ML20082R177
| ML20082R177 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/05/1991 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082R179 | List: |
| References | |
| NUDOCS 9109130243 | |
| Download: ML20082R177 (16) | |
Text
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'n UNITED STATES
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i NUCLEAR REGULATORY COMMISSION Ee%
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i COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMEN'l TO FACILITY OPERATING LICENSE L
Amendment No. 114 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
Tise application for amendment by the Commonwealth Edison Company (the licensee) dated November 28, 1988, June 26, 1989, October 23, 1989, March 23, 1990, and July 26, 1991, 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 (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 emended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B. of Facility Operating License No. DPR-19 is hereby amended to read as follows:
e 9109130P43 910905 PDR ADOCK 05000237 P
2 B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 114, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W
Richard J. Barrett, Director Project Directorate 111-2 Division of Reastor projects - lil/1V/V Office of Nucleir Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 5, 1991
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. ATTACHMENT TO LICENSE AMENDMENT N0. 114_
FACILITY OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 Revise the Aptendix 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
'NSERT viii viii 3/4.6-2 3/4.6-2
/4.6-23 3/4.6-23 6 3/4.6-26 B 3/4.6-26 B 2/4.6-26a
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DRESDEN 11 DPR-19 Amendment No. 114 List of Tabler (continued)
P,ag Table 3. 2-1 Fire Detection Instruments B 3/4.12-17 Table 3.12-2 Sprinkler Systems B 3/4.12-18 Table 3.12-3 C0 Systems B 3/4.12-19 p
Table 3.12-4 Fire Hose Stations B 3/4.12-20 & 21 Table 6.1.1 Minimum Shiit Manning Chart 6-4 Table 6.6.1 Special Reports 6-23 List of Figures Figure 2.1-3 APRM Bias Scram Relationship to Normal Operating Conditions B 1/2.1-17 Figure 4.1.1 Graphical Aid in the Selection of an Adequate Interval Between Tests B 3/4.1-18 Figure 4.2.2 Test Interval vs System Unavailability B 3/4.2-3B Figure 3.4.1 Deleted 3/4.4-4 Figure 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 Figure 3.6.1 Minimum Reactor Vessel Metal Temperature 3/4.6-23 Figure 3.6.2 Thermal Power vs Core Flow limits for Thermal Hydraulic Stability Surveillance in Single Loop Operating 3/4.6-24 Figure 4 I
Minimum Reactor Pressurization Temperature B 3/4.ti-29 Figure 4.6.2 Chlor de Stress Corrosion Test Results at 500 F B 3/4.6-31 i
Figure 4.8.1 Owner Controlled / Unrestricted Area Boundary B 3/4.8-38 Figure 4.8.2 Detail t,f Central Corralex B 3/4.8-39 Figure 6.1-1 Offsite Organization - Deleted Figure 6.1-2 Station Organization - Deleted viii
- _ __. - _ _ _._._ _ ~
5 DRESDEN II DPR-19 Amendment No. 114 "3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT (Cont'd.)
(Cont'd.)
B.
Pressurization Temperature B.
Pressurization Temperatore 1.
The reactor vessel-shall 1.
Reactor Vessel shell be vented and power-temperature and reactor operation shall not be coolant pressure shall conducted unless the-be permanently recorded reactor vessel at 15 minute intervals tercerature is equal to whenever the shell or. greater than that temgeratureisbelow shown in Curve C of 220 F and the reactor Figure 3.6.1.
Opera-vessel is not vented.
tion for-hydrostatic or leakage tests, during heatup or cooldown ana withthecorecritIcal shall be-conducted only when: reactor vessel metal temperature is equal to or above that shown in the appropriate-curve of Figure 3.6.1.
Figure 3.6.1
- l is effective through 16 effective full power i
- years.
At least six months
- l prior to 16 effective full power years new-curves will be submitted.
2.
The reactor vessel head-2.
When the reactor vessel bolting studs shall not be head bolting studs are under tension unless the tightened or loosened -
temperature of the vessel the reactor vessel shell immediately below shell temperature the. vessel flange is immediately below the greater than or equal to_
head flange si.al_be
_ l._
-80 F.
permanently recorded.
J 3/4.6-2
l 1
DRESDEN 11 DPR-19 Amendment No. 114 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) 1600 _
E VAUD TO E
A - SYSTEM HYDROTEST Uu!T g
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50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( F)
FIGURE 3.6.1.
3/4.6-23 I
DRESDEN 11 DPR-19 Amendment No. IM
- 3. 6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
B.
Pressurization Temperature - The reactor vessel is a primary ier against the release of fTssion products to the environs.
In order to provide assurance that this barrier is maintained at a high degree of integrity, pressure-temperature limits have been established for the operating conoitions to which the reactor vessel can be subjected Figure 3.6.1 presents the pressure-temperature curves for those opera-ting conditions; Inservice Hydrostatic Testing (Curve A), Non-Nucleat Heatup/Cooldown (Curve B), and Core Critical Operation (Curve C).
These curves have been established to be in conformance with Appendix G to 10 CFR 50 and Regulatory Guide 1.99, Revision 2, and take into account the change in reference nil-ductility transition temperature (RTNDT) as a result of neutron embrittlement.
The adjusted reference temperature (ART) of the limiting vessel material is used to account for irradiation effects.
Three vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non-beltline region (other than the closure flange region); and 3) the closure flange region. The beltline region is defined as that region of the reactor vessel that directly surrounds the effective height of the reactor core (between the bottom and the top of active fuel), and is subject to an RT adjustment to account for irradiation embrittle-NDT ment.
The non-beltline and closure flange regions receive insufficient fluence to necessitate an RTg37 adjustment.
These regions contain components which include; the reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do not directly surround the reactor core.
Although the closure flange region is a non-beltline regien, it (the closure flange region) is treatea separately far th; deve v nt of the pressure-temperature curves to address 10 LFR 50..c;,e-c1x G requiremer..s.
In evaluating the adc. aacy cf tne steol which comprises the reactor vessel, % is necessar,y that the following be established:
- 1) the RT t r all vessel and adjoining materials; 2) the relationship NDT between RT and integrated neutron flux (fluence, at energies NDT greater than one Hev); and 3) the fluence at the location of a postu-lated flaw.
Boitup Temperature The initial RT f the main closure flanges, the shell and head NDT materials connecting to these flanges, the connecting welds and the vertical electrNiag welds which terminate immediately below the vessel flange are all 20 ' or lower.
Therefore, the minimum allowable boltup temperature is established as 80 F (RTNDT + 60 F)
B :/4.6-26 l
DRESDEN 11 DPR-19 Amendment
_3. 114 l
l which includes a 60 F conservatism required by the original ASME Code of construction.
Curve A - Hydrotesting As indicated in Curve A of Figure 3.6-1 for system hydrotesting, the minimum metal temperature of the reactor vessel shell is 80 F for reector pressures less than 312 psig.
This 80*F minimum boltup tem-perature is based on an RT f 20 F for the top head plate (most NDT limiting material) and a 60 F conservatism required by the original ASME Code of construction.
At reactor pressures greater than 312 psig the minimum vessel metal temperature is established as 110 F.
The 110 F minimum temperature is based on a closure flange region RT of 20 F and a 90'F conservatism NDT required by 10 CFR 50 Apperdix G for pressure in excess of 20% of the preservice hydrostatic test pressure (1563 psig).
At ap3roximately 620 psig reactor pressure the effects of pressuriza-tion secomo more limiting than the boltup stresses at the closure flange region, as shown by the non-linear portion of Curve A intersect-ing ti'e vertical 110*F line. The non-linear portion of the curve is dependent on the non-beltline region (which is actually more limiting than the celtline region through a vessel exposure of 22 effective full power years), and based on an RT f 40 F.
NDT Curve B - Non-Nuclear Heatup/Ccoldown Ct.rve B of Figure 3.6.1 aplies during heatups with non-nuclear heat (e.g., recirculation pump Teat) and during cooldowns when the reactor is not critical (c.u., following a scram).
The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress (non-beltine stresses).
As indicated by the vertical 30 F line, the boltup stresses at the closure flange region are most limiting below approximately 80 psig.
ribove appproximately 80 psig, aressuri ation and th rmal stresses become more limiting than the >oltup stresses, which is reflected by the non-linear portion of Curve B.
The non-linear portion of the curve is dependent on non-beltline region (which is actually more limiting than the beltline region through a vessel exposure of 22 effective full power years), and based on an RT f 40 F.
NDT Curve C - Co.e Critical Operatior Curve C, the core critical operation curve shown in Figure 3.6.1, is generated in accordance with 10 CFR 50 Appendix G which requires core c
critical pressure-temperature limits to be 40 F above any Curve A or B limits.
Since Curve B is more limiting, Curve C is Curve B plus 40 F.
i I
P 3/4.6-26a
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gfN UNITED STATES
{<3,g,1 NUCLEAR REGULATORY COMMISSION D,, ;
g WASHINGTON. D.C. 20655
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COMMONWEALTH EDISON COMPANY DOCKET FO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 AMEtlDMENT TO FACILITY OPERATING LICENSE Amendment No. 111 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
i A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated November 28, 1988, June 26, 1989, October 23, 1989, March 23, 1990, and July 26, 1991, complies
- tith the standards and requirements of the Atomic Energy Ac+ of 1954, as amended (the Act), and the Commission's rules and regulations set forth in i
10 CFR Chapt - 1; B.
The facility will o)erate in conformity with the application, the provisions of tie 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 amendn.ent 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 Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B. of Facility Operating License No. DPR-25 is hereby amended to read as follows:
9 4
2 B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 111, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective es of the date of its issuai.ce.
FOR THE NUCLEAR REGULATORY COMMISSION Richard J. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - lil/lV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 5, 1991
ATTACHMENT TO LICENSE AMENDMENT NO.
111 FACILITY OPERATING LICENSE NO. DPR-25 DOCKET NO. 50-249 i
i Revise the 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 viii viii 3/4.6-2 3/4.6-2 3/4.6-23 3/4.6-23 B 3/4.6-26 B 3/4.6-26 B 3/4.6-26a 9
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t DRESDEN III DPR-25
~
Amendment No. 111 List of Tables (continued)
.P,.a.g.e.
T$ble3.12-2 SprinklerSystsms B 3/4.12-18 Table 3.12-3 00 Systems B 3/4.12-19 2
Table 3.12-4 Fire Hose Stations B 3/4.12-20 & 21 Table.6.1.1' Minimum Shift Manning Chart 6-4 Table.6.6.1 Special Reports 6-22
- List of Figures Figure 2.1 -APRM Bias Scram Relationship to Normal LOperating Conditions B 3/2.1-17 Figure 4.1.1-Grarhtcal Aid in the Selection of
. an Adequate Interval Between Tests B 3/4.1-18
- Figtire -4.~2. 2 Test Interval vs System. Unavailability-B 3/4.2-38 Figure 3.4.1:
Deleted-3/4.4-4 Figure 3.4;2' Sodium Pentaborate Solution Temperature Requirements
~3/4.4-5 Figure 3.6.1' 1 Minimum Reactor Vessel Metal Temperature 3/4.6 -Figure 3.6.2 Thermal P1wer vs Core Flow Limits for Thermal-
~ Hydraulic Stability Surveillance in Single Loop Operating 3/4.6-24 Figure 4.6.1 Minimum Retctor Pressurization Terperature B 3/4.6-29 o
Figure '4. 6._-2
_ Chloride Stress-Corrosion Test Results at 500 F
.B 3/4.6-31 E
Figure 4.8,1 Owner Controlled / Unrestricted Area Boundary B 3/4.8-38 Figure 4.8.2 Detail of Central Complex-B 3/4.8-39 Figure.6.1-1 Offsite Organization # Deleted Figure 6.1-2.
Station. Organization - Deleted viii
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r DRESDEN III DPR-25 Amendment No. 111
- 3. 6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT l
(Cont'd.)
(Cont'd.)
B.
Pressurization Temperature B.
Pressurization Temperature 1.
The reactor vessel shall 1.
Reactor Vessel shell be vented and power temperature and reactor operation shall I.ot be coolant pressure shall conducted unless the be permanently recordeG reactor vessel at 15 minute intervals temperature is equal to whenever the shcIl or greater than that 220geratureisbelow tem shown in Curve C of F and the reactor Figure 3.6.1.
Opera-vessel is not vented.
tion for hydrostatic or leakage tests, during heatup or cooldown, and l
with the core critical shall be conducted only-I when reactor vessel metal temperature is equal to cr above that shown in the appropriate curve of Figure 3.6.1.
Figure 3.6.1 is effective t
I through 16 effective full power years.
At least six months prior l
to 16 effective full power years, new curves will be submitted.
i 2.
The reactor-vessel 2.
When the reactor vessel-l-
head bolting studs head bolting studs are L
shall not be under tightened or loosened tension unless the the reactor vessel temperature of the shell temperature l:
-vessel shell immediately below the L
immediately below head flange shall be l
the vessel flange is permanently racorded.
greater than or equal to 100 F.
f I
3/4.6-2
DRESDEN III DPR-25 Amendment No.
111 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('f) 1600=
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50-100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) i FIGURE 3.6.1.
3/4.6-23 l.
r DRESDEN III DPR-25 Amendment No. 111 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
8.
Pressurization Temperature - The reactor vessel is a primary barrier against the release of fission products to the environs.
In order to provide assurance that tnis barrier is maintained at a high degree of integrity, pressure-temperature limits have been established for the operating conditions to which the reactor vessel can be subjected.
Figure 3.6.1 presents the cressure-temper 6ture curves for those operating conditicns; Inservice Hydrostatic Testing (Curve A), Non-Nuclear Heatup/Cooldown (Curve B), and Core Critical Operation (Curve C).
These curves have been established to be in conforn nce with Appendix G to 10 CFR 50 and Regulatory Guide 1.99 Revision _, and take into accountthechangeinreferencenil-ductilitytransitiontempera-ture (RTNDT) as a result of neutron embrittlement.
The adjusted reference ter:mrature (ART) of the limiting vessel material is used to account for irradiation effects.
Three vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non-beltline region (other than the closure flange region); and 3) the closure flange region.
The beltline recion is defined as that region of the reactor vessel that directly surrcunds the effective height of the reactor core (between the bottom and top of active fuel), and is subject to an RT adjustment NDT to account for irradiation embrittlement.
The non-beltline had closure flange regions receive insufficient fluencc to necessitate an RTNDT adjust-ment.
These regions contain components which include; the rehctor vessel nozzles, closure flanges, top and bottom head plates, control rod drive penetrations, and the11 plates that do not directly surround the reactor core.
Although the closure flange regicn is a non-beltline region, it (the closure flange region) is treated separately for the development of the
-pressure-temperature curves to address 10 CFR 50 Appendix G requirements.
In evaluating the adequacy of the steel which comprises the reactor vessel, it is necessary that the following be established:
- 1) the RT fraH NDT ves. 1 and adjoining raterials, 2) the relationship between RT and in" NDT tegrated neutron flux (fluence, at ene gie3 greater than one Mev); and 3) the fluence at the location of a postulated flaw Boltup Temperature The initial RT f the main closure flanges, the shell and head NDT materials connecting to these flanges, and connecting welds is 10 F; however, the vertical electroslag welds which terminate i:amediately belos the vessel flange have an RT of 40"F.
Therefore, the minimum NDT allowable boltup temperature is established as 100 F (RTNDT + 60*F)
[
B 3/4.6-26
l 9:
DRESDEN III DPR-25 Amendment No. 111 which: includes a 60'F conservatism required by the original ASME Code of construction.
Cy ve A - Hydrotesting As indicated in Curve A of Figure 3.6.1 for system hydrotesting, the minimum metal temperature of the reactor vessel shell is 100 F for reactor pressures less than 312 psig.
This-100'F n.inimum boltup temperature is based on an RT f 40*F for the electroslag weld NDT immediately below the vessel flange and a 60"F conservatism required by the original ASME Code of construction.
At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 130'F.
The 1?O F minimum temperature is based on a closure flange region RTNDT of 4u F ond a 90 F conservatism required by 10 CFR 50 Appendix G for pressure in excess of 20% of the preservice-hydrostatic test pressure (1563 psig).
At approximately 650 psig the effects of pressurization are more limiting than the boltup stresses at the closure flance region, hence a family of non-linear curves intersect the 330 F vertical line.
Belt-line as'well as non-beltline curvec have been provided to allow sepa-rate monitoring of the two regions, Beltline curves as a function of vessel exposure for 12,14 and 16 effective full power years (ErPY) are presented to allow ~the use of the appropriate curve up to 16 EFPY of operation.
Curve B
.Non-Nuclear Heatup/Cooldown Curve B of Figure 3.6.1 applier during heatups with non-nuclear heat (e.g... recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).
The curve provides the minimum. reactor vessel metal temperatures based on the most limiting vessel stress.
As indicated by the vertical 100 F line, the boltup stresses at the closure flange region are most-limiting for reactor pressures below approximately 110 psig. 'For reactor pressures greater than approxi-mately 110 psig, pressur.ization and thermal stresses become more limiting than the boltuo stresses, which is reflected by the non-linear portion of curve B.
The non-linear portion of the curve is dependent on non-beltline and beltline egions, with the beltine region temperature limits having been adjusted to account for vessel irradi -
ation (up to a vessel exposure of 16 EFPY).' The non-beltline region is limiting between a) proximately 110 psig and 830 psig.
Above approxi-mately 803 psig, tie beltline region becomes limiting.
Curve C - Core Critical Operation' Curve C, the core critical operation curve shown in Figure 3.6.1, is generated in accordance with 10 CFR 50 Appendix G which requires core critical pressure-temperature limits to be 40 F above any Curve A or B
. limits.
Since Curve B is more limiting, Curve C is Curve B plus 40'F.
B 3/4.6-26a 9
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