ML20070W027

Proposed Tech Spec Change Request 90-17 Re pressure-temp Limits for Reactor Vessels

ML20070W027
Person / Time
Site:	Peach Bottom
Issue date:	03/28/1991
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20070W025	List: ML20070W023 ML20070W027
References
NUDOCS 9104120288
Download: ML20070W027 (14)

Text

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i ATTAC!IMENT 2 I PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 4

DPR-56 REVISED TECHNICAL SPECIFICATION PAGES List of Attached Pages iv (Unit 3) iva (Unit 3) 143 (Units 2 and 3) 144 (Units 2 and 3) 152 (Unit 3) 152a (Unit 3) 164 (Unit 3) 164a (Units 2 and 3) 164b (Unit 3) 164c (Unit 3) 2095b. doc l

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1 9104120288 910328 '

I i PDR ADOCK 05000277

P PDR I

Unit 3 PBAPS LIST UT~7TGURES Figure Title Pace 1.1 1 APRM Flow Bias Scram Relationship To Normal 16 Operating Conditions 4.1.1 Instrument Test Interval Determination Curves 55 4.2.2 Probability of System Unavailability vs. Test Interval 98 3.3.1 SRM Count Rate vs. Signal-to-Noise Ratio 103a 3.4.1 DELETED 122 3.4.2 DELETED 123 3.5.K.1 DELETED 142 3.5.K.2 DELETED 142 3.5.1.A DELETED 3.5.1.B DELETED 3.5.1.C DELETED 3.5.1.0 DELETED 3.5.1.E DELETED 142 3.5.1.F DELETED 142 3.5.1.G DELETED 142 3.5.1.H DELETED 142 3.5.1.1 DELETED 142-3.5.1.J DELETED 142 ,

3. 5.1. K - DELETED 142-3.6.1 Minimum Temperature for Pressure Tests such 164 as required by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup or 164a Cooldown following Nuclear Shutdown ,

3.6.3 Minimum Temperature for Core Operation (Criticality) 16ab 3.6.4 DELETED 164c e

3.6.5 Thermal Power Limits of Specifications 3.6 f.3, 164d

-3.6.F.4, 3.6.F.5. 3.6.F.6 and 3.6.F 7

3. 8.1 = Site Boundary and Effluent Release Points 216e

6. 2-l = DELETE 0 s 2-2 DELETED

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' PBAPS Unit 3 l

LIMI. TING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6- PRIMARY SYSTEM BOUNDARY Applicability: Applicability:

Applies to the operating status Applies to the period P 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 operation of the reactor coolant reactor coolant system and the system. operation of the safety devices related to it.

Specification: Specification:

A. Thermal and Pressurization A. Thermal and pressurization Dmitations Limitations

1. The average rate of reactor 1. During heatups and cool-downs, 4 coolant temperature change the following temperatures duringnormalheatuporcgol~ 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 d 45 minute period is less than 50 F.

2. The reactor vessel shall not be pressurized for inservice hydro-static testing above the pressure (a)Bottomheaddrain 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 pressucized during heatup by non- and reactor coolant pres-nuclear ceans, during cooldown sure shall be permanently l following t.eclear shut down or logged at least every 15 during low itvel physics tests above the preisure allowable by minuteswhenevertheshe}l 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 vessel base, weld.and a critical core above the pressure heat affected zone metal allowable by Figure 3.6.3, based were installed in the

-on the temperatures recorded under reactor vessel adjacent 4.6.A. to the vessel wall at the core midplane level. The specimens and sample progr6m shall conform to ASTM E 185-66 to the degree discussed in the FSAR.

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PBAPS Unit 2 4

MMTINGCONDITIONSFOROPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY 1 Applicability: Applicability l

, Applies to the operating status Applies to the periodic examination of the reactor coolant system. and testing requirements for the '

i reactor coolant system. j Objective: _ Objective:

To assere the integrity and safe To determine the condition of t' operation of the reactor coolant reactor coolant system and the system, operation of the safety devict 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, I coolant temperature change the ft.11owing temperatures duringnormalheatuporegol- shall-be permanently logged down shall not exceed 100 F .at least every 15 minuter increase (ordecrease)in until the. difference between any one-hour period. any 2 readings taken.over a 45 minute _ period-is less than 60 F.

2. The reactor vessel shall not be i 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-nuclear means, during cooldown sure shall be permanently followirs nuclear shut down or . logged at least every 15 l

during low level physics tests above the pressure allowable by minuteswhenevertheshejl 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 vessel base, weld and.

a critical core above the pressure _- heat-affected zone metal-allowable by Figure 3.6.3, based- were installed:in the-on the temperatures recorded under reactor; vessel adjacent

[ 4.6.A. to.the vessel wall at the- ,

i core midplane-level. .The

l. specimens and_ sample

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-program shall conform-to-ASTM E/185-66 to the' degree-  !

discussed in the FSAR.: ,

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I Unit 3 PBAPS LIMITING CONDITIONS FOL OPERATION SURVEILLANCE REQUIREMENTS 3.6.A Thermal and Pressurization 4.6.A Therma; and pressurization Limitations (Cont'd) Limitations (Cont'd)

Selected surveillance specimens shall be removed and tested in accordance with 10 CFR 50 Appendix H, to experimently verify or adjust the calculated values of integrated neutron flux and irradiation embrittlement that are used to determine the RTND" for Figures '

3.6.L. 3.6.2 and 3.6.3, and the figures shall be updated based on the results.  ;

3. The reactor vessel head bolting studs shall nc,t be under tension 3 When the reactor vessel head bolting studs are tensioned and unless the temperatures of the closure flanges and adjacent the reactor is in a Cold Condition, the reactor vessel vessel and head materials are shell temperature immediately greater than 70g F.

below the head flange shall be permanently recorded.

4. The pump in an idle recirculation 4. Prior to and during startup of loop sht11 not be started unless an idle recirculation loop, the the terperatures of the coolant temperature of the 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 rea: tor 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.

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Unit 2 I PBAPS

, LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

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' 3.6.A _ Thermal and Pressurization 4.6.A Thermal and Pressurization Limitations (Cont'd) Limitations (Cont'd) l Selected surveillance specimens shall be removed and tested in i

accordance with 10 CFR 50 Appendix H, to experimently verify or 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 updated 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 greater than 70 0 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 opercting coolant in the operating and regirculationIcopsarewithin 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.

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.. PBAPS Unit 3 3.6.A & 4.6.A PASES (Cont'd) l Operating limits on the reactor pressure and temperature were developed after consideration of Section III of the ASME Boiler and Pressure Vessel Code and Appendix G to 10 CPR 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 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 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-duct.lity (RTNDT). Generic methods are used until two surveillance capsules are removed and tested, at which time the surveillance test results may be uned to develop plant-specific relationships of RTNDT 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 and tested in 1989 tc 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 90-50 of DRP Bll-00494. The results of vessel material testing will not ne 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 (FyPY)).

The flux wire test results provide the flux at one location in the vessel. The flux distribution can be determined analytically from the core phycies data. The ratio of the flux at the peak vessel location to that at the flux wira location, known as the lead f actor, 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 EPPY of fluence was taken into account. However, in. comparing the beltline operating limits (uith 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 irradiation, the pressure-temperature limits in Figures 3.6.1, 3.6.2 and 3.6.3 apply, without any RTNDT shifting, through 32 EFPY of operation.

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 transiert 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 500 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.

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PBAPS Unit 3 3.6.A & 4.6.A BASES (Cont'd)

The design basis event for protection from pressure in excess of vessel design pressure, as raqi'*ed by the ASME Boiler and Pressure Vessel Code, is the closure of all MS!Vs resultiog in a high flux scram (the slowest indirect scram due to high pressure). The reactor vessel pressure Code limit of 1375 psig is well abcVe the peak pressure produced by this most limiting overpressure event. This is ciscussed in more detail in Section 4.4.6 of the FSAR and GE safety analyses NE00 2Cll-P-A, l

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