ML18092A865

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Proposed Tech Spec 3.4.9.1 Updating RCS Pressure/Temp Limits & Replacing Present Heatup Limit Curves
ML18092A865
Person / Time
Site: Salem PSEG icon.png
Issue date: 10/16/1985
From:
Public Service Enterprise Group
To:
Shared Package
ML18092A863 List:
References
NUDOCS 8510220326
Download: ML18092A865 (11)
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Text

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SALEM NUCLEAR GENERATING STATION UNIT NO. 1 LCR NO. 85-06 Description of Change A.

Technical Specifications Section 3/4.4.9 Pressure/Temperature Limits -

Reactor Coolant System:

B.

Replace present Heatup Limit Curves, Figure 3-4.2, with attached Heatup Limit Curves (Figure 3-4.2).

Replace present Cooldown Limit Curves, Figure 3-4.3, with attached Cooldown Limit Curves (Figures 3-4.3A and 3-4.3B).

Technical Specifications, Bases~ 3/4.4.9 Pressure/Temperature Limits:

Change last line on page B 3/4 4-6 to read -

"The heatup and cooldown curves were prepared based u~on the most limiting value-of the predicted adjusted Leference temperature at the end of 10 EFPY".

Change last line of first paragraph on page B 3/4 4-7 to read -

"The heatup and cooldown limit curves (Figures 3.4-2, 3.4-3A and 3.4-3B) include predicted adjustments for this shift in RTijDTat the end of 10 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments".

Replace present neutron f luence versus service life curves, Figure B 3/4.4-1, with attached curves.

Replace existing curves for shift in reference transition temperature ( A RTNDT), °F versus fluence, Figure B 3/4.4-2, with attached curves.

Replace existing Table B 3/4.4-1, Reactor Vessel Toughness Data, with the attached table

  • I

Reason for Change As required by Technical Specifications Section 4.4.9.1.2, the heatup and cooldown limit curves have been updated to reflect the results of analyses and tests conducted on material specimens contained in surveillance capsule Y, the second capsule removed from Unit l reactor vessel.

The results of the evaluation show that the weld metal transition temperature increase was as predicted, and that the increase for plate material was slightly less than predicted.

Since the surveillance weld metal is not identical to the limiting intermediate to lower shell seam weld, the revised heatup and cooldown curves are based on the upper limit of Regulatory Guide 1.99, Revision 1, prediction curves (0.35% Cu, 0.012%

p)

  • The test results and conclusions are summarized in Report WCAP-10694, "Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 1.Reactor Vessel Radiation Surveillance Program", dated December 1984.

Significant Hazards Consideration Analysis The proposed license change request does not involve any significant hazards consideration.

The revised heatup and cooldown curves are based on the upper limit of Regulatory Guide 1.99, Revision l for transition temperature shifts: and are more conservative than the existing curves.

The reactor vessel operation in accordance with these limitations will result in lower stresses to the reactor vessel during heatup and cooldown of the reactor coolant system.

The core design on Salem Units 1 and 2 has been modified to a low leakage configuration to reduce the embrittlement rate.

The Salem End of Life (EOL) RTNDT will be recalculated using the embrittlement rate predicted by the upper limit of Regulatory Guide 1.99 and the recalculated lower EOL flux.

When completed, this information will supersede the data* for Salem transmitted to the NRC by WOG-82-290 dated December 31, 1982 on Pressurized Thermal Shock.

In summary, the use of more limiting curves will offset the higher predicted RTNDT and will result in continued conservative operation.

There is no increase in the probability or consequences of any accidents previously evaluated, nor are any new accidents introduced. Additionally, safety margins are increased. This change conforms to example (ii) of the guidance provided by the Commission for changes th~t are Not Likely To Involve Significant Hazards Corisiderations in that it increases the stringency of a LIMITING CONDITION FOR OPERATION.

e MATERIAL PROPERTY BASIS UPPER LIMIT OF REG. GUIDE TREND CURVES (FIGURE B3/4 4-2)

COPPER CONTENT PHOSPHORUS CONTENT RT INITIAL RT~~i AFTER 10 EFPY 0.35 WU 0.012 wn; 0°F l/4T, 236°F 3/4T, 107°F CURVES APPLICABLE FOR HEATUP RATES UP TO 60°F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY 3000 (32000 Q. -

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100 INOICRTEO TE"PEARTUA! (°F) 383.4 Figure 3.4-2 Salem Unit l Reactor Coolant System Heatup Limitations Applicable up to 10 EFPY SALEM -

UNIT l 3/4 4-26

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CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100°F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY I

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Figure 3.4-3A Salem Unit 1 Reactor Coolant System Cooldown Limitations Appli-cable up to 10 EFPY (Excluding Instrument Error Margins)

SALEM - UNIT l

. 3/4 4-27

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COPPER CONTENT 0.35 WT%

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RT INITIAL 0°F RT~~+ AFTER 10 EFPY l/4T, 236°F 3/4T, 107°F CURVES APPLICABLE FOR COOLOOWN RATES UP TO 100°F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY I I Includes instrument error margins (51 PSIG and 15.4 OF)

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(°F) 500 Figure 3.4-38' Salem Unit l Reactor Coolant System Cooldown Limitations Appli-cable up to 10 EFPY {Including Instrument Error Margins)

SALEM - UNIT l 3/4 4-27a

REACTOR COuLANI SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and. pressure changes.

These cyclic loads are introduced by nonna1 ioad transi~nts, reactor trips, and startup and shutdown operations. The various cat~aries of load cycles used for design purposes are provided in Section 4.1.5 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

.During heatup, the thennal gradients in the reactor vessel wall produce

- thennal* stresses which vary from compressive at the inner wall to tensile at tQe outer wall. These thennal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve. based on steady state -conditions (i.e., no thennal stresses) represents a lower bound of all s1m11ar curves for finite heatup rates when the inner wall of the vessel is treated as the governing 1 ocati on.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thennal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present.

The thennal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an ind4vidual basis.

The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by detennining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60°F per hour.

The cooldown limit curves, Figure 3.4-3, are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cool down. thennal gradients *tend to produce tensile stresses while producing compressive stresses at the outside wall.

The heatup and cooldown cu~ves were prepared based upon the most limiting value of the predicted adJusted reference temperature at the end of 10 EFPY.

SALEM - UN IT l B 3/4 4-6

REACTOR COOLANT SYSTEM BASES The reactor vessel materi.a.ls have been tested to determine their initial RTN

the results of these tests are shown in Table B 3/4.4-1.

Reactor ope~ltion and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RTN

. Therefore, an adjusted reference temperature, based upon the ~Tuence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.

The heatup and cooldown limit curves (Figures 3.4*2 and 3.4-3) include predicted adjustments for this shift in RTN at the end of 10 EFPY, as well as adjustments for possible errors in ~~e pressure and tanperature sensing instruments.

The actual shift in RTNRT of the vessel material ~ill ~ established periodically during operatic by removing and evaluating, in accordance with ASTM ElSS-70, reactor vessel material irradiation surveillance specimens installed near the inside wall of the.reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ~RT determined fran the surveillance capsule is different fran the cal~BTated ~RTNDT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure canpliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and

  • the frequencies for removing and testing these specimens are in accordance with the requirenents of Appendix H to 10 CFR Part 50.

The-limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements

  • SALEM - UNIT l B 3/4 4-7

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10 12 14 16 18 20 22 24 26 28 30 32 SERVICE LlrE (EFFECTIVE FULL POWER YEARS)

Figure B3/4.4-l Fast Neutron Fluence (E>lMeV) as a Function of Full Power Service Life (lTPY)

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Figure B3/4. 4-2 Effect of Fluence and Copper and Phosphorus Contents on 6RTNDT for Reactor Vessel Steels Weld Metal (Based on Capsule Y results)**

e Shell Plate 82402-3 (Based on Capsule Y result~)

TABLE B 3/4.4-1

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SALEM UNIT l REACTOR VESSEL TOUGHNESS DATA (UNIRRAOIATED)

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50 ft lb 1NOT 35-Mil RTNDT

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Material Cu p

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Com onent tleat Uo.

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Of Of Of (ft lb)

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Cl Hd Dome A0610 82407-1 A533B, Cl.1 0.20 0.011

-30.

99*

39 110 Cl Hd Segment Cl544 82406-l A533B, Cl.1 0.13 0.010

-20 89*

29 125 Cl Hd Segment Cl544 62406-2 A5330, Cl.1

0. 16 0.012

-30 85*

25 122*

Cl Hd Segment 85852 82406-3 A5338, Cl. 1 0.10 0.009

-50 66*

6 132 Cl Hd Flange l23P409 82811 A508, Cl.2 0.010 28*

22*

28 199 Vessel Flange 5Pll91 82410 A508, Cl. 2 0.009 (jQ*

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50 145 4P1o19 Enlet Nozzle 123P403 82408-1 A508 1 Cl. 2 0.010 50*

43*

50 144 llnlet Nozzle 125P544 82408-2 A508, Cl.2 0.011 46*

26*

46 157 Inlet Nozzle 123P403 82408-3 A508, Cl. 2 0.010 47*

37*

47 161 Inlet Nozzle l25P544 82408-4 A508, Cl.2 0.010 9*

17*

9 167 tD Outlet Nozzle ZT2550 82409-1 A508 1 Cl.2 0.010 60*

95*

60 75 Outlet Nozzle ZT2t>ti0 82409-2 A508, Cl.Z 0.011 60*

95*

60 78 w

Outlet Nozzle ZT2585 82409-3 A508, Cl.2 0.013 60*

10*

60 121

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Outlet Nozzle ZT2585 82409-4 A508, Cl.2 0.012 60*

13*

60 126

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Upper Shell A0497 82401-1 A5338 1 Cl.1 0.22 0.012

-30 87*

27 114 I

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Upper Shell A0495 82401-2 A5338, Cl.l

o. 19 0.011 0

80*

20 122 0

Upper Shell A0512 82401-3 A5338, Cl.1 0.24 0.011

-10 114*

34 96 Inter Shell C1354 82402-1 A533B, Cl.1 0.24 0.010

-30 105 45 73.0 97 Inter Shell C1354 82402-2 A5338, Cl. l 0.24 0.010

-30 55

-5

91. 5 112 Inter Shell C1397 82402-3 A533B,. Cl. l
0. 22 0.011

-40 57

-3 104.0 127 lower Shell Cl356 82403-1 A533B, Cl. l

o. 19 0.011

-40 70 10 99.0 143 lot11er Shell C1356 82403-2 A5338 1 Cl. 1 0.19 0.012

-70 86 26 94.0 128 lower Shell Cl J"S6.

82403-3 A533B, Cl.1

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-40 90 30 102.0 131 Bot Hd Seg111ent A0705 82404-1 A5338, Cl.1 0.10 0.009 10 48*

10 120 Bot Hd Segment A0705 82404-2 A533B, Cl.1

o. 11 0.010

-50 60*

0 132 Bot Hd Segment A0705 82404-3 A5338, Cl.1 0.12 0.008 10 47*

10 126 Ont Hd Dolle*

A0705 82405-1 A5338, Cl. l

o. 15 0.010

-20 57*

-3 106 Survei 1 larice 0.16 0.019 0*

-38 **

0 104**

JMID - Normal to Major WOrking Direction Miio

- Major Working Direction

- Estiinated per NRC Standard Review Plan Branch Technical Position KTEB 5-2

- Actual transverse data obtained from surveillance program (from minimum data points).

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Either POPS has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50°F above the RCS cold leg temperatures, or (2) the start of a safety injection pump and its injection into a water solid RCS.

3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components wi11 be maintained at an acceptable 1eve1 throughout the life of the plant.

To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

SALEM - UNIT 1 B 3/4 4...:11

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