ML19350D189
| ML19350D189 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/01/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19350D190 | List: |
| References | |
| NUDOCS 8104130694 | |
| Download: ML19350D189 (12) | |
Text
9 arq UNITED STATES
[ i NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
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- u.....9 t:0RTHERN STATES POWER COMPANY DOCKET N0. 50-282 PRAIRIE ISLAND NUCLEAR GENERATIl1G PLAf1T UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICEllSE
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Amendment No. 47 License No. DPR-42
~. The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated November 24, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth
. in 10 CFR Chapter I; E.
The facility Will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) th'at the activities attthorized by this amendment can be conducted without endangering the health and safety of 'the public, and (ii) that such activities will be conducted in cocpliance with the Comission's regulations; E.
The issuance of this amendment will ndt be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 47, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION t.
cu t R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: April 1, 1981
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.,..,~,'g UNITED STATES E
ft NUCLEAR REGULATORY COMMISSION g
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C WASWNGTON, D. C. 20555
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NOPTHERN STATES POWER COMPANY DOCKET N0. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 41 License No. OPR-60 l.
The Nuclear Regulatory Comnission (the Commission) has found that:
A. -The application for amendment by Northern States Power Company (the licensee) dated November 24, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Cocmission's rules and regulations set forth in 10 CFR Chapter I; 5.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commissio'n;
- . 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 corpliance with the Com:nission's regulations; J.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Conmission's regulations and all applicable requirements I
have been satisfied.
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, 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendaent, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 41, are hereby incorporated in the license.
The licensee shall operate the. facility in accordance with the Technical Specifications.
3.
This license anendaent is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
.h Cw R. A. Clark, Chief
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Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 1, 1981 a
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ATTACHMENT TO LICENSE AMENDMENTS NOS. 47 AND 41 FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET-NOS. 50-282 AND 50-306 e;1a:e the following pages of the Appendix A Technical Specifications with the a ttached pages..The changed arets on the revised pages are reflected by marginal Tires.-
T5.3.1-1 TS.3.1-1A TS.3.1-3 TS.3.8-1 TS.3.8-2 TS.3.8-4 Table TS.4.1-2A b
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v TS.3.1-1 3.0' LIMITING CONDITIONS FOR OPERATION 3.11 REACTOR COOLANT SYSTEM Applicability Applies to the operating. status of the reactor coolant system.
Obiective To specify those limiting conditions for operation of the reactor coolant system which must be met to assure safe reactor operation.
Specification A.
' Operational Components 1.
Coolant pumps a.
Both reactor coolant pumps shall be operable whenever average reactor coolant system temperature is above 350 F.
b.
Both reactor coolant pumps shall be in operation whenever a reactor is critical, except during low power physics tests.
c.
eat least one reactor coolant pump or one residual heat removal pump shall be in operation at all times. All l
l pumps may be shutdown for up to one hour provided the reactor is subcritical.-no operations are permitted that
.l would cause dilution of the reactor coolant boron j
concentration and core outlet temperature is maintained a
at least 10 F below saturation temperature, j
d.
If a reactor coolant pump becomes inogerable with average reactor coolant temperature above 350 F, restore the pump (s) 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 reduce the g
average reactor coolant temperature below 350 F.
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Dl'R Amendment No. 47 DPR Amendn.ent No. 41 i
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r P00R ORIGINAL
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TS.3.1-1A 2.
Both steam generators shall be operable whenever average reactor coolan:, system temperature is above 350 F.
b.
Ifasteamgeneratorbecomesinoperabgewithaverage reactor coolant temperature ab.ove 350 F, restore the steam generator 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 reduce the average reactor coolant temperature below 350*F.
In the event of excessive primary to secondary leakage take the actions required by Specification 3.1.C.6.
3.
Requirements for Decay Heat Removal Below 350 F lihenever the average reactor coolant temperature is below a.
350 F, except during refueling shutdown with the vessel head unbolted, at least two methods for removing decay heat shall be operable.
Acceptable methods for removing decay heat are an operable steam generator or a residual heat removal loop including pump and associated heat exchanger.
b.
If two methods of removing decay beat are not operable, initiate corrective action immediately to restore the inoperable equipment or be in a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, c.
Prior to draining the reactor coolant level below the reactor vessel flange during maintenance, both residual heat removal loops, each consisting of a pump and associated heat exchanger, shall be operabic. ~ If a residual heat renoval loop becomes inoperable, immediate action shall be taken to restore reactor evolant system level above the vessel flange.
- ?F. Amendment No. 47
- ?? Amendment No. 41
TS.3.1-3 The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following 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 6
350 F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure.
If no residual heat were removed by any of the means available, the amount of steam whic- >ould be generated at safety valve relief pressure would be less than half t.ie 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 both safety valves {s greater than the maximum surge resulting from complete loss of load.
" 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 tests submitted to the staff have demonstrated that tubes having a wall thickness greater than 0.025-inch have adequate margins of safety against failure dug to loads imposed by normal plant operation and design basis accidents.
The Specifications require that at least two methods of removing decay heat are available for each reactor.
Above 350 F, both steam generators must be operable to serve this function.
Below 350 F, either a steam generator or a residual heat removal loop are capable of removing decay heat and any combination of two loops is specified.
If redundant means are not available, the reactor is placed in the cold shutdown condition.
References I FSAR, Section 14.1.9 Testimony by J Knight in the Prairie Island Public Hearing on Janua'ry 28, 1975.
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i DFR Amendment No. 47 DPR Amendment No. 41 i
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TS.3.8-1 3.8 REFUELING AND FUEL HANDLING Aeolicability
- Applies to operating limitations during fuel-handling and refueling opera-tiens.
Objective To ensure that no incident could occur during fuel handling and refueling operations that would af fect public health and safety.
Soecification A.
During refueling operations the following conditions shall be sat is fied :
1.
The equipment hatch and at least one door in each personnel air lock shall be closed.
In addition, at least one isolation valve shall be operable or locked closed in each line which penetrates l
the containment and provides a direct path frem containment atmosphere to the outside.
2.
Radiation levels in fuel handling areas, the containment and the spent fuel storage pool areas shall be monitored continuously.
3.
The core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment, which are in service whenever core gecmetry is l
being changed. When core geometry is not bein; changed, at least one neutron flux monitor shall be in service.
I 4.
During reactor vessel heaa removal and while loading and unloading fuel from the reactor, the minimum boron concentration of 2000 ppm shall be =aintained in the reactor coolant system. The required boron concentration shall be verified by chemical analysis daily.
5.
During movement of fuel assemblies or control rods out of the reactor vessel, at least 23 feet of water shall be maintained above the reactor vessel flange. The required water level shall be verified prior. co moving fuel assemblies or control rods and at least once every day while the cavity is flooded.
6.
At le as t one residual heat removal pump shall be operable and running.
The pump may be shutdown for up to one hour to facilitate movement of fuel or core components.
7.
If the water level above the top of the reactor vessel flange is less than 20 feet, except when the cavity is being drained for head replacement or cer. trol rod 1:tching and anlatching operations, cath residual hect removal loops shall be operable.
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8.
If Specification 3.8.A.6 or 3.8. A.7 cannot be satisfied, all fuel handling operations in containment shall be suspended, the containment integrity requirements of Specification 3.8.A.1 shall be satisfied, and no reduction in reactor coolant boron concentra-l tien shall be made.
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47 O!"'- 'iO - A.*c ndre n L No.
41
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P00R ORIGlNM 9.
Direct communication between the control room and the l
operating floor of the containment shall be available whenever changes in core geometry are taking place.
10.
No movement of irradiated fuel in the reactor shall be l
nade until the reactor has been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, 11.
The radiation monitors which initiate isolation of the l
Containment Purge System shall be tested and verified to be operable immediately prior to a refueling operation.
1.
During fuel handling operations, the following conditions shall be satisfied:
1.
No heavy loads will be transported over or placed in either part of the' spent fuel pool when irradiated fuel is stored in that part.*
2.
Prior to spent fuel handling in the auxiliary building, tests shall be made to determine the operability of the spent fuel pool special ventilation systen including the radiation monitors in the normal ventilation system that actuate the special system and isolate the normal systems.
3.
Prior to fuel handling operations, fuel-handling cranes shall be load-tested for operability of limit switches, interlocks, and alarms.
4.
When the spent fuel cask contains one or more fuel assemblics, it will not be saspended more than 30 feet above any surface until the fuel has decayed more than 90 days.
If any of the specified conditions in 3.8.A or 3.8.B above are not tet, refueling or fuel-handling operations shall cease.
Work shall te initiated to correct the violated conditions so that the specifications are met, and no operations which may increase the reactivity of the core shall be performed.
I t ar the purpose of complet ing the f uel stora);e pool modif ication, the e n '. and placement of loads described in the installatien procedures
- taia modifiention are permitted at, uescribed in the licensee's si.it ta in of Novemhe 24 1976, April 14 and ?/, 1977, and hearing
.t rJ of. lune 14 through 17, 1977.
??. Amendment No. 11, 22, 47
- ?? Amendment No. 11, 16, 41
T3.3.8-4 the cask into a carrier, there is a potential drop of 66 feet (
The cask will not be loaded onto the carrier for shipment prior to a 3-month s torage pe riod. At this time, the radioactivity has decayed so that a release of fission products from all fuel assemblies in the cask would result in of f-site doses less than 10 CFR Part 100.
It is assumed, for this dose analysis that 12 assemblies rupture af ter storage for 90 days.
Other assumptions are the same as those used in the dropped fuel assembly accident in the SER, Section 15.
The resultant doses at the site boundary are 94 Rems to the thyroid and 1 Rem whole body.
Fuel Pool Special Ventilation System (
is a safeguards system The Spent which maintains a negative pressure in the spent fuel enclosure upon detection of high area radiation. The Spent Fuel Pool Normal Ventilation s;.s tem is automatically isolated and exhaust air is drawn thro, ugh filter mocules containing a roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of the Shield Building exh aus t stacks. Two completely redundant trains are provided. The exhaust f an and filter of each train are shared with the corresponding train of the Containment In-service Purge System. High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine absorbers in each SFPSVS filter train. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment.
The in place test results should indicate a HEPA filter leakage of less than 1% through DOP testing and a charcoal adsorber leakage of less than 1% through halogenated hydrocarbon testing. The laboratory carbon sample test results should indicate a. radioactive metnyl iodide recaval efficiency of at least 90% under test conditions which are more severe than accident conditions.
The satisfactory completion of these periodic tests combined with the qualification testing conducted on new filters and adsorber provide a high level of assurance that the emergency air treatment systems will perform as predicted in the accident analyses.
During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provide suf ficient shielding.
The specifications require that at least one res idual heat removal loop be in operation. This assures that sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor below 140 F and that sufficient coolant circulation is =aintained through the core to minimize the effect of a boron dilution incident and prevent baron stratifica-tion.
The requirement to have two residual heat re= oval loops operable when there is less than 20 feet of water above the vessel flange ensures that a single f ailure of the operating loop will not result in a complete loss of residual heat removal capacility. With the reactor vessel head l
removed and 20 feet of water above the vessel flange, a large heat sink is available for core cooling.
In the event of a f ailure of the operating RHR loop, adequate time is provided to initiate repairs or emergency procedures l
to cool the core.
t References a
(1)
FSAR Section 9.5.2 r
(2)
FSAR Table 3.2.1-1 (3)
FSAR Section 14.2.1 (A)
FS AR Section 9.6 L
(5)
FSAR Page 9.5-20a DPR Amendment No. 25, 47 j
DPR Amendment No. 19, 43 l
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TS.4.1-2A MINIMUM FREQUENCIES FOR EOUIPMENT TESTS FSAR Section Test Frecuency Reference Control Rod Assemblies Rod drop times All rods during each 7
of full length refueling shutdown rods or following each removal of the reac-tor vessel head; affected rods following maintainance on or modification to the control rod drive system which could affect performance of those specific rods la. Reactor Trip 3reakers Open trip Monthly 2.
Control Rod Assemblies Partial move-Every 2 weeks 7.
ment of all rods 3.
Pressurizer Safety Set point Each refueling 4
Valve s shutdown Main Steam Safety Set point Each refueling 10 Valves shutdown 5.
Reactor Cavity Water level Prior to moving fuel assemblies or control rods and at least once every day while the cavity
- is flooded.
6.
(Deleted)
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(Deleted) 5.
(Deleted) 9.
Primary System ~eakage Evaluate Daily 4
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' 3. (Dele ted)
.l. Turbine stop valves, Functional Monthly 10 governor valves, and l
intercept valves.
(Part of turbine overspeed protection.)
'2. (Deleted) e
'70~IS :
= See Specification 4.1.D.
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~ ?E Amendment No. J/, /s, 7$, 47
- ?F. Amendment No. 11, 19, ld, 41 i
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