ML20034B969

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Proposed Tech Specs Re Figure 3.4.6.1 Concerning Pressure Temp Limit
ML20034B969
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 04/26/1990
From:
SYSTEM ENERGY RESOURCES, INC.
To:
Shared Package
ML20034B968 List:
References
NUDOCS 9005010247
Download: ML20034B969 (9)
v • d • e


Text

WL 90/ol

' REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS HACTORCOOLANTSYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system pressure and reactor _ vessel metal temperature shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydrostatic or leak testing; (2) curve B for heatup by non nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:

' A maximum reactor coolant heatup of 100'F in any one hour period, a.

b.

A maximum reactor coolant cooldown of 100'F in any one tour period, A maximum temperature change'of _less than or equal to 10'F in any c.

one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and.

d.

The reactor vessel flange and head flange temperature greater than or equal to 70'F when reactor vessel head bolting studs are under' tension.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perfom an engineering evaluation to of the reactor coolant system; determine that the reactor co remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During s testing operations,ystem heatup acidown and inservice leak and hydrostatic the reactor t.)olant system temperature and pressure shall be determined to be within the tMye_ required heatup and cooldown limits and the reactor coolant system preswre and reactor vessel metal temperature'shall be determined to be to the right of the limit lines of Figure 3.4.6.1-1 curves A or B and-F, as applicable, at least once per 30 minutes, 1

i GRAND GULF-UN]T 1 3/4 4-19 MON

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A/L 90/o/

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system pressure and reector vessel metal temperature shall be determined to be to the right of the cri drawal of control rods to bring the reactor to criticality and at least'once per 30 minutes during system heatup.

4.4.6.1.3 verified to be greater than or equal to 70'F:The reactor vessel flange and hea In OPERATIONAL CONDITION 4 when reactor coolant system temperature a..

is:

1.

I 100'F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

5 80*F, at least once per 30 minutes, b.

Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

4.4.6.1.4 The reactor vessel material specimens shall be removed and examined as a function of time and THERMAL POWER as required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1, 4.4.6.1.5 Tra r;nter fle-.ir; ;;;; inn; ; tall be is..e4 et it.e fire; e !=;ti=-;f ti= =d ;;nr !=;l =d and t; --difj Tigurer;f;; ling ;;t;;; ad R rn alt; ;f tra flu n;e d;t; mirati =;, in ; = jur.;;ien.itti N.se00/4.0.01.

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T h e Press u re.- Tem p era +a< e. 16 &

Fic)u.re.3.H,6,1-1 is va.t\\ L Thrws h 10 Mys and shall be. re e valuafed Friarte e.vcee 4 n3 I s s I= P y s.

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NINI648N MACTOR YESSEL ETAL TEWERATURE VS. MACTOR VESSEL PRES $URE FIGURE 3.4 6 1 1

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j 3/4. 4. 5 SptCIFIC hCTIVITY (Continued)

The surveillance requimments provide esegnate assurance that escessive specific activity levels in the reacter coolant wiH be estected in sufficient l

time to take corrective action.

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3/4.4.6 pett9_**/TEW ERATURE LIMITS the effects of cyclic leads due to systes ta p erstde and pre l

These cyclic loses em introduced by mamal lead transients, reacter trips,.

s and starte and shutepun operations, used for desi h various categoties of lead cycles and shuteoun,gn purposes em provided in lection 11.3 of %e FSAR.

1 i

the Petes of temperature and pressue changes are lietted se thatDuring star assisuptions and satisfy the stress limits for cyclic operation,the t

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a thermal stresses which very fees coepressive at the inner w j

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the outer well.. These thema) induced capressive str p we.tensde stresses induced by the internal pressure. esses tend to.elleviate k

the i

t Theref nture curve nu.d.n swee saw conditi.m i.e., m = ore, a pressure-3 th. inner m M of the m sei is t t.d = = p o rning im.ti.e.Mpm.nu a io., bound.f 4H si.nar curm for s.,

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N heate analysis aise covers the detemination of pmssure temperatum d

'liettations for the case in which the outer weH of the vessel bece

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trolling locatten. N themel tensne stress which am aires.gredients establi.shed dort3.heetup produce p. ent. m ou ed stre o. at d the outer wall of the vessel are hasHe and are dependent en both the rate of y

heatup and the ties along the beste ramp 1 therefore, a louer bound cerve similar y

to that described for the heatup of the inner well cannet be defined. Subse-A i

quently controlling location, each hatte rate of interst; aust.be analysed 8

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individual basis.

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Tw The reacter vesset materials have been tested to determine their o

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g RT The RT for welds and base asterial in.the closure flange region is fI gy.

gy e < it'F.

j:g These tests e m shown in TableThe initta) hydrostatic test pressure was 1863 psis.

i 8 3/4.4.64.

g fast neutron, I greater than 1 Nov, irresistion wH 1 cause an incrosse iSoester opera tre RT Therefore, an adjusted reference temperature, based een the fluence, jl n the

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o gy.

M ;t::;dNtent and capper centent offthe esterial in question, can be WjpredictedusingBasesFip!?fRevisten $

re 8 3/4.4.6 1 4fGuide1.pg and the.roceanendations of Regulatory Seesse-'teEsectorWesselMaterials.";^.; ;7 ".ntt' W r2 c. 7;C^:: Radiation puriftle.*feATWr Figure 3.4.8.1 1, curses A", t and C. -he pressure / temperature limit curve, 4

snfft in RTgfe ^t: n: ;7 'ife 7'xx.. ludes predicted adjustments for this 0.;... O' r.; i m nt = t m^.

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seses Figure s 3/4.4.61 het been revloed to reflect the enetysis of the flus wire sosteeter i

teilch was removed Wring the firtt refueling autepe. The never bound curve s 3/4.4.6 1 wee determining the Pressure issperature tielt curves in Figure 3.4.61 1 I

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PRES $URE/ TEMPERATURE LIMITS (Continued)

(g The actual shift in RT NOT of the vessel asterial will be estabitshed period-

.E9 itally during operation by removing and evaluatig in accordance with ASTM E185 73 i

.E and 10 CFR 50, Appendix H, irradiated reactor vassel material specimens installed.

c near the inside well of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material.

d transition o.asperature shift.

The operating limit curves of Fi ute 3.4.6.1 1 shall be ad ustea,as required, on the basis of t sendations o,f. Reguistory Guide 1.99, Revision / ;he specimen da a and recom-L The pressure-temperature limit lines shown in Figures 3.4.6.1 1, curves C,4 and A, for reactor criticality and for inservice leak and hydrostatic requirem,ents of Appendix G to 10 CFR Part 50 for reacto l

inservice leak and hydrostatic testing.

3/4.4.7 MAIN STEAM LINE !$0LAT!0N VALVES Double isolation valves are provided on each of the sain steam lines to sintette the potential leakate paths from the containment in case of a line break.

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

history of this type valve.The survet11ance requirements are based on the operatin The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following i

line breaks, s

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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 syntes were desianed to provide access i

to permit inservice inspections in accordance with sectTon KI of the ASME Boiler and Pressure vessel Code,1977 Edition, and Addenda through Svaner 1978 The inservice inspection program for ASME Code Class 1 2 and 3 components will be performed in accordance wtth Section XI of the ASE, toiler 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 i

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

3/4.4.9 RE5100AL NEAT REMOVAL v

A si le shutdown cooling mode loop provides sufficient heat removal capabilit l

for removing core decay heat and mixing to assure accurate tempera-

't ture indication; however, single failure considerations require that two loops l

be OPERABLE or that alternate methods capable of decay heat removal be 7

demonstrated and that an alt. ornate method of coolant sixine be in operation.

GRAND GULF-UNIT 1 8 3/4 4-5 Amendment Nu. 32 R

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AIL;Jolol De.leke. t h is -figu r e-REACTOR COOLANT SYSTEM Re. place win rev' Sed bases Finare 6 3/4 H.6-I

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Fast Neutron Fluence (E>l MeV) at-1/4 T As a Function of Service Life

  • BASES FIGURE B 3/4 4.6-1 FAST NEUTRON FLUENCE (E>l Me AT 1/4 T AS A FUNCTION OF SER CE LIFE *
  • At 90% of RATED THERMAL POWER AND 90% availability.

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AT VESSEL 1.0. AS A FUNCTION OF EFPY.

1 GRAND GULF UNIT 1 B % 4-7 A mendmenf No._

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