ML20205R750

Amend 32 to License NPF-29,changing Tech Specs & Associated Bases for Reactor Pressure Vessel Pressure & Temp Limits to Be Consistent W/Limits Provided by GE for NSSS

ML20205R750
Person / Time
Site:	Grand Gulf
Issue date:	03/31/1987
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205R753	List: ML20205R750 ML20205R773
References
TAC-63558, NUDOCS 8704060431
Download: ML20205R750 (10)

Text

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/'pm Riog#'o UNITED STATES l'

NUCLE AR REGULATORY COMMISSION

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sv MISSISSIPPI POWER & LIGHT COMPANY SYSTEM EhERGY RESOURCE 5. INC.

SOUTH MISSIS 5IPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 32 License No NPF-29 1.

The Nuclear Regulatory Comission (the Commission) has found that A.

The application for amendment by Mississippi Power & Light Company, System Energy Resources, Inc. (fonnerly Middle South Energy, Inc.)

and South Mississippi Electric Power Association, (the licensees) dated November 11, 1986 as revised January 20, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commi:;sion's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; e.nd E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 32, are hereby incorporated into this license.

1 System Energy Resources, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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. 3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuar.ce: March 31, 1987 1

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This license amendrrent is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION AbC?EL Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensino

Attachment:

Changes to the Technical Specifications Date of Issuance: March 31, 1987 i

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ATTACHMENT TO LICENSE AFFNDMENT t:0. 32 J ' CILITY OPERATING LICENSE NO. NPF-?9 DOCKET NO. 50-416 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf page(s) provided to maintain document completeness.*

Remo,v_e Insert e

3/4 4-21 3/4 4-21 3/4 4-22 3/4 4-22*

B 3/4 4-3 B 3/4 4-3*

B 3/4 4-4 P 3/4 4-4 8 3/4 4-5 8 3/4 4-5 B 3/4 4-6*

B 3/4 4-6*

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4 A BB' CC' 1400 A - INITIAL SYSTEM HYDROTEST LIMIT

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B - INITIAL NON-NUCLEAR HEATING LIMIT L__

C - INITIAL NUCLEAR (CORE CRITICAL) 1 I

1200 A',B',C'-A,B.C LIMITS AFTER AN l_j '-

/ f ASSUMED 26'F CORE BELTLINE TEMP.

SHIFT FROM AN INITIAL SHELL PLATE

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RT OF O'F. A' IS NOT SHOWN NDT

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(NOT L!ti! TING)

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B' and C' are coincident 1000 with B and C, respectively.

f f ACore Belt I.-

Line after l__ _ f

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I 800

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'J Curves A, B and C are predicted to a

l be applicable for service periods i

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up to 32 EFPY.

c FEEDWATER 600 N0ZZLE r

f LIMITS g

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R Acceptable region of operation 400

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applicable curve.

312 psig i

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f BOLTUP l '4 LIMIT -

FEEDWATER 200 70*F N0ZZLE f {Q

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LIMITS

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0 100 200 300 400 S00 600 RPVMetalTemperature*(*F)

MINIMUM REACTOR FRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE Figure 3.4.6.1-1 GRAND GULF-UNIT 1 3/4 4-21 Amendment No. 32 l

1 TABLE 4.4.6.1.3-1 E

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE g

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CAPSULE VESSEL LEAD WITH0RAWAL TIME a

NUMBER LOCATION FACTOR (EFPY) 5

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1.

131C8981G1-N01 3*

0.46 8

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131C8981G1-N02 177*

0.46 24 3.

131C8981G1-NO3 183*

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REACTOR COOLANT SYSTEM BASES 3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormsl conditions. When the conduc-tivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the Ifmits will be detected in sufficient time to take corrective action.

3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 micro-curies per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 must be restricted to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year, approximately 10 percent of the unit's yearly operating time, since these activity levels increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam line rupture. The reporting of cumulative operating time over 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any 6 month consecutive period with greater than 0.2 micro-curies per gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

I Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

GRAND GULF-UNIT 1 B 3/4 4-3

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b REACTOR COOLANT SYSTEM I

BASES 3/4.4.5 SPECIFIC ACTIVITY (Continued)

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand 1

the effects of cyclic loads due to system temperature and pressure changes.

i 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 f

thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-

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temperature curve based on steady state conditions, i.e., nc thermal stresses, represents a lower bound of all similar curves for finite heatup rates when d

the inner wall of the vessel is treated as the governing location.

j The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the con-trolling location. The thermal gradients established during heatup produce tensile stresses which are already present. The thermal 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. Subse-j quently, for the cases in which the outer wall of the vessel becomes the stress j

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 RT for welds and base material in the closure flange region is NDT.

NDT 1 10*F.

The initial hydrostatic test pressure was 1563 psig. The results of these tests are shown in Table B 3/4.4.6-1.

Reactor operation and resultant fast neutron, E greater than 1 Mev, irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, ba<sd upon the fluence, NDT.

phosphorus content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recomme1dations of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor vessel Materials." The pressure / temperature limit curve, Figure 3.4.6.1-1, curves A', B' and C', includes predicted adjustments for this i

shift in RT for the end of life fluence. Curves B' and C' are coincident NDT with curves B and C, respectively.

GRAND GULF-UNIT 1 B 3/4 4-4 AmendmentNo.32l t

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REACTOR COOLANT SYSTEM BASES i

PRESSURE / TEMPERATURE LIMITS (Continued)

The actual shift in RT of the ve'ssel material will be established period-NDT l

ically during operation by removing and evaluating in accordance with ASTM E185-73 i

and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed 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 j

shall be adjusted, as required, on the basis of the specimen data and recom-mendations of Regulatory Guide 1.99, Revision 1.

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves j

C, C', and A, for reactor criticality and for inservice leak and hydrostatic testing, have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for j

inservice leak and hydrostatic testing.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES 4

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Double isolation valves are provided on each of the main steam lines to j

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 i

4 the containment. The surveillance requirements are based on the operating l

history of this type valve. The maximum closure time has been selected to i

contain fission products and to ensure the core i' not uncovered following s

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j line breaks.

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3/4.4.8 STRUCTURAL INTEGRITY

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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 j

acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access I

to permit inservice inspections in accordance with Section XI of the ASME j

Boiler and Pressure Vessel Code, 1977 Edition, and Addenda through Summer 1978.

i The inservice inspection prcgram 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 t

l where specific written relief has been granted by the NRC pursuant to 10 CFR i

Part50.55a(g)(6)(1).

3/4.4.9 RESIDUAL HEAT REMOVAL j

A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication; however, single failure considerations require that two loops l

be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

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I GRAND GULF-UNIT 1 B 3/4 4-5 Amendment No. 32 i

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BASES TABLE B 3/4.4.6-1 O

REACTOR VESSEL TOUGHNESS g

cs Minimum i

Startigg Maxi m,"*

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"'*'""".EOL cn Beltline Weld Seam I.D. or Heat No.-Slab No. or RTNOT ( F) ARTNOT ( F)

(ft-Ib)

NOT RT I II E

Component Material Type Heat No./ Lot No.

Cu % P (%)

E Plate SA-533 Gr.B. CL.1 C2594-2 0.04 0.012 0

+26

+26*

SA-533 Gr.B. CL.1 (C2594-2)

Weld

1. 2 Shell Long.

627260/B322A27AE 0.06 0.020

-30 44 N/A

+14 w

Seams Non-Beltline Material Type or Heat No.-Slab No. or I.II liighest Starting RTMT Component Weld Seam I.0.

Heat No./ Lot No.

Shell Ring SA-533 Gr.B. CL.1 C2815-2, C2779-2,

+10 C2779-1, C2788-2, C2788-1, C2741-1 t

Botton Head SA-533 Gr.B. CL.1 Alll3-1 0

Dollar Plate C2630-2 4

Bottom Head SA-533 Gr.B. CL.1 C2539-2, All45-1

+10 cz, Radial Plates w

D Top Head SA-533 Gr.B. CL.1 C2448-3

-30 Dollar Plate 3

E Top Head SA-533 Gr.B. CL.1 C2944-1

+10 Side Plates Top Head SA-508 CL.2 4801682

-30 Flange Vessel Flange SA-508 CL.2 4801141

-30 feedwater SA-508 CL.2 Forging No. 249A-1,

-20 Nozzle 2, 3, 4, 5, & 6, Q2Q65W Weld N/A N/A

-20***

Closure Stud SA-540 Gr.824 84025, 84299

+10

" Combination of the highest starting RTNOT plate and the highest ARINDI plate.

• • These values are given only for the benefit of calculating the end-of-life (f 0L) Riggy.

• • • Based on purchase spec. requirements.

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