ML20100Q201

Amend 140 to License DPR-35,revising Thermal & Pressurization Limit Curves of Figure 3.6.1 & 3.6.2 of Tech Specs

ML20100Q201
Person / Time
Site:	Pilgrim
Issue date:	01/29/1992
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20100Q206	List: ML20100Q201 ML20100Q207
References
NUDOCS 9202060233
Download: ML20100Q201 (21)

Text

{{#Wiki_filter:_ _ _ _ _ m. -. - _ _ _. 8[pp htCg$'o, ) + UNIT E D ST ATE S '? NUCL E AR REGUL ATORY COMMISSION i I WASHINGTON, D. C. 205b6 i o h,, / P,01TCPEDIS0tJCO((A[Y DOCLE1 f40. 50-293 PIlGRlti tillCL E AR P0}lE[,,51[T1p[ AMEt;Dt/EllT TO FAC1LITY M ATitG LICEN$,E. Amendment th.140 License No. DPR-35 1. The toclear Pegulatory Concission (the Commission or the NPC) has found that: A. The application for aniendnent filed by the Bosten Edison Con,pany (the 14rentre) dated Jure 11, Inal erd letters of August 5 and October 4, if 91, requesting bases changes con. ply with the standards and requirenents of the Itor.dc Energy Act of 1954, as amended (the Act), and the Cornission's rules and regulations set forth in 10 CfE Chaptu 1; E. The f acility will operate in conforn.ity v.ith the application, the provisions of the Act, and the rules and regulations of the Commission; There is reescnable assurance: (4) that the activities authorized by this anendn.ent can be conducted without endangering the health and se-(ty of the public, and (ii) that such activities will be conducted it: contplianct with the Concission's regulations set forth in 10 CFR Chaptei 1; D. The issuance of this enendnient will not be inirnical to the comon defense and security or to the health and safety of the public; and E. The issvence of th4 5 enendnent is in accordance w th 10 CFR Part 51 of i the Conmission's regulations and all applicable requiren<ents have been satistied. 2. Accordingly, the license is aniended by changes to the Technical Specifica-tic +t as dr4d'cated do the attachrent to this license amendment, and paragr oph 3.B of f acility Operating License bo. DF h-35 is hertby irrended to read as follows: Techt.ical Spe,cjfj_ cat 4c3 0r,s cord 3 ned in Appendh A, as reviseC The Techriical 5[ecificht4 4 through Ar endrent No.1,- e hereby incorporated in the license. i n accorcm ce v.d tl. t h 114 itcu ste sht 1 citr6 h stie fac'1't) lethri cal Specification:. d 9802060233 920129 PDR ADOCK 05000293 P PDR $)

t i t 1-k 3. This lict:hst. anendrent is ef fective as of its date of issuance and i shall t>e h.pleniented w thiri ?O days, j i FOR THE NUCLEAR REGULAT0fa t,0t71SS10N 4 /4 / '/ . V Waltfr R. Butle. Director Project Directorate 1-3 su k ision of Reactor Projects - 1/11 Offica of Nuclear Reetter Regulation Att achn ent: Changes to the Technical Specif_icathns Date of issuerce: January 29, 1992 i f (. l l

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ATTACPf'E!JT TO LICIta$i Alitt$t'Itil fl0.140 FAtli11Y OPERATit!G I100flSE 110._DPR 35 POCKET 110. 60-?93 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revisert pages are identified by Anendo.ent nurit4r and contain vertical lines indicating the area of change. Remove Insert 40b 40b ) 70 70 1 89A 09A 89C 89C 110 110 117 117 118 110 1T3 123 i 174 124 1141. 124t. 125 125 178 120 128A 128A 128B

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3.1 (LMIS (Cont'd) tigja_11 tim._Line Hio!LEadia tioD High radiation levels in the main stes,line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal backgrounrt. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination, Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors which cause an isolation of the main condenser off-gas line. Egactor Mode Syltth A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status. Ref. Section 7.2.3.9 FSAR. l Manual Scram The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation. Scram Discharae Instrument Volume The control rod drive scram system is designed 50 that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. The two scram discharge volumes accommodate in excess of 39 gallons of wacer each and are at the low points of the scram discharge piping. No credit was taken for these volumes in the design of the dis:harge piping as concerns the amount of water which mus'c be accommodated during a scram. During normal operation the scram discharge volume system is empty; however, should it fill with water, the water discharged to the piping could not be accommodated, which would result in slow scram times or partial control rod insertion. To preclude this occurrence, redundant and diverse level Getection devices in the scram discharge instrument volumes have been provided which will alarm when water level reaches 4.5 gallons, initiate a control rod block at 18 gallons, and scram the reactor when the water level reaches 39 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This ft;nction shuts the reactor down while sufficient volume rema.ijs to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately. A source range monitor (SRM) system is also provided to supply additional neutron level information during start-up but has no scram functions. Ref. Section 7.5.4 FSAR. The APRH's cover the " Refuel" and "Startup/ Hot Standby" modes with the APRM 15% scram, end the power range with the flow W nd ent No. J ?.t. M D

~ 3.2 se U (Cont'd) dent. With the established setting of 7 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference FSAR Section 14.5.1 and Appendix R.3.2.5. l Pressure instrumentation is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 880 psig. In the Refuel and Startup Mode this function is replaced by high reactor water level. This function is provided primarily to provide protection against a pressure regulator malfunction which results in the contrc,1 and/or bypass valves opening. With the trip settings specified, inventory loss is limited so that fuel is not uncovered. The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation resu'ts in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic, and all sensors are required to be operable. Temperature is monitored at three (3) locations with four (4) temperature sensurs at each location. Two (2) sensors at each location are powered by "A" direct current control bus and two (2) by "B" direct current control bus. Each pair of sensors, e.g., "A" or "B", at each location are. physically separated and the tripping of either "A" or "B" bus sensor will actuate HPCI isolation valves. The trip settings of 13001 of design flow for high flow and 200'F or 170*F, depending on sensor location, for high temperature are such that core uncovery is prevented and fission product release is within limits. The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 13001 for high flow and 200*F, 170*F and 150*F, depending on sensor location, for temperature are based on the same cri*eria as the HPCI. The Reactor Hater Cleanup System high flow and temperature instrumentation are arranged similar as that for the HrCI. The trip settings are such that core uncovery is prevented and fission product release is within limits. The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this } fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing is,being performed. An exception to this is when logic functional testing is being performed. Amendant N?. U. D 70

= _ - - -. - -.- -_- ---- Ll_an_d 4 3 fL%U: basis is given is subsection 3.5.2 of the TSAR, and the safety evaluation is given in subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving fcree to rapidly eject a drive housing. Additionally, the support is not required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain subtritical even in the event of complete ejection of the strongest control rod. 3. In the course of performing normal startup and shutdown procedures, the reactor operator follows a pre-specified sequence for the { withdrawal or Insertion of control rods. The specified sequences are characterized by homogeneous, scattered patterns of control rods 1 2 elected for withdrawal or insertion. The maximum control rod worths encountered in these patterns for the initial core load are presented in FSAR Figure R.3-1. These sequences are developed prior to initial operation of the unit to limit the reactivity worths of individual control rods in the core. These control rod sequences, l together with the integral rod velocity limiters which will limit the velecity during free fall to less than five feet per second, limit the potentiel reactivity insertion such that the consequences of a control rod drop accident will not exceed a peak calculated enthalpy of 280 calories / gram generated in the fuel. The design limit of 280 calories / gram is selected for limiting peak enthalples in U02 and is assumed to be the lower threshold at which rapid fuel dispersal and damaging pressure pulses to the primary system might occur. As discussed in FSAR Section 14.5.1.3, the calculated radiological l consequences of a control rod drop accident are well within the guideline values of 10 CFR Part 100. k.mac o t t la 89A I

1 3 and 4.J BSIS1 Above 10% of design power assuming a single operator error, it will not be possible for the maximum rod worth to exceed 0.020 delte K in accordance with Sp0cification 3.3.0.3.b(2). Specification 4.3.B.3 requires a sequence of checks and tests on the RHH to verify its orerability before startup and before reducing power to less than 10% of design power. These checks and tests assure that the actions of the control operator are always monitored and blocked when in error should they lead to a condition which might cause fuel damage during the control rod drop accident. Und6r these specification limits, the maximum energy deposition in the fuel and the number of fuel rods damaged resulting from a control rod drop accident, assuming Technical Specification limits on scram times (Specification 3.3.C) and rod drop velocity (5 feet /second), is established to be below the consequences calculated by the licensee for the hot standby critical case.

Reference:

FSAR Section 14.5.1. Therefore, the assumptions used by the licensee and the NRC in estimating the number of failed fuel rods and fuel damage resulting from the excursion energy generated by the rod drop accident appear conservative within the LCO. p "I.,'f, E.{ fl i g[

ILE S: 3.5.C LLPC1 The limiting conditions for operating the HPCI System are derived from the Station Nuclear Safety Operational Analysis (FSAR Appendix G) and a detailed functional analysis of the HPCI System (FSAR Section 6). The HPCIS is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss-of-coolant which does not result in rapid depressurization of the reactor vessel. The HPCIS permits the reactor to be shut down while maintain sufficient reactor vessel water level inventor) until the vessel is depres,srized. The HPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core Spray System operation maintains core cooling. The capacity of the system is selected to provide this required core cooling. The HPCI pump is designed to pump 4250 gpm at reactor pressures between 1100 and 150 psig. Two sources of water are available. Irltially, demineralized water from the condensate storage tank is used insteat. of injecting water from the suppression pool into the reactor. When the HPCI System begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI System. As the r uctor vessel pressure continues to decrease, the HPCI flow momentarily reached equilibrium with the flow through the break, Continued depressurization causes the break flow to decrease below the HPCI flow and the liquid inventory begins to rise. This type of response is typical of the small breaks. The core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of the HPCI. The analysis in FSAR Appendix G shows that the ADS provides a single failure l proof path for depressurization for postulated transients and accidents. The RCIC is rtquired as an alternate source of makeup to the HPCI only in the case of loss of all offsite A-C power. Considering the HPCI and the ADS plus RCIC as redundant paths, and considering judgments of the reliability of the ADS and RCIC systems, a 7-day allowable repair time is specified. The requirement that HPCI be operable when reactor coolant temperature is greater that 365'F is included in Specification 3.5.C.1 to clarify that HPCI need not be operable during certain testing (e.g., reactor vessel hydro testing at high reactor pressure and low reactor coolant temperattire). 365'F is approximately equal to the saturation steam temperatyre at 150 psig. A endment No, M/ 1/f.140 IM

M115: 3.5.0 RCIC system The RCIC is designed to provide makeup to the nuclear system as part of the planned operation for periods when the normal heat sink is unavailable. The Station Ncclear Safety Operational Analysis. FSAR Appendix G, shows that RCIC l also serves as redundant makeup system on total loss of all offsite power in the event that HPCI is unavailable. In all other postulated accidents and transients, the ADS provides redundancy for the HPCI. Based on this and judgments on the reliability of the HPCI system, an allowable repair time of seven days is specified. The requirement that RCIC be operable when reactor coolant temperature is greater than 365'f is included in Specification 3.5.0.1 to clarify that RCIC r.eed not be operable during certain testing (e.g., reactor vessel hydro i testing at high reactor pressure and low reactor coolant temperature). 365'T is approximately equal to the saturation steam temperature at 150 psig. t knendaent Na, 7;'s, J / y,, 14 0 117 1 ,r-r --~. ,,.y v ~,w,,- -,,---e --e ,,v,,w-.,,w=we~- m---,--,,wn,,,e---e -,w,,---w e v--- m -~--,vor-- r- - - +, en---

ILEIS: 3.5.E Autoratic Deoressurization System (AD5l The limiting conditions for operating the ADS are derived from the Station Nuclear Safety Operational Analysis (FSAR Appendix G) and a detailed functional analysis of the ADS (FSAR Section 6). This specification ensures the operability of the !.D: onder all conditions fc which the automatic or manual depressurization of the t1 clear system is an essential response to station abnormalities. The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low pressure coolant injection (LPCI) and the core spray systems can operate to protect the fuel barrier. Because the Automatic Depressuri7ation System does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS. Performance analysis of the Automatic Depressurization System is considered only with respect to its depressurizing effect in conjunction with LPCI or Core Spray. There are four valves provided and each has a capacity of 800,000 lb/hr at a reactor pressure of 1125 psig. The allowable out of service time for one ADS vsive is determined as seven days because of the redundancy and because of HPCIS operability; therefore, redundant protection for the core with a small break in the nuclear system is still available. The ADS test circuit permits contint.ed surveillance on the operable relief valves to assure that they will be ava'lable if required. Amenckent No.17. Uf,1 0

.L1 HATING CONDITION FOMPERATION WRVEILLANCE R(29JREMENTS 3.6 PRIMA [u' SYSTEH BOUNkMI' 4.6. ERIMARY SYSTEH BOUNDARY Anjhability: ADD 1icability: Applies to the operating status of the Applies to the periodic examination anj rcactor coolant system. testing requirements for the reactor cooling system. Obiettive: Objective: To assure the integrity and safe To determine the condition of the operation of the reactor coolant system reactor coolant system and the operation of the safety devices related to it. Soecification: Satification: A. Thermal and Pressurh ation A. Thermal and Pressurizt11gn Limitations Ljmitations 1. The average rate of reactor 1. During heatups and cooldowns, with coolant temperature change during the reactor vessel temperature less normal heatup or cooldown shall than or equal to 450'F, the not exceed 100'F/hr when averaged temperatures at the following over a one-hour period except !ocations shall be permanently when the vessel temperatures are logned at least every 15 minutes above 450'f. The reactor vessel until the difference between any flange to adjacent reactor vessel two readings at individual shell temperature differential locations taken over a 45 minute shall not exceed 145'f. period is less than 5'F: a. Reactor vessel shell adjacent to reactor vessel flange { b. Reactor vessel shell flange c. Recirculation loops A and B 2. The reactor vessel shall not be 2. Reactor vessel shell temperatures, pressurized for hydrostatic including reactor vessel bottom and/or leakage tests, and head, and reactor coolant pressure critical core operation shall not shall be permanently logged at be conducted unless the reactor least every 15 minutes whenever the vessel temperatures are above shell temperature is below 220'T those defined by toe appropriate and the reactor vessel is not curves on figures 3.6.1, 3.6.2, vented. and 3.6.3. (Linear interpolation between curves is permitted). At Test specimens of the reactor stated pressure, the reactor vessel base, weld and heat affected vessel bottom head may be zone metal subjected to the highest maintained at temperatures below fluence of greater than I Hev those temperatures cor responding neutrons shall be installed in the to the adjacent reactor vessel reactor vessel adjacent to the shell as shown in figures 3.6.1 vessel wall at the core midplane and 3.6.2. level. The specimens and sample program shall conform to the Amendmcnt No. f2, I G 123 l

LIHfTING CONDITIQN FOR OPERATION SqPyD 1 LANCE R[OWIREHENTS s 3.6.A lhtrJpal and Preslutin110D 4.6.A lttermal and PressurJR1120 l Limitations (Cont'd) LlLn11Al10D1 (Cont'd) g In the event this requirement requirements of ASTM E 185-66, is not met, achieve stable Selected neutron flux specimens reactor conditions with shall be removed at the frequency reactor vessel temperature required by Table 4.6.3 and tested above that defined by the to experimentally verify appropriate curve and obtain adjustments to Figures 3.6.1, an engineering evaluation to 3.6.2, and 3.6.3 for predicted NDT determine the appropriate temperature irradiation shifts. course of action to take. 3. The reactor vessel head 3. When the reactor vessel head bolting studs shall not be bolting studs are tensioned and the under tension unless the reactor is in a Cold Condition, the temperature of the vessel reactor vessel shell temperature head flange and t.he head is immediately below the head flange greater than 55'F. shall be permanently recorded. 4. The pump in an idle 4. Prior to and during startup of an recirculation loop shall not idle recirculation loop the ^ be started unless the temperature of the reactor coolant temperatures of the coolant in the operating and idle loop' within the idle and operating shall be pertranently logged. recirculation loops are within 50'F of each other. 5. The reactor recirculation 5. Prior to starting a recirculation pumps shall not be started rump, the reactor coolant unless the coolant temperatures in the dome and in the temperatures between the dome bottom head drain shall be compared and the bottom head drain are and permanently logged. within 145'F. 6. Thermal-Hydraulic Stability Core thermal power shall not exceed 25% of rated thermal power without forced recirculation. B. Coolant Chemistrv B. Cool anLChemi s t rv 1. The teattor coolant system 1. a. A reactor coolant sample shall radioactivity concentration be taken at least every 96 in water shall not exceed 20 hours and analyzed for microcuries of total iodine radioactivity content. per ml of water, b. Isotopic analysis of a reactor j coolant sample shall be made at least once per month. I Amendment No,' 22,140 124

~.-_ TABLE 4.6.3 REACTOR VESSEL HATERIAL SURVEILLANCE PROGRAM HITHDRAHAL SCHEDULE i Effective full Capsule Power Years MVEttr (EFPY) 1 4.17 2 15 (approx.) 3 32 (End of Life) 1 I (It Yak O ! ). f1 140 124A

LIMITING. @ NDITION E9R_9PERATIQtf 50RjllLLANCE REQ 91REMENTS 3.6.B Coolant Chemisity (Cont'd) 4.6.B Coolant _Chemli_in (Cont'd) 2. The reactor coolant water shall 2. During startups and at steaming not exceed the following limits rates less than 100,000 pounds with steaming rates less than per hour, a sample of reactor 100,000 pounds per hour, except coolant shall be taken every as specified in 3.6,0.3: four hours and analyzed for chloride content. Conductivity.. 2 pmho/cm Chloride ion.. 0.1 ppm 3. For reactor startups and for the

3. a.

Hith steaming rates of 100,000 first 24 hours after placing the pounds per hour or greater, a reactor in the power operating reactor coolant sample shall be condition, the following ilmits taken at leatt every 96 hours shall not be exceeded. and analyzed for chloride ion content. Conductivity.. 10 pmho/cm Chloride ion.. 0.1 ppm b. When all continuous conductivity monitors are 4. Except as specified in 3.6.B.3 inoperable, a reactor coolant above, the reactor coolant water sample shall be taken at least shall not exceed the following daily and analyzed for limits when operating with conductivity and chloride ion steaming rates greater than or content. equal to 100,000 pounds per hour. Conductivity.. 10 pmho/cm Chloride ion.. 1.0 ppm 5. If Specification 3.6.0 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in Hot Shutdown within 24 hrs, and Cold Shutdown within the next 8 hours. 3.6.C Coolant leakaat 4.6.C Coolant LeakM.g Any time irradiated fuci is in Any time irradiated fuel is in tha reactor vessel and coolant the reactor vessel and coolant temperature is above 212"f, temperature is above 212'f, the the following limits shall be following surveillances shall observed: be performed: 1. Ooerational Leakjigg 1. Doerational Leakaag a. Reactor coolant system Demonstrate drywell leakage is leakage shall be limited within the limits spEcified in to: 3.6.C.1 by monitoring the coolant leakage detection 1. No Pressure Boundary systems required to be operable Leakage by 3.6.C.2 at least once every 2. 15 gpm Unidentified 8 hours. Leakage 3. 525 gpm Total Leak 3ge averaged over any 24 hour period. Amendmtnt No. 42, 127, 140 125

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Dnn: LLAEL1.idd lhermal and Preswrizalien t.imitatienj (Cont'd) The reactor coolant system is a primary barrier against the release of fission products to the environs, in order to provide 3ssurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected. Appendix G to 10CFR50 defines the temperature-pressurization restrictions for hydrostatic and leak tests, pressurization, and critical operation. These limits have been calculated for Pilgrim and are contained in figures 3.6.1, 3.6.2, and 3.6.3. The bottom head, defined as the spherical portion of the reactor vessel located below the lower circumferential weld, was also evaluated. Reference transition temperatures (Rig ) were developed for the bottom head and the resulting pressure vs. temperature curves nlotted on figures 3.6.1 and 3.6.2. It has been determined that the bottom head temperatures are allowed to lag the vettel shell terperatures.

Reference:

Telodyne Engineering Services (TES) report TR-6051C-1, dated June 27, 1986. The referenced analysis utilizes the strest results established in the Combustion Engineering Inc., Pilgrim Reactor Vessel Design Report, No. CENC 1139, dated 1971, and combines the stress analysis results, specific to the bottom head, with the pressurization temperatures necessary to maintain fracture toughness requirements in accordance with the ASME Doiler and Pressure Vessel Code, Section 111, the criteria of 10CfR Part 50, Appendix G, and the supplementary guidelines of Reg. Guide 1.99, Rev. 2. i for Pilgrim pressure-temperature restrictions, two locations in the reactor vessel are limiting. The closure region controls at lower pressures and the b'Itline controls at higher pressures. The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner. Radiation exposure f rom f ast neutrons (>l mev) above about 10" nyt may shif t the NDT temperati.e of the vessel metal above the initial value. Impact tests from the first material surveillance capsule removed at 4.17 EfPY indicated a maximum Ri shift of 55'f for the weld specimens. g Neutron flur wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The wires and samples will be periodically removed and testeu to experimentally verify the values used for figures 3.6.1, 3.6.2, and 3.6.3. The withdrawal schtdule of l Table 4.6.3 has been established as required by 10CFR50, Appendix H. The Ri of the cloture region is - 5'f. The initial Ri for the beltline g g_ weld and basemetal are -50'f and O'f respectively. These RTm temperatures are based upon unirradiated test data, adjusted for specimen orientation in accordance with USNRC Branch Technical Position HTEB 5-2. 139 Amendment No. P/,92.140

flaid : 3.6.A and 4.6.A T helm 3 1 and Pr_tinLr.intion Limitations (Cont'd) The closure and bottom head regions are not exposed to neutron fluence () 1 Hev) over the vessel life sufficient to cause a shift in RT The pressure-temperature limitations (figures 3.6.1,3.6.2,and3.6.b3 of the closure and bottom head regions will therefore remain constant throughout vessel life. Only the beltline region of the reactor vessel will experience a shift in RT with a resultant increase in Pressure-Temperature limits. ug The curves apply to 100% bolt preload condition, but are conservative for lesser bolt preload conditions. For critical core operation when the water level is within the normal range for power operation and the pressure is less than 20% of the preservice system hydrostatic test pressure (313 psi), the minimum permissible temperature of the highly stressed regions of the closure flange is RTyp + 60 - 55'f. A conservative cutoff limit of 75'f was chosen as shown on Figurc 3.6.3 and as permitted by 10CFR50 Appendix G, paragraph IV. A.3. This same cutof f is included in the limits for hydrostatic and leak tests and for non-critical operation, as shown on figures 3.6.1 and 3.6.2 respectively, in order to be consistant with the limits for critical operation. The closure region is more limiting than the feedwater nozzle with regards to both stress intensity and RT Therefore the pressure-temperature limits yg. of the closure are controlling. The adjusted reference temperature shift is based on Regulatory Guide 1.99, Revision 2 dated May 1988; the analytical results of General Electric Report MDE 277-1285, Revision 1, dated January 21, 1985, regarding projected neutron fluence; and Teledyne Engineering Services Reports, TR-6052B-1, Revision 1, dated June 26, 1986, as supplemented by TR-7487, dated April 16, 1991, for RT vs. fluence as a function of temperature and pressure, and TR-6052C-1, yg dated June 27, 1986, for the RPV bottom head pressurization temperatures. Ar e nd m n t N o. 2 2, l m 139A

~ BASES: 3.9 The general objective of this Specification is to assure an adequate source of electrical power to operate the auxiliaries during plant operation, to operate facilities to cool and lubricate the plant during shutdown, and to operate the engineered safeguards following an accident. There are three sources of a-c electrical energy available; namely, the startup transformer, the diesel generators and the shutdown transformer. The d-c supply is required for switchgear and engineered safety feature systems. Specification 3.9.A states the required availability of a-c and d-c power; 1.e., an active off-site a-c source, a back-up source of off-site a-c power and the maximum amount of on-site a-c and d-c sources. The diesel fuel supply consists of two (2) 25,000 gallon tanks. The level alarms will be set to ensure a minimum supply of 19,800 gallons in each tank. Auxiliary power for PNPS is supplied from two sources; either the unit auxiliary transformer or the startup transformer. Both of these transformers are sized to carry 100% of the avviliary load. If the startup transformer is lost, the unit can continue to operate since the unit auxiliary transformer is in service, the i shutdown transformer is available, and both diesel generators are operational. If the startup and shutdown transformers are both lost, the reactor power level must be reduced to a value whereby the unit could safely reject the load and continue to supply auxiliary electric power to the station. In the normal mode of operation, the startup transformer is energized, two diesel generators and the shutdown transformer are operable. One diesel generator may be allowed out of service based on the availability of power from the startup I transformer, the shutdown transformer and the fact that one diesel generator carries sufficient engineered safeguards equipment to cover all breaks. With the shutdown transformer and one diesel generator out of service, both 345kV supply l lines must be available for the startup transformer. l Upon the loss of one on-site and one off-site power source, power would be available from the other immediate off-site power source and the one operable on-site diesel to carry sufficient engineered safeguards equipment to cover all breats. In addition to these two power sources, removal of the Isolated Phase Bus 4 flexible connectors would allow backfeed of power through the main transformer to { the Unit auxiliary transformer and provide power to carry the full station auxiliary load. The time required to perform this operation is comparable to the time the reactor could remain on RCIC operation before controlled depressurization r.ccd be initiatcd A battery charger is supplitd with each of the 125 and 250 volt batteries and, in additicn, (1) A 125 clt shared back-up battery charger is supplied which f,i f_,2 f I ! g

3.10 SASIS: B. Core Monitorina The SRH's are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup. Requiring two operable SRM's in or adjacent to any core quadrant where fuel or control rods are being moved ensures adequate monitoring of that quadrant during such alterations. The requirement of 3 counts per second provides assurance that neutron flux is being monitored and ensures startup is conducted only if the source range flux levt'. is above the minimum assumed in the control rod drop accident. The limiting conditions for operation of the SRM subsystem of the Neutron Monitoring System are derived from the Station Nuclear Safety Operational Analysis (FSAR Appendix G) and a functional analysis of the neutron monitoring system. The specification is based on the Nuclear Safety Requirements for Plant Operation in Subsection 7.5.10 of the FSAR. A spiral unloading program is one by which the fuel is in the outermost cells (four fuel bundles surrounding a control blade) is removed first. Unloading continues by removing the rema The center cellwillbethelastremoved.gngoutermostfuelcellbycell. A spiral loading program is one by which fuel is loaded on the periphery of the previously loaded fueled region beginning around a single SRM. Spiral unloading and reloading will preclude the creation of flux traps (moderator filled cavities surrounded on all sides by fuel). During spiral unloading, the SRH's shall have an initial count rate of 2 3 cps with all rods fully inserted. The count rate will diminish during fuel removal. Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRH's will drop below 3 cps before all of the fuel is unloaded. Since there will be no reactivity additions, a lower number of counts will not present a hazard. When all of the fuel has been removed to the spent fuel storage pool, the SRH's will no longer be required. Requiring the SRM's to be operational prior to fuel removal assures that the SRH's are operable and can be relied on even when the count rate may go below 3 cps. During spiral relohd, SRM operability will be verified by using a portable external source every 12 hours until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, up to two fuel assemblies will be loaded in different cells containing control blades around each SRH to obtain the required 3 cps. Until these assemblies have been loaded, the 3 cps requirement is not necessary. U)During selected refueling outages, prior to initiating spiral unloading, the central controlled cell will be removtd to facilitate inspection of the Core Spray Spargers. Amendrent to. JP. ??/.140 20M --}}