ML20151Z271
| ML20151Z271 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 02/07/1986 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20151Z269 | List: |
| References | |
| NUDOCS 8602140118 | |
| Download: ML20151Z271 (16) | |
Text
.
ATTACH" INT A REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
a.
A maximum heatup of 100*F in any one hour period, b.
A maximum cooldown of 100*F in any one hour period, and c.
A maximum temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICA8ILITY: At all times.
~
~
ACTION:
With any of the acove limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine tnat the Reactor Coolant System remains acceptable for continued operation or ce in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 8
200*Fand500psig,respectively,withintneYE11owing30 hours.
SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals requirec Oy 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5.
The results of these examinations sna11 De used to update l
Figures 3.4-2 and 3.4 3, SUMMER - UNIT 1 3/4 4-29 8602140118 860207 PDR ADOCK 05000395 P
PDR I
vs TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE m
2 CAPSULE VESSEL LEAD WITIGRAWAL TIME
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IDENTIFICATION LOCATION FACTOR fffY U
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Figure 3 4 3 Reactor Coolant System Pressure Temperature Limits Versus Cooldown Rates SUMER - UNIT 1 I
3/4 4 32
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 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 the activity levels allowed by Figure 3.4-1 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 generator tube 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 1.0 microcuries/ 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.
Reducing T,yg to less than 500*F prevents the release of activity shoult a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
,g3 The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.
A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE LIMITS l
The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G.
1)
The reactor coolant temperature and pressure and system neatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.f;r th: '* :t 'u!'-ecre te-" ice re-ied.
a)
Allowable comoinations of pressure and temperature for specific j
temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates oetween those presented i
may be obtained by interpolation.
l l
i SUMMER - UNIT 1 B 3/4 4-6 l
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE 1.IMITS (Continued) b)
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For nonnal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be acnieved over certain pressure-temperature ranges.
2)
These limit lines shall be calculated periodically using methods provided below.
3)
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.
4)
The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200'F/hr respectively.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 625'F.
5)
System in-service leak and hydrotests shall be perfonned at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.
~
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1972 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code.
Heatup and Coolcow1 limit Curves are Calculated using the most limiting value of RT (reference nil-cuttility temperature). The most limiting RT g of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART
. RT is designated as tne higher of eitner the crop wetgnt nil-ductility tram.sition temperature (NOTT) or the tem:eratu e r
at whicn the material exhibits at least 50 ft Ib of impact energy anc 35-mil lateral expansion (normal to the major working cirection) minus 60*F.
in s, RT increases as tne material is exposec to fast-neutron ractation.
u to finc the mest limiting RT at any time perioc in the reactor's life.
ART cue to the raciation exposure associatec with that time peatoc wit g
be accec to tae original unir*aciated RT The extent of the snift in g;.
RT is enra* cec Oy certa'n chemical elements (suca as co;;ee) preseat ia T
reactor vesse' steels.
Ces'gr c. ses a'cn sr: tre e'fect o' f'ueace aa:
l copper content on ART f or reactor vessel steeis a e sno.n in Fig. e 8 3/4 4-2.
7 8 3/4 4-7 SUMER - UNIT 1 5
e i
I l
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
Given tne copper content of tne most limiting material, tne radiatten-inea:eg t.R*g. ca-ce est ate: fac-5'q.*e W.% tast acutr:n fluer:e (E > 1 u,,;
at tne vessel inner surface, the 1/4 T (wall tnickness), and 3/4 T (wall thickness) vessel locations are given as a function of full-Dower service 11fe in Figure se.*.I. Tne cata f or all otner f erritic materials in the reactor coolant pressure boundary are examineo to insure that no other comoonent wtil De limiting witn respect to 8'g 7-The preirradiation f racture-toughness properties of the V. C. Sunener unit I reactor vessel materials are presented in Table t$.4-1. The f racture-tougnness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan.UI The postirradiation fracture-tougnness properties of the reactor vessel beltline material were obtained directly f rom the V. C. Sunumer Urdt i
~
Vessel Meterial Surveillance Program.
The ASME approacn for calculating tne allowaele limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatus g
or cooldown cannot De greatea than the Pe'ereace stress intensity factor, E,, f or t*ie metal temperature at inat time. K, is obtained from the g
ref erence f racture tougnness curve, cefined in Appencis G of the A5ME Code.E 3 The K, curve is given ey tne equation.:
g K
- 26.78 + 1.223 exo (0.0145 (T-AT
+ 160)]
rp%n (1) g g
1.
" Fracture Tougnmess Requirernents,' Branch Technical Position MTEB 5-2, Chapter S.3.2 in Standard eeview plan for the Review of Safety Analysis Reports f or N. clear Power Plants, LWR Ecition NuREG-0000,1981.
2.
ASME Boiler and peessure vessel Code, Section !!! Division 1 -
Appendices. 'Reles for Construction of Nuclear vessels,' AD:eacia 3
' Protection Against NonduClite Failure,* pp. $$9-66a, 1983 Edition, American Society of Mechanical Engineers, New Yort, 1983.
SupO9ER - UNIT 1 B 3/4 4-8
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SUMMER. UNIT 1 3 3/4 4 10a
REACTOR C001. ANT SYSTEM BASES where K, is the reference stress intensity factor as a function of the g
metal temperature T and tne metal reference nil-ductility temperatare R T@ T.Thus, the gewerning ecuation for !"e heatsc-coo'do n analysis is defined in Appendia G of tne ASME Code as follows:
CK,+Kgg g
g, Eysh=,
(Q iK where
- 1M 35 ine stress intensity factor caused my mem:rane (pressure) stress K
is the stress intensity factor caused by sne inermal gradients C = 2.0 for Level A and Level 8 service limits C = 1.5 for hydrostatic and leak test conditions during which the reactoe Core is not Critical At any time during the heatup or cooldown transient, K is determined by g
the metal temperature at the tip of the postulated flaw, the appropriate value for RT and the reference fracture toughness curve.
The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thernal) stress intensity factors, K!t, f r the reference flaw are computed.
From Equation 2, the pressure stress intensity f actors are obtained and, f rom these, the allowable pressures are calculated.
COOLDOWN For the calculation of the allowable Dressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of 1
the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the iPside, w"ich ircrease wits increasing cooldown rates.
Allow 3:le pressure-tem:erature relations are generated for Doth steady-state and finite cooldown rate situations.
From these relations, com:osite limit curves are constructed f or each cooldo.n rate of interest.
SupWER - UNIT 1 8 3/4 4-11 i
l
{
REACTOR COOLANT SYSTEM BASES The use of the composite curve in the coolcown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature,'whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.
During cooldown, tne 1/4 T vessel location is at a higner temperature than the fluid adjacent to the vesse; 10.
This condition, of course, is not true for the steady-state situation.
It f ollows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K, at the 1/4 T tocation for finite cooldown rates than for steady-state g
operation! Furthermore, if conditions exist such that the increase in K,
g eveeeds rg, the ralrulated allowable pressure durino cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The therwal gradients during heatup i
produce compressive stresses at the inside of the wall that alleviate the I
tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the. coolant temperature; therefore, the E for the 1/4 T gg crack during heatup is lower than the K, for the 1/4 T crack during g
steady-state conditions at the same coolant temoerature.
During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive therwal stresses and lower K;p's do not offset each other, and the pressure-temperature curve based on steady-state conditions no i
SUMMER - UNIT 1 B 3/4 4-12 l
'E"
- EA
- *00 :::_2.-
BASES longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered.
Therefore, both cases have to De analyzec in order to insure that at any coolant temperature the lower value of the 3110wable pressure calculated f or steady-state anc finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 7 deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surf ace during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses 'present. These therwal stresses are dependent on both the rate of
~
heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for buth the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves undee consideration.
The use of the compostte curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup esrp, the controlling condition svitches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Then the composite curves for the heatup _r. ate data and the cooldc.n,
rate data are adjusted for possible errors in t9e pressure and temperature l
sensing instruments by the values indicated on the respective curves.
Finally, the new 10CFR50( 3 rule which addresses the metal temperature of the closure head flange aad vessel flange regions is considered.
This 10CFR50 rule states that the metal temperature of the closure flange regions must exceed the material RT by at least 120*F for normal operation when the i
3.
Code of Federal Regulations, 10CFR50, Appendia G, ' Fracture Toughness
\\
Requirements," U.S. Nuclear Regulatory Commission. Washington, D.C.,
Amended May 17, 1983 (48 Federal Register 24010).
SUMMER - UNIT 1 8 3/4 4-13 I
i
__________._--_.__m
REACTCR 200.4N' i s'il BASES pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig f or V. C. Summer Unit 1).
Table 9$4.4]indicatesthatthelimitingRT NOT of 10'F occurs in the head flange of V. C. Summer Unit 1, and the minimum allowable temperature of this region is ' 0*F at pressures greater than 621 psig.
Limit curves for normal heatup and cooldown of the primary Reactor Coolant i
The System have been calculated using the methods discussed derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.I'l Transition temperature shif ts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure Charpy test specimens from Capsule U indicate vessel surveillance program.
that both the surveillance weld metal and core region intermediate shes' pItte f 30*F at a fluence of 6.39 x code no. A9154-1 exhibited shifts in RTNOT IO 2
This shift is well within the appropriate design curve 10 n/cm.
(Figure 6%4At) prediction. Therefore, the heatup and cooldown curves in Figures 331 and sx 3 are based on the trend curve in Figurelskal and these curves are The heatup curve in applicable up to 8 ef f ective f ull power years (EFPY). However, the Cooldown Figure 39-1 is not impacted by the new 10CFR$0 rule.
curve in Figure 3N-315 impacted by this 10CFR50 rule.
Allowable combinations of tempe ature and pressure for specific temperature change rates are"below and to the right of the limit lines shown on the heatup The reactor must not be made critical'until and cooldown curves.
pressure-temperature combinations are to the right of the criticality limit line shown in Figure 3.1-1. This is in accition to other criteria which must be met bef ore the reactor is made critical.
The leak test l' 't c. ve s*:<a ir Siga-e IM-t rep eseats e'ai a-te pe at. e requirements at tne lear test pressure specified by applicable codes.
The leak test limit curve was cetermined by methods of References 2 and a.
l Figures 191 and 3.8-3 defiae limits for insuring prevention of nonductile failure.
4.
' Pressure-Temperature Limits
- Chapter 5.3.2 in Stancare neview plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
SUMMER - UNIT 1 83/44-14 Amendment No. 26
~
~v
REAC*:R : C.P.' i*i~i" 8ASES Although the pressurizer operates in temoerature ranges above tnose for which there is reason for concern of non-ductile f ailure, operating limits are provided to assure compatibility of operation witn tne f atigue analysis performed in accordance with the ASME Code reautrements.
The OPERABILITY of two RHRSRVs or an RCS vent opening of at least 2.7 souare inches ensures that tne RCS will be protected fecm pressure transients -nicn could exceed tne limits of Appendix G to 10 CFR part 50 nen one or more cf the RCS cold legs are less than or equal to 300*F.
Either RHRSRV nas adequate relieving capacility to protect the RCS from overpressurization wnen the transient is limited to either (1) the start of an idle RCP with the seconcary water temperature of tne steam generator less than or equal to 50*F acove tne RCS cold leg temperatures or (2) the start of a HPSI oumo and its injection into a water solid RCS.
i i
i SLMMER - UNIT 1 g 3/4 4 14,
ATTACHMENT B No Significant Hazarda Determination The proposed changes to the Technical Specifications include revisions to section 3/4.4.9, " Pressure / Temperature Limits -
Reactor Coolant System," and its bases.
These changes are being requested as a result of the information obtained from the review of the first surveillance capsule removed from the reactor vessel.
Results of this review are contained in the Westinghouse Topical Report WCAP 10814, " Analysis of Capsule
'U' from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program" transmitted to the NRC Staff by a letter dated November 8,1985 from Mr. D.
A. Nauman to Mr. H. R.
Denton.
SCE&G has determined that the proposed changes involve a no significant hazards determination.
The amendment will not:
1) involve a significant increase in the probability or consequences of an accident previously evaluated because the changes are being made to make the Technical Specifications more accurate as a result of the data obtained from the review of the first reactor vessel specimen; 2) create the possibility of a new or different kind of accident from any accident previously evaluated because the physical plant design is not being changed; or 3) involve a significant reduction in a margin of safety because the change will make the Technical Specifications reflect the requirements dictated by the predicted service life conditions of the reactor vessel.