ML20057A601
| ML20057A601 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/03/1993 |
| From: | Bill Dean Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20057A602 | List: |
| References | |
| NUDOCS 9309150029 | |
| Download: ML20057A601 (42) | |
Text
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UN11ED STATES 0
Q
^ jf NUCLEAR REGULATORY COMMISSION W A5HINGTON. D.C. 205S0001 1
NORTHERN STATES POWER COMPANY i
DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. I j
l AMENDMENT TO FACILITY OPERATING LICENSE I
i Amendment No. 10B License No. DPR-42 j
1.
The Nuclear Regulatory Commission (the rammission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated June 11, 1993, as revised June 30, 1993, 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 j
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 Tcchnical 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 i
to read as follows:
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9309150029 930903 PDR ADDCK 05000282 i
P PDR i
. = = -. -.. -
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2-Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 1W, 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.
FOR THE NUCLEAR REGULATORY COMMISSION
[kH(1
?-
William H. Dean, Acting irector 4
Project Directorate 111-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation i
Attachment-Changes to the Technical 1
Specifications l
Date of Issuance:
Septeter 3,1993 i
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l ATTACHMENT TO LICENSE AMENDMENT NO. 108 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 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.
REMOVE INSERT TS-lii TS-iii TS-vii TS-vii TS-xiii TS-xiii l
TS.3.3-1 TS.3.3-1 TS.3.8-4 TS.3.8-4 TS.3.8-5 Figure TS.3.8-1 Table TS.4.1-2B (Page 1 of 2) Table TS.4.1-2B (Page 1 of 2) l Table TS.4.1-2B (Page 2 of 2) Table TS.4.1-2B (Page 2 of 2)
TS.5.3-1 TS.5.3-1 TS.5.6-1 TS.5.6-1 15.5.6-2 TS.5.6-2 TS.5.6-3 Figure TS.5.6 4 Figure T5.5.6-2 B.3.8-2 B.3.8-2 B.3.8-3 B.3.8-4 1
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TS-iii TABLE OF CONTENTS (Continued)
TS SECTION TI.TLE PACE 3.6 Containment System TS.3.6-1 A. Containment Integrity TS.3.6-1 B. Vacuum Breaker System TS.3.6-1 C. Containment Isolation valves TS.3.6-1 D. Containment Purge System TS.3.6-2 E. Auxiliary Building Special Ventilation Zone Integrity TS.3.6-2 F. Auxiliary Building Special Ventilation System TS.3.6-3 C. Shield Building Integrity TS.3.6-3 H. Shield Building Ventilation System TS.3.6-3 I. Containment Internal Pressure TS.3.6-3 J. Contain=ent and Shield Building Air Temperature TS.3.6-4 K. Containment Shell Temperature TS.3.6-4 L. Electric Hydrogen Recombiners TS.3.6-4 M. Containment Air Locks TS.3.6-4 3.7 Auxiliary Electrical Systes TS.3.7-1 3.8 Refueling and Fuel Handling TS.3.8-1 A. Core Alterations 75.3.8-1 B. Fuel Handling Operations TS.3.8-3 C. Small Spent Fuel Pool Restrictions TS.3.8-4 D. Spent Fuel Fool Special Ventilation System TS.3.8-4 E. Spent Fuel Pool Storage TS.3.8-4 l
3.9 Radioactive Effluents TS.3.9-1 A. Liquid Effluents TS.3.9-1
- 1. Concentration TS.3.9-1
- 2. Dose TS.3.9-1 I
- 3. Liquid Radvaste System TS.3.9-2 4 Liquid Storage Tanks 75.3.9-2
- 3. Caseous Effluents TS.3.9-3
- 1. Dose Rate TS.3.9-3
- 2. Dose from Noble Cases 75.3.9-3
- 3. Dose from I-131, Tritiu= and Radioactive Particulate TS.3.9-4
- 4. Caseous Radvaste Treatment System and Ventilation Exhaust Treate.ent Systems TS.3.9-4
- 5. Containment Purging TS.3.9-5 C. Solid Radioactive Waste TS.3.9-6 D. Dose from All Uranium Fuel Cycle Sources TS.3.9-6 E. Radioactive Liquid Effluent Monitoring Instrumentation TS.3.9-7 F. Radioactive Caseous Effluent Monitoring Instrumentation TS.3.9-7 l
Prairie Island Unit 1 Amendment No. 97, 97, 108 i
Prairie Island Unit 2 Amendment No. Ef, 97, 101
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TS-vii TABLE OF CONTENTS (Continued)
TS SECTION TITLE PAGE 5.0 DESIGN FEATURES TS.5.1-1 i
5.1 Site TS.S.1-1 5.2 A.
Containment Structures TS.5.2-1 i
- 1. Containment vessel TS.S.2-1
- 2. Shield Building TS.S.2-2
- 3. Auxiliary Building Special Ventilation Zone B.
Special Ventilation Systems TS.S.2-2 C.
Containment System Functional Design TS.5.2-3 5.3 Reactor TS.S.3-1 A. Reactor Core TS.S.3 1 B. Reactor Coolant System TS.5.3-1 C. Protection Systems TS.5.3-1 5.4 Engineered Safety Features TS.S.4-1 i
5.5 Radioactive Waste Systems TS.5.5-1 l
A. Accidental Releases TS.5.5-1 i
B. Routine Releases TS.5.5 1
- 1. Liquid Vastes TS.S.5-1
{
- 2. Gaseous Wastes TS.S.5-2
- 3. Solid Wastes TS.S.5-3 i
C. Process and Effluent Rad'slogical Monitoring TS.S.5-3 i
System 5.6 Fuel Handling TS.S.6-1
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A. Criticality Consideration TS.5.6-1 B. Spent Fuel Storage Structure TS.S.6-1 C. Fuel Handling TS.S.6-2 t
D. Spent Fuel Storage Capacity TS.5.6-3 l
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Prairie Island Unit 1 Amendment No. 97, PP, 108 l
?inirie Island Unit 2 Amendment No. Ef, PE, 101 j
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TS-xiii b
APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FICURES TS FIGURE TITLE i
2.1-1 Safety Limits. Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-z Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER vith the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.8-1 Spent Fuel Pool Unrestricted Region Minimum Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Caseous Effluents 3.10-1 Rsquired Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6 2 Spent Fucl Fool Checkerboard Region Minimum Burnup Requirr+ ants 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group Prairie Island Unit 1 Amendment No. 97, Jps, 108 Prairie Island Unit 2 Amendment No. b3, ps, 101 1
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a TS.3.3-1 i
3.3 ENGINEERED SAFETY FEATURES Arrlic abili ty Applies to the operating status of the engineered safety features.
Obiective To define those limiting conditions that are necessary for operation of engineered safety features:
(1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.
Specifications A.
Jafety Iniection and Residual Heat Removal Systems 1.
A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200*F unless the following conditions are satisfied (except as specified in 3.3.A.2 below):
a.
The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 2500 ppm.
b.
Each reactor coolant system accumulator shall be OPERABLE when reactor coolant system pressure is greater than 1000 psig.
OPERABILITY requires:
L (1) The isolation valve is open (2) Volume is 1270 i 20 cubic feet of borated water (3) A minimum boren concentration of 1900 ppm (4) A nitrogen cover pressure of 740 30 psig Two safety injection pumps are OPERABLE except that pump c.
control switches in the control room shall meet the require-ments of Section 3.3.A.3, 3.3.A 4 and 3.1.A.l.d.(2) whenevet the reactor coolant system temperature is less than 310*F*.
d.
Two residual heat removal pumps are OPERABLE.
i e.
Two dual heat exchangers are OPERAn r, i
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- Valid until 20 EFPY Prairie Island Unit 1 Amendment ho. 77, 9J, 108 Prairie Island Unit 2 Amendment No. 79, E9, 101
TS.3.8-4 3.8.C. Small Seent Fuel Pool Restrictions No more than 45 recently discharged assemblies shall be located in the small pool (pool No. 1).
I D.
Spent Fuel Pool Special Ventilation System 1.
Porh trains of the Spent Fuel Pool Special Ventilation System shall be OPERABLE at all tires (except as specified in 3.8.D.2 and 3.8.D.3 below).
2.
With one train of the Spent Fuel Pool Special Ventilation System e
inoperable, fuel handling operations and crane operations with loads over spent fuel (inside the spent fuel pool enclosure) are permissible during the following 7 days, provided the redundant train is demonstrated OPERABLE prior to proceeding with those operations.
3.
With both trains of the Spent Fuel Fool Special Ventilation System inoperable, suspend all fuel handling operations and crane operations with loads over spent fuel (inside the spent fuel pool enclosure).
4 The provisions of specification 3.0.C are not applicable.
E.
Spent Fuel Pool Storare 1.
Fuel Assembly Storage To be stored without restriction in the spent fuel pool, the a.
burnup and initial enrichment of a fuel assembly shall be within the unrestricted range of Figure TS.3.8-1.
b.
Fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 shall be stored in accordance with Specification 5.6.A.1.d.
c.
If the requirements of 3.8.E.1.a and 3.8.E.1.b are not met, ic=ediately initiate cetion to move any noncomplying fuel assembly to an acceptable location.
d.
The provisions of Specification 3.0.C are not applicable.
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Prairie Island Unit 1 Amendment No. 99, 97, 108 Prairie Island Unit 2 Amendment No. SJ, SJ, 101 l
TS.3.8-5 l
J 3.8.E.2. Spent Puel Pool Boron Concentration The spent fuel pool boron concentration shall be a 1,800 ppm a.
when 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 and a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel pool.
b.
If the requirements of specification 3.8.E.2.a are applicable and the spent fuel pool boron concentration is not within its limit, then immediately:
I 1.
Suspend movement of fuel assemblies in the spent fuel pool, and 2.
Either initiate action to restore spent fuel pool boron concentration to within its limit or perform a spent fuel pool verification.
The provisions of Specification 3.0.C are not applicable.
c.
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Prairie Island Unit 1 Amendment No. 108 Prairie Island Unit 2 Amendment No. 101 I
FIGURE TS.3.8-1 12000 11000 4
l l
/
10000 l
/
/
/
DhD'l KID
/
/
,/
m M
l s
v t
f 7000
=
t
/
s f
na 5000
/
O
/
b
,r s
/
U 5000
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o
/
/
/
/
4000
/
z
/
/
MN 3000 j
/
I y
s 1000 b.5 4.0 4.5 5.0 INITIAL NOMINAI. U-235 ENRICIDENT (w/o)
FIGURE TS.3.8-1 Spent Fuel Fool Unrestricted Region Minimum Burnup Requirements Prairie Island Unit 1 Amendment No. 108
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Prairie Island Unit 2 Amendment No. 101 l
Table TS.4.1-2D (Page 1 of 2) l TABLE TS.4.1-2B l
MINIMUM FREQUENCIES FOR SAMPLING TESTS FSAR Section TEST FREOUENCY Reference 1.
RCS Gross 5/ week Activity Determisation i
2.
RCS Isotopic Analysis for DOSE 1/14 days (when at power)
EQUIVALINT I-131 Concentration 3.
RCS Radiochemistry E determination 1/6 months (l) (when at power) 4 RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE _
EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.
RCS Radiochemistry (2)
Monthly 6.
RCS Tritium Activity Weekly 7.
RCS Chemistry (Cl*,F*, 02) 5/ Week B.
RCS Boron Concentration *(3) 2/ Week (4) 9.2 9.
RWST Boron Concentration Weekly
- 10. Boric Acid Tanks Boron Concentration 2/ Week
- 11. Caustic Standpipe NaOH Concentration Monthly 6.4
- 12. Accumulator Boron Concentration Monthly 6
- 13. Spent Fuel Pit Boron Concentration Monthly /Veeklyons) 9.5.5 i
Prairie Island Unit ;
Amendment No. ER, SP, 108 Prairie Island Unit 2 Amendment No, pp, P2, 101
Tchle TS.4.1-2B (Page 2 of 2)
TABLI TS.4.1-2B
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i MINIMUM FREOUENCIES FOR SAMPLING TESTS l
FSAR Section TEST FREOUENCY Reference 14 Secondary Coolant Gross Weekly Beta-Ga=ma activity l
15.
Secondary Coolant Isotopic 1/6 months (5) i Analysis for DOSE EQUIVALENT I-131 concentration 16.
Secondary Coolant Che=istry pH 5/ week (6) pH Control Additive 5/ week (6) j Sodium 5/ week (6) 1 Notes:
1.
Sample to be taken after a minimum of 2 EFFD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l
or longer.
2.
To determine activity of corrosien products having a half-life greater than 30 minutes.
3.
During REFUELING, the boron concentration shall be verified by chemical analysis daily.
4 The maximum interval between analyses shall not exceed 5 days.
5.
If activity of the samples is greater than 10% of the limit in Specification 3.4.D. the frequency shall be once per month.
6.
The maximum interval between analyses shall not exceed 3 days.
7.
The minimum spent fuel pool boron concentration from Specification 3.8.B.1.b shall be verified by chemical analysis weekly while a spent fuel i
cask containing fuel is located in the spt.nt fuel pool.
8.
The spent fuel pool boron concentration shall be verified weekly, by chemical analysis, to be within the limits of Specification 3.8.E.2.a when fuel assesblies with a coibination of burnup and initial enrichment in the restricted ranee of Figure TS.3.8-1 are stored in the spent fuel pool and a
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spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel pool.
Prairie Island Unit 1 Amendment No. 9J, SP, 108 Prairie Island Unit 2 Amendment No. 59, 97, 101
TS.5.3-1 5.3 REACTOR A.
Reactor Core 1.
The reactor core contains uranium in the form of natural or slightly
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enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 or ZIRLD tubing to form fuel rods. The reactor core is made l up of 121 fuel assemblies.
Each fuel assembly contains 179 fuel rods (Reference 1).
2.
The maximum enrichment will be 5.0 veight 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 clad with stainless steel (Reference 2).
B.
Resetor Coolant Syste=
1.
The design of the reactor coolant system complies with all appli-cable code requirements (Reference 3).
2.
All high pressure piping, cor.ponents of the reactor coolant system and their supporting structures are designed to Class I requirements, and have been designed to withstand:
1 The design seismic ground acceleration, 0.06g acting in the a.
horizontal and 0.04g acting in the vertical planes simultane-ously, with stresses maintained within code allevable working stresses.
b.
The maximu= potential seismic ground acceleration, 0.12g, acting in the horizontal and 0.08g acting in the vertical 4
planes simultaneously with no loss of function.
3.
The nominal liquid volume of the reactor coolant systes, at rated operating conditions, is 6100 cubic feet.
C. Protection Systems The protection syste=s for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of change of neutron flux as measured by the excore nuclear instruments (Reference 4).
The system is intended to trip the reactor upon the abnormal dropping of more than one control rod (Reference 4).
If only one control rod is dropped, the core can be operated at full power for a short time, as permitted by Specification 3.10.
References i
1.
USAR, Section 3.4.2 3.
USAR, Table 4.1-11 2.
USAR, Section 3.5.2 4.
USAR, Section 7.1 Prairie Island Unit 1 Amendment No. EP, 99, 108 Prairie Island Unit 2 Amendment No. 77, SJ, 101
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1 TS.5.6-1 i
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5.6 FUEL HANDLING A.
Criticality Consideration 1.
The spent fuel storage racks are designed (Reference 1) and shall be
(
maintained with:
l Fuel assemblies having a maximum U-235 enrichment of 5.0 weight a.
percent; I
b.
14rr s 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference 2; New or spent fuel assemblies with a combination of burnup and c.
initial enrichment in the unrestricted range of Figure TS.3.8-1 allowed unrestricted storage in the spent fuel racks; and d.
New or spent fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1
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stored in compliance with Figures TS.S.6 1 and TS.S.6-2.
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.2.
The new fuel storage racks are designed (Reference 1) and shall be j
maintained with:
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Fuel assemblies having a maximum U-235 enrichment of 5.0 veight a.
percent; i
b.
K.tr s 0.95 if fully flooded with unborated water, which includes i
an allowance for uncertainties as described in Reference 2; and l
c.
K.fr s 0.98 if accidentally filled with a low density moderator l
vhich resulted in optimum low density moderation conditions.
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3.
Fuel will not be inserted into a spent fuel cask in the pool, unless a i
minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that k rr for the spent fuel cask, including statistical e
uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask. The criticality analysis for i
the TN-40 spent fuel storage cask was based on fresh fuel enriched to j
i 3.85 weight percent U-235.
L B.
Svent Fuel Storare Structure l
b The spent fuel storage pool is enclosed with a reinforced concrete building having 12. to 18-inch thick walls and roof (Reference 1).
l The pool and pool enclosure are Class'I (seismic) structures that afford protection against loss of integrity from postulated tornado 3
missiles. The storage compartments and the fuel transfer canal are connseted 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
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vertically in specially constructed racks.
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Prairie Island Unit 1 Amendment No. pp, pp, 108 Prairie Island Unit 2 Amendment No. 53, 97, 101 1
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TS.S.6-2 The spent fuel pool has a reinforced concrete bottod slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident. Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies.
The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (seismic) structures. The spent
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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.
C.
Fuel Handline 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 crane, 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 lif ting device are used for preparing the reactor for refueling and for assembling the reactor after refueling.
Upon arrival in the storage pit, spent fuel vill 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 vill be loaded into storaEe casks for storage in the Independent Spent Puel Storage Installation or into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane.
Spent fuel casks will be handled by a single failure proof handling system meeting the requirements of Section 5.1.6 of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants", July 1980. The auxiliary building crane has been upgraded to conform with the single failure proof requirements of Section 5.1.6 of NUREG-0612. The auxiliary building crane is designed to not allow a load drop as a result of any single failure. The improved reliability of the auxiliary building crane is achieved through increased factors of safety and through redundancy or duality in certain active components.
Prairie Island Unit 1 Amendment No. 99, 97, 108 Prairie Island Unit 2 Amendment No. SJ, 97, 101 l
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TS.S.6-3 D.
Spent Fuel Storare Capacity The spent fuel storage facility is a two-compartment pool that,.i f completely filled with fuel storage racks, provides up to 1582 stora6e 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 j
used, four racks are removed from the small pool. With the four i
storage racks in the southeast corner of pool 1 removed, a total of -
1386 storage locations are provided. To allow insertion of a spent fuel l
cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor.
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i References
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1.
USAR, Section 10.2 1
2.
' Criticality Analysis of the Prairie Island Units 1 & 2 Fresh and Spent Fuel Racks," Westinghouse Commercial Nuclear Fuel Division, February 1993.
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Prairie Island Unit 1 Amendment No. pp, pp, 108 i
Prairie Island Unit 2 Amendment No. EP, pg, 101
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FIGURE TS.5.6-1
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'x ii.. ' -
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..b55 NE. ;iii 5 5'S25
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PATTERN FOR CHECKERBOARD REGION I
l M M PA M M MMVA WAV&/"AMMV45%
VW a/VAg
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y 77 7 j7 1 - [ J ///)
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E BOUNDARY BEIVEEN CHECKERBOARD AND UNRESTRICTED REGIONS Fresh Fuel:
Enrichments w to 5.0 m/o U-235, no restrictions on turn, q.g Checkerboard Region mei!! Burned Fuel:
M.st settsfy minima turne reg.strements of Figure Ts.5.6 2.
j Unrestreited Region Burned Fuel:
Must settsfy einimum turn, requirements of tipure Ts.3.s-1.
1 i
Note: The Checkerboard and unrestricted regions can alternatively be separated by a single row of vacant cells on each adjacent face, s
d FIGURE TS.5.6-1 Spent Fuel Fool Burned / Fresh Checkerboard Cell Layout Prairie Island Unit 1 Amendment No. 108 j
Prairie Island Unit 2 Amendment No. 101
TICURE TS.S.6-2 30000 1-I
/
i 7
/
/
25000 f
4 ACCEPTABLE l
/
/
/
g
( 20000
/
I l
l, l
j v
/
l 1
I
/
B I
I f.
w 15000 7
j a
l f
u s
5 l
/
NOT ACCEPTABLE !
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/
g*
l
/
E 10000 g
/
I
/
/
I l
I l/
I A
5000 t
/
\\
/
l/
/
S./ l 5
3.0 3.5 4.0.
4.5 5.0 INITIAL NOMINAL U-235 EKRICEMENT (w/c)
FIGURE TS.5.6-2 Spent Fuel Fool Checkerboard Region Minimum Burnup Requirements Prairie Island Unit 1 Amendment No. 108 Prairie Island Unit 2 Amendment No. 101 l
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B.3.8-2 3.8 PlFUELING AND FUEL HANDLING I
i Bases continued i
During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provide sufficient shielding.
I The water level may be lowered to the top of the RCCA drive shafts for latching and unistching. The water level may also be lowered below 20 l
1 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 j
core, (2) during latching /unistching and upper internals removal / replace-l ment the level is closely monitored because the activity uses this l
1evel as a reference point, (3) the time spent at this level is minimal.
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The Prairie Island spent fuel storage racks have been analyzed (Reference 4) to j
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 i
4 criticality analysis utilized the following storage configurations or regions l
to ensure that the spent fuel pool vill remain suberitical during the storage i
of fuel assemblies with all possible combinations of burnup and initial i
enrich =ent:
j 1.
The first region utilizes a checkerboard loading pattern to accommodate new l
or low burnup fuel with a maximum enrichment of 5.0 vtt U-235.
This I
configuration stores " burned" and " fresh" fuel assemblies in a 2x2 checkerboard pattern. Fuel assemblies stored in " burned" cell locations i
must have an initial enrichment less than 2.5 wet 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 i
j in the " fresh
- cell locations can have enrichments up to 5.0 vtt U-235 with
{
no requirements for burnup or burnable absorbers, i
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 the restricted range of Figure TS.3.8-1 must be stored in the checkerboard region in accordance with Specification 5.6.A.1.d.
i I
The burned / fresh fuel checkerboard region can be positioned anywhere within the spent fuel racks, but the boundary between the checkerboard region and the unrestricted region must be either:
j I
1.
separated by a vacant row of cells, or the interface must be configured such that there is one row carryover of
{
2.
the pattern of burned assemblies from the checkerboard region into the first row of the unrestricted region (Figure TS.S.6-1).
j i
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J Prairie Island Unit 1 Amendment No. 9J, 99, 108 Prairie Island Unit 2 Amendment No. H, 97, 101 i
i
B.3.8-3 3.8 PETUELING AND FUEL HAW LING Bases continued 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 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 MVD/MTU. Therefore, Figure TS.3.8-1 has been drawn to require that fuel with an initial enrichment of less than 3.87 weight 1
percent U-235 have 2000 MVD/M7V burnup or greater before unrestricted storage l
j in the spent fuel pool will be allowed.
1 The water in the spent fuel pool normally contains soluble boron, which results i
in large suberiticality margins under actual operating conditions.
- However, i
the NRC guidelines, based upon the accident condition in which all soluble poisen is assumed to have been lost, specify that the limiting k.rr of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which ensures that each region is
=aintained in a suberitical condition during normal operation with the regions i
fully loaded.
Most accident conditions do not result in a significant increase in the i
activity of either of the two regions.
Exa=ples 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 vall (rack
]
design precludes this condition). However, accidents can be postulated that could increase the reactivity.
For these accident conditions, the double j
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 abnormat 1
or accident conditions, since only a single accident need be considered at ene time.
For example, the cost severe accident scenario is the accidental sisloading of a fuel assembly into a rack location for which the restrictions on location, enrich =ent or burnup are not satisfied. This could potentially i
increase the reactivity in spent fuel racks. To mitigate these postulated criticality related accidents, Specification 3.8.E.2 ensures the spent fuel pool contains adequate dissolved boron anytime fuel assemblies with a co=bination 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.
The boron concentration requirements of Specification 3.8.E.2 are no longer imposed when no fuel novements are occurring and a spent fuel pool verificaeion has been completed, because the storage requirements of Specifications 3.8.t.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 ec=bination of burnup and initial enrich =ent 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 T5.3.8-1 it is not possible to cause an inadvertent criticality by mispositioning a fuel assembly in the spent fuel pool.
L Prairie Island Units 1 and 2 Amendment Nos.108 and 101 I
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l B.3.8-4 3.8 REFUELING AND FUEL HANDLING j
Bases continued 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 j
of boron is restored simultaneously with suspending movement of fuel assemblies. An acceptable alternative is to complete a spent fuel pool i
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 verification is required following the last movement of fuel l
assemblies in the spent fuel pool, if fuel assemblies with a combination of I
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 l
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.l.d.
+
t a
\\
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References l
1 1.
USAR, Section 10.2.1.2 2.
USAR, Section 14.5.1
(
3.
USAR, Section 10.3.7 4.
" Criticality Analysis of the Frairie Island Units 1 & 2 Fresh and Spent Fuel I
Racks," Westinghouse Commercial Nuclear Fuel Division, February 1993.
i l
Prairie Island Unit 1 Amendment No. PJ, pp, 108
)
Prairie Island Unit 2 Amendment No. EJ, 97, 101 l
r-n,
o ** %
[ r%
b UNITED STATES 5%
NUCLEAR REGULATORY COMMISSION
%; v j W ASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 101 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 June 11, 1993, as revised June 30, 1993, 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 4
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 (1) 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-graoh 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
. Technical Specifications i
The Technical Specifications contained in Appendix A, as revised through Amendment No.101, 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.
FOR THE NUCLEAR REGULATORY COMISSION i
ldd(AtingDirector-l' W i liam
. Dean, Project Directorate 111-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: Septerter 3,1993 l
1 4
=.
I I
ATTACHMENT TO LICENSE AMENDMENT NO. 101 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 t
i Revise Appendix A Technical Specifications by removing the pages identified i
below and inserting the attached pages. The revised pages are identified by
}
amendment number and contain vertical lines indicating the area of change.
i REMOVE INSERT TS-iii TS-iii TS-vii TS-vii l
TS-xiii TS-xiii TS.3.3-1 TS.3.3-1 i
TS.3.8-4 TS.3.8-4 l
TS.3.8-5 j
Figure TS.3.8-1 Table TS.4.1-2B (Page 1 of 2) Table TS.4.1-28 (Page 1 of 2) l Table TS.4.1-2B (Page 2 of 2) Table TS.4.1-2B (Page 2. of 2)
TS.S.3-1 TS.5.3-1 TS.5.6-1 TS.S.6-1 TS.5.6-2 TS.5.6-2 TS.S.6-3 Figure TS.5.6-1 Figure TS.S.6-2 B.3.8-2 B.3.8-2 B.3.8-3 j
B.3.8-4 t
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f 1
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t l
75-111 TABLE OF C05' TENTS (Centinued)
TS SECTION TITLE PACE _
l 3.6 Containment System TS.3.6-1 A. Containment Integrity TS.3.6-1 B. Vacuum Breaker System TS.3.6-1 C. Containment Isolation Valves TS.3.6-1 D. Containment Purge System TS.3.6-2 i
E. Auxiliary Building Special Ventilation Zone Integrity TS.3.6-2 F. Auxiliary Building Special Ventilation System TS.3.6-3 l
C. Shield Building Integrity TS.3.6-3 H. Shield Building Ventilation System TS.3.6-3 I. Containment Internal Pressure TS.3.6-3 J. Contain=ent and Shield Building Air Temperature TS.3.6-4 K. Containment Shell Temperature 75.3.6-4 L. Electric Hydrogen Recombiners TS.3.6-4 M. Containment Air Locks T5.3.6-4 3.7 Auxiliary Electrical Systen TS.3.7-1 3.8 Refueling and Fuel. Handling TS.3.8-1 A. Core Alterations TS.3.8-1 l
B. Tuel Handling Operations TS.3.8-3 C. Small Spent Fuel Fool Restrictions TS.3.8-4 D. Spent Tuel Pool Special Ventilation Syste=
TS.3.8-4 E. Spent Puel Pool Storage TS.3.8-4 l
L 3.9 Radioactive Effluents T5.3.9-1 A. Liquid Effluents TS.3.9-1 s
- 1. Concentration TS.3.9-1
- 2. Dese 75.3.9 1
- 3. Liquid Radweste System TS.3.9 2 4 Liquid Storage Tanks TS.3.9-2 B. Caseous Effluencs TS.3.9-3
- 1. Dose Rate TS.3.9-3
- 2. Dose from Noble Cases TS.3.9-3
- 4. Gaseous Radvaste Treatment System and Ventilation Exhaust Treatment Systems TS.3.9-4
- 5. Containment Purging TS.3.9-5 C. Solid Radioactive Waste TS.3.9-6 i
D. Dose from All Uranium Puel Cycle Sources TS.3.9-6 E. Radioactive Liquid Effluent Monitoring Instrumentation TS.3.9-7 F. Radioactive Caseous Effluent Monitoring Instrumentation TS.3.9-7 f
i Prairie Island Unit 1 Amendment No. 9J, 97, 108 Prairie I: land Unit 2 Amendment No. Ef, PE, 101
TS-vii e
TA3LE OF CONTENTS (Continued)
TS SECTION TITLE PAC [__
5.0 DESIGN FEATURES 75.5.1-1 5.1 Site 7S.5.1 1 5.2 A.
Containment structures TS.5.2-1
- 1. Containment Vessel 75.5.2-1
- 2. Shield Building TS.5.2-2
- 3. Auxiliary Building Special Ventilation Zone B.
Special Ventilation Systems TS.S.2-2 C.
Containment System Punctional Design TS.5.2-3 5.3 Reactor TS.5.3-1 A. Reactor Core 7S.5.3-1 t
B. Reactor Coolant System TS.S.3-1 C. Protection Systems 7S.5.3-1 5.4 Engineered Safety Features 75.5.4-1 5.5 Radioactive Waste Systems 75.5.5-1 A. Accidental Releases TS.5.5-1
- 5. Routine Releases 75.5.5-1
- 1. Liquid Vastes 75.5.5-1
- 2. Caseous vastes TS.S.5-2
- 3. Solid Wastes T5.5.5-3 C. Process and Effluent Radiological Monitoring TS.5.5-3 System 5.6 Fuel Handling TS.S.6-1 A,
Criticality Consideration TS.S.6-1 B. Spent Fuel Storage Structure TS.S.6-1 C. Fu=1 Handling TS.S.6-2 D. Spent Fuel Storage Capacity TS.5.6-3 l
i i
Prairie Island Unit 1 Amendment No. 9J, pp, 108 i
Prairie Island Unit 2 Amendment No. Ef, 97, 101
?
' ' - ~
e TS-xiii APPE'CIX A TECHNICAL SPECIFICATIONS LIST OF FICl*RES i
TS TIC 17E TITLE j
2.1-1 i
Safety Linits, Reactor Core, Thermal and Hydraulic Two loop
]
Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT 1-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THEFEAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT 1-131 3.8-1 Sp nt Fuel Pool Unrestricted Region Minimum Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Caseous Effluents i
i 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4 1 Shield Building Design In Leakage Rate 5.6-1 Spent Fuel Fool Burned / Fresh Checkerboard Cell Layout 5.6 2 Spent Fuel Fool Checkerboard Region Minimum Burnup Requirements 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating Organizations 6.1 2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group i
i Prairie Island Unit 1 Amendment No. 97, JPE, 108 Prairie Island Unit 2 Amendment No. SE, PE, 101
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TS.3.3-1 j
i 3.3 ENGINEERED SAFETY FEATURES i
Arolfeability Applies to the operating status of the engineered safety features.
f Obiective i
To define those limiting conditions that are necessary for operation of i
engineered safety features:
(1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.
l Specificatiens i
A.
Saferv Inicetion and Residusi Heat Renoval Systems t
1.
A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200*F unless j
the following conditions are satisfied (except as specified in 3.3.A.2 below):
a.
The refueling water tank contains not less than 200,000 j
gallons of water with a boron concentration of at least 2500 ppm.
l 1
- b. ' Each reactor coolant system accumulator shall be OPERAELE when reactor coolanc system pressure is greater than 1000 psig.
i OPERABILITY requires:
(1) The isolation valve is opwn
-l (2) Volume is 1270 1 20 cubic feet of borated water 4
(3) A minimum boron concentration of 1900 ppm j
(4) A nitrogen cover pressure of 740 1 30 psig Two safety injection pumps are OPERABLE except that pump c.
control switches in the control room shall meet the requira-ments of Section 3.3. A.3, 3.3.A 4 and 3.1. A.1.d. (2) whenever the reactor coolant system temperature is less than 310*F*.
d.
Two res. "Lal heat removal pumps are OPERABLE.
e.
Two resid;al heat exchangers are OPERABLE.
j!
3
- Valid until 20 EFFY j
a Prairie Island Unit 1 Amendment No. 77, PJ, 108 Prairie Island Unit 2 Amendment No. 79, Ef, 101
l 1
TS.3.8 4 l
3.8.C. Small Seent Tuel Pool Restrictions No more than 45 recently discharged assemblies shall be located in the small pool (pool No. 1).
D.
Spent Puel Pool Seccial Ventilation System 1.
Both trains of the Spent Puel Pool Special Ventilation System shall be OPERA 3LI at all times (except as specffied in 3.8.D 2 and 3s8.D.3 below).
2.
With one train of the Spent Puel Pool Special Ventilation System inoperable, fuel handling operations and crane operations with loads over spent fuel (inside the spent fuel pool enclosure) are permissible during the following 7 days, provided the redundant train is demonstrated OPERAELE prior to proceeding with those operations.
3.
Vit5 both trains of the Spent Tuel Fool Special Ventilation System ino; table, suspend all fuel handling operations and crane oper.csons with leads over spent fuel (inside the spent fuel pool cr.elosure).
The provisions of specification 3.0.C are not applicable.
a.
E.
Spent Puel Pool Storare 1.
Tuel Assembly Storage To be stored without restriction in the spent fuel pool, the a.
burnup and initial enrichment ; a fuel assembly shall be within the unrestricted rsnge of Figure TS.3.8-1.
b.
Tuel assesblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 shall be stored in accordance with Specification 5.6..
1.d.
t If the requirements of 3.8.E.1.a and 3.8.E.1.b are not met, c.
im=ediately initiate action to move any noncomplying fuel assesbly to an acceptable location.
d.
The provisions of Specification 3.0.C are not applicable.
i i
Prairie Island Unit 1 Amendment No. pp, SJ, 108 s
Prairie Island Unit 2 Amendment No. S), SJ,101
- - ~
~ --
l TS.3.8-5 l
?
l 4
3.8.E.2. Spent Puel Pool Boron Concentration The spent fuel pool boron concentration shall be k 1,800 ppm I
a.
when 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 and a spent fuel pool 4
verification has not been performed since the last movement of any fuel assembly in the spent fuel pool.
j b.
If the requirements of specification 3.8.E.2 a are applicable anc the spent fuel pool boron concentration is not within its limit, then immediately.
i 1.
Suspend novement of fuel assemblies in the spent fuel pool, and i
2.
Either initiate action to restore spent fuel pool boren concentration to within its limit or perform a spent fuel
{
pool verification.
c.
The provisions of Specification 3.0.C are not applicable.
i i
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I f
k t
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Prairie Island Unt 1 Amendment No. 108 Prairie Island Utr 2 Amendment No. 101 l
i
~-
I FIGURE 75.3.8 1 12000 11000 i
/
10000 1
/
r
/4 OD~l KID
/
g
/
^
I I
A
/
g
/
/
v 7000
/
t-R
/
R
/
g
/
sa2 M
/
u
, f f
f/
G 5000 e
-a
/
/
h
/f X
/ 4 1
y (D)Q
/
7 l
E f
/
/
M A KIM A
/
/
/
/
2003 1
Z~i' 1000 3.5 4.0 4.5 5.0 INITIAL NOMINAL U-235 ENPJCEMENT (w/o)
FIGUPI TS.3.8-1 Spent Fuel Fool Unrestricted P.sgion Minimum Burnup P.equirements Prairie Island 'J-*'t 1
Amendment No. 108 Prairie Island Unit 2 Amendment No. 101
Table TS.4.1-2B (Paga l' of 2)
TABl.E TS.4.1-2B MINIMUM TREOUENCIES FOR SAMPLING TESTS FSAR Section TEST TREOUENCY Reference 1.
RCS Gross 5/ week Activity Determination 2.
RCS Isotopic Analysis for DOSE 1/14 days (when at power)
EQUIVA1.ENT I-131 Concentration I
l 3.
RCS Radioche=istry E determination 1/6 months (1) (when at power) 4 RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including I-131, 1-133, and I-135 the specific activity ex-caeds 1.0 uCi/ gram DOSE _
EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following thermal PO'T.R change exceeding 15 percent of the RATED THERMAL POWER vithin a one hour period ( above hot shutdown) 5.
RCS Radiochenistry (2)
Monthly 6.
RCS Tritium Activity Weekly 7.
RCS Chemistry (Cl*,F*, 02) -
5/Veek E.
RCS Boron Concentration *(3) 2/ Week (4) 9.2 l
9.
RVST Boron Concentration Weekly j
- 20. Boric Acid Tanks Boron Concentration 2/ Week
- 11. Caustic Standpipe NaOH Concentration Monthly 6.4 l
- 22. Accumuistor Boron Concentre. tion Monthly 6
- 13. Spent Fuel Pit Boron Concentration Monthly /WeeklyNU 9.5.5 Y
Prairie Island Unit 1 Amendment No. 57, PP, 108 Prairie Island Unit 2 Amendment No. fp, P7, 101
Tchio TS.4.1 2B (Page 2 of 2)
I l
TABLE TS.4.1-2B l
~
MINIMUM TREQUENCIES FOR SAMPLING TESTS FSAR Section TEST FREOUENCY Reference 14 Secondary Coolant Cross Weekly Beta-Ca==a activity 15.
Secondary coolant Isotopic 1/6 months (5)
Analysis for DOSE EQUIVALENT I 131 concentration 16.
Secondary Coolant Chemir tyy pH 5/ week (6) pH Control Additive
$/ week (6)
Sodium 5/ week (6)
Notes-1.
Sample to be taken after a minimum of 2 EFFD and 20 days of pJ.'ER OPERATION have e. lapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
2.
To determine activity of corrosion products having a half-life greater than 30 minutes.
3.
During RETUELING, the boron concentration shall be verified by chemical analysis daily.
4.
The maximum interval between analyses shall not exceed 5 days.
5.
If activity of the samples is greater than 10% of the limit in Specification 3.4.D, the frequency shall be once per month.
6.
The maximum interval between analyses shall not exceed 3 days, 7.
The minimum spent fuel pool boron concentration from Specification 3.8.B.1.b shall be verified by chemical analysis weekly while a spent fuel cask containing fuel is located in the spent fuel pool.
B.
The spent fuel pool boron concentratier. shall be verified weekly, by chemical analysis, to be within the limits of Specification 3.8.E.2.a when fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure 7S.3.8-1 are stored in the spent 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.
Prairie Island Unit 1 Amendment No. PJ, 99, 108 Prairie Island Unit 2 Amendment No. Ef, 92, 101
TS.5.3-1 5.3 REACTOR A.
Reactor Core 1.
The reactor core contains uranium in the form of natural or slightly l
enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 or ZIRLD tubing to form fuel rods. The reactor core is made l up of 121 fuel assemblies. Each fuel assembly contains 179 fuel rods (Reference 1).
2.
The maximum enrichment will be 5.0 veight percent U-235.
3.
In the reactor core, there are 79 full-length RCC assemblies that contain a 142-inch length of silver-indium cad =ium alloy clad with stainless steel (Reference 2).
B.
Resetor Coolant Svilig 1.
The design of the reactor coolant system complies with all appli-cable code requirements (Reference 3).
2.
i.11 high pressure piping, co=ponents of the reactor coolant syste= and their supporting structures are designed to Class I requirements, and have been designed to withstand:
The design seis=ic ground acceleration, 0.06g acting in the a.
horizontal and 0.04g acting in the vertical planes simultane-ously, with stresses maintained within code allowable verking stresses.
b.
The maximu= 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 no=inal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.
C. Protection Systees The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1s68. The design includes a reactor trip for a high negative rate of change of neutron flux as measured by the excore nuclear instruments (Reference 4).
The system is intended to trip the reactor upon the abnormal dropping of more than one control rod (Reference 4).
If only one 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, Section 3.4.2 3.
USAR, Table 4.1-11 2.
USAR, Section 3.5.2 4
USAR, Section 7.1 Prairie Island Unit 1 Amendment No. EP, 90, 108 Prairie Island Unit 2 Amendment No. 77, EJ, 101
b TS.5.6-1 5.6 FUEL HANDLING A.
Criticality Consideratip.D 1.
The spent fuel storage racks are designed (Reference 1) and shall be maintained with:
Fuel assemblies having a maximum U 235 enrichment of 5.0 veight a.
percent; b.
14rr s 0.95 if fully flooded with unborated water, which iticludes an allowance for uncertainties as described in Reference 2; Now or spent fuel assemblies with a combination of burnup and c.
initial enrichment in the unrestricted range of Figure 75.3.8-1 allowed unrestricted storage in the spent fuel racks; and d.
New or spent fuel assemblies with a combination of burnup and 1
initial enrichment in the restricted range of Figure 75.3.8 1 secred in co=pliance with Figures TS.5.6-1 and TS.5.6-2.
2.
The new fuel storage racks are designed (Reference 1) and shall be maintained with:
Fuel assemblies havinE a maximum U 235 enrichment of 5.0 veight a.
percent; b.
K,tr s 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference 2; and r
c.
K.tr s 0.98 if accidentally filled with a lov density moderator which resulted in optimum low density moderation conditions.
3.
Fuel vill not be inserted into a spent fuel cask in the pool, unless a mini =um boren concentration of 1800 ppa is present. The 1800 ppm vill ensure that k rt for the spent fuel cask, including statistical o
uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask. The criticality analysis for the TN 40 spent fuel storage cask was based on fresh fuel enriched to 3.85 vei ht percent U-235.
E B.
Seent Puel Storare Structure The spent fuel 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 are Class 1 (seismic) structures that afford protection against loss of integrity from postulated tornado I
missiles. The storage compartments and the fuel transfer canal are connected by fuel transfer slots that can he 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.
Prairie Island Unit 1 Amendment No. pp, pp, 108 Prairie Island Unit 2 Amendment No. S7, PE, 101
)
75.5.6-3 6
The spent fuel pool has a reinforced concrete botton slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident. Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies.
The new and spent fuel pit structures are designed to withstand th.
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.
C.
Fuel Handline 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 co=ponents of the fuel handling system are the manipulation crane, 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 liftin5 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 vill 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. Af ter sufficient decay, the fuel vill be loaded into storage casks fer storage in the Independent Spent Fuel Storage Installation or into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane.
Spent fuel casks vill be handled by a single fai'.cre proof handling system meeting the requirements of Section 5.1.6 r.f NULIG-0612, " Control of Heavy Loads at Nuclear Power Plants", July 1980. CL: auxiliary building crane has been upgraded to conform with the single failure proof requirements of Se: tion 5.1.6 of NURIG 0612. The auxiliary building crane is designed to not allow a load drop as a result of any single failure. The improved reliability of the auxiliary building crane is achieved through increased factors of safety and through redundancy or duality in certain active components.
Prairie Island Unit 1 Amendment No. pp, 99, 108 Prairie Island Unit 2 Amendment No. SJ, 92, 101
L a
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TS.S.6-3 D.
Spent Fuel Storare Capac(ty 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 spent fuel cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor.
i r
l References 1.
USAR, Section 10.2 2.
" Criticality Analysis of the Prairie Island Units 16 2 Fresh and Spent Fuel Racks," Westinghouse Commercial Nuclear Fuel Division, February 1993.
Prairie Island Unit 1 Amendment No. pp, 99, 108 Prairie Island Unit 2 Amendment No. 57, 97, 101
FIGURE 75.5.6-1 rT i-n
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separated by a single row of vacant cells on each adjacent face.
FICURE 7S.5.6-1 Spent Fuel Fool Burned / Fresh Checkerboard Cell Layout Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 108 Amendment No. 101
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4.5 5.0 IRITIAL NOMINAL U-235 ENRICEMENT (w/o)
I TICURI TS.5.6-2 Spent Fuel Fool checkerboard Region Minimum Burnup Requirements 1
Prairie Island Unit 1 Amendment No. 108 Prairie Island Unit 2 Amendment No. 101 I
i
B.3.8-2
=
3.8 REFUELING AND FUEL HANDLING Esses continued During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provida 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 asergency procedures to cool the core, (2) during latching /unistching and upper internals removal / replace-ment 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 veight percent U-235 while maintaining )ber s 0.95 including uneartainties. 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 asse=blies with all possible combinations of burnup and initial enrichsent:
1.
The first region utilizes a checkerboard loading pattern to accommodate new or low burnup fuel with a maximum enrichment of 5.0 vtt 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 vtt 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 vet 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 the restricted range of Figure TS.3.8-1 must be stored in the checkerboard region in accordance with Specification 5.6.A.1.d.
The burned / fresh fuel checkerboard region can be positioned anywhere within the spent fuel racks, but the boundary between the checkerboard region and the unrestricted reglen 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 75.5.6-1).
Prairie Island Unit 1 Amendment No. PJ, PS, 108 Prairie Island Unit 2 Amendment No. EJ, PE, 101
t B.3.8-3 3.8 PEFUELING AND WEL FANDLING Eases continued 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 veight percent U-235.
Enrichsents lower than 3.87 weight percent are conservatively bounded by the minimum burnup requirement for 3.87 weight percent U 235 which is 2000 M7D/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 M7D/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 suberiticality 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 er of 0.95 be e
evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which ensures that each region is maintained in a suberitical condition during normal operation with the regions fully leaded.
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 assu=e two unlikely, independent, concurrent events to ensure protection against a criticality accident.
The double contingency principle allows credit for soluble boron under abnor=al 4
or accident conditions, since only a single accident need be considered at one ti=e.
For exacple, the most severe accident scenario is the accidental
=isleadin5 of a fuel asse=bly into a rack location for which the restrictions on location, enrichment or burnup are not satisfied. This could potentially 4
increase the reactivity in spent fuel racks. To mitigate these postulated criticality related accidents, Specification 3.8.E.2 ensures 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 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 1
the increased reactivity caused by a mispositioned fuel assembly.
The boron concentration requirements of Specification 3.8.E.2 are r.o longer impesed when no fuel movements are occurring and a spent iuel 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 ec=bination of burnup and initial enrichsent 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.s are not required in that case because with only fuel asse=blies 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.
Prairie Island Units 1 and 2 Amendment Nos.108 and 101
{
i e
l B.3.8-4 3.8 REFUELING A'?D FUEL HANDLING Bases continued 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 i
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 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.
i t
i Peferences 1.
USAR, Section 10.2.1.2 i
2.
USAR, Section 14.5.1 3.
USAR, Section 10.3.7 l
4
- Criticality Analysis of the Prairie Island Units 1 6 2 Fresh and Spent Fuel i
Racks," Vestinghouse Commercial Nuclear Fuel Division, February 1993.
l i
Prairit: Island Unit 1 Amendment No. SJ, 97, 108 Prairia Island Unit 2 Amendment No. EJ, 92, 101
.