ML20247G774

Proposed Tech Specs,Modifying pressure-temp Limits for Reactor Vessel

ML20247G774
Person / Time
Site:	Peach Bottom
Issue date:	05/15/1989
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20247G771	List: ML20247G767 ML20247G774
References
NUDOCS 8905310068
Download: ML20247G774 (10)

Text

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ATTACHMENT 2 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 Docket No. 50-277 License No. DPR-44 REVISED TECHNICAL SPECIFICATION PAGES List of Attached Pages iva.

143 144 152 152a 164 164a 164b 164c i

8905310068 890515 PDR ADOCK 05C,00277 P

PDC

[=o_____----_--_---

t PBAPS Unit 2 LIST OF FIGURES Figure Title Page 3.5.1.I MAPLHGR vs. Planar Average Exposure 142h Unit 2, P8X8R Fuel, Type P8DRB284H, 80 mil & 100 mil channel & 120 mil channels I

3.5.1.J MAPLHGR vs. Planar Average Exposure 1421 Unit 2, P8X8R and BP8X8R Fuel, Type P8DRB299 and BP8DRB299, 100 mi) 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 Minimum Temperature for Pressure Tests 164 such as required by Section XI 3.6.2 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 l

3.6.5 Thermal Power and Core Flow Limits 164d 3.8.1 Site Boundary and Effluent Release Points 216e I

6.2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant 0perations 245 j

l l

-iva-l 1

________________m__-------

IMITING'CO'N'DITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 13.6 L PRIMARY-SYSTEM BOUNDARY

'4.6 PRIMARY SYSTEM BOUNDARY'

~ App *ticability:/

Applicability:

(Applies..to the operating status Applies t'o the periodic examination of:the. reactor coolant system.

'and testing requirements for;the' reactor coolant system..

l~

20bjective:

Objective:

To assure tlie integrity and safe-To determine the: condition of the

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operation of the reactor' coolant'

. reactor coolant system and the system..

op.eration of<the safety devices related:to it.

Specification:

Specification:

-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 duringnormalheatuporcgol-shall be permanently logged down shall not exceed 100. F at least every 15 minutes' increase (ordecrease)in-until the difference betwo.n-any'.one-hour period.

any 2 readings taken over a.45 0

minute period is less than 5 F. -

?2.

The reactor vessel shall not be pressurized for inservice hydro-

. static' testing above the pressure (a) Bottom head drain

- allowable for a given temperature (b) Recirculation loop

~

by Figure 3.6.1.

A and B.

The reactor vessel shall not be 2.

Reactor vessel temperature pressurized during heatup by non-and reactor coolant pres-l nuclear means, during cooldown sure shall be permanently following nuclear shut down or logged at least every 15 during low level physics tests minutes whenever the shell 0

.above the pressure allowable by temperature is below 220 F Figure 3.6.2, based on the tem-and the reactor vessel is peratures recorded under 4.6.A not vented.

The reactor vessel shall not be Test specimens of the reac-pressurized during operation with tor vesse'. base, weld and a critical core above the pressure heat effected zone metal allowable by Figure 3.6.3, based subjected to the highest on the temperatures recorded under fluence of greater than 1 Mev 4.6.A.

neutrons shall be installed in the reactor 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 FSAR.

- 143 -

)

u

F

-PBAPS Unit 2 i

1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

'3.6.A-- Thermal and Pressurization-4.6.A Thermal and Pressurization Limitations (Cont'd)

Limitations (Cont'd)_

Selected surveillance i

specimens shall be removed

• and tested to experimentally verify or l

adjust the' calculated values of integrated neutron flux and irradiation embrittlement that are used to determine the-RTND" for Figures 3.6.:., 3.6.2 and 3.6.3, and the figures shall be uodated based on the results.

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 U

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 tha reactor within the idle and operating coolant in the operating and regirculationloopsarewithin 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 domeandthebgttomheaddrain the bottom head drain shall be are within 145 F.

compared and permanently logged.

• Specimen Removal Schedule 1

Removed at 7.53 EFPY actual 2

15-18 EFPY 3

Standby

- 144 -

PBAPS Unit 2 l

3.6.A & 4.6.A BASES (Cont'd)

Operating limits on the reactor pressure and temperature were developed after consideration of Section III of the ASME Boiler and l

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 400 F margin required by 10 CPR 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 necessary to account for this change.

Regulatory Guide 1.99, Revision 2 provides methods for predicting decreased fracture toughness, in terms of shift in reference temperature of nil-ductility (RTNDT).

Generic methods are used until two surveillance capsules are removed and tested, at which time the surveillance test results may be used to develop plant-specific relationships of R'iNDT shift versus fluence.

Three capsules of neutron flux wires and samples 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 RTNDT predicted by Regulatory Guide 1.99, Revision 2 methods.

The results of the testing 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)).

l The flux wire test results provide the flux at one location in the l

vessel.

The flux distribe', ion can be determined analytically from the l

core physics data.

The ratio of the flux at the peak vessel location i

to that at the flux wire location, 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.

j Since the feedwater nozzles do not experience significant changes in fracture toughness due to irradiation, the pressure-temperature limits

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in Figures 3.6.1, 3.6.2 and 3.6.3 apply, without any RTNDT shifting, through 32 EFPY of operation.

l 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 i

steady state and transient conditions with respect to material fatigue.

The results of these transients are compared to allowable stress limits.

Requiring the coolant temperature 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.

- 152 -

6 PBAPS-Unit 2

' 3.6.A & 4.6.A. BASES (Cont'd)

' -The design basis event.for protection from pressure in excess'of vessel

.' design pressure as required by the ASME Boiler and Pressure Vessel Code, 13 the closure of all MSIVs resulting in a high flux scram (the slowest.' indirect scram due to high pressure).. The reactor vessel pressure Code limit of 1375 psig is wel. above the peak pressure i

produced by this most limiting overpressure event. This is discussed in more detail in Section 4.4.6 of the FSAR and GE safety analyses NEDE-24011-P-A.

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- 152a -

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-164-

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-164a-

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-164b-

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-164c-

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