ML20010B728
| ML20010B728 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/28/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010B729 | List: |
| References | |
| NUDOCS 8108170473 | |
| Download: ML20010B728 (54) | |
Text
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UNITED STATES EY%
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NUCLEAR REGULATORYCOMMISSION y9 y WASHINGTON, D. C. 20555
,;%;<dj NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIC ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. DPR-42 1.
The Nuclear Regulatory Comnission (the Commission) has found that:
A.
The applications for amendment by Northern States Power Company (the licensee) dated February ~20, 1980, May 16, 1980, and July 31,1980, comply w.ith the. standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will.not be inimical te the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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I 8108170473 810728 PDR ADOCK 05000282 P
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. 2.
Accordingly, Facility Operating License No. OPR-42 is amended by revising paragraphs 2.C.(2) and 2.C.(3) to read as follows:
(2) Technical Specifications The Technical Specifications contained in' Appendices A and B, as revised through Amendment No. 49, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the following Commission approved documents, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). These approved documents consist of infor-mation withheld from public disclosure pursuant to 10 CFR 2.790(d):
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" Prairie Island Nuclear Generating Plant Security Plan" --
a.
Revision ~4 filed on flarch 3,1978 and Reiti'sion 5 filed September 25, 1978.
b.
" Prairie Island Nuclear Generating Plant Safeguards Contingency Plan" dated itarch 23, 1979, as revised by submittal dated August 20, 1980 which contained revised pages dated July 1, 1980, submitted pursuant to 10 CFR 73.40. The Cor.tingency Plan shall be fully impler.ented, in accordance with 10 CFR 73.40(b), within 30 days of approval by the Commission (February 25, 1981).
c.
" Prairie Island Nuclear Generating Plant Security Guard Force Training and Qualification Plan", subnitted by letter dated August 17, 1979 as amended by Revision 1 submitted itay 16,1980. This Plan shall be followed in accordance witn 10 CFR 73.55(b)(4), 60 days after approval by the Commission (February 25, 1981). All security personnel, as required in the above plans, shall be qualified within two years of this approval (February 25,1981). The licensee may make changes to this plan without prior Commission approval if the changes do not decrease the safeguards effectiveness of the plan. The licensee shall maintain records of and submit reports concerning such changes in the same manner as required for changes made to the Safeguards Contingency Plan pursuant to 10 CFR 50.54(p).
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3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0!4t4ISSION Y888.
R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical
- Specifications Date of Issuance: July 28, 1981 L
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UNlTED STATES s
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.y.. ?"i, NUCLEAR REGULATORY COMMISSION
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E WASMNGTON, D. C. 20555 NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 43 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Northern States Power Company (the licensee) dated February 20, 1980, May 16, 1980, and July 31,1980, comply with the standards and requirements of the Atomic Energy-Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the aoplication, the provisions of the Act, and the rules and regulations of the Conmission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendnent will not be inimical to the common l
defense and security or to the health and safety of the public; and 1
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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e 2.
Accordingly, Facility Operating License No. DPR-42 is amended by revising paragraphs 2.C.(2) and 2.C.(3) to read as follows:
(2) Technical Specifications The Technical Specifications contained in' Appendices A and B, as revised through Amendment flo. 43, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the following Commission approved documents, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). These approved documents consist of infor-mation withheld from public disclosure pursuant to 10 CFR 2.i90(d):,
" Prairie. Island fluclear Generating Plant.. Security Plan" --
a.
Revision 4 filed on March 3,1978 and Revision 5 filed September 25, 1978.
- b. '" Prairie Island !!uclear Generating Plant Safeguards Contingency Plan" dated March 23, 1979, as revised by submittal dated August 20, 1980 which contained revised pages dated July 1, 1980, subnitted pursuant to 10 CFR 73.40. The Contingency Plan shall be fully implemented, in accordance with 10 CFR 73.40(b), within 30 days of approval by the Commission (February 25, 1981).
" Prairie Island !!uclear Generating Plant Security Guard c.
Force Training and Qualification Plan", submitted by letter dated August 17, 1979 as amended by Revision 1 submitted May 16, 1980. This Plan shall be followed in accordance I
l with 10 CFR 73.55(b)(4), 60 days after approval by the Commission (February 25,1981). All security personnel, as required in the above plans, shall be qualified within two years of this approval (February 25,1981). The licensee may make changes to this plan without prior Commission approval if the changes do not decrease the safeguards effectiveness of the plan. The licensee shall maintain records of and submit report., concerning such changes in the same manner as required for changes made to the Safeguards Contingency Plan pursuant to 10 CFR 50.54(p).
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY' COMMISSION f
R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
2 Changes to the Technical Specifications Date of Issuance: July 28, 1981 n
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ATTACHMENT TO LICENSE AMErlDMErlTS AMENDMENT N0. 49 TO FACILITY OPERATING LICEllSE N0. DPR-42 AMErlDMENT N0. 43 TO FACILITY OPERATIriG LICENSE N0. OPR-60 DOCKET tt05. 50-282 AND 50-306 Replace the following pages and insert the new pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
TS-tii TS.1-4 TS.1-5 TS.3.1-2 TS.3.1-3 TS.3.1-3A TS.3.3-4
- TS.3. 5-1 TS.3.5-3 Table TS.3.5al (Page 2"of 2) (New Page)
Table TS.3.5-6 (New Page)
TS.3.10-6 Table TS.3.12-1 (Page 1 of 8) (4 flew Pages)
Through (Page 8 of 8)
TS.3.14-1 TS.3.14-2 TS.3.14-3 TS.3.14 4 TS.3.14-5 TS.3.14.6 Table TS.3.14-1 (pg 1 of 3)
Table TS.3.14-1 (pg 2 of 3)
Table TS.3.14-1 (pg 3 of 3)
Table TS.4.1-1 (Page 1 of 5)
Table TS.4.1-1 (Page 5 of 5)
Table TS.4.1-2A TS.4.4-4 TS.4.5-2 TS.4.5-3A TS.4.6-1 TS.4.16-2 TS.4.16-3 TS.4.16-4 TS.4.16-5 TS.4.16-6 (New Page)
TS.6.1-1 Figure TS.6.1-1 Figure TS.6.1-2 TS.6.2-1 TS.6.2-3 TS.6.2-5 TS.6.2-6 TS.6.4-1 TS.6.5-2
TS-iii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.1 1 Unit 1 Reactor Vessel Toughness Data 3.1-2 Unit 2 Reactor Vessel Toughness Data 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System l
3.9-1 Radioactive Liquid Waste Sampling and Analysis 3.9-2 Radioactive Gaseous Waste Sampling and Analysis 3.12-1 Safety Related Shock Suppressors (Snubbers) j 3.14-l' Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of
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Instrument Channels 4.lJ2A Minimum Frequencies for Equi'pment Tests ~
4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.4-1 Unit 1 and Unit 2 Penetration Designation for Leakage Tests 4.10-1 Prairie Island Nuclear Generating Plant-Radiation Environ = ental Monitoring Program Sample Collection and Analysis Environmental Monitoring Program 4.12-1 Steam Generator Tube Inspection 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition 6.7-1 Special Reports Prairie Island 'Init 1 Amendment No. 43, 46, 49 Prairie Island Unit 2 Amendment No. 27, 40, 43
IS.1-4 G.
Limiting Safety Svstem Settings Limiting safety system settings are settings on protective instrumenta-tion that initiate automatic protective action at a level such that safety limits will not be exceeded.
H.
Limiting Conditions for Operation Limiting conditions for operation are those restrictions on unit operation resulting from equipment performance capability that must be met in order to assure safe operation of the unit.
I.
Operable A system, subsystem, train, component or device shall be Operable or have Operability 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, 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).
When a system, subsystem,' train, component or device is determined to be
, inoperable solely because its emergency power source is inoperable, or solely because it's 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 requirements of this paragraph.
The operability of a system or component shall be considered to be estab-lished when: (1) it satisfies the Limiting Conditions for Operation in Specification 3.0, (2) it has been tested periodically in accordance with Specification 4.0 and has met its performance requirements, and (3) its condition is consistent with the two paragraphs above in this TE scetion 1.1.
J.
Power Operation Power operation of a unit is any operating condition that results when the reactor of that unit is critical, and the neutron flux power range instru-mentation indicates greater than 2% of rated power.
K.
Protection Instrumentation and Logic 1.
Protection System The protection system consists of both the reactor trip system and the engineered safety feature system. The protection system encompasses all electrical and mechanical devices and circuitry (from sensors through the actuating devices) which are required to operate in order to produce the required protective function. Tests of protection systems will be considered acceptable when overlapped if run in pscts.
Prairie Island - Unit 1 Amendment No. 49 Prairic Island - Unit 2 Amendment No. 43
TS.1-5 2.
Protection System Channel A protection system channel is an arrangement of components and modules as required to generate a single protective action signal when required by a unit condition.
The channel loses its identity where single action signals are combined.
3.
Logic Channel A logic channel is a group of relay contact matrices which operate in response to anatog channel signals to generate a protective action signal.
L.
Quadrant Power Tilt Quadrant power tilt is the ratio of the maximum quadrant power indicated by an upper excore detector to the average reactor power indicated by the upper excore detectors or the ratio of the maximum quadrant power indicated by a lower excore detector to the average reactor power indicated by the lower excore detectors, whichever is
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greater. Power is proportional to excore detector current times
'its calibration factor. P :rcentage quadrant power tilt is 100 times the amount the quadrant power tilt ratio exceeds one.
M.
Rated Power Rated power of a uni: is the steady state heat output of 1650 megawat'es thermal (MWt) from the reactor core of that unit.
N.
Reactor Critical A reactor is critical when the neutron chain reaction is self-sustain-ing and k,gg = 1.0, 3.
Refueling Operation Refueling operation of a unit is any operatioa involving movement of those core components that could af fect the reactivity of the core when the reactor vessel head is unbolted or removed.
P.
Shutdown 1.
Hot Shutdown A reactor is in the hot shutdown condition when the reactor is suberitical by an amount greater than or equal to the margin as specified in Figurg TS.3.10-1 and the reactor coolant average temperature is 547 F or greater.
2.
Cold Shutacwn A reactor is in the cold shutdown condition when the reactor is subcritical by at least 1% Ak/k and the reactor coolant average temperature is less than 200'F.
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Prairie Island Unit 1 Amendment No. 49 Prairie Island Unit 2 Amendment No. 43
e TS.3.1-2 4.
Pressurizer a.
Whenever average reactor coolant system temperature is above 350 F or the reactor is critical, the pressurizer shall be operable with:
1.
Steam bubble 2.
Pressurizer heater groups "A" and "B" and their associated safeguards power supplies operable 3.
At least one operable spray b.
With the pressurizer inoperable due to an inoperable heater group restore the equipment to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or olace the reactor
'n at least Hot Shutdown within the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
With t,he pressurizer inoperable for any other reason than c.
(b) above, the reactor shall be placed in at least Hot Shutdown within the following-12 hours.-
d.
At least one pressurizer safety valve shall be cperable whenever the head is on the reactor vessel, except during hydrostatic tests.
Both pressurizer safety valves shall be~
operable whenever average reactor coolant system temperature is above 350 F or the reactor is critical. Pressurizer safety valve lift setting shall be 2485 psig ! 1%.
Except as specified in (f) and (g) below, two power operated e.
relief v.alves (PORV's) and their associated block valves shall be operable whenever average reactor coolant system temperature is above 350 F or the reactor is critical.
f.
With one or more PORV's inoperable, within one hour either restore the PORV(s) to operable status or close the associated block valve (s).
If this cannot be done, place the reactor in the Cold Shutdown condition within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
g.
With one or more block valves inoperable, within one hour either restore the block valve (s) to operable status or close the valve..If this cannot be done, place the reactor in the Cold Shutdown condition within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
DPR Amendment No. 4y, 49 DPR Amendeant No. $@,43
TS.3,1-3 Basis When the boron concentration of the reactor coolant system is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the primary system volume in approximately one-half hour.
" Steam Generator Tube Surveillance", Technical Specification 4.12, identifies steam generator tube imperfections having a depth >50% of the 0.050-inch tube wall thickness as being unacceptable for power operation. The results of steam generator burst and tube collapse tects submitted to-the staff have demonstrated that tubes having a wall thickness greater than 0.025-inch have adequate margins of safety against failure plant operation and design basis accidents.fue to loads imposed by normal Part A of the specification requires that both reactor coolant pumps be operat-ing when the reactor is critical to _ provide core cooling in the event that a loss of flow occurs.
In the event of the worst credible coolant flow loss (loss of both pumps,from-100% power) the minimum calculated DNBR remains well above 1.30.
Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Critical operation, except for low power physics tests, with less than two pumps is not planned. Above 10% power, an automatic reactor trip will occur if flow from either pu=p is lost.
Below 10% power, a shutdown under administrative control will be made if flow from either pump is lost.
The pressurizer is needed te maintain acceptable system pressure during nor=al plant operation, including surges that may result follcuing anticipated transients. Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point. Below 350*F and 450 psig in the reactor coolant system, the residual heat removal system can re=ove decay heat and thereby control system temperature and pressure.
If no residual heat were removed by any of the means available, the a cunt of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization of the reactor coolant system for reactor coolant temperatures less than 350*F.
The combined capacity of bort safety valve greater than the maximum surge rate resulting from complete 1.oss of load.g is i
1 References.
FSAR, Section 14.1.9.
Testi=ony by J. Knight in the Prairie Island'Public Hearing on January 128, 1975.
Prairie Island Unit 1 Amendment No. 47, 49 Prairie Island Unit 2 Amendment No. 41, 43 f.
n.
e TS.3.1-3A Basis (continue i)
The requirement that c>o groups of pressurizer heaters be operable provides assurarte that at least one group will be available during a loss of of fsite power to maintai.s natural circulation. Backup heater group "A" is normally supplied by one safeguards bus.
Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bos. Tests have confirmed the ability of either group to caintain natural circulation condit ions.
The prescurizer power operated relief valves (PORV's) operate to relieve reactor coolant system pressure below the setting of the pressurizer Code safety valves.
These relief valves have remotely operrted block valves to provide a positive shutoff capability should a relief valve become inoperable.
The FORV's are pneumatic valves operated by instrument air. They fail closed on loss of air or loss of power co their DC solenoid valves. The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses.
The Specifications require that at least two methods of removing decay hec.c are available for each_ reactor. Above 350*F, both steam gererators must be operable to serve this function. Below 350*F, either a steam generator or
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a-residual heat removal loop are capable.of removing. decay heat and any combination of two loops is specified.
If redundant means are not avail-able, the reactor is placed in the cold shutdown condition.
References 1FSAR, Section 14.1.9 2Te s ti=ony by J Knigh t in the Prairie Island Public Hearing on January 28, 1975.
DPR Amendment No. 4$, 49 DPR Amendment No. $Q, 43
b TS.3.3-4 c.
Any redundant valve or damper required for functioning of the containment air cooling system and the containment spray system during and following accident conditions may be inoperable provided it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Prior to initiating repairs, all valves in the system that provide redundancy shall be demonstrated to be operable. ~
C.
Component Cooling Water System 1.
Single Unit Operation a.
A reactor shall not be made or maintained critical nor shall it be heated or maintained above 200 F, unless the following conditions are satisfied, except as permitted in Specification 3.3 C.1.b. below.
(1) The two component cooling pumps assigned to that unit are operable.
(2) The two component cooling heat exchangers assigned to that unit are operable.
(3) All valves, interlocks, instrumentation and piping associated with the above components, and required for the functioning of the system during accident conditions, are operable.
b.
During startup operation or power operation, any one of the following conditions of inoperability may exist provided startup operation is discontinued until operability is restored.
The reactor shall be placed in the hot shutdown condition if during power operation operability is not restored within the time specif.' d, and it shall be placed in the cold shutdown condition if operability is not restored within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
(1) One of the assigned component coolink pumps may be out of service for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(2) One of the assigned component cooling heat exchangers may be out of service for a period not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
2.
Two-Unit Operation A second reactor shall not be made or maintained critical nor a.
shall it he heated or maintained above 200 F, unless the I
following conditions are satisfied, except as provided by Specification 3.3 C.2.b. below.
Prairie Island Unit 1 Amendment No. 49 Prairie Island Unit 2 Amendment No. 43 i
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o TS.3.5-1 3.5 INSTRUMENTATION SYSTEM Applicability Applies to protection system instrumentation. -
OLjectives To provide for automatic initiation of the engineered safety features in the event that principal process variable limits are exceeded, and to delineate the conditions of the reactor trip and engineered safety feature '.nstrumentation necessary to ensure reactor safety.
Specification A.
Limiting set points for instrumentation which initiates operation of the engineered safety features shall be as stated in Table TS.3.5-1.
B.
For on-line t'esting or in the event of failure of a sub-system instrumentation channel, plant operation shall-be permitted to continue at rated power in accordance with Tables TS.3.5-2 through TS.3.5-6.
C.
If the number of channels of a particular sub-system in service falls below the limits given in the column entitled Minimum Operable Channels, or if the specified Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in the column titled Operator Action of Tabler TS.3.5-2 through TS.3.5-6.
D.
In the event of sub-system instrumentation channel f ailure permitted by Specification 3.5.B, the requirements of Tables TS.3.5-2 through TS.3.5-6 need not be observed during the short period of time the operable sub-system channels are tested where the failed channel must be blocked to prevent unnecessary reactor trip.
If the test time exceeds four hours, operation shall be limited according to the requirement shown in the column titled Operator Action of Tables TS.3.5-2 through TS.3.5-6.
Basis Instrumentation has been provided to sense accident conditions and to initiate reactor trip and operation of the Engineered Safety Features (1).
Prairie Island Unit 1 Amendment No. 49 Prairie Island Unit 2 Amendment No. 43
- o IS.3.5-3 o
Steam Line Isolation In the event of a steam line break, the steam line stop valve of the affected line is automatically isolated to prevent continuous, uncontrolled steam release from more than one steam generator. The steam lines are isolated on high containment pressure (Hi-Hi) or high steam line flow in coincidence with low T and safety injection or high steam flow (Hi-Hi)
In coincidence with sa$e$y injection. Adequate protection is afforded for breaks inside or outside the containment even when it is assumed tnat the steam line check valves do not function properly.
Containment Ventilation Isolation Valves in the containment purge and inservice purge systems automatically close on receipt of a Safety Injection signal or a high radiation signal.
Gaseous and particulate monitors in the exhaust stream or a gaseous monitor in the exhaust sta:k provide the high radiation signal.
Ventilation System Isolation In the event of a high energy line rupture outside of containment, redundant isolation dampers in certain ventilation ducts are closed.
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Safeguards Bus Voltage Relays are provided on buses 15, 16, 25, and 26 to detect loss of voltage and degraded voltage (the voltage level at which safety related equipment may not operate properly). On loss of voltage, the automatic voltage restoring scheme is initiated immediately, tihen degraded voltage is sensed, the voltage restoring scheme is initiated if acceptable voltage is not restored within a short time period.
This time dalay preven.s initiation of the voltage restoring scheme when large loads are started and bus voltage
=cmentarily dips below the degraded voltage setpoint.
Auxiliary Feedwater System Actuation The following signals aute=stically start the pumps and open the steam admission control valve to the turbine driven pump of the affected unit:
1.
Low-low water level in either steam generator 2.
Trip of both main feedwater pumps 3.
Safety injection signal 4.
Undervoltage on both 4.26 KV normal buses (turbine driven pump only)
Manual control from both the control room and the Hot Shutdown Panel are also available.
The design provides assurance that water can be supplied to the steam generatcrs for decay heat removal when the normal feedwater system is not available.
Prairie Island Unit 1 Amendment No. 46, 49 Prairie Island Unit 2 Amendment No. 40, 43
TS.3.5-4 Limiting Instrument Setpoints 1.
The high containment pressure limit is set at about 10% of the maximum internalpgsure.
Initiation of Safety Injection protects against loss of coolant or steam line break accidents as discussed in the safety analysis.
2.
The Hi-Hi containment pressure limit is set at about 50%'of the maximum internal pressure for iniriation of containment spray and at about 30%
for initiation of steam line isolation.
Initiation of Containm g Spray andSteamLineIsolationpgectsagainstlargelossofcoolant or steam line break accidents as discussed in the safety analysis.
3.
The pressurizer low pressure limit is set substantially below system operating pressure limits. However, it is sufficiently high to pro againstalossofcoolantaccidentasshowninthesafetyanalysis.gt 4.
The steam line low pressure signal is lead / lag compensated and its set-point is set well above the pressure expected in the event steam line break accident as shown in the safety analysis. g a large 5.
The high steam line flow limit is set at approximately 20% of nominal full-load flow at the no-load pressure and the high-high steam line flow
. limit is set at approximately 120%.of nominal full-load flow at the full load pressure in order to protect against large steam break accidents.
The coincident low T setting limit for steam line isolation initiation is set below its hot"sftutdown value. The safety analysis shows settingsprovideprotectionintheeventofalargesteambreak.gtthese 6.
Steam generator low-low water level and 4.16 KV Bus 11 and 12 (21 and 22 in L* nit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System.
Selection of these setpoints is discussed in Section 2.3 of the Technical Specifications.
7.
High radiation signals providing input to the Containment Ventila01on Isolation circuitry are set in accordance with the Radioactive Effluent Technical Specifications.
The setpoints are established to prevent exceeding the limits of 10 CFR Part 20 at the site boundary.
8.
The degraded voltage protection setpoint is 901 2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the degraded voltage setpoint. The degraded voltage protection time delay of 612 seconds has been shown by testing j
and analysis to be long enough to allow for voltage dips resulting from the starting of large loads. This time delay is also consistent with the maximum time delay assumed in the ECCS analysis for starting of a safety injection pump. A maximum limit on the degraded voltage setpoint has been i
established to prevent unnecessary actuation of the voltage restoring scheme.
(
The loss of voltage protection setpoint is approximately 55% of nominal 3
4160 V bus voltage.
Relays initiate a rapid (less than two seconds)
I transfer to an alternate source on loss of voltage.
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l Prairie Island Unit 1 Amendment No. 4, 49 l
l Prairie Island Unit 2 Amendment.No. M, Q
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TAlil.E TS.3.5-1 (continued)
ENGINEERED SAFL"rY INITIATION INSTitHMENTATION LIMITING SET POINTS
' yy FilNCTIONAl, UNIT CliANNEl.
LIMITING SET POINTS rr QQ 10.
4KV Safeguards ilusses a.
llegraded Voltage Voltage Restoration yp Voltage (% nominal) 90 1 2%
EE gg Time Delay 612 sec cc-g g-13 l.oss of Voltage
- nn we J.
Voltage (% nominal) 55% i 10%
Time Delay-2 1 2 see 2.
Voltage (% nominal) 90 1 2%
Time Delay 212 sec
?
d b'
b aa R
- a. a.
El El H
bb Ny nn
& D.
Y e
00 N
P%
I c.
TAltl.E TS. 3:5-6 yy INSTRUMENT OPERATING CONDITIONS FOR AUXILIARY EI.ECTRICAI. SYSTEM SS 1
3 4
SS MINIMUM MINIMUM PERMISSIBLE C?ERATOR ACTION IF
.~
OPERABLE DECREE OF 11YPASS CONDITIONS OF COLUMN Y5 FilNCY 10NAI. UNIT CllANNEl.S REDllNDANCY CONDITIONS 1 OR. CANNOT BE MET i
EE ao a
p.
1.
Degraded Voltage 1/ilus 1/Itus Place inoperable channel in the y g:
4KV Safeguards liusses tripped condition within one hour p.p of hot shutdown.***
u e 2.
a.
Loss of voltage 1/ilus 1/llus Place inoperable channel in the tripped. condition within one hour 4KV Safeguard i
lius (90%)
of hot shutdown.***
b.
l.oss of voltage 1/llus 1/isus Place inoperable channel in the 4KV Safeguard tripped condition within one hour lius (55%)
of hot shutdown.***
4 l
- 1f minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps $hallbetakentop.lacetheunit in cold
{
gg shutdown conditions.
N
- n. n.
n n g
b
[
n n w
O WW
TS.3.10-6 e
E.
Rod Misalignment Limitations 1.
If a full-length rod cluster control assembly (RCCA) is misaligned from its bank by more than 15 inches, the rod will be realigned or the core power peaking factors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and Specffication 3.10.B applied.
If peaking factors are not deter-mined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the high neutron flux trip setpoint shall be reduced to 85 percent of rating.
2.
a.
If the bank demand position is greater than or equal to 215 steps, or less than or equal to 30 steps and the rod position indicator channel differs by more than 24 steps, that rod control cluster assembly (RCCA) shall be considered misaligned.
b.
If the bank demand position is between 30 and 215 steps and the rod position indicator channel differs by more than 12 steps, that RCCA shall be considered misaligned.
3.
If the misaligned RCCA is not realigned within a total of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, l
the RCCA shall be declared inoperable.
F.
Inoperable Rod Position Indicator Channels 1.
If a rod position indicator (RPI) channel is out of service then a.
For operation between 50% and 100% of rating, the position of the RCCA shall be checked directly by core instrumentation (excore detector and/or thermocouples and/or movable incore
-detectors) every shift or subsequent to rod motion exceeding a total of 24 steps, whichever occurs first.
b.
During operation below 50% of rating, no special monitering is required.
2.
The plant shall be brought to the hot shutdown condition should more than one RPI channel per group or more than two RPI channels per bank be found to be inoperable during power operation.
3.
If a full length rod having a rod position indicator channel out of service is found to be misaligned from 1.a. above, then apply Specification 3.10.E.
G.
Inoperable Rod Limitations 1.
An 1. )perable rod is a rod which (a) does not trip, (b) is declared inoperable under Specification 3.10.E or 3.10.H or (c) cannot be moved by its drize mechanism and cannot be corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Prairie Island Unit 1 Amendment No. 32, 44, 49 Prairie Island Unit 2 Amend =ent No. 26, 38, 43
~.
TABLE TS.3.12-1 (Page 1 of 8)
SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Area During No.
Location Elevation (A or I) to Remove Shutdown UNIT I AFSH-22 A&B Main and Aux-773'-4-1/4" A
AFSH-36 iliary Steam 745'-7-1/4" A
AFSH-39 699'-10-1/4" A
AFSH-48 699'-6-1/4""
A MSDH-25 736'-6-7/16" A
X MSDH-26 756'-7-1/4" A
X "dDH-29 756'-7-1/4" A
MSDH-30 736-6-7/16" A
MSH-48 739'-1-11/16" A
X MSH-62 A&B 735'-6" A
MSH-68 A&B' 755'-8" A
UNIT II AFSH 2, Main and Auxiliary 749'-4" A
2 AFSH-19 Steam 745'-7-1/4"
- A AFSH-20 745'-7 '.f4" A
AFSH-24 7/5 -6" A
AFSH-29 A&B 721'-1-9/16" A
AFSH-33 707'-5" A
AFSH-39 696'-6-1/4" A
AFSH-40 696'-6-1/4" A
AFSH-44 750'-7-1/2" A
AFSH-46 750'-7" A
MSDH-17 739'-0" A
X MSDH-18 759'-0" A
X MSDH-19 739'-0" A
KSDH-20 759'-0" A
Prairie Island Unit 1 Amendment No.14, 49 Amendment No. 8, 43 Prairie Island Unit 2
_~
o, TABLE TS.3.12-1 (Page 2 of 8)
Snubbers In High Accessible or Especially Radiation Soubber Inaccessible Difficult Areas During No Location Elevation (A or I) to Remove Shutdown UNIT II MSE-23 Main and Auxiliary 739'-1-3/16" A
X MSK-54 A&3 Steam 756'-0-1/16" I
MSH-81 A&B 735'-9" A
e MSH-82 A&3 755'-8" A
MaH-83 761'-13/16""
I UNIT I RERRE-5 Safety Injection 723'-4-1/4" I
RERRH-41 698'-11" I
RERRH-58 670'-0" A
KERRE-60 670'-0" A
RPCH-160 718'-1/2" I
RSIK-92 714'-11" I
RSIK-93 714'-11" I
RSIE-95
~
711'-1" I
RSIH-96 711'-2" I
RSIH-98 701 ' -2" --
I-RSIH-163 717'-9" I
RSIH-l67 717'-9" I
RSIR-413 A&B 722'-8" A
RS!H-411 716'-10" I
RSIR-442 717'-9-1/2" I
RSIE-469 707'-6-1/2" I
RSIH-476 707'-1-3/~"
I SIRE-9 737'-0" I
SIRE-11 718'-6""
I SIRE-17 730'-0" I
SIRK-13 730'-0" I
SIRH-22 711'-4" I
SIRE-23 A&B 711'-4" I
Amendment No.14, 49 Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 8, 43
TABLE TS.3.12-1 (Page 3 of 8)
SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or Espe'cially Radiation Snubber Inaccessible Difficult Areas During Mo.
Location Elevation (A or I) to Remove Shutdown UNIT II RHRH-13 Safety Injection 673'-9" A
RHRH-14 674'-0" A
RHRH-52 670'-6" A
RHRK-54 670'-6" A
RHRRR-19 700'-11" I
RHRRH-23 711'-2" I
RHRRH-28 707'-4" I
RSIH-265 699'-9" I
RSIH-268 713'-9-3/16" I
RSIH-343 719'-8-11/16" 1
RSIH-349 703'-11" I
RSIH-350 703'-11" I
RSIF-353 A&B 701'-9" I
~
SIH 720'-0" A
SIH-49 A&B 737'-3" A
SIH-53 710'-3" A
SIRH-4A 711'-6-1/8" I
sIRH-4B 711'-3" I
SIRH-7 716'-3-1/16" I
SIRH-18 722'-6" I
l r
l Prairie Island Unit 1 Amendment.No. 14, 49 Prairie Island Unit 2 Amendment No. 8, 43
o TABLE TS.3.12-1 (Page 4 of 8)
SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas Duris No.
Location Elevation (A or I) to Remove Shut.down UNIT I RCRH-5 A&B Reactor Coolant 732'-6" I
RCRH-12 A&B 720'-7" I
RCRH-76 762'-8" I
RCRH-27 A&B 7 61 -7" I
RCRH-34 764'-7" I
RCRR-45 765'-1" I
RE RH-46 765'-1" I
RCRH-47 745'-10" I
'RHRRH-15 705'-6" I
RHRRH-27 705'-6" 1
RHRRH-29 A&B 705'-6" I
UNIT II.
~
RCRH-5 Reactor Coolant 731'-6" I
RCRH-8 717'-6"-
I RCRH49 712'-0" I
RCRH-14 705'-9" I
RCRR-20 71.* ' - 7 "
I RCRH-25 732'-2" I
RCRH-26 757'-7" I
RCRH-31 764'-1" I
RCRH-45 724'-6" I
RCRH-46 758'-3" I
RCRH-47 760'-3" I
RCRH-48 765'-1" I
RCRE-49 765'-1" I
RRCH-279 A&B 724'-9" I
RRCH-282 723'-2" I
RRCH-284 A&B 725'-8" I
RHRRH-2 699'-0" I
RHRRR-4 705'-11" I
ERRRH-9 705'-11" I
RHRRH-15 699'-0" I
Prairie Island Unit 1 Amendment No.14, 49 Prairie Island Unit 2 Amendment No. 8, 43
a TABLE TS.3.12-1 (Page 5 of 8)
SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or E. specially Radiation Snubber Inaccessible
. Difficult Areas During No.
Location Elevation (A or I) to Remove Shutdown UNIT I CWH-359 Cooling Water 705'-8" A
CWH-380 706'-11" A
CWH-385 709'-0" A
CWH-394 731'-0" A
CWH-395 746'-6" A
CWH-405 707'-10" A
CWH-429 722'-11" A
CWH-432 722'-11" A
CWH-433 735'-11" A
CWH-434 735'-11" A
CWH-436 737'-11" A
~
UNIT II CWH Cooling Water 709'-3" A -.
CWH-35 746'-8" A
CWH-39 710'-6" A
CWH-40 710'-6" A
CYH-44 730'-11" A
CWH-45 709'-0" A
CWH-49 723'-0" A
CWH-50 723'-10" A
CUH-52 736'-0" A
CWH-54 738'-0" A
Prairie Island Unit 1 Amendment No.14, 49 Prairie Island Unit 2 Amendment No. 8, 43
o TABLE TS.3.12-1 (Page 6 of 8) 1 SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No.
Location Elevation (A or I) to Remove Snutdown UN IT_ _I_
AFWH-72 Feedwater 752'-0" I
AFWM-82 728'-11" A
AFWH-84 728'-11" A
UNIT II AFWH-72 A&B Feedwater 706'-3/4" A
FWH-72 A&B 751'-0" I
UNT.'t I 25.12620.003 3
Steam Generator 760'-9-1/2" I
X 25.12620.003 - 4 760'-9-1/2" I
X 25.'12620.003 - 5 760'-9-1/2" I
X 25.12620.003 - 6 760'-9-1/2" I
X 25.12620.003 - 7 760'-9-1/2" I
X 25.12620.003 - 8 760'-9-1/2" I
X 25.12620-003 - 10 760'-9-1/2" I
X 25.12620.003 - 15 760'-9-1/2" I
X
~
UNIT II 25.12620.003 - 1 760'-9-1/2" I
X 25.12620.003 - 2 760'-9-1/2" I
X 25,12620.003 - 9 760' 9-1/2" I
X 25.12620.003 - 11 76T 1/2" I
X 25.12620.003 - 12 760'-9-1/2" I
X 25.12620.003 - 13 760'-9-1/2" I
X 25.12620.003 - 14 760'-9-1/2" I
X
- 25. 2620.003 - 16 760'-9-1/2" I
X UNIT I CVCH-182 Chemical & Vol 707'-6" A
RCRH-16 A&S Control 705'-2" I
RCRH-19 705'-2" I
RCRH-21 705'-7" I
RCRH-23 A&B 715'-11" I
RCVCH-907 A&B 717'-11" I
RCVCH-1293 712'-0" I
RPCH-22 703'-1" I
RPCH-23 703'-1" I
i RPCH-121 707'-9" I
i RPCH-139 704'-4" I
RPCH-140 707'-7" I
PRCH-146 714'-7" I
RPCH-147 714'-10" I
WDRH-24 707'-9" I
Prairie Island Unit 1 Amendment No. 14, 49 Prairie Island Unit 2 Amendment No. 8, 43 l
TABLE TS.3.12-1 (P(22 7 of 8)
SAFETY RELATED SHCCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No.
Location Elevation (A or I) to. Remove Shutdown UNIT 11-RCVCH-1396 Chemical & Vol 702'-10" 1
RCVCH-1505 Control 708'-6" I
RCVCH-1513 710'-1" 1
RCVCH-1524 719'-1" I
RCVCH-1574 721'-0" I
RCVCH-1668 705'-5" I
RCVCH-1373 722'-11" I
RCVCH-1389 706'-1" I
RRCH-253 704'-4" I
RRCH-255 704'-8" 1
RRCH-261 707'-2" I
RRCH-288 707'-2" I
RRCH-291 704'-6" I
RRCH-292 704'-7" I
UNIT I -
CCH-304, comp Cooling 717'-7" A
CCH-373 712'-4" A
CCH-376 ASB 700'-5" A
CCH-377 703'-0" A
CCH-378 708'-4" A
CCH-3b 0 670'-8" A
CCH-381 A&B 671'-4" A
CCH-397 699'-3" A
CCH-398 A&B 671'-4" A
UNIT II CCH-161 Comp Cooling 717'-7" A
CCH-166 719'-11" A
CCH-167 720'-0" A
CCH-172 720'-0" A
CCH-173 708'-5" A
CCH-176 705'-3" A
CCH-179 A&B 671'-4" A
CCH-180 670'-8" A
CCH-181 708'-4" A
CCH-182 704'-2" A
CCH-185 A&B 671'-4" A
CCH-186 670'-10" A
UNIT I RCSH-81 Containment Spray 76"'-9" I
RCSH-62 760'-8" I
RSCH-83 A&B 732'-1" I
UNIT II CSH-75 A&B Containment Spray 731'-i0" I
CSH-76 752'-7" I
CSH-79 751'-9" I
CSH-82 A&B 731'-11" I
CSH-83 767'-2" I
CSH-84 767'-2" I
CSH-210 698'-0" I
CSH-215-698'-0" A
CSH-224 710'-6" A
49 Prairie Island Unit 1 Amendment No. If Prairie Island Unit 2 Amendment No. 5,'43
b O
TABLE TS.3.12-1 (Page f of 8)
SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)
Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No.
Location Elevation (A or I) to Remove Shutdown-UNIT I RRHH-20 RHR 704'-3" A
RRHH-62 705'-10" A
UNIT II CVCRH-6 RHR 711'-0" I
RRHH-21 704'-5" A
I i
i Prairie Island Unit 1 Amendment No. 14, 49 Prairie Island Unit 2 Amendment No. 8, 43 b -
m c.---
. ---..---. - ~.. _.., - _ _ _.,. _, -. -
b 0
TS.3.14-1 3.14 FIRE DETECTION AND T IECTION SYSTEMS Aeplicability Applies to instrumentation and plant systems used for fire detection and protection of the nuclear safety-related structures, systems, and componeats of the plant.
Objective To insure that the structures, systems, and components of the plant important-to nuclear safety are protected from fire damage.
Specification A.
Fire Detection Instrumentation 1.
Excspc as specified below, the minimum fire detection instrumentati s for each fire detection zone shown in Table 3.14-1 shall be operab!
whenever equipment in that fire detection :one is required to be operable. Eire ' detection instruments located within containment are not required to be operable during the performance of Type A
~
containment leakage rate tests.
2.
If Specification 3.14.A.1 cannot be met:
a.
Within one hour, establish a fire watch patrol to inspect the
- ene with the insperable instruments at least once per hour.
Fire zones located inside pri=ary containment are exempt fren 5 is raqui rement when contain=er.t integrity is required.
b.
Restore the inoperable instrucents to operable status within 14 days or submit a 30-day written report outlining the cause of the malfunction and the plans for rescoring the instruments to operable status.
3.
Fire Suooression Water System 1.
Except as specified in 3.16.3.2 or 3.14.B.3 below, the system shall be operable at.111 ti=es with:
a.
The following pu=ps, including automatic initiation logic, operable and capable of delivering at least 2000 gpm at a dischar** pressure of 108 psig.
1.
Diesel-driven fire pump Motor-driven fire pu=p 3.
Screen wash pu=p Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43
o TS.3.14-2 o b.
An operable flow path capable of taking suction from the river and transferring the water through distribution piping with operable sectionalizing control or isolation valves to the yard hydrant valves and the firs t valve ahead of each deluge valve, hose station, or sprinkler system required to be operable.
2.
With one or two of the pumps required by Specification 3.14.B.1.a inoperable, restore the inoperable equipment to operable status within seven days or provide a 30-day written report outlining the plans and procedures to be used to provide for the loss of redundancy in the Fire Suppression Water system. With an inoperable pump, perform the surveillance required by Specification 4.16.B.2.
3.
With the fire suppression water system otherwise inoperable:
a.
Establish a backup Fire Suppression Water System within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
Provide prompt notification with a written follevup report outlining the actions taken and the plans and schedule for restori,ng the system to opera'.
.tatus.
c.
If Specification 3.14.B.3.a cannot be met, the reactors shall be placed in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdewn within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
C.
Scray add Serinkler Systems 1.
Whenever equipment protected by the following spray and sprinkler systems is required to be operable, the spray and sprinkler system shall be operable:
a.
Auxiliary Feed Pump Room WP-10 b.
Diesel Generator Areas FA-1 c.
Unit No. 1 Electrical Penetration Area FA-3 d.
Unit No. 1 Electrical Penetration Area PA-4 e.
Unit No. 2 Electrical Penetration Area PA-6 f.
Unit No. 2 Electrical Penetration Area PA-7 g.
Screenhou'se PA-9 2.
If Specification 3.14.C.1 cannot be =et, a continuous fire watch with backup fire suppression equipment shall be established within one hour. Restore inoperable spray and sprinkler systems to operable status within 14 days or submit a 30-day written report outlining the cause of inoperability and the plans for restoring the system to operable status.
Prairie Island Unit 1 Amendnent No. 26, 49 Prairie Island Unit 2 A=endment No. 20, 43
- a
,o TS.3.14-3 D.
Carbon Dioxide System 1.
Except as specified in 3.14.D.3 below, the CO system protecting 2
the relay and cable spreading roca area shall be operable with a minimum level of 60% in the CO se rege tank.
2 2.
During those periods when the relay and cable spreading room area is normally occupied, automatic initiation of the CO 87'*** **7 2
be bypassed. During those periods when the area is normally unoccupied, the CO system shall be capable of automatic initiation 2
unless there are personnel actually in the area.
3.
If specification 3.14.D.1 cannot be met, a continuous fire watch with backup fire suppression equipment shall be stationed in the relay and cable spreading room within one hour.
Restore the system to operable status within 14 days or submit a 30-day written report outlining the cause of inoperability and the plans for restoring the system to operable status.
E.
Fire Hose Stations. -
~1.
Whenever equipment protected by hose stations ~ in the following areas is required to be operable, the hose station (s) protecting that area shall be operable:
a.
Diesel Generator Roems 5.
Safety Related Switchgear Roems c.
Safety Related Area.s of Screenhouse d.
Auxiliary Building e.
Control Roem f.
Relay & Cable Spreading Room g.
Battery Roems h.
Auxiliary Feed Pump Room 2.
If Specification 3.14.E.1 cannot be met, within one hour hoses supplied frem operable hose stations shall be made available for routing to each area with an inoperable hose station.
Restore the inoperable hose station (s) to Operable status within 14 days or submit a 30-day written report outlining the cause of the inoperability and the plans and schedule for restoring the stations to Operable status.
Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43
b TS.3.14-4 F.
Yard Hydrant Hose Houses 1.
Whenever equipment in the following buildings is required to be operable, the yard hydrant hose houses in the main yard loop a31= cent to each building shall be operable:
a.
Unit No. 1 Reactor Building b.
Unit No. 2 Reactor Building c.
Turbine Building d.
Auxiliary Building e.
Screen house 2.
If Specification 3.14.F.1 cannot be met, within one hour have sufficient additional lengths of 2-1/2 inch diameter hose located in adjacent operable yard hydrant hose house (s) to provide service to the unprotected area (s).
Restore the yard hydrant hose house (s) to Operable sta619 within 14 days or submit a 30-day written report outlining the cause of the inc?erability and the plans and schedule for restoring the houses to Operable status.
G.
All penetration fire barriers in fire area boundaries protecting equipment tequired to be operable shall be operable.
2.
If Specification 3.14.G.1 cannot be met, a continuous fire watch shall be estabif.shed on at least one side of the affected penetra-tion (s) within one hour.
Restore the inoperable penetration fire barriers to Operable status within 14 days or submit a 30-day written report outlining the cause rf the inoperability and the plans and schedule for restoring the barriers to Operable status.
Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No.'20, 43
'o TS.3.14-5 3asts Ionization, photoelectric, and thermal type fire detectors are, located l
throughout safety related structures.
These detectors sense,the products of combustion during the very early stages of a fire or the heat emitted by a fire. The detectors in each area initiate an alarm in the control room.
The specifications require a minimum number of detectors to be operable in each area.
If this number is not operable, except for fire detectors located in primary containment, a patrolling fire watch is established in the affected area.
If an area is found to have an inoperable detector, the alarm for the af fected zone may be bypassed while the detector is being repaired.
Primary containment detectors are unique since (1) they arr. inaccessible during normal operation, and (2) no significant fire hazard exists inside containment during normal operation.
Inoperable fire datectors located inside containment will be repaired during the firs t scheduled outage following discovery. Safety reis id fire detection instruments are listed in Table TS.3.14.1.
The-fire suppression water system is supplied fems the Missis,sippi River by two horizottal contrifugal tire pu=ps rated at 2000 gpm at 120 psig. One pump is =otor driven and the other pu=p is diesel driven. A third pump also rated at 2000 gpm at 120 psig, is assigned to the screen wash system, and serves as a backup to the fire suppression water system.
Header pressure is =aintained between 108 and 113 psig ly a jockey fire pump.
If the water demand is such that the jockey pe=p cannot =aintain the header pressure, the screen wash pu=p will start (if not running) and the screen vash to fire hender bypass valve will open at 102 psig.
The bypass line is orificed to restrict flow to 450 g;=.
On further demand, the =ocer driven fire pu=p will autr=acically start at 95 psig.
If further demand of water is called for and the header pressure drops to 90 psig, the diesel driven fire pu=p will s tart. Pumps are designed to pump 2000 gpm and maintain a
=ini=um of 65 psig in the fire header, =easured at the highest point in the system. The screen wash pu=p may be directly aligned to the fire header by
=anual action frem the control roes.
Any one fire pump, or the screen wash pu=p, can be used to supply all fire fighting water requirements.
In the event that a pu=p is inoperable, up to seven days are allowed to restore I
the pump to operability or a report must be submitted to the Commission explaining the circu= stances.
If all pumps are inoperable, or if the fire suppression water system is incapable of supplying water to a safety I
related area, a backup fire suppression water system =ust be established l
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the Come.ission must be informed.
i The cooling water system, also supplied by the Mississippi River, provides additional radundancy to the fire suppression water system.
Crossover water supplies from the cooling water system to the fire protection system are provided for the safety related areas.
Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Ame-dment No. 20, 43 i
L
e TS.3.14-6 Basis (continued)
- 4ater deluge or vet pipe sprinkler systens are provided in safety related areas where a significant fire hazard exists, except for the relay and cable spreading room.
Due to the nature of the equipment in.the relay and cable spreading area, a carbon dioxide system is provided. Whenever a deluge or sprinkler system is inoperable, a continuous fire watch with backup fire suppression equipment available is stationed in the area until operability is restored. Whenever the relay and cable spreading room carbon dioxide systems become inoperable, up to 14 days are allowed to complete
=aintenance.
If the system cannot be restored to operable status within this time period, a report outlining the situation is submitted to the Commission.
Whenever the carbon dioxide system is ineperable, a continuous fire watch with backup fire suppression equipment is sentioned in the room.
Since the relay and cable spreadtng area is occupied during normal working hours, the aute=stic initiation feature of the CO system is bypassed 2
during this period and whenever entry is Lade during other times.
The system is initiated manually in the event fire is detected when the room is occupied.
In addition to deluge and sprinkler systems, hydrant hose houses are located in the yard and hose stations are located throughout the plant.
These hose stations p'rovided primary and backup protection for safety related syste=s and cceponents.
Normally all yard hydrant hose houses and hose stations are operable when a reactor is above cold shutdown.
If a hose hcuse er stati:n pectacting safety : 41ated equipment beccces inoperable, additional hose cust be available for routing to the unprotected area.
This hose =av be sunolied fres an operable hydrant hose house, hose s:stion, or brigade locker.
Piping and electrical penetrations are provided with seals where required by the fire neverity.
If a seal is =ade or found to be inoperable for any reason, the penetration area is continucusly attended until an ef fective fire seal is restored.
Seals have been qualified for the maximum fire severity present on either side of the barrier.
Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43 l-
TABLE TS.3.14-1 (Pg 1 of 3)
TABLE TS.3.14-1 SAFETY RELATED FIRE DETECTION INSTRUMENTS MINIMUM TOTAL NO.
ZONE NO.
LOCATION TYPE OF DETECTOR NO REQUIRED INSTALLED-1 Battery Rooms Ion 2
2 2
Air compressor &
- Ion, 2
9 Auxiliary Feed Thermal 0
3 Pump Area 6
D-2 Diesel Generator
- Ion, 2
3 Room Flame 0
1 8
Auxiliary Building,
- Ion, 10 46 Unit No. 1, Ground Smoke 0
1 Floor Thermal 0
2 10 Reactor. Building *
- Ion, 2
18 Unit No. 1, Ground Smoke 0
1
~
Floor 11 Bus 15 & 16 Switch-Ion 2
6 gear Rooms i
12 Relay & Cable Ion 8
17 Spreading Room Ion 2
4 IL Computer Room 19 Auxiliary Building, Unit No. 1, Mezzanine Ion 5
31 20 Reactor 3uilding, Ion 4
15 Unit No. 1, Mezzanine 21 Reactor Building,
- Ion, 2
12 Unit No. 1, Annulus Flame 0
4 Mezzanine 26 3us 110 & 120 Switch-Ion 2
2 gear Rooms l
23 Auxiliary 3uilding, Ion 2
15 Unit No. 1, Operating Floor i
29 Reactor Building, Ion 2
14 Unit No. 1, Operating Floor Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43 I
TABLE TS.3.14-1 (pg 2 of 3),
TABLE TS.3.14-1 (CONTINUED)
SAFETY RELATED FIRE DETECTION INSTRUMENTS
_Z_0NE NO.
LOCATION MINIMUM TOTAL NO.
TYPE OF DETECTOR NO REQUIRED INSTALLED 30 Auxiliary Building, Ion 7
28 Unit No. 1, Fan Deck 31 Control Room Chiller Ion 2
6 Unit Room 32 Reactor Building, Ion 2
4 t
Unit No.1, Fan Floor 33 Spent Fuel Handling Ion 4
13 Area 35 3attery Rooms Ion 2
2 40 Auxiliary Building.
Ion 5
14 Unit No. 2, Ground Floor 42 Reactor Building,
- Ion, 2
16 Unit No. 2, Ground Smoke 0
1 Floor 43 Bus 25 & 26 Switch-Ion 2
6 gear Roces 46 Auxiliary 3uilding, Icn 5
22 Unic No. 2, Mezzanine 47 Reactor 3uilding, Ion 2
12 Unit No. 2, Annulus, Flame 0
4 Meazanine 50 3us 210 & 220 Svitch-Ion 2
2 gear Roems 51 Auxiliary Building Ion 1
10 Unit No. 2, Operating Floor 52 Reactor Building, Ion 3
14 Unit No. 2, Operating Ficor 33 Auriliary 3uilding, Ion 3
23 Unit No. 2, Fan Dech 54 Reactor 3uilding, Ion 2
4 Unit No. 2, Fan Deck 36 Reactor Buildine Ion 4
15 Unit No. 2. Mezzanine Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43
TABLE TS.3.14-1 (pg 3 of 2)
MINIMUM TOTAL NO.
ZONE NO.
LOCATION TYPE OF DETECTOR NO REQUIRED INSTALLED 57 Control Room Ion 7.
30 1
11 74 Screenhouse, Ground Floor Ion 75 Screenhouse, Operating Floor Ion 2
20 82 D-1 Diesel Generator Room
- Ion, 2
3 Flame 0
1 j
l l
{
i l
I r
l Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43 l
= ~
TABLE TS.4;l-1 (Page
- of 5)
~
MINIItUM FREQUENCIES FOR CllECKS, CALIBRATIONS AND y
TEST OF IllSTRU!!ENT CHANNELS gg 47 Es Channel Functional
Response
g Description Check Calibrate Test Test Remarks 5 m EE
- 1) Once/shif t when in service 5 S. 1.
Nuclear Power S(l)
D(2)
M(3)
R
- 2) Heat balance g:g:
Range M(4)
Q(4)
M(5)-
- 3) Signal toaT; bistable action (permissive, rod stop, trips),
!!(6) pp M(7) with the exception of the Nr items covered in Remark #7.
- 4) Upper and lower chambers for axial of f-set using in-core detectors
- 5) Simulated signal for testing positive and negative rate bistable action
- 6) Quadrant Power Tilt Monitor i
- 7) P8 and P10 permissives and the 25% High Flux Low Setpoint Trip.
2.
Nuclear Inter-
- S(l)
NA T(2)
R
- 1) Once/ shift when in service mediate Range
- 2) Log Level; bistable action (permissive, rod s' top, trips) 3.
Nuclear Source
- S(l)
NA T(2)
R
- 1) Once/ shift when in service Range
[
- 2) Bistable action (alarm, trips) l 4.
Reactor Coolant S(1,2)
R(1,2,3)
M(1)
R(1)
- 1) Ovartemperature AT y
Temperature H(2)
R(2)
- 2) Overpower AT g!
@g T(3)
- 3) Control Rod Bans Insertion p!
@g Limit Monitor M
] g 5.
R M
NA hj l
' n Flow p
t g y 6.
Pressurizer S
R H
NA Water Level g;g; 7.
Pressurizer S
R H
NA
- y Pressure n a p.
W M3 o
Me
TAl;l,E TS.4.1-1 (Page 5 of 5)
Channel Functional
Response
Descriptinn Ched Calihrate Test Test Remarks II u u gg
.35.
Event Monitoring M
lt NA NA Includes all those in FSAR g-p Instrumentation Tabic 7.7-2 and Table TS.3.15-1 not included cicewhere in this YY Table E7 o u o.
o.
36.
Steam Exclusion W
R H
NA See FSAR Appendix I, Section yp Actuation System 1.14.6 r
r.
n n 37.
Pressurize. PORV NA R
H NA Instrument Channels for PORV Control Control Including Overpresiure Mitigation System 38.
Degraded Voltage NA R
H NA 4CV Saf eguard liusses 39.
I,ons of Voltage NA R
M NA 4KV Safeguard liusses S
Each Shift I
D Daily' I$
+4 m m W
Weekly M
Monthly Q
-Quarterly hu 2 y o
o' R
Each refueling shutdown a
P Prior to each startup if not,done previous week P
gg n
o Prior to each startup following uhutdown in excess of 2 days if not done in the previous 30 days y
T oa NA Not Applicable
, U' m -
- v. u o
we See Specification 4.1.1) m
TABLE TS.4.1-2,A MINIMUM FREQUENCIES FOR EQUIPMENT TESTS FSAR Section Test Frequency Reference 1.
Centrol Rod Assemblies Rod drop times All rods during each 7
of full length refueling shutdown rods or following each removal o' the reac-tor vessel head; affected rods follow-ing maintenance on or modification to the control ro<! drive system which could affect performance of those specific rods la. Reactor Trip Breakers Open trip Monthly 2.
Control Rod Assemblies Partial move-Every 2 weeks 7
ment of all rods
- 3. - Pressurizer Saf ety Set point Each refueling 4
Valves shutdown d
4.
Main Steam Safety Sec point Each refueling 10 Valves shutdown 5
Reactor Cavity h'ater level Prior to moving fuel assemblies or control rods and at least once every day whfle the cavity is flooded.
6.
Pressurizer PORV Functional Quarterly Block Valves 7.
Pressurizer PORV's Functional Every 18 months 8.
(Deleted) 9.
Primary System Leakage Evaluate Daily 4
10.
(Deleted) 11.
Turbine stop valves, Functional Monthly 10 governor valves, and intercept valves. (Part of turbine overspeed p ro t ec tion. )
12.
(Deleted)
Prairie Island Unit 1 Amendment No. 26, 47,49 Prairie Island Unit 2 Amendment No. 20, 47, 43
TS.4.4-4 B.
Emergency Charcoal Filter Svstems 1.
Periodic tests of the shield building ventilation s3 - tem shall be performed at quarterly intarvals to demo.otrate operability.
Each redundant train shall be initiated from the control rocm l
and determined to be operable at the time of ire periodic test if-it meets drawdown performance computed,far the test conditions with 75% of the shield building inleakage specified in Figure TS 4.4-1 after initiation.
2.
Periodic tests of the auxiliary building special ventilation system shall be performed at approximateJy quarterly intervals to demonstrate its operability. Eac:t redundant train shall be initiated from the control room and determined to be operable at j
the time of periodic test if it isolates the normal ventilation system and produces a measureable negative pressure in the ABcVZ within 6 minutes after initiation.
3.
At least once per operating cycle, or once each 18 months, which-ever comes first, tests of the filter units in the Shield Building Ventilation System and the Auxiliary Building Special Ventilation System shall be performed as' indicated balow:
The pressure drop across the coEbined HEP filters and c.
the charcoal adsorbers shall be demonstrated to be less than 6 inches of water ar. system design flow rate (+10%).
b.
The inlet heaters and associated controls for each train shall be determined to be operab'e.
c.
Verify that each train of each ver.cilation system. automatically starts on a simulated signal of safety injection and high radiation (Auxiliary Building Special Ventilation only).
4.
a.
The tests of Specification 3.6.E.2 shall be performed at lease once per operating cycle, or once every 18 months whichever occurs first, or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or following painting, fire or chemical release in any ventilation zone communicating with the system that could contaminate the HEPA filters or charcoal adsorbers.
Prairie Island Unit 1 Amendment No. 23,49 Prairie Island Unit 2 Amendment he. 19, 43 1
TS.4.5-2 3.
Containment Fan Coolers Each fan cooler unit shall be tested during each reactor refueling shutdown to verify proper operation of all essential features including low motor speed, cooling water valves, and normal ventilation system dampers.
Individual unit performance will be monitored by observing the terminal temperatures of the fan coil unit and by verifying a cooling water flow rate of greater than or equal to 900 gpm to each fan coil unit.
4.
Component Cooling Water System a.
System tests shall be performed during each reactor refueling shutdown.
Operation of the system will be initiated by tripping the actuation instrumentation.
b.
The test will be considered satisfactory if control board indica-tion and visual observations indicate that all components have operated satisfactorily.
5.
Cooline Water System a.
System tests shall be~ performed at each refueling shutdown. Tests shall consist of an automatic start,of each diesel engine and
^
automatic operation of valves required to mitigate accidants including those valves that isolate non-essential equipment from the system. Operation of the system will be initiated by a simulated accident signal to the actuation instrumentation. The tests will be considered satisfactory if centrol board indication and visual observations indicate that all components have operated satisfactorily anc if cooling water flow paths required for accident mitigation have been establis*_ed.
b.
Each diesel engine shall be inspected at each refueling shutdown.
S.
Comoonent Te igs 1.
Pumps e
a.
The safety injection pumps, residual heat removal pumps and contain-ment spray pumps shall be started and operated at intervals of one month. Acceptable levels of performance shall be th'at the pumps start and reach their required developed heat on minimum recircula-tion flow and the control board. indications and visual ooservations
[
' indicate that the pumps are operating properly for at least 15 minutes.
b.
A test consisting of a manually-initiated start of each diesel engine, and assumption of load within one minute, shall be conducted monthly.
Prairie Island Unit 1 Amendment No. 49 Prairie Island Unit 2 Amendment No. 43 1
i
-~
O O
e TS.4.5-3A h.
Following completion of high head safety injection system or RHR system modifications that alter system flow characteristics a flow balance test shall be performed during shutdown to confirm the following injection flow rates are achieved:
1.
High Head Safety Injection System:
(a) Flow through all four injection lines plus miniflow shall not exceed 835 gpm with one pump in operation.
(b) The minimum flow through loop A & B cold legs shall be 670 gpm with one pump in operation.
The flow rates in each leg shall be within 20 gpm of each other with one pump in operation.
(c) Flow orifices and throttling valves will be used to limit and balance flow through the reactor vessel injection lines to a maximum of the total flow limit in Specification 4.5.B.3.h.1.(a) above, with one pump in ope rat ion. During this flow test the flow rates in each leg shall be within 50 gpm of each other.
2.
RHR System:
The minimum flow through each RHR Reactor Vessel Injection line shall be at least 1800 gpm.
Basis The Safety Injection System and the Containment Spray System are principal plant Safety Systems that are normally inoperative during reactor operation.
Complete systems tests cannot be performed when the reactor is operating-because a safetf injection signal causes containment isolation and a Containment Spray System test requires the system to be temporarily disabled.
The method of assuring operability of these systems is therefore to combine systems tests to be performed during refueling shutdowns, with more frequent component tests which can be performed during reactor operation.
The systems tests demonstrate proper automatic operation of the Safety Injection and Containment Spray Systems. With the pumps blocked from starting, a test signal is applied to initiate automatic action and verifica-tion made that the components receive the safety injection in the proper sequence.
The t :st demonstrates the operation of the valves, pump circuit b re ake rs, and automatic circuitry.
Prairie Island Unit 1 Amendment No.30, 49 Prairie Island Unit 2 Amendment No. 24. 43
i l
TS.4.6-1 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM Applicability Applies to periodic testing and surveillance requirements of the emergency power system.
Objective To verify that the emergency power sources and equipment are operable.
Specification The following tests and surveillance shall be performed:
A.
Diesel Generators 1.
At least once each month, for each diesel generator:
a.
Verify the fuel level in the day and engine-mounted tank.
b.
Verif.y the fuel level in the fuel storage tank.
Verify that a sample of diesel fuel from the fuel c.
storage tank is wichin the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water, and sediment.
d.
Verify the fuel trans fer pump can be started and transfers fuel from the storage system to the day tank.
e.
Verify the diesel starts from the normal standby condition.
f.
Verify the generator synchronizes, is loaded to at least 1375 kw, and operated for at least one hour.
2.
At least once each 18 months:
a.
Subject each diesel generator to a thorough inspection in accordance with procedures prepared in conjunction with the manufacturer's recommendati ns for this class of standby
- service, b.
For each unit, simulate a loss of of fsite power in con-junction with a safety injection signal, and:
1.
Verify de-energization of the emergency busses and load shedding from the emergency busses.
2.
Verify the diesels start from the normal standby con-dition on the auto-start signal and energize the emergency busses in one minute.
3.
Verify that the diesel generator system trips, except those for engine overspeed, ground fault, and generator differential current, are sutomatically bypassed.
4.
Verify that the auto-connected loads do not exceed 3000 kw.
c.
Verify the capability of each generator to operate at least one hour while loaded to 3000 kw.
d.
Verify the capability of each generator to reject a load of at least 650 kw without tripping.
e.
During this test, operation of the emergency lighting system shall be ascertained.
Prairie Island Unit' 1 Amendment No. 25, 49 Frairic Island Unit 2 Amendment No.19, 43
TS.4.16-2 b.
The motor-driven fire pump shall be started every month and run for at least 15 minutes on recirculation flow.
The diesel-driven fire pump shall be started every. month c.
from ambient conditions and run for at least 20 minutes on recirculation flow.
d.
The level in the diesel-driven fire pump fuel storage tank shall'be checked every month and verified to contain at least 500 gallons of fuel.
e.
Every three months verify that a sample of fuel from the diesel-driven fire pump fuel storage tank is within the accept-able limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water and sediment.
f.
Every 18 months subject the diesel-driven fire pu=p engine to an inspection in accordance with procedures prepared in conjunc-tion with the manuf acturer's recemmendations for this class of standby,se rvice. -
g.
A simulated automatic actuation-of each fire pump and the screen wash pump, including verification of pump capability, shall be conducted every 18 months.
~
h.
The header system shall be flushed every 12 months.
i.
System flow tests shall be perfor=ed every three years.
- j. Valves in flow paths supplying fire suppression water to safety related stru'etures, systems, and ce=ponents shall be cycled every 12 months.
k.
Each valve (=anual, power operated, or automatic) in the flow path c c safety-related areas and areas posing a fire ha:ard to safety-releted areas, shall be verified to be in its correct position e;ery =onth and the =ethod of securing the valve in its correct position shall be verified every month.
l 2.
When it is determined that one of the fire pumps required by specification 3.14.B.I.a is inoperable, the remaining operable pumps shall be run daily for at least 15 minutes (motor driven pumps) or 20 minutes (diesel driven pump) until specification 3.14.B.l.a can be met.
Prairie Island Unit 1 Amendment No. 39,49 Prairie Island Unit 2 Amendment No. 33,43
TS.4.16-3 C.
Soray and Sorinkler Systems Each spray and sprinkler system specified in 3.14.C.1 shall be demonstrated operable by performing a nozzle inspection and system functional test, which includes simulated automatic actuation of the system, every 18 sonths.
D.
Carbon Dioxide System The relay and cable spreading room carbon dioxide system shall be demonstrated operable by the following actions:
1.
Verify CO se rage tank level and pressure every week.
2 2.
Verify that the system is operable by performing a aystem functional test which includes simulated automatic actuation of the system every 18 months and'a puff test every three years.
E.
Fire Hose Stations
,The fire hose stations specified in 3.14.E.1 shall be demonstrated oper-
~
able as follows:
l.
Each =onth a visual inspection shall be conducted to assure all eculpment is available.
2.
Every 18 months the hose shall be removed for inspection and re-racking and all gaskets in the couplings shall be incpected and replaced if necessary.
3.
Every three years, partially open each hose station valve to verify valve operability and no blockage.
4 Every three years each hose shal be hydrostat.cally tested at a i
pressure at least 50* psig greate; than the maximum pressure available at that hose station.
Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43
\\
i TS.4.16-4 F.
Yard Hvdrant Hose Houses The yard hydrant hose houses specified in 3.14.F.1 shall be demonstrated operable as follows:
1.
Each month a visual inspection shall be conducted of the yard hydrant hose houses to assure all required equipment is available.
2.
Every six months (in the spring and fall) visually inspect each yard fire hydrant and verify that the hydrant barrel is dry and that the hydrant is not da= aged.
3.
Eva ry year conduct a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any yard hydrant hose house and conduct an inspection of all gaskets in the
, couplings. All degraded gaskets shall be replaced.
G.
Penetration Fire Barriers Fenetration fire barfiers 'in fire' area boundaries protecting safety
.related equipment shall be demons trated_ operable as, follows :
1.
A visual inspection of fire barrier penetration fire barriers shall be conducted every 18 months.
2.
Folleving repair or =aintenance of a penetration fire barrier a visual inspection of the seal shall be conducted.
I i
i i
i l
l Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No. 20, 43 l
l
TS.4.16-5 Basis The minimum number of fire detectors required to be operable in each fire zone are functionally tested following the manufacturer's recommendatiens each six months, axcept for those located inside the primary containment which are tested during each cold shutdown exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 'unless performed during the previous six months. These tests are performed by the plant staff.
Other fire detectors will be tested at an interval which experience has shown to be necessary to assure reliable operation. Every six months an alarm circuit check is performed. This check can be performed in conjunction with, detector functional tests. All circuitry is also provided with automatic supervision for opens and ground faults.
Fire pumps are tested each month to verify operability. Test starting of the screen wash pump is not required since it is normally in service. Each fire pump is manually started and operated for at least 15 or 20 minutes with l
pump flow directed througn the recirculation test line. Every 18 months the cperability of the automatic actuation logic for the fire pumps and the screen wash pump is verified and the performance of each pump is verified to meet system requirements. The specified flush and valve lineup check provide assurance that the piping system is capable of supplying fire suppression water tb all safety related areas. When one of the pumps is inoperable the operab.le pumps are run daily. co verify._ operability until all pumps are once again available.
Fire suppression water system flow tests will be done at least every three years to verify hydraulic performance.
The testing will be performed using Sectica 11, Chapter 5 of the Fire Protection Handbook, 14th Edition, as a procedural guide. The test is generally performed in conjunction with l
insurance inspections.
l.
Surveillance specified for each spray and sprinkler system is intended to assure that the systems will function as designed when they are needed.
Functional tests are conducted at 18 month intervals on those systems' provided with test facilities.
The testing specified for the relay and cable spreading room CO, system
-provides assurance that the CO invent ry is adequate to extingGish a 2
fire in this area and that the system is capable of automatic actustion.
Hose stations and yard hydrant hose houses are inspected monthly to verify that all required equipment is in place. Gaskets in hose couplings are inspected periodically and the hose is pressure tested. Pressure testing of outdoor hose is conducted more frequently than indoor hose because of the less favorable storage conditions. Operability of hose station isolation valves is verified every three years by partially opening each valve to verify flow.
All of these tests provide a high degree of assurance that each hose station will perform satisfactorily after periods of standby service.
Prairie Island Unit 1 Amendment No. 26, 49 Prairie Island Unit 2 Amendment No.-20, 43
TS.A.16-6 Plant fire barrier walls are provided with seals for pipes and cables where necessary. Where such seals 'are installed, they =ust be maintained intact to perform their function. Visual inspection of each-installed seal is required every 18 months and af ter seal repair. A visual' inspection following repair of a seal in the secondary containment boundary is sufficient to assure that seal leakage will be within acceptable limits.
J I
1 Prairie Island Unit 1 A=enament No. 26, 49 43 Prairie Island Unit 2 Amendment No. 20,
,. -..~,
l TS.6.1-1 6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION A.
The Plant Manager has the overall full-time onsite responsibility for safe operation of the f acility. During periods when the Plant Manager is unavailable, he may delegate this responsibiliity to other qualified supervisory personnel.
B.
The Northern States Power corporate organizational structure relating to the operation of this plant is shown on Figure TS.S.1-1.
C.
The functional organization for operation of the plant shall be as shown in Figure TS.6.1-2 and:
l 1.
Each on duty shif t shall be composed of at least the minimum shift crew composition shown on Table TS 6.1-1, 2.
For each reactor that contains fuel:
a licensed operator in the control room.
3.
At least two licensed operators shall be present in the control room during a reactor startup, a scheduled reactor shutdown, and during recovery from a reactor trip.
These operators are in addition to those required for the other reactor.
4 An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor.
5.
All refueling operations shall be directly supervised by a licensed Senior Reactor Operator of Senior Reactor Operator Limited to Fuel Handling who has no other concurrent respons-ibilities during this operation.
6.
A fire brigade of at least five members shall be maintained on site at all t imes.*
The fire brigade shall not include the six members of the minimum shif t crew for safe shutdown of the reactors.
- Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of Fire Brigade members provided irrediate action is taken to restore the Fire Brigade to within the minimum requirements.
Prairie Island Unit 1 Amendment No. 39, 49 Prairie Island Unit 2 Amendment No. 33, 43
PRE!ilD%!lT NN E5 77 SliNIOR VICE PRr:0II)EITIS gg POWER SUPPI.Y g
r ca to S.R.
??
RR I
uw VICE PRESIDENT VICE PRESIDEi!T ii PIAtlT E!!GIllEERIN0 PCWER PRODUCTION <
g.g AND CONSTRUCTION e o S.E.
RR EE
.P P GENERAL MA!!AGER GENERAL GENERAL MANAGER ww*
NUCLEAR PIAllTS SUPERINTENDENT HEADQUARTERS OPERATIONAL QA NUCLEAR SERVICES l
,y we I
I PIANT MAllAGERS TRAINING MANAGER liUCLEAR PIAllTS 4-i
. MANAGER NUCLEAR SUPPORT SERVICES y
a
- lias the responsibility for the l
SAC 11NNISTRATION fire protection program g
g i
SAFETY AUDIT h
L, I
COMMITTEE (SAC) 8 b
I I
AUDIT AIID L_.
REVILV FIGURE TS.6.1 I NSP CORPORATE ORGA!!I7.ATIONAL RETATIONSilIP TO ON-SITE OPEi!ATIllG ORGAtlIZATION
4
/
OPERATIONS Pl. ANT q
MANAGER
- C0!!!!ITTEE NN I
a mo EE
.-r
- _ BEV11M AND _AUIMT -.
m*
pg TT oa Pl. ANT SilPERINTENDENT SUPT.
PLANT SUPERINTENDENT OPERATIONS &
QUALITY ENGINEERING &
ao NAINTENANCE*+
ENGINEEKlNG*
RADIATION PROTECTION *(LS0) rr nn wr SUPPORT FOR QA SUPERVISOR OF AND QC FUllCTIOli SECURITY & SERVICES
[
l I
[
SUPT. OF SUPT. OF SUPT.
SUPT'.
SUPT.
SENIOR tlA I NTENANCEA OPERATIONS RAD 1ATION TECilNICAL OPERATIONS NUCLEAR
- (LS0)
PROTECTION
- ENGINEERING
- ENGINEERING
- ENGINEER
- I SillFT SUPER-VISOR (LS0)
I MECil AN ICisl, LEAD Pl ANT EQLilP TEPilNICAI.
TECilNICAL l'ECllNICAL TECilNICAL MEi4T & REACTOR SUPPORT h SUPPORT &
iUPPORT &
SUPPORT &
ELECTRICAL OPERATOR (LO)
RADIATION ENGINEERS.'OR ENGINEERS FOR ENGINEERS flAINTENANCE PROTECTION INSTRUMENTS, JPERATION,
FOR NUCLEAR gg GROUP SPECIAl,lSTS CONTROLS, &
'tAIN'I ENANCE, ENGINEERING y
gg Pl. ANT EQUIPMENT
& REACTOR
. CollPUTER; 3URVEII. LANCE, g
INSTRUMENT &
& TESTING g
OPERATOR (S)(1.0)
CONTROLS SPEC g
n n
% 'A ASSISTANT Pl. ANT m
EQUIPl!ENT FIGURE TS'.6.1-2 PRAIRIE ISLAND NUCLEAR GENERATING PIANT
'i' ww
??
OPERATOR (S)
FUNCTIONAL ORGANIZATION FOR ON-SITE GROUP AND 33 PI. ANT ATTENDANTS Key Supervisor I.0 Licenr.ed Oper* tor FIRE 11RICADE (AS REQUIRED)
LSO Licensed Senior Operator
+
lias responsibility for impicmentation of the fire protection prcgram
TS.6.2-1 6.2 Review and Audit Organizational units for the review and audit of' facility operations shall be constituted and have the responsibilities and authorities outlined below:
A.
Safety Audit Committee'(SACl The Safety Audit-Committee provides theLindependent review of plant operations from a nuclear safety standpoint. Audits of plant. operation are conducted under the cognizance of the SAC.
1.
Membership a.
The SAC shall consist of at least five (5) persons.
b.
The SAC chairman shall be an NSP representative, not having line responsibility for plant operation, appointed by.the Vice President - Power Production. Other SAC members shall be appointed by the Vice President - Power Eroduction or by such orb.er person as he may' designate. The Chairman shall appoint a Vice Chairman'from the SAC membership to act in
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his absence.
c.
No more than two members of the SAC shall be from groups holding line responsibility for operation of the plant.
d.
A SAC member may appoint an alternate to serve in his absence, with concurrence of'the Chairman. No more than one alternate shall serve en the SAC at any one time.
The alternate member shall have voting rights.
2.
Qualifications a.
The SAC members should collectively have the capability required to review activit'ies in the following areas:
nuclear power plant operations, nuclear. engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and-other appro-priate fields associated with the unique' characteristics of the nuclear power plant.'
Prairie Island Unit 1 Amendment No.13,.49 Prairie Island Unit 2 Amendment No. 7, 43
&a TS.6.2-3 f.
Investigation of all events which are required by regula-tion or technical specifications (Appendix A) to be reported to NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, g.
Revisions to the Facility Emergency Plan, Facility Security Plan, and the Fire Protection Program, h.
Operations Committee minutes to determine if matters con-sidered by that Committee involve unreviewed or unresolved safety questions.
i.
Other nuclear safety matters referred to tLe SAC by the Operations Committee, plant management or company manage-ment.
- j. All recognized indications of an unanticipated deficiency in 4
some aspect of design or operation of safety-related structures systems, or components.
k.
Reports of special inspections and audits conducted in accor-dance with specification 6.3.
Audit - Th'e operation of the nuelear power p'lant shall be audited
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6.
formally under the cognizance of the SAC to assure safe facility operation.
a.
Audits of selected aspects of plant operation, as delineated in Paragraph 4.4 of ANSI N18.7-1972, shall be performed with a frequency commensurate with their nuclear safety significance and *n a manner to assure that an audit of all nuclear safety-related activities is completed within a period of two years.
The audits shall be performed in I
accordance with appropriate written instructions and procelures.
b.
Periodic review of the audit program should be performed by the SAC at least twice a year to assure its adequacy.
Written reports of the audits shall be reviewed by the Vice c.
President - Power Production, by the SAC at a scheduled meeting, and by members of management having responsibility in the areas audited.
7.
Authority The SAC shall be advisory to the Vice President - Jower Production.
8.
Records Minutes shall be prepared and retained for all scheduled meetings of the Safety Audit Co=mittee. The minutes shall be distributed within one month of the meeting to the Vice President - Power Production, the General Manager, Nuclear Plants each member of the SAC and others designated by the Chairman. There shall be a for=al approval of the minutes.
Prairie Island Unit 1 Amendment No. If, 49 Prairie Island Unit 2 Amend:ent No. 20,22
TS.6.2-5 B.
Operations Committee (OC) 1.
Membership The Operations Committee shall consist of at least six (6) members drawn from the key supervisors of the onsite staf f.
The Plant Manager shall serve as Chairman of the OC and shall appoint a Vice Chairman from the OC membership to act in his absence.
2.
Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.
3.
Quorum A majority of the permanent members, including the Chairman or Vice Chairman.
4 Responsibilities - The following subjects shall be reviewed by the Operations Committee:
a.
Proposed tests and experiments and their results, b.
Modifications to plant systems or equipment as described in the Final Safety Analysis Report and having nuclear safety significance or which involve an unreviewed safety question as defined in Paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations, c.
Proposals which would ef fe.:t permanent changes to normal and emergency operating procedures and any other proposed changes or procedures that will affect nuclear safety as determined by the Plant Ma.ager.
d.
Proposed changes to the Technical Specifications or operating
- licenses, e.
All reported or suspected violations of Technical Specifica-tions, operating license requirements, administrative procedures, operating precedures. Results of investigations,.
including evaluation and re.,mmendations to prevent recurrence will be reported in writing to the General Manager - Nuclear Plants and to the Chairman of the Safety Audit Committee.
Prairie Island Unit 1 Amendment No. 49 Prairie Island Unit 2 Amendment No. 43 L
TS.6.2-6 f.
All eve - chich are requireo by regulations or Technical Spec.tications to be reported to the NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
g.
Drills on emergency procedures (including plant evacuation) and adequacy of communication with offsite support groups.
h.
All procedures required by these Technical Specifications, including implementing procedures of the Emergency Flan, and the Security Plan, shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two years.
i.
Special reviews and investigations, as requested by the Safety Audit Committee.
5.
Authority The OC shall be advisory to the Plant Manager.
In the event of a disagreement between the recommendations._of the OC and the Plant Mandger, the course determined by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the General Manager and the Chairman of the SAC for review.
6.
decercs Minutes shall be recorded for all meetings of the 00 and shall identify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committe, the General '!anager Nuclear Plants and others designated by the OC Chairman or Vice Chairman.
7.
Proc <Jures A written charter for the OC shall be prepared that contains:
a.
Responsibility and authority of the group b.
Content and method of submission of presentations to the Operations Committee-c.
Mechanism for scheduling meetings d.
Provision for meeting agenda Prairie Island Unit 1 A=endment No.9, 49 Prairie Island Unit 2 Amendment No. 4, 4 3
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TS.6.4-1 6.4 SAFE 7 LIMIT VIOLATION If a safety limit is exceeded, the reactor shall be shut down and,:he Commission shall be notified immediately.
It shall also be promt ly reported to the General Manager Nuclear Plants and the. Chairman 22 the Safety Audit Committee, or their designated alternates. A safety limit violation report shall be prepared. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) the basis for corrective action taken to preclude recurrence. The report shall be reviewed by the Operations Committee. The safety limit viola-tion report shall be submitted to the Commission, the General Manager Nuclear Plants;and the:. Safety' Audit Committee within two weeks of the event.
Operation shall not be resumed until authorized by the Nuclear j
Regulatory Commission.
I 1
Prairie Island Unit 1 Amendment No. 9, 49 Prairie Island Unit 2 Amendment No. 4, 43
TS.6.5-2 1.
a.
Paragraph 20.203 " Caution signs, labels, signals and controls".
In lieu of the " Control device" or alarm signal required by paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and con 9picuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (or continuous escort by a qualified person for the purpose of making a radiation survey) and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rste in the area.
b.
The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 nRem/hr, except that. doors shall be locked or attended to prevent unauthorized entry into these areas and the keys or key devices for locked doors shall be maintained under the administrative control of the Plant Manager.
2.
A program shall be implemented to reduce leakage from systems outside containment that would or coul! contain highly radioactive fluids during a serious transient or accident to as low as practical le vels.
This program shall. include the following:
a.
Provisions. establishing preventive. maintenance and periodic visual inspection requirements, and b.
Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
A program acceptable to the Commission was described in letters from L 0 Mayer, NSP, to Director of Nuclear Reactor Regulation, dated December 31, 1979 " Lessons Learned I=plecentacien" and Mar:h 13, 1980, "1/1/30 Lessons Learned I=plementation Additional Informacien".
3.
A program shall be i=ple=ented which will ensure the capability to accarately determine the airborne iodice concentration in essential plant areas under accident conditions. This program shall include the following:
a.
Training of personnel, b.
Procedures for monitoring, and d.
Provisions for maintenance of sampling and analysis equipment.
A program acceptable to the Ce=sission was described in letters from L 0 Mayer, NSP, to Director of Nuclear Reactor Regulation, dated Decemoer 31, 1979 " Lessons Learned I=plementation" and Marca 13, 1979, "l/1/80 Lessons Learned Implementation Additional Infor=ation".
Prairie Island Unit 1 AmenCaent No. 25, 46, 49 Prairie Island Unit 2 Amendment No. 19, 40, 43
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