ML20024G262
| ML20024G262 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/13/1973 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G258 | List: |
| References | |
| NUDOCS 9102080441 | |
| Download: ML20024G262 (29) | |
Text
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.stnaisnart im*cepve yllamon gnirud re noitarepo lamron gnirud SCCE eht fo noitautca tneverp ot dehcilutse nigrac eht c uder dluow ti tub ,noitcerid ef as a ni eb dluow tniop ter noitaitini SCCE eht esiar oT .noiretirc rieht gniteem norf stnenopecc SCCE eht tneverp itme SCCE eht f c noitaitini rer tniop tca eht rewol oT .tniop tes noitaitini SCCE eht dna
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- deunitnoC secap 3 .
\ s SAFETY 1,I? FITS LU*ITING OAFETY SYSTD; SETTINGS 2.0 R"AC'ICR WCIANT 9YSTET* 2.? 4 RGCTOR CDOIMrr SYSTE'4 P.2 i
Applicability: Applicability: Applies to limits on roacter coolan'. system Applies to t rip settings of the instruments and devicas which are provided to prevent the press ve. reactor sys'a9 safety limi*.s from beine ex-ceedad. Objective: Objective: To define the level of the process variables To es tablish e. limit t elow wnich th<- i nte.:;r ity of the reactor coolant system is not threate".ed at which automatic protective action is due to an overpressure condition. initiated to prevent the safety limits fro . l being exceeded. l Speci fica tion: Specification:
'Ihe reactor vessel pressure shall not exceed A. Reactor Coolant High Pressure Scram shall 1335 psi 8 at any time when irradiated fuel is be $ 1075 psig.
present in the raactor vessel Reacter Coolant Syste= Safety / Relief Valvas F. Initiation shall tw as follows: h valves at 61080 psig. C. Peactor Coolant System Safety Valves Noninal Settings shall be as follows: h V*tlves at 512h0 psig. I 23 2.2/2.l* FE7
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2.2 The reactor coolant' system integrity is an important barrier in the prevention of uncontrolled f' release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is
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[ irradiated fuel in the reactor vessel. l The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 l psig at the lowest elevation of the reactor coolant system. The 1375 psig value was derived from j the design pressures of the reactor pressure vessel, coolant piping, and recirculation pump casing. The respective design pressures are 1250 psig at 575'F.1148 psig at 562*F, and 1400 psig at 575*F. ' The pressure safety limit was chosen as the lower of the pressure transients permitted by the l l applicable design codes: ASME Boiler and Pressure Vessel Code Section III-A for the pressure vessel, + ASME Boiler and Pressure Vessel Code Section III-C for the recirculation pump casing, and the USAS l Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure i transients up to 10 percent over the vessel design pressure (110% x 1250 = 1375 psig) and the USAS ; Ccde permits pressure transients up to 20 percent over the piping design pressure (120% x 1148 = t i 1378 psig). , !- l ! The design basis for the reactor pressure vessel makes evident the substantial margin of protection ~! against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a , general membrane stress no greater than ~ 26.700 psi at an internal pressure ci 1250 psig and temper- l ature of 575'F; this is more than a factor of 1.5 below the yield strength of 42,300 psi at this ! temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29.400 [ psi, still safely below the yield strength. ! 0 The reactor coolant system piping provides a comparable margin of protection at the established { , pressure safety limit. l t 2.2 BASC 2b RE7 [, 4 -a-- -m & -
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Bases: 2.4 Tha settings on the reactor high presstae scram, reactor coolant system safety / relief valves, turbira control valve fast closure scran, and tarbine stop valve closure scram have been established to ass ure never reaching the reactor coolant system pressure safety limit as well as assurir.g the sys-ten pressure does not exceed the rnnre of the fuel claddint; integrity safety limit. The AIBM neutron flux scran and the turbine bypass system also provide protection for these safety limits. In nddition to preventing power operation above 1075 psig, the pressure scram backs up the AI?M neutron flux scram for steam line isolation typo transients. The reactor coolant system safety valves of fer yet anotbei protective feature for the reactor coolant system pressure safety Ilmit. la compliance with Section III of the ASME noIIer and pressere Vessel Code, l'365 editioq, the safety valves nust he set t o open at a pressure no higlier (han 105 percent of design pressure, and they nest ]Init the reactor pressure to no rore than 110 percent of design pressure. The saf ety valves are sized according to the code for a condition of MSIV closure while operating at 1670 IMt. followed by no MSIV closure scram but scran fron an indirect (high flux) means. With the safety valves set as specified herein, the maximum vessel pressure (at the botton of the pressure vessel) would be about 1306 psig. See FSAR Section 4.4.3 and supplemental information submitted February 13, 1973. Evaluations presented indicate that a total of eight valses ( h safety valves and h dual purpose safety / relief valves) set at the specified presstires maintain the peak pressure during the transient within the code of allowable and safety limit pressure. The operator will set the reactor coolant high pressure scran trip setting at 1075 psig or lower. However, the actual setpoint can be as much as 10 psi above t he 1075 psig indicated set point due - to the deviations discussed in the basis of Specification 2 3 on pace 18. In a like nanner, the l operator will set the reactor coolant system safety / relief valve initiation trip setting at 1080 _ psig or lover. However, the actual set point can be as much as 11 psi above the 1080 psig indicated l set point due to the deviations discussed in the basis of Specification 2 3 on page 18s A violation of this specification is ass ~uned to occur only when a device is knowingly set outside of the limiting trip setting, or when a sufficient number of devices have been affected by any means 2.4 BASES 26 REV
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e' Bases Continued 3.3 and 4.3: consequences of reactivity accidents are functions of the initial neutron flux. The require-ment of at least 3 counts per second assures that any tran:.ient, should it occur, begins at ' or above the initial value of 10% of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A mirimum of two operable SPJi's are provided as an added conservatism.
- 5. The consequences of a rod block monitor f ailure have been evaluated and reported in the Dresden 11 SAR Amendments 17 and 19. These evaluations, equally applicable to Monticello, show that during reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with MCHFR's less than 1.0.
During use of such patterns, it is judged that test h g of the RBM system prior to withdrawal of such rods to assure its operability will assure that imprcper withdrawal does not occur. It is the responsibility of the Engineer, Nuclear, to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to i the occurrence of inoperable rods in other than limiting patterna. C. Scram Insertion Times The control rod system is designed to bring the reactor saberitical at a rate fast enour,b to prevent fuel dama6e; i.e., to pmvent the MCHFR from becoming lens than 1.0. This mquires the ne6ative , mactivity insertion in any local region of the core and in the over-all core to be equivalent to at I least one dollar within 0.75 second. The required average senun timer for three control rods in all - two by two urrays and the required average scram tin 2s for all control rods art based on inserting this runount of negative reactivity locally and in the overall core, respectively, within 0.75 second. Under these conditions, the thermal limits are never reached durin6 the transiente requirin6 control rod scram as presented in the E AR. 'ibe limiting operational transiert is that resulting from a turbire stop valve closure with failure of the turbine bypara cyrten. Analysis of this transient shows that the , negative reactivity rates resulting from the scram with the avera6e response of all the drives as given . in the abow Specification, provide the requimd protection, and MCHFR Mnn greater than 1.8. In I the analytical treatment of the transients, 290 milliseccuds are allovmi between a neutmn sensor reaching the scran point and the start of motion of the evntrol rods. 3 3/h.3 CASm 85 RE7 l
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30 LIMITIIIG COITDITIOIC FOR Ol'ERATICII h.O CURVEIIJ/WCE RRUIRDIEITIG tion of four safety /relier valves shall every two refueling outages. The n>minal be eperable. 'Ihe solenoid activated pcpping point of the four safety vnives i relief function of the safety / relief val- 'sh^ll be set at i 12hO psig. ves :: ball be operable ac required by Spa c-ification 3 5.E. i
- 2. If specification 3 6.E.1 is not n:et, Ini- 2. a . A minimum of two safety / relief valvos shall tinte an crderly shutdevn and have coolant te banch checked or replaced with a bench prescure and temerature reduced to 119 checked valve each yefueling cutace. All psig or less and 345 F or less within Ph four valves shall be checked or replaced hours. evary two refueling outages. The popping point of the safety / relief valves shall be set as follows:
fiumber of Valves Set Point (psig) h _< 1060
- b. At least one of the s=fety/relier v,1ves shall be disasse= bled and inspacted each i refueling outage.
- c. The inte6rity of the safety / relief valve bellows shall be continuously moni*ored.
d. The operatility of the believs monitoring i 3 6/h.6 119 REV
_7 Bases Continued 3.6 and 4.6: D. Coolant Leakage ne forner 15 gpm limit for Icaks from unidentified sources was established assuming such leakage was comirtg frem the prima g systen. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the ptMability that the crack will propagate. From the crack size a leakage rate can be detemined. For a crack size which gives a leakage of 5 gpm, the probability of rapid propagation is less than 10-5 Eus, nn unidentified Icak of 5 gpm when assumed to be from the primary system had less than one chance in 100,000 of propa-gating, which provides adequate margin. A leakage of 5 gpm is detectable and measurable. Le 24 hour period allowed for determination of leakage is also based on the low pro *> ability of the crack propagating. The capacity of the drywell sump pumps is 100 gpm and the capacity of the drywell equipment drain tank pumps is also 100 gpm. Removal of 25 gpm from either of these numps can be accomplished with considerable margin. The perfomance of the reactor coolant leakage detection system, including an evaluation of the speed and sensi-tivity of detection, will be evaluated during the first 18 months of plant operating, and the conclusions of this evaluation will be reported to the AEC. Modifications, if required, will be perfomed during the first refueling outage af ter AEC review. In addition, other techniques for detecting Icaks and the applicability of these techniques to the Monticello Plant will be the subject of continued study. E. Safety and Relief Valven Experience in safety valve operation shows that a testing of 5'J7. of the safety valves per refueling outage is adequate to detect failures or deterioration. A tolerance value is specified in Section III of the ASME Boiler and l Pressure Vessel Code as +17. of the set pressure. An analysis has been performed which shows that with all safety valves set 17. higher than the set- pressure, the reactor coolant pressure safety limit of 1375 peig is not exceeded. j Safety / relief valves are used to minimize activation of the safety valves. The operator will set the pressure settings at or below the settings listed. However, the actual setpoints can van as listed in the basis of Specification 2.4. The required safety valve steam flow capacity is detemined by analyzing the pressure rise accompanying the main steam flow stoppage resulting from a MSIV closure with the reactor at 1670 MWt. Le analysis assumes no MSIV closure scram, but a reactor scram from indirect means (high flux). The relief and safety valve capacity is assumed to total 83.97. (477. relief and 36.97 safety) of the full power steam generator rate. H is capacity corresponds to assuming that four safety / relief valves (477,) and four safety valves (36.97.) operated. 3.6/4.6 BASES 134 REV
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- 1. INTRODUCTION Cv Analysis of the recent change to the exposed core scram reactivity U
insertion curve (GE December 1972 curve, curve C, Fig.1) has resulted in the inability of Mon ticello to satisfy, near the end of cycle, the GE recommended 25 psi margin between the " worst case" pressurization type t ransient (tu rbine trip without bypas s, i. e. , relief valve si, ing transient) and the setpoint of the first spring safety valve (1210 psig). Calculations performed previously have shown that the current scram reactivity insertion curve (GE Generic '72, curve B, Fig. 1) will remain valid until a core exposure (R-1 reload) of 2400 MWD /T. The tr ansient analyser based on this curve and submitted to the AEC cin February 13 197 3, adequately describes the effectr of the transients. Beyond an exposure of 2400 MWD /T, additional measures are required to ensure the margins of the February 13, 1973 analysis are met. Of the short range remedies studied, a reduction in reactor power to 84% will adequately meet the requirement. A second alternative, increasing the spring safety valve setpoints, could also assist in maintaining the transient ma r gin s . Liis report is intended to provide the analytical and administrative iustifications for the safety valve setpoint change.
- 11. CONCLUSION AND RECOMMENDATIONS The setpoints of all four spring safety valves can and shouid be reset to 1240 prig prior to a core exposure (R-1 reload) of 2400 MWD /T. This
,4 f $ h pdf setpoint inc rease can be effected with'.n appropriate codes and regulations, fM2 and maintaint the required and recommended margins described in earlier
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l l With the four spring safety valves set at 1240 psig and the four combina- ! I tion relief / safety valves set at 1080 psig (present setpoint), the reactor l can be operated at 91% power from which the margin between peak pressure resulting from the relief valve sizing transient and the first safety valve will be 26 psi. This satisfies the GE recommended design j guideline margin (23 psi). I l The safety valve sizing analysis assumed a power level of 100% and
. yields a margin of 67 psi from the 1375 psig limit; operation at 91%
(limited by the RV sizing transient) would result in larger margins. In conjunction with the setpoint change, Tech Specs must also be modified to include the settings and their bases. III. DISCUSSION A. Has.is for Change On February 13, 1973, NSP submitted to the AEC a report prepared l by General Electric entitled "Results of Transient Reanalysis for Monticello Nuclear Generating Plant with End-of-Cycle Core l Dynamic Characteristic s. " This report described the changes to the abnormal operational transient analysis as described in the FSAR caused by a shift in the scram reactivity feedback curve for exposed core conditions. Also included in that report were the proposed changes necessary to meet the General Electric guideline which is to maintain a margin of 25 psi between the peak pressure resulting from the " worst case" pressurization type trnnsient (turbine trip without bvpass, i. c. , relief valve sizing transient) and the setpoint of the first spring safety valve. On June 1,1973, NSP submitted a proposed change to the Technical Specifications incorporating the results and recommendations of the February 13 report.
3 Gene ral Elec t ric, on the ba sis of refiner' analytic al techniques, informed 1:SP earlier this year that the shift in the neram reactivity curve is a function of core exposure, i. e. , the present curvt (l'i g . 1, GlW gene t ic "72, Cu rve B) does not fully represent the final end-of-cycle core condition. GE has determined that Curve 14 will define the sc ram reactivity function to a core exposure (11-1 reload) of 2400 MWD / T. l'outulated transients using the assumptions from the .nalyses and occurring up to that exposure would not result in peak pressures in excess of that des cribed in previous submittals. lleyond an exposure of 2400 MWD /T, additional measures are required to ensure satisfaction of the GE reconunended design objective of a 25 psi margin between peak transient pressure and the safety valve setpoints. The margin between peak pressure and the reactor vessel pressure limit following the safety valve sizing transient (MSIV closure) remains well above the 25 poi required design margin; therefore, no safety limit is affected whether or not additional measures are taken. General Electric has determined an "all rods out" scram reactivity curve (Fig. 1, Curve C) to define the wore,t case core conditions between 2400 MWD /T and the end of the current cycle for Monticello. Although actual conditions do not reach Curve C until the all-rods-out end-of-cycle exposure, Curve C was applied to determine what measures were necessary to ensure maintenance of the effects of 12 ansient s throuphout the remainder of the cyc1 ? In conformance with tla earlie r analysis. E va l ua t ion t h c4 b< n n ma de fo r a rml n.overne nt "f rec 7 e" until power coasts down to 847 at 2400 MWD /T: this is sufficient to meet the 25 psi snargin in the relief s ve sizing t ransient. This operational restriction maintains the validity of the earlier transient a naly.< c u. Ot he r inea s urc e; such as the change discussed in this I
l 4 report, a re being developed; their application may aid further in mitirating the ocerall effects of the newer curves. B. Assumptions Used in Analysis The same generic assumptions as those used in the February 1973 submittal were applied to the transients described in this report. Conservative assumptions, such as a multiplier on the void coefficient, a multiplier on the scram reactivity curve, and average control rod scra m times equivalent to the '67 Product Line BWR, were used. Because the nev, scram reactivity curve (Curve C) represents an exposed core condition, the new analyses were done with end-of-cycle inputs for consistency and to ensure that a realistic worst case would he defined. For example, the void coefficient is reduced at the end of the cycle and this will tend to reduce the peak of the pressurization transients. The scram reactivity curve (Curve C) is a function of core exposure and will not be approached until near the end of cycle; however the curve is assumed to q;1y from 2400 hBVD/T to the cycle end for the determination of 100% power transient effects. In the analyses of this report, the four cornbination relief / safety valves are assumed operable as described in the February 1973 report. The four spring safety valves are also assumed operable with a setpoint of 1240 psig. Because a complete set of transient analyses is not required, only the tra nsients of most cone. rn were redone. These were the turbine trip without bypass transient for checking relief valve adequacy and the MSLlY closure with indirect scram for checking safety valve a dequa cy.
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.. C. Transients Not Reanalyzed I l
The FSAR incaded about 20 analyses of worst case abnormal transic..ts in six categories of events. These categories a re prima ry syst . m essu re ir c rer. sc s, made rator temperature decreases, ; reacta . h ins e, tions., . ore coolant. anventory decreases, core coolant ) p r/ is .rr .: ( md cort .oolar nr d. creases. These were all redewed te del.cr sne those i.ith adght be significantly affected by the new end- / ,yclt cere. charar eristic s assumptions. The break. downer ca gories, evertr. , and legic for those in which only a review wss deer- 'o be adequc.te, is described in the analysis submitted in Fcbruary 1973. Reite ' tion of that discussion is unnecessary here.
^
D. Results of Transients Reanalyses l
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- 1. Scope of Reanalyser The following t ransients were reanalyr.ed in order to determine the specific changes that might occur to the previous analytical results:
Turbine trip without bypass (Relief valve adequacy check) 1 Main Steam Isolation Valve Closure, (includes delayed scram case for safety valve adequacy check) Specific write-ups for these analyses are included herein. The original FSAR analysis used the turbine trip without bypass with flux scram for the safety valve sizing transient. Ilowever, analyses of later plants revealed that the main steam line 4 isolation with flux scram could be more severe. Daring the reanalyses work reported in February 1973, this possibility was checked by performing both analyt es and the result:: showed a somewhat higher peak pressure with main steam isolation valve closu re. This analysis is used for checking safety valve adequacy in this report as well.
. . _ . _ . . _ _ _ __ .._.. _ _ _ . _ . ,. __ _ _ _ ___ __ . ~ . _ _ . , _ .. ._
- . - (, .
- 2. Turbine Trip Without Bypass - Relief Valve Adequaev T ra n s ie rit Reactor operating at 91% of rated, core flow 100%, 67
> ptu duct line scram times (data interpolated from several cases run from 85 to 100% power):
I A scram signal is initiated at the same time a t $1ne trip occurs by position switches on the turbine stop valves. This transient causes a rapid pressure increase in the reactor pressure vessel. Primary system relief valves are provided in remove se'ficient energy from the reactor to prevent safety valves from lif ting. R ea nalys is . showed that peak pressure in the steam line at the safety valve location did not meet the GE margin of 25 psi to the first safety valve setpoint (1210 peig). 110 wever, a i peak pressure in the steam line of 1214 psig at the safety valve location provides an adequate margin of 26 psi to 1240 psig, the recommended first safety valve setpoint. Thus, the adequacy of the four relief valves was confirmed f ar these conditions. Using the parameters associated with the end of life conditiont,, four relief / safety valver, a re required to operate as described in the February 1973 report to prevent this pressure transient from exceeding the safety valve setpoint. The rapid pressure rise due to
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i i l 4 1 rapid closure (O.10 sec. ) of the turbine stop valve . l I without bypass operation causes core volds to ; I collapse and neutron flux peaks at 262 percent of l des ign l'n 0. 9 2 s ec. (Figure 3) before the scram shuts down the reactor. Peak surface heat flux is 100,3% at 1. 36 sec. MCIIFR and other pertinent parameters remain within acceptable limits.
- 3. Clos.ure of All Main Steam Line Isolation Valves (Flux Scram) - Safety Valve Adequacy Transient The ASME Nuclear Boller and Pressure Vessel Code requires that, each vesse' designed to meet Section III-be protected from the consequence of pressure and temperature in excess of design conditions. The ASA Code for Pressure Piping also requires overpressure protection. The setpoints of the safety valves comply with the ASME pressure vessel code taking into accoun' static heada and dynamic 1 sses.
Thb was discussed at length in the February 1973 report and is included here for completeness. The change in safety valve setpoints described in this i ! report will meet appropriate codes and regulations l (ASMC NB & PV Code Sect, XI), i
- - . .-_ - .- - . - - ..-. ~- .. . _ _ -
1
. 8-l To determine the required flow capacity of the safety valves, it l i
is assumed that: I
- a. The reactor is at 1670 MWt, l
l
- b. The reactor experiences its worst main steam isolation j transient,
- c. Direct reactor scram is neglected (based on isolation valve position switches),
- d. The backup scram due to high neutron flux shuts down the reactor,
- c. The Target Rock relief valves act as safety valves with low serpoints.
1 I
.I30th a turbine trip without bypass and closure of all main stearn j
'line isolation valves produce severe overpressure transients.
Analyses for these two events have shown that the 3 second l closure of the isolation valves is slightly more severe for'ihe final plant configuration when direct reactor scram is neglected. This results because the longer steam lines, allowing more volume for steam compression, more than compensates for the l faster acting turbine stop valves in the former transient when i i compared with MSLIV closure. The latter transient is therefore provided here as the basis for determining the adequacy of the safety valves. l , j Pressure increases follow this reactor isolation until limited by j the opening of the safety valves. The peak allowable pressure is. 1375 psig (according to ASME Section III, equal to 110 perc ent l of the vessel design pressure of 1250 psig). The Target F.ock l l setpoints are < 1080 psig and the spring safety valve setpoints are at 1240 psig (4 valves). Thus the ASME code specifications that the lowest safety valve be set at or below vessel design
- - - - -- --- ._ __ . .-.~.- - .. . - - _ - . - - -
. I 9 .
pressure, and the highest safety valve be set to open at or below 105 percent of vessel design pressure are satisfied. The four spring valves together have a capacity of greater than 35 percent of turbine design flow. 1 The resulting transient assuming the capacity of the 1 safety / l relief valves (47% of main steam generation rate) and the 4 safety val.ves (36. 9% of main steam generation rate) is shown I in Figure 4. An abrupt pressure and power rise occurs as soon as the isolation becomes effective. Neutron flux causes the scram at approximately 1.8 seconds thereby initiating reactor shutdown. Flux peaks at a value of 644 percent in 2.14 sec. The assumed safety valve capacity (Target Rock plus. ! spring safety capacities) keeps the peak vessel pressure 67 psi below the peak allowable ATME overpressure of 1375 psig.
- Therefore, the relief valves plus the spring safety valves provide adequate protection against excessive overpressuriza-i tion of the nuclear system process barrier with an adequate ma rgin.
IV TECHNICA L SPECIFICATION CHANGES' A. Seopec of Chays The principle changes of interest concern the safety valve setpoints. Thin. is needed to be consistent with the new assumptions used in the transient reanalyses and is dircussed in detail in Section III. D. Other changes are those associated with the results of the transient reanalysco discusced in Section 111. D. None of these are of a crucial safety nature and mostly affect statements about margins for various ! pressuritation transients. Tech Spec changes submitted to the AEC June 1, 197 3, are conside red to be in effect; c rrors or omissions I related to the February 1973 report and June 1, 197 3 submittal have l I been corrected or added. i
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B. Specific Changes ,
- Ite m Location Change Reason Bases statement for 2.1 Pg. 16 Add reference to the Indicates documentation of I'ebruary 1973 submittal ' discussions on this topic.
- and this analysis Bases statement for Pg. 20 - end Change ". . . Pg. 22. " Corrects typograohical error
- 2. 3. A of third para. to "... Pg. 18. "
Dases etatement for Pg. 21 - end - Change ". . . Pg. 22. " Corrects typographical error { Z. 3. B of second para. to .' . . Pg. 18. " !
- Dases statement for Pg. 21 - end Change ". . . Pg. 22. " Corrects typngraphical error. '
! 2. 3. C of second para. t o ". . . Pg. 18. " i i , Tech. Spec. 2.4. C Pg. 23 Change "2 valves at 3 1210 This cht.nge reflects an ; psig. " and "2 valves at 3 assumption of this analysis. c 1220 psig" to "4 valves at l 5 1240 psig. " J Bases statement for 2. 2 Pg. 24 last p: ra. Change to read as follows: This change provides the basi Pg. 25 top of pg. "The normal operating pres- for the valve configurati { sure of the reactor coolant used in this analysis . sf stem is approximately. ! l 025 psig. ' The turbine
; trip from 91% power with l failure of the bypass system j represents the most severe j primary system pressure increase resulting from an abnormal operational transient.
The peak pressure in this transient is 1214 psig.
f Location Change Reason Item In addition, the safety valves are sized on the basis of a closure of all hiain Steam Isolation Valves (MSIV Closure) where erram is assumed to be indirect (high flu.:) rather 1 ban from the MSIV position
'ci tc he s . In this transient, amuming rated power, the pro s < u re at the bottom of the vessel is 1308 psig.
Reactor pressure is continuously monitored in the control room 6 aring operation on a 1500 psig full-scale pres sure recorder. - 9 Pg. 26, Para 2, Change 1283 psig to 1308psig This change reflects the Besis statement for 2.4 results of this analysis. Line 9 Pg. 26, Pa ra 2, Change to read: This change reflects an Line 10 " . . . a total of eight valves assumption of this analysis. (4 safety valves and 4 dual purpose safety relief valves) s et at. . . " Pg. 26, Para 3, Change ". . . Page 22" to Corrects typographical error Lines 3 and 6 ". . . Pa ge 18. "
i 1 Location Change Reason Item Change to read: Restores originnl statement j Basis statement for Pg. 85,
- 3. 3. C /4. 3. C Lines 8 and 9 "The limiting power transient erroneously charged in !
is that resulting from a February 1973 rebmittal. l turbine stop valve closure l with failure of the turbine ) l bypass system. " Tech. Spec. 3.6.E,1 Page 118, Change ". . . three safety This change reficcts an Last Line valve s . . . " to ". . . fou r assumption in this analyses safety valves. . . " Tech Spec 4. 6. E.1 Pn. 119 Change to read as follows: This change refl~ cts an
". . . every two refueling assumption in t1 - s analysis outages. The nominal pop-ping point of the four safety valves shall be set at <.1240 psig. " ,
N Change RV/SV capacity Pg. 134, These changes r-flect ' I3ases statement for " ...to total assumptions in this analysis 3.6.E/4.6.E Last Para. , as follows: Line 4 83. 9% (47% relief and
- 36. 9% safety) of . . . "
Line 5 ". . . assuming that four relief safety valves (47%) and four safety valves (36. 9%) operated. " Lines 6 and 7 Delete entire last sentence. g -
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AEC DISTRIBUTION FOR PART 50 DOCKET MATFRTAL Y (TDiPORARY F0141) I CONTROL NO: 7190 FILE: DATE OP DM DATE REC'D LTR l'!2:0 RPT OTHER TRCN: bort hern Sta tes f ou r Company Minneapolis, Minnesota $5401 L. O. Mayer 4-13-73 9-26-73 X TO: ORIG CC OTHER SENT AEC PDR X Mr. O' Leary 3 signed CLASS VliCLASc PROP IhTO INPUT NO CYS REC'D DOCKET NO: XX XXX 40 50-263 DESCRIPTION: ENCLOSUPES: Ltr trans the following: Request for Authorization to Change Appendix A of the Tech Specs, notarized 9- 13-7 3. . . . . . .W/ A;. tache d Rp t , "Monticello - Safety Valve Setpoint Increase". ( eu cys rec'd) 1 n .. h i w- > - PLANT NAME: Monticello Sg FOR ACTION /INFORMATION 9-26-73 AB BUTLER (L) SClk'ENCER(L) y ZID1 ANN (L) REGAN(E) W/ Copier V/ (. . p f e r W/9 Copics P/ Copics CLAP 1(L) STOL2(L) DICKER (E) W/ Copics W/ Copies W/ Copies W/ Copics GOLLER(L) VASSALLO(L) K!11GitTON(E) W/ Copict W/ Copics W/ Copies W/ Copies 131IEL(L) SCHEMEL(L) YOUNGELOOD(E) V/ Copics W/ Copics W/ Copics W/ Copics
._ - T NT..'? " AL DI STRI BUTI ON REG FILE ' TECH REVIF'A DC; TON A/T IND ./ AtA IN HENDRIE GRIMES LIC ASST BRAIIMAN /0GC, ROOM P-506A SCHROEDER GA!0 1LL /DIGGS (L) SALTZMAN t/ MUNTZING/STAFP liACCARY KASTNER GEARIN (L)
CASE 12;IG3T BALLARD COULEOUPJ'E (L) PLA?S GIA!!BUSSO PAWLICKI SPANGLER LEE (L) MCDONALD BOYD SHA0 MAICRET (L) / DUBE MOORE (L)(174R) STELLO ENVIRO SERVICE (L) DEYOUNG(L)(l"JR) HOUSTON MULLER SHEPPARD (E) INFO
/SKOVHOLT (L) NOVAK DICKER SMITH (L) C. 11T LES P. COLLINS ROSS KNIGHTON TEETS (L) / ALLEN CABELL(Ltr)
IPPOLITO YOUNGr,LOOD WADE (E) REG OPR TEDESCO REGAN WILLIAMS (E) y FILE & REGION (3) LONC PROJECT LDR WILSON (L) LORR1S LAn, i STEELE BEhAROYA HARLESS VOLL5 'S' _ E: .it Pn n inTIOS n [ - LOCAL PDR Minnenpolis,_Minn. Wf
*/1 - DTI E ( A"d'. ;.ill!Y ) (1)(2)'10bMATIONit LAB'S VI - NSIC(E"Onm As) 1- R .f c hoonrak e r ,00,GT,D-323 1-PDR-S 1-GERALD LELLOU AN /LA /IP[CHE 1 - ASLE(YORP%.,c/ 1 - \l , l' E o ' ; I "GT D ., , la L-201 GT E'100KEAVEN iMT LAP C00DAG/"h" ST. 1 -CO:iSULT! UT' S 1- AGMED(WALTER E0 ESTER 6 - CYS ACRS E0L'J11:G 1:lM1ARE/ ELUltr / ACB AnI AN PM- C- 4 2 7- CT l-GERALD ULRI RSO:;. . .ORNL l-RD.. MULLER. P-309 CT
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