ML20006E664
| ML20006E664 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 02/08/1990 |
| From: | Jocelyn Craig Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20006E666 | List: |
| References | |
| NUDOCS 9002260195 | |
| Download: ML20006E664 (16) | |
Text
{{#Wiki_filter:1 s. ' ' " " %( j ...-- - - - (y UNITED STATES e,([jN NUCLEAR REGULATORY COMMISSION. .y 3 .t ~ r,, WASHINGTON, D. C. 20555 ] 7 44,4.... q i COMMONWEALTH EDISON COMPANY DOCKET NO. 50-454 -l BYRON STATION, UNIT 1 i AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37-l License No. NPF-37 l '.. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Comonwealth Edison Company (thelicensee)datedNovember 17,1989,. supplemented January ~ 10, y 1990,. complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations-set forth in 10 CFR Chapter I; I B.- The facility will-operate in conformity with the application, the provisions of the Act, and the rules and regulaticas of the ' Comission; C. Thereisreasonableassurance(1)thatthe-activitiesauthorized 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; and 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 l Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows: 9002260195 900200 [~ PDR ADOCK 05000454 P PNU
'. I., ~ - 2 '- -(2) Technical Specifications The Technical Specifications contained in Appendix A as. revised through-Amendment No. 37 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and-the Environmental Protection Plan. i 3.. This license amendment is effective as of the date of its issuance. FDP,THE NUCLEAR REGULATORY COMMISSI0tt i ,JohnW.Craig, Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special. Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: February 8,-1990
a F ' Wio .j ug [ ' : o,j UNITED STATES :
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NUCLEAR REGULATORY COMMISSION-7
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. WASHINGTON, D. C. 20555 x j; ....+ COMMONWEALTH EDISON COMPANY DOCKET NO. 50-455 2 BYRON STATIOL UNIT 2- . AMENDMENT TO FACILITY OPERATING. LICENSE Amendment No. 37 License No. NPF-66 - 1 ~. .The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Comonwealth Edison Company. (thelicensee)datedNovember 17,1989, supplemented January 10, 1990, complies with the standards and requirements of the Atomic. Energy Act of 1954, as amended (the Act) and the Comission's rulesL and regulations set forth in 10 CFR Chapter 1; B. The facility will operate in conformity with the application,. the provisions of the Act, and the rules and regulations of.the Comission; C. There is reasonable assurance (i) that the activities-authoriz'ed by~.this amendment can be conducted withott 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;. and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all. applicable requires 4nts 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-66 is hereby amended to read as follows:
y a :-.:
- n. a.
1 1 _y_ (2) T,e,chnical-Specifications The= Technical Specifications contained.in Appendix A-(NUREG-1113), j ~as revised through Amendment No."37 and revised by Attachment 2 1 to NPF-60,'and the Environmentil Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37.. dated February 14, 1985, are hereby incorporated into this license. ~ contains a revision to Appendix A which is hereby. ' incorporated into this' license.- The licensee shall operate ~the facility in-accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date of its issuance. KOR HE NUCLEAR REGULATORY COMMISSION - i /'hohnW.Craig,. Director Project Directorate III-2 i Division of Reactor Projects - III, } IV, V and Special Projects
Attachment:
Changes to the Technical . Specifications Date of.lssuance: February 8, 1990 i -j 9 i l i 1 --J
x; c .5 '., s U, ? ' ATTACHMENT TO LICENSE AMENDMENT NOS. 37 AND 37 l - FACILITY OPERATING LICENSE N05. NPF-37 AND NPF-66 DOCKET NOS.'50-454 AND 50-455 I Revise Appendix A'as follows: . /2 Remove Pages Insert Pages 3/4 4 3/4'4-33 3/4 4-34 3/4 4-34 3/4 4-36' 3/414 ' 3/4 4 3/4 4-39 3/4 4-40 3/4 4-40a 3/4 4-40b B 3/4 4-8 B 3/4 4-8 B 3/4 4-11 B 3/4 4-11 B 3/4 4-12 B 3/4 4-12 B 3/4-4-15 B 3/4 4-15 B 3/4 4-16 B 3/4 4-16 4 . l l-l' I,. t .)^
l c1.. S' e .i 11 ll Hill 11l ll l lll l - I l-l l i 1 i d FEst M ' ii i i ll It ii l i t. 1 ii 1 - ACCEPTABLE FORI i= essia6 ruerisiv saass -{ ( HYDR 0!TE5T51 coarsettias mTEasat. av vPip 4f i f i j 1 j j sistLL SP = leta i i p ( .( ] CePPER Conftet:steestev&fii'tL7
== --, assimED as e.le si t' ,j j y i ,et,,, initial: see, J ,87ast affte 33 (FPTt 8 g e 1/4f. iggey 'j sist, es*p l I' s_ i A 1 wavt &PPLICABLE Fat ut&fsP tart $ ve to ise rtaa pee in stefitt ,j y; e E. ,Ptales or to 33 trPt. Aas contains j j ,,,,masia er new ano se Psis een f1 / e possiett instavutar tasons. / I / I Ii / / I ACCEPTA8d-- .I UNACCEPTABLE f / OPERATI0tt..). t 1 QPCRATIONI f f t 1000 i i-J i g ll ll f l f 1 I statur aaitsm I / I" F or to see*F/ne N / I" -l l l l if pd i l 11 1 I 4r I,' ames,en IMERT:CE* " ili -I A / areasstATIC 7t874- = + ftenettaiset (27t*F) i Catticatiri e' iI fa8 fut SERTiet I . timitt L l Ptsies vP fo arittre l l l l l \\\\1 111 l l i '. l jl;I l l 1 0' I i l 0 0 100 200 300 e00 llelCATED TEMPEA&TURE (*F) FIGURE 3.4-2A REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY*(UNIT-1) + l i
- Applicability date has been reduced per Regulatory Guide 1.99 Rev. 2 to 29.5 EFPY. The calculation BYRON - UNITS 1 & 2 3/4 4-33 to determine applicability utilised actual copper content of 0.05 wt%.
AMENDMENT NO. 37 1
= -.. - 4 5, MATERI AL PROPERTY BASIS - CONTROLLING MATERIAL:. CIRCUNFERENTIAL WELD . 1 RT AMER 16lEFM : 1/4T, H6.5 4 NDT ~ l, 3/4T 122.8'F e k CURVES APPLICABLE FOR HEATUP RATES UP TO 100'F/HR FOR THE-SERVICE PERIOD UP TO 16 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.. l l. 2500 u.i - .ii iiii,
- i. iii i iiiiieii iiiiei i
Leak lest l l 2250 Limit r 1 1 1 r j i J f f 6
- I J
J ' 4 f A E I l' 2000 l f f ' i J J t i 37;n .i r ,i i,,, ,I 'i J i e i 'l ! Acceptable 7 15c0 i, Operation i 4 i, f; 6 j; y, Unacceptable / W 33gei E Operation, ,,1 g i. i 10 0 8 Heatup-Rates Up*To r 1-y 750 inn *F/Hr I l / Criticality Limit I Based on Ir. service I 500 / Hydrostatic Test i Temperature (292'F) ~ ~ for the Service 250 Period Up to 16 EFPY-0 O 60 100 160 200 250 300 350 400 450 600 ) IMDicATED TCWPERATUet (DEC.r) FIGURE 3.4-2b REACTOR COOLANT SYSTEli HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) BYRON - UNITS 1 & 2 3/4 4 34 MENMENT NO. M
- . 7
.. = w .:.. MATERIAL PROPERTY BASIS N ', J 9 CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD RT AFTER 16 EFPY: 1/4T 146.5'F NDT 3/4T,122.8'F- - CURVES APPLICABLE FOR C00LDOWN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ~ ERRORS. asoO ar ~ I I azSe i
- 6 F
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l l t ,,,I i*,1 I 4, 6I l ,,i e 1750 l ll ll. l 6 ,i i# 3 1 .r _o 1500 . i 4,, r, i i ,.1 i g i 4,i ,6 i 4 1 w l Unacceptable lf; lll1 l l ltaso; l Operation f Acceptable e -1600l ....lll!ll,; , ; ; l ;'., ~0perat10n l,.. l n. ,i,, .s.... ii, w ... ii m,i !..ii, ii.. , i i.. s. i .....i, Y ISO P C00ldown P-i se-i t-- Rates . zzm i ep,g7 sss,r +, , s sss,
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i 20,, s 25e 40 e , no 100 i,ii e O 50 100 150 200 250 300 350 400 450 500 IN01cATCO TCWPERAfunt (Orc.r) e l FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LDOWN LIti!TATIONS i APPLICABLE UP TO 16 EFPY (UNIT 2) BYRON - UNITS 1 & 2 3/4 4-36 AMENDMENT NO. 37 l-L
r. Q: REACTOR COOLANT SYSTEM ?"
- 0VERPRESSURE PROTECTION SYSTEMS-LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:-
a. Two residual-heat removal (RHR) suction' relief valves each with a Setpoint of 450 psig i 1%, or b. Two power-operated relief valves (PORVs) with lift Setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4a for Unit 1 (Figure 3.4-4b for Unit 2), or l c. The Reactor Coolant Syster., (RCS) depressurized with an RCS vent of greater' than or equal to 2 square inches. APPLICABILITY: MODES 4 and 5, and MODE 6 with the reactor vessel head on. ACTION: a. With one PORV and one RHR suction relief viive inoperable, either restere two PORVs or two RHR suction relief valves to OPERABLE status-within 7 days or depressurize and vent the RCS through at least a 2_ square inch vent within the next 8 hours, f b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent.the RCS through at least a 2 square inch vent within 8 hours. I c. In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence. d. The provisions of Specification 3.0.4 are not applicable. I 1 l 1 f f BYRON - UNITS 1 & 2 3/4 4-39 AMENDMENT NO. 37 I I-
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.i -= ,,.1 900 t_ =. ~. _...,.. _ Selected Points on the Curve:.. - T PS RTD MAX. 70 535 i i -~ 97 535 ... a.__. s00 147-550 2 227 580 T---- g 277 800 i: 327 800 1 1 1 i
- 100-
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==g e _ m==== Es s= a======= s= ======= = = = === = - r r E=E= E=== E=== r r . = = 50 100 200 300 400 TRTD.LOWESTCOMSRfDTEMPERATURE(CESF) h FIGURE 3.4-4b NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM (UNIT 2) BYRON - UNITS 1 & 2 3/4 4-40b AMENDMENT NO. 37 o
3 REACTOR COOLANT SYSTEM-BASES PRESSURE / TEMPERATURE LIMITS (Continued) Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 32 effective full power years for Unit 1 (16 effective full power years for Unit 2) of service life. The 32 EFPY for Unit 1 (16 EFPY.for Unit 2) service life period is chosen such that the limiting RT at the 1/4T location in the NDT core region is greater than the RT f the limiting unirradiated material. NDT The selection of such a limiting RT assures that all components in the NDT Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to' determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, NDT. based upon the fluence, copper content and nickel content of the material l in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by either Regulatory Guide 1.99, Revision 2, " Radiation NDT Embrittlement of Reactor Vessel Materials"_ or the Westinghouse Copper Trend Curves shown in-Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shif t in RTNDT at the end of 32 EFPY for Unit 1 (16 EFPY for Unit 2) as well as adjust-ments for possible errors in the pressure and temperature sensing instruments. Revised heatup and cooldown curves have been generated for Unit 2 in accordance with Regulatory Guide 1.99 Revision 2. For Unit 1, the curves remain the same. However, the applicability date has been reduced per RG 1.99 Revision 2 to 29.5 EFPY for heatup. The Byron Unit 1 applicability date of 32 EFPY for cooldown remains the same. Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from-the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ART NDT NDT for the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975. BYRON - UNITS 1 & 2 B 3/4 4-8 AMENDMENT N0. 37
= C .bf . TABLE B 3/4.4-la -w
- =
.J,. REACTOR VESSEL TOUGHNESS (UNIT 1) c 5' Average Upper- .d-Shelf Energy. Normal to-50 ft-lb Principal Principal-35 mil Working-Working-w . T RT Cu P NOT Temp. NDT Direction Direction COMPONENT Heat No. Grade (%) (%) ( F) (F*) (*F): (ft-lb) (ft-Ib) Closure Head Dome C3486-1 A533B CL1 .10 016 -10 < 40 -10 -151 Closure Head Ring IV4566 A508 CL2 .11 007 20 < 80 20 125 Closure Head Flange 124K358VA1 .011 ' 60- <100 60 145 Vessel Flange 123J219VA1 012 10 < 70 10 152 Inlet Nozzle IV4684/3V1320 .12 .008 -10 < 40 -10 117 g" 1V4684/3V1320 .12 008 -20 < 40 20 116 1V4695 .13 . 007 -20 < 10- -20 116 I~ " IV4695 .12 .006 -20 < 10 -20 119 Outlet Nozzle IV4656 .11 .007 0- < 10 0 131 1V4656 .11 007 -20 < 10 -20 131 2V2557 .11 007 -20 < 10 '-20 112 2V2557 .11 ' 008 -10 50 -10 94 Nozzle Shell 123J218 .05 010 20 < 70 20 138 -184 Upper Shell** SP-5933 .05 010(.73) 40 <100 40. 139-156~ Lower Shell** SP-5951 .04 014(.64). 10 < 70-10 150 160 g g Bottom Head Ring IV4672 .012 0 < 60 0 115 gj Bottom Head Dome .C2815-1 A533B CL1 .19 009 -30 40 ~ 20 ~118 - - ~ 5 Upper to Lower WF336 .024-010(.70)- -30 30 -30
- 77*
- -- - l. g Shell Girth Weld ** RI CNormal to Principal Welding Direction RCalculations per Regulatory Guide 1.99 Revision 2.use nickel content shown in parenthesis. l
- 'Fr*'
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- Principal e
35 mil Working l Working T RT w Cu P NOT Temp. NOT' Direction ~ Direction-Component Heat No. Grade (%) (%) .( F) j(F* (*F) (ft-lb) (ft-lb); Closure Head Dome C4375-2 A533 B, C1.1.12 .013 -40 <20 -40 114-Closure Head Ring 48C1300-1-1 A508 C1.3 .05 .007- -30 <30 -30 108 - - - ~ ~ Closure Head Flange 2029-V-1 A508 C1.2 .011- .O <60 0 157 Vessel Flange 124L556 val' .008 30 <90 30 129 Inlet Nozzle 51-2979 .07 .010 -10 <50 -10 130 [" 51-2979 . 07. .009 -20 <40 -20 -121. 2 42-5105 .07 .008 0 <60 0 122 [" 42-5105 .07 .011 0 <60 0 121 "' Outlet Nozzle 11-5052 .09 .010 -10 <50 -10 108 11-5052 .08 .007 -10 <50 -10 121 4-2953 .09 .010 -20. ' 40 -20 133 4-2956 .09 .009 -10 <50 121 Nozzle Shell 4P-6107 .014 '10 <70 10 155 Upper Shell* 490329/ A508 C1.3 .01 .007(.70)- -20 <40 -20 149 149' l' 1-1 49C297 g Lower Shell* 49D330/ A508 C1.3 .05 .008(.73) -20 <40 -20 127 159 -l. 1-1 @ Bottom Head Ring 49C298 g 4801566 1-1 A508 C1.3 .07 .007 -30 <30 -30 126 '5 Battom Head Dome C3053-1 A533B, C1.1 .06 .004 -30 ~40 -20 121 5 Upper Shell to Lower WF447 SAW .059 .009(.62) 10 <70 10 80 l ( Shell Girth Weld * " Weld HAZ -60 <0 -60 143
- Calcelations per Regulatory Guide'1.99 Revision 2 use nickel content'shown in parenthesis.
1 /.; REACTOR COOLANT SYSTEM 1 BASES PRESSURE / TEMPERATURE LIMITS (Continued) A notch in the cooldown curve of Figure 3.4-3 may be present due to the l added constraint on the vessel closure flange given in Appendix G of 10 CFR 50. This constraint requires that, at pressures greater than 20% of the preservice system hydrostatic test pressure, the flange regions that are highly stressed by the bolt preload must exceed the RT f the material by at least 120 F. NDT The flange RTNDT + 120 F impinges on the cooldown curves and therefore the notch is required. If no notch is present, this indicates that the vessel closure flange region has been determined not to be limiting. HEATUP 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 for the 1/4T crack IR during heatup is lower than the K for the 1/4T crack during steady-state IR conditions-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 different K 's for steady-state and finite heatup rates IR 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. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and 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 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface dud ng heatup produce ~ stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest 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 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 under consideration. BYRON - UNITS 1 & 2 8 3/4 4-15 AMENDMENT N0.37 r ,r
.s - ,c ,g i REACTOR COOLANT SYSTEM . BASES-PRESSURE / TEMPERATURE LIMITS (Continued) The use of the composite curve is necessary to set conservative heatup limitations because it-is possible for conditions to exist such that over the 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. Finally, 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. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350 F. Either PORV has adequate relieving capability to protect the RCS from-overpres-surization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or ' equal to 50 F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water solid RCS. These two scenarios are analyzed to determine the resulting overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from full closed to full open. Figure 3.4-4 is based upon this analysis and represents the maximum allowable PORV variable setpoint such that, for the two overpres-surization transients noted, the resulting pressure will not exceed the Appendix G reactor vessel NDT limits (nominal 10 effective full power years l for Unit 1 only). I RHR RCS suction isolation valves 8701A and 8702A are interlocked with an "A" train wide range pressure transmitter and valves 8701B and 8702B are inter-locked with a "B" train wide range pressure transmitter. Removing power from valves 8701B and 8702A, prevents a single failure from inadvertently isolating both RHR suction relief valves while maintaining RHR isolation capability for both RHR flow paths. 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). BYRON - UNITS 1 & 2 B 3/4 4-16 AMENDMENT N0.37}}