ML20138L978
| ML20138L978 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/12/1985 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20138L980 | List: |
| References | |
| NUDOCS 8512200157 | |
| Download: ML20138L978 (20) | |
Text
.
/
4 UNITED STATES NUCLEAR REGULATORY COMMISSION L
- E WASHINGTON, D. C. 20555
%,...../
COP 940NWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 91 License No. DPR-19 l
1.
The Nuclear Regulatory Commission (the Commission)-has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated August 13, 1985 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.
l 8512200157 851212 PDR ADOCK 05000237 p
4.
6 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraphs 3.B and 3.N of Provisional Operating License No. DPR-19 are hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.91, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
N.
Deleted.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATO COMMISSION John
. Zwolinski, Director BWR Pioject Directorate #1 Division of BWR Licensing
Attachment:
Changes to License No. DPR-19 and the Technical Specifications Date of Issuance: December 12, 1985 N
D
1 l
ATTACHMENT TO LICENSE AMENDMENT NO. 91 PROVISIONAL OPERATING LICENSE DPR-19 DOCKET NO. 50-237 1.
For your convenience a revised copy of page 6 to Provisional Operating License DPR-19 is attached.
For administrative purposes the text on page 7 has been relocated to page 6.
2.
Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3/4.10-8 B 3/4.10-8 8 3/4.10-9*
B 3/4.10-9 8 3/4.10-10*
B 3/4.10-10 B 3/4.10-11*
B 3/4.10-11 B 3/4.10-12 5-1 5-1 m
~
h
- Pagination change only
.---,-----e
Amendment No. 91
.- DPR-19 M.
3.
The MCPR Limiting Condition for Operation (LCO) will be increased 0.03 (TS 3.5.K and Fig. 3.5-2)
Am. 75 4/07/83 4.
The Maximum Average Planar Linear Heat Generation' Rate (MAPLHGR) limits will be reduced to 70% of current values for all fuel types.
(T.S. reference 3.5.1) 5.
The APRM Scram and Rod Block Setpoints and the RBM j
Setpoints shall be reduced by 3.5% to read as follows:
Am. 75 T.S. 2.1.A.1 8 s (.58 WD + 58.5) 4/07/83 T.S. 2.1.A.1* 3 1 (.58 WD + 58.5) FRP/MFLPD Corrected by T.S. 2.1.3 8 s (.58 WD + 46.5)
Letter dated T.S. 2.1.B*
8 s (.58 WD + 46.5) FRP/MFLPD 10/5/83 T.S. 3.2.C (Table 3.2.3):
APRM Upscale s (.58 WD + 46.5) FRP/MFLPD RBM Upscale 1 (.65 WD + 41.5)
- In the event that NFLPD exceeds FRP for General Electric fuel.
Am. 63 6.
The suction valve in the idle loop is closed and 7/9/81 electrically isolated until the idle loop is being prepared for return to service.
7.
APRM flux moise will be measured once per shift and the recirculation pump speed will be reduced if the flux moise averaged over 1/2 hour exceeds 5% peak to peak, as measured on the APRM chart recorder.
I 8.
The core plate delta p noise will be measured once per shift and the recirculation pump speed will be reduced if the moise exceeds 1 psi peak to peak.
N.
Deleted.
i 4.
This license is effective as of the date of issuance and shall I
DRL expire December 22, 1972, unless extended for good cause shown, Order or upon the earlier issurance of a superseding operating 6/10/71 license.
FOR THE ATONIC ENERGY CONNISSION Original Signed by Peter A. Morris, Director Division of Reactor Licensing
Attachment:
Appendix A - Technical Specifications Date of Issuance: December 22, 1969 5039N 8403D
^
i DRESDEN II DPR-19 i
Amendment No. $d, 91 l
l l
3.10 MgITING CONDITIONS FOR OPERATION 4.10 SURVEILLANCE REQUIREMENTS
(
(Cont'd.)
(Cont'd.)
)
1 f
~
f C.
Fuel Storage Reactivity Limit C.
Fuel Storage' Reactivity Limit 1
1.
The new fuel storage 1.
Prior to storing Fuel in the f
facility shall be such new fuel storage facility, i
that the E gg dry is an analysis must be
[
1ess than 0.90 and flooded performed to demonstrate j
is less than 0.95.
that the criteria in i
I 3.10.C.1 are satisfied.
2.
Whenever a fuel assembly is 2.
Pelor to storing Fuel in the i
stored in the spent fuel the spent fuel storage pool, I
storage pool, the peak assembly an analysis must be reactivity in a reactor lattlee performed to demonstrate distribution shall be limited that the criteria in l
to less than or equal to the 3.10.C.2 are satisfied.
following values:
Assembly Type E nf i
GE 7z7 1.26 GE 8x8 1.32 ENC 8x8 1.33 INC 9E9 1.27 Whenever storing other assembly types or fuel rods in the spent fuel storage pool, their peak reactivity shall be bounded by the most limiting Eint value listed above.
r I.
Loads Over Spent Fuel Storage Fool 1
No loads heavier than the waight of l
a single spent fuel assembly and j
handling tool shall be carried over
{
fuel stored in the spent fuel l
storage pool.
l l
i
.~
I l
3/4.10-8 3694a I
~
3124A l
\\
l
\\
l l
t DRgSDEN II DPR-19
)
AmsAdment No. p2,91 p',
3.10 LIMITING CONDITION FOR OPERATION BASFJ A.
Refueling Interlocks During refueling operations, the reactivity potential of the core _is being altered.
1.t is.necessary to require certain interlocks and restrict ~certain refueling procedures such that there is assurance that inadvertent criticality does not occur.
To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality. The core reactivity limitation of specifications 3.2 lihits the core alterations to assure that the resulting core loading can be controlled with the reactivity control system and interlocks at any time during shutdown or the following operating cycle.
Addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up l
by refuellag interlocks on' rod withdrawal and movement of the refueling platform. When the mode switch is in the " Refuel" position, interlocks prevent the refueling platform from being I
moved over 1.he core if a control rod is withdrawn and fuel is i
on a hoist. Likewise, if the refueling platform is over the coca with fuel on a hoist, control rod agtfon is blocked by tha interlocks. With the mode switch in the refuel position only one control rod can be withdrawn.
For a new core the dropping of a fuel assembly into a vacant i
l fuel location adjacent to a wit *4diawn control rod does not l
result In an excursion or a eritical configuration, thus adequate margin is provided.
{
3.
Core Monitorieg t
l The EEM's are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup. Requiring two operable ERM's i
.in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. Requiring a miniana of 3 counts pei*
second whenever criticality is possible provides assurance that eutron fluz is being monitored. Criticality is con-siderei to be tapossible ifsthere are no more than two assen-blies la a guadrant and if thsee are in locations adjacent to the SEk. In this case only, the 334 or dunking type detector eguetrateispermittedtobelessthan3countspersecond.
B 3/4.10-9
~
36941 3124A t
3 1
DRESDEN II DPR-19 Amendment No. pf, 91 3.10 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
C.
Fuel Storage Pool Water Level To assure that there is adequate water to shield and cool the irradiated fuel assemblies stored in the pool, a minimum pool water level is established. The minimum water level of'33 feet is established because it would be a significant change from the normal level (37'9") well above a level to assure adequate cooling (just above active fuel) and above the level at which the CSEP action is initiated (5' ancontrolled loss of level with level decreasing).
D.
During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same
~
time. This specification provides assurance that inadvertent criticality does not occur during such maintenance.
The maintenance is performed with the mode switch in the re-fuel" position to provide the re-fueling interlocks a
normally available during re-fueling operations as explained in Part A of these Bases. In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the re-fueling interlock on the first control rod I
which prevents more than one control rod from being withdrawn I
at the same* time. The requirement that an adequate shutdown i
margin be demonstrated with the control rods remaining in service insures that inadvertent criticality cannot occur during this maintenance. The Shutdown margin is verified by demonstrating that the core is shut down even if the strongest 7
control rod remaining in service is fully withdrawn.
i, l
Disarming the directional control valves does not inhibit control rod scram capability.
6 The requirement for SEN operability during the maintenance is i
covered in part 3 of these Bases.
E.
The intent of this specification is to permit the unloading of a significant portion of the reactor core for such purposes as in-service inspection requirements, examination of the core support plate, etc. This specification provides assurance
.that inadvertent criticality does not occur during such operation.
This operation is performed with the mode switch in the "re-fuel" position to provide the re-fueling interlocks normally available darias re-fueling as explaimed la part A of these Bases. In order to withdraw more than one control rod, it is necessary to bypass the re-fueltag interlock.oa each B 3/4.10-10 3694a 3124A
DRESDEN II DPR-19 Amendment No. pf, 91 3.10 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
withdrawn control rod which prevents more than one control rod from being withdrawn at a time. The requirement that the fuel assemblies in the cell controlled by the contror rod be removed from the reactor. core before the inter, locks can be bypassed insures that withdrawal of another control rod does not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with that control rod.
Thus, removal of an entire cell (fuel assemblies plus control rod).results in a lower reactivity potential of the core.
The requirement for SEN operability during these operations is covered in part B of these Bases.
F.
The operation of the redundant crane in the Restricted Mode during fuel cask bandling operations assures that the cask remains within the controlled area once it has been removed from its transport vehicle (i.e., once it is above the 545' elevation). Mandling of the cask on the Refueling Floor in the Unrestricted Mode is allowed only in the case of equipment i
failures or emergency conditions when the cask is already suspended. The Unrestricted Mode of operation is allowed only to the extent necessary to get the cask to a suitable stationary position so the required repairs can be made.
Operation with a failed controlled area microswitch will be allowed for a 48-hour period providing an Operator is on the 3
floor in addition to the crane operator to assure that the i-cask handling is limited to the controlled area as marked on
~
1 the floor. This will allow adequate time to make repairs but still will not restrict cask handling operations saduly.
The Surveillance Requirements specified assure that the redundant crane is adequately inspected in accordance with the accepted ANSI Standard (3.30.2.0) and manufacturer's reconnendations to determine that the equipment is in satisfactory condition. The testing of the controlled area 35mit switches assures that the crane operation will be l
limited to the designated area in the Restricted Mode of operation. The test of the "two-block" limit switch assures I
the power to the hoisting motor will be interrupted before an actus1 "two-blocking" incident can occur. The test of the inching hoist assures that this mode of load control is available when requirac.
Requiring the lifting and holding of the cask for 5 minutes during the initial lift of each series of eask handling 3 3/4.10-11 3694a 3124A
i DRESDEN II DPR-19 Amendment No. pf, 91 3.10 LIMITING CONDITION FOR OPEKATION BASES (Cont'd.)
operations. puts a load test on the entire crans lifting mechanism as well as the braking system.
Performing this test when the cask is being lifted initially from the cask car assures that the system is operable prior to lifting the load to an excessive height.
I G.
The acceptance criteria for the reactivity in the new fuel storage vault (K,gg dry < 0.90 and K,gg flooded < 0.95) and in the spent fuel storage pool (K,gg 1 0.95) are based on the overall uncertainties associated with the calculational methods. These limits provide sufficient margin to prevent criticality.
To avert any unreviewed increase, or increased uncertainty, in the calculated K,gg in the spent fuel storage pool above the limit of K,gg 1 0.95, a limit on the calculated peak reactivity of the fuel assembly design when loaded in an infinite array in the reactor lattice (Einf) is required.
1 l
The peak assembly reactor lattice Eint is adequate in j
protecting the limit of K,gg s 0.95, since it factors in the effects of'U-235 loading, burnable poisons loading, and other
/
minor assembly perturbations. Therefore, if assmablies stor'ed in the spent fuel storage pool have a peak assembly reactor lattice Eint less than or equal to the corresponding values in Section 3.10.G.2., then the spent fuel storage pool reactivity will be less than or equal to 0.95.
I H.
By prohibiting the movement of loads with weight in excess of the nominal weight of a fuel assembly and handling tool over i
fuel assemblies in the spent fuel storage. pool, so more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident.
4.10 SURVEILLANCE REQUIRENINT SASES None i
l l
B 3/4.10-12 7
3694a 3124A
DRESDEN II DPR-19 Amendment No. 92, 94, 91 i
5.0 DESIGN FEATURES 5.1 Site i
I Dresden Unit 2 is located at the Dresden Nuclear Power Station which consists of a tract of land of approximately 953 acres located in the northeast quarter of the Morris 15-minute quadrangle (as designated by the United States Geological Survey), Goose Lake Township Grundy l
County, Illinois. The tract is situated in portions of Sections 25, 26, 27, 34, 35, and 36 of Township 34 North, Range 8 East of the Third Principal Meridian.
5.2 Reactor A.
The core shall consist of not more than 724 fuel assemblies B.
The reactor core shall contain 177 cruciform-shaped control rods.
The control material shall be boron carbide powder (B C) compacted 4
i to approximately 70% of theoretical density, or Hafnium metal.
5.3 Reactor Vessel The reactor vessel shall be as described in Table 4.1.1 of the SAR.
The applicable design codes shall be as described in Table 4.1.1 of the SAR.
5.4 Containment A.
The principal design parameters and applicable design codes for the primary containment shall be as given in Table 5.2.1 of the SAR.
The secondary containment shall be as described in Section 5.3.2 of B.
the SAR and the applicable codes shall be as described in Section 12.1.1.3 of the SAR.
C.
Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set i
forth in Section 5.2.2 of the SAR and the applicable codes shall be i
as described in Section 12.1.1.3 of the SAR.
5.5 Fuel Storage l
A.
A maximum of 320 new fuel assemblies are permitted to be stored in l
the new fuel storage facility. New fuel storage reactivity limits are specified in Section 3.10.G.I.
I B.
Fuel storage is permitted in up to 33 high density fuel storage racks. This will allow a storage capacity of 3537 fuel assemblies.
Spent fuel storage reactivity limits are specified in Section 3.10.G.2.
5-1 3696a 3124A
o,,
UNITED STATES
-8 NUCLEAR REGULATORY COMMISSION p,
g
- j WASHINGTON D. C. 20555
't, * *... sd COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.85 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated August 13, 1985 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 l
common defense and security or to the health and safety of the public; and l
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.
r
l i.
l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraphs 3.B and 3.M of Facility Operating License No. DPR-25 are hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 85, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
M.
Deleted.
~
3.
This license amendment is effective as of the date of its issuance.
FOR TH NUCLEAR REGULATORY COMMISSION
-1k w]'
John Zwolinski, Director BWR P ject Directorate #1 Division of BWR Licensing
Attachment:
Changes to License DPR-25 and the Technical Specifications
~
Date of Issuance: December 12, 1985 S
.a l
1
ATTACHMENT TO LICENSE AMENDMENT NO. 85 FACILITY OPERATING LICENSE DPR-25 DOCKET NO. 50-249 1.
For your convenience a revised copy of page 6 to Provisional Operating License DPR-19 is attached.
For administrative purposes the text on page 7 has been relocated to page 6.
2.
Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3/4.10-8 B 3/4.10-8 B 3/4.10-9*
B 3/4.10-9 B 3/4.10-10*
B 3/4.10-10 B 3/4.10-11*
B 3/4.10-11 B 3/4.10-12 5-1 5-1 i
t
- Pagination change only 1
Amendment NoJB5
' DPR-25 1
L.
5.
The APRM Scram and Rod Block Setpoints and the RBN Setpoints shall be reduced by 3.5% to read as 1
follows:
s Am. 63 T.S. 2.1.A.1 8 s (.58 WD + 58.5) 4/29/82 T.S. 2.1.A.1* 8 s (.58 WD + 58.5) FRP/MFLPD T.S. 2.1.5 s s (.58 WD + 46:5) 7.8.-2.1.B*
8 s (.58 WD + 46.5) FEP/MFLPD T.S. 3.2.C (Table 3.2.3):
APRM Upscale 1 (.58 WD + 46.5) FRP/MFLPD i
RBN Upscale s (.58 WD + 41.5)
- In the event that NFLPD exceeds FRP.
6.
The suction valve in the idle loop is closed and electrically isolated until the idle loop is being prepared for return to service.
7.
APEN flux moise will be measured once per shift and the recirculation pump speed will be reduced Am. 54 if the fluz moise averages ovee 1/2 hour exceeds 7/9/81 5% peak to peak, as measured or the APRM chart recorder.
8.
The core plate delta y noise will be measured once per shift and the recirculation pump speed will be reduced if the noise exceeds 1 psi peak to peak.
M.
Deleted (Renumbered) 4.
This license is effective as of the date of issuance, (Per Am. 2) and shall expire at Midnight October 14, 2005.
(1-26-73)
FOR TNE ATONIC ENERGY COMNISSION Original Signed by Peter A. Norris Director Division of Reactor Licensing
Enclosures:
Appendix A - Technical specifications Date of Issvances, January 12, 1971 5039N 8403D
l DRESDEN III PPR-25 Amendment No. ?$ 85 3.10 LIMITING CONDITIONS FOR OPERATION 4.10 SURVEILLANCE REOUIREMENTS (Cont'd.)
(Cont'd.)
)
t C.
Fuel Storage Reactivity Limit C.
Fuel Storage Reactivity Limit 1.
The new fuel' storage 1.
Prior 'to storing Fuel in the facility shall be such new fuel storage facility, that the K,gg dry is an analysis must be less than 0.90 and flooded performed to demonstrate
{
is less than 0.95.
that the criteria in 3.10.G.1 are satisfied.
2.
Whenever a fuel assembly is 2.
Prior to storing Fuel in the stored in the spent fuel the spent fuel storage pool, storage pool, the peak assembly an analysis mest be reactivity in a reactor lattice performed to demonstrate distribution shall be limited that the criteria in to less than or equal to the 3.10.G.2 are satisfied.
following values:
Assembly Type E nf i
GE 7x7 1.26 CE Sz8 1.32 ENC Ss8 1.33 j
ENC,929 1.27
)_
Whenever storing other assembly types or fuel rods in the spent
'l '
fuel storage pool, their peak it reactivity shall be bounded by the most limiting Kinf value listed above.
l N.
Loads Over Spent Fuel Storage pool No loads heavier than the weight of a single spent fuel assembly and handling tool shall be carried over fuel stored in the spent fuel storage pool.
i l
.~
3/4.10-8 3901a 3124&
DRgSDEN III DPE-25 Amendment No. /,85 3.10 LIMITING CONDITION FOR OPERATION BASES A.
Refueling Interlocks l
During refueling operations,'the reactivity potential'of the core is being altered. It is necessary to require certain interlocks and restrict certain refueling procedures such that there is assurance that inadvertent criticality does not occur.
To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core.
This requirement assures that during refueling the refueltag interlocks, as designed, will prevent inadvertent criticality. The core reactivity limitation of specifications 3.2 limits the core alterations to assure that the resulting core loading can be controlled with the reactivity control system and interlocks at any time during shutdown or the following operating cycle.
Addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling
- platform. When the rode switch is in the " Refuel"
/
position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist. Likewise, if the refuellas platform is ovet the core with fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the refuel position only one control rod can be withdrawn.
For a new core the dropping of a fuel assembly into a vacant i
fuel location adjacent to a withdrawn control rod does not l
result la an escursion or a critical configuration, thus adequate margin is provided.
\\
3.
Core Monitoring The sEM's are provided to monitor the core during periods of station shsidown and to guide the operator during refueling operations and station startup.
Requiring two operable stM's in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations.
Requiring a minimum of 3 counts per second whenever criticality is possible provides assurance that mostron fluz is belas monitored.
Criticality is son-sidered to be impossible if there are no more than'two assen-blies in a quadrant and if these are in locations adjacent to the BRM.
In this case only, the SRM or dunking type detector count rate is permitted to be less than 3 counts per second.
B 3/4.10-9 3901a 3124A
. - - - -.--~
~-
DRESDEN III DPE-25 l
Amendment No. }f,85 3.10 J,IMITING CONDITION FOR OPERATION BASES (Cont'd.)
C.
Fuel Storage Fool Water Level To assure that there is adequate water to shield and cool the irradiated fuel assemblies stored in the pool, a minimum pool water level is established. The minimum water level of 33 feet is established because it would be a significant change from the normal level (37'9") well above a level to assure adequate cooling (just above active fuel) and above the level at which the GSEP action is initiated (5' uncontrolled loss of level with level decreasing).
6.
During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same
~
time. This specification provides assurance that inadvertent criticality does not occur during such maintenance.
~
The maintenance is performed with the mode switch in the "re-fuel" position to provide the re-fueling interlocks normally available during re-fueling operations as explained in Part A of these Bases. In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass,the re-fueling interlock on the first control rod which prevents more than one control rod from being withdrawn
/
at the same time. The requirement that an adequate shutdown margin be demonstrated with the control rods remaining in service insures that inadvertent criticality cannot occur during this maintenance. The shutdown margin is verified by
~
demonstrating that the core is shut down even if the strongest l'
control red remaining in service is fully withdrawn.
Disarming the directional control valves does not inhibit control rod scram capability.
The requirement for SEN operability during the maintenance is covered in part 3 of these Bases.
E.
The intent of this speelfication is to permit the unloading of a significant portion of the reactor core for such purposes as in-service inspection requirements, examination of the core support plate, etc. This specification provides assurance that inadvertent criticality does not occur during such operation.
This operation is performed with the mode switch in the "re-fuel" position to provide the re-fueling interlocks normallyavailableduringre-fuelingasexplainedJapartAof these Bases. In order to withdraw more than one control rod, it is necessary to bypass the re-fueling interlock on each B 3/4.10-10 i
3901a i
3124A
DRESDEN III DPR-25 Amendment No pA85 i
i 3.10 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
withdrawn control rod which prevents more than one control rod from being withdrawn at a time. The requirement that.the fuel assemblies in the cell controlled by the control rod be i
removed from the reactor core before the interlocks can be
^
bypassed insures that withdrawal of another control rod does
{
1 not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with that control rod.
Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.
i The requirement for BRM operability during these operations is covered in Part 3 of these Bases.
i F.
The operation of the redundant crane in the Restricted Mode during fuel cask handling operations assures that the cask remains within the controlled area once it has been removed from its transport vehicle (i.e., once it is above the 545' elevation).
[
Mandling of the cask on the Refueling Floor in the Unrestricted Mode is allowed only in the case of equipment failures or emergency conditions when the cask is already suspended.' The Unrestricted Mode of operation is allowed only to the extent necessary to get the cask to a suitable stationary position so the required repairs can be made.
Operation with a failed controlled area microswitch will be allowed for a 48-hour period providing an Operator is on the floor in addition to the crane operator to assure that the cask handling is limited to the controlled area as marked on the floor. This will allow adequate time to make repairs but still will not restrict cask handling operations naduly.
The Surveillance Requirements specified assure that the redundant crane is adequately inspected in accordance with the accepted ANSI Standard (B.30.2.0) and manufacturer's recommendations to determine that the equipment is in satisfactory condition. The testing of the controlled area limit switches assures that the crane operation will be
. limited to the designated area in the Restricted Mode of operation. The test of the "two-block" limit switch assures the power to the hoisting motor will be interrupted before an actual "two-blocking" incident can occur. The test of the inching hoist assures that this mode of lead sentrol is available when required.
Requiring the lifting and holding of the cask for 5 minutes during the initial lift of each series of cask hand 11ag 3 3/4.10-11 3901a 3124A
- ~.
1 DRESDEN III DPR-25 Amendment No. %, 85 3.10 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
operations puts a load test on the entire crane lifting mechanism as well as the braking system.
Performing this test when the cask is being lifted initially 4
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from the cask car assures that the system is operable prior to lifting the load to an excessive height.
G.
The acceptance criteria for the reactivity in the new fuel storage vault (K,gg dry < 0.90 and K,gg flooded < 0.95) and in the spent fuel storage pool (K,gg 1 0.95) are based on the overall uncertainties associated with the calculational methods. These limits provide sufficient margin to prevent critical,ity.
To avert any unreviewed increase, or increased uncertainty, in the calculated K,gg in the spent fuel storage pool above the limit of K,gg 1 0.95, a limit on the calculated peak reactivity of the fuel assembly design when loaded in an infinite array in the reactor lattice (Kinf) is required.
The peak assembly reactor lattice Einf is adequate in protecting the limit of K,gg 1 0.95, since it factors in the effects of U-235 loading, burnable poisons loading, and other
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minor assembly perturbations. Therefore, if assemblies stcred in the spent fuel storage pool have a peak assembly reactor lattice Einf less than or equal to the corresponding values in Section 3.10.G.2., then the spent fuel storage pool 7
reactivity will be less than or equal to 0.95.
E.
By prohibiting the movement of loads with weight in excess of the nominal weight of a fuel assembly and handling tool over fuel assemblies in the spent fuel storage pool, ao more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident.
4.10 SURVEILLANCE REQUIREIGNT BASES l
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B 3/4.10-12 3901a 3124A
DRESDEN III DpR-25 Amendment No. 7,4, 76, 85 5.0 DESIGN FEATURES 5.1 Site Dresden Unit 3 is located at the Dresden Nuclear power Station w~ich consists of a tract of land of approximately 953 acres located in the norcheast quarter of the Morris 15-minute quadrangle (as designated by the United States Geological Survey),
Goose Lake Township, Grundy County, Illinois. The tract is l
situated in portions of Sections 25, 26, 27, 34, 35, and 36 of Township 34 North, Range 8 East of the Third principal Meridian.
5.2 Reactor A.
The core shall consist of not more than 724 fuel assemblies B.
The reactor core shall contain 177 cruciform-shaped control rods. The control material shall be boron carbide powder (8 C) compacted to approximately 70% of theoretical density, 4
or Hafnium metal.
5.3 Reactor Vessel The reactor vessel shall be as described in Table 4.1.1 of the SAR. The applicable design codes shall be as described in Table 4.1.1 of the SAR.
5.4 Containment i -
A.
The principal design parameters and applicable design codes for the primary containment shall be as given in Table 5.2.1 of the SAR.
B.
The secondary containment shall be as described in Section 5.3.2 of the SAR and the applicable codes shall be as described in Section 12.1.1.3 of the SAR.
C.
penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.2 of the SAR and the applicable codes shall be as described in Section 12.1.1 3. of the SAR.
5.5 Fuel Storage A.
A maximum of 320 new fuel assemblies are permitted to be stored in the new fuel storage facility. New fuel storage reactivity limits are specified in Section 3.10.G.1, B.
Fuel storage is permitted in up to 33 high density fuel storage racks. This will allow a storage capacity of 3537 fuel assemblies. Spent fuel storage reactivity limits are specified in Section 3.10.G.2.
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