ML20054M278
| ML20054M278 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/30/1982 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054M279 | List: |
| References | |
| NUDOCS 8207120219 | |
| Download: ML20054M278 (15) | |
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UNITED STATES E
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DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.53 License No. DPR-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company (the licensees) dated April 5,1982 complies with the standards and requirements of the Atomic Energy A,ct of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
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 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
(
i E.
The issuance of this amendment is in accordance with 10 CFR Part l
51 of the Commission's regulations and all applicable requirements have been satisfied.
~
8207120219 820630 PDR ADOCK 05000334 P
. 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. DPR-66 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 53, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
All EGULATORY COMMISSION FpTHENUC
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[ Operating Reactors B anch #1 Division of Licens g
Attachment:
Changes to the Technical Specifications Date of Issuance: June 30,1982
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 53 TO FACILITY OPERATING LICENSE NO. DPR-66 l
DOCKET NO. 50-334 4
i Revise Appendix A as follows:
Remove Pages Insert Pages 3/4 4-24 3/4 4-24 3/4 4-25 3/4 4-25 3/4 6-14 3/4 6-14 B3/4 4-6 B3/4 4-6 83/4 4-6a B3/4 4-6b i
B3/4 4-7 B3/4 4-7 B3/4 4-7a 83/4 4-7b B3/4 4-8 B3/4 4-8 B3/4 4-8a B3/4 4-9 B3/4 4-9 t
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3MB CURVE APPLICAsLE POR HEATUP RATES UP TO 80*FMR POR THE SERVICE PERIOD UP TO 8 EFPY AND CCNTAINS MARGINS OF 10*F Afe 80 PSIG POR POSS4SLE INSTRUMENT ERRORS 3 53 LEAK TE5T LIMIT 5 MATERIAL PROPERTY SASIS 2EMM3
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CONTROLLING MATERIAL: PLATE METAL E
CDPPER CONTENT:
0.20W7%
I PHOSPHORUS CONTENT:
0.010 WT%
E RTMDTINmAL:
27'F j
RTNOTAFTER 8 EFPY:
1/47,274*F g
m T.i e r w
g 1500 O
~
fi!
1000 MATUP RATES UP TOEl*FMR 500 l
mmCALITY LIMrr-sAsED ON INSERVICE j
HYDROSTATIC TErf TEMPER ATURE 1414*Fl l
POR THE sERvicT PERIOO UP TO e EFPY l
l l
l l
0 0
100 2M 300 400 500 1
1 INDICATED TEMPER ATURE (*F) l Figure 3.4-2 Beaver Valley Unit No.1 Reactor Coolant System Heatup Limitations Applicable for 19e First 6 EFPY 3/4 4-24 Amendment No. 53
l i
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3300 MATERIAL PROPERTY BA315 CINsTROLLING MATERIAL:
PLATE METAL 2500 COPPER CONTENT:
0.20 WT%
PHOSPHORUS CONTENT:
0.010 WT%
RTNDTINITIAL:
27'F RTNUTAFTER S EFPY:
1/47,274*F 3/47,144'F O
g 2000 E
CURVE APPLICABLE F<R COOLDCNN RATES E
UP TO 1DO*FMR FOR THE SERVICE PERIOO '
3 UP TO S EFPY AND CONTAINS dAHGINS OF E
10*F AND 00 PSIG FOR POSSIBL2 ImmutENT g
ERRORS i
e.
c 1500 U
5 1000 cOOLDOwN g
_ RATE 3 *F MR O
2n so en 100 0
0 100 200 300 400 500 INDICATED TEMPERATURE FF) i Figure 3.4-3 Beaver Valley Unit No.1 Reactor Coolant System Cooldown Limitations Applicable for the First 6 EFPY 3/4 4-25 A:nendment Nc. 53 1
)
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
At least once per 18 months by verifying that on a Containment Pressure--High-High signal, the recirculation' spray pumps start automatically as follows:
RS-P-LA and RS-P-2B 210 2 5 second delay RS-P-2A and RS-P-1B 225 2 5 second delay c.
At least once per 18 months, during shutdown, by verifying, that on recirculation flow, each outside recirculation spray pump develops a discharge pressure of 2115 psig at a flow of 2:2000 3pm.
d.
At least once per 18 months during shutdown, by:
1.
Cycling each power operated (excluding automatic) valve in the flow path not testable during plant operation, through at least one complete cycle of full travel.
2.
Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
e.
At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
BEAVER V ALLEY-U NIT 1 3/4 6-14 Amendment No. 53 i
REACTOR COOLANT SYSTEM BASES The heatup analysis also covers 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 at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer vall 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 curse similar to that described for the heatup of the inner wall cannot be defined. Subsequently, 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 heatup limit curve, Figure 3.4-2, is a composite curve which was pre-pared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 6 EFPY.
The reactor vessel materials have been tested to determine their initial RT' DT; t e results of these tests are shown in Table B 3/4.4-1.
Reactor N
operation and resultant fast neutron (E> l Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based NDT.
upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.
The heatup and cool-down limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT as well as adjustments for possible errors in the pressure NDT and temperature sensing instruments.
BEAVER V ALLEY-U NIT 1 B 3/4 4-6 Amendment No. 53
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ad Gol Mm Effect of Fluence, Cooper Content, and Phosphorus Content on 4RTNDT for Reactor Vessel Steels per Regulatory Guide 1.99 Figure B 3/4.4-2 Amendment No. 53 BEAVER VALLEY - UNIT 1 B 3/4 4-6b
TABLE B 3/4.4-1 REACTORVESSELT00GilNE55OATA(UNIRRADIATEO) w tlp Shelf Ene (Ft-lb)
T Component Heat No.
Code No.
Material Type g) g)
ny Closure Head. C6213-18B6610 A5338 CL.1
.15.010
-40 0*
121 4
Dome Closure Head 9.
Seq.
A5518-2 B6611 A5338 CL. 1
.14.015
-20
-20*
131 sH Closure Head A508 CL. 2
.08..C07 60*
60*
> 100 Flange ZY3758 A500 C1. 2
.12.010 60*
60*
166 Vessel Flange ZY3661 A508 C1. 2
.10 008 60*
60*
82.5 Inlet Nozzle 9-5443 A500 C1. 2
.10 010 60*
60*
94 Inlet Nozzle 9-5460 A500 C1. 2
.08 007 60*
60*
97 Inlet Nozzle 9-5712 A508 C1. 2
-.008 60*
60*
97 Outlet Norzle 9-5415 A508 C1. 2 007 60*
60*
112.5 m
Outlet Nozzle 9-5415 A508 Cl. 2
.09.007 60*
60*
103 t'
Outlet Nozzle 9-5444 Upper Shell 123V339 A508 C1. 2 010 40 40*
155 S
Inter. Shell C4 381-2 86607-2 A5330 C1.1
.14.015
-10 73 123 82.5 Inter. Shell C4 381-1 B6607-1 A5338 C1.1
.14 015
-10 43 128.5 90 Lower Shell C6317-1 B6903-1 A533B C1.1
.20 010
-50 27 134 80 tower Shell
.C6293-2 B7203-2 A5338 Cl. 1
.14 015
-20 20 129.5 83.5 30 30*
143 A508 Cl. 2 Trans. Ring 123V223 g
Bottom Hd.
g Seg.
C4423-3 B$618 A5338 C1. 1
.13 008
-30
-29*
124 Bottom Ha.
Oome C4482-1 86619 A5338 C1. 1
.13.015
-50
-33*
125.5 g
Core Re31on
> 100 g
.304.37.013 0*
136.5
-40
- Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWD - Major Working Ofrection HMWD - Nonnal to Major Working Direction
REACTOR COOLANT SYSTEM BASES Heatup and cooldown limit curves are calculated using the mst limiting value of RT (reference nilductility temperature). The most limiting RT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties RT is designated as the and estimating the radiation-induced ARTg.
higher of either the drop weight nil-ductility transition temperature (Tg) or the temperature at which the material exhibits at least 50 f t Ib of ' impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.
RT increases as the material is exposed to fast-neutron radiation.
g Thus, to find the most limiting RT at any time period in the reactor's g
life, A RT due' to the radiation exposure associated with that time period must be added to the original unirradiated RT The extent of the shift in RT is enhanced by certain chemical elements (such as copper and phosphorus) present in reactor vessel steels. The Regulatory Guide 1.99 trend curves which show the effect of fluence and copper and phosphorus contents on A RT for reactor vessel steels are shown in Figure B 3/4.4-2.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ART can be estimated from Figure B 3/4.4-2.
Fast-neutron fluence (E > 1 Mev) at ~ the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel locations are given as a function 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 1_.niting with respect to RTNDT*
B 3/4 4-7a Amendment No. 53 BEAVER V ALLEY-U NIT 1
BASES The preirradiation fracture-toughness properties of the Beaver Valley 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 E1I plan.
The postirradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Beaver Valley Unit i 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 heatup 7
and cooldown cannot be greater than the reference stress intensity factor, K
- E*##
s obtained hom the IR' IR reference fracture toughness curve, defined in Appendix G to the ASME Il Code.
The K cme s ghen by the equadon:
IR K
W IR "
- NDT where K s the reference stress intensity factor as a hnction of de IR metal temperature T and the metal reference nilductility temperature RT Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Code as follows:
CK
+K
-<K g
_ IR 1.
" Fracture Toughness Requirements," Branch Technical Position MTEB No. 5-2, Section 5.3.2-14 in Standard Review Plan, NUREG-75/087,1975.
2.
ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendices,
" Rules for Construction of Nuclear Vessels," Appendix G.
" Protection Against Nonductile Failure," pp. 461-469, 1980 Edition, American Society of Mechanical Engineers, New York,1980.
', BEAVER V ALLEY-U NIT 1 B 3/4 4-7b Amendment No. 53
REACTCR COOLANT SYSTEM BASES where K
is the stress intensity factor caused by membrane (pressure) stress 73 K
is the stress intensity factor caused by the thermal gradients 7
K is a function of temperature to the RT IR NDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value t e re erence frac ure ug ness curve. The thermal stresses for RTNDT, an resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K7, for the reference flaw are computed. From equation (4-2), the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
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. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown 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, the 1/4T 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 s itua tion.
It follows that, at any given reactor coolant tempe ra ture, the AT developed d,uring cooldown results in a higher value of K at the 1/4T location IR for finite cooldown rates than for steady-state operation.
Furthermore, if conditicos exist such that the increase in K exceeds K the calculated 7g 7t, allowable pressure during cooldown will be greater than the steady-state va lue.
BEAVER V ALLEY-U NIT 1 B 3/4 4-8 Amendment No. 53
REACTOR COOLANT SYSTEM BASES s
The above procedures are needed because there is no direct control on temperature at the 1/4T 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 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K f r the 1/4T crack during IR heatup is lower than the K rte 1 cra ur ng steady-state conditions IR at the same coolant temperature.
During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower KIR's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefo re, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
iBEAVER V ALLEY-U NIT 1 B 3/4 4-8a Amendment No. 53
6 REACTOR COOLANT SYSTDi BASES The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 'l/4T deep outside sur-face flaw is assumed. Unlike the situation at 6he vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal 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 both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed bases on a peint-by-point comparison of the steady-state and finite heatup rate data. At any given tem-perature, the allowable pressure is taken to be the lesser of the 'Aree 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, over the course of the heatup ramp, the controlling condition swit hes from the ingide to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Then, com-posite 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.
The actual shif t in NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and 3 3/4 4-9 Amendment No. 53 BEAVER V ALLEY-U NIT 1
-