ML20043A902

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Proposed Tech Specs Revising Min Temp Vs Reactor Pressure, Curves Contained in Section 3/4.4.6 & Associated Bases
ML20043A902
Person / Time
Site: River Bend Entergy icon.png
Issue date: 05/14/1990
From:
GULF STATES UTILITIES CO.
To:
Shared Package
ML20043A901 List:
References
NUDOCS 9005230261
Download: ML20043A902 (12)
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Text

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ENCLOSURE 1-Marked-up Pages of RBS. Tech. SFecs. for Requested Changes l

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FIGURE 3.4.6.1 1 MINIMUM TEMPERATURE REQUIRED VS REACTOR PRESSURE RIVER BEND - UNIT 1 3/4 4-23.

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100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF)

FIGURE 3.4.6.1 - 1 MINIMUM TEMPERATURE REQUIRED VS REACTOR PRESSURE RIVER BEND - UNIT 1 3/4 4-23

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TABLE 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITH0RAWAL SCHEDULE CAPSULE VESSEL-LEAD WITHDRAWAL TIME-NUMBER LOCATION TOR 9 k (EFPY) 1 3*

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2 177*

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'l RIVER BEND - UNIT 1 3/4 4-24

J REACTOR' COOLANT SYSTEM l

BASES

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3/4.4.6 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 normal load transients, reactor trips, '

and startup and-shutdown operations..The various categories of load cycles used for design purposes are provided in Section_3.9 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 thermal. gradients.in the reactor vessel wall produce thermal stresses which vary from compressive.at the inner wall to tensile at the outer wall.

These thermally induced compressive stresses tend'to alleviate the tensile stresses induced'by the internal pressu~re.

Therefore, a pressure-temperature curve based on steady state conditions, i.e., no thermal stresses :

represents a lower bound of all similar curves. for finite heatup rates _when the inner wall of the vessel is treated as the governing location.

The heatup analysis alic ccvers the determination of pressure-temperature

-limitations for'the case'in which the outer wall of-the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermally induced stresses at t

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 :urve 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 individual basis.

The reactor vessel materials have'been tested to determine their initial RT The results of these tests are shown in Table B 3/4.4.6-1.

Reactor NDT.

operation and resultant-fast _ neutron (E greater than 1 MeV) irradiation will cause an increase in the RT Therefgre an adjusted reference temperature, NDT.

based upon the fluencabeepnertr$(c'ontent and copper content of the material l

in question, can be predicted using Bases figure B 3/4.4.6-1 and the recommen-l dations of Regulatory Guide 1.99, RevisiorQ, ":ff;;t; ;f %:id;:1 El;;;nt: :n

"redi;;;d %dicti;; t-:;;; t: %;;;;r '!;::.1 ":t;rt:1;." The pressure /

temperature limit curve, Figure 3.4.6.1-1, curves A', B and C', includes pre-dicted adjustments for this shift in RT f r they n: ;7-M f; f ;;;;;, ;; -ell NDT

:dj":* rt: f:r ;:::151: :rr:r: in th: pr:::;r: :nd t;;;, r:tur; i:n:ing l=: = =::.

d,dow ' Mdhhed of Reacher Ve*d Mj@

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RIVER BEND - UNIT 1 B 3/4 4-5

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The actual shift in RTNDT of the vessel material will be determined periodicall during operation by removing and evaluating, in accordance with ASTM ElB and 10 CFR 50, Appendix:H, irradiated reactor vessel material l

specimens nstalled near the inside wall of.the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift.-

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis.of the specimen a

dataandrecommendationsofRegulatoryGuide1.99, Revision 7.A l

0 The pressure / temperature. limit lines shown in Figure 3.4.6.1-1, curves C, and C', and A and A'; for reactor criticality and for. inservice leak and hydro-static testing have been provided to assure' compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criti-cality and for inservice leak and hydrostatic testing.

L 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize _the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of-the containment.

However, single failure considerations require that two valver be OPERABLE.

The surveillance requirements are based on the operating history-of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line. breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

o 3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that' the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to pennit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1977 Edition and Addenda through Summer 1978.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

RIVER BEND - UNIT 1 8 3/4 4-6 t

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BASES TABLE B 3/4.4.6-1

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COMPONENT MAT't TYPE HEAT / LOT CU(%)

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NOTE:* These values are given only for the benefit of calculating the(; d M if: (COL))RTNDT 32.~ EFPY) l HEAT / SLAB HIGHEST NON-BELTLINE OR STARTING RT COMPONENT MT'L TYPE HEAT / LOT.

MDT(*F) i R

Shell Ring SA 533 Gr8 C1.1 ALL HEATS

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Botton Head bone.

SA 533 Gr8 C1.1.

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Top Head Dome SA 533 Gr8 C1.1

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SA 533' Gr8 C1.1 -

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+10 Top Head Flange SA 508 C1.2 ALL HEATS.

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Feedwater Nozzle SA 508 C1.2

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LOW ALLOY-STEEL-ALL HEATS

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-or 824 Meets requirement of 45 ft-Ibs and 25 mils lateral expansion at +.10*F l

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10 20 30 40 SERVICE LIFE (YEAR $*)

sAsas FlouRE a 3/4.4.s.1

' FAST NEUTRON FLUENCE (E>1 MEV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE' i

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Factor change in River. Bend! Technical Specifications"t j

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". Implementation..of.:RegulatorfJ:1 Guide?

'i 1.99, Revision /2 for Ri.ver Bend Station Unit 1".J

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GE Nuclear Energy i

G-LD-9-160 s

l October 6, 1989 Mr. J.C.

Deddens L

Senior Vice President - RBNG Gulf States Utilities River Bend Station Post Office Box 220 i

St. Francisville, Louisiana 70775 Attention:

Erwin Zoch

Subject:

EXPLANATION OF LEAD FACTOR CHANGE-IN RIVER BEND TECHNICAL-SPECIFICATIONS

Reference:

GE Letter G-LD-9-152 dated September 19,. 1989 j

transmitting GE Report SASR89-20, Implementation of Reg' Guide 1.99 Rev. 2

Dear Erwin:

The.following is in response to your questions.on the lead. factor values which you raised after reviewing the Referenced report, l

The' lead factor currently in the Technical Specifications, 0.86 i

at the-1/4 T location, was calculated by'a combination of one-dimensional and two-dimensional neutron transportLfinite difference computer codes at-the time lthe~FSAR was prepared.

As computational capability improved over the years, the one-dimensional calculations were. replace by a second-two-dimensional calculation, resulting in theJrevised lead factor of 0.89 at the 1/4 T location.

The combination of two two-dimensional computationsLis.more l

accurate than the earlier method involving one-dimensional computations.

The small reduction in conservatism. associated with the revised lead factor is the result of improved calculation i

accuracy, and is therefore justified.

1 If you have any questions or comments, please contact Tom Caine at-(408) 925-4047 or myself.

Sincerely, lr h

John E.

Dale

, Nuclear Services Manager L{504) 295-8670 JED10062 cc: GE Nuclear Enercy T.A. Caine C.E. McGee y

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March 12, 1990 1

Mr. Erwin Zoch Gulf States Utilities Company Highway 61, North Access Road St. Francisv111e, LA 70775

Subject:

REVISION OF RIVER BEND P-T CURVES. REPORT SASR 89-20

References:

l'.

Caine. T. A.,-" Implementation of R'egulatory Guide l'.99, Revision 2 for River Bend Station Unit 1," GE Report SASR 89 20, Revision 1,-

March 1990.-

2.

Caine, T.

A., " Implementation _of Regulatory Guide 1,99, Revision 2 for River Bend Station Unit 1," GE Report SASR 89-20,) Revision 0,.

September 1989.

1

Dear Erwin,

Enclosed.are the following:

.one - bound ~ copy and ' one - loos e copy o f Re fe r ence 1,.a set of the pages changed in Reference 1, with. changes l

l highlighted, and the vessel outline' drawing which shows the vessel diameter and thickness used in the pressure-temperature (P-T) curve calculations.

As we have discussed by phone, the - thickness 5. 81 ~ inches used in calculating previous P-T curves, including those in Reference 2 and.those currently in the Tech Specs,.was incorrect.

Reference 1-is based on the correct value of 5.41 inches.

The most'significant change in Reference 1 compared to Reference 2 is that the P-T curves A', B' and C' have increased by about 7'F.

If you have any questions on the enclosed information', please give me a call at the number.below.

Regards,

/

In

~

T. A. Caine, Senior Engineer Materials Monitoring & Structural Analysis Services (408)'925-4047, Mail Code 747 cc: (w/o enclosures)

J. Dale, GE NSM 5

1

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