ML20210C187
| ML20210C187 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 09/09/1986 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210C191 | List: |
| References | |
| NUDOCS 8609180267 | |
| Download: ML20210C187 (19) | |
Text
.
- uag(o UNITED STATES
+
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NUCLEAR REGULATORY COMMISSION M'
E WASHINGTON, D. C. 20555 i
V/
.n SOUTH CAP 0 LINA ELECTRIC & GAS COMPANY SOUTP CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMMED NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. NPF-17 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by South Carolina Electric & Gas Company and South Carolina Public Service Authority (the licensees) dated February 7, 1986, complies with the standards and requirements of the Atomic Energy Act of 1054, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFD Chapter I; B.
The facility will operate in conformity with the application,-
the provisions of the Act, and the rules and regulations of the Conunission; C.
There is reasonable assurance (i) that the activities authorized by this amerdment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conduc'ted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 l
of the Commission's regulations and all applicable requirements l
have been satisfied.
l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license t
amendment, and paragraph 2.C.(2) of Facility Operating License.
No. NPF-12 is hereby amended to read as follows:
8600h P
PDR P
i
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 53, are bereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This amenoment is effective as of its date of issuance, and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGl!LATORY COMMISSION h
W Lester S. Rube stein, Director PWR Pro,iect D rectorate #2 Division of PWR Licensing-A Office of Nuclear Reactor Regulation
Attachment:
Changes to the Techrical Specifications Date of Issuer.ce: September 9,1986 i
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 53 TO FACILITY OFERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines. indicating the areas of change. Corresponding overleaf pages are also provided to maintain document completeress.
Remove Pages Insert Pages 3/4 4-29 3/4 4-?9
-3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 B3/4 4-6 B3/4 4-6 B3/4 4-7 B3/4 4-7 B3/4 4-8 83/4 4-8 B3/4 d-10 B3/4 4-10 83/4 4-10a B3/4 4-10a B3/4 4-11 B3/4 4-11 B3/4 4-12 B3/4 4-12 B3/4 4-13.
83/4 4-13 B3/4 4-14 B3/4 4-14 83/4 4-14a
REACTOR' COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coclant System (except th'e 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.
APPLICABILITY:
At all times.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engir.eering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBi 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 200 F and 500 psig, respectively, within the Y8110 wing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
a 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 required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5.
The results of these examinations shall be used to update Figures'3.4-2 and 3.4-3.
I l
SUMMER - UNIT 1 3/4 4-29 Amendment No.53 i
i
TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITH0RAWAL SCHEDULE
=
CAPSULE VESSEL LEAD WITH0RAWAL TIME, E
IDENTIFICATION.
LOCATION
. FACTOR EFPY a
U 343*
3.7 1st Refueling e
V 107*
3.1 3rd Refueling X
287*
3.1 5th Refueling W
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2.7 10th Refueling Y
290 2.7 17th Refueling Z
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SumER - UNIT 1 3/4 4-31 Amendtent No. 53
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Reactor Cootent Sysem Proesure. Tomoorsture Limis vmus Ceoiseen Raus Amendment f;o. 53 StamER - UNIT 1 3/4 4-32
r-REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
There-fore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining the chemistry within the Steady State Limits provides ' adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen,
~ hloride and fluoride limits are time and temperature dependent.
Corrosion cstudies show that-operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on The time interval the structural integrity of the Reactor Coolant System..
permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady State Limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY i
The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.
The values for the limits l
on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the Virgil C. Summer site, such as site boundary location and meteorological conditions, were not considered in this evaluatien, l
SUMMER - UNIT 1 B 3/4 4-5
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, to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
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 1
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 heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.
a)
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation.
SUMMER - UNIT 1 B 3/4 4-6 Amendment No. 53
o REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) b)
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, mav limit the heatup and cooldown rates that can be achieved 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 abcve 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 respective _ly.
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 performed 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 cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature).
The most limiting RT f the material in the core region of the reactor vessel is deter-l NDT mined by using the preservice reactor vessel material properties and I
estimating the radiation-induced ART RT is designated as the NOT.
NDT higher of either the drop weight nil-ductility transition temperature t
(NDTT) or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35-mil lateral. expansion (normal to the major work-ing direction) minus 60 F.
l RT increases as the material is exposed to fast-neutron radiation.
NDT Thus, to find the most limiting RT at any time period in the reactor's NDT life, SRT due t the radiation exposure associated with that time NDT period must be added to the original unirradiated RT The extent of NDT.
the shift in RT is enhanced by certain chemical elements (such as NDT copper) present in reactor vessel steels.
Design curves which show the effect of fluence and copper content on ART f r reactor vessel steels NDT are shown in Figure B 3/4 4-2.
l l
SUMf2R - UNIT 1 B 3/4 4-7 Amendment No. 53 l
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REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
Given the copper content of the most limiting material, the radiation-induced ART can be estimated from Figure B 3/4.4.2.
Fast neutron NDT fluence (E > 1 Mev) at the vessel inner surface, the 1/4 T (wall thick-ness), and 3/4 T (wall thickness) vessel locations are given as a func-tion of full power service life in Figure B 3/4.4.1.
The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RTNOT' The preirradiatior, fracture-toughness properties of the V. C. Summer Unit I reactor vessel materials are presented in Table B 3/4.4-1.
The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan.2 The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the V. C. Summer Unit 1 Vessel Material Surveillance Program.
The ASME approach for calculating the allowable 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 3
heatup or cooldown cannot be greater than the reference stress intensity factor, Kyp, for the metal temperature at that time.
K is obtained from IR the reference fracture toughness curve, defined in Appendix G of the ASME Code.2 The K curve is given by the equation:
IR K
= 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]
Equation (1)
IR 1" Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
2ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendices,
" Rules for Construction of Nuclear Vessels," Appendix G, " Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983.
SUMMER - UNIT 1 B 3/4 4-8 Amendment No. 53
TABLE B 3/4.4-1 E
3 REACTOR VESSEL TOUGHNESS 9
MIN. 50 FT-LB t
Cu NOTT 35 MIL TEMP.
RT AVG. UPPER c:
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114.5 Inlet Nozzle SA508 Class 2 Inlet Nozzle SA508 Class 2
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-20 135 Outlet Nozzle SA508 Class 2
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V REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) where K is the reference stress intensity factor as a function of the IR metal temperature T and the metal reference nil-ductility temperature RT Thus, the governing equation for the heatup-cooldown analysis is NDT.
defined in Appendix G of the ASME Code as follows:
Equation (2)
C K;g + kit 1 IR where K
is the stress intensity factor caused by membrane (pressure)
IM stress K
is the stress intensity factor caused by the thermal gradients It C = 2.0 for Level A and tevel B service limits C = 1,5 for hydrostatic and leak test conditions during which the reactor core is not critical is determined by At any time during the heatup or cooldown transient, Kyg
-the metal temperature at the tip of the postulated flaw, the appropriate The thermal value for RTNDT, and the referance fracture toughness curve.
stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, g, for the reference flaw are computed.
From Equation 2, the pressure K
stress intensity factors are obtained and, from these, the allowable pressures are calculated.
COOLDOWN For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of 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 inside, which increase with increasing cooldown rates.
Allow-able pressure-temperature relations are generated for both steady-state and finite cooldown rate situations, From these relations, composite limit curves ate constructed for each cooldoon rate of ir.terest.
SUMMER - UNIT 1 B 3/4 4-11 Amendment No.53
REACTOR COOLANT SYSTEM BASES-COOLDOWN (Continued)
The use of the composite curve in the cooldown analysis is necessary be-cause control of the cooldown procedure is based on measurement of reactor j
J coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.
During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID.
This condition, of course, is not true for the steady-state situation.
It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4 T location for finite cool--
IR down rates than for steady-state operation.
Furthermore, if conditions exist.
such that the increase in K exceeds Ky, the calculated allowable pressure IR during cooldown will be greater than the steady-state value.
l 1
The above procedures are needed because the-e is no direct control on tem'-
perature at the 1/4 T location and, therefore, ellowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. Th( use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period.
Three separate calculatiens are required to dete mine the limit curves for finite heatup rates.
As is done in the cooldown analysis, allowable pressure-i 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 thermal gradients during heatup produce com-pressive stresses at the inside of the wall that alleviate the tensile stresses i
produced by internal pressure.
TN? metal temparature at the crack tip lags tte coolant temperature; therefore, the K for the 1/4 T crack during heatup is IR lower than the K f r the 1/4 T crack during steady-state conditions at the IR same coolant temperature.
During heatup, especially at the end of the tran-sient, conditions may exist such that the affects of compressive thermal t
I stresses and lower Kgg's do not offset each otter, and the pressuee-temperature curve based on steady-state conditions no longer represents a lower bound of SUMMER - UNIT 1 B 3/4 4-12 Amendment No.53
REACTOR COOLANT SYSTEM BASES all similar curves lfor finite heatup rates when the 1/4 T flaw is considered.
Therefore, both cases have to be analyzed in order to insure that at any cool-ant temperature the lower value of the allowable pressure calculated for steady state and.ficite heatup rates is obtained.
I The second poetion of the heatup analysis concerns the calculation of
}
pressure-ten.perature limitations for the case in which a 1/4 T deep outside sur-face flaw is assumed.
Unlike the situation at the vessel inside surface, the thermal gracients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses pregent.
These thermal stresses are deoendent 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 both 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 a,1d finite heatup rate data.
At any given temperature, the allow-able pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, ever t.he course of the heatup ramp, the controlling condition switches 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 rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
Finally, the new 10 CFR 503 rule which addresses the metal temperature of the clotJre head flange and vessel flange regions is considered.
The 10 CFR 50 rule States that the metal temperature of th closure flange regions must exceed the,materlat RT by at least 120'F for normal operation when the pressure l
NDT 3 Code of Federal Regulations, 10 CFR 50, Appendix G " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,
Amended May 17, 1983 (48 Federal Register 24010).
SUMMER
- UNIT 1 B 3/4 4-13 Amendment No.53
\\
[
' BASES exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for I
LV. C. Summer Unit 1).
Table B 3/4.4.1 indicates that the limiting RTNDT "I 10*F occurs in the head flange of V. C. Summer Unit 1, and the minimum allow-able temperature of this region is 130*F at pressures greater than 621 psig.
Limit curves for normal heatup and cooldown of the primary Reactor Coolant i
System have been calculated using the methods discussed.
The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.*
I i
Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program.
Charpy test specimens from Capsule U indicate that both the surseillance weld metal and core region intermediate shell plate code no. A9154-1 exhibited shifts in RT f 30*F at a fluence of 6.39 x NDT
~
101s n/cm.
This shift is well within the appropriate design curve 2
(Figure B 3/4.4.2) prediction.
Therefore, the heatup and cooldown curves in 1
l Figures 3.4-2 and 3.4-3 are based on the trend curve in figure B 3/4,4.2 and these curves are applicable up to 8 effective full power years (EFPY). The heatup curve in Figure 3.4-2 is not impacted by the new 10 CFR 50 rule.
How-ever, the cooldown curve in Figure 3.4-3 is impacted by this 10 CFR 50 rule.
1 Allowable combinations of temperature and pressure for specific temperature i
change rates are below and to the right of the limit lines shown on the heatup and cooldown curves.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown i
in Figure 3.4-2.
This is in addition to other criteria which must be met before the reactor is made critical.
i i
The leak test limit curve shown in Figure 3.4-2 represents minimum tempera-
'j ture requirements at the leak test pressure specified by applicable codes.
The t
leak test limit curve was determined by methods of References 2 and 4.
(
Figures 3.4-2 and 3.4-3 define limits for insuring prevention of nonductile failure.
6 r
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4" Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition, 6
NUREG-0800, 1981.
SUMMER - UNIT 1 B 3/4 4-14 Amendment No. 53 j
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