ML18153A837

Proposed Tech Specs,Incorporating Revised Pressure/Temp Limits & Associated Ltops Setpoint That Will Be Valid to end-of-license

ML18153A837
Person / Time
Site:	Surry
Issue date:	06/08/1995
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18153A836	List: ML18153A835 ML18153A837
References
NUDOCS 9506120339
Download: ML18153A837 (17)

Text

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• ATTACHMENT 2 Proposed Technical Specifications Change Revised Pressure/Temperature Operating Limits and L TOPS Setpoint Surry Units 1 and 2

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9506120339 950608 PDR ADOCK 05000280 P PDR

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• TS 3.1-3 Reactor power shall not exceed 50% of rated power with only two pumps in operation unless the overtemperature AT trip setpoints have been changed in accordance with Section 2.3, after which power shall not exceed 60% with the inactive loop stop valves open and 65% with the inactive loop stop valves closed.

f. When all three pumps have been idle for > 15 minutes, the first pump shall not be started unless: (1) a bubble exists in the pressurizer or (2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

2. Steam Generator A minimum of two steam generators in non-isolated loops shall be operable when the average reactor coolant temperature is greater than 350°F.

3. Pressurizer Safety Valves

a. Three valves shall be operable when the head is on the reactor vessel and the reactor coolant average temperature is greater than 350°F, the reactor is critical, or the Reactor Coolant System is not connected to the Residual Heat Removal System.

b. Valve lift settings shall be maintained at 2485 psig +/- 1 percent*

• For the remainder of Cycle 1O and Cycle 11 operation for both units, the valve lift settings shall be maintained at 2485 psig (+5, -1 percent.)

Amendment Nos.

,*

B.

• HEATUPAND COOLDOWN

• TS 3.1-6 Specification

1. Unit 1 and Unit 2 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figures 3.1-1 and 3.1-2.

Heatup:

Figure 3.1-1 may be used for heatup rates of up to 60°F/hr.

Cooldown:

Allowable combinations of pressure and temperature for specific cooldown rates are below and to the right of the limit lines as shown in TS Figure 3.1-2. This rate shall not exceed 100°F/hr.

Cooldown rates between those shown can be obtained by interpolation between the curves on Figure 3.1-2.

Core Operation:

During operation where the reactor core is in a critical condition (except for low level physics tests), vessel metal and fluid temperature shall be maintained above the reactor core criticality limits specified in 10 CFR 50 Appendix G. The reactor shall not be made critical when the reactor coolant temperature is below 522°F as specified in T.S. 3.1.E.

2. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70°F.

Amendment Nos.

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• TS 3.1-9 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 28.8 Effective Full Power Years (EFPY) and 29.4 EFPY for Units 1 and 2, respectively. The most limiting value of RTNDT (228.4°F) occurs at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RTNDT of the limiting unirradiated material.

This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNoT; the results of these tests are presented in report BAW-2222, "Response to Closure Letters to NRC Generic Letter 92-01, Revision 1,"

dated June, 1994 and are reproduced in Tables 3.1-1 and 3.1-2. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTNDT* Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively (as well as adjustments for location of the pressure sensing instrument).

Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 28.8 EFPY or 29.4 EFPY for Units 1 and 2, respectively, prior to a scheduled refueling outage.

Amendment Nos.

f

• TS 3.1-10 Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section Ill of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of one and one half T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section Ill as the reference flaw, amply exceed the current capabilities of inservice. inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ~RTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1 R, for the metal temperature at that time. K1 R is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K1R curve is given by the equation:

K1R = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) where K1 R is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RTNDT Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C K1M + K1t ~ K1R (2) where, K1M is the stress intensity factor caused by membrance (pressure) stress.

Amendment Nos.

• TS 3.1-11 Kit is the stress intensity factor caused by the thermal gradients K1R is provided by the code as a function of temperature relative to the RTNDT of the material.

C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, K1R is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K1t, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining 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 of Figure 3.1-2 are composite curves which were prepared based upon the sam~ type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The cooldown limit curves are valid for cooldown rates up to 100°F/hr. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively. The adjusted reference temperature was calculated using materials properties data from the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) documented in the most recent revision to BAW-1543 and reactor vessel neutron fluence data obtained from plant-specific analyses.

Amendment Nos.

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• TS 3.1-12 THIS PAGE HAS BEEN INTENTIONALLY DELETED Amendment Nos.

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• E. Minimum Temperature for Criticality

• TS 3.1-18 Specifications

1. Except during LOW POWER PHYSICS TESTS, the reactor shall not be made critical at any Reactor Coolant System temperature above which the moderator temperature coefficient is more positive than the limit specified in the CORE OPERATING LIMITS REPORT. The maximum upper limit for the moderator temperature coefficient shall be:

a. + 6 pcm/°F at less than 50% of RATED POWER, or

b. + 6 pcm/°F at 50% of RATED POWER and linearly decreasing to O pcm/°F at RATED POWER.

2. In no case shall the reactor be made critical with the Reactor Coolant System temperature below the limiting value of RTNDT + 10°F, where the limiting value of RTNDT is as determined in Part B of this specification.

3. When the Reactor Coolant System temperature is below the minimum temperature as specified in E-2 above*, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to primary coolant depressurization.

4. The reactor shall not be made critical when the Reactor Coolant System temperature is below 522°F.

Basis During the early part of a fuel cycle, the moderator temperature coefficient may be calculated to be slightly positive at coolant temperatures in the power operating range. The moderator coefficient will be most positive at the beginning of cycle life, when the boron concentration in the coolant is the greatest. Later in the cycle, the boron concentration in the coolant will be lower and the moderator coefficient will be less positive or will be negative in the power operating range. At the beginning of cycle life, during pre-operational physics tests, measurements are made to determine that the moderator

. coefficient is less than the limit specified in the CORE OPERATING LIMITS REPORT.

Amendment Nos.

• TS 3.1-19 The requirement that the reactor is not to be made critical when the moderator coefficient is greater than the low power limit specified in the CORE OPERATING LIMITS REPORT has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requirement is waived during LOW POWER PHYSICS TESTS to permit measurement of reactor moderator coefficient and other physics design parameters of interest.

During physics tests, special operation precautions will be taken. In addition, the strong negative Doppler coefficient (2)(3) and the small integrated Delta k/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below the limiting value of RTNDT + 10°F provides increased assurance that the proper relationship between Reactor Coolant System pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps.

The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below 522°F provides added assurance that the assumptions made in the safety analyses remain bounding by maintaining the moderator temperature within the range of those analyses.

If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

(1) UFSAR Figure 3.3-8 (2) UFSAR Table 3.3-1 (3) UFSAR Figure 3.3-9 Amendment Nos.

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• TS 3.1-23a (3) Two Power Operated Relief Valves (PORV) shall be OPERABLE with a lift setting of ::;; 390 psig, or (a) When the PORVs are providing the overpressure protection verify the block valves are open at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

(4) A bubble shall be established in the pressurizer with a I*

maximum pressurizer narrow range level of 33% shall be maintained. After the* initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; two PORVs must also be OPERABLE, or (5) The RCS shall be vented through one opened PORV or an equivalent size opening.

(a) With the RCS vented through an unlocked vent path verify the path open at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(b) With the RCS vented through a locked vent path verify the path bpen at least once every 31 days.

2.

• The requirements of Specification 3.1.G.1.b (3) may be modified as follows:

a. One PORV may be inoperable in INTERMEDIATE SHUTDOWN with the RCS average temperature > 200° F but < 350°F for a period not to exceed 7 days. If the inoperable PORV is not

• restored to OPERABLE status within 7 days, then completely depressurize the RCS and vent through one open PORV or an equivalent size opening within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. One PORV may be inoperable in COLD SHUTDOWN or REFUELING SHUTDOWN with the reactor vessel head bolted for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inoperable PORV is not restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> then completely depressurize the RCS and vent through one open PORV or an equivalent size opening within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Amendment Nos.

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TABLE 3.1-1 UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d)

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NMwo(b)

UPPER SHELF HEAT OR MATERIAL Cu Ni TNDT RTNDT ENERGY MATERIAL CODE NO, SPEC, NO, 00 00 w w (EI LB)

Closure head dome C4315-2 A53313 Cl. 1 .14 .59 0 0 75 Head flange FV-1894 A508 Cl. 2 .13 .64 1o(a) 10 125 Vessel flange FV-1870 A508 Cl. 2 .10 .65 1o(a) 10 74 Inlet nozzle 9-5078 A508 Cl. 2 .87 so(a) 60 64 Inlet nozzle 9-4819 A508 Cl. 2 .84 so(a) 60 68 Inlet nozzle 9-4787 A508 Cl. 2 .85 6o(a) 60 64 Outlet nozzle 9-4762 A508 Cl. 2 .83 so(a) 60 85 Outlet nozzle 9-4788 A508 Cl. 2 .84 so(a) 60 72 Outlet nozzle 9-4825 A508 Cl. 2 .85 so(a) 60 68 Upper shell 122V109 A508 Cl. 2 .09 .74 40 40 83 Intermediate shell C4326-1 A533B Cl. 1 .11 .55 10 10 115(c)

=,, Intermediate shell C4326-2 A533B Cl. 1 .11 .55 0 0 94 3

rt)

,

Q. Lower shell C4415-1 A533B Cl. 1 .11 .50 20 20 103(c) 3 rt)

,

("lo Lower shell C4415-2 A533B Cl. 1 .11 .50 0 0 83 -l V,

z

.

0 Bottom head ring 123T338 A508 Cl. 2 .69 50 50 86 .....

w I

Bottom dome C4315-3 A533B Cl. 1 .14 .59 0 0 85 N C'I Inter. & lower shell 8T1554 & Linde 80 flux .18 .63 o(a) -5 77/EMA(e) vertical weld seam .

L1, L3, & L4

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TABLE 3.1-1 (Continued)

UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d)

NMwo(b)

UPPER SHELF HEAT OR MATERIAL Cu Ni TNDT RTNDT ENERGY MATERIAL COPE NO, SPEC, NO, 00 00 m m (EI LB)

Lower shell vertical 299L44 & Linde 80 flux .35 .68 o(a) -7 70/EMA(e) weld seam, L2 Inter. to lower 72445 & Linde 80 flux .21 .59 o(a) -5 77(a)1EMA(e) shell girth seam Upper shell to Inter. 25017 & SAF 89 flux .33 .10 o(a) 0 EMA(e) shell girth seam

• NOTES:

i:,, (a) Estimated per NRC standard review plan, NUREG-0800, Section MTEB 5-2 3

('1)

,

0. (b) Normal to major working direction - estimated per NRC standard review plan, NUREG-0800, Section MTEB 5-2 3

('1)

, (c) Actual values

("to 2

? (d) Reactor Vessel Fabricator Certified Test Reports (e) The approved equivalent margins analysis in the Topical Reports BAW-2192PA and BAW-2178PA demonstrates compliance with the requirements of I viw 10 CFR 50, Appendix G. *

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l

....I TABLE 3.1-2

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• UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) .

NMwo(b)

UPPER SHELF HEAT OR MATERIAL Cu Ni TNDT RTNDT ENERGY MATERIAL COPE NO, SPEC, NO, 00 00 ro ro (EI LB)

Closure head dome C4361-2 A533B Cl. 1 .15 .52 -20 7 81 Head flange ZV-3475 A508 Cl. 2 .11 .60 <1 o(a) <10 129 Vessel flange ZV-3476 A508 Cl. 2 .10 .64 _55(a) -65 129 e

Inlet nozzle 9-4815 A508 Cl. 2 .87 5o(a) 60 66 Inlet nozzle 9-5104 A508 Cl. 2 .84 50(a) 60 73 Inlet nozzle 9-5205 A508 Cl. 2 .86 50(a) 60 66 Outlet nozzle 9-4825 A508 Cl. 2 .85 50(a) 60 74 Outlet nozzle 9-5086 A508 Cl. 2 .86 50(a) 60 79 Outlet nozzle 9-5086 A508 Cl. 2 .87 5o(a) 60 73 Upper shell 123V303 A508 Cl. 2 .09 .73 30 30 104

t:,

=

11)

,

Q.

=

11)

Intermediate shell Intermediate shell Lower shell C4331-2 C4339-2 C4208-2 A533B Cl. 1 A533B Cl. 1 A533B Cl. 1

.12

.11

.15

.60

.54

.55

-10

-20

-30

-10

-20

-30 84 83 94 105(c)

, Lower shell C4339-1 A533B Cl. 1 .11 .54 -10 -10 c-1'

-I z 10 10 101 V,

0 Bottom head ring 123T321 C4361-3 A508 Cl. 2 A533B Cl. 1 .15

.71

.52 -20 -15 80

......

w Bottom dome I N

co Intermediate shell 72445 & Linde 80 flux .21 .59 -5 77(a)1EMA(d) vertical weld seams Lot 8579 L3 (100%), L4 (OD50%)

L4 (ID50%) 8T1762 & Linde 80 flux 8597 .20 .55 -5 EMA(d)

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i TABLE 3.1-2 (Continued)

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UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

NMwo(b)

UPPER SHELF HEAT OR MATERIAL Cu Ni RTNDT ENERGY MATERIAL COPE NO, SPEC, NO, 00 00 w (FT LB)

Lower shell vertical welds Seam L2 (ID 63%) 8T1762 & Linde 80 flux 8597 .20 .55 -5 EMA(d)

Seam L1 (100%) 8T1762 & Linde 80 flux 8597 .20 .55 -5 EMA(d)

Seam L2 (OD37%) 8T1762 & Linde 80 flux 8632 .20 .55 -5 EMA(d)

Inter. to lower 0227 and Grau Lo Flux LW320 .19 .56 o(a) 0 90(C)/EMA(d) shell girth seam Upper shell to Inter. 4275 & SAF 89 flux .35 .10 o(a) 0 EMA(d) shell girth seam NOTES:

i:a 3

rD

,

C.

z 3

rD

,

r+

(a) Estimated per NRC standard review plan, NUREG-0800, Section MTEB 5-2 (b) Normal to major working direction - estimated per NRC standard review plan. NUREG-0800, Section MTEB 5-2 (c) Actual value based on surveillance tests normal to the major working direction

.0 (d) The approved equivalent margins analysis in the Topical Reports BAW-2192PA and BAW-2178PA demonstrates compliance with the requirements of 10 CFR 50, Appendix G.

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. Figure 3 ..1-1

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l 'O I Surry Units 1 and 2 Reactor Coolant System Heatup Limitations Material Property Basis Limiting Material: Surry Unit 1 Intermediate to Lower Shell Circ Weld Limiting Adjusted RT(NDT) (Surry 1 at 28.8 EFPY):

228.4F (1/4-T), 189.5 F (3/4-T) 2500.00 I I I

Leak Test Limit I I I

tn 2000.00 I

• -tn I

-Q.

Cl) a..

s I 17 u,

tn 17 Cl) a.. 1500.00 a.. I/

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ca IUnacceptable I/

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"'C 1000.00 J

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/;

"'C I-"'. l7 17 Cl)

+,I }/ I/ I Acceptable ca l / 1,.,,..Y l Operation

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• - Heatup Rates 1---:: ,

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"'C (F/hr) L--- f/

t::-- v l/

L--- L..- I/

C: 20 _J--

500.00

~ t ---

40 60

- L---L..--

-

0.00 0 50 100 150 200 250 300 350 400 Indicated Cold Leg Temperature (Deg. F)

Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60 F/hr) Applicable for the First 28.8 EFPY for Surry Unit 1 and the First 29.4 EFPY for Surry Unit 2 Amendment No.

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• Figure 3.1-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations Material Property Basis Limiting Material: Surry Unit 1 Intermediate to Lower Shell Circ Weld Limiting Adjusted RT(NDT) (Surry 1 at 28.8 EFPY):

228.4 F (1/4-T), 189.5 F (3/4-T) 2500.00 I I I

I I

t>> 2000.00 I

• -

-...

. tn Q.

Q)

s 7 7

tn tn I a.

...

Q) 1500.00 I Q) Unacceptable Operation I I C,

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"C 1000.00 I

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==

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+,I Q) cu /~ ~'i I/ V L-::: ~ ~ l

• -"C

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Cooldown Rates

,--

v '/ / Acceptable Operation

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C: 500.00 (F/hr) 0 c-

- -- -

,__ c- L,-- L,-- y v

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20 40 -- - ,__ c-- L,-- v

,__ c-- /

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60 100 I-- I-- -- V" 0.00 0 50 100 150 200 250 300 350 400

• ;Indicated Cold Leg Temperature (Deg. F)

Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates* up to 100 F/hr) Applicable for the First 28.8 EFPY for Surry Unit 1 and the First 29.4 EFPY for Surry Unit 2 Amendment No.

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• Figures 3.1-3 and 3.1-4 have been deleted.