ML20148N453

Amend 37 to Provisional Oper Lic DPR-22,changing Tech Spec to Improve Simmer Margin of Safety/Relief Valves

ML20148N453
Person / Time
Site:	Monticello
Issue date:	11/06/1978
From:	Grimes B Office of Nuclear Reactor Regulation
To:
Shared Package
ML20148N445	List: ML20148N453 ML20148N471
References
NUDOCS 7811270103
Download: ML20148N453 (33)

Text

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[ga as c UNITED STATES e f NUCLEAR REGULATORY COMMISSION h1'y ' E 3 WASH:NGTON. D. C. 20555

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NORTHERN STATES POWER COMPANY '

I DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE 4

Amendment No. 37 License No. DPR-22 h i

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1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the Northern States Power Company (the licensee), dated March 21, 1978, as supplemented i August 10 and September 28, 1978 and applications dated .

September 30, 1977 and August 16, 1978, comply with the [

standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. Th'e facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; p

C. There is reasonable assurance (i) that the activities authorized [

by this amendment can be conducted without endangering the  !

heal th and safety of the oublic, 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 l common defense and security or to the health and safety of l the public; and l l

E. The issuance of this amendment is in accordance with 10 CFR Part d 51 of the Commission's regulations and all applicable requirements  !

have been satisfied.  !

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Provisional License No. DPR-22 is hereby amended to read as follows:

3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 37, 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.

FOR THE NUCLEAR Rt'GULATORY "0MMISSION l

, ,- s

,- M - c7 - it M,rrtfM-Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: November 6, 1978 l

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1 ATTACHMENT TO LICENSE AMENDMENT NO. 37 l PROVISIONAL OPERATING LICENSE NO. DPR-22 DOCKET l.J. 50-263 Replace the following pages of the Technical Specifications contained in Appendix A of the above indicated license with the attached pages. The changed areas on the revised pages are reflected by a marginal line.

i thru vi vii (DELETED) viii (DELETED) ix (DELETED) 2 3

6 I 7

8 9

10 (DELETED) 11 (DELETED) 12 (DELETED) 13 14 19 20 l 21 23 ,

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26 119 134 189B 189C 189D 189E 189F 189G 189H l

1891 (DELETED) 189J (DELETED) 199K (DELETED) id9L (DELETED) l 190 t

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TAHl.E OF CONTEN.TS-Page 1.0 Dl:FI N IT 10NS 1 2.0 S A FETY LIMITS AND LIttlTING SAFETY SYSTEM SEl' TINGS 6 l

2 .1 a nd 2 . 3 Fuel Cladding Integrity 6 2.1 Bases 13 2.3 Bases 17 2.2 and 2.4 Reactor Coolant System 23 l l

2.2 Bases 24 2.4 Bases 26 30 lit!lTING CONDITIONS FOR OPERATION AND 4.0 SURVEILLANCE REQUIREMENTS 28 l

3. i a nd 4.1 Reactor Protect ion System 28 3.1 Bases 37 4.1 Bases 43 3.2 a nd 4.2 Pr ot ec t ive Ins t ru me nta t io n 47 l A. Primary Containment Isolation Functions 47 l B. Eme rge ncy Core Cooling Subsystems Actuation 48 l C. Control Rod Block Actuation 48 D. Air Ejector Of f-Gas System 48 E. Reactor Building Ventilation Isolation and 49 Standtiy Gas Treatment System Initiation F Recirculation Pump Trip initiation 49 3.2 Bases 64 4.2 Bases 71

3. 3 a nd 4. 3 Control Rod System 75 i

A. React ivi ty Limitat ions 75 B. Control Rod Withdrawal 76 C. Sc ram Insetion Times 79 D. Control Rod Accumulators 80 C. Reactivity Anomalies 81 3 3 and 4.3 Bases 82 1

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Amendment No. 37

3.4 a nd 4.4 Standby L.iquid Control System 88 A. Normal Operation 88 B. Operat ion with Inope rable Components 89 C. Volume-Conce nt rat ion Requiretaents 90 3.4 and 4.4 Bases 94 3.5 a nd 4.5 Core and Containment Cooli ng Sys tems 96 l l

A. Core Spray Sys tem 96 f

B. LPCI Subsystem 98 i C. RilR Service Water System 101 D. HPCI System 103 E. Automatic Pressure Relief System 104 q F. RCIC System 106  !

G. tiinimum Core and Containment Cooling System 107 l Availability

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11 . Deleted

1. Recirculation System '

108A 3.5 Bases 109 4.5 Bases 114 1

3.6 a nd 4.6 Primary Sys tem Bounda ry 115 A. Reacto. Coola nt lleatup and Cooldown 115 B. Reactor Vessel Temperature and Pressure 116 C. Coola nt Chemistry 116A D. Co ola nt Le akage ll8A E. Safety / Relief Valves 119 F. Structural Integrity 120 C. Jet Pumps 120 11 . Snubbers 121 3.6 and 4.6 Bases 130 g

3. 7 a nd 4. 7 Containment Systems 139 ,

A. Primary containment B. Standby Cas Treatment System 139 )

148 i C. Seconda ry Containment 150 D. Primary Containment Isolation Valves 151 .

t 3.7 Bases 156 I 4.7 Bases 161 { ,

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i Amendment No. 37 I il ;1 i'

'- 4 4.8 Radiat lon Environmental Monitoring Program i68 3.8 and 4.8 Bases 175 3.9 and 4.9 Auxiliary Electrical Systems 180 A. Operational Requirements for Startup 180 B. Operational Requirements for Continued Operation 181

1. Transmission Lines 181 I

2. Rese rve Trans forme rs 182

3. Standby Diesel Generators 182

4. Station Battery System 183 3.9 Bases 185 4.9 Bases 186 3.10 and 4.10 Refueling 187 A. Ref ueling Interlocks 187 B. Core Monitoring 188 C. Fuel Storage Pool Water Level 188 D. Movement of Fuel 188 E. Ext ended Core and Control Rod Drive Maintenance 188A 3.10 and 4.10 Bases 189 3.1 I a nd 4.11 Reactor Fuel Assemblics 189B i

A. Ave rage Planar Linear lleat Generation Rate 189B B. Linear lleat Generation Rate 189C l C. Minimum Critical Power Ratio 189D 1

3.11 Bases 189E 4.11 Bases 189G 3.12 and 4.12 Scaled Source Contamination 189N A. Contaminatlon 189N B. Records 189P l '1 1 3.12 and 4.12 Bases 189Q l

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Amendment No. 37  !

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3.13 and 4.13 Fire Detection and Protection Systems 189R A. Fire Detection Ins t ru me nt a t ion 189R B. Fire Suppression Water System 189S C. Hose Stations 189U I D. Fire Barrier Penetration Fire Seals 189V l

3.13 Bases 189W  !

4.13 Bases 189X i

!1 5.0 DESIGN FEATURES 190 1

5.1 Site 190 4 5.2 Reactor 190 l 5.3 Reactor Vessel 190 5.4 Containment 190 .

5.5 Fuel Storage 191 5.6 Seismic Design 191  !;

H 6.0 ADMINISTRATIVE CONTROLS 192 6.1 Organization 192 i; 6.2 Review and Audit 195 6.3 Special Inspections and Audits 201 I

6.4 Act ion to be taken if a Safety Limit is Exceeded 201 6.5 .Pla n t operating Procedures 202 6.6 Plant Ope rating Records 209 l 6.7 Repo rt ing Requirements 211  ;

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Amendment No. 37 L il l

LIST OF FIGURES F l y,u r e No. Page No.

2.3.1 APRM Flow Referenced Scram and Rod Block Trip Settings 11 2.3 2 Relationship Between Peak fleat Flux and Power for Peaking 12 Factors of 3.08 ( 7x 7 fuel) and 3.04 (8x8 fuel) 4.1.1 'M' Factor - Graphical Aid in the Selection of an Adequate 46  ;

Interval Between Tests 4.2.1 System Unavailability 74 3.4.1 Sodium Pentaborate Solution Volume-Concent ration Requirements 92 3.4.2 Sodium Pentaborate Solution Tempera ture Requirements 93 3.6.1 Change in Charpy V Transition Temperature versus Neutron Exposure 122 3.6.2 llinimum Tempe rature ve rsus Pressure for Pressure Tests 122A 3.6.3 Minimum Temperature versus Pressure for Mechanical lleatup or 122B Cooldown Fo11ouine, Nuclear Shutdown J. 6 Mi n inuim Tempt iatire versus Pr essure for Core O pe ra t i c;n 122C 4 . ti . 2 Ch l o r i de Stress Lorrosion Test. Results @ 500 F 123 4.8.1 bampling I,ocations - Monticello Nuclear Generating Plant 173 Radiation Environmental Monitoring Program 3.11.3 K Factor versus Pe rc ent of Rated Core Flow 189M 6.1.1 Nfi l' Corpo ra t e Organiza tional Rela t ionship to On-Site Operating 19 3 Urganization 6.1.2 Monticello Nuc lear Gene rat ing Plant Functional Organiza tion for 194 Un-Site Operating Group i

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Amendment No. 37 j l

LIST OF TAllLi'S Table No. Page No.

3.1.1 Reactor Protection System (Scram) Instrument Requirements 30 4.1.1 Sc ram Ins trument Functional Tests - Minimum Functional Test 34 Frequencies for Safety Instrumentation and Control Circuits 4.1.2 Scram Ins trument Calibration - Minimum Calibration Frequencies 36 for Reactor Protection Instrument Channels 3.2.1 Instrumentation that Initiates Primary Containment Isolation 50 i Functions 3.2.2 Ins t rument ation that Initiates Emergency Core Cooling Systems 53 l 3.2.3 Ins tr'imentat ion that Initiates Rod Block 57 j 3.2.4 Instrumentation that Initiates Reactor Building Ventila tion 60  ;

Isolation and Standby Gas Treatment System Initiaticn l.2.5 Trip Funct ions and Deviations 69 4.2.1 Minimum Test and Calibra tion Frequency for Core Cooling, Rod Block 61 and I so la t ion ins t rume nt a t ion )

3.6.1 Saf ety Related Snubbers 121B j 4.6.1 In-Service Inspection Requirements for Monticello 124 3.7.1 l'rlma ry Containment Iso la tion 153 4.8.1 Monticello Nuclear Plant - Environmental Monitoring Program 169 Sample Collection and Analysis 3.11.1 Maximum Average Planar Linear Heat Generation Rate 189E l 6.1.1 Minimum Shift Crew Composition 194A l 6.5.1 Protection Factors for Respirators 206 l

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l Amendment No. 37 vi

D. Irrediate - It=ediate means that th e required action will be initiated as soon as practicable conside ring the safe operation cf the unit nr he inpertance of the required action.

E. Instrurent Functional Test - An instrua 1 test neans the injection of a si=ulated signal into the primary sensor to ve rify s t ruce nt channel respense, alar =, and/or initiating action.

F. Instrument Calibration - An ins truce nt calibration neans the adjustrent of an instrument signal cutput so that it corresponds, within acceptable range, accuracy, and response tire to a inewn value (s) of the paraceter which the instrunent monitors. Calibratica shall encorpass the entire in s tru re nt including actuation, ala rm or trip. Response time is not part of the routine ins trume nt calibration but will be checked once per cycle.

G. Li-it ic~ Cerdit icns for C-eration (LCO) - Th2 liniting cenditiens fer opera tien specify the ninicun acceptcble levels cf sys tem pe rf e rnance necessary to assure safe startup and operation of the facility. When these conditions are ret, the plant can be operated safely and abnorcal situations can be saf ety controlled.

H. Liritine Safetv_yvsten Settin~ ( LS SS) - The limiting safety systen settings are settings on ins trumentnt ion wnich initiate the autcastic protective act ion at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents cargin with normal operation lying below these settings. The nargin has been es tablished so that with proper operation of the instrumentation, the safety linits will never be exceeded.

I. Maxinu- Fraction of Limitine Pcwer Density (MFLPD) - The maxinun fraction of limiting power density is the highest value in the core of the ratio of the existing to the design linear heat generation rate.

I J. Minicum Critical Power Ratio (MCPR) - The minimun critical power ratio is the value of critical pewe r ratio asscciat ed with the mos t limiting assembly in the reactor core. Critical power ratio (CPR) is the ratio of that power in a fuel assembly which is calculated by the GEXL correla tion to cause some point in the assenbly to experience boiling transition to the actual assenbly operating power.

l K. Mode - The reactor code is that which is established by the mode-selector switch.

l L. Operable - A sys ten or component shall be considered operable when it is capable of perforcing its intended f unction in its required manne r.

! I M. Onorating - Operating ecans that a system or component is perforning its required functions in its required manner.

Amendment No. 37 2 10

I *: . Om rari-r Cvela - In t e rva l between the end of one refueling cutage and the end cf the next s u3 s eq ue nt refueling cutcge.

O. Power Operation - Power Operation is any operation with the node switch in the " Start-Up" or "Run" position with the reactor critical and above 1% rated thermal power.

P. Pricary Centaincent Intecrity - Primary Containment Integrity means that the drywell and pressure suppression chacber are intact and all of the following conditions are satisfied.

1. All canual containcent isolation valves on lines connecting to the reactor coolant system or containment which are not required to ba open during accident conditions are closed.

2. At least one door in the airlock is closed and scalec.

3. All autcratic containment isolation valves are operable or are deactivated in the closed position or at least ene valve in each line having an inoperable valve is closed

4. All blind flanges and manways are closed.

Q Protective Instrumentation Lcgic Definition <

l. Instrucent Channel - An instrument channel means an arrangement of a senser and auxiliary equiptent required to generate and transmit to a trip system, a single trip signal related to the plant parameter monitored by that instrucent channel.

2. Trio Systeu - A trip system means an arrangerent of instrupent channel trip signals and auxili vry equipment required to initiate a prctection action. A trip system may require one or more instrument channel trip signals related to one or core plant paraceters to initiate trip system action. Initiation of the protective function may require tripping of a single trip system (e.g., HPCI system isolation, off gas system isolation, reacter building isolation and standby gas treatment initiatien, and rod block), or the coincident trippin of two trip systems (e.g., initiation of scran, reactor isolation, and primary containment isolation).

3 Protective Action - An action initiated by the protection system when a limit is exceeded.

A protective action can be at channel or system level.

Amendment No. 37 3

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Aonlicability:

. Applies to the interrelated variables Applies to trip settings of the instruments and aa actated with fuel thermal bchavior. devices vi.ich are provided to prevent the reactor system safety limits free being exceeded.

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Objective:

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To establish licits below which the To define the level of the process variables in:c;;rity of the fuel cladding is preserved. at which autecatic prctective action is initiated to prevent the sciety limits frc=

being exceeded.

SL ification: Specification:

A. Core Thermal Power Limit (Reactor The Limiting safety system settings shall be as

?rcssure > 800 Psia and Core Flow is specified below:

> 10% of Rated)

A. Neutron Flux Scran

• When the reactor pressure is > 800 Psia

! and core ficw is > 10*.' of rated, the 1. APR"i - The APRM flux scran trip setting

existence of a minimum critical power shall be:

ra tio (MCPR) less than 1.0 7 for 8x3 f uel cad less than 1.07 for 8x3R fuel shall S$ 0.65 W + 55%

' constitute violation of the fuel cladding wh e re ,

integrity safety limit S= Setting of percent of rated the rmal p ,we r , rated power being 16 70 m.T U= recirculation drive flow in

. percent Amendment No. 37 6 2.1/2.3

.f 2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 4

B. Core Thernal Power Limit (Reactor except in the event of operation with a maximum f raction of limiting power density Pressure < 800 Psia or Core Flow <g 10% of Rated) for any fuel type in the core greater than the fraction cf rated pcwer, when the setting shall be rodified as follcus:

When the rea: tor pressure is $[ 800 psia or core flov is :s 10% of rated, the core 7g, therral power shall rot exceed 25% of S < (0.65 W + 55%) 37{gD rated the rnal power.

where, FRP = fraction of rated thermal power, C. Power Transients rated power being 1670 MWt MFLPD = maxinus fraction of limiting To insure that the safety limit established in Speciiication 2.1.A is not exceeded, each pcwer density for any fuel type in the core.

required scran shall be initiated by its ,

printry source signal as indicated by the plant process cecputer

2. IRM - Flux Scram setting shall be 5; 20% of rated neutron flux B. APRM Rod Block - The APRM red block setting shall be:

3 S S 0.65 W + 43%

where, S= Setting of percent of rated thermal power, rated power being 1670 MWT U= recirculation drive flow in percent Amendment No. 37 7 l 2.1/2.3 s

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i i l 2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS A E. Turbine Control Valve Fast Closure Scram shall

, initiate upon loss of pressure at the accele ration relay with tetrbine first stage pressure 2 301.

F. Turbine Stop Valve Scram shall be < 10% valve clos ure from full open with turbine first stage pressure 2 30%.

G. Main Steamline Isolation Valve Closure Scram shall be 5 10% valve closure from full open.

H. Main Steamline Pressure initiation of main steam-line isolation valve closure shall be 2 825 psig.

Amendment No. 37 9

, 2.1/2.3 i NEXT PAGE IS 13

Scses:

2.1 The fuel cladding integrity linit is set such that no calculated fuel damage could occur as a result of an abnormal operational transient. Because fuel dacage is not directly observable, a step-back l appreach is used to establish a Safety Linit such that the MCFR is no less than 1.07. This licit represents a conser.ative margin relative to the conditions required te raintain fuel claddir integrity. The fuel cladding is one of the phyaical barriers which sepa ra te radioactive materials fron the environs. The integrity of this cladding ba rrier is related to its relative freeden frc=

perfora tions or cracking. Although scoe cerrosion or use related cracking tay occur during the life of the cladding, fission product migration from this ecurce is incrementally cunulative and continuously reasurable. Fuel cladding perforations, however, can result f roc the rral s tresses which occur frca reacter eperation significantly above design conditions and the protection systen safety settings.

, .lile fissien preduct cigration f rca cladding perf oration is j ust as neasurhale as that from use related cracking, the thermally caused cladding perforations signal a threshold, beyond which still

, greater thernal stresses may cause gross rather than incremental cladding de teriora tion. Therefore, i

the feel cladding Safety Limit is defined with margin to the conditions which would produce onset of f trans it ien boiling. (MCPR of 1.0). These conditions represent a significant departure frem the

, condi t io n intended by design for planned operation. The concept of MCPR, as used in the GETAS/CEXL i

critical power analysis, is discussed in Reference 1.

i A. Core Thermal Power Limit (Reactor Pressure > 800 psia and Core Flow > 10% of Rated.) Onset of I

transition boiling results in a decrease in heat transfer from the clad and. therefore, elevated clad t aperature and the possibility of clad failure. Ecwever, the existence of critical power, i

or boiling transition, is not a directly observable pa rame te r in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel e

assenbly is characterized by the critical power ratio (CPR) which is the ratio of the bundle l power which would produce onset of transition boiling divided by the actual bundle power. The minicut value of this ratio for any bundle in the core is the miniaun critical power ratie j (MCPR). It is assumed tha t the p la nt oper . ion is controlled to the nominal protective setpoints

' via the ins trumented va riables. The Safety Limit (T.S.2.1.A) has sufficient conserva tism to assure that in the event of an abnormal operational transient initiated from the Operating MCFR

} Liait (T.S.3.11.C) more than 99 9% of the fuel rods in the core are expected to avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the Safety Limit i

j Amendment No. 37

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{ 2.1 Bases

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is derived f rem a detailed statistical analysis considering all ci the uncertaintres in coaitoring the core operating s tate including uncertainty in the boiling transition correlatica as described in Reference 1 The uncertainties enployed in dariving the Safety Licit are provided at the beginning of each fuel cycle.

Because the boiling transition correlation is based on a large quantity of full scale data, there is a very high confidence that operation of a fuel asse=bly at the MCFR Safety Limit would not produce boiling transition. Thus, although it is not required to establish the Safety Limit, additional cargin exists between the Safety Limit and the actual occurrence of loss of

'I cladding integrity.

3 I Howeve r , if boiling transition were to occur, clad perforation would not be expected. Cladding

'l temperatures would increase to approximatley 1100 r which is below the perforation temperature of the cladding caterial. This has been verified by tests in tre General Electric Test Reactor i

(CETR) where fuel similar in design to Monticello operated above the boiling transition for a j sign ificant period of time (30 minutes) without clad perforation.

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! If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of

' applicability of the boiling transition correlation) it would he assumed that the fuel cladding integrity Safety Limit has been violated.

In addition to the MCPR Safety Limit, operation is constrained to a maximum design linear heat i generation rate for any fuel type in the core.

B. Core Thermal Power Limit (Reactor Pressurc $ 800 psia or Core Flow $ 10% of Rated) At I pressure below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 f psi. At low powers and all core flows, this pressure differential is maintained in the bypass i region of the core.

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Amendment No. 37 ,,

2.1 BASES L

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3ases Centinued:

' that indicated by the neutron flux at the scram setting. Analyses ceconstrate that, with a 120%

scram trip setting, none of the abnor=al operational transients analyzed violate the fuel Safety Linit and there is a substantial cargin fren fuel damage. Therefore, the use of flow referenced scra: trip provides even additional cargin.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Linit is rer :hed. The APRM scram trip setting was determined by an analysis of targins reqcired to provide a reasonable range for caneuverirg during cperation. Eeducing thison cperating cargin would increase the frequency of spuricus scrams which have an adverse effect reccter safety because of the resulting thernal s tresses. Thus, the AFRS ceram trip setting was selected because it prevides adequate cargin for the fuel cladding integrity Safety Licit yet allows operating targin that reduces the possibility of unnecessary scrams. Therefore, it is intended to ultimately replace (with prior !:RC approval) the autenatic flow referenced scran with a fixed 120 percent scrar setting.

Tne scrat trip setting must be adjusted to ensure that the LEGR transient peak is not increased for any cochination of maximum f raction of limiting power density and reactor core ther=al power. The scram setting is adjusted in accordance with the forcula in Specification 2.3.A.1, when the raxitum fraction of liciting power density is greater than the frnction of rated power. If the APRM scram setting should require a change due to an abnorral peaking cendition, it will be done by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced scram curve by the reciprocal of the APRM gain change. Analyses of the limiting transients show that no scram adj ustme nt is required to assure that the MCPR Safety Limit (T.S.2.1.A) is not exceeded when the t ra ns ie nt is initiated from the Operating MCPR Limit (T.S.3.11.C).

Fo r ope ra tion in the startup code while the reactor is at low pressure, the IRM scram setting of 20%

of rated power provides adequate thermal margin between the setpoint and the safety limit, 25% of rated. The cargin is adequate to acconcodate anticipated maneuvers associated with power plant s tart up. Ef fects of increasing pressure at ze ro or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, tenperature coefficients are saali, and control rod patterns are constrained to be uniform by operating procedures Amendment No. 37 19 2.3 BASES

4 Bases Continued:

backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod unifor= control rod withdrawal is pattern. Thus, of c11 possible sources of reactivity input, the most prebable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power In an assumed rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.

uniform red withdrawal approach to the scram level, the rate of power rise is no more than to assure a scram 5% of rated power per minute, and the IRM system would be more than adequate before the power could exceed the safety limit. The IRM scram remains active until the mode switch is placed in the run position. This switch occurs when reactor pressure is greater than 850 psig.

The analysis to support operation at various power and flow relationships has considered opera-tion with either one or two recirculation pumps. During steady-state operation with one recircula-tion pump operating the equalizer lina shall he open. Analysis of transients from this operating condition are less severe than the same transients from the two pump operation.

The operator will set the APRM neutron flux trip setting no greater than that stated in Specifica-tion 2.3.A.I . However, the actual setpoint can be as much as 3% greater than that stated in l Specification 2.3. A.1 for recirculation driving flows less than 50% of design and 2% greater the deviations than that shown for recirculation driving flows greater than 50% of design due to discussed on page 18.

B. APRM Control Rod Block Trips Reactor power level may be varied by moving control rods or. by varying the recirculation flow rate. The APRM system provides a control rod block to prevent

ate, and thus to protect red withdrawal beyond a iS ven point at constant recirculation flo This red block trip against the condition of a MCPR less than the Safety Linit (T.S.2.1.A).

setting, which is automatically varied with recirculation loep flow rate, prevents The an incrc2se in the reactor power 1cvel to excessive values due to centrol red withdrawal. flow varieble trip setting provides substantial margin from fuel damage, assuming a steady-state The margin to the Safety operation Limit at the trip setting, over the entire recirculation flow range.

Amendment No. 37 20 7.' zes. ;

Bases Continued:

increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored by the in-core LPRM system. When the maximum fraction of limiting power density exceeds the fraction of rated thermal reactor power, the rod block setting is adjusted in accordance with the formula in Specification 2.3.B. If the APRM rod block setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced rod block curve by the reciprocal of the APRM gain change.

The operator will set the APRM rod block trip settings no greater than that stated in Specification 2.3.3 However, the actual setpoint can be as cuch as 3% greater than that stated in Specification 2.3.S for recirculation driving flows less than 50% of design and 2% greater than that shown for recirculation driving flows greater than 50% of design due to' the deviations discussed on Page 18 C. Reactor Low Unter Level Scran The reactor low water lev'1 scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained.

The operator will set the low water level trip setting no lower than 10'6" above the top of the active fuel. However, the actual setpoint can be as cuch as 6 inches lower due to the deviations discussed on page 18.

D. Reactor Low Low Water Invel ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit fuel clad temperature to well belew the clad telting temperature to assure that core geonetry remains intact and to limit any clad retal-water reaction to less than 1%.

The design of the ECCS components to meet the above criterion was dependent on three previously set pa rame te rs : the maximum break size, the Icw water level scran setpoint, and the ECCS initiation setpoint. To lower the setpoint for initiation of the ECCS could prevent the ECCS conponents fren 21 2.3 EASES Amendment No. 37

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.2 REACTOR COOLANT SYSTEM 2.4 REACTOR COOLANT SYSTEM Applicability: Applicability:

Applies to limits on reactor coolant Applies to trip settings of the instruments system pressure. and devices which are provide i to prevent the reactor system safety limi s from being exceeded.

Objective: Objective:

To es tablish a limit below which the To define the level of the process variables integrity of the reactor coolant at which automatic protective action it sys tem is not threatened due to an initiated to prevent the safety limits from overpressure condition. being exceeded.

5pecification: , Specification:

The reactor vessel pressure shall A. Reactor Coolant High Pressure Scram shall not exceed 1335 psig at any time be f: 1075 psig.

when irradiated fuel is present in the reactor vessel. B. The self-actuation function of at leas t seven Reactor Coolant System safety relief valves shall be operable.

Valves shall be set as follows:

8 valves at [= 1108 psig.

Amendment No. 37 23 2.2/2.4

Bases Continued:

2. 2. The normal operating pressure of the reactor coolant system is approximately 1025 psig. The turbine trip with failure of the . bypass systen represents the mos t severe prinary system pressure increase resulting frca an abnormal operational transient. The peak pressure in this transient is limited to 1207 psig. The safety / relief valves are sized assuming no direct scram during MSlv closure. The cnly scran assuced is from an indirect means (high flux) and the pressure at the bottom of the vessel is limited to 1248 psig in this case. The analysis assumed that only seven of the eight valves are operable and that they open at 1% over their setpoint with a 0.4 second delay. Reactor pressure is continuously monitored in the control roco during operation on a 1500 psig full scale pressure recorder.

s Amendment No. 37 25 7 2.2 BASES

Bases:

2.4 The settings en the reactor high pressure scram, reactor coolant system safety / relief valves, turbine control valve fast closure scra=, and turbine step valve closure scram have been established to assure never reaching the reactor coolant system pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety limit. The APRM neutron flux scram and the turbine bypass system also provide protection for these safety limits.

In addition to preventing power operation above 1075 psig, the pressure scram backs up the APRM neutron flux scram for steam line isolation type transients.

The reactor coolant system safety / relief valves assure that the reactor coolant system pressure safety limit is never reached. In cocpliance with Section III of the ASME Boiler and Pressure Vessel Code, 19 65 edition, the safety / relief valves must be set to open at a pressure no higher than 105 percent of design pressure, and they must limit the reactor pressure to no more than 110 percent of design pressure. The safety / relief valves are sized according to the Code for a condition of MSIV closure while operating at 1670 MWt, followed by no MSIV closure scram but scram f rom an indirect (high flux) means. With the safety / relief valves set as specified herein, the maximum vessel pressure (at the bottom of the pressure vessel) would be about 1248 psig. Only seven of the eight valves are assumed to be ope rab le in this analysis and the valves are assumed to open at 1%

above their setpoint with a 0.4 second delay.

The operator will set the reactor coolant high pressure scram trip setting at 1075 psig or lower.

However, the actual setpoint can be as much as 10 psi above the 1075 psig indicated set point due to the deviations discussed in the basis of Specification 2.3 on Page 18. In a like manner, the operator will set the reactor coolant system safety / relief valve initiation trip setting at 1108 psig or lower. However, the actual set point can be as much as 11.1 psi above the 1108 psig indicated set point due to the deviations discussed in the basis of Specification 2.3 on Page 18.

A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting, or when a suf ficient number of devices have been af fected by any means Amendment No. 37 26 2.4 BASES

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILI>dCE REQUIREMENTS E. Safety / Relief Valves E. Safety / Relief Valves

1. During power operating conditions 1. a. A cinimum of seven safety / relief and whenever reactor coolant pressure valves shall be bench checked or is greater than 110 psig and replaced with a bench checked te=perature is greater than 345 F. valve cath ref ueling cu tage.

The nominal setpoint of all opera-

a. The safety valve function (self- tional safety / relief valves shall actuation) of seven safety / be 1108 psig.

relief valves shall be operable.

b. At least two of the safety / relief

b. The solenoid activated relief valves shall be disassembled and function (Autcoatic Pressure inspected each ref ueling outage.

Relief) shall be operable as required by Specification 3.5.E. c. The integrity of the safety / relief valve bellows shall be continuously t,o nit o r ed .

d. The operability of the bellows monitoring system shall be demon-strated at least cnce every three months.

Amendment No. 37 ,

119 3.6/4.6

Bases Continued 3.6 and 4.6:

~

w D. Coolant Leakage The former 15 gpc limit for leaks frca unidentified sources was established assuming such a leakage was cc=ning f rem the primary system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will prepagate. From the crack size a leakege rate can be determined. For a crack size which gives a leakage of 5 gpm, the probability

~#

of rapid prepagation is less than 10 . Thus, an unidentified leak of 5 gpm when assumed to be from i the primary systen had less than one chance in 100,000 of propagating, which provides adequate margin. ,

A leakage of 5 gpc is detectable and measurable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed for determination of leakage is also based on the low probability of the crack propagating.

The capacity of the drywell su=p pumps is 100 gpm and the capacity of the dryvell equipment drain tank pumps is also 100 gpm. Removal of 25 gpo from either of these sumps can be ac~couplished with consider-able margin.

E. Sefetv/Eelief Valves Tes ting of all required safety / relief valves each refueling cutage ensures that any valve deteriora tion is detected. A tolerance value of 1% for safety / relief valve setpoints is snecified in Section 1I1 ef the ASME Boiler and Pressure Vessel Code. Analyses have been performed with all valves assumed sct 1%

higher (1138 psig - 1%) than the nominal setpoint; the 1375 psig ccdc limit is not excceded in any case.

Ihe safety / relief valves are used to limit reactor vessel overpressure and fuel thermal duty.

The required safety / relief valve steam flow capacity is determined by analyzing the transient accomp any-ing the main steam flow stoppage resulting from a postulated FSlV Closure from a power of 1670 Mwt.

The analysis assumes a multiple-failure wherein direct scram (valve position) is neglected. Scram is assumed to be from indirect means (high flux). In this event, the safety / relief valve capacity is assumed to be 83.2% of the full power steam generation rate.

Amendment No. 37 134 3.6/4.6 BASES

2 2 3 0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRE:IENTS 3.11 REACTO3 FUEL ASSEMBLIES 4.11 REACICR FUZL ASSEMELIES Applicability Aralicability The Limiting Conditions for Operation associated The Surveillance Requirenents apply to with the fuel rods apply to these parancters the parameters which renitor the fuel which conitor the fuel red operating conditions. rod operating conditions.

Objective Obiective The objective of the Limiting Conditions for Opera-  ! The cbjective of the Surveillance Require-tion is to assure the perforcarcc of the fuel rods.  !' ncats is to specify the type and frequency of surveillance to be applied to the fuel

rods.

I i Specifications Specifications

.i A. Average Planar Linear Heat Generation Rate (APLEGR) A. Averane Planar Linear Heat Genera-tion Rate ( AP LEGR)

During power operati 'n, the APLEGR for each type of fuel as a fu- on of average planar The APLEGR for each type of fuel as l exposure shall not ext .1 the limiting value a function of average planar exposure

! given in Table 3.11.1 based on a straight shall be determined daily during line interpolation between data points. When reactor operation at 2 25% rated l

core flow is less than 90% of rated core flow, the rnal power.

I the APLEGR shall not exceed 95% of the limit-

{ l ing value given in Table 3.11.1. When core flow is less than 70% of rated core flow,

{

the APLEGR shall not exceed 90% of the limit-9 ing value given in Table 3.11.1. If any l

,. t ime during operation it is de termined

'I that the limit for APLEGR is being ex-lf ceeded, action shall be initiated within 15 i

1 t

Amendment Nos. ,,27T'37 3.11/4.1' 1

- l 3 0 LIMITING CONDITICNS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS minutes to res to re operation to within the presc ribed limits. Surveillance and corres-pending action shall continue until reactor operation is within the prescribed limits.

If the A?LHGR is net returned to within the prescrib2d limits within two (2) hours, the reactor shall be brcught to the Cold Shutdown condition within 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />s-3 Linear Heat Generation Rate (LHGR) B. Line.-r Heat Generation Rate (L3GR)

The LHGR as a function of core height During power operation, the LiiGR as a function of core height shall be limited to:

shall be checked daily during reactor operation at 2 25% of rated thermal

.022 X/L) power.

LEGR 513 4(1 where, X = Elevation from the bottom of the core L = Fuel Column Length i

If at any time during operation it is de-termined tha t the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the presc ribed linit s. Surveillance and corresponding action shall continue until reactor operation is within the pre-scribed licits. If the LiiGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.

Amendment No. 37 189C l 3.11/4.11

3.0 LIMITING CONDITIONS POR OPERATION 4.0 SURVEILLANCE REQUIREMENTS C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio (MCPR)

1. During power operation, the Operating MCPR MCPR shall be determined daily during Limit shall be Erl.33 for 8x8 fuel and reactor power operation at 2'25% rated EB1.33 for 8x8R fuel at rated power and thermal power and following any change flow. If at any time during operation it in power level or distribution which is de termined that the limiting value for has the potential of bringing the core MCPR is being exceeded, action shall be to its operating MCPR limit.

initiated within 15 minutes to restore ope ra tion to within the prescribed limits.

Surveillance and corresponding action

.shall continue until reactor operation is within the prescribed limits. If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For core flows other than rated the Operating MCPR Limit shall be the above applicable MCPR value times K7 where Kg is as shown in Figure 3.11.3.

2. If the gross radioactivity release ra te of noble gases at the steam jet air ejector monitors exceeds, for a period greater than 15 minutes, the equivalent of 236,000 uCi/sec following a 30-minute decay, the Operating MCPR Limits specified in 3.11.C.1 shall be adjusted to jb 1.45 for 8x8 fuel and 2t1.40 for 8x8R fuel, l times the appropriate Kr. Subsequent oper-ation with the adjusted MCPR values shall be per paragraph 3.ll.C.1.

l t

l 189D 3 11/4.11 AmendrentNos.20jj'f'37 O

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. Bases 3.11 4

A. Averc9e Planar Linear Heat Generation Rate ( AP L"GR)

This specification assures that the peak cladding terperature follcwing the postulated design bcsis

Ics s-o f-c ocia nt cccident vill not exceed the licit specified in the ICCFR30, Appendix K.

I i

The peak cladding te=perature following a postula ted Icss-of-ccciant sccident is pricarily a function n' the vern ;e hect generatien rete of all the reds cf a fuel ascerbly et any axici location and is c: ' dependent recondcril:. on t'r e rod to rod power distribution within an assenbly. Since expected Iccal variatio,- f j power digtribution within a fuel casenbly af fcct the calculated peak cladding te perature by less than +20 relative to the peak tcaperature for a typical f uel design, the linit on the average linear heat generation rate is sufficient to assure thaf calculated temperatures are within the 10CFR50 Appendix E licit. The liciting value for APiEGR is given by this specification.

Reference 6 denonstrates that for lower initial core ficw rates the potential exists for earlier D:iB during pos tulated LOCA's. Therefore a mo re restrictive limit for APLHGR is required during reduced flow conditions.

e i Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic I

reactor scran are not considered a violation of the LCO. Exceeding APLEGR limits in such cases need

, not be reported.

4 i B. LEGR i

i This specification assures that the linear heat generation rate in any rod is less than the design j '

linear heat generation if fuel pellet densification is pos tulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference I and in References 2 and 3, and j assures a linearly increasing variation and axial gaps between core botton and top and assures with a

] 95% conf idence , that no more than one fuel red exceeds the design linear heat generation rate due to

! I power spiking.

. Those abnormal operational transients, analyzed in FSAR Section 14.5, which rreult in an autcsatic j reactor scran are not considered a violation of the LCO. Exceeding LEGR limits in such cases need not be reported.

i i

i 3 Amendment No. 37 4

j l 189F J 3.11 na sr.s

___________________.___-_m _ _ _ __--m- _ . _

Eases Continued C. Minimum Critical Power Ratio (MCPR)

The ECCS evaluation presented in Reference 4 assumed the steady state MCPR prior to the postulated loss-of-coolant accident to be 1.18 for all f uel types. In addition, the ECCS analysis presented in Reference 6 assumed an initial MCPR of 1.24 for reduced flow conditions. The Operating MCPR Limit of l.33 for 8x8 fuel and 1.33 for 8x8R fuel is determined from the analysis or- transients discussed in

{ By maintaining an operating MCPR above these limits, the Safety Limit Bases Sections 2.1 and 2.3 (T.S . 2.1. A) is maintained in the event of the mos t limiting abnormal operational transient.

For operation with less than rated core flow the Operating MCPR Limit is adjusted by multiplying the above limit by K . Reference 5 discusses how the transient analysis done at rated conditions 7

encompasses the reduced flow situation when the proper K factor is applied.

Noble gas activity levels above that stated in 3.ll.C.1 are indicative of fuel failure. Since the failure mode cannot be positively identified, a more conservative Operating MCPR Limit must be applied to accounti for a possible fuel loading error.

Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO. Exceeding MCPR limits in such cases need not be reported.

6 Amendment Nos. [ 37 3.I1 BASES

_ _ _ _ :_ _ _ _ _ r _z__rz_==__  : _ _ __ _ _ __ _ _ _ :_ _ _ __ __ z_r_: _ __: r -- - - - - - - - - - - - - - - - - -

References .

1. " Fuel Densification Ef fects in General Electric Boiling k'ater Reactor Fuel," Supplements 6, 7, and 8, NEDM-10735, August, 1973.

2. Supplement I to Technical Report on Densification of General Electric Reactor Fuels, Dececber 14, 1974 (USAEC Regulatory Staf f) .

3. Communication: VA Moore to IS Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27, 19 74.

4. "Los s-o f-c oo la nt Accident Analysis Report for the Monticello Nuclear Generating Plant," NEDO-24050, September 1977, L 0 Mayer (NSP) to V Stello (USNRC), Sep t emb e r 15, 1977.

5. " General Electric Bk'R Generic Reload Application for 8x8 Fuel," NEDO-20360, Revision 1, November 19 74.

6. " Revision of low Core Flow Effects on LOCA Analysis for Operating Bk'R's," R L Gridley (GE) to D G Eisenhut (USNRC) , Septembe r 28, 1977.

Bases 4.I1 The APLHGR, LilGR and MCPR shall be checked daily to determine if fuel burnup, or control rod movement have caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are remved daily, a daily check of power distribution is adequate. For a limiting value to occur below 25% of rated thermal power, an unreasonably large peaking f actor would be required, which is not the case fo. operating control rod sequences. In addition, the MCPR is checked whenever changes in the core power level or distribution are made which have the potential of bringing the fuel rods to their thermal-hydraulic limits.

Amendment No. 7 gggg 4.11 BASES l

NEXT PAGE IS 189M

=

5.0 DESIGN FEARTRES 5.1 Site A. The reactor center line is located at app roxira tely 850,810 feet North and 2,038,920 feet East as deternined on ghe Minnesota State Grid, Seuth Zone. The nearest site boundary is approxinatelv '

1630 feet S 30 W of th e reactor center line and the exclusion area is defined by the minitun fenced area shewn in FSAR Figure 2.2.22. Luc to the prevailing wind pattern, the direction of saximun integrated desage is SSE. The southern property line follevs the northern boundary of the right-of-way for the Burlington Northern Railway.

5.2 Reactor

! A. The reactor core shall consist of not more than 484 fuel assemblies.

B. The reactor core shall contain 121 crucifo rn-shaped control rods. The control rod material shall be boron carbide powder (B,C)

• compacted to approximately 70% of theoretical density.

5.3 Recctor Vessel A. The pressure vessel shall be designed for a pressure of 1250 psig and a tecperature of 575 F.

The coolant recirculation system shall be designed for a pressure of 1148 psig on suction side of pump and 1248 psig at pump discharge. Both the pressure vessel and recirculation system shall be designed in accordance with the ASME Boiler and Pressure Vessel Code Sections Ill and IX.

3.4 Containment A. The primary containment shall be of the pressure suppression type having a drywell and an gbsorption j charber cons tructed of steel. The drywell shall have a volume of approximately 134,200 ft and is designed to confort to ASME Boiler and Pressure Vessel Code Section III Class B for an internal i

pressure of 56 psig at 281Fandanexternalpressureof2psjgat 281 F. The absorption chamber shall have a total volu=e of approximately 176,250 ft

!i s

Amendment No. 37 l, 19 0 ll l.

5.0

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