ML20086E210
| ML20086E210 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  (DPR-42-A-119, DPR-60-A-112) |
| Issue date: | 07/03/1995 |
| From: | Wetzel B NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20086E213 | List: |
| References | |
| NUDOCS 9507110274 | |
| Download: ML20086E210 (18) | |
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UNITED STATES
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NUCLEAR REGULATORY. COMMISSION t
WASHINGTON, D.C. 20665 4001 liORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.- 119 License No. DPR-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated December 5, 1994, as supplemented January 9, 1995, and May 15, 1995, complies 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 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 to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been i
satisfied.
I 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:
1 9507110274 950703
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PDR AD00K 05000282 P
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Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 119, 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 issuance, with full implementation within 30 days.
I FOR THE NUCLEAR REGULATORY COMMISSION
{
Ja kk y Beth A. Wetzel, Project Manager Project Directorate III-I Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 3, 1995 i
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ATTACHMENT TO LICENSE AMENDMENT NO.119 FACILITY OPERATING LICENSE NO. DPR-42 i
DOCKET NO. 50-282 l
i Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
l REMOVE INSERT i
TS.3.8-1 TS.3.8-1 l
TS.3.8-2 TS.3.8-2 B.3.8-1 B.3.8-1 B.3.8-2 B.3.8-2 B.3.8-3 B.3.8-3 B.3.8-4 B.3.8-4 i
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TS.3.8-1 3.8 REFUELING AND FUEL HANDLING Applicability Applies to operating limitations associated with fuel-handling operations, CORE ALTERATIONS, and crane operations in the spent fuel pool enclosure.
Obiectives To ensure that no incident could occur during fuel handling, CORE ALTERATIONS and crane operations that would affect public health and safety.
Specification A.
Core Alteratiqng 1.
During CORE ALTERATIONS the following conditions shall be satisfied (except as specified in 3.8.A.2 and 3 below):
a.
- 1) The equipment hatch shall be closed. In addition, at least one isolation valve shall be OPERABLE or locked closed in each line which penetrates the containment and provides a direct path from containment atmosphere to the outside.
- 2) Airlock doors a) At least one door in each air lock is closed, or b) Both doors in each air lock may be open if:
1.
The containment (high flow) purge system is isolated, ii. The inservice (low flow) purge system is capable of l
automatic isolation, fii.At least one door in each air lock is OPERABLE, under procedural control, and capable of being closed within 30 minutes following a fuel handling accident in containment, and iv. At least two containment fan coil unit fans are capable of operating in the high speed mode following a fuel handling accident in containment.
b.
Radiation levels in the fuel handling areas of the containment shall be monitored continuour.ly.
Prairie Island Unit 1 Amendment No. 73,91,119 Prairie Island Unit 2 Amendment No. 66, 84,112 I
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TS.3.8-2 4
3.8.A.1.c.
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 t
the' containment, which are in service whenever core geometry is being changed. When core geometry is not-being changed, at least one neutron flux monitor shall be in service.
d.
The plant shall be in the REFUELING' condition.
e.
During movement of fuel assemblies or control rods out.of the j
reactor vessel, at least 23 feet of water shall be maintained above the reactor vessel flange. The required water level shall be verified prior to moving fuel assamblies or control rods and-at least once every day while the cavity is flooded.
i f.
At least one residual heat removal pump shall be OPERABLE and running. The pump may be shut down for up to one hour to j
facilitate movement of fuel or core components.
g.
If the water level above the top of the reactor vessel flange is l
1ess than 20 feet, except for control rod unlatching / latching operations or upper internals removal / replacement, both residual heat removal loops shall be OPERABLE.
i h.
Direct communication between the control room and the operating floor of the containment shall be available whenever-CORE i
ALTERATIONS are taking place.
1.
No movement of irradiated fuel in the reactor shall be made until j
the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
i
- j. The radiation monitors which initiate isolation of the Containment j
Purge System shall be tested and verified to be OPERABLE prior to l
l
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2.
If any of the above conditions are not met, CORE' ALTERATIONS shall cease. Work shall be 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.
3.
If Specification 3.8. A.1.f or 3.8. A.13 cannot be ' satisfied, all fuel handling operations in containment shall be suspended, the-requirements of Specification 3.8.A.1.a.1) shall be satisfied, at least one door in each personnel air lock shall be closed, and no reduction in reactor coolant boron concentration shall be made.
Prairie Island Unit 1 Amendment No. 73, 7f, 97, gg9 Prairie Island Unit 2 Amendment No. 66,67,54. 112
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B.3.8-1 3.8 REFUELING AND FUEL HANDLING Bases Core alteration containment isolation specifications are provided to minimize releases following a fuel handling accident (FRA). Allowing both airlock doors open during core alterations will facilitate evacuation of containment following a FHA and help maintain the seals in good working order. The FHA does not cause containment pressurization, however, with an assumed single failure the operating purge system supply fan is assumed to continue supplying air to containment. To maintain post-FHA releases well within the limits of 10CFR100, only the inservice purge system is allowed to be operating during core alterations. Two containment fan coil unit fans are required to operate in the high speed mode following a fuel handling accident in containment to assure that radioactive material in containment is well mixed and any releases will leave containment at a lower concentration over the duration of the accident. The provision that one door is OPERABLE and under procedural control will ensure that at least one door will be closed in within 30 minutes as required, thus assuring radioactive releases are well within the limits of 10CFR100.
F The equipment and general procedures to be utilized during refueling are discussed in the FSAR.
Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during CORE ALTERATIONS that would result in a hazard to public health and safety (Reference 1).
Whenever changes are not being made in core geometry, one flux monitor is sufficient.
This permits maintenance of the instrumenta-tion.
Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.
Under rodded and unrodded conditions, the K.tr of the reactor must be less than or equal to 0.95 and the boron concentration must be greater than or equal to 2000 ppm.
Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.
3.8.A.1.h allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel. The delay time is consistent with the fuel handling accident analysis (Reference 2).
Fuel will not be inserted into a spent fuel cask unless a minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that k.tr for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask.
The number of recently discharged assemblies in Pool No. 1 has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.
Prairie Island Unit 1 Amendment No. 91, 99, 119 Proirle Island Unit 2 Amendment No. 84' 92' 112
B.3.8-2 3.8 REFUELING AND FUEL HANDLING Bases continued The Spent Fuel Pool Special Ventilation System (Reference 3) is a safeguards
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system which maintains a negative pressure in the spent fuel enclosure upon detection of high area radiation. The Spent Fuel Fool Normal Ventilation I
System is automatically isolated and exhaust air is drawn through filter modules containing a roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of the Shield Building exhaust stacks. Two completely redundant trains are provided. The exhaust fan 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 adsorbers in each SFPSVS filter train.
The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment.
During movement of irradiated fuel assemblies or control rods, a water J
1evel of 23 feet is maintained to provide sufficient shielding.
The water level may be lowered to the top of the RCCA drive shafts for latching and unlatching. The water level may also be lowered below 20 feet for upper internals removal / replacement. The basis for these allowance (s) are (1) the refueling cavity pool has sufficient level to allow time to initiate repairs or emergency procedures to cool the core, (2) during latching / unlatching and upper internals removal / replacement the level is closely monitored because the activity uses this level as a reference point, (3) the time spent at this level is minimal.
The Prairie Island spent fuel storage racks have been analyzed (Reference 4) to allow for the storage of fuel assemblies with enrichments up to 5.0 weight percent U-235 while maintaining K.tr s 0.95 including uncertainties. This criticality analysis utilized the following storage configurations or regions to ensure that the spent fuel pool will remain suberitical during the storage of fuel assemblies with all possible combinations of burnup and initial enrichment:
1.
The first region utilizes a checkerboard loading pattern to accommodate new or low burnup fuel with a maximum enrichment of 5.0 wet U-235.
This configuration stores " burned" and " fresh" fuel assemblies in a 2x2 checkerboard pattern.
Fuel assemblies stored in " burned" cell locations must have an initial enrichment less than 2.5 wtt U-235 (nominal) or satisfy a minimum burnup requirement. The use of empty cells is also an acceptable option for the " burned" cell locations.
Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0 wtt U-235 with no requirements for burnup or burnable absorbers.
2.
Thi second region does not utilize any special loading pattern.
Fuel aseemblies with burnup and initial enrichments which fall into the unrestricted range of Figure TS.3.8-1 can be stored anywhere in the region with no special placement restrictions.
Fuel assemblies which fall into the restricted range of Figure TS.3.8-1 must be stored in the checkerboard region in accordance with Specification 5.6.A.l.d.
Prairie Island Unit 1 Amendment No. 97, 99, 108, 113 Prairie Island Unit 2 Amendment No. 84, 92, 101, 112
B.3.8-3
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3.8 ' REFUELING AND FUEL HANDLING Agggg continued The burned / fresh fuel checkerboard region _ can be positioned anywhere within the i
spent fuel racks, but the boundary between the checkerboard region and the
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unrestricted region must be either, 7
i 1.
separated by a vacant row of cells, or 2.
the interface must be configured such that there is one row carryover of the pattern of burned assemblies from the checkerboard region into the j
first row of the unrestricted region (Figure TS.5.6-1).
i Figure TS.3.8-1, which specifies the minimum burnup requirements for unrestricted storage in the spent fuel pool, is based on enrichments from 3.87 l
to 5.0 weight percent U-235.
Enrichments lower than 3.87 weight percent are conservatively bounded by the minimum burnup requirement for 3.87 weight percent U-235 which is 2000 MWD /MTU. Therefore, Figure TS.3.8-1 has been drawn to require that fuel with an initial enrichment of less than 3.87. weight percent U-235 have 2000 MWD /MTU burnup or greater before unrestricted storage in the spent fuel pool will be allowed.
The water in the spent fuel pool normally contains soluble boron, which results_.
l in large subcriticality margins under actual operating conditions. However, t
the NRC guidelines, based upon the accident condition in which all soluble _
poison is assumed to have been lost, specify that the limiting k,cc of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions j
is based on the use of unborated water, which ensures that each region is j
maintained in a suberitical condition during normal operation with the regions fully loaded.
Most accident conditions do not result in a significant increase in the 1
activity of either of the two regions.
Examples of these accident conditions
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are the loss of cooling, ths dropping of a fuel assembly on the top of the rack, and the dropping of a fuel assembly between rack modules and wall (rack l
design precludes this condition). However, accidents can be postulated that could increase the reactivity.
For these accident conditions, the double l
contingency principle of ANSI N16.1-1975 can be applied. This states that one l
is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.
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The double contingency principle allows credit for soluble boron under abnormal or accident conditions, since only a single accident need be considered at one l
time.
For example, the most severe accident scenario is the accidental misloading of a fuel assembly into a rack location for which the restrictions on location, enrichment or burnup are not satisfied. This could potentially i
increase the reactivity in spent fuel racks. To mitigate these postulated j
criticality related accidents, Specification 3.8.E.2 ansures the spent fuel pool contains adequate dissolved boron anytime fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure i
TS.3.8-1 are stored in the fuel pool and a spent fuel pool verification has not
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been performed since the last movement of any fuel assembly in the spent fuel pool.
The negative reactivity effect of the soluble boron would compensate for the increased reactivity caused by a mispositioned fuel assembly.
Prairie Island Unit 1 Amendment No. 105, 119 Prairie Island Unit 2 Amendment No. 107,112 l
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i 3.3.8-4 i
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.i 3.8. REFUELING AND FUEL HANDLING f
Amasa contineed i
The boron concentration requirements of Specification 3.8.E.2 are no longer l
imposed when no fuel movements are occurring and a spent fuel pool verification has been completed, because the storage requirements of Specifications 3.8.E.1 and 5.6.A.l.d are then adequate to prevent criticality.
Specification 3.8.E.2.a is not imposed when only fuel assemblies with a
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combination of burnup and initial enrichment in the unrestricted range of i
Figure.TS.3.8-1 are stored in the spent fuel pool. The requirements of 1
Specification 3.8.E.2.a are not required.in that case because with only' fuel assemblies that have burnup and initial enrichment in the unrestricted range of Figure TS.3.8-1 it is not possible to cause an inadvertent criticality by mispositioning a fuel assembly in the' spent fuel pool.
l When the requirements of Specification 3.8.E.2.a are applicable, and the
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. concentration of boron.in the spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or:to mitigate i
the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
The concentration of boron is restored simultaneously with suspending movement of fuel t
assemblies. An acceptable alternative is to complete a spent fuel pool verification. However, prior to resuming movement of fuel assemblies,-the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.
A spent fuel pool verification is required following the last movement of fuel assemblies in the spent fuel pool, if fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are' stored in the spent fuel pool. This verification will' confirm that any fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in accordance with the requirements of Specification 5.6.A.1.d.
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References I
1.
USAR, Section 10.2.1.2 2.
USAR, Section 14.5.1 3.
USAR, Section 10.3.7 j
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" Criticality Analysis of the Prairie Island Units 1 & 2 Fresh and Spent Fuel Racks", Westinghouse Commercial Nuclear Fuel Division, February 1993.
Prairie Island. Unit 1 Amendment No. 91, 99, 108.119 e
Prairie Island Unit 2 Amendment No. 84, 92, 101, 112
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UNITED STATES 4
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NUCLEAR REGULATORY COMMISSION
'2 WASHINGTON. D.C. 20ea6-0001 o
%,.....,o NORTHERN STATES POWER COMPANY DOCKET N0. 50-306 PRAIRIE ISLAND NVCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.112 Licensa No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated December 5, 1994, as supplemented January 9, 1995, and May 15, 1995, complies 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 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 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 I
the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
i I
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". Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. -114.are hereby incorporated in the license.'
1 The licensee shall operate the facility in accordance with the
.i Technical Specifications.
t 3.
This license amendment is effective as of the date of issuance, with full
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implementation within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
%1h 7 AJJ Of Beth A. Wetzel, Project Manager" Project Directorate III-I Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation'
Attachment:
Changes to the Technical Specifications Date of Issuance: July 3, 1995 l
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ATTACHMENT TO LICENSE MiENDMENT NO. -112 i
FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306-i Revise Appendix A Technical Specifications by removing the pages~ identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical. lines indicating the area of change.
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REMOVE INSERT TS.3.8-1
-TS.3.8-1 TS.3.8-2 TS.3.8-2 B.3.8-1 B.3.8-1
-i B.3.8-2 B.3.8-2 B.3'8-3 B.3.8-3 B.3.8-4 B.3.8-4 i
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TS.3.8-1 3.8 REFUELIIIG AND FUEL HANDLING L.
Aanlicability Applies to operating limitations associated with fuel-handling operations, CORE ALTERATIONS, and crane operations in the spent fuel pool enclosure.
Obiectives To ensure that no incident could occur during fuel handling, CORE ALTERATIONS and crane operations that would ' affect public health and safety.
l Scecification A.
During CORE ALTERATIONS the following conditions shall be satisfied (except as specified in 3.8.A.2 and 3 below):
a.
- 1) The equipment hatch shall be closed. In addition, at least one isolation valve shall be OPERABLE or locked closed in each lina which penetrates the containment and provides a direct path from containment atmosphere to the outside.
- 2) Airlock doors a) At least one door in each air lock is closed, or b) Both doors in each air lock may be open if:
1.
The containment (high flow) purge system is isolated, ii. The inservice (low flow) purge system is capable of automatic isolation, iii. At least one door in each air lock is OPERABLE, under procedural control, and capable of being closed within 30 minutes following a fuel handling accident in
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containment, and iv. At least two containment fan coil unit fans are capable of operating in the high speed mode following a fuel handling accident in containment.
b.
Radiation levels in the fuel handling areas of the containment shall be monitored continuously.
i Prairie Island Unit 1 Amendment No. 73, 9J,119 Prairie Island Unit 2 Amendment No. 66, BA,112
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r TS.3.8-2 h
'3.8.A.1.c.
The core suberitical neutron flux shall be continuously monitored
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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 geometry is being changed. When core geometry is not being changed, at least 1
one neutron flux monitor shall be in service, d.
1he plant shall be'in the REFUELING condition.
E e.
During movement of fuel assemblies or contro1Lrods 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,to moving fuel assemblies or control. rods and at l
1 east once every day while the cavity is flooded.
f.
At least one residual heat removal pump shall be OPERABLE and j
running. The pump may be shut down for up to one hour to' facilitate movement of fuel or core components.
g.
If the water level above the top of the reactor vessel. flange is less than 20 feet, except for control rod unlatching / latching operations or upper internals removal / replacement, both residual
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heat removal loops shall be OPERABLE.
j h.
Direct communication between the control room and the operating floor of the containment shall be available whenever CORE ALTERATIONS are taking place,
- i. No movement of irradiated fuel in the reactor shall be made until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
J.
The radiation monitors which initiate isolation of the Containment
-l Purge System shall be tested and verified to be OPERABLE prior to CORE ALTERATIONS.
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2.
If any of the above conditions are not met CORE ALTERATIONS. shall cease. Work shall be 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.
3.
If Specification 3.8. A.1.f or 3.8. A.1.g cannot be satisfied, all fuel handling operations in containment shall be suspended, the requirements of Specification 3.8.A.1.a.1) shall tHe satisfied, at i
least.one' door in each personnel air lock shall be closed, and no j
reduction in reactor coolant boron concentration shall be made.
Prairie Island Unit 1 Amendment No. 73, 7f, 91, gg9 Prairie Island Unit 2 Amendment No. 66, 67, $4,112 1
B.3.8-1 3.8 REFUELING AND FUEL HANDLING Bases Core alteration containment isolation specifications are provided to minimize releases following a fuel handling accident (FHA). Allowing both airlock doors open during core alterations will facilitate evacuation of containment following a FHA and help maintain the seals in good working order. The FHA does not cause containment pressurization, however, with an assumed single failure the operating purge system supply fan is assumed to continue supplying air to containment. To maintain post-FHA releases well within the limits of 10CFR100, only the inservice purge system is allowed to be operating during core alterations. Two containment fan coil unit fans are required to operate in the high speed mode following a fuel handling accident in containment to assure that radioactive material in containment is well mixed and any releases will leave containment at i
a lower concentration over the duration of the accident. The provision that one door is OPERABLE and under procedural control will ensure that at least one door will be closed in within 30 minutes as required, thus assuring radioactive releases are well within the l
l limits of 10CFR100.
The equipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during CORE ALTERATIONS that would result in a hazard to public health and safety (Reference 1).
Whenever changes are not being made in core geometry, ona flux monitor is sufficient.
This permits maintenance of the instrumenta-tion.
Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.
Under rodded and unrodded conditions, the K rr of the reactor must be e
less than or equal to 0.95 and the boron concentration must be greater than or equal to 2000 ppm.
Feriodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.
3.8.A.l.h allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel.
The delay time is consistent with the fuel handling accident analysis (Reference 2).
Fuel will not be inserted into a spent fuel cask unless a minimum boron concentration of 1800 ppa is present. The 1300 ppm will ensure that k,cr for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask.
The number of recently discharged assemblies in Pool No. I has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.
Prairie Island Unit 1 Amendment No. 97, 99, 119 Prairie Island Unit 2 Amendment No. 84. 92, 112
l B.3.8-2 i
3.8 REFUELING AND FUEL HANDLING Bases continued The Spent Fuel Pool Special Ventilation System (Reference 3) is a safeguards system which maintains a negative pressure in the spent fuel enclosure upon detection of high area radiation. The Spent Fuel Pool Normal Ventilation System is automatically isolated and exhaust air is drawn through filter modules containing a roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of the Shield Building exhaust stacks. Two completely redundant trains are provided.
The exhaust fan 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 adsorbers in each SFPSVS filter train.
The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment.
During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provide sufficient shielding.
The water level may be lowered to the top of the RCCA drive shafts for latching and unlatching.
The water level may also be lowered below 20 feet for upper internals removal / replacement.
The basis for these allowance (s) are (1) the refueling cavity pool has sufficient level to allow time to initiate I
repairs or emergency procedures to cool the core, (2) during latching / unlatching and upper internals removal / replacement the level is closely monitored because the activity uses this level as a reference point, (3) the time spent at this level is minimal.
The Prairie Island spent fuel storage racks have been analyzed (Reference 4) to allow for the storage of fuel assemblies with enrichments up to 5.0 weight percent U-235 while maintaining K rt s 0.95 including uncertainties.
This criticality o
analysis utilized the following storage configurations or regions to ensure that the spent fuel pool will remain suberitical during the storage of fuel assemblies with all possible combinations of burnup and initial enrichment:
l 1.
The first region utilizes a checkerboard loading pattern to accommodate new or low burnup fuel with a maximum enrichment of 5.0 wet U-235.
This configuration stores " burned" and " fresh" fuel assemblies in a 2x2 checkerboard pattern.
Fuel assemblies stored in " burned" cell locations must have an initial enrichment less than 2.5 wet U-235 (nominal) or l
satisfy a minimum burnup requirement. The use of empty cells is also an l
acceptable option for the " burned" cell locations.
Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0 wtt U-235 with no requirements for burnup or burnable absorbers.
2.
The second region does not utilize any special loading pattern.
Fuel assemblies with burnup and initial enrichments which fall into the unrestricted range of Figure TS.3.8-1 can be stored anywhere in the region with no special placement restrictions.
Fuel assemblies which fall into i
the restricted range of Figure TS.3.8-1 must be stored in the checkerboard region in accordance with Specification 5.6.A.l.d.
Prairie Island Unit 1 Amendment No. 97, 99, 108,'l13 Prairie Island Unit 2 Amendment No. 84, 92, 101, 112
C*M
'B.3.8-3 c
1 3.8 - REFUELING AND FUEL HANDi.ING 13,g31 continued' The' burned / fresh fuel checkerboard region can be positioned anywhere within the' i
spent fuel racks, but,the boundary between the checkerboard region and the unrestricted region must be either:
1.
separated by a vacant row of cells, or
~2.
the' interface must be configured such that there is one row carryover of.
the pattern of burned assemblies from the checkerboard region into. the first row of the unrestricted region (Figure TS.5.6-1).
Figure TS.3.8-1, which specifies the minimum burnup requirements for.
unrestricted storage in the spent fuel pool, is based on enrichments from 3.87 to 5.0 weight percent U-235.
Enrichments lower than 3.87 weight percent are 4
conservatively bounded by the minimum burnup requirement for 3.87 weight percent U-235 which is 2000 MVD/MTU. Therefore, Figure TS.3.8-1 has,been drawn.
l to require that fuel with an initial enrichment of less than.3.87 weight percent U-235 have 2000 MWD /MTU burnup or greater before unrestricted storage.-
in the spent fuel pool will be allowed.
The water in the spent fuel pool normally contains soluble boron, which results-in large subcriticality margins under actual operating conditions. However,
-the NRC guidelines, based upon the accident condition in which all. soluble,
poison is assumed to have been lost, specify that the limiting k.cc of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions i
is based on the use of unborated water, which ensures that each region is
'l maintained in a suberitical condition during normal operation with the regions fully loaded.
i Most accident conditions do not result in a significant increase in the activity of either of the two regions.
Examples of these accident conditions are the loss of cooling, the dropping of a. fuel assembly on the top of the rack, and the dropping of a fuel assembly between rack modules and wall (rack design precludes this condition). However, accidents can be postulated that could increase the reactivity.
For these accident conditions, the double contingency principle of ANSI N16.1-1975 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.
The double contingency principle allows credit for soluble boron under abnormal or accident conditions, since only a single accident need be considered at one time.
For example, the most severe accident enenario is the accidental misloading of a fuel assembly into a rack loca. ion'for which the restrictions j
on location, enrichment or burnup are not satisfied. This could potentially increase the reactivity in spent fuel racks. To mitigate these postulated' criticality related accidents, Specification 3.8.E.2 ensures the spent fuel i
pool contains adequate dissolved boron anytime fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure
~
TS.3.8-1 are stored in the fuel pool and a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel pool. The negative reactivity effect of the soluble boron would compensate for the increased reactivity caused by a mispositioned fuel assembly.
Prairie Island Unit 1 Amendment No. 105,119 Prairie Island Unit 2 Amendment No. 101, 112 t
B.3.8-4 3.8 REFUELING AND FUEL HANDLING Bases continued The boron concentration requirements of Specification 3.8.E.2 are no longer imposed when no fuel movements are occurring and a spent fuel pool verification has been completed, because the storage requirements of Specifications 3.8.E.1 and 5.6. A.1.d are then adequate to prevent criticality.
Specification 3.8.E.2.a is not imposed when only fuel assemblies with a combination of burnup and initial enrichment in the unrestricted range of Figure TS.3.8-1 are stored in the spent fuel pool. The requirements of Specification 3.8.E.2.a are not required in that case because with only fuel assemblies that have burnup and initial enrichment in the unrestricted range of Figure TS.3.8-1 it is not possible to cause an inadvertent criticality by mispositioning a fuel assembly in the spent fuel pool.
When the requirements of Specification 3.8.E.2.a are applicable, and the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. An acceptable alternative is to complete a spent fuel pool verification. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.
A spent fuel pool verification is required following the last movement of fuel assemblies in the spent fuel pool, if fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the spent fuel pool. This verification will confirm that any fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in accordance with the requirements of Specification 5.6.A.1.d.
References 1.
USAR, Section 10.2.1.2 2.
USAR, Section 14.5.1 3.
USAR, Section 10.3.7 4.
" Criticality Analysis of the Prairie Island Units 1 & 2 Fresh and Spent Fuel Racks", Westinghouse Commercial Nuclear Fuel Division, February 1993.
Prairie Island Unit 1 Amendment No. 97, 99, 108,119 Prairie Island Unit 2 Amendment No. 84, 92, 101, 112