ML20246N962
| ML20246N962 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/28/1989 |
| From: | Yandell L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246N966 | List: |
| References | |
| NUDOCS 8909110009 | |
| Download: ML20246N962 (13) | |
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.-90 License No. DPR-42 1.
The Nuclear Regulatory Commission (the Commission) has found thst:
A.
The application for amendment by Northern States Power Company (the licensee) dated April 6, 1989, 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 satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:
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.2 Technical Specifications
'The-Technical Specifications contained-in Appendix A, as revised
- through Amendment No. 90,. are hereoy 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 Lawrence A. Yandell, Acting Director Project Directorate III-1 Division of Reactor Projects - III.
IV, V & Special Projects Office of Nuclear Reactor Regulation
Attachment:
' Changes to the Technical Specifications Date~of Issuance: August 28, 1989 m_____-__________
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NUCLEAR REGULATORY COMMISSION UNITED STATES o
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License 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 April 6, 1989, 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 satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
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Technical Specifications The.-Technical Specifications contained in Appendix A, as revised
'.through Amendment No.83, 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 Lawrence A. Yandell, Acting Director T+
Project Directorate'III-1 Division of Reactor Projects.- III, IV, V & Special Projects Office of Nuclear Reactor Regulation
Attachment:
t Changes to the Technical Specifications Date of Issuance: August 28, 1989 j
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ATTACHMENT TO LICENSE AMENDMENT NOSO 90 AND 83 FACILITY OPERATING LICENSE NOS.'DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 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 marginal lines indicating the area of change.
REMOVE INSERT TS.3.8 TS.3.8-3
'TS.3.8-4 TS.3.8-4 TS.3.8-5 TS.3.8-5 TS.3.8-6 TS.5.3-1 TS.5.3-1 TS.S.6-1 TS.5.6-1 TS.S.6-2
.T5.5.6-2 TS.5.6-3
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TS.3.8-3 D.
.Soent Fuel Pool Snecial Ventilation System
- 1. Except as specified in Specification 3.8.D.3 below, both trains of the Spent Fuel Pool Special Ventilation System and the diesel generators required for their operation sh: 1 be operable at all times.
- 2. a.
The results of in-place DOP and halogenated hydrocarbon tests at design' flows on HEPA filters and charcoal adsorber banks respec-l tively shall show 2 996 DOP removal for particles having a mean diameter of 0.7 microns and 2 994 halogenated hydrocarbon removal.
b.
The results of laboratory carbon sample analysis shall show 2 90% radioactive methyl iodide removal efficiency (130*C, 95% RH).
c.
The Spent Fuel Pool Special Ventilation System fans shall operate within i 10% of 5200 cfm per train.
- 3. From and after the date that one train of the Spent Fuel Pool Special Ventilation.Syrtem_is made or found inoperable for any reason, fuel handling operations are permissible only during the succeeding seven l
days (unless such train is made operable) provided that the redundant train is verified to be operable daily.
1
- 4. If the conditions for operability of the Spent Fuel Pool Special l
Ventilation-System cannot be met, fuel handling operations in the Auxiliary Building shall be terminated immediately.
E.
Storare of Low Burnuo Fuel I
- 1. The following restrictions shall apply whenever fuel with an average assembly burnup less than 5,000 MWD /MTU is stored in the spent fuel pool (except as specified in 3.8.E.2 and 3.8.E.3 below):
i a.
The boron concentration in the spent fuel pool shall be maintained
-l greater than or equal to 500 ppm, and j
b.
Fuel with an average assembly burnup less than 5,000 MWD /MTU shall not be stored in more than three storage locations of every four by four storage rack array.
~
- 2. If the conditions in 3.8.E.1.a above are not met, verify that the spent fuel pool storage configuration meets the requirements of specification 3.8.E.1.b and suspend all actions involving the movement of fuel in the spent fuel pool until the boron concentration is increased to 500 ppm i
or greater.
i
- 3. If the conditions in 3 8.E.1.b above are not met, suspend all actions involving movement of fuel in the spent fuel pool, verify the spent fuel pool boron concentration to be greater than or equal to 500 ppm I
and initiate corrective actions. Mis-positioned fuel assemblies shall be moved to acceptable locations prior to the resumption of other fuel movement in the spent fuel pool.
Prairie Island Unit 1 - Amendment No. 17, 25, 73, 74, 90 Pr Island Unit 2 - Amendment No. JJ,19, E6, E7, 83 1
TS.3.8-4 i
Basis The equipment and general procedures to be utilized during refueling are dis-cussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in inter-locks and safety features, provide assurance that no incident could occur during the ueling operations that would result in a hazard to public health and safety.
Wenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation.
Continuous monitoring of radiation levels (B. above) 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,ff of the reactor must be < 0.95 and the boron concentration must be > 2000 ppm as indicated in A.4.
Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.
A.9 above 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 analysis. g The delay time is consistent with the fuel handling accident in the fu The spent fuel assemblies will be loaded into the spent fuel cask after sufficient decay of fission products. Wile inserting and withdrawing the cask into pool No. 1, the cask will be suspended above the bottom of the pool up to a maximum of 42 feet. The consequence potential load drops have been evaluated in accordance with NUREG-0612 Follow hg is a dis-cussion of the basis for the limitations which resulted from that evaluation.
The cask will not be inserted into the pool until all fuel stored in the pool has been discharged from the reactor a minimum of 5 years.
Supporting analysis indicated that fuel stored in the pool for a period as short as 50 days would allow sufficient decay of the fission products such that their release would result in off-site doses less than 25% of the 10 CFR Part 100 guidelines. The five year decay period was selected in following the general principle that spent fuel with the longest decay time would result in the least off-site doses in the event of an accident, while providing the plant operational flexibility.
The cask will not be inserted or withdrawn from the pool unless a minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that if fuel is crushed by a cask drop, K,ff will be less than or equal to 0.95.
The cask will not be inserted or withdrawn from the pool unless a cask impact limiter, crash pad, or combination thereof is in place with the capability to absorb energy of a cask drop such that no significant amount of water leakage results from pool structural damage. This is to ensure that at no time will water level drop below the top of the spent fuel stored in the In loading 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 storage period.
Prairie Island Unit 1 - Amendment No. 17, 25, 73, 7 A, 90 Prairie Island Unit 2 - Amendment No. JJ, IP, EE, E7,83 1
j l
I TS.3.8-5 At this time, the radioactivity has decayed so that a release of fission-l-
products from all fuel assemblies in the car.:k would result in off-site doses l
1ess than 10 CFR Part 100. It is assumed, for this dose analysis, that 12 assemblies rupture after 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.
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 capabil-ity, at least eight hours are available under worst case conditions to make l
repairs until the onset of boiling.
The Spent Fuel Pool Special Ventilation System (3) is a safeguards system which maittains a negative pressure in the spent fuel enclosure upon detection of high area radiation. The Spent Fuel Pool Normal-Ventilation System is auto-matica11y 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 j
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. 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 methyl iodide removal 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 quali-fication 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 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 bases for these allowances 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 l
is minimal.
Prairie Island Unit 1 - Amendment No. 25, 47, 63, 74, 90 Prairie Island Unit 2 - Amendment No.19, 42, E7, 67, 83
_ = _. __
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TS.3.8-6 1
The requirements for the storage of low burnup fuel.in the spent fuel pool ensure that the spent fuel pool will remain suberitical during fuel storage.
Fuel stored in the spent fuel pool will be limited to a maximum enrichment of 4.25 weight percent U-235.
It has been shown by criticality analysis that the use of the three out of four storage configuration will assure that the K,ff will remain less than 0.95, including uncertainties, when fuel with a maximum enrichment of 4.25 weight percent U-235 and average assembly burnup of less than 5,000 MWD /MTU is stored in the spent fuel pool.
The requirement for maintaining the spent fuel pool boron concentration
. greater than 500 ppm whenever -fuel with average assembly burnup of less than 5,000 MWD /MTU is stored in the spent fuel pool ensures that K,fg for the spent fuel pool will remain less than 0.95, including uncertainties, even if a fuel assembly is inadvertently inserted in the empty cell of the three out of~four storage configuration.
References (I) FSAR Section 9.5.2 (2) FSAR Section 14.2.1 (3) FSAR Section 9.6 (4) FSAR Page 9.5-20a (5) Exhibit C, NSP License Amendment Request Dated December 21, 1984 l
Prairie Island Unit 1 - Amendment No. 90 Prairie Island Unit 2 - Amendment No. 83 l
'4 TS.S.3-1 5.3 REACTOR A.
Reactor Core 1.
The reactor core contains uranium in the form of sliga. ly enriched uranium dioxine pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor r. ore is made up of 121 fuel assemblies. Each fuel assembly contains 179 fuel rods (Reference 1).
2.
The maximum enrichment will be 4.25 weight percent U-235.
3.
In the reactor core, there are 29 full-length RCC assemblies that contain a 142-inch length of silver-indium-cadmium alloy c'_ad with stainless steel (Reference 2).
B.
The e-a'gn of the reactor coolant system complies with all appl-cable vode requirements (Reference 3).
All hi h pressure piping, components of the reactor coolant 2.
S system and their supporting structures are designed to Class I requirements. and have been designed to withstand:
a.
The design seismic ground acceleration, 0.06g acting in the horizontal and 0.04g acting in the vertical planes simultane-ously, with stresses maintained within code allowable workit S stresses.
b.
The maximum potential seismic ground acceleration, 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.
3.
The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.
C. Protection Systems The' protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design. includes a rcactor trip for a high negative rate of change of neutron flux as meas ared by the excore nuclear instruments (Reference 4).
The system is intended to trip the reacter upon the abnormal dropping of more than one control rod (Reference 4).
If only one l
control rod is dropped, the core can be operated at full power for a short time, as permitted by Specification 3.10.
References 1.
USAR, Scction 3.4.2 3.
USAR, Table 4.1-11 2.
USAR, Section 3.5.2 4
USAR, Section 7.1 Prairie Island Unit 1 - Amendmer.t No. 3E, 48, 80,90 Prairie Island Unit 2 - Amendment No. 29, A2, 73,83 lw
TS.5.6-1 5.6 FUEL HANDLING A.
Criticality Consideration The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (seismic) structures. The spent fuel pit has a stainless steel liner to ensure against loss of water (Reference 1).
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations.
The design of the new fuel storage pit and racks (Reference 1) ensures a new fuel pit K,ff of less than or equal to 0.95, including uncertain-ties, even if unborated water were used to fill the pit.
The new fuel rackconfigurationalsoensuresK,7f less than or equal to 0.98, including uncertainties, even if cae new fuel racks were accidentally filled with a low density moderator which resulted in optimum low density moderation conditions.
Fuel stored in the new fuel storage racks will have a maximum enrichment of 4.25 weight percent U-235.
The spent fuel storage rack design (Reference 1) and the limitations on the storage of low burnup fuel contained in Technical Specification Section 3.8.E ensure a spent fuel pooi K,ff of less than or equal to 0.95, including uncertainties. The maximum enrichment of fuel to be stored in the spent fuci pool will be 4.25 weight percent U-235.
The criticality considerations as they relate to the dropping of a spent fuel cask (i.e., heavy load) drop onto the racks has been evalu-ated.
The maximum K,ff has been calculated to be less than 0.900 for water /UO2 ratios of between 2.0 and 2.3 with a boron concentration of 1800 ppm.
B.
Spent Puel Storare Structure The spent furi storage pool is enclosed with a reinforced concrete building having 12-to 18-inch thick walls and roof (Reference 1).
The pool and pool enclosure arc Class I (seismic) structures that afford protection against loss of integrity from postulated tornado missiles. The storage compartments and the fuel transfer canal are connected by fuel transfer slots that can be closed off with pneumatically sealed gates. The bottoms of the slots are above the tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks.
The. spent fuel pool has a reinforced concrete bottom slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident.
In addition, the spent fuel cask will have an innact limiter attached or a crash pad will be in place in the pool i
Prairie Island Unit 1 - Amendment No. 77, 22, AB, 74, 80,90 Prairie Island Unit 2 - Amendment No.11, J5, AZ, 67, 73,83 Correction Letter of July 26, 1985 L________
TS.S.6-2 l-l -
I which will have the capability to absorb energy of impact due to a l
cask drop.
This will result in no structural damage taking place to the pool which would result in significant leakage from the pool.
i Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies.
C.
Fuel Handling The fuel handling system provides the means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves after post-irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system.
Major components of the fuel handling system are the manipulation tone, the spent fuel pool bridge, the auxiliary building crane, the fuel transfer system, the spent fuel storage racks the spent fuel cask, and the rod cluster control changing fixture. The reactor vessel stud tensioner, the reactor vessel head lifting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assembling the reactor after refueling.
Upon arrival in the storage pit, spent fuel will be removed from the transfer system and placed, one assembly at a time, in storage racks using a long-handled manual tool suspended from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane.
The load drop consequences of a spent fuel cask for Prairie Island have been evaluated. It is not possible, due to physical constraints, for a cask to be dropped into the large pool (pool no. 2).
A load path has been defined which provides for safe movement of the cask. Travel interlocks and mechanical stops prevent cask movement outside of this path. The only safety-related equipment that can be impacted directly during a cask drop along this path is the fuel stored in the small pool (pool no. 1).
The consequences of this drop have been evaluated and found to meet the NRC staff criteria contained in NUREG-0612 if at least 50 days have elapsed since reactor shutdown for fission gas release considerations and the pool water contains at least 1800 ppm boron for criticality considerations. While 50 days was determined adequate, a minimum decay period of 5 years has been incorporated into these technical specifications to provide additional margin in meeting the criteria specified in NUREG 0612 for fission gas releases, while not restricting the plant's operational flexibility. A cask impact limiter or crash pad prevents significant structural damage to the pool floor.
Prairie Island Unit 1 - Amendment No. 48, 61, 74, 80, 90 Prairie Island Unit 2 - Amendment No. #2, EE, E7, 73, 83
h l
TS.S.6-3 o
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i The' spent fuel cask will be' lowered 66 feet from the auxiliary building to the railroad car for offsite transportation. Specification 3.8 will limit this loading operation so that if the cask-drops 66 feet, there will not be a significant release of fission products from the fuel in w
the cask.
D.
Soent Fuel'Storare Caoacity The spent fuel storage facility is a two-compartment pool that, if completely filled with fuel storage racks, provides up to 1582 storage locations. The southeast corner of the small pool (pool no.'1) also serves as the cask lay down area. During times when the cask is being used, four racks are removed from the small pool. With the four storage racks in the southeast corner of pool 1 removed, a total of 1386 storage locations are provided. To allow insertion of a shipping cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor.
Reference 1.
USAR, Section 10.2 Prairie Island Unit 1 - Amendment No.90 Prairie Island Unit 2 - Amendment No.83
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