ML19325E346
| ML19325E346 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/25/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19325E347 | List: |
| References | |
| NUDOCS 8911060300 | |
| Download: ML19325E346 (12) | |
Text
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3 NUCLEAR REGULATORY COMMISSION
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PHILADELPHIA ELECTRIC COMPANY f
PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY l
ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMFNT TO FACILITY OPFP.ATING LICENSE Amendment No. 153 l
License No. DPR-44 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et al. (the licensee). dated May 15, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I.
I B.' The facility will operate in conformity with the application, the provisions of the Act, and the rules'and regulations of the l
Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangerir.g the health and safety of the public, and (ii) that such activities will be conducted in ecmpliance with the Commission's regulations; l
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health or safety of the public; and E.
The issuance of this anendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been t.
satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attcchment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-44 is hereby f
amended to read as follows:
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Technical S;)ecifications e
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.153, are here>y incorporated in the license. PECO shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical l
Specifications Date of Issuance:
October 25, 1989 i
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ATTACHMENT TO LICENSE AMENDMENT N0.153 FACILITY OPERATING LICENSE NO. DPR-44 i
DOCKET NO. 50-277 t-Replace the following pages of the Appendix A Technical Specifications with~
l the enclosed pages. The revised areas are indicated by marginal lines.
Pages Iva 143 144 152 152a l
164 164a 164b 164c 1:,
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LIST OF FIGURES Figure Title Page 3.5.1.I MAPLHGR vs. Planar Average Exposure 142h Unit 2,.P8X8R Fuel, Type P80RB284H, 80 mil & 100 mil channel & 120 mil channels
'3.5.1.J MAPLHGR vs. Planar Average Exposure 1421 i
Unit 2, P8K8R and BP8X8R Fuel, Type P80RB299 and BP80RB299, 100 mil channels
~3.5.1.K-MAPLHGR vs. Planar Average Exposure 142j Unit 2, P8X8R Fuel (Generic) 3,5.1.L MAPLHGR vs. Planar Average Exposure 142k Unit 2, BP8X8R Fuel. Type BP8DRB299H 3.5.1.M MAPLHGR vs. Planar Average Exposure 1421 Unit 2, GE8X8EB Fuel, Type BD319A 3.5.1 N MAPLHGR vs. Planar Average Exposure 142m Unit 2, GE8X8EB, Type BD321A 3.5.1.0 MAPLHGR vs. Planar Average Exposure 142n Unit 2, GE8X8EB, Type LTA310 3.6.1 Hinimum Temperature for Pressure Tests 164 such as required by Section XI
'l.6.2 Minimum Temperature for Mechanical Heatup 164a or Cooldown following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation 164b (Criticality) l.3.6.4 Deleted 164c l3.6.5 Thermal Power and Core Flow Limits 164d 3.8.1 Site Boundary and Effluent Release Points 216e 6.2-1 Management Organization Chart 244 l
6.2-2 Organization for Conduct of Plant 245 Operations P
k Amendment No, M,102, H4, 153
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M PBAPS Unit 2 o
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
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L 3.6. PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY j
Applicability:
Applicability:
L Applies to the operating status Applies to the periodic examination of the reactor coolant system.
and testing requirements for the-reactor coolant system.
l-Objective:
Objective:
To-assure the integrity and safe To determine the condition of the r
operation of the reactor coolant reactor coolant system and the system.
operation of the safety devices L
related to it.
' Specification:
Specification:
I A.
Thermal and Pressurization A.
Thermal and Pressurization Limitations Limitations 1.
The average rate of reactor 1.
During heatups and cool-downs, coolant temperature change the following temperatures during normal heatup or cool-shall be permanently logged down shall not. exceed 100 F at least every.15 minutes increase (or decrease) in until the difference between any one-hour period, any 2 readings taken over a 45 minute period is less than 5' F.
2.
The reactor vessel shall not be pressurized for inservice hydrostatic (a) Bottom head drain testing above the pressure allowable (b) Recirculation loop A and B.
for a given temperature by Figure 3.6.1.
The reactor vessel shall not be 2.
Reactor vessel temperature and pressurized during heatup by non-reactor coolant pressure shall l
nuclear means, during cooldown shall be permanently logged at following nuclear shut down or least every 15 minutes whenever during low level physics tests the shell temperature is above the pressure allowable by below 220 F and the reactor i
Figure 3.6.2, based on the tem-vessel is not vented.
l peratures recorded under 4.6.A.
1 h
The reactor vessel shall not be Test specimens of the reactor l
pressurized during operation with vessel base, weld and heat a critical core above the pressure effected zone metal subjected allowable by Figure 3.6.3, based to the highest fluence of on the temperatures recorded under greater than 1 Mev neutrons 4.6.A.
shall be installed in the reac-vessel adjacent to the vessel wall at the core midplane level.
The specimens and sample program shall conform to ASTM E 185-66 to the degree discussed in the I
FSAR.
Amendment No. @ t'r,153
-143-
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS b
l-3.6.A Thermal and Pressurization 4.6.A Thermal and Pressurization Limitations (Cont'd)
Limitations (Cont'd)
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Selected surveillance specimens' shall be removed
- and tested to experimentally verify or adjust-L the calculated values of.
integrated neutron flux and t
irradiation embrittlement that are used to determine the RT NDT for Figures 3.6.1, 3.6.2 and i,,
3.6.3, and the figures shall be, updated based on the results.
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3.
~ The. reactor vessel head bolting 3.
When the reactor vessel head studs shall not be under tension bolting studs are tensioned and unless'the temperatures of the the reactor is in a Cold closure flanges and adjacent Condition, the reactor vessel vessel and head materials are shell temperature immediately greater than 70' F.
below the head flange ^shall be.
permanently recorded.
4.
The pump in an idle recirculation 4.
Prior to and during startup'of-loop shall not be' started unless an idle recirculation loop, the the temperatures of the coolant temperature of the reactor within the idle and operating coolant in the operating and recirculation loops are within idle loops shall'be permanently 50 F of each other.
logged.
5.
The reactor recirculation pumps 5.
Prior to starting a recircula-shall not be started unless the tion pump, the reactor coolant coolant temperatures.between the temperatures in the dome and in dome and the bottom head drain the bottom head drain shall be
- are within 145 F.
compared and permanently
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- Specimen Removal Schedule 1
1 Removed at 7.53 EFPY actual 2
15-18 EFPY 3
Standby i
1 Amendment No. 'tt, 4t, fr3, M, 153 344
L PBAPS Unit 2 l
3.6.A & 4.6.A BASES (Cont'd)
Operating limits on the reactor pressure and temperature were developed after l
l consideration of Section III of the ASME Boiler and Pressure Vessel Code and Appendix G to 10 CFR Part 50.
These considerations involved the reactor vessel beltline and certain areas of discontinuity (e.g. feedwater nozzles and vessel head flange).
These operating limits (Figures 3.6.1, 3.6.2 and 3.6.3) assure that a postulated surface flaw can be safely accommodated.
Figure 3.6.3 includes an additional 40 F margin required by 10 CFR 50 Appendix G.
The fracture toughness of the vessel low alloy steel in the core region, referred to as beltline, gradually decreases with exposure to neutrons, and it is necess-ary tn account for this change.
Regulatory Guide 1.99, Revision 2 provides methods for predicting decreased fracture toughness, in terms of shif t in refer-ence temperature of nilductility (RTsurveillancecapsulesareremovedan$ die)s.
Generic methods are used until two ted, at which time the surveillance test results may be used to deve'iop plant-specific relationships of RT shift NDT versus fluence.
Three capsules of neutron flux wires and samoles of vessel material were installed in the reactor vessel adjacent to the vessel wall at the core midplane level.
The first capsule of wires and samples was removed at the end of Cycle 7 and tested in 1988 to experimentally verify the irradiation shift in RT pre-dicted by Regulatory Guide 1.99, Revision 2 methods.
Theresultsofth$D[esting are documented in GE Report SASR 88-24 of DRF B13-01445.
The results of vessel material testing will not be factored into Figures 3.6.1, 3.6.2 and 3.6.3 until the second capsule is tested.
However, the flux wire results were used to predict the design fluence (valid to 32 effective full power years (EFPY)).
The flux wire test results provide the flux at one location in the vessel.
The flux distribution can be determined analytically from the core physics data.
_The ratio of the flux at the peak vessel location to that at the flux wire loca-tion, known as the lead factor, was calculated to relate the flux wire test results to the maximum value for the vessel.
In developing Figures 3.6.1, 3.6.2 and 3.6.3, the shift predicted by Regulatory Guide 1.99, Revision 2 methods for 32 EFPY of fluence was taken into account.
However, in comparing the beltline operating limits (with 32 EFPY shift) to the feedwater nozzle limits, it was determined that the feedwater nozzle was more limiting.
Since the feedwater nozzles do not experience significant changes in fracture toughness due to l
irradiation, the pressure-temperature limits in Figures 3.6.1, 3.6.2 and 3.6.3 apply, without any RT shifting, through 32 EFPY of operation.
NDT As described in paragraph 4.2.5 of the Final Safety Analysis Report, detailed stress analyses have been made on the reactor vessel for both steady state and transient conditions with respect to material f atigue.
The results of these transients are compared to allowable stress limits.
Requiring the coolant tem-perature in an idle recirculation loop to be within 50 F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.
Amndment No. Sfr -P, *iF,153
-152-
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o-PBAPS Unit 2 3.6.A &-4.6.A BASES (Cont'd) v The design basis event for protection.from pressure in excess of vessel design
-pressure,las required by the ASME Boiler and Pressure Vessel Code, is the closure.
of all MSIVs resulting in a.high flux scram.(the slowest indirect. scram due to.
l' high pressure). :The reactor vessel pressure Code limit of 1375 psig is well p
above the peak pressure produced by this most limiting overpressure event. fThis is discussed in more detail in Section 4.4.6 of the FSAP. and GE safety analyses NEDE-24011-P-A.
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Amendment No. *trl 153
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