING INSTRUMENTATION A ON A MNS
- b. With the number of OPERABLE channels less than required by the minimum channels OPERABLE requirements, initiate the pre-planned alternate method of monitoring the appropriate parameter(s) within 72 hours, and:
- 1) either restore the inoperable channel(s) to OPERABLE status within seven days of the event, or
- 2) prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
ACTION - 4 34.
- a. With the number of OPERABLE channels less than the total Number of Channels shown in Table Q. ll-l, prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
- b. With the number of OPERABLE channels less than required by the minimum channels OPERABLE requirements, initiate the pre-planned alternate=method of monitoring the appropriate parameter(s) within 72 hours, and:
J') either restore the inoperable channel(s) to OPERABLE status within seven days of the event, or
- 2) prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE sgst~sk&u5 C. If the pre-planned alternate method of monitoring the appropriate parameter(s) is not available, either restore the inoperable channel(s) to OPERABLE status within seven days or be in at least HOT SHUTDOWN within the next 12 hours.
Amendment No. 72 Nine Mile Point - Unit 1 3/4 6-99
TABLE 4.6.11 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREHENT Instrument Channel Parameter Test Instrument Channel Calibration (1) Relief valve position indicator Once per month Once during each major refueling outage (Primary Acoustic) Relief valve position indicator Once per month Once during each major refueling outage (Backup - Thermocouple) (2) Safety valve position indictor Once per month Once during each major refueling outage (Primary - Acoustic) Safety valve position indicator Once per month Once during each major refueling outage (Backup - Thermocouple) (3) Reactor vessel water level Once per month Once during each major refueling outage (4) Drywell Pressure Monitor Once per month Once during each major refueling outage (5) Suppression Chamber Water Level Once per month Once during each major refueling outage Monitor (6) Containment Hydrogen Monitor Once per month Once per quarter (7) Containment High Range Once per month Once during each major refueling outage Radiation Monitor (8) Suppression Chamber Water Temperature, Once per month Once during each major refueling outage Amendment No. 72, 76 Nine Mile Point - Unit 1 3/4 6-100
LIHITING CONOITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 12 REACTOR PROTECTION SYSTEM MOTOR 4.6. 12 REACTOR PROTECTION SYSTEM MOTOR N A H N N Applies to the operability of instrumen- Applies to the survei llance of instrumen-tation that provides protection of Motor tation that provides protection of the Generator sets;and the maintenance bus reactor protection Motor Generator 'sets that supplies power to the reactor pro- and maintenance bus that supplies power to tection system'and reactor trip system. the reactor protection system and reactor trip system.
~0b'ective: ~0b'ective:
To assure the operability of the instru- To verify the operability of protection mentation required for safe operation of instrumentation on the Motor Generator sets the Motor Generator sets and the mainte- and maintenance bus that supplies power to nance bus that supplies power to the the reactor protection and reactor trip reactor protection system and reactor buses. trip system. e'
- a. Except as specified in specifica- At least once ever six months tions b and c below, two protective emonstrate operabs sty o the over-relay systems shall be operable for voltage, undervoltage and under-each Motor Generator set and the frequency protective instrumentation maintenance bus. by performing an instrument channel test. This instrument channel test will consist of simulating abnormal Motor Generator Set conditions by applying from a test source, an overvoltage signal, an undervoltage signal and an underfrequency signal to verify that the tripping logic up to but not including the output contactors functions properly.
Amendment No. 62 Nine Mile Point - Unit 1 3/4 6-101
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 12 REACTOR PROTECTION SYSTEM MOTOR 4.6. 12 REACTOR PROTECTION SYSTEM MOTOR N A M N N ont d) N RA 0 S M N R N ont d)
- b. With one protective relaying system b. At least once er refuel in c cle inoperable, restore the inoperable emonstrate operab>>ty o the over-system to an operable status within voltage, undervoltage and under-72 hours or remove the Motor Gen- frequency protective instrumentation erator set or maintenance bus from by performing an instrument channel service. test. This instrument channel test will consist of simulating abnormal Motor Generator Set conditions by applying from a test source an over-voltage signal, an undervoltage signal and an underfrequency signal to verify that the tripping logic including the output contactors functions properly at least once. In addition, a sensor calibration will be performed to verify the following setpoints.
- i. Overvoltage <132 volts, <4 seconds ii. Undervoltage >108 volts, <4 seconds iii. Underfrequency >57 hertz, <2 seconds
- c. With both protective relaying sys-tems inoperable, restore at least one to an operable status within 30 minutes or remove the associated Motor Generator sets or maintenance bus from service.
Amendment No. 62 Nine Mile Point - Unit 1 3/$ 6-)P2
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.13 REMOTE SHUTDOWN PANELS 4. 6. 13 REMOTE SHUTDOWN PANELS Applies to the operating status of the remote Applies to the periodic testing requirements shutdown panels. for the remote shutdown panels.
~0b'ective: ~Qb 'ective:
To assure the capability of the remote To assure the capability of the remote shutdown panels to provide 1) initiation of shutdown panels to provide 1) initiation of the emergency condensers independent of the the emergency condensers independent of the main/auxiliary control room 2) control of the main/auxi liary control room 2) control of the motor-operated steam supply valves independent motor-operated steam supply valves independent of the main/auxiliary control room and 3) of the main/auxiliary control room and 3) parameter monitoring outside the control parameter monitoring outside the control room. room. The remote shutdown panels surveillance shall be performed as indicated below:
- a. During power operation and whenever the a. Each remote shutdown panel monitoring reactor coolant temperature is greater instrumentation channel shall be than 212'F, at least one remote shutdown demonstrated operable by performance of panel shall be operable. the operations and frequencies shown in Table 4.6.13-1.
Amendment No. 71 Nine Mile Point - Unit 1 3/4 6-103
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 13 REMOTE SHUTDOWN PANELS (Cont'd) 4.6. 13 REMOTE SHUTDOWN PANELS (Cont'd)
- b. A remote shutdown panel shall be b. Durin each ma or refuelin outa e considered inoperable if either the emergency condenser condensate return Each remote shutdown panel shall valve control switch is inoperable, either be demonstrated to initiate the motor-operated steam supply valve control emergency condensers independent switch is inoperable, or the number of of the main/auxiliary control room.
operable instrumentation channels is less than that required by Table 3.6.13-1. 2. Each remote shutdown panel shall be demonstrated to open both the motor-operated steam valves.
- c. If Specification 3. 6. 13. a cannot be met, commence an orderly shutdown within 24 hours and be in cold shutdown within 36 hours.
Amendment No. 71 Nine Nile Point - Unit 1 3/4 6-104
0 0
TABLE 3.6.13-1 REMOTE SHUTDOWN PANEL MONITORING Limitin Condition for 0 eration MINIMUM NUMBER OF INSTRUMENT OPERABLE CHANNELS Reactor Pressure Reactor Water Level Reactor Water Temperature Torus Water Temperature Drywell Pressure Emergency Condenser Water Level Drywell Temperature "All Rods In" Light Amendment No. 71 Nine Mile Point - Unit 1 3/4 6-105
TABLE. 4.6. 13-1 REMOTE SHUTOOWN PANEL MONITORING Surveillance Re uirement Instrument Channel Parameter Sensor Check Calibration Reactor Pressure Once per day Once per 3 months (a) Reactor Mater Level Once per day Once per 3 months (a) Reactor Mater Temperature Once per day Once per refueling cycle Torus Mater Temperature Once per day Once per refueling cycle Drywell Pressure Once per day Once per 3 months (a) Emergency Condenser Water Once per day Once per refueling cycle Level Orywell Temperature Once per day Once per refueling cycle "All Rods In" Light Once per refueling cycle N/A (a) The indicator located at the remote shutdown panel will be calibrated at the frequency listed in Table 4.6.13-1. Calibration of the remaining channel instrumentation is provided by Specification 4.6.2. Amendment No. 71 Nine Mile Point - Unit 1 3/4 6-106
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 14 RADIOACTIVE EFFLUENT INSTRUMENTATION 4.6. 14 RADIOACTIVE EFFLUENT INSTRUMENTATION A 1 icabi lit Applies to the operability of plant instru- Applies to the surveillance of instrumenta-mentation that monitors plant effluents. tion that monitors plant effluents.
~bb 'ective: ~0b'ective:
To assure the operability of instrumentation To verify operation of monitoring to monitor the release of radioactive plant instrumentation. ef f1uents.
~Sifts
- a. Liquid Effluent a. Liquid Effluent The radioactive liquid effluent monitoring Each radioactive liquid effluent moni-instrumentation channels shown in Table toring instrumentation channel shall 3.6.14-1 shall be operable with their be demonstrated operable by performance alarm setpoints set to ensure that the of the sensor check, source check, in-limits of Specification 3.6.15. a. 1 are not strument channel calibration and channel exceeded. The alarm setpoints of these test operations at the frequencies shown channels shall be determined and adjusted in Table 4.6.14-1.
in accordance with the methodology and parameters in the Offsite Dose Calculation Records - Anditable records shall be Manual. ma>nta>ned, in accordance with proce-dures in the Offsite Dose Calculation With a radioactive liquid effluent moni- Manual, of all radioactive liquid toring instrumentation channel alarm set- effluent monitoring instrumentation point less conservative than a value which alarm setpoints. Setpoints and setpoint will ensure that the limits of 3. 6. 15. a. 1 calculations shall be available for review are met, immediately suspend the release to ensure that the limits of Specification of radioactive liquid effluents monitored 3.6.15.a.l are met. by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-107
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 14 RADIOACTIVE EFFLUENT INSTRUMENTATION (Cont')
- a. Liquid Effluent (Cont'd)
With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels operable, take the action shown in Table 3.6.14-1. Restore the instruments to OPERABLE status within 30 days, or outline in the next Semi-Annual Radioactive Effluent Release Report the cause of the inoperability and how the instruments were or will be restored to operable status. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-108
TABLE 3.6. 14-1 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION Limitin Condition for 0 eration Minimum Channels Instrument ~0erabl e A 1 icabi1 it
- 1. Gross Radioactivity Monitors ( )
A. Liquid Radwaste Effluent Line 1(c) At all times B. Service Mater System Effluent Line 1(d) At all times
- 2. Flow Rate Measurement Devices A. Liquid Radwaste Effluent Line 1(e) At all times B. Discharge Canal
- 3. Tank Level Indicating Devices A. Outside Liquid Radwaste Storage Tanks At all times
**Pumps curves or rated capacity will be utilized to estimate flow.
Amendment No. (9', $$ Nine Mile Point - Unit 1 3/4 6-109
NOTES FOR TABLE 3.6.14-1 (a) Provide alarm, but do not provide automatic termination of release. (b) An operator shall be present in the Radwaste Control Room at all times during a release. (c) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases may continue provided that prior to initiating a release:
- 1. At least two independent samples are analyzed in accordance with Specification 4. 6. 15. a, and
- 2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valying.
Otherwise suspend release of radioactive effluents via this pathway. (d) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided that, at leas't once per 12 hours, graf samples are collected and analyzed for gamma radioactivity at a lower limit of detection of at least 5xl0- microcurie/ml. (e) Ouring discharge, with the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. (f) With the number of channels operable less than required by the minimum channels operable requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during liquid additions to the tank. (g) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents. (h) With the number of channels operable less than required by the minimum channels operable requirement, steam release via this pathway may commence or continue provided vent pipe radiation dose rates are monitored once per four hours. (i) Yi(eked n0 'g'h ha condu.ch~ unHnuaguy Q~'gq~ Bul4Linp Mviea, u>lL4I <ckufA Jlkuvr Q aB<gd~ S~Pg 44 !~ 4/ kpplhli'Nd~ (g ~,'pg~ 'elwda4,
~'I(L'g Amendment No. Pg SE Nine Mile Point - Unit 1 3/4 6-110 J
Table 4.6. 14-1 RADIOACTIVE LI UID EFFLUENT MONITORING INST UMENTATION, Surveillance Re uirement Sensor Source Channel Channel Instrument Check Check(f) Test Calibration
- 1. Gross Beta or Gamma Radioactivity Monitors
- a. Liquid Radwaste Effluent Line Once/day" Once/ Once/3 months Once/year discharge*
- b. Service Mater Effluent Line Once/day Once/month Once/3 months Once/year
- 2. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line Once/day None None Once/year
- b. Discharge Canal Hone None None Once/year
- 3. Tank Level Indicating Devices
- a. Outside Liquid Radwaste Storage Tanks Once/day"" None Once/3 months Once/18 months
*Required prior to removal of blank flange in discharge line and until blank flange is replaced.
""During liquid addition to the tank. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-111
NOTES FOR TABLE 4.6.14-1 (a) The channel test shall also demonstrate that control following conditions exist: room alarm annunciation occurs if any of the
- 1. Instrumentation indicates measured levels above the alarm setpoint.
- 2. Instrument indicates a downscale fai lure.
- 3. Instrument controls not set in operate mode.
(b) The channel calibration shall be performed using one or more reference standards certified by the National Bureau of Standards or using standards that are traceable to the National Bureau of Standards or using actual samples of liquid waste that have been analyzed on a system that has been calibrated with National Bureau of Standard traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. (c) Sensor check shall consist of verifying indication of flow= during periods of release. Sensor check shall be made at least once per 24 hours on days on which continuous, periodic or batch releases are made. (d) Pump performance curves or rated data may be used to estimate flow. (e) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents. (f) Source check may consist of an installed check source, response to an external source, or (for liquid radwaste monitors) verification within 30 minutes of commencing discharge of monitor response to effluent. Nile Point Unit Amendment No. 66 Nine 1 3/4 6-112
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 14 RADIOACTIVE EFFLUENT INSTRUMENTATION (Cont') 4.6. 14 RADIOACTIVE EFFLUENT INSTRUMENTATION (Cont')
- b. Gaseous Process and Effluent b. Gaseous Process and Effluent The radioactive gaseous process and Each radioactive gaseous process and effluent monitoring instrumentation effluent monitoring instrumentation channels shown in Table 3.6. 14-2 shall be channel shall be demonstrated operable by operable with their alarm setpoints set to performance of the sensor check, source ensure that the limits of Specifica- check, instrument channel calibration and tion 3.6. 15. b. 1 are not exceeded. The instrument channel test operations at the alarm setpoints of these channels shall be frequencies shown in Table 4.6.14-2.
determined and adjusted in accordance with the methodology and parameters in the Auditable records shall be maintained of Offsite Dose Calculation Manual. the calculations made, in accordance with procedures in the Offsite Dose Calculation With a radioactive gaseous process and Manual, of radioactive gaseous process and effluent monitoring instrumentation effluent monitoring instrumentation alarm channel alarm setpoint less conservative setpoints. Setpoints and setpoint than required by the above specification, calculations shall be available for review immediately suspend the release of to ensure that the limits of Specifica-radioactive gaseous effluents monitored by tion 3. 6. 15. b. 1 are met. the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative. With less than the minimum number of radioactive gaseous process and effluent monitoring instrumentation channels operable, take the action shown in Table 3.6.14-2. Restore the instruments to OPERABLE status within 30 days or outline in the next Semi-Annual Radioactive Effluent Release Report the cause of the inoperability and how the instruments were or will be restored to operable status. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-113
TABLE 3.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Limitin Condition for 0 eration Minimum Channels Instrument ~0erab le A 1 icabi1 it Parameter
- 1. Stack Effluent Monitoring
- a. Noble Gas Activity Monitor Radioactivity Rate Measurement
- b. Iodine Sampler Cartridge 1(b) Verify presence of cartridge
- c. Particulate Sampler Filter 1(b) Verify presence of filter
- d. Sample Flow Rate Measuring Device 1(c) Sampler flow rate measurement
- e. Stack Gas Flow Rate Measuring Device 1( ) Effluent flow rate measurement
- 2. Main Condenser Offgas Treatment Explosive Gas Monitoring System
- a. Hydrogen Monitor 1(e) Hydrogen V
"Required prior to removal of blank flange in discharge line and until blank flange is replaced.
""During liquid addition to the tank. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-114
TABLE 3.6. 14-2 (Cont'd) RADIOACTIYE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Limitin Condition for 0 eration Minimum Channels Instrument ~berabl e A 1 icabi 1 it Parameter
- 3. Condenser Air Ejector Radioactivity Monitor (Recombiner discharge or air ejector discharge)
- a. Noble Gas Activity 1(g) Noble gas radioactiviy rate measurement
- b. Offgas System Flow Rate Measuring 1(c) System flow rate Devices measurement
- c. Sampler Flow Rate Measuring Devices 1 Sampler flow rate measurement
- 4. Emergency Condenser System
- a. Noble Gas Activity Monitor 1 per vent( ) Noble gas radioactivity rate measurement
*""During operation of the main condenser air ejector
- ~"During reactor power operating condition Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-115
NOTES FOR TABLE 3.6.14-2 (a) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken once per 12 hours and these samples are analyzed for gross activity within 24 hours. (b) Mith the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment in accordance with the requirements of Table 4. 6. 15-2. (c) With the number of channels operable less then required by the minimum channels operable requirements, effluent releases via this pathway may continue provided the flow rate is estimated once per 8 hours. (d) Stack gas flow rate may be estimated by exhaust fan operating configuration. (e) Mith the number of channels operable less then required by the minimum channels operable requirement, operation of the main condenser offgas treatment system may continue provided gas samples are collected and anlyzed once per 8 hours. (f) One monitor on each recombiner. The system is designed to withstand the effects of a hydrogen explosion. (g) With the number of channels operable less than required by the minimum channels operable requirement, gases from the main condenser offgas treatment system may be released provided:
- 1. Offgas grab samples are collected and analyzed once per 12 hours.
- 2. The stack monitor is operable.
- 3. Otherwise, be in at least hot shutdown within 12 hours.
(h) Mith the number of channels operable less than required by the minimum channels operable requirements, steam release via this pathway may commence or continue provided vent pipe radiation dose rates are monitored once per four hours. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-116
C t >>~>> TABLE Surveillance 4.6. 14-2 Sensor Re uirements Source T Channel Channel Instrument Check Check Test Calibration
- 1. Stack Effluent Monitoring System
- a. Noble Gas Activity Honitor Once/day( ) Once/month Once/3 months Once/year
- b. Iodine Sampler Cartridge None None None None
- c. Particulate Sampler Filter None None None None
- d. Sampler Flow Rate Measuring Device Once/day( ) None None Once/year
- e. Stack Gas Flow Rate Measuring Device Once/day None None Once/year
- 2. Main Condenser Offgas Treatment System Explosive Gas Monitoring system (for system designed to withstand the effects of a hydrogen explosion)
- a. f{ydrogen Monitor Once/day( ) None Once/month Once/3 months
- 3. Condenser Air Ejector Radioactivity Monitor (Recombiner Discharge or Air Ejector Discharge)
- a. Noble Gas Activity Honitor Once/day( ) Once/month Once/~p~rating Once/year cycle
- b. 'Flow Rate Monitor Once/day None None Once/year
- c. Sampler Flow Rate Monitor Once/day( ) None None Once/year
- 4. Emergency Condenser System
- a. Noble Gas Activity Monitor Once/day Once/month Once/3 months Once/o~~~ating cycle Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-117
NOTES FOR TABLE 4. 6. 14-2 (a) At all times. (b) The channel calibration shall be performed using one or more of the reference standards certified by the National Bureau of Standards, standards that are traceable to the National Bureau of Standards or using actual samples of gaseous effluent that have been analyzed on a system that has been calibrated with National Bureau of Standards traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. (c) The channel function test shall demonstrate that control the following conditions exist: room alarm annunciation occurs if either of
- 1) Instrument indicates measured levels above the Hi or Hi Hi alarm setpoint.
- 2) Instrument indicates a downscale failure.
The channel function test shall also demonstrate that automatic isolation of this pathway occurs either of the following conditions exist: if
- 1) Instruments indicate two channels above Hi Hi alarm setpoint.
- 2) Instruments indicate one channel above Hi Hi alarm setpoint and one channel downscale.
(d) During main condenser offgas treatment system operation. (e) The channel calibration shall include the use of standard gas samples containing a nominal:
- 1. One volume percent hydrogen, balance nitrogen.
- 2. Four volume percent hydrogen, balance nitrogen.
(f) ~ During operation of the main condenser air ejector. (g) The channel test shall produce upscale and downscale annunciation. (h) During reactor power operating condition. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-118
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 15 RADIOACTIVE EFFLUENTS 4.6. 15 RADIOACTIVE EFFLUENTS A 1icabilit Applies to the radioactive effluents from Applies to the periodic test and recording the station. requirements of the station process ef f1 uents.
~0b'ective: ~0b ective:
To assure that radioactive material is not To ascertain that radioactive effluents released to the environment in any uncon- from the station are within allowable trolled manner and is within the limits values of 10 CFR 20, Appendix B and of 10 CFR 20 and 10 CFR 50 Appendix I. 10 CFR 50, Appendix I.
'e if'.
Liquid a. Liquid (1) Concentration (1) Concentration The concentration of radioactive Radioactive liquid wastes shall material released in liquid effluents be sampled and analyzed according to unrestricted areas shall be limited to the sampling and analysis pro-to the concentrations specified in gram of Table 4.6.15-1. 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than The results of the radioactivity dissolved or entrained noble gases. analyses shall be used in accord-For dissolved or entrained noble ance with the methodology and gases, the concentration shall be parameters in the Offsite Dose limited to 2 x 10-4 microcuries/ml Calculation Manual to assure that total activity. the concentrations at the point of release are maintained within Should the concentration of radioac- the limits of Specification tive material released in liquid 3.6.15. a. (1). effluents to unrestricted areas exceed the above limits, restore the concen-tration to within the above limits immediately. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-119
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.15 RADIOACTIVE EFFLUENTS (Cont') 4.6. 15 RADIOACTIVE EFFLUENTS (Cont')
- a. Liquid (Cont') a. Li qui d (Cont')
(2) Dose (2) Dose The dose or dose commitment to a Cumulative dose contributions from member of the public from radioactive liquid effluents for the current materials in liquid effluents re- calendar quarter and the current leased, from each reactor unit, to calendar year shall be determined in unrestricted areas (see Figures accordance with the methodology and
-5.1-1) shall be limited: parameters in the Offsite Dose Calculation Manual, prior to each (a) During any calendar quarter to release of a batch of liquid waste.
less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and (b) During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.3 a Special Report that identifies the cause(s) for ex-ceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-120
TABLE 4.6.15-1 urves ance e usrement Minimum Lower Limit Sampling Analysis Type of Activity of Dectection Li uid Release T e ~Fee uenc ~Fee uenc Anal sis LLD Ci/ml A. Batch Waste Tanks Each Batch Each Batch Principal Gamma 5 x 10-7 Emitters I-131 1 x 10-6 1 x 10-s
~
Each Batch~ Each Batch~ ~ Dissolved and Entrained Gases (Gamma Emitters) Monthly H-3 1 x 10-s Each Batch Composite Gross Alpha 1 x 10-7 quarterly~ Sr-89, Sr-90 5 x 10-8
~
Each Batch Composite Fe-55 1 x 10-6 B. Service Water System Effluent Once/month Once/month Principal Gamma Emitters 5 x 10-~ I-131 1 x 10-6 Dissolved and Entrained 1 x 10-s Gases H-3 1 x 10-s Gross Alpha 1 x 10-7 Once/quarter Once/quarter Sr-89, Sr-90 5 x 10 s Fe-55 1 x 10-6
- Completed prior to each release Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-121
NOTES FOR TABLE 4.6.15-1 (a) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system which may include radiochemical separation: 4.66 Sb LLD E.V 2.22 x 10 AY. exp (-Mt) Mhere: LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume,
- 2. 22 x 10 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A, is the radioactive decay constant for the particular radionuclide, and ht for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.
Typical values of E, V, Y and ht should be used in the calculation. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact for a particular measurement. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-122
NOTES FOR TABLE 4. 6. 15-1 (Cont') (b) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling. (c) The principal gamma emitters for which the LLO specification applies exclusively are the following radio-nuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, to-gether with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radio-active Effluent Release Report. (d) If more than one batch is released in a calendar month, only one batch need be sampled and analyzed during that month. (e) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. (f) If the alarm setpoint of the service water effluent monitor, as determined by the method presented in the Offsite Oose Calculation Manual, is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists. Frequency of analysis shall be increased to daily for principal gamma emitters (including dissolved and entrained gases) and an incident composite for H-3, gross alpha, Sr-89, Sr-90 and Fe-55. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-123
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 15 RADIOACTIVE EFFLUENTS (Continued) 4.6.15 . RADIOACTIVE EFFLUENTS (Continued)
- b. Gaseous b. Gaseous 1
(1) Dose Rate (1) Dose Rate The dose rate due to radioactive The dose rate due to noble gases materials released in gaseous in gaseous effluents shall be effluents from the site to areas determined to be within the at and beyond the site boundary limits of Specification 3. 6. 15 in shall be limited to the following: accordance with the methodology and parameters in the Offsite (a) For noble gases: Less than or Dose Calculation Manual. equal to 500 mrems/year to the total body and less than or The dose rate due to iodine-131, equal to 3000 mrems/year to the iodine-133, tritium and all radio-skin, and nuclides in particulate form with half lives greater than 8 days (b) For iodine-131, iodine-133, in gaseous effluents shall be tritium and all radionuclides in determined to be within the particulate form with half lives limits of Specification 3.6. 15 in greater than 8 days: Less than or accordance with methodology and equal to 1500 mrems/year to any parameters in the Offsite Dose organ. Calculation Manual by obtaining representative samples and . Mith the dose rate(s) exceeding performing analyses in accordance the above limits, without delay with the sampling and analysis restore the release rate to within program specified in Table the above limit(s). 4. 6. 15-2. Amendment No. 66 Nine -Mile Point - Unit 1 3/4 6-124
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 6.15 RADIOACTIVE EFFLUENTS (Cont'd). 4.6. 15 RADIOACTIVE EFFLUENTS (Cont')
- b. Gaseous (Cont'd) b. Gaseous (Cont')
(2) Air Dose (2) Air Dose The air dose to noble gases released Cumulative dose contributions for in gaseous effluents, from each reac- the current calendar quarter and tor unit, to areas at and beyond the current calendar year for noble site boundary shall be limited to the gases shall be determined monthly following: in accordance with the methodology and parameters in the Offsite Dose (a) During any calendar quarter: Less Calculation Manual. than or 'equal to 5 milliroentgen for gamma radiation and less than or equal to 10 mrads for beta radiation and, (b) During any calendar year: Less than or equal to 10 milliroentgen for gamma radiation and less than or equal to 20 mrads for beta radiation. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6. 9. 3, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the correc-tive actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-125
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 15 RADIOACTIVE EFFLUENTS (Cont') 4.6. 15 RADIOACTIVE EFFLUENTS (Cont')
- b. Gaseous (Cont'd) b. Gaseous (Cont'd)
(3) Tritium, Iodines and Particulates (3) Tritium, Iodines and Particulates The dose to a member of the public Cumulative dose contributions for from iodine-131, iodine-133, tritium the current calendar quarter and and all radionuclides in particulate current calendar year for form with half lives greater than iodine-131, iodine-133, tritium 8 days in gaseous effluents released, and radionuclides in particulate from each reactor unit, to areas at form with half lives greater than and beyond the site boundary shall be 8 days shall be determined monthly limited to the following: in accordance with the methodology and parameters in the Offsite Dose (a) During any calendar quarter: Less Calculation Manual. than or equal to 7. 5 mrems to any organ and, (b) During any calendar year: Less than or equal to 15 mrems to any organ. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-126
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.15 RADIOACTIVE EFFLUENTS (Cont'd)
- b. Gaseous (Cont'd)
With the calculated dose from the release of iodine-131; iodine-133, tritium and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6. 9. 3, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed correc-tive actions to be taken to assure that subsequent releases will be in compliance with the above limits. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-127
TABLE 4. 6.15-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Re uirements Minimum Lower Limit Gaseous Release Sampling Analysis Type of Activity of Dectection
~pre uenc ~pre uenc Anal sis LLD Ci/ml A. Containment Purge Each Purge Prior to each Principaf )amma 1 x 10-4 release Emmiters Grab Sample Each Purge Principaf tjamma Emitters H-3 B. Stack Once/Month Once/Month Principagamma Emitters Once/Month Once/Month H-3 1 x 10-e C. Stack Continuous Once/Week~ ~ 1 x 10-'2 Charcoal Sample Continuous Once/Meek Principagamma 1 x 10-"
Particulate Emitters Sample Continuous Once/Month Gross alpha 1 x 10-~~ Composite Sr-89, Sr-90 Particulate Sample Continuous Noble gas monitor Noble Gases, Gross 1 x 10-Gamma or Princj~ql Gamma Emitters' Amendment No. 66 Nine Nile Point - Unit 1 3/4 6-128
NOTES FOR TABLE 4. 6. 15-2 (a) The LLD is defined in notation (a) of Table 4.6. 15-1. (b) Purge is defined in Section 1. 23. (c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59; Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, I-131, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identi.fiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report pursuant to Specification 6. 9. 1. (d) Sampling and analysis shall also be performed following shutdown, startup or an increase on the recombiner discharge monitor of greater than 50 percent, factoring out increases due to changes in thermal power level or dilution flow; or when the stack release rate is in excess of 1000 pCi/second and steady-state gaseous release rate increases by 50 percent. (e) The sample flow rate and the stack flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3. 6. 15. b.(1).(b) and 3. 6. 15. b.(3). (f) When the release rate is in excess of 1000 pCi/sec and steady state gaseous release rate increases by 50 percent. The iodine and particulate collection device shall be removed and analyzed to determine the changes in iodine-131 and particulate release rate. The analysis shall be done daily following each change until it is shown that a .pattern exists which can be used to predict the release rate; after which it may revert to weekly sampling frequency. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10. (g) When RAGEMS is inoperable the LLD for noble gas gross gamma analysis shall be 1 x 10- ; (h) Tritium grab samples shall be taken weekly from the station ventilation exhaust'(stack) when fuel is offloaded until stable tritium release levels can be demonstrated. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-129
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 15 RADIOACTIVE EFFLUENTS . (Cont') 4. 6.15 RADIOACTIVE EFFLUENTS (Cont'd)
- c. Hain Condenser c. Main Condenser The gross radioactivity (beta and/or The radioactivity rate of noble gases gamma) rate of noble gases measured at the at the recombiner discharge shall be recombiner discharge shall be limited to continuously monitored in accordance less than or equal to 500,000 uCi/sec. with Table 3. 6. 14-2.
This limit can be raised to 1 Ci/sec. for a period not to exceed 60 days provided the The gross radioactivity (beta and/or offgas treatment system is in operation. gamma) rate of noble gases from the recombiner discharge shall be deter-With the gross radioactivity (beta and/or mined to be within the limits of gamma) rate of noble gases at the recombiner Specification 3.6. 15 at the following discharge exceeding the above limits, restore frequencies by performing an isotopic the gross radioactivity rate to within its analysis of a representative sample limit within 72 hours or be in at least Hot of gases taken at the recombiner Shutdown within the next 12 hours. discharge: Monthly. Within 4 hours following an increase on the recombiner dis-charge monitor of greater than 50K, factoring out increases due to changes in thermal power level and dilution flow changes. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-130
4' LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6.15 RADIOACTIVE EFFLUENTS (Cont'd) 4. 6. 15 RADIOACTIVE EFFLUENTS (Cont')
- d. Uranium Fuel Cycle d. Uranium Fuel Cycle The annual (calendar year) dose or dose Cumulative dose contributions from liquid commitment to any member of the public and gaseous effluents shall be determined due to releases of radioactivity and to in accordance with Specifications radiation from uranium fuel cycle sources 4.6.15.a. (2), 4.6.15.b. (2) and from Nine Mile Point Unit 1 shall be 4.6.15.b.(3) and in accordance with the limited to less than or equal to 25 mrems methodology and parameters in the Offsite to the total body or any organ, except the Dose Calculation Manual.
thyroid, which shall be limited to less than or equal to 75 mrems. Cumulative dose contributions from direct radiation from the reactor units and from With the calculated doses from the release radwaste storage tanks shall be determined of radioactive materials in liquid or in accordance with the methodology and gaseous effluents exceeding twice the parameters in the Offsite Dose Calculation limits of Specifications 3. 6. 15. a. 2(b), Manual. This requirement is applicable
- 3. 6. 15. b. 2(b) and 3. 6. 15. b. 3(b), calcula- only under conditions set forth in tions shall be made including direct Specification 3. 6. 15. d.
radiation contributions from the reactor units and from outside storage tanks to determine whether the above listed 40CFR190 limits have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6. 9. 3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources, including Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-131
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.15 RADIOACTIVE EFFLUENTS (Cont')
- d. Uranium Fuel Cycle (Cont'd) all effluent pathways and direct radi-ation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the esti-mated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in ac-cordance with the provisions of 40CFR 190.
Submittal of the report is considered a timely request and a variance is granted until staff action on the request is complete. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-132
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Applies to the operating status of the liquid, Applies to the surveillance requirements gaseous and solid effluent treatment systems. for the liquid, gaseous and solid effluent treatment systems.
~0b 'ective: ~0b'ective:
To assure operability of the liquid, gaseous To verify operability of the liquid, gaseous and solid effluent treatment, system. and solid effluent treatment system.
- a. Liquid a. Liquid The liquid radwaste treatment system shall Doses due to liquid releases to unre-be used to reduce the radioactive mate- stricted areas shall be projected rials in liquid wastes prior to their dis- prior to the release of each batch of charge as necessary to meet the require- liquid radioactive waste in accordance ments of Specification 3. 6. 15. with the methodology and parameters in the Offsite Dose Calculation Manual.
- b. Gaseous b. Gaseous The gaseous radwaste treatment system Doses due to gaseous releases to areas shall be operable. The gaseous radwaste at and beyond the site boundary shall treatment system shall be used to reduce be calculated monthly in accordance radioactive materials in gaseous waste with the methodology and parameters in prior to their discharge as necessary to the Offsite Dose Calculation Manual.
meet the requirements of Specification 3.6. 15. With gaseous radwaste from the main con-denser air ejector system being discharged without treatment for more than 7 days, Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-133
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (Cont'd) 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (Cont'd)
- b. Gaseous (Cont'd) prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.3, Special Report that identifies the inoperable equipment and the reason for its inoperability, actions taken 'to re-store the inoperable equipment to OPERABLE status, and a summary description of those actions taken to prevent a recurrence.
- c. Solid c. Solid The solid radwaste system shall be used in The process control program, shall be accordance with a Process Control Program used to verify the solidification of to process wet radioactive wastes to meet at least one representative test shipping and burial ground requirements. specimen from at least every tenth batch of each type of wet radioactive With the provisions of the process control waste (e.g., filter sludges and program not satisfied, suspend shipments evaporator bottoms).
of defectively processed or defectively packaged solid radioactive wastes from the (1) If any test specimen fails to site. verify solidification, the solidi-fication the batch may then be resumed using the alternative solidification parameters deter-mined by the process control program. (2) If the initial test specimen from a batch of waste fails to verify solidification, the process con-trol program shall provide for the collection and testing of repre-sentative test specimens from Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-134
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.6. 16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (Cont'd)
- c. Solid (Cont'd) each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate solidification.
Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-135
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.17 EXPLOSIVE GAS MIXTURE 4.6.17 EXPLOSIVE GAS MIXTURE A 1 icabi1 it: Applies to the operability of instrumentation Applies to the surveillance of instrumentation to monitor hydrogen concentration in the main that monitors hydrogen concentration in the condenser off-gas treatment system. main condenser off-gas treatment system. O~becti ve: ~0b ective: To assure the operability of the hydrogen To verify operation of monitoring monitoring instrumentation in the main instrumentation. condenser off-gas treatment system. e if'he concentration of hydrogen in the main The concentration of hydrogen in the main condenser off-gas treatment system shal.l be condenser off-gas treatment system shall be limited to 4 percent by volume. determined to be within the above limits by continuously monitoring the waste gases in the If the concentration of hydrogen in the main main condenser off-gas treatment system in condenser off-gas treatment system exceeds accordance with Table 3. 6. 14-2 of this limit, restore the concentration to Specification 3 ';14. within the limit within 48 hours. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-136
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 18 MARK I CONTAINMENT 4. 6. 18 HARK I CONTAINMENT Applies to the venting/purging of the Mark I Applies to the surveillance requirement for Containment. venting and purging of the Mark I Containment when required to be vented/purged through the Emergency Ventilation System.
~0b 'ecti ve: ~0b'ective:
To assure that the Hark I Containment is To verify that the Mark I Containment is vented/purged so that the limits of specifi- vented through the Emergency Ventilation cations 3.6.15.b.l and 3.6.15.b.3 are met. System when required. e if'he e if'he Mark I Containment drywell shall be containment drywell shall be determined to vented/purged through the Emergency be aligned for venting/purging through the Ventilation System unless Specification Emergency Ventilation System within four hours
- 3. 6. 15. b. 1 and 3. 6. 15. b. 3 can be met without prior to start of and at least once per 12 use of the Emergency Ventilation System. hours during venting/purging of the drywell.
If these requirements are not satisfied, suspend all venting/purging of the drywell. Amendment No. 66 Nine Mile Point Unit 1 3/4 6-137
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. 6. 19 LI UID WASTE HOLDUP TANKS" 4.6.19 LI UID WASTE HOLDUP TANKS Applies to the quantity of radioactive Applies to the surveillance requirements for material that may be stored in an outdoor outdoor liquid waste holdup tanks.
liquid waste holdup tank.
~0b 'ective: ~0b'ective:
To assure that the quantity of radioactive To verify the quantity of radioactive material material stored in outdoor holdup tanks does stored in an outdoor liquid waste holdup tank. not exceed a specified level. The quantity of radioactive material contained The quantity of radioactive material contained in an outdoor liquid waste tank shall be in each of the tanks listed in Specification limited to less than or equal to 10 curies, 3.6. 19 shall be determined to be within the excluding tritium and dissolved or entrained limit of Specification 3. 6. 19 by analyzing a noble gases. representative sample of the tank's contents at least weekly when radioactive materials are With the quantity of radioactive material in being added to the tank. any such tank exceeding the above limit, immediately suspend all additions of radio-active material to the tank. Within 48 hours reduce the tank contents to within the limit and describe the events leading to this condi-tion in the next Semi-Annual Radioactive Effluent Release Report.
*Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents and that do not have tank over-flows and surrounding area drains connected to the liquid radwaste treatment system.
Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-138
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.6. 20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Applies to radiological samples of station Applies to the periodic sampling and environs. monitoring requirements of the radiological environmental monitoring program.
~bb 'ective: ~0b'ective:
To evaluate the effects of station operations To ascertain what effect station operations and radioactive effluent releases on the and radioactive effluent releases have had environs and to verify the effectiveness of upon the environment. the controls on radioactive material sources. e if'he radiological environmental monitoring The radiological environmental monitoring program shall be conducted as specified in samples shall be collected pursuant to Table Table 3. 6. 20-1. 3.6.20-1 from the specific locations given in the table and figure(s) in the Offsite Dose With the radiological environmental monitoring Calculation Manual and shall be analyzed program not being conducted as specified in pursuant to the requirements of Table 3.6.20-1 Table 3.6.20-1, prepare and submit to the Com- and the detection capabilities required by mission, in the Annual Radiological Environ- Table 4.6. 20-1. mental Operating Report, a description of the reasons for not conducting the program as re-quired and the plans for preventing a recurrence. Deviations are permitted from the required sample schedule if samples are unobtainable due to hazardous conditions, seasonal unavai 1-ability, theft, uncooperative residents or to malfunction of automatic sampling equipment. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-139
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ont d In the event of the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period. With the level of radioactivity (as the result of plant effluents), in an environmental sampling medium exceeding the reporting levels of Table 6.9.3-1 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant to Specification 6. 9. 3. The Special Report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Specifications 3. 6. 15. a.(2), 3. 6. 15. b.(2) and 3.6. 15.b.(3). When more than one of the radionuclides in Table 6. 9.3-1 are detected in the sampling medium, this report shall be submitted if: r concentration 1 + concentration (2) +.
~....> 1.0 Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-140
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ont When radionuclides other than those in Table 6.9.3-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specification
- 3. 6. 15. a.(2), 3. 6. 15. b.(2) and 3. 6. 15. b.(3).
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. With milk or fruit and/or vegetables no longer available at one or more of the sample locations specified in Table 3. 6. 20-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Semi-Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the Offsite Dose Calculation Manual reflecting the new location(s). Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-141
TABLE. 3. 6. 20-1 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Limitin Condition for 0 eration Exposure Pathway Number of Samples (a) Sampling and Collection Type of Analysis and/or Sam le and Locations Fre uenc (a and Fre uenc Radioiodine 8 Samples from 5 locations: Continuous sampler operation Radioiodine Canisters Particulates with sample collection weekly ana yze once week or
- 1) 3 samples from off-site or as required by dust I-131.
locations in different loading, whichever is more sectors of the highest frequent Particulate Sam lers calculated site average Gross beta radio-D/g (based on all site activity following licensed reactors) filter change, (b) composite (by loca-
- 2) 1 sample from the vicinity tion) for gamma of an established year round isotopic analysis (c) community having the highest once per 3 months, (as calculated site average D/g a minimum)
(based on all site licensed reactors)
- 3) 1 sample from a control location 10-17 miles distant and in a least prevalent wind direction (d)
Amendment No. 66 Nine Mile Point - Unit 1 3/4 6"142
TABLE 3.6.20-1 (Cont'd) OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Limitin Condition for 0 eration ~/ Exposure Pathway Number of Samples and Locations (a) Sampling and Collection Fre uenc (a) Type of Analysis and Fre uenc Direct Radiation(e) 32 stations with two or more Once per 3 months Gamma dose once per dosimeters to be placed as 3 months follows: an inner ring of stations in the general area of the site boundary and an outer ring in the 4 to 5 mile range from the site with a station in each land based sector. " The balance of the stations should be placed in special interest areas such as population centers, nearby residences, schools and in 2 or 3 areas to serve as control stations. WATERBORNE Surface (f) 1) 1 sample upstream Composite sample over 1 Gamma isotopic
- 2) 1 sample from the site's month period (g) analysis (c) once/month.
downstream cooling water Composite for once intake per 3 months tritium analysis. Sediment from Shoreline 1 sample from a downstream Twice per year Gamma isotopic area with existing or potential analysis (c) recreational value At this distance, 8 wind rose sectors are over Lake Ontario. Nine Mile Point Unit Amendment No. 66 1 3/4 6-143
TABLE 3. 6. 20-1 (Cont'd) OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Limitin Condition for 0 eration Exposure Pathway Number of Samples (a) Sampling and Collection Type of Analysis and/or Sam le and Locations Fre uenc a and Fre uenc INGESTION Milk 1) Samples from milk sampling Twice per month, April-December Gamma isotopic (c) locations in 3 locations (samples will be collected in and I-131 analysis within 3.5 miles distance January-March if I-131 is twice per month when having the highest calcu- detected in November and animals are on lated site average D/g. December of the preceding pasture (April-If there are none, then 1 year) December); once/ sample from milking animals month at other times in each of 3 areas 3 '-5. 0 (January-March) if mi les distant having the required highest calculated site average D/g (based on all site licensed reactors)
- 2) 1 sample from a milk sampling location at a control location (9-20 miles distant and in a least prevalent wind direction) (d)
Fish 2 samples of commercially or Twice per year Gamma isotopic recreationally important analysis (c) on species in the vicinity of edible portions a site discharge point (h) twice per year
- 2) 1 sample each of the same species (or of a species with similar feeding habits) from an area at least 5 miles distant from the site. (d)
Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-144
TABLE 3.6.20-1 (Cont'd) OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM-Limitin Condition for 0 eration Exposure Pathway Number of Samples (a) Sampling and Collection Type of Analysis and/or Sam le and Locations Fre uenc a and Fre uenc Food Products 1) 6 samples total (utilizing Once per year during Gamma isotopic at least 2 sectors) of fruits harvest season analysis of edible and/or vegetables will be portions (isotopic collected from available off- to include I-131) site locations of highest Once during the calculated site average D/g harvest season (based on all licensed site reactors)
- 2) 1 sample of each of similar vegetation grown 9-20 miles distant in a less prevalent wind direction Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-145
NOTES FOR TABLE 3.6.20-1 (a) It is recognized that, at times, it may not be possible or practical to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and may be substituted. Actual locations (distance and directions) from the site shall be provided in the Annual Radiological Environmental Operating Report. Highest 0/g locations are based on historical meteorological data for all site licensed reactors. (b) Particulate sample filters should be analyzed for gross beta 24 hours or more after sampling to allow for radon and thoron daughter decay. If the gross beta activity in air is greater than 10 times a historical yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. (c) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility. (d) The purpose of these samples is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites, such as historical control locations which provide valid background data may be substituted. (e) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges shall not be used for measuring direct radiation. (f) The "upstream sample" should be taken at a distance beyond significant influence of the discharge. The "downstream sample" should be taken in an area beyond but near the mixing zone, if possible. (g) Composite samples should be collected with equipment (or equivalent) which is capable of collecting an aliquot at time intervals which are very short (e. g. hourly) relative to the compositing period (e. g. monthly) in order to assure obtaining a representative sample. (h) In the event commercial or recreational important species are not available as a result of three attempts, then other species may be utilized as available. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-146
i TABLE 4.6.20-1 OETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (a, b) LOWER L M 0 0 C ON LLO c Surveillance Re uirement ~Anal aia Water
~Ci//1 (c) Airborne Particulate or Gases Ci/m Fish Ci/k wet)
Milk 1/1 ~/ Food Products Sediment Ci/k dr ) gross beta 0. 01 H-3 3000 Mn-54 15 130 Fe-59 30 260 Co-58, Co-60 15 130 Zn-65 30 260 Zr-95, Nb-95 15 J I-131 (d) 0. 07 60 Cs-134 15 0. 05 130 15 60 150 Cs-137 18 0. 06 150 18 80 180 Ba/La-140 15 15 Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-147
NOTES FOR .TABLE 4.6. 20-1 (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6. 9. 1. d. (b) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in ANSI N.545 (1975), Section 4.3. Allowable exceptions to ANSI N.545 (1975), Section 4.3 are contained in the Nine Mile Point Unit 1 Offsite Dose Calculation Manual (ODCM). (c) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 Sb LLD- -Abt
~
.exp Where:
LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, S is the standard deviation of the background counting rate or of the counting rate of a blank sample a5 appropriate, as counts per minute. E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, where applicable. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-148
NOTES FOR TABLE 4.6.20-1 (Cont'd) A, is the radioactive decay constant for the particular radionuclide, and ht for environmental samples is the elapsed time between sample collection, or end of the sample collection period and time of counting Typical values of E, V, Y and ht should be used in the calculation. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for the particular measurement. Analyses shall be performed in such a manner that the stated LLDs wi 11 be achieved under routine conditions. Occasionally, background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6. 9. 1. d. (d) LLD for drinking water samples would be 1 pCi/1. No drinking water pathway exists at the Nine Mile Point Site under normal operating conditions due to the direction and distance of the nearest drinking water intake. Therefore, the LLD of the gamma isotope analysis may be used. Amendment No. 66 Nine Mile Point Unit 1 3/4 6-149
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT
- 3. 6. 21 INTERLABORATORY COMPARISON PROGRAM 4. 6. 21 INTERLABORATORY COMPARISON PROGRAM Applies to participation in an interlaboratory Applies to testing the validity of comparison program on environmental sample measurements on environmental samples.
analysis.
~0b ective: ~0b ective:
To ensure the accuracy of measurements of To verify the accuracy of measurements on radioactive material in environmental samples. radioactive material in environmental samples. e if'nalyses shall be performed on radioactive The Interlaboratory Comparison Program materials supplied as part of an Interlaboratory shall be described in the Offsite Dose Comparison Program which has been approved by Calculation Manual. A summary of the the Commission. Participation in this program results obtained as part of the above shall include media for which environmental required Interlaboratory Comparison samples are routinely collected and for which Program shall be included in the Annual intercomparison samples are available. Radiological Environmental Operating Report. Participants in the EPA Cross with analyses not being performed as required Check Program may provide the EPA above, report the corrective actions taken to program code designation in lieu of prevent a recurrence to the Commission in the providing results. Annual Radiological Environmental Operating Report. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-150
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.22 LAND USE CENSUS 4.6.22 LAND USE CENSUS A 1icabil it Applies to the performance of a land use Applies to assuring that current land use census in the vicinity of the Nine Nile Point is known. Nuclear Facility.
~0b'ective: ~0b 'ecti ve:
To determine the utilization of land within a To verify the appropriateness of the distance of three miles from the Facility. environmental survei1 lance program. A land use census shall be conducted and shall The land use census shall be conducted identify within a distance of three miles the during the growing season at least once location in each of the 16 meteorological per 12 months using that information that sectors the nearest residence and within a will provide the best results, such as distance of three miles the location in each conducting a door-to-door survey, aerial of the 16 meteorological sectors of all milk survey or consulting local agriculture animals. In lieu of a garden census, authorities. The results of the land use specifications for vegetation sampling in census shall be included in the Annual Table 3.6.20-1 shall be followed, including Radiological Environmental Operating analysis of appropriate controls. Report. Mith a land use census identifying a milk animal location(s) that represents a calculated D/g value greater than the D/g value currently being used in Specifica-tion 4. 6. 13. b. 3, identify the new location(s) in the next Semi-Annual Radioactive Effluent Release Report. Amendment No. 66 Nine Mile Point - Unit 1 3/4 6-151
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUiREMENT 3.6.22 LAND USE CENSUS (Cont'd) If the 0/Q value at a new milk sampling location is significantly greater (50K) than the D/Q value at an existing milk sampling location, add the new location to the radiological environmental monitoring program within 30 days. The sampling location(s) excluding the control station location, having the lowest calculated D/Q may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Specification 6. 9. l. e identify the new location(s) in the next Semi-Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the Offsite Dose Calculation Manual reflecting the new location(s). Amendment No. 66 Nine Mile Point - Unit I 3/4 6-152
Document Name: NNP"1 TS SEC B 3/4 1 Requestor's ID: CYNTHIA Author's Name: JAMERSON) C. Document Comments: ETPB Rev. 9/10/86 PLEASE RETURN THIS SHEET WITH REVISIONS
BASES FOR 3.1.8 AND 4.1.8 HIGH PRESSURE COOLANT INJECTION The High Pressure Coolant Injection System (HPCI) is provided to ensure adequate core cooling in the unlikely event of a small reactor coolant line break. The HPCI System is required for line breaks which exceed the capability of the Control Rod Drive pumps and which are not large enough to allow fast enough de-pressuriza-tion for core spray to be effective. One set of high pressure coolant injection pumps consists of a condensate pump, a feedwater booster pump and a motor driven feedwater pump. One set of pumps is capable of delivering 3,800 gpm to the reactor vessel at reactor pressure. The performance capability of HPCI alone and in conjunction with other systems to provide adequate core cooling for a spectrum of line breaks is discussed in the Fifth Supplement of the FSAR. In determining the operability of the HPCI System the required performance capability of various components shall be considered.
- a. The HPCI System shall be capable of meeting its pump head versus flow curve.
- b. The motor driven feedwater pump shall be capable of automatic initiation upon receipt of either an auto-matic turbine trip signal or reactor low-water-level signal.
- c. The Condenser hotwell level shall not be less than 57 inches (75,000 gallons).
- d. The Condensate storage tanks inventory shall not be less than 105,000 gallons.
e. During reactor start-up, operation, and shutdown the condensate and feedwater booster pumps are in operation. At reactor pressures up to 450 psig, these pumps are capable of supplying the required 3,800 gpm. Above 450 psig a motor-driven-feedwater pump is necessary to provide the required flow rate. The capability of the condensate, feedwater booster and motor driven feedwater pumps will be demonstrated by their operation as part of the feedwater supply during normal station operation. Stand-by pumps will be placed in service at least quarterly to supply feedwater during station operation. An automatic system initiation test will be performed at least once per operating cycle. This will involve automatic starting of the motor driven feedwater pumps and flow to the reactor vessel. Yi,k,< suqha Q ggv n 9M~~ q
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cg Lho~. Nine Mile Point - Unit 1 B 3/4 1-22
Document Name: NMP"I TS SEC B 3/4 2 Requestor's ID: NORMA Author's Name: Jamerson,C. Document Comments: ETPB Rev. 9/22/86 KEEP THIS SHEET WITH DOCUMENT
BASES FOR 3. 2. 1 REACTOR VESSEL HEATUP AND COOLDOWN Design calculations reported in Volume I, Section V-A,4. 0 (p. V-6)" have demonstrated that the heatup and cooldown rate of 100F/hr considered in the fatigue analysis will result in stresses well within code limits. A series of calculations have demonstrated that various extreme heatup and cooldown transients result in thermal strains well within the ASME Code limits stated in Volume I, Section V-C,3.0 (p. V-19)~. Cooldown incidents include: failure of the pressure regulator leading to a cooldown of 215F in 5. 5 minutes (Appendix E-I,3.15 (p.E-45))", inadvertent opening of a single solenoid-actuated pressure relief valve leading to a cooldown of 1050F/hr sustained for 10 minutes (Vol. I, Section V-B,1. 3 (p.V-ll))", and finally, opening all six of the solenoid-actuated relief valves leads to a cooldown of 250F in 7. 5 minutes (Volume IV, Section I-B)". Reactor vessel heatup of 300F/hr (Volume IV, Section I-B)" also demonstrates stresses well within the code requirements. In view of the reported results, the specified heatup and cooldown rates are believed to be conservative.
- FSAR Nine Nile Point - Unit 1 B 3/4 2-1
BASES FOR 3.2.2 AND 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figure 3. 2. 2a is a plot of pressure vs. temperature for a heatup and cooldown rate of 100F/hr. maximum.. (Specification 3. 2.1). This curve is based on calculations of str s intensity factors according to Appendix G of Section III of the ASME Boiler and Pressure Vessel de. The temperature limit of 135F represents the minimum permissable temperature for the inservic hydrostatic pressure test at the test pressure of 1210 psig. This limit is based on Section XI of t ASME Boiler and Pressure Vessel Code. The basic data for Figure 3.2.2.c for A302B/A533B - Class 1 steel is based on 30 ft. - lb. Charpy'-notch energy transition temperatures which have been correlated w h drop weight specimen nil ductility transition for this steel. At the design exposure of 5 x 10~~ nvt th change in NDTT is 65F. This shift is applicable to only the beltline region material. The reactor vessel head flange and the vessel flange combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted to ether. When the vessel head is placed on the reactor vessel, only that portion of the head flange near e inside of the vessel rests on the vessel flange. the head bolts are replaced and tensioned, the ve sel head is flexed slightly to bring together the entire
's contact surfaces adjacent to the "0" rings of t head and vessel flange. Both the head and vessel flange have a NDT temperature of 40F and they are not subject to any appreciable neutron radiation exposure.
Therefore, the minimum vessel head and head ange temperature for bolting the head flange and vessel flange is established as 40 + 60F or lOOF. The integrated neutron flux at the vess wall is calculated form core physics data and will be measured using flux monitors installed inside t vessel. This measured flux will be used to check and correct the calculated data to deter ne an accurate flux. From this data a conservative NDTT temperature if necessary can be determined. Since no shift 11 occur until an integrated flux of 10'7 nvt is reached the confirmation can be made well in a vance of any shift. Vessel material surveillance sa les are located within the core region to permit periodic monitoring of exposure and material properti s relative to control samples. The material sample program conforms with ASTM E 185-66 with material withd wal schedule as specified in Specification 4.2.2.c. In addition, samples will iso be installed to monitor the sensitized stainless steel components. Samples consisting of sensitized stainless steel forgings and strips and annealed material will be located in the steam, mixture, and wa r phases inside the reactor vessel. Detailed laboratory examination of these samples would be requ ed if inspections and/or analyses of other conditions, e. g., substantial deviations in primary coolant c emistry, indicate that stress corrosion cracking of the sensitized stainless steel hfil!. +num~ Nine Mile Point - Unit 1 B 3/4 2-2
0 BASES FOR 3.2.2 AND 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figures 3. 2. 2. a and 3.2. 2. b are plots of pressure versus temperature for a heatup and cooldown rate of 100'F/hr. maximum. (Specification 3. 2. 1). Figure 3. 2.2. c is a plot of pressure versus temperature for hydrostatic testing. These curves are based on calculations of stress intensity factors according to Appendix G of Section III of the ASME Boiler and Pressure Vessel Code 1980 Edition with Minter 1982 Addenda. In addition, temperature shifts due to integrated neutron flux at eleven effective full power years of oper-ation were incorporated into the figures. These shifts were calculated from the formula presented in Regula-tory Guide 1. 99, proposed Revision 2 d e co er~osp oru o reactor vess . These curves are applicable to the beltline region avow and e eva e temperatures and tTte vessel ange at intermediate temperatures. Reactor vessel flange/reactor head flange boltup is governed by other criteria as stated in. Specification 3.2.2.d. The pressure readings on the figures have been adjusted to reflect the calculated elevation head difference between the pressure sensing instrument locations and the pressure sensitive area of the core beltline region. The reactor vessel head flange and vessel flange in combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted together. Mhen the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flange. Both the head and vessel and flange have a NDT temperature of 40F and they are not subject to any appreciable neutron radiation exposure. Therefore, the minimum vessel head and- head flange temperature for bolting the head flange and vessel flange is established as 40 + 60F or lOOF. Figures 3. 2. 2. a, 3. 2. 2. b and 3. 2. 2. c have incorporated a temperature shift due to the calculated integrated neutron flux. The integrated neutron flux at the vessel wall is calculated from core physics data and has been measured using flux monitors installed inside the vessel. The curves are applicable for up to eleven . effective full power years of operation. Vessel material surveillance samples are located within the core region to'permit periodic monitoring of exposure and material properties relative -.to control samples. The material sample program conforms with ASTM E 185-66 except for the material withdrawal scheduled which is specified in Specification 4.2.2.b. Amendment No. SH, 85 Nine Mile Point - Unit 1 B 3/4 2-3
BASES FOR 3.2.3 AND 4.2.3 COOLANT CHEMISTRY Haterials in the primary system are primarily 304 stainless steel and the Zircaloy fuel cladding. The reactor water chemistry limits are established to prevent damage to these materials. Limits are placed on chloride concentration and conductivity. The most important limit is that placed on chloride concentration to prevent stress corrosion cracking of the stainless steel. When the steaming rate is less then 100,000 pounds per hour, a more restrictive limit of 0. 1 ppm has been established. At steaming rates of at least 100,000 pounds per hour, boiling occurs causing deaeration of the reactor water, thus maintaining oxygen concentration at low levels. A short term spike is defined as a rise in conductivity such as that which could arise from injection of additional feedwater flow for a duration of approximately 30 minutes in time. When conductivity is in its proper normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range. When and if conductivity becomes abnormal, then chloride measure-ments are made to determine whether or not they are also out of their normal operating values. This would not necessarily be the case. Conductivity could be high due to the presence of a neutral salt, e. g., Na SO , which would not have an affect on pH or chloride. In such a case, high conductivity alone is not a 3au$ e for shutdown. In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives. In the case of BWR's, however, where no additi.ves are used and where neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water. Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include, operation of the reactor clean-up system, reducing the input of impurities and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the clean-up system to re-establish the purity of the reactor coolant. During start-up periods, which are in the category of less than 100,000 pounds per hour, conduc-tivity may exceed 2 pmho/cm because of the initial evolution of gases and the initial addition of dissolved metals. During this period of time, when the conductivity exceeds 2 pmho (other than short term spikes), samples will be taken to assure that the chloride concentration is less than 0. 1 ppm. Nine Hile 'Point - Unit 1 8 3/4 2-4
BASES FOR 3.2.3 AND 4.2.3 COOLANT CHEMISTRY The conductivity at the reactor coolant is continuously monitored. The samples of the coolant which are taken every 96 hours will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors. If conductivity is within its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. However, if the conductivity changes significantly, chloride measure-ments will be made to assure that the chloride limits of Specification 3. 2. 3 are not exceeded. Nine Nile Point - Unit 1 B 3/4 2-5
BASES FOR 3.2.4 AND 4.2.4 REACTOR COOLANT ACTIVITY The primary coolant radioactivity concentration limit of 25 pCi total iodine per gram of water was calculated based on a steamline break accident which is isolated in 10.5 seconds. For this accident analysis, all the iodine in the mass of coolant released in this time period is assumed to be released to the atmosphere at the top of the turbine building (30 meters). By limiting the thyroid dose at the site boundary to a maximum of 30 Rem, the iodine concentration in the primary coolant is back-calculated assuming fumigation meteorology, Pasquill Type F at 1 m/sec. The iodine concentration in the primary coolant resulting from this analysis is 25 pCi/gm. A radioactivity concentration limit of were near the limit based 25 pCi/g total iodine could only be reached on the assumed effluent isotopic content (Table A-12 of if the gaseous effluents the FSAR) and the fact that the primary coolant cleanup systems were'noperative. When the cleanup system is operating, expected that the primary coolant radioactivity would be about 12 pCi/g total iqdine. The concentrations it is expected during operations with a gaseous effluent of about O. lpCi/sec would be about 1.5 pCi/g total iodine. The reactor water sample will be used'to assure that the limit of Specification 3.2.4 is not exceeded. The total radioactive iodine activity would not be expected to change rapidly over a period of 96 hours. In addition, the trend of the stack offgas release rate, which is continuously monitored, is a good indicator of the trend of the iodine activity in the reactor coolant. Since the concentration of radioactivity in the reactor coolant is not continuously measured, coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures. However, as discussed in the bases for Specification 3. 6. 2, some capability to detect gross fuel element failures is inherent in the radiation monitors in the offgas system and on the main steam line. Nine Nile Point - Unit 1 B 3/4 2-6
BASES FOR 3.2.5 AND 4.2.5 REACTOR COOLANT SYSTEM LEAKAGE RATE Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite a-c power. The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits. The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study). Mork utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth. This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. However, the establish-ment of allowable unidentified leakage greater than that given in 3. 2. 5 on the basis of the data presently available would be premature because of uncertainties associated with the data. For leakage of the order of 5 gpm as specified in 3. 2. 5, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propaga-tion. Leakage of the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the plant should be shut down to allow further investigation and corrective action. Inspection and corrective action is initiated when unidentified leakage increases at a rate in excess of 2 gpm, within a 24 hour period or less. This minimizes the possibi lity of excessive propagation of intergranular stress corrosion cracking. A total leakage of 25 gpm is well within the capacity of the control rod drive system makeup capability (page III-7 of the First Supplement). " As discussed in 3. 1. 6 above, for leakages within this makeup capability, the core will remain covered and automatic pressure blowdown will not be actuated. The primary means of determining the reactor coolant leakage rate is by monitoring the rate of rise in the levels of the drywell floor and equipment drain lines. Checks will be made every four hours to verify that no alarms have been actuated due to high leakage. For sump inflows of one gpm, changes on the order of 0.2 gpm can be detected within 40 minutes. At inflows between one and five gpm, changes on the order of 0.5 gpm can be detected in eight minutes. "FSAR Amendment No. 70 Nine'Mile Point - Unit 1 8 3/4 2-7
h BASES FOR 3. 2.5 AND 4.2.5 REACTOR COOLANT SYSTEM LEAKAGE RATE Leakage is detected by having all unidentified leakage routed to the drywell floor drain tank and identified leakage routed directly to the drywell equipment drain tanks. Identified leakage includes such items as recirculation pump seal leakage and recirculation pump suction and discharge valve packing leakoff. Another method will monitor the time required to When fill the tanks between two accurately determined levels. the level in the tank reaches the low-level switch setting, a timer will start and operate for a preset time interval. If the timer resets before the high-level switch setting is reached indicating a leakage rate within allowable limits, no action will result, and the system resets for the next filling and timing cycle. If the leakage is high enough to cause the level to reach the high level switch setting before the timer resets automatically, an alarm is actuated indicating leak rate above the predetermined limit (First and Fifth Supplements)." Additional information is available to the operator which can be used for the shift leakage check drywell sumps level alarms are out of service. The integrated flow pumped from the sumps to the waste if the disposal system can be checked. gualitative information is also avai lable to the operator in form of indication of drywell atmospheric condi-tions. Continuous leakage from the primary coolant system would cause an increase in drywell temperature. Any leakage in excess of 15 gpm of steam would cause a continuing increase in drywell pressure with resulting scram (First Supplement) * ~ Either the rate of rise leak detection system, the timer, leak detection system or the integrated flow can be utilized to satisfy Specification 3. 2. 5.b.
- FSAR Amendment No. 70 Nine Mile Point - Unit 1 B 3/4 2-8
0 BASES FOR 3.2.6 AND 4.2.6 INSERVICE INSPECTION AND TESTING The inservice inspection and testing program for the Nine Mile Point Unit 1 plant conforms to the require-ments of 10 CFR 50, Section 50. 55a(g). Mhere practical, the inspection of components, pumps and valves classified into NRC equality Groups A, B and C conforms to the requirements of ASME Code Class 1, 2 and 3 components, pumps and valves, respectively, contained in Section XI of the ASME Boiler and Pressure Vessel Code. If a Code required inspection is impractical for the Nine Mile Point Unit 1 facility, a request for a deviation from that requirement is submitted to the Commission in accordance with 10 CFR 50, Section 50.55a(g)(6)(i). Deviations which are needed from the procedures prescribed in Section XI of the ASHE Code and applicable Addenda will be reported to the Commission prior to the beginning of each 10-year inspection period are known to be required at that time. Deviations which are identified during the course of inspection if they will be reported quarterly throughout the inspection period. The augmented inservice inspection program for the Nine Mile Point Unit 1 plant conforms to the schedules contained in NUREG 0313 Revision l. It is performed in order to detect and survey intergranular stress corrosion cracking of ASME Code Class 1, 2 and 3 pressure boundary piping. Inspections shall be performed by individuals qualified to (1) the ASME Boiler and Pressure Vessel Code, Section XI, as specified to the NRC, and (2) Ultrasonic Testing Operator Training for Intergranular Stress Corrosion Cracking developed by the EPRI Non-Destructive Examination Center, as specified to the NRC. References (1) Letter form the Nuclear Regulatory Commission (D. B. Vassallo) to Niagara Mohawk Power Corporation (G. K. Rhode), dated September 19, 1983. (2) Letter from Niagara Mohawk Power Corporation (D. PE Disc) to the Nuclear Regulatory Commission (T. A. Ippolito), dated August 7, 1981. Amendment No. 57 Nine Mile Point - Unit 1 B 3/4 2-9
BASES FOR 3.2.7 AND 4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Oouble isolation valves are provided in lines which connect to the reactor coolant system to assure isolation and minimize reactor coolant loss in the event of a line rupture. The specified valve requirements assure that isolation is already accomplished with one valve shut or provide redundancy in an open line with two operative valves. Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable of automatic initiation and the closure times presented in Table 3. 2. 7. These closure times were selected to minimize coolant losses in the event of the specific line rupturing. Using the longest closure time on the main-steam-line valves following a main-steam-line break (Appendix E-II. 1.0"), the core is still covered by the time the valves close. Following a specific system line break, the cleanup and shutdown cooling closing times will upon initiation from a low-low level signal limit coolant loss such that the core is not uncovered. Feedwater flow would quickly restore coolant levels to prevent clad damage. Closure times are discussed in Section VI-C. 1.0". The valve operability test intervals are based on periods not likely to significantly affect operations, and are consistent with testing of other systems. Results obtained during closure testing are not expected to differ appreciably from closure times under accident conditions as in most cases, flow helps to seal the valve. The test interval of once per operating cycle for automatic initiation results in a failure probability of
- 1. 1 x 10-~ (Fifth Supplement, p. 115)* that a line will not isolate. More frequent testing for valve operability results in a more reliable system.
AFSAR Nine Mile Point Unit 1 B 3/4 2-10
4 BASES FOR 3.2.8 AND 4.2.8 PRESSURE-RELIEF SYSTEM"SAFETY VALVES The required number of operable safety valves is based on the analysis presented in Appendix E-I.3.7" which assumed reactor isolation with no scram. Operation of all 16.safety valves will limit reactor pressure below the safety limit of 1375 psig. Partial redundancy is provided by the solenoid-actuated pressure relief valves as the relieving capacity of each of these valves is approximately the same as a safety valve, as discussed in 2.2.2 above. The safety valve testing and intervals between tests are based on manufacturer's recommendations and past experience with spring actuated safety valves. ~FSAR Nine Mile Point - Unit 1 B 3/4 2-11
4 BASES FOR 3.2.9 AND 4.2.9 PRESSURE RELIEF SYSTEM - SOLENOID ACTUATED PRESSURE RELIEF VALVES As discussed in 2. 2. 2 and 3. 2.8 above, the solenoid-actuated pressure relief valves are used to avoid actuation of the safety valves. The set points of the six relief valves are staggered. Two valves are set at 1090 psig, two are set at 1095 psig, and two are set at 1100 psig. The operator will endeavor to place the set-point at these figures. However, a set-point error for each valve can be as much as +12 psig. Six valves are provided for the automatic depressurization function, as described in 3.1.5. However, only five valves are required to prevent actuation of the safety valves, as discussed in the Technical Supplement to Petition to Increase Power Level, Section II. XV, letter, T. J. Brosnan to Peter A. Morris dated February 28, 1972, and letter, Philip D. Raymond to A. Giambusso, dated October 15, 1973. The basis for the surveillance requirement is given in 4.1. 5. Nine Nile Point - Unit 1 B 3/4 2-12
Document Name: NMP-1. TS SEC B 3/4 3 Requestor's ID: CYNTHIA Author's Name: Jamerson, C. Document Comments: ETPB Rev. 9/10/86 KEEP THIS SHEET WITH DOCUMENT
BASES FOR 3.3.1 AND 4..3.1 OXYGEN CONCENTRATION The four percent oxygen concentration eliminates the possibility of hydrogen combustion following a loss-of-coolant accident (Section VII-G.2.0 and Appendix E-11.5.2)." The only way that significant quantities of hydrogen could be generated would be if all core spray systems failed to sufficiently cool the core. As discussed in Section VII-A.2.0 and illustrated in Figure VII-2," the core spray system is capable of design flow of 3400 gpm at a reactor pressure of 113 psig. In addition to hydrogen generated by metal-water reaction, significant quantities can be generated by radiolysis. (Technical Supplement to Petition for Conversion from Provisional Operating License to Full Term Operating License.) At reactor pressures of 110 psig or less, the reactor will have been shut down for more than an hour and the decay heat will be at sufficiently low values so that fuel rods will be completely wetted by core spray. The fuel clad temperatures would not exceed the core spray water saturation temperature of about 344'F. The occurrence of primary system leakage following a major refueling outage or other, scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified, oxygen concen-tration limit is based. Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus to preclude the possibility of starting the reactor and operating for extended periods of time with signifi-cant leaks in the primary system, leak inspections are scheduled during startup periods when the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be reasonable to perform the leak inspection and establish the required oxygen concentration. The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the containment but air could not leak in to increase the oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentra-tion is necessary. However, at least once a week, the oxygen concentration will be determined as added assurance that Specification 3. 3. 1 is being met. "FSAR Nine Nile Point - Unit 1 B 3/4 3-1
0 4
BASES FOR 3.3.2 AND 4.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL The values specified for suppression chamber water temperature, maximum downcomer submergence, and system pressures are based on the effect these parameters have on the short-term post-accident system pressure following a loss-of-coolant accident. The combinations shown on Figures 3. 3. 2 a and b and the water level required are based on maintaining the post-accident pressure below the design value of 35 psig and the maxi-mum suppression chamber water temperature below 140 F in the containment design basis loss-of-coolant accident (Appendix E-11.2. 2. 3)." The calculational basis for the pressure suppression system initial conditions, Figures 3. 3. 2 a and b are presented in the Fifth Supplement." The three foot minimum and the four and one-half foot maximum submergence are a result of the Mark I Contain-ment Long Term Program. The 215~F limit for the reactor is specified, since below this temperature the containment can tolerate a blowdown without exceeding the 35 psig design pressure of the suppression chamber without condensation. Actually, for reactor temperatures up to 312'F the containment can tolerate a blowdown without exceeding the 35 psig design pressure of the suppression chamber, without condensation. Some experimental data suggests that excessive steam condensing loads might be encountered if the bulk temper-ature of the suppression pool exceeds 160'F during any period of relief valve operation with sonic conditions at the discharge exit. This can result in local pool temperatures in the vicinity of the quencher of 200'F. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings. In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. As a minimum, this action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool. "FSAR Amendment No. 2, 76 Nine Mile Point - Unit 1 B 3/4 3-2
BASES FOR 3.3. 2 AND 4.3. 2 PRESSURE SUPPRESSION SYSTEH PRESSURE AND SUPPRESSION CHAHBER WATER TEHPERATURE AND LEVEL Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high load-ings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress. Continuous monitoring of suppression chamber water level and temperature and pressure suppression system pressure is provided in the control room. Alarms for these parameters are also provided in the control room. To determine the status of the pressure suppression system, inspections of the suppression chamber interior surfaces at each major refueling outage with water at its normal elevation will be made. This will assure that gross defects are not developing. Nine Nile Point - Unit 1 B 3/4 3-3
BASES FOR 3.3.3 AND 4. 3. 3 LEAKAGE RATE The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident. The peak drywell pressure would be 35 psig which would rapidly reduce to 22 psig within 100 seconds following the pipe break. The total time the dry-well pressure would be above 22 psig is calculated to be about 10 seconds. Following the pipe break, the suppression chamber pressure rises to 22 psig within 10 seconds, equalizes with drywell pressure and there-after rapidly decays with the drywell pressure decay. (1) The design pressures of the drywell and absorption chamber are 62 psig and 35 psig, respectively. (2) The design leak rate is 0.5X/day at a pressure of 35 psig. As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after, about 10 seconds. Based on the calculated primary containment pressure response discussed above and the suppression chamber design pressure; primary containment preoperational test pressures were chosen. Also, based on the primary containment pres-sure response and the fact that the drywell and a suppression chamber function as a unit, the primary contain-ment will be tested as a unit rather than testing the individual components separately. The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1. 9X/day at 35 psig. The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90 percent for halogens, 95 percent for particulates, and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about
- 6. 0 rem and the maximum total thyroid dose is about 150 rem at the site boundary considering fumigation con-ditions over an exposure duration of two hours. The resultant doses would occur for the duration of the accident at the low population distance of 4 miles are lower than those stated due to the variability of meteorological conditions that would be expected to occur over a 30-day period. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident. These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct re-lease of fission products from the primary containment through the filters and stack to the environs. There-fore,- the specified primary containment leak rate and filter efficiency (Specification 4.4.4) are conserva-tive and provide margin between expected offsite doses and 10 CFR 100 guideline limits.
The maximum allowable test leak rate as specified in 4. 3. 3. b is 1. 5X/day at a pressure of 35 psig. This value for the test condition was derived from the maximum allowable accident leak rate of about 1.9X/day when cor-rected for the effects of containment environment under accident and test conditions. In the accident case, the containment atmosphere initially would be composed of steam and hot air depleted of oxygen whereas under test conditions the test medium would be air or nitrogen at ambient conditions. Considering the differences in mixture composition and temperatures, the appropriate correction factor applied was 0.8 and determined from the guide on containment testing. (3) Nine Mile Point - Unit 1 B .3/4 3-4
BASES FOR 3. 3. 3 AND 4.3. 3 LEAKAGE RATE Although the dose calculations suggest that the allowable test leak rate could be allowed to increase to about
- 3. OX/day before the guideline thyroid dose limit given in 10 CFR 100 would be exceeded, establishing the limit at 1.5X/day provides an adequate margin of safety to .assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime.
Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate. The operational limit is derived by multiplying the allowable test leak rate by 0.75 thereby providing a 25K margin to allow for leakage deterioration which may occur during the period between leak rate tests. The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. The leak rate test frequency is based on the AEC guide for developing leak rate testing and surveillance of reactor containment vessels. (4) The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure. It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized. If the leakage rates of the double-gasketed seal penetrations, testable penetration isolation valves, containment air purge inlets and outlets and the vacuum relief valves are at the maximum specified, they will total 90 percent of the allowed leak rate. (2) Hence, 10 percent margin is left for leakage through walls and untested components. Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a practical minimum. (1) Appendix E, FSAR. (2) Volume 1, Section VI, FSAR. (3)- TID-20583, Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Rate Determinations. (4) 10 CFR 50 Appendix J, "Reactor Containment Leakage Testing for Water Cooled Power Reactors." Nine Mile Point - Unit 1 B 3/4 3-5
BASES FOR 3.3.4 AND 4.3.4 PRIMARY CONTAINMENT ISOLATION VALVES Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Except where check valves are used as one or both of a set of double isolation valves, the isolation closure times are presented in Table 3.3.4. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accidentt Details of the isolation valves are discussed in Section VI-C. " For allowable leakage rate specification, see Section 3.3.3 above. As illustrated in Figure E-34 of Appendix E" fuel rod perforation does not occur until about 150 seconds following the loss-of-coolant accident. A required closing time of 60 seconds for all primary containment isolation valves will be adequate to prevent fission product release through lines connecting to the primary containment. For reactor coolant system temperatures less than -312'F, the containment could not become pressurized due to a loss-of-coolant accident. The 215~F limit is based on preventing pressurization of the reactor building and rupture of the blowout panels. The test interval of once per operating cycle for automatic initiation results in a failure probability of . 1.1 x 10- that a line will not isolate (Fifth Supplement, p. 115).* More frequent testing for valve opera-bility results in a more reliable system. In addition to routine surveillance as outlined in First Addendum to Technical Supplement to Petition to increase Power Level, each instrument-line flow check valve will be tested for operability. All instruments on a given line will be isolated at each instrument. The line will be purged by isolating the flow check valve, opening the bypass valves, and opening the drain valve to the equipment drain tank. When pur'ging is sufficient to clear the line of non-condensibles and crud the flow-check valve will be cut into service and the bypass valve closed. The main valve will again be opened and the flow-check valve allowed to close. The flow-check valve will be reset by closing the drain valve and opening the bypass valve depressurizing part of the system. Instruments will be cut into service after closing the bypass valve. Repressurizing of the individual instruments assures that flow-check valves have reset to the open position. "FSAR Nine Mile Point - Unit 1 B 3/4 3-6
BASES 3.3.5 AND 4.3.5 ACCESS CONTROL Access to the containment during operation is expected to be infrequent. However, each door of the two double-doored access locks is designed to withstand 62 psig drywell pressure. It is, therefore, possible to open one door at a time and still maintain containment integrity. Access door design is discussed in Sec-tion VI-A.2.2 of the FSAR. The equipment hatch and drywell head and other flanged openings are provided with double "0" rings and must be secure in order to maintain the integrity of the primary containment system. Maintaining the pressure suppression system integrity when above the stated pressure and temperature will ensure that a reactor coolant system rupture will not result in an overpressurization of the reactor building. Nine Mile Point - Unit 1 B 3/4 3-7
BASES FOR 3.3.6 AND 4.3.6 VACUUM RELIEF Four vacuum relief valves are provided between the drywell and suppression chamber (Section VI-A.1.5 and 2.6"). Each valve is capable of opening on a differential pressure of 0. 25 + 0. 10 psi. The operation of any one valve will prevent damage to the drywell under the accident conditions expected following the loss-of-coolant accident due to a recirculation line break. As discussed in Section VI-F,* one valve operation will limit maximum pressure differential between the two chambers to approximately 3 psi, well below the maximum allowable pressure differential of 8. 94 psi. At a coolant temperature of 215 F, the steam generated during a loss-of-coolant accident would not be sufficient to purge the drywell or suppression chamber. Three sets of vacuum relief valves are provided between the primary containment and atmosphere (Section VI-A.1.5 and 2.6"). Each valve is capable of opening on a differential pressure of 0.25 + 0.10 psi. As discussed in Section VI-A.2.6,* operation of all three relief valve sets will prevent containment pressure from dropping below the vacuum ratings of the drywell and the suppression chamber. The selection of these valves is based on the conservative assumption that the ventilation valves on the suppression chamber were left open during a postulated loss-of-coolant accident, permitting the pressure suppression system to blow down to atmospheric pressure. Closure of the ventilation valves followed by startup of the containment spray and core spray pumps leads to a rapid condensation of the steam in the drywell and a consequent drop in pressure below atmospheric. Normally, the ventilation valves are locked closed and there is little likelihood of this series of events occurring. Subsequent calculations showed that with only two valve sets operating, the worst vacuum in the suppression chamber is -3.0 psig. At this pressure a safety factor of about 1.70 still exists to incipient buckling. Near ly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required. Replacement of principal system components could necessitate outages of more than 15 days. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months. Using an analysis which is the same as used in the Fifth Supplement (page 115)" results in a failure probabil-ity of 1.8 x 10- for the drywell to suppression chamber valves and a failure probability of 9. 5 x 10- for the valves between the containment and the atmosphere.
- FSAR Nine Nile Point - Unit 1 B 3/4 3-8
BASES FOR 3.3.6 AND 4.3.6 VACUUM RELIEF Each drywell-suppression chamber vacuum breaker is equipped with two independent switches to indicate the opening of the valve disk. Redundant control room alarms are provided to permit detection of any drywell-suppression chamber vacuum breaker opening in excess of the described allowable limits. The containment design has been examined to establish the allowable bypass area between the drywell and suppression chamber as 0.053 square feet. The limit on each individual valve will be set such that with all valves at their limit, the maximum value, of cumulative leakage will not exceed the maximum allowable. The value will be at approximately 0. 06 inch of disk travel off its seat and will be alarmed in the control room. The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chamber and between the suppression chamber and reactor building so that the structural integrity of the containment is maintained. The vacuum relief system from the pressure suppression chamber to the reactor building consists of three vacuum relief breakers (3 parallel sets of 2 valves in series). Operation of either system will maintain the pressure differential less than 1 psig; the external pressure is 2 psig. The leak rate testing program is based on AEC guidelines for development of leak rate testing and surveillance schedules for reactor containment vessels. Surveillance of the suppression chamber-reactor building vacuum breakers consists of operability checks and leakage tests (conducted as part of the containment leak-tightness tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition. Therefore, a testing frequency of three months for operability is considered justified for this equipment. Inspections and calibrations are performed during the refueling outages, this frequency is based on equipment quality, experience, and engineer-ing judgment.- During each refueling outage, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by approximately 1 psi with respect to the suppression pool pressure and then held constant. The subsequent suppression chamber transient will be monitored with a sufficiently sensitive pressure instrument. If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure, it would indicate exis-tence of a significant leakage path which will be identified and eliminated before further drywell vacuum breaker testing. Nine Mile Point - Unit 1 B 3/4 3-9
BASES FOR 3.3.7 AND 4.3.7 CONTAINMENT SPRAY SYSTEM For reactor coolant temperatures less than 215'F not enough steam is generated during a loss-of-coolant acci-dent to pressurize the containment.= In fact, for coolant temperatures up to 312 F, the resultant loss-of-coolant accident pressure would not exceed the design pressure of 35 psig. Operation ef only one containment spray pump is sufficient to provide the required containment spray flow. The specified flow of 3000 gpm (approximately 95 percent to the drywell and the balance to the suppression chamber) is sufficient to remove post-accident core energy released including a substantial chemical reac-tion involving hydrogen generation and will also limit pressure and temperature rises in the pressure sup-pression system to below design values (Appendix E-II 2, 2. 3 p. E-78 and the Fifth Supplement). " Each containment spray system is considered operable when both pumps are capable of delivering at least 3000 gpm at a pump developed head of 375 feet of water at 60'F. Requiring both pumps in both systems operable (400 percent redundancy) will assure the availability of the containment spray system. All'owable outages are specified to account for components that become inoperable in both systems and for more than one component in a system. t The correspondin~ raw water cooling system is designed to maintain containment spray water temperature no greater than 140 F under the most limiting operating conditions. The containment spray raw water cooling system is considered operable when the flow rate is not less than 3000 gpm and the pressure on the raw water side of the containment spray heat exchangers is not less than 160 psig. The higher pressure on the raw water side will assure that any leakage is into the containment spray system. Electrical power for all system components is normally available from the reserve transformer. Upon loss of this service the pumping requirement will be supplied from the diesel generator. At least one diesel generator shall always be available to provide backup electrical power for one containment spray system, corresponding raw water cooling system and associated electronic equipment required for automatic system initiation. Automatic initiation of the containment spray system assures that the containment will not be overpressurized due to hydrogen generation. This automatic feature would only be required if all core spray system malfunc-tioned and significant metal-water reaction occurred. For the normal condition of 90F suppression chamber water and 2 psig containment pressure, containment spray actuation would not be necessary for about 15 minutes. Raw water cooling affects the temperature of the spray water and the suppression chamber pool. Taking into AFSAR Nine Mile Point - Unit 1 8 3/4 3-10
BASES FOR 3.3. 7 AND 4.3. 7 CONTAINMENT SPRAY SYSTEM account the reduced steam condensation capability and increased suppression chamber vapor pressure, the raw water cooling would not be required for more than 20 minutes for initial suppression chamber temperatures up to 110'F. This assumes that all core spray systems fail. Therefore, manual initiation of the raw water system is acceptable. Nearly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required. Replacement of principal system components could necessitate outages of more than 15 days. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months. In conjunction with containment spray pump operation during each operating cycle, the raw water pumps and associated cooling system performance will be observed. The containment spray system shall be capable of automatic initiation from simultaneous low-low'reactor water level and high containment pressure. The associated raw water cooling system shall be capable of manual actuation. Operation of the containment spray system involves spraying water into the atmosphere of the containment. Therefore, periodic system tests are not practical. Instead separate testing of automatic containment spray pump startup will be performed during each operating cycle. During pump operation water will be recycled to the suppression chamber. Air tests to determine flows to spray headers will also be performed at this time and compared to initial pre-operational air testing, verifying that piping and/or nozzle conditions have not changed significantly. Design features are discussed in Volume I, Section VII-B.2.0 (page VII-19*). The valves in the containment spray system are normally open and are not required to operate when the system is called upon to operate. The test interval between operating cycle results in a system failure probability of 1.1 x 10 (Fifth Supplement, page 115*) and is consistent with practical considerations. Pump operability will be demon-strated on a more frequent basis and will provide a more reliable system. The intent of Specification 3. 3. 7f is to allow control rod drive maintenance and instrument replacement at the time that the suppression chamber is unwatered and to perform normal fuel movement activities in the refuel mode with an unwatered suppression chamber. Based on the limited time involved in performance of the concurrent maintenance tasks, procedural controls to minimize the potential and duration of leakage from the control rod drive housing or instrument penetra-tion and available coolant makeup provides adequate protection against drainage of the vessel while the suppression chamber is drained.
- FSAR Amendment No. 83 Nine Mile Point - Unit 1 B 3/4 3-11
Document Name: NMP"1 TS SEC B 3/4 4 Requestor's ID: CYNTHIA Author's Name: Jamerson, C Document Comments: ETPB REV 9/10/86 Please return this sheet with revisions
0 BASES FOR 3.4. 1- and 4.4. 1 LEAKAGE RATE In the answers to guestions II-3 and IV-5 of the Second Supplement and also in the Fifth Supplement," the relationships among wind speed direction, pressure distribution outside the building, building internal pres-sure, and reactor bui lding leakage are discussed. The curve of pressure in Figure 3.4. 1 represents the wind direction which results in the least building leakage. It is assumed that when the test is performed, -the wind direction is that which gives the least leakage. If the wind direction was not from the direction which gave the least reactor building leakage, building internal pressure would not be as negative as Figure 3.4.1 indicates. Therefore, to reduce pressure, the fan flow rate would have to be increased. This erroneously indicates that reactor building leakage is greater than if wind direction were accounted for. If wind direction were accounted for, another pressure curve could be used which was less negative. This would mean that less fan flow (or measured leakage) would be required to establish building pressure. However, for simplicity it is assumed that the test is conducted during con-ditions leading to the least leakage while the accident is assumed to occur during conditions leading to the greatest reactor building leakage. As. discussed in the Second Supplement and Fifth Supplement, the pressure for Figure 3.4.1 is independent of the reactor building leakage rate referenced to zero mph wind speed at a negative differential pressure of 0.25 inch of water. Regardless of the leakage rate at these design conditions, the pressure versus wind speed relationship remains unchanged for any given wind direction. By requiring the reactor building pressure to remain within the limits presented in Figure 3.4.1 and a reac-tor building leakage rate of less than 2000 cfm, exfiltration would be prevented. This would assure that the leakage from the primary containment is directed through the filter system and discharged from the 350-foot stack.
"FSAR Bases Change of 5-16-85 Nine Mile Point - Unit 1 B 3/4 4-1
BASES FOR 3.4.2 and 4.4.2 REACTOR BUILDING INTEGRITY ISOLATION VALVES Isolation of the reactor building occurs automatically upon high radiation of the normal building exhaust ducts or from high radiation at the refueling platform (See 3.6.2). Isolation will assure that any fission products entering the reactor building will be routed to the emergency ventilation system prior to discharge to the environment (Section VII-H.3.0 of the FSAR). Nine Nile Point - Unit 1 B 3/4 4-2
BASES fOR 3.4.3 and 4.4.3 ACCESS CONTROL The reactor building serves as a secondary containment during normal Station operations and as a primary con-tainment during refueling and other periods when the pressure suppression system is open. Maintaining the building integrity and an operative emergency ventilation system for the conditions listed will ensure that any fission products inadvertently released to the reactor building will be routed through the emergency ven-tilation system to the stack. The worst such incident is due to dropping a fuel assembly on the core during refueling. The consequences of this are discussed in Appendix E-II.3. 0 of the FSAR. As discussed in Section VI-F" all access openings of the reactor building have as a minimum two doors in series. Appropriate local alarms and control room indicators are provided to always insure that reactor building integrity is maintained. Maintaining closed doors on the pump compartments ensures that suction to the core and containment spray pumps is not lost in case of a gross leak from the suppression chamber. "FSAR Nine Mile Point - Unit 1 B 3/4 4-3
BASES FOR 3.4.4 and 4.4.4 EMERGENCY VENTILATION SYSTEM The emergency ventilation system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. Both emergency ventilation system fans are designed to automatically start upon high radiation in the reactor building ventilation duct or at the refueling platform and to maintain the reactor building pressure to the design negative pressure so as to minimize in-leakage. Should one system fail to start, the redundant system is designed to start automatically. Each of the two fans has 100 percent capacity. High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 per-cent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed. Opera-tion of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. Only one of the two emergency ventilation systems is needed to clean up the reactor building atmosphere upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the contain-ment system performance and reactor operation or refueling operation may continue while repairs are being made. If neither circuit is operable, the plant is brought to a condition where the emergency ventilation system is not required. Pressure drop across the combined HEPA filters and charcoal adsorbers of -less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability and pressure drop should be determi ned at least once per operating cycle to show system performance capability. The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified in Table 5-1 of ANSI 509-1980. Amendment No. 73 Nine Mile Point - Unit 1 B 3/4 4-4
BASES FOR 3.4.4 and 4.4.4 EMERGENCY VENTILATION SYSTEM The replacement charcoal for the adsorber tray removed for the test should meet the same adsorbent quality. Any HEPA filters found defective shall be replaced with filters qualified pursuant to ANSI 509-1980. All elements of the heater should be demonstrated to be functional and operable during the test of heater . capacity. Operation of the inlet heater will prevent moisture buildup in the filters and adsorber system. With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear pe-riphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated. If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use. The determination of significant shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination. Demonstration of the automatic initiation capability and operabi lity of fi lter cooling is necessary to assure system performance capability. If one emergency ventilation system is inoperable, the other system must be tested daily. This substantiates the availability of the operable system and thus reactor operation or refueling operation may continue during this period of time. Amendment No. 73 Nine Mile Point - Unit j. 8 3/4 4-5
BASES FOR 3.4.5 and 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM The control room air treatment system is designed to filter the control room atmosphere for intake air. A roughing filter is used for recirculation flow during normal control room air treatment operation. The con-trol room air treatment system is designed to automatically start upon receipt of a high radiation signal from one of the two radiation monitors located on the ventilation intake and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage. High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, adequate radiation protection will be provided such that resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the system is found to be inoperable, there is no immediate threat to the control room and reactor opera-tion or refueling operation may continue for a limited period of time while repairs are being made. If the makeup system cannot be repaired within seven days, the reactor is shut down and brought to cold shutdown within 36 hours or refueling operations are terminated. Pressure drop across the combined HEPA filters and charcoal adsorbers of less than six inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operating cycle to show system perfor-mance capability. In addition, air intake radiation monitors will be calibrated and functionally tested each operating cycle, not to exceed 24 months, to verify system performance. The frequency of tests and sample analysis are necessary to show the HEPA filters and charcoal adsorbers can perform as evaluated. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent quali-fied according to Table 5-1 of ANSI 509-1980. The replacement charcoal for the adsorber tray removed for the test should meet the same adsorbent quality. Any HEPA filters found defective shall be replaced with filters qualified pursuant to ANSI 509-1980. Amendment No. 73 Nine Mile Point - Unit 1 B 3/4 4-6
BASES FOR 3.4.5 and 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM Operation of the system for 10 hours every month will demonstrate operability of the filters and adsorber system and remove excessive moisture bui lt up on the adsorber. If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign materials, the same tests and sample analysis shall be performed as required for operational use. The determination of significant shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination. Nine Mile Point - Unit 1 B 3/4 4-7
Document Name: NMP-1 TS SEC B 3/4 5 Requestor's ID: NORMA Author's Name: Jamerson, C Document Comments: ETPB Rev 9/22/86 Please return this sheet with revisions
BASES FOR 3.5.1 AND 4.5.1 SOURCE RANGE MONITORS The SRM's are provided to monitor the core during periods of Station shutdown and to guide the operator dur-ing refueling operations and Station startup. Requiring three operative SRM s will ensure adequate coverage for all possible critical configurations produced by fuel loading or dispersed withdrawals of control rods during Station startup. Allowing withdrawal of the SRM while maintaining a high count rate will extend the operating range of the SRM'. Evaluation of the SRM operation is presented in Section VIII-C.l. 2. 1 of the FSAR. Nine Mile Point - Unit 1 B 3/4 5-1
BASES FOR 3.5.2 AND 4.5.2 REFUELING PLATFORH INTERLOCK The control rod withdrawal block and refueling platform'travel blocks are provided to back up normal pro-cedural controls to prevent inadvertent large reactivity additions to the core. These interlocks are pro-vided even though no more than one control rod can be removed from the core at a time during refueling with the mode switch in the "refuel" position. Even in the fresh fully loaded core if a new assembly is dropped into a vacant position adjacent to the withdrawn rod, no excursion would result. This is discussed in detail in Appendix E-II.3.0 of the FSAR. There are normally two Station personnel directly involved in 'refueling the reactor, one in the control room and one at the platform. If the interlocks are inoperable, an additional person will check that "a" and "b" are not violated. Nine Hile Point - Unit 1 8 3/4 5-2
0 BASES FOR 3. 5. 3 EXTENDED CORE AND CONTROL ROD DRIVE MAINTENANCE The intent of this specification is to permit the unloading of a significant portion of the reactor core for such purposes as removal of temporary control curtains, control rod drive maintenance, in-service inspection requirements, examination of the core support plate, etc. When the refueling interlock input signal from a withdrawn control rod is bypassed, administrative controls will be in effect to prohibit fuel from being loaded into that control cell. These operations are performed with the mode switch in the "Refuel" position to provide the refueling inter-locks normally available during refueling. In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod. The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed insures that withdrawal of another control rod does not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with the control rod. Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core. The SRH's are provided to monitor the core during periods of station shutdown and to guide the operator dur-ing refueling operations and station startup. Requiring two operable SRH s, one in and one adjacent to any core quadrant where fuel or control rods are being moved, assures adequate monitoring of that quadrant during such alterations. The requirement of 3 counts per second provides assurance that neutron flux is being monitored. A spiral unloading pattern is one by which the fuel in the outermost cells (four fuel bundles surrounding a control blade) is removed first. Unloading continues by removing the remaining outermost fuel by cell. The center cell will be the last removed. Spiral reloading is the reverse of unloading. Spiral unloading and reloading will preclude the creation of flux traps (moderator filled cavities surrounded on all sides by fuel) ~ During spiral unloading, the SRM's shall have an initial count rate of 3 cps with all rods fully inserted. The count rate will diminish during fuel removal. After'll the fuel is removed from a cell, the refueling interlock will be bypassed on the corresponding control rod. Prior to withdrawal of that rod, one licensed operator and a member of the reactor analysis staff will verify that the interlock bypassed is on the correct control rod. Once the control rod is withdrawn, it will be valved out of service. Under this special condition of complete spiral core unloading, it is expected that the count rate of the SRH's will drop below 3 cps before all of the fuel is unloaded. Since there will be no reactivity additions, a lower number of counts will not present a hazard. When all of the fuel has been removed to the spent fuel storage pool, the SRH's will no longer be required. Requiring the SRM's to be operational prior to fuel re-moval assures that the SRM's are operable and can be relied on even when the count rate may go below 3 cps. Amendment No. 27 Nine Mile Point - Unit 1 B 3/4 5-3
0 BASES FOR 3.5.3 EXTENDED CORE AND CONTROL ROD DRIVE MAINTENANCE During spiral reload, SRH operability will be verified by using a portable external source every 12 hours until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, two fuel assemblies will be loaded in different cells containing control blades around each SRH to obtain the required 3 cps. Until these two assemblies have been loaded, the 3 cps requirement is not necessary. Amendment No. -27 Nine Hile Point - Unit 1 B 3/4 5-4
Document Name: NMP-1 TS SEC B 3/4 6 Requestor's ID: NORMA Author's Name: Jameson C. Document Comments: ETPB Rev. 9/22/86 PLEASE RETURN THIS SHEET WITH REVISIONS
BASES FOR 3.6.1b AND 4.6.1b MECHANICAL VACUUM PUMP ISOLATION The purpose of isolating the mechanical vacuum pump line is to limit release of activity from the main con-denser during a control rod drop accident.. During the accident, fission products would be transported from the reactor through the main-steam lines to the main condenser. The fission product radioactivity would be sensed by the main-steam line radioactivity monitor s and initiate isolation. Nine Mile Point - Unit 1 B 3/4 6-1
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The reactor protection system abtomatically initiates a reactor scram to prevent exceeding established limits. In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents or terminates operator error. The reactor protection system is a dual channel type (Table 3. 6. 2. a). Each trip system except .the manual scram has two independent instrument channels. Operation of either channel will trip the trip system, i. e., the trip logic of the channel is one-out-of-two. A simultaneous trip of both trip systems will cause a reactor scram, i. e., the tripping logic of the trip systems is two-out-of-two. The tripping logic of the total system is referred to as one-out-of-two taken twice. This system will accommodate any single failure and still perform its intended function and in addition, provide protection against spurious scrams. The reliability of the dual channel system or probability that it will perform its intended function is less than that of a one-out-of-two system and somewhat greater than that of a two-out-of-three system (Sec-tion VIII-A.1.0 of the FSAR). The instrumentation used to initiate action other than scram is generally similar to the reactor protection system. There are usually two trip systems required or available for each function. There are usually two instrument channels for each trip system. Either channel can trip the trip system but both trip systems are required to initiate the respective action. Where only one trip system is provided only one instrument chan-nel is required to trip the trip system. All instrument channels except those for automatic depressurization are normally energized. De-energizing causes a trip. Power to the trip systems for each function is from reactor protection system buses ll and 12. The signals for initiating automatic blowdown and rod block differ from other initiating signals in that only one of the two trip systems is required to start blowdown or initiate rod block. Both instrument channels in the trip system must trip to initiate automatic blowdown. This difference is due to the requirement that automatic depressurization be prevented unless A. C. power is available to the emergency core cooling systems. The instrument channels in the trip system for automatic depressurization are normally de-energized. In order to cause a trip both instrument channels must be energized. Power to energize the instrument channels is from power boards 102 and 103. If A. C. power is lost to one power board, one trip system becomes inoperable but the other trip system remains operable and capable of initiating automatic blowdown. If both power boards have lost A.C. power neither trip system can be energized and automatic blowdown is prevented. Only one instrument channel is required to initiate rod block. Amendment No. 37 Nine Mile Point - Unit 1 B 3/4 6-2
BASES FOR 3. 6. 2 AND 4. 6. 2 PROTECTIVE INSTRUMENTATION (Continued) Each reactor operating condition has a related reactor mode switch position for the safety system. The instrumentation system operability for each mode switch position is based on the requirements of the related safety system. For example, the specific high drywell pressure trip systems must be tripped or operable any time core spray, containment spray, automatic depressurization or containment isolation functions are required. In instrumentation systems where two trip systems are required to initiate action, either both trip systems are operable or one is tripped. Having one trip system already tripped does not decrease the reliability in terms of initiating the desired action. However, the probability of spurious actuation is increased. Cer-tain instrument channels or sensor inputs to instrument channels may be bypassed without affecting safe operation. The bas~ for allowing bypassing of the specified SRM's, IRM's, LPRM's and APRM's is discussed in ~~~I w Volume I (Section II C. 1.2)*. The high area temperature isolation function for the cleanup system has one trip system. There are three instrument channels; each has four sensor inputs. Only two instrument channels are required since the area covered by any one sensor is also covered by a sensor in one of the other two instrument channels. The shutdown system also has one trip system for high area temperature isolation. However, since the area of concern is much smaller, only one instrument channel is provided. Four sensors provide input to the channel. Since the area covered is relatively small only three of the four sensors are required to be operable in order to assure isolation when needed. Manual initiation is available for scram, reactor isolation and containment isolation. In order to manually initiate other systems, each pump and each valve is independently initiated from the control room. Contain-ment spray raw water cooling is not automatically initiated. Manual initiation of each pump is required as discussed in 3.3.7 above.
*FSAR; Letter, R. R. Schneider to A. Giambusso, dated November 15, 1973 Nine Mile Point - Unit 1 B 3/4 6-3
BASES FOR 3. 6.2 AND 4. 6. 2 PROTECTIVE INSTRUMENTATION (Continued)
- a. The set points included in the tables are those used in the transient analysis and the accident analysis.
The high flow set point for the main steam line is 105 psi differential. This represents a flow of approximately 4.4x10s lb/hr. The high flow set point for the emergency cooling system supply line is 19 psi differential. This represents a flow of approximately 8.7xl0 at rated conditions. Normal background for the main steam line radiation monitors is defined as the radiation level which exists in the vicinity of main steam lines after 1 hour or more of sustained full rated power. The dose rate at the monitor due to activity from the control rod drop accident of Appendix E or from gross failure of one rod with complete fission product release from the rod would exceed the normal background at the monitor. The automatic initiation signals for the emergency cooling systems have to be sustained for more than 10 seconds to cause opening of the return valves. If the signals last for less than 10 seconds, the emergency cooling system operating will not be automatically initiated. The high level in the scram discharge volume is provided to assure that there is still sufficient free volume in the discharge system to receive the control rod drives discharge. Following a scram, bypassing is permitted to,allow draining of the discharge volume and resetting of the reactor protection system relays. Since all control rods are completely inserted following a scram and since the bypass of this particular scram initiates a control rod block, it is permissible to bypass this scram function. The scram trip associated with the shutdown position of the mode switch can be reset after 10 seconds. The condenser low vacuum, low-low vacuum and the main steam line isolation valve position signals are bypassed in the startup and refuel positions of the reactor mode switch when the reactor pressure is less than 600 psig. These are bypassed to allow warmup of the main steam lines and a heat sink during startup. Amendment No. 60 Nine Mile Point - Unit 1 B 3/4 6-4
BASES FOR 3. 6. 2 AND 4. 6. 2 PROTECTIVE INSTRUMENTATION (Continued) The set points on the generator load rejection and turbine stop valve closure scram trips are set to anti-cipate and minimize the consequences of turbine trip with failure of the turbine bypass system as described in the bases for Specification 2. 1.2. Since the severity of the transients is dependent on the reactor operating power level, bypassing of the scrams below the specified power level is permissible. The primary containment monitoring system is provided to alert the operator of conditions which could reduce safety margins during a postulated Loss-of-Coolant Accident. Appropriate operator corrective action is described in Specification 3. 3. 2, should Limiting Conditions for Operation be exceeded. This monitoring instrumentation does not automatically initiate engineered safeguards systems. Although the operator will set the setpoints at the values indicated in Tables 3.6.2.a-l, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. The de-viations include inherent instrument error, operator setting error and drift of the set point. These errors are compensated for in the transient analyses by conservatism in the controlling parameter assumptions as discussed in the bases for Specification 2.1.2. The deviations associated with the set points for the safety systems used to mitigate accidents have negligible effect on the initiation of these systems. These safety systems have initiation times which are orders of magnitude greater than the difference in time between reaching the nominal set point and the worst set point due to error. The maximum allowable set point devia-tions are listed below: Neutron Flux APRM, +2.7X of rated neutron flux IRM, +2.5X of rated neutron flux Recirculation Flow, +1% of rated recirculation flow Reactor Pressure, +15.8 psig Containment Pressure, +0.053 psig Reactor Water Level, +2.6 inches of water Main Steam Line Isolation Valve Position, +2.5X of stem position Scram Discharge Volume, +0 and -1 gallon Amendment No. N, 76 Nine Mile Point - Unit 1 8 3/4 6-5
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION (Continued) Condenser Low Vacuum, +0. 5 inches of mercury High Flow-Main Steam Line, +1 psid High Flow-Emergency Cooling Line, +1 psid High Area Temperature-Main Steam Line, +10F High Area Temperature-Clean-up and Shutdown, +6F High Radiation-Hain Steam Line, +100K and -50K of set point value High Radiation-Emergency Cooling System Vent, +100X and -50X of set point High Radiation-Reactor Building Vent, +100K and -50K of set point High Radiation-Refueling Platform, +100K and -50K of set point High Radiation-gffgaa Line, +50K of aet point, (Appendix 0)* Suppression Chamber Water Level, +1.8 inches The test intervals for the trip systems result in calculated failure probabilities <10-4 which correspond to the proposed IEEE Criteria for System Failure Probability. (IEEE SG-3, InformatTon Docket ¹1 - Protection System Reliability, April 24, 1968). The test intervals for the trip systems result in calculated failure probabilities ranging from 6.7 x 10-~ to 1. 76 x 10- (Fifth Supplement, p. 115). " The more frequent sensor checks result in even less probabi lity that the particular system will fail. Because of local high radiation, testing instrumentation in the area of the main steam line isolation valves can only be done during periods of Station shutdown. These functions include high area temperature isolation, high radiation isolation and isolation valve position scram. Testing of the scram associated with the shutdown position of the mode switch can be done only during periods of Station shutdown since it always involves a scram. "FSAR Amendment No. 28, 76 Nine Mile Point - Unit 1 B 3/4 6-6
BASES FOR 3. 6. 2 AND 4. 6. 2 PROTECTIVE INSTRUMENTATION (Continued)
- b. The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR is maintained greater than the SLCPR. The trip logic for this function is 1 out of n; e.g., any trip on one of the eight APRH's, eight IRH's or four SRM's will result in a rod block. The minimum instru-ment channel requirements provide sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the rod block may be reduced by one for a short period of time to allow maintenance, testing, or calibration. This time period is only ~3K of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawals The APRM rod block trip is flow biased and prevents a significant reduction in MCPR especially during operation at reduced flow. The APRH provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the SLCPR.
The APRH rod block also provides local protection of the core; i.e., the prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst case single control rod withdrawal error has been analyzed and the results show that with the specified trip settings rod withdrawal is blocked before the MCPR reaches the SLCPR, thus allowing adequate margin. Below ~60K power the worst case withdrawal of a single control rod results in not required. a MCPR > SLCPR without rod block action, thus below this level it is The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the SLCPR. A downscale indication on an APRH or IRM is an indication the instrument has failed or-the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented. The downscale rod blocks are set at 5 percent of full scale for IRH and 2 percent of full scale for APRM (APRM signal is generated by averaging the output signals from eight LPRH flux monitors). Nine Hi le Point Unit Amendment No. 9, 8L, 87, 41 1 B 3/4 6-7
0 BASES FOR 3. 6. 3 ANO 4. 6. 3 EMERGENCY POMER SOURCES Other than the Station turbine generator, the Station is supplied by four independent sources of a-c power; two 115 kv transmission lines, and two diesel-generators. Any one of the required power sources will provide the power required for .-the worst loss-of-coolant accident. The required loads of 2500 kva and 2750 kva for the loss-of-coolant are calculated in detail in the First Supplement to the FSAR. This loading is greater than that required during a Station shutdown condition. The monthly test run paralleled with the system is based on the manufacturer's recommendation for these units in this type of service. The testing during oper-ating cycle will simulate the accident conditions under which operation of the diesel-generators is required. A detailed tabulation of the equipment comprising the maximum diesel-generator load is given in the answer to question V-10 of the First Supplement to the FSAR. As mentioned above, a single diesel-generator is capable of providing the required power to equipment fol-lowing a major accident. Two fuel oil storage tanks are provided with piping interties to permit supplying either diesel-generator. A two-day supply will provide adequate time to arrange for fuel makeup The full capacity of both tanks will hold a four-day supply. if needed. It has been demonstrated -in Appendix E-I.3.21" that even with complete d-c loss the reactor can be safely isolated and the emergency cooling system will be operative with makeup water to the emergency cooling system shells maintained manually. Having at least one d-c battery available will permit: automatic makeup to the shells rather than manual, closing of the d-c actuated isolation valve on all lines, from the primary system and the suppression chamber, maintenance of electrical switching functions in the Station and providing emer-gency lighting and communications power. A battery system shall have a minimum of 106 volts at the battery terminals to be considered operable. "FSAR Nine Nile Point - Unit 1 B 3/4 6-8
0 BASES FOR 3.6.4 AND 4.6.4 SHOCK SUPPRESSORS (SNUBBERS) Snubbers are required to be operable to ensure that the structural integrity of the reactor coolant system and other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the number of observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25K) may not be used to lengthen the required inspection interval. Any inspec-tion whose results require a shorter inspection interval will override the previous schedule. Hydraulic or mechanical, accessible or inaccessible, snubber s may each be treated as a different entity for the above surveillance programs. Amendment No. 74 Nine Nile Point - Unit 1 B 3/4 6-9
BASES FOR 3.6.5 AND 4.6.5 RADIOACTIVE MATERIAL SOURCES The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage - from the source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. quantities of interest to this specification which are exempt from the leakage 'testing are consistent with the criteria of 10 CFR Parts 30. 11-20 and 70. 19. Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested. Amendment No. 11 Nine l1ile Point - Unit 1 B 3/4 6-10
BASES FOR 3.6.6 AND 4.6 ~ 6 FIRE DETECTION The basic function and capabilities of the system are to provide the means to detect fires with a visual indication of its location and an audible alarm at central points, and also to control certain ancillary actions, such as extinguishment and ventilation subsequent to the detection of fire. The system is comprised of seven (7) Local Fire Alarm Control Panels (LFACP) located throughout the Reactor Building and Turbine Building, primarily in a central location to the zones of fire detection for which each panel serves. In addition there is a Main Fire Alarm Control Panel (MFACP-2), in the Control Room to which all seven (7) panels report, and their indications and control functions are duplicated. Five types of detection instruments are employed in the system: a) Ionization Smoke b) Photoelectric Smoke c) Infrared d) Thermal e) Thermistor Mire The configuration of the fire detection instrument locations has been examined and found satisfactory to detect a fire with the minimum number of detectors operable as indicated in Table 3. 6. 6a. Amendment No. 2Z, 53 Nine Mile Point - Unit 1 8 3/4 6-11
BASES FOR 3.6.7 AND 4.6.7 FIRE SUPPRESSION The fire water supply is provided by two vertical turbine fire pumps, one electric and a diesel-driven unit which are design rated at 2500 gpm at 125 psig pump discharge head. These pumps are located in the screen house and take suction from the station cooling water intake tunnel and have relief valves set at 140 psig. The automatic initiation logic for each fire pump indicated in Specification 3. 6. 7. a. 2 requires that these pumps are automatically started together upon a drop in discharge header pressure. Each pump can also be manually started. In addition, the diesel fire engine will be started on low air pressure at alternate testing intervals to verify the adequacy of the low air pressure start system. A bypass of the starting air solenoid valves is provided for additional assurance in starting the diesel fire engine. The verification of the hydraulic performance of the fire suppression water system required once per 3 years in Surveillance Requirement 4. 6.7. a. 5 will be done by means of a measured hydrant flow test. The redundant components in the fire water supply system are the fire pumps, which discharge to the same header. They are the only components addressed in Specification 3.6.7.b. The backup water supply system referenced in Specification 3. 6.7. c. 1 is the Oswego City water system, which can be connected to the fire main if required. The water spray systems provide fire protection for the safety-related reserve transformers 101N and 101S. Supply for these systems is provided by the fire line. The systems employ open nozzles and are controlled by deluge valves. Valve actuation is by pneumatic type rate-of-rise devices installed over the protected equipment. In addition to the automatic operation, systems may be tripped manually either at the deluge valves on elevation 250'r at remote cable pull stations on elevation 261'. The fire control panel annunciator records system operation, low supervisory air pressure and valve closure. In addition to the spray systems described above, a closed head wet pipe automatic sprinkler system is provided for the diesel fire pump room in the Screen House on Elevation 254. The sprinkler heads used have fusible elements rated at 165 F. The system has flow alarms connected to the fire control panel annunciator. Fourteen pre-action type systems are used for various hazards throughout the plant. These systems employ closed heads, under an air pressure of 20 psig, and are controlled by a pre-action type valve. Amendment No. 22, 53 Nine Nile Point - Unit 1 B 3/4 6-12
BASES FOR 3.6.7 AND 4.6.7 FIRE SUPPRESSION (Continued) Valve actuation is automatic by ionization type detectors installed over the protected equipment. In addition to the automatic operation, systems may be tripped manually either at the pre-action valve or from the Hain Fire Panel in the Control Room. System operation, low supervisory air pressure and valve closure is monitored on both the Main Fire Control and Local Fire Panels. Amendment No. gg, 53 Nine Mile Point - Unit j. 8 3/4 6-13
BASES FOR 3.6.8 AND 4.6.8 CARBON DIOXIDE SUPPRESSION SYSTEM A low pressure carbon dioxide system is installed to serve seven different safety-related hazard points in the station indicated in Specification 3. 6. 8. a. Supply is provided by a 10 ton tank of liquid carbon dioxide located on elevation 261 feet. The self-contained refrigeration unit maintains the liquid at O'F with a resultant pressure of 300 psig. Carbon dioxide to the individual hazards is controlled by a series of carbon dioxide operated, pilot type master valves at the tank. Each of these valves serve a group of hazard valves of similar construction located at the individual areas. Fire extinguishment by carbon dioxide is either by total flooding or local application. In total flooding, sufficient CO< is injected into a closed room or space to inert the atmosphere and suppress combustion. Lo-cal application is employed for unenclosed areas and involves application of CO< on the equipment protected to extinguish the fire with additional discharge to permit cooling and inhibit reflash. The automatically actuated CO> systems employ either thermostats set at 225'F or smoke detectors to trip a timer located in the main cardox control cabinet. One or more sirens and a strobe light in the hazard area are initially operated for a pre-discharge period of 30 seconds to enable personnel to leave the area. The related master and hazard valves are then opened for a timed discharge period. Restoration of the CO~ hazard area to service is accomplished manually by pushbutton at the fire control panel. Manual pushbutton stations are also located at the individual areas to initiate the cycle. The control switch for each area on the fire control panel has three positions and is normally set for "Automatic" operation. An "Alarm only" position permits greater safety when men are working in the hazard area and the 30 second delay may be insufficient. A "Manual" position permits the operator to actuate the discharge cycle on his own initiative. An area pushbutton station will override the "Alarm only" setting on the Fire Control Panel. Due to the high rate of personnel access, and thus safety requirements, the Auxiliary Control Room CO> system is a manual system, used to back up a total flood automatic 6X Halon system. All CO> systems except hose reels are provided with odorizing devices as a safety measure. A glass flask of wintergreen concentrate is inserted in a capped tee beyond each hazard valve. This flask ruptures upon operation of the hazard and must be replaced after each use. In the event of total loss of D. C. control power to the CO~ system, all master valves wi ll open since their pilot valve solenoids are normally energized. The COz system hazard valves remain closed since their pilot valve solenoids are normally de-energized. CO> can be discharged into an area by operating the manual lever provided in each pilot valve cabinet. This is a manual operation within predischarge alarm or timer. Amendment No. g2, 53 Nine Mile Point - Unit 1 B 3/4 6-14
0 BASES FOR 3.6.8 AND 4.6.8 CARBON DIOXIDE SUPPRESSION SYSTEM (Continued) The flow test (" Puff Test" ) of the C02 system is performed by closing the CO< tank valve. This allows only the CO< vapor in the line to be discharged to the various designated areas in the plant. Carbon dioxide hose reels are provided at various points throughout the Turbine Building. These reels are provided with 150 feet of 1" high pressure hose with manual shutoff at the nozzle. Removal of the nozzle from its mounting bracket trips a switch which opens the master valve serving the hose reels. Carbon dioxide then flows to the nozzles of all hose reels. No odorant capsules are provided for hose reels. Certain hose stations are provided with timer operated bleeder valves to discharge vapor.and speed arrival of liquid C02 at the hose station. All system operations are monitored on the annunciator on the fire control panel. Amendment No. ZZ, 53 Nine Nile Point - Unit 1 B 3/4 6-15
BASES FOR 3.6.9 AND 4.6.9 FIRE HOSE STATIONS Standpipe risers at various locations in the turbine, reactor, waste and diesel buildings serve hose connec-tions. This equipment is located to permit hose stream coverage of safety-related equipment in the buildings. Each hose connection is equipped with 100 feet of 1 1/2 inch hose mounted on a reel. All hand line nozzles are of the adjustable spray type which can be varied down to 100 minimum spray pattern to render them safe for use on-electrical equipment. Eight foot long applicator spray nozzles and foam induction nozzles with five gallon cans of foam solution are also provided for use on hose lines as required. Amendment No. 22 Nine Nile Point - Unit 1 B 3/4 6-16
BASES FOR 3.6.10.1 AND 4.6.10.1 FIRE BARRIER PENETRATION FIRE SEALS Cable penetrations of the primary containment (drywell and pressure suppression chamber), reactor building, auxiliary control room and the cable room have been designed to provide adequate fire stop and to fire from spreading through the penetration. Drywell and pressure suppression chamber penetrations prevent are a double-sealed, 12-inch pipes that are inerted with nitrogen. Reactor building penetrations consist of standard conduit (pipe) sleeves, which vary in diameter from 3/4" to 4" and which are sealed at both ends. The auxiliary control room and the cable room have formed pipe sleeves and cable tray penetrations. These sleeves and penetrations are sealed at the ends with rock-wool filler and externally applied fire-resistant material for fire proofing. The local leak test required in Surveillance Requirement 4. 6. 10. 1. a. 2 will be performed by a non-hazardous method to ensure penetration integrity (an example of an acceptable local leak testing method is the "Downy Mand Test" or equivalent). Amendment No. gg, 53 Nine Mile Point - Unit 1 B 3/4 6-17
BASES FOR 3.6.10.2 AND 4.6.10.2 HALON SUPPRESSION SYSTEM The Halon 1301 fire protection systems are a gaseous fire suppressant system used in the Auxiliary Control and the Emergency Condenser Isolation Valve Rooms. The Halon 1301 fire protection system is comprised of a fire detection system status monitoring network and a fire suppression system. The fire detection system's status monitoring network monitors the areas covered by the Halon 1301 systems for fire conditions and system abnormalities. The fire suppression system consists of storage tanks of Halon 1301 and a delivery system to route Halon 1301 to the affected area in the event of a fire. Fire extinguishment by Halon 1301 is by total flooding. In total flooding, sufficient Halon is injected into the area to extinguish the fire. Both Halon systems are pr ovi'ded with odorizing devices as a safety measure. A glass flask of wintergreen concentrate is inserted in a capped tee in the main line piping. This flask ruptures upon operation of the system and must be replaced after each operation. A siren and strobe light in the protected area are initially operated for a pre-discharge period of 30 sec-onds to enable personnel to leave the area. Both systems may be manually tripped, either from the Main Fire Control Panel or at the storage banks. Amendment No. 53 Nine Mile Point - Unit 1 B 3/4 6-18
BASES FOR 3. 6. 11 AND 4. 6. 11 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" and/or NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980 and NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program." Amendment No. 72, 76 Nine Mile Point - Unit 1 B 3/4 6-19
BASES FOR 3.6;12 AND 4.6.12 REACTOR PROTECTION SYSTEM MOTOR GENERATOR SET MONITORING To eliminate the potential for undetectable single component failure which could adversely affect the operability of the reactor protection system, protection relaying schemes installed on MG sets 131, 141, 162, 172 and maintenance bus 130, provide for overvoltage, undervoltage and underfrequency protection. Amendment No. 62 Nine Mile Point - Unit 1 B 3/4 6-20
BASES FOR 3.6. 13 AND 4.6.13 REMOTE SHUTDOWN PANELS The remote shutdown panels provide 1) manual initiation of the emergency condensers 2) manual control of the steam supply valves and 3) parameters monitoring independent of the main/auxiliary control room. Two panels are provided, each located in a separate fire area, for added redundancy. Both panels are also in separate fire areas from the main/auxiliary control room. One. remote shutdown panel provides the necessary capabili-ties consistent with 10 CFR 50 Appendix R. Therefore, only one remote shutdown panel is required to be oper-able. The electrical design of the panels is such that no single fire can cause loss of both emergency condensers. Each remote shutdown panel is provided with controls for one emergency condenser loop. The emergency con-densers are designed such that automatic initiation is independently assured in the event of a fire 1) in the Reactor Building (principle relay logic located in the auxiliary control room or 2) in the main/auxiliary control room or Turbine Building (redundant relay logic located in the Reactor Building). Each remote shut-down panel also has controls to operate the two motor-operated steam supply valves on its respective emer-gency condenser loop. A key operated bypass switch is provided to override the automatic isolation signal to these valves. Once the bypass switch is activated, the steam supply valves can be manually controlled from the remote shutdown panels. Since automatic initiation of the emergency condenser is assured, the remote shutdown panels serve as additional manual controlling stations for the emergency condensers. In addition, certain parameters are monitored at each remote shutdown panel. The remote shutdown panels are normally de-energized, except for the monitoring instrumentation, which is normally energized. To energize the remaining functions on a remote shutdown panel, a power switch located on each panel must be activated. Once the panels are completely energized, the emergency condenser condensate return valve and steam supply valve controls can be utilized. Amendment No. 71 Nine Mile Point - Unit 1 8 3/4 6-21
BASES FOR RADIOACTIVE EFFLUENT INSTRUMENTATION 3.6.14 AND 4.6.14 The radioactive liquid and gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid and gaseous effluents during actual or potential releases of liquid and gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the Offsite Dose Calculation Manual to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes pro-visions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the main condenser offgas treatment 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 purpose of tank level indicating devices to assure the detection and control of leaks that result in the transport of radioactive materials to unrestricted areas. if not controlled could potentially Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-22
I BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS LI UID CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a member of the public and (2) the limits of 10 CFR Part 20. 106 (e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assump-tion that Xe-135 is the controlling radioisotope and its maximum permissible concentration in air. (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A., "Limits for Rualitative Retection and ()uantitative Determination - Applscation to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-23
BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS (Continued) Li uid Dose This specification is provided to implement the requirements of Section II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation expressed as quarter and annual limits are set at those values found in Section II.A. of Appendix I, in accordance with Section IV.A. The Limiting Conditions for Operation provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to unrestricted areas will be kept "as low as is reasonably'chievable." There are no drinking water supplies that can be potentially affected by plant operations. The dose calculation methodology and parameters in the Offsite Dose Calculation Manual implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculation procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially under-estimated. The equations specified in the Offsite Dose Calculation Manual for calculating the doses due to the actual release rates of radioactive mater ials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1. 109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-24
BASES FOR 3.6.15 AND 4.6.15-RADIOACTIVE EFFLUENTS (Continued) Gaseous Dose Rate This specification is provided to ensure that the dose at any time at and beyond the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a member of the public in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits spe-cified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20. 106(b)). For members of the public who may at times be within the site boundary, the occupancy of that member of the public will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above back-ground to a member of the public at or beyond the site boundary to less than or equal to 500 mrems/year to the total body or to less than or equal to 3000 mrems/year to the skin. These release rate limits also re-strict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less- than or equal to 1500 mrems/year. The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A., "Limits for gualitative Detection and Duantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), T Company Report ARH-SA-215 (June 1975). Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-25
BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS (Continued) Dose - Noble Gases This specification is provided to implement the requirements of Sections II. B, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation expressed as quarter and annual limits are set at those values found in Section -II.B of Appendix I in accordance with the guidance of Section IV.A. The action statements provide the required operating flexibility and at the same .time implement the guides set forth in Section IV-A of Appendix I to assure that the releases of radioactive material in gaseous effluents to un-restricted areas will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I be shown by calcu-lational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the Offsite Dose Calculation Manual for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1. 109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Mater Cooled Reactors," Revision 1, July 1977. The Offsite Dose Calculation Manual equations provided to determine the air doses at and beyond the site boundary are based upon the historical average atmospheric conditions. Amendment No. 66 Nine'i le Point - Unit 1 B 3/4 6-26
BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS (Continued) Dose - Iodine-131 Iodine-133 .Tritium and Radionuclides in Particulate Form This specification is provided to implement the requirements of Sections II. C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation expressed as quarter and annual limits are set at those values found in Section II. C of Appendix I in accordance with the guidance of Section IV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to un-restricted areas will be kept "as low as is reasonably achievable." The Offsite Dose Calculation Manual cal-culational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The Offsite Dose Calculation Manual calculational methodology and parameters-for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1. 109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, Octo-ber 1977 and Regulatory Guide 1. 111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the site boundary. The pathways that were examined in the development of these calculations were: 1) indi-vidual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man and 4) deposition on the ground with subsequent exposure of man. Main Condenser Restricting the gross radioactivity rate of noble gases from the main condenser provides assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a very small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environ-ment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. The primary purpose of providing this specification is to limit buildup of fission product activity within the station systems which would result if high fuel leakage were to be per-mitted over extended periods. Amendment No. 66 Nine Mile Point - Unit 1 8 3/4 6-27
BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS (Continued) Total Dose - Uranium Fuel C cle This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46FR 182525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant. generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the reactor units and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a member of the public to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to a member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contribution from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190. 11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specification 3 6.15.a. (1) and
~
3.6.15.b(1). An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. Amendment No. 66 Nine Nile Point - Unit 1 B 3/4 6-28
BASES FOR 3.6.16 AND 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Li uid Radwaste Treatment S stem The requirement that the appropriate portions of this system be used provides assurance that the releases of radioactive materials in liquid=-effluents will be kept "as low as is reasonably achievable." This specifica-tion implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. Gaseous Radwaste Treatment S stem The requirement that this system be used provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50. 36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. Since the capability exists to operate within specification without use of the sy'tem, it is conceivable that due to unforeseen circumstances, limited operation without the system may be made sometime during the life of the pl ant. Solid Radioactive Waste This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the process control program may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents and mixing and curing times. Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-29
BASES FOR 3. 6. 17 AND 4. 6. 17 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures con-tained in the waste gas treatment system is maintained below the flammabi lity limits of hydrogen. Automatic control features are included in the system to prevent the hydrogen concentration from reaching these flam-mability limits. Maintaining the concentration of hydrogen below flammability limits provides assurance. that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-30
BASES FOR 3,6.18 AND 4.6.18 MARK I CONTAINMENT This specification provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas. Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6-31
BASES FOR 3.6.19 AND 4.6.19 LIQUID HOLDUP TANKS This specification applies to any outdoor tank that is not surrounded by liners, dikes or walls capable of holding the tank contents and that does not have tank overflows and surrounding areas drains connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks'ontents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area. Amendment No."66 Nine Mile Point - Unit 1 8 3/4 6-32
BASES FOR 3.6.20 AND 4.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PLAN The radiological environmental monitoring program required-by this specification provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the station operation. This monitoring program implements Section IV. B. 2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental'onitoring. The ini-tially specified monitoring program will be effective for at least the first three years of commercial opera-tion. Following this period, program changes may be initiated based on operational experience. 'The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4. 6.20-1 are considered optimum for routine environ-mental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement. Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A., "Limits for qualitative Detection and quantitative Determination-Application to Radiochemistry," Anal. Chem 40, 586-93 (1968) and Hartwell, J. K., "Detection Limits for ( Amendment No. 66 Nine Mile Point - Unit 1 B 3/4 6"33
BASES FOR 3.6.21 AND 4.6.21 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring for the purposes of Section IV. B. 2 of Appendix I to 10 CFR Part 50. Amendment No. 66 Nine Mile Point Unit 1- B 3/4 6-34
BASES FOR 3.6!QO AND 4.6@2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best survey information such as from a door-to-door survey(s), from an aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV. B. 3 of Appendix I to 10 CFR Part 50. In lieu of a garden census, the significance of the exposure via the garden pathway can be evaluated by the sampling of vegetation as specified in Table 3.6. 20-1. A milk sampling location, as defined in Section 1, requires that at least 10 milking cows are present at a designated milk sample location.. It has been found from past experience, and as a result of conferring with local farmers, that a minimum of 10 milking cows is necessary to guarantee an adequate supply of milk twice per month for analytical purposes. Locations with less than 10 milking cows are usually utilized for breeding purposes which eliminates a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry. Amendment No. 66 Nine Nile Point - Unit 1 B 3/4 6-35
Document Name: NMP-1 TS SEC 5 Requestor's ID: NORMA II Author's Name: Jamerson C. Document Comments: ETPB REV 9/22/86 KEEP THIS SHEET WITH DOCUMENT
- 5. 0 DESIGN FEATURES 5.1 Site The Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant site comprising approximately 1500 acres, is located on the shores of Lake Ontario, about seven miles northeast of Oswego, New York. An exclusion distance of nearly 4000 feet is provided between the Station and the nearest site boundary to the west, a mile to the boundary on the east, and a mile and a half
'to the southern site boundary (as described in the Sixth Supplement of the FSAR).
Figure 5.1-1 is a Site Boundary Map of Nine Mile Point which allows the identification of gaseous and liquid waste release points. Figure 5.1-1 also defines the unrestricted area within the site boundary that is accessible (except for fenced areas) to members of the public. 5.2 Reactor The reactor core consists of no more than 532 fuel assemblies containing enriched uranium dioxide pellets clad in Zircaloy-2. The core excess reactivity will be controlled by movable control rods and burnable poisons. The core will be cooled by circulation of water internally and external to the pressure vessel through recirculation loops. 5.3, Reactor Vessel The pertinent features of the reactor vessel other than those referred to in the technical specifications are as follows: Internal Height 63'-10" Internal Diameter 17'-9" Vessel Design Lifetime 40 years Materials of Construction Base Metal SA3028 Clad Weld Deposited 308L Electrode Amendment No. 66 Nine Mile Point - Unit 1 5-1
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SITE BOUNDAR IES 4 NIAGARA MOHAWK POWER CORPORATION NINE MlLE POINT-UNIT I E
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NOTES TO FIGURE 5.1-1 (a) NMP1 Stack (height is 350') (b) NHP2 Stack (height is 430') (c) JAFNPP Stack (height is 385') (d) NHPl Radioactive Liquid Discharge (Lake Ontario, bottom) (e) NHP2 Radioactive Liquid Discharge (Lake Ontario, bottom) (f) JAFNPP Radioactive Liquid Discharge (Lake Ontario, bottom) (g) Site Boundary (h) Lake Ontario Shoreline Additional Information: NHP2 Reactor Building Vent is located 187 feet above ground level JAFNPP Reactor and Turbine Building Vents are located 173 feet above ground level JAFNPP Radwaste Building Vent is 112 feet above ground level Amendment No. 66 Nine Mile Point - Unit 1 5-3
5.4 Containment The containment system consists of a drywell, suppression chamber and a reactor building. The pressure suppression system consists of a drywell with a volume of approximately 243,000 cubic feet and an interconnected suppression chamber with a volume of 209,000 cubic feet. Of this total volume some 180,000 and 120,000 cubic feet of free space are available in the drywell and suppression chamber, respectively. The pertinent design features not discussed elsewhere in the technical specifications are as follows: Pressure Internal Design Pressure 62 psig 35 psig Internal Design Temperature 310 ~F 205 OF External Design Pressure 2 pslg 1 pslg Material of Construction A-201 and A-212 Grade "B" Firebox Steel made to A-300 requirements. For long-term post-accident recovery, the pressure suppression system is designed to permit flooding to a level at least six feet above the core. The reactor building is designed for a maximum in-leakage rate of 100 percent per day at 0.25 inch of water internal vacuum and zero wind speed. Exterior loadings for wind, snow and ice meet all applicable codes. The roof and supporting structures are designed to withstand a loading of 40 psf of snow or ice. The'alls and building structure are designed to withstand an external or internal loading of 40 psf which is approximately equivalent to that caused by a wind velocity of 125 mph 30 feet above the ground level. Pressure relief is provided to prevent damage to the superstructure due to the break of any primary system line in the reactor building. In this event, blowout panels will fail, relieving pressure in the event of a major line rupture. Nine Mile Point Unit 1
Ill q 5.5 Stora e of Unirradiated and S ent Fuel Unirradiated fuel assemblies will normally be stored in critically safe new fuel storage racks in the reactor building storage vault. Even when flooded with water, the resultant k Fresh fuel may also be stored in shipping containers. f is less than 0. 95. The unirradiated fuel storage vault is designed and shall be maintained with a storage capacity limited to no more than 200 fuel assemblies. The spent fuel storage facility is designed to maintain fuel in a geometry such that k is less than 0.95 under conditions of optimum water moderation. The spent fuel storage facility is Msigned and shall be maintained with a storage capacity limited to no more than 2776 fuel assemblies. Fuel assemblies stored in the 1066 spent fuel storage locations of the non-poison flux trap design are limited to 15.6 grams (3.0 weight percent) of Uranium-235 per axial centimeters of assembly. Fuel assemblies stored in the 1,710 spent fuel storage positions of the poison type which use Boraflex as the neutron absorber are limited to 18. 13 grams (3.75 weight percent) of Uranium-235 per axial centimeters of assembly. Calculations for k values have been based on methods approved by the Nuclear Regulatory Commission covering special always (10 CFR 70. 55). The reactor building and all contained engineered safeguards are designed for the maximum credible earth-quake ground motion with an acceleration of 11 percent of gravity. Dynamic analysis was used to deter-mine the earthquake acceleration, applicable to the various elevations in the reactor building. Amendment No. 54 Nine Nile Point - Unit 1 5-5
Document Name: NMP-1 TS SEC 6 Requestors ID: NORMA Author's Name: Jamerson C. Document Comments: ETPB Rev. 9/22/86 KEEP THIS SHEET WITH DOCUMENT
- 6. 0 ADMINISTRATIVE CONTROLS 6.1.1 The General Superintendent Nuclear Generation shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.
6.1.2 The Station Shift Supervisor - Nuclear (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A manage-ment directive to this effect, signed by the Vice President - Nuclear Generation shall be re-issued to station personnel on an annual basis.
.2 Offeite 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.-
6.2.2 The facility organization shall be as shown on Figure 6.2-2 and:
- a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
- b. At least one licensed Operator shall be in the control room when fuel is in the reactor. During reactor operation, this licensed operator shall be present at the controls of the facility.
- c. At least two licensed Operators shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips.
- d. An individual qualified in radiation protection* procedures shall be on site when fuel is in the reactor.
"The Radiation Protection qualified individual and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions. Amendment No. 65 Nine Mile Point - Unit 1
- e. A licensed Senior Reactor Operator shall be required in the Control Room during power operations, hot shutdown, and when the emergency plan is activated. This may be the Station Shift Supervisor - Nuclear or the Assistant Station Shift Supervisor - Nuclear during power operations. Mhen the emergency plan is activated, the Assistant Station Shift Supervisor Nuclear becomes the Shift Technical Advisor and the Station Shift Supervisor - Nuclear is restricted to the control room until an additional licensed Senior Reactor Operator arrives.
- f. -
A licensed Senior Reactor Operator shall be responsible for all movement of new and ir-radiated fuel within the site boundary. All core alterations shall be directly super-vised by a licensed senior reactor operator who has no other concurrent responsibilities during this operation. A Licensed Operator will be required to manipulate the controls of all fuel handling equipment except movement of new fuel from receipt through dry storage. All fuel moves within the core shall be directly monitored by a member of the reactor- analyst group.
- g. A Fire Brigade of five (5) members~ shall be maintained on site as defined by 5.1 at all times.
- h. Administrative procedures shall be developed and implemented to limit the working hours of facility staff who perform safety-related functions; e.g., licensed Senior Operators, licensed Operators, health physicists, auxiliary operators and key maintenance personnel.
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the facility is operating. However, in the event that unforeseen. problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications on a temporary basis, the fol-lowing guidelines shall be followed: "The Radiation Protection qualified individual and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence, pro-vided immediate action is taken to fill the required positions. Amendment No. HS, 80 Nine Nile Point - Unit 1 6-2
- 1) An individual should not be permitted to work more than 16 hours straight (excluding shift turnover time).
- 2) An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period (all excluding shift turnover time).
- 3) A break of at least 8 hours should be allowed between work periods (including shift turnover time).
- 4) Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the General Superintendent-Nuclear Generation or designee, or higher levels of management, in accordance with estab-lished procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Station Superintendent - Nuclear Generation or designee to assure that excessive hours
'have not been assigned. Routine deviation from the above guidelines is not authorized.
Amendment No. 65 Nine Mile Point - Unit 1 6-3
FIGURE 6.2 I NINE MILE POINT NUCLEAR STATION MANAGEMENT ORGANIZATION CHART PRESIDENT Vice President Quality Assurance Senior Vice President Manager Vice President Nuclear Engineering Nuclear Generation 8c Licensing General Superintendent Nuclear Generation Station Superintendent Nuclear Generation Unit 1 Amendment Ho. 58,77, 89
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Amendment No. 65 Nine Mile Point - Unit 1
TABLE 6.2"1 MINIMUM SHIFT CREW COMPOSITION (1) (6) Operation (3) Reactor License Normal Operation Shutdown Condition W/0 Process Computer Startups Senior Operator 1 (5) Operator Unlicensed (2) Asst. Station Shift Supevisor (Shift Technical Advisor Function) 1 (4) (Senior Operator License) (7) Notes: (1) At any one time, more licensed or unlicensed operating people could be present for maintenance, repairs, refuel outages, etc. (2) Those operating personnel not holding an "Operator" or "Senior Operator" License. (3) For operation longer. than eight hours without process computer. (4) Hot shutdown condition only. (5) An additional senior reactor operator who has no other concurrent responsibilities shall supervise all core alterations. (6) The Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed two hours 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 require-ments of Table 6. 2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. (7) The Assistant Station Shift Supervisor performs the Shift Technical Advisor function and shall hold a senior reactor operator. license. Amendment No. 65 Nine Mile Point - Unit 1
- 6. 3 Facilit Staff ualifications
- 6. 3. 1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.
- 6. 4 ~Trainin 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Superintendent - Training Nuclear and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A training program for the Fire Brigade 'shall be maintained under the direction of the Superintendent - Training Nuclear and Supervisor - Fire Protection, Nuclear and shall meet or exceed the requirements of Appendix R to 10 CFR 50. 6.5 Review and Audit
- 6. 5. 1 Site 0 erations Review Committee (SORC Function 6.5.1.1 The Site Operations Review Committee shall function to advise the General Superintendent - Nuclear Generation on all matters related to nuclear safety.
Com osition 6.5. 1. 2 The Site Operations Review Committee shall be composed of the: Chairman: General Superintendent - Nuclear Generation Member: Station Superintendent - Nuclear Generation Member: Technical Superintendent - Nuclear Generation Member: Superintendent Technical Services - Nuclear Member: Site Superintendent Maintenance - Nuclear Member: Supervisor Instrument and Control - Nuclear Member: Superintendent Chemistry and Radiation Management Amendment No. 65 Nine Mile Point - Unit 1 6-7
Alternates 6.5. 1.3 Alternate members shall be appointed in writing by the SORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate in SORC activities at any one time. Meetin Fre uenc 6.5.1.4 The SORC shall meet at least once per calendar month and as convened by the SORC Chairman. uorum 6.5.1.5 A quorum of the SORC shall consist of the Chairman and four members including alternates. Res onsibilities 6.5.1.6 The Site Operations Review Committee shall be responsible for:
- a. Review of all REPORTABLE EVENTS.
- b. Review of facility operations to detect potential safety hazards.
C. Performance of special reviews investigations or analyses and reports thereon as requested by the Chairman of the Safety Review and Audit Board.
- d. Investigation of violations of the Technical Specifications and shall prepare and forward a report covering evaluation and recommendations to prevent recurrence to the Vice President - Nuclear Generation and to the Chairman of the Safety Review and Audit Board.
Amendment No. 65 Nine Mile Point - Unit 1
~Authorit 6.5.1.7 The Site Operations Review Committee shall:
- a. Render determinations in writing with regard to whether or not each item con-sidered under 6. 5.1. 6(a) through (d) above constitutes an unreviewed safety question.
- b. Provide immediate written notification to the Vice President - Nuclear Genera-tion and Chairman of the Safety Review and Audit Board of disagreement between the SORC and'the General Superintendent - Nuclear Generation; however, the General Superintendent - Nuclear Generation shall have the responsibility for resolution of such disagreements pursuant to 6. 1. 1 above.
Records 6.5.1.8 The Site Operations Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Vice President - Nuclear Generation and Chair-man of the Safety Review and Audit Board.
- 6. 5. 2 Technical Review and Control Activities 6.5.2.1 Each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, shall be prepared by a qualified individual/organization. Each such procedure, and changes thereto, shall be re-viewed by an individual/group other than the individual/group which prepared the procedure, or changes thereto, but who may be from the same organization as the individual/group which prepared the procedure, or changes thereto. Approval of procedures and programs and changes thereto and their safety evaluations, shall be controlled by administrative procedures.
6.5.2.2 Proposed changes to the Technical Specifications shall be prepared by a qualified individual/organization. The preparation of each proposed Technical Specifications change shall be reviewed by an individual/group other than the individual/group which prepared the proposed change, but who may be from the same organization as the individual/group which prepared the proposed change. Proposed changes to the Technical Specifications shall be approved by the General Superintendent - Nuclear Generation. Amendment No. 65 Nine Mile Point - Unit 1 6-9
Activities (Cont'd) 6.5.2.3 Proposed modifications to unit structures, systems 'and components that affect nuclear safety shall be designed by a qualified individiual/organization. Each such modi-fication shall be reviewed by an individual/group other than the individual/group which designed the modification, but who may be from the same organization as the individual/group which designed the modification. Proposed modifications to struc-tures, systems and components and the safety evaluations shall be approved prior to implementation by the General Superintendent - Nuclear Generation; or the Station Superintendent -,Nuclear Generation; or the Technical Superintendent Nuclear Generation as previously designated by the General Superintendent - Nuclear Generation. 6.5.2.4 Individuals responsible for reviews performed in accordance with Specifica-tions 6.5.2.1, 6.5.2.2 and 6.5.2.3 shall be members of the station supervisory staff, previously designated by the General Superintendent - Nuclear Generation to perform such reviews. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary such review shall be performed by the appropriate designated station review personnel. 6.5.2.5 Proposed tests and experiments which affect station nuclear safety and are not ad-dressed in the FSAR or Technical Specifications and their safety evaluations shall be reviewed by the General Superintendent - Nuclear Generation; or by the Station Superintendent - Nuclear Generation, or the Technical Superintendent Nuclear Generation as previously designated by the General Superintendent - Nuclear Generation. 6.5.2.6 The General Superintendent - Nuclear Generation shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President - Nuclear Generation. 6.5.2.7 The facility security program, and implementing procedures, shall be reviewed at least every 12 months. Recommended changes shall be approved by the General Superintendent - Nuclear Generation and transmitted to the Vice President Nuclear Generation, and to the Chairman of the Safety Review and Audit Board. 6.5.2.8 The facility emergency plan, and implementing procedures shall be reviewed at least every 12 months. Recommended changes shall be approved by the General Superinten-dent Nuclear Generation and transmitted to the Vice President - Nuclear Generation and to the Chairman of the Safety Review and Audit Board. Amendment No. 65 Nine Nile Point - Onit 1 6-10
Activities (Cont'd) 6.5.2.9 The General Superintendent - Nuclear Generation shall assure the performance of a review by a qualified individual/organization of changes to the Radiological Waste Treatment systems. 6.5. 2. 10 Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President - Nuclear Generation and to the Safety Review and Audit Board.
- 6. 5. 2. 11 Review of changes to the Process Control Program and the Offsite Oose Calculation Manual. Approval of any changes shall be made by the General Superintendent - Nuclear Generation or his designee before implementation of such changes.
6.5.2. 12 Reports documenting each of the activities performed under Specifications 6. 5. 2. 1 through 6.5.2.9 shall be maintained. Copies shall be provided to the Vice Presi-dent - Nuclear Generation and the Safety Review and Audit Board. 6.5.3 Safet Review and Audit Board SRAB Function 6.5.3.1 The Safety Review and Audit Board shall function to provide independent review and audi,t of designated activities in the areas of:
- a. nuclear power plant operations'.
nuclear engineering
- c. chemistry and radiochemistry
- d. metallurgy
- e. instrumentation and control
- f. radiological safety g., mechanical and electrical engineering
- h. quality assurance practices
- i. (other appropriate fields associated with the unique characteristics of the nuclear power plant)
Amendment No. N, 66 Nine Mile Point - Unit 1 6-11
Com osition
- 6. 5.3. 2 The Safety Review and Audit Board shall be composed of- the:
Chairman: Staff Engineer or Manager or Vice President Member: General Superintendent - Nuclear Generation Member: Staff Engineer - Nuclear Member: Staff Engineer - Mechanical or Electrical Member: Staff Engineer - Environmental Member: Consultant (See 6.5.3.4) Alternates 6.5.3.3 Alternate members shall be appointed in writing by the SRAB Chairman to serve on a temporary basis; however, no more than two alternates shall participate in SRAB activities at any one time. Consultants 6.5.3.4 Consultants shall be utilized as determined by the SRAB Chairman to provide expert advice to the SRAB. Meetin Fre uenc 6.5.3.5 The SRAB shall meet at least once per six months. uorum 6.5.3 ' A quorum of SRAB shall consist of the Chairman or his designated alternate and a majority'f SRAB members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility. Amendment No. 65 Nine Mile Point - Unit 1 6-12
Review
- 6. 5. 3. 7 The SRAB shall review:
- a. The safety evaluations for 1) changes to procedures, equipment or systems and
- 2) tests or experiments completed under the provision of Section 50. 59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
- b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- d. Proposed changes in Technical Specifications or operating license.
- e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
- f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
- g. All REPORTABLE EVENTS.
- h. Any indication of an unanticipated deficiency in some aspect of design or opera-tion of safety related structures, systems, or components.
- i. Reports and meeting minutes of the Site Operations Review Committee.
Amendment No. HS, 66 Nine Nile Poirit - Unit 1 6-13
Audits 6.5.3.8 Audits of facility activities shall be performed under the cognizance of the SRAB. These audits shall encompass: The conformance of facility operation to all provisions contained within the Technic'al Specifications- and applicable license conditions at least once per year.
- b. The performance, training and qualifications of the entire facility staff at I
least once per year. C. The results of actions taken to correct deficiencies occurring in facility equip-ment, structures, systems, or method of operation that affect nuclear safety at least once per six months.
- d. The performance of all activities required by the equality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per two years.
- e. The Facility Emergency Plan and implementing procedures at least once every 12 months.
The Facility Security Plan and implementing procedures at least once every 12 months. The Facility Fire Protection Program and implementing procedures at least once per two years. Any other area of facility operation considered appropriate by the SRAB, the Vice President - Nuclear Generation or the Vice President - Nuclear Engineering and Licensing. The radiological environmental monitoring program and,the results thereof at least once per 12 months. The Offsite Oose Calculation Manual and implementing procedures at least once per 24 months. The Process Control Program and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months. Amendment No. 85, 88, 69 Nine Mile Point - Unit 1 6-14
A~uthorit 6.5.3.9 The SRAB shall report to and advise the Vice President - Nuclear Generation and Hanager Nuclear Engineering 5 Licensing on those areas of responsibility specified in Section 6.5.3.7 and 6.5.3.8. Records 6.5.3. 10 Records of SRAB activities shall be prepared, approved and distributed as indicated below:
- a. Hinutes of each SRAB meeting shall be prepared, approved and forwarded to the Vice President Nuclear Generation and Hanager Nuclear Engineering 5 Licensing within 30 days following each meeting.
- b. Reports of reviews encompassed by Section 6.5.3.7 e,f,g and h above, shall be prepared, approved and forwarded to the Vice President Nuclear Generation and Manager - Nuclear Engineering 5 Licensing within 14 days following completion of the review.
- c. Audit reports encompassed by Section 6.5.3.8 above, shall be forwarded to the Vice President Nuclear Generation and Hanager Nuclear Engineering 5 Licensing within 90 days following completion of the review.
Amendment No. g5,86,89
- 6. 6 Re ortable Event Action 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
- a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.72 and 50.73 to 10 CFR Part 50, and
- b. Each REPORTABLE EVENT shall be reviewed by the SORC and the results of this review sub-mitted to the SRAB and the Vice President - Nuclear Generation.
6.7 Safet Limit Violation 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. The provisions of 10 CFR 50.36(c)(l)(i) shall be complied with immediately.
- b. The Safety Limit violation shall be reported to the Commission, the Vice President-Nuclear Generation and to the SRAB immediately.
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the Commission, the SRAB and the Vice President - Nuclear Generation within 10 days of the violation.
6.8 Procedures 6.8.1 Mritten procedures and administrative policies shall be established, implemented and main-tained that meet or exceed the requirements and recommendations of Sections 5. 1 and 5. 3 of ANSI N18.7-1972 and Appendix "A" of USAEC Regulatory Guide 1.33 except as provided in 6.8.2 and 6.8.3 below. 6.8. 2 Each procedure and administrative policy of 6. 8. 1 above, and changes thereto, shall be re-viewed and approved by the General Superintendent - Nuclear Generation or designee prior to implementation and periodically as set forth in each document. Amendment No. 66 Nine Nile Point Unit 1 6-16
6.8 Procedures (Cont'd) 6.8.3 Temporary changes to procedures of 6.8. 1 above may be made provided:
- a. The intent of the original procedure is not altered.
- b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
- c. The change is documented, reviewed and approved by the General Superintendent - Nuclear Generation or designee within 14 days of implementation.
6.9 Re ortin Re uirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of Inspection and Enforcement Re-gional Office I, King of Prussia, Pennsylvania 19406, unless otherwise noted. 6.9. 1 Routine Reports be submitted following (1) receipt of an operating license,'2) amendment to the license involving a planned increase power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these'values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this repor t. Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed. Amendment No. 66 Nine Nile Point - Unit 1 6-17
- b. Annual Occu ational Ex osure Re ort. A tabulation shall be submitted on an annual basis which includes the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr arid their associated man rem exposure ac-cording to work and job functions, 1/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20K of th'e individual total dose need not be accounted for. In the aggregate, at least 80X of the total whole body dose received from external sources shall be assigned to specific major work functions. C. Monthl 0 eratin Re ort. Routine reports of operating statistics and shutdown exper-ience including documentation of challenges to the safety relief valves or safety valves, shall be submitted on a monthly basis, which will include a narrative of operating experience,, to the Director, Office of Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office of I8E, no later than the 15th of each month following the calendar month covered by the report.
- d. Annual Radiolo ical Environmental 0 eratin Re ort". Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1, 1985.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological envi-ronmental surveillance activities for the report period, including a comparison with operational controls as appropriate, and with environmental surveillance reports from the previous 5 years, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3. 6. 22. 1/This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
"A single submittal may be made for a multiple unit station.
Corrected May 2, 1985 Amendment No. HS, 66 Nine Mile Point - Unit 1 6-18
The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the Table and Figures in the Offsite Oose Calculation Manual, 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 shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps"" covering all sampling loca-tions keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.6.21; discussion of all deviations from the sampling schedule of Table 3.6.20-1; and discussion of all analyses in which the LLO required in Table 4.6.20-1 was not achievable.
- e. Semiannual Radioactive Effluent Release Re ort. Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin on January 1, 1985.
The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as out-lined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. ""One map shall cover stations near the site boundary; a second shall include the more distant stations. Amendment No. 66 Nine Mile Point - Unit 1 6-19
The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and pre-cipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the Offsite Dose Calculation Manual. The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protec-tion Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in the Offsite Dose Calculation Manual. The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:.
- a. Container volume,
- b. Total curie quantity (specify whether determined by measurement or estimate),
- c. Principal radionuclides (specify whether determ'ined by measurement or estimate),
- d. Source of waste and processing employed (e. g., dewatered spent resin, compacted dry waste, evaporator bottoms),
- e. Type of container (e.g., LSA, Type A, Type 8, Large guantity), and,
- f. Solidification agent or absorbent (e. g., cement)
Amendment No. 66 Nine Mile Point - Unit 1 6-20
The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP) and to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.6.20. Changes to the Process Control Program (PCP) shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change with-out benefit of additional or supplemental information;
- b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation of the fact that the change has been reviewed and found acceptable.
Changes to the Offsite Dose Calculation Manual (ODCM): Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change with-out benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the Offsite Dose Calculation Manual to be changed, together with appropriate analyses or evaluations justifying the change(s);
- b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable.
Amendment No. 66 Nine Mile Point - Unit 1 6-21
0 6.9.2 Fire Protection Pro ram Re orts
- a. Submit a special report to the appropriate Regional Office as follows:
Notify the Director of the appropriate Regional Office by telephone within 24 hours. Confirm by telegraph, mailgram or facsimile transmission no later than the first working day following the event, and Follow up in writing within 14 days after the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to an operable status.
- b. Submit a special report to the Director of the appropriate Regional Office within 30 days following the event outlining the plans and procedures to be used to restore the inoperable equipment to an operable status.
Special reports shall be submitted to the Director of Regulatory Operations Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
- a. Reactor Vessel Material Surveillance Specimen Examination, Specification 4. 2. 2(c)
(12 months)
- b. Safety Class 1 Inservice Inspections, Specification (See Table 4. 2. 6(a)) (Three months)
- c. Safety Class 2 Inservice Inspections, Specification (See Table 4.2.6(b)) (Three months)
- d. Safety Class 3 Inservice Inspections, Specification (See Table 4. 2. 6(c)) (Three months)
- e. Primary Containment Leakage Testing, Specification 3.3.3 (Three months)
- f. Secondary Containment Leakage Testing, Specification 3.4. 1 (Three months)
Amendment No. 66 Nine Mile Point - Unit 1 6-22
. 9. 3
- g. Sealed Source Leakage in Excess of Limits, Specification 3~ 6.5.2 (Three months)
- h. Calculate Dose from Liquid Effluent in Excess of Limits, Specification 3. 6. 15. a(2)(b)
(30 days from the end of the affected calendar quarter).
- i. Calculate Air Dose from Noble Gases Effluent in Excess of Limits, Specification 3.6.15.b(2)(b), (30 days from the end of the affected calendar quarter).
- j. Calculate Dose from I-131, I-133, H-3 and Radioactive Particulates with half lives greater than eight days in Excess of Limits, Specification 3.6 '5.b(3)(b) (30 days from the end of the affected calendar quarter).
- k. Calculated Doses from Uranium Fuel Cycle Source in Excess of Limits Specification 3.6.15.d (30 days from the end of the affected calendar year).
- l. Inoperable Gaseous Radwaste Treatment System, Specification 3. 6. 16. b (30 days from the event).
- m. Environmental Radiolo ical Re orts. With the level of radioactivity (as the result of p ant e uents , sn an environmental sampling medium exceeding the reporting level of Table 6.9.3-1, when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within thirty (30) days from the end of the calendar quarter a special report identifying the cause(s) for exceeding the limits, and define the corrective action to be taken.
Amendment No. 66 Nine Mile Point - Unit 1 6-23
TABLE 6.9.3-1 REPORTING LEVEL FOR RAOIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS Water Airborne Particulate Fish Milk Food Products ~Anal aia ~Ci/1 or Gases ( Ci/m3) ( Ci/k wet ~(Ci/1) ( Ci/k wet) H-3 30,000 Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr/Nb-95 400 I-131 0.9 100 Cs-134 30 10. 0 1,000 60 1,000 Cs-137 50 20. 0 2,000 70 2,000 Ba/La-140 200 300 Amendment No. 66 Nine Mile Point - Unit 1 6-24
6.10 Record Retention 6.10.1 The following records shall be retained for at least five years: m
- a. Records and logs of facility operation covering time interval at each power level.
- b. Records and logs of principal maintenance activities, inspections, repair and replace-ment of principal items of equipment related to nuclear safety.
- c. REPORTABLE EVENT REPORTS.
- d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
- e. Records of reactor tests and experiments.
- f. Records of changes made to Operating Procedures.
- g. Records of radioactive shipments.
- h. Records of sealed source leak tests and results.
- i. Records of annual physical inventory of all source material of record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License:
- a. Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.
- b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
- c. Records of facility radiation and contamination surveys.
- d. Records of radiation exposure for all individuals entering radiation control areas.
Amendment No. 8$ , 66 Nine Mile Point - Unit 1 6-25
- 6. 10 Record Retention (Cont'd)
- e. Records of gaseous and liquid radioactive material released to the environs.
- f. Records of transient, or operational cycles for those facility components designed for a limited number of transients or cycles.
- g. Records of training and qualification for current members of the plant staff.
- h. Records of in-service inspections performed pursuant to these Technical Specifications.
- i. Records of Quality Assurance activities required by the QA Manual.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50. 59.
- k. Records of meetings of the SORC and the SRAB.
- 1. Records of analyses required by the radiological environmental monitoring program that .
would permit evaluation of the accuracy of the analysis at a later date. This should in-clude procedures effective at specified times and Quality Assurance records showing that these procedures were followed.
- 6. 11 Radiation Protection Pro ram Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.
6.12 Hi h Radiation Area
- 6. 12. 1 In lieu of the "control device" or "alarm signal" required by Paragraph 20. 203(c)(2) of 10 CFR.20, each high radiation area normally accessible" by personnel in which the intensity of radiation is greater than 100 mrem/hr"" but less than 1000 mrem/hr*~ shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit in accordance with site approved procedures.
"by accessible passage and permanently fixed ladders
""measurement made at 18" from source of radioactivity Amendment No. 66 Nine Mile Point - Unit 1 6-26
6.12 Hi h Radiation Area (Cont'd) 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 which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rates in the area have been established and personnel have been made knowledgeable of them
- c. An individual qualified in radiation protection, with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Protection Supervisor or designate in the Radiation Work Permit
- 6. 12. 2 In addition to the requirements of 6. 12. 1 areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem*" shall be provided with locked doors'b prevent unauthorized entry, and the hard keys or access provided by magnetic keycard shall be maintained under the administrative control of the Station Shift Supervisor or designate on duty and/or the Radiation Protection Supervisor or designate. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify in accordance with site approved procedures accordingly, the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, continuous surveillance, direct or remote, such as use of closed circuit TV cameras may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. for individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem"" that are located within large areas, such as the drywell, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off conspicuously posted and a flashing light shall be activated as a warning device
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""measurement made at 18" from source of radioactivity Amendment No. +87 Nine Nile Point - Unit 1 6-27
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6.13 Fire Protection Ins ection 6.13.1 An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site licensee personnel or an outside fire protec-tion firm. 6.13.2 An inspection and audit by an outside qualified fire consultant shall be performed at intervals no greater than 3 years. Procedures shall be established, implemented and maintained to meet or exceed the requirements and recommendations of Section 2. 1. 6. a of NUREG-0578.
- 6. 15 Iodine Honitorin Procedures shall be established, implemented, and maintained to meet or exceed the requirements and recommendations of Section 2.1.8. c of NUREG-0578.
Amendment No. 8$ , 66 Nine Mile Point - Unit 1 6-28
Document Name: NMP-1 TS APP B Requestor's ID: 3815 Author's Name: Jameson C. Document Comments: PH"364 KEEP THIS SHEET WITH DOCUMENT}}