ML20245D294
| ML20245D294 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/14/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245D289 | List: |
| References | |
| NUDOCS 8904280213 | |
| Download: ML20245D294 (35) | |
Text
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UNITED STATES.
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NUCLEAR REGULATORY COMMISSION p
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WASHINGTON, D. C. 20655
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l NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.29 License No. DPR-21 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Northeast Nuclear Energy Company (the licensee), dated January 20, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reascnable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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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 2.C.(2) of Facility Operating License No. DPR-21 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
g, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Lub d j;fJohn F. Stolz, Director
/
Project Directorate I-4 Division of Reactor Projects I/II Office of Nuclear Reactor Reculation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 14, 1989
ATTACHMENT TO LICENSE AMENDMENT NO. 29 FACILITY OPERATING LICENSE NO. DPR-21 DOCKET NO. 50-245 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert i
i 11 11 iii 111 1-2 1-2 1-B 2-1 2-1 2-4 2-4 2-5 2-5 B 2-1 B 2-1 B 2-la B 2-2 B 2-2 B 2-4 8 2-4 B 2-6 8 2-6 B 2-7 B 2-7 B 2-8 B 2-8 B 2-11 B 2-11 3/4 3-3 3/4 3-3 3/4 6-13 3/4 6-13 3/4 11-1 3/4 11-1 3/4 11-2 J
3/4 11-3 3/4 11 4 3/4 11-5 3/4 11-2 3/4 11-6 3/4 11-3 3/4 11-7 3/4 11-8 3/4 11-4 8 3/4 2-4 B 3/4 2-4 B 3/4 3-4 8 3/4 3-4 B 3/4 3-5 B 3/4 3-5 B 3/4 6-6 B 2/4 6-6 B 3/4 11-1 B 3/4 11-1 B 3/4 11-la B 3/4 11-2 B 3/4 11-2 B 3/4 11-2a 6-19 6-19 6-19a
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. TABLE OF CONTENTS Paae No.
1.C DEFINITIONS...............................................
1-1 SAFFTY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.1 FUEL CLADDING INTEGRITY.............
2.1.2.................
2-1 1
2.2.1 REACTOR COOLANT SYSTEM..............
2.2.2................
2-7 i
l LIMIT 1NG CONDITIONS FOR OPERATION SURVEILLANCE RE001REMENT 3.0 GENERAL..................................... 4.0.........
3/4 0-1 3.1 REACTOR PROTECTION SYSTEM................... 4.1.........
3/4 1-1 3.2 PROTECTIVE INSTRUMENTATION.................. 4.2.........
3/4 2-1 A. Primary Containment Isolation functions.................
3/4 2-1 B. Emercency Core Cooling Subsystems Actuation.............
3/4 2-1 C. Control Rod Bl ock Actuation....................
3/4 2-1 D. Air Ejector Off-Gas System...............................
3/4 2-12 E.' Reactor Building Ventilation, Steam Tunnel Ventilation Isolation, and Standby Gas Treatment System Initiation......................................
3/4 2-1 3.3 REACTIVITY CONTROL.......................... 4.3.........
3/4 3-1 A. Reactivity Limitations..................... A 3/4 3-1 B. Control Rod Withdrawal......................
B 3/43-3
\\
C. Scram Insertion Times...................... C 3/4 3-5 D. Control Rod Accumulators................... D 3/4 3-7 E. Reactivity Anomalies.......................
E 3/4 3 8 F. Shutdown Requirements...................................
3/4 3-8 G. Thermal Power - Core Flow...............................
3/4 3 8 3.4 STANDBY LIQUID CONTROL SYSTEM................ 4.4........
3/4 4-1 A. No rmal Ope r at i on........................ A 3/4'4-1 B. Operation with Inoperable Components....................
3/4 4-3 C. Bo ror. Requi rement s..................... C 3/4 4-4 D. Shutdown Requirement....................................
3/4 4-4 l
i Abendment No. J, 29
Surveillance Pace No.
3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 3/4 5-1 A. Core Spray and LPCI Subsystems.............. A......
3/4 5-1 B. Containment Cooling Subsystems............... B......
3/4 5-3 C. FWC I S u b sy s t em.............................. C......
3/4 5-5 D. Automatic Pressure Relief Subsystems........ D......
3/4 5-6 E. Isolation Condenser System.................. E......
3/4 5-7 F. Minimum Core and Containment Cooling System Availability....................... F......
3/4 5-8 3.6 PRIMARY SYSTEM BOUNDARY 4.6 3/4 6-1 A. T h e rm al L i mi t a t i o n s......................... A......
3/4 6-1 B. Pressurization Temperature.................. B......
3/4 6-2 C. C o o l a n t C h e mi s t ry........................... C......
3/4 6-5 D. C o ol a n t L e a k a g e.............................. D......
3/4 6-11 E. Safety and Relief Val ves..................... E......
3/4 6-11 F. Structural Integrity........................ F......
3/4 6-12 G.JetPumps...................................G......
3/4 6-13 H. Recircul at ion Pump Flow Mismatch............ H......
3/4 6-14
- 1. Snubbers...................................
1......
3/4 6-15 J. Condensate Demineralizers................... J......
3/4 6-18 K. Mechanical Condenser Vacuum Pump............ K......
3/4 6-18 3.7 CONTAINMENT SYSTEMS 4.7 3/4 7-1 A. Primary Containment......................... A......
3/4 7-1.
B. Standby Gas Treatment System................ B......
3/4 7-10 C. Secondary Containment....................... C......
3/4 7-13 D. Primary Containment Isol ation Valves........ D......
3/4 7-14 3.8 RADI0 ACTIVE MATERIALS 4.8 3/4 8-1 A. Radioactive Liquid Effluent l~ instrumentation. A......
3/4 8 1 B. Radioactive Gaseous Effluent Instrumentation. B......
3/4 8-6 C. Radioact ive Liquid Effluents................ C......
3/4 8-12 D. Radioactive Gasious Effluents................ D......
3/4 8-14 3.g AUXILIARY ELECTRICAL SYSTEM 4.9 3/4 9-1 3.10 REFUELING 4.10 3/4 10-1 i
A. Re f uel i ng Inte rl ocks........................ A......
3/4 10-1 B. C o re Mon i t or i ng............................. B......
3/4 10-2 l
C. Fuel Storage Pool Water Level............... C......
3/4 10-3 D. C r a n e Op e ra bil i ty........................... D......
3/4 10-3 E. Crane Interlocks and Switches................ E......
3/4 10-3 ii Amendment No. J, 29
p j;
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Surveillance Pace No.
3.1]
REACTOR. FUEL ASSEMBLY 4.11-3/4 11-1 l
l
-A Average Planar. Linear Heat Generation Rate (APLHGR)................... A..... 3/4 11-1 B. Linear Heat Generation Rate (LHGR)........... B..... 3/4 11-2 C. _ Minimum Critical Power Ratio (MCPR).......... C..... 3/4 11-3 3.12 FIRE PROTECTION SiSTEMS
'3/4 12-1 l
A. Fire Suppression Water System...............'A..... 3/4 12 l B. Spray and/or Sprinkler Systems............... B..... 3/4 12-5 '
l C. Carbon Dioxide and Halon 1301 Systems........C'.....
3/4 12-7 D. Fi re Hose Stat i ons.......................... D..... 3/4 12-9 E. Fire Protection. Instrumentation............. E....... 3/4 12-13 F. Penetration Fire Barriers................... F...... 3/4 12-17 3.13 INSERVICE. INSPECTION 4.13 3/4 13-1 3.14 PLANT SYSTEMS 4.14 3/4 14-1 E.0 DESIGN FEATURES 5-1 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Responsibility.................................. 6-1 6.2 Organization (Offsite and Onsite)................
F '
6.3.
Uni t Staff Quali fications....................... t 6.4 1 raining....................................... 6-6 6.5 Review and Audit............................... 6-6 l
6.6 Reportable Event Action........................ 6-15 1
6.7 Safety Limit Violation......................... 6-15 6.8 Procedures..................................... 6-16 6.9 Reporting Requirements......................... 6-17 6.10 Record Retention............................... 6-19a I
6.11 Radiation Protection Program................... 6-20 6.12 High Radiation Area............................ 6-21 6.13 Systems Integrity.............................. 6-21 6.14 lodine Monitoring.............................. 6-22 4
6.15 Radiological Effluent Monitoring and Offsite Dose Calculation Manual........... 6-22 6.16 Radioactive Waste Treatment.................... 6-22 l
ili Amendment No. J, 29 i
l
1.0 H.
M,nimum Critical Power Ratio (MCPR) i Minimum Critical Power Ratio (MCPR) is tho value of critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated by application of an NRC approved critical power correlation to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.
1.
Mode The reactor mode is that which is established by the mode-selector-switch.
J.
Operable - Operability A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).
Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling er seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
K.
Oreratinc Operating means that a system or component is performing its intend-ed function in its required manner.
L.
Oreratina (vcle Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
M.
Fraction of Limitina Power Density The ratio of the linear heat generation rate (LHGR) existing at a given location to the maximum allowable LHGR for that bundle type.
l Maximum Fraction of Limitino Power Density The Maximum Fraction of Limiting Power Density (MFLPD) is the highest value existing in the core of the Fraction of Limiting Power Density (FLPD).
N.
Primary Containment inteority Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied.
Millstone Unit I l-2 Amendment No. p 29
r l
W.
Sucolemental Rebad Licensino Submittal The SUPPLEMENTAL RELOAD LICENSING SUBMITTAL (SRLS), its supplements and revisions, are unit and cycle specific document (s) containing the power' distribution limits (MCPR, LHGR, and APLHGR) for the.
current operating cycle. The limits in the SRLS, including supple-ments and revisions, are only applicable during the reload / cycle number (s) given in the title. These cycle specific limits shall be determined for each reload cycle in accordance wit,h Specification 6.9.1.9.
PF.
Averaae Flanar Linear Heat Generation Rate (APLHGR)
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be i
applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod for all the j
fuel rods in the specified bundle at the specified height, divided by the number of fuel rods in the fuel bundle at that height.
l 0;.
Linear Heat Generation Rate (LHGR)
The LINEAR HEAT GENERATION RATE (LHGR) shall be applicable to a specific rod at a specific height and is equal to the heat genera-tion rate per unit length for the specific rod at that specific height.
fii11 stone Unit I l-8
,StrETY LIMITS 2.1.1 FUEL CLADDING INTEGRITY Annlicability:
l Applies to the interrelated variables associated with fuel thermal behavior.
Obiettive:
To establish limits below which the integrity of the fuel cladding is pre-served.
Specification:
A.
When the reactor pressure is greater than 800 psia and the core flow is greater than 10% of rated design, a minimum critical power ratio (MCPR) less than 1.04 shall constitute a violation of the fuel cladding integrity I
safety limit.
LIMITING SAFETY SYSTEM SETTINGS 2.1.2 FUEL CLADDING INTEGRITY Arrlicability:
Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
Obiective:
To define the level of the process variables at which automatic protective action is initiated to prevent the safety limits from being exceeded.
Specification:
The limiting safety system settings shall be as specified below:
A.
Neutron Flux Scrl 1.
APRM Fluy Scram Trio Settino (Run Mode) a.
When the Mode Switch is in the RUN position, the APRM flux scram trip setting shall be as shown on Figure 2.1.2 and shall be:
S <0.58 W + 62 Millstone Unit 1 2-1 Amendment No. J, 29
u.
1 l
L il'J T D a SAFETY SYSTEM SETTINGS (Continued) 2.2.2 FUEL CLADDING INTEGRITY A.I.b.
MFLPD =
maximum fraction of limiting.
power density.
l The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
c.
During power ascensions with power levels less than or equal to 90%, APRM Flux Scram Trip Setting adjust-ments may be made as described below, provided that the change in scram settirg adjustment is less than 10% and, a notice of the adjustment is posted on the reactor control panel.
The APRM meter indication is adjusted by:
ARPRM -
MFLPD p
FRP where:
APRM = APRM Meter Indication P
= % Core Thermal Power For no combination of loop recirculation flow rate and' core thermal power shall the APRM flux scram trip setting.
be allowed to exceed 120% of RATED THERMAL POWER.
2.
APRM Reduced Flux Trio Settina (Refuel or Startuo/ Hot Standby Pode)
When the mode switch is in the REFUEL or STARTUP/ HOT STANDBY position, the APRM scram shall be setdown to less than or equal to 15% of RATED THERMAL POWER. The IRM scram trip setting shall not exceed 120/125 of full scale.
l 1
thlistone Unit 1 2-4 Amendment No. E, 29
{..
4
. ':T1la SAFETY SYSTEM SETTINGS (Continued) 2 -. l. 2 FUEL CLADDING INTEGRITY B.
1.
APRM Rod Block Trio Settina a.
The APRM rod block trip setting shall be as shown in Figure 2.1.2 and shall be:
(Run Mode)
SRB s 0.58W + 50 where:
S RB Rod block setting in percent of rated thermal power
=
(2011 MWt).
W Total recirculation flow in percent of design.
=
(Note 1, Page 2-3).
b.
In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP),
the setting shall be modified as follows:
Sgg 1 (0.58W + 50)
FRP fraction of rated thermal power (2011 MWt)
=
MFLPD =
maximum fraction of limiting power density.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
During power ascensions with power levels less than or equal to 90k c.
APRM Rod Block Trip Setting adjustments may be made as described below, provided that the change in scram setting adjustment is less than 10% and a notice of the adjustment is posted on the reactor control panel:
tiil l st ont Unit 1 2-5 Amendment No. E, 29
l y.
l 2.1.1 FUEL CLADDING INTEGRITY E:SES The fuel cladding integrity limit is set such that no calculated fuel damage would o'. cur as a result of an abnormal operational transient. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the minimum critical power ratio (MCPR) is no less than 1.04.
MCPR equal to or greater than 1.04 represents a conser-l vative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signif-icantly above design conditions and the protection system safety settings.
While fission product migration from cladding-perforation is just as measur-able as that from use related cracking, the thermally caused cladding perfora-tions signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel claddino Safety Limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation.
Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boiling transition, is not a directly observable parameter 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 distribu-tion. The cargin for each fuel assembly is characterized by the critical power ratic (CPE) which is the ratio of the bundle power which would produce onset ef transiticn boiling to the actual bundle power. The minimum value of this ratio for any bunole in the core is the minimum critical power ratio (MCPR).
It is Ossumed that plant operation is controlled to the nominal protective setpoints via the instrumented variable, i.e.,- normal plant operation present-ed on Figure 2.1.2 by the nominal expected flow control line. The Safety Limit (MCPR of 1.04) has sufficient conservatism to assure that in the event l
of an abnormal operational transient, initiated from a normal operating condition, 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 (MCPR = 1.04) is derived from a detailed statis-l tical analysis considering all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as described in Reference 1.
1.
NEDO-l%M. General Electric BWR Thermal Analysis Bases (GETAB) Data Correlation and Desion Acolication.
Millstone Unit ]
B 2.]
Amendment No. 29
'O 2.1.1.
FUEL CLADDING INTEGRITY EASEF Generic approval for the MCPR safety limit of 1.04 is given in Reference 2.
This reference states that the MCPR safety limit of 1.04 may be generically ap;1ied to D-lattice plants with a core which is operated with two successive reloads of high bundle initial R-factor GE B/P8 x 8R and/or GE8 x BEB fuel.
tii11 stone Unit No. I meets these criteria.
i i
l i
I 2.
NRC letter A.C. Thadani to S. S. Charnley, Safety Evaluation Report for Amendment 14 to GE Licensing Topical Report NEDE-24011-P. A., dated December 27, 1957.
l Amendment No. 29 Pd11 stone Unit 1 B 2-la
r 2.1.1 FUEL CLADDING INTEGRITY EASES Because the b)iling Mansition correlation is based on a large quantity of full scale date', there is a very high confidence that operation of a fuel assembly at the cundition of MCPR = 1.04 would not produce boiling transition.
l However, i' boiling transition were to occur, clad perforation would not be expected. 1,ladding temperatures would increase to approximately 1100*F which is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Test Reactor (GETR) where fuel similar in design to Millstone operated above the critical hea.t. flux for a significant period of time (30 minutes) without clad perforation. Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity. The limit of applicability of the boiling transition correlation is 1400 psia during normal power operation. However, the reactor pressure is limited as per Specification 2.2.1.
1 The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density and reactor core thermal power. The scram setting is adjusted I
in accordance with the formula in Specification 2.1.2.A.1, when the maximum fraction of limiting power density is greater than the fraction of rated power. ]f the APRM scram setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain thus reducing l
the slope and intercept point of the flow referenced scram curve by the i
reciprocal of the APRM gain change.
At pressures below 800 psia, the core evaluation pressure drop (0 power, O flow) is greater than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows yill always be greater than 4.56 psi.
Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Tgus, I
the bundle flow with a 4.56 psi driving head will be greater than 28 x 10 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical l
power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a core thermal power of more than 50%. Thus, a 1
core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.
Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 2.1.lA or 2...lB will not be exceeded. Scram times are checked periodically to assure i
the insertion times are adequate. The thermal power transient resulting when i
a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux fc11owing closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit Millstore Unit 1 B 2-2 Amendment No. 29 l
i j
)
2..
2 FUEL CLADDING INTEGRITY BASES The transients expected during operation of Millstone Unit I have been analyzed up to the thermal power condition of 2011 MWt.
The analyses were based upon gnt operation in accordance with the operating map given in Figure 3.3.1 In addition, 2011 MWt is the licensed maximum steady-state power level of Millstone Unit 1.
This maximum steady-state power will never be knowingly exceeded.
Conservatism was incorporated by conservatively estimating the control-ling factors such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, axial power shapes, etc. These factors are all selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for the evaluation of reactor dynamics performance. Compar-isons have been made showing results obtained from a General Electric boiling water reactor and the predictions ma The comparisons and results are summarized in Ref erence 2.g) by the model.
ine void reactivity coefficient utilized in the analysis is conser-vatively estimated to be about 25% larger than the most negative value expect-e: te occur during the core lifetime.
The scram worth used has been derated to be equivalent to the scram worth of about 80% of the control rods.
The scram delay time and rate of rod insertion are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifica-tions. 'he effect of scram worth, scram delay time and rod insertion rate, all conserv, tively applied, are of greatest significance in the early portion of the necitive reactivity insertion. The rapid insertion of negative reactivity strong { turns the transient and the stated 5% and 20% insertion times conser-vativeb accomplished this desired initial effect.
The time for 50% and 90%
inserti'n are given to assure proper completion of the insertion stroke, to further assure the expected performance in the earlier portion of the tran-sient, and to establish the ultimate fully shutdown steady-state condition.
For analyses of the thermal cons-equences of the tansients, MCPRs speci-fied in the Supplemental Reload Licensing Submittal are conservatively assumed l
to exist prior to initiation of the transients.
i 1
(1)
" Extended Load Line Limit Analysis, Millstone Point Nuclear Power Station, Unit 1", NEDO-24366 and NEDO 24366-1. (See TS Page B2-5)
(2)
- Linford, E.S.,
" Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor,* NED0-10802.
Millstone Unit 1 B 2-4 Amendment No. 29 i
l' 2.1.2 FUEL CLADDING INTEGRITY USES Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.04 when the transient is initiated from-MCPR's specified in the Supplemental Reload Licensing Submittal.
In order to assure adequate core margin during full load rejections in the event of failure of the select rod insert, it is necessary to reduce the APRM scram trip setting to 90i; of rated power following a full load rejection incident. This is necessary because, in the event of failure of the select rod insert to function, the cold feedwater would slowly increase the reactor power level to the scram trip setpoint. A trip setpoint of 905; of rated has been established to provide substantial margin during such an occurrence. The trip setdown is delayed to prevent scram during the initial portion of the transient. The specified maximum setdown delay of 30 seconds is conservative because the cold feedwater transient does not produce significant increases in reactor power before approximately 60 seconds following the load rejection.
For operP. ion in the REFUEL or STARTUP/ HOT STANDBY modes while the reactor is at 'ow pressure, the APRM reduced flux trip scram setting of
< 355; of rated power provides adequate thermal margin between the maximum power and the safety limit, 259; of rated power. The margin is adequate te accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are
. minor. Cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are small and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant porer rise.
In an assumed uniform rod withdrawal approach to the scram level, the APRM system would be more than adequate to assure a scram before the '
power could exceed the safety limit. The APRM reduced trip scram remains active until the mode switch is placed in the RUN position. This switch occurs when the reactor pressure is greater than 880 psig.
The IRM trip at < 120/125 of full scale remains as a backup feature.
The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.
During steady-state operation, with one recirculation pump operating, the equalizer line shall be isolated. Analyses of transients from this operating condition are less severe than the same transients from the two pump operation.
f'illstcr:e Ur.it ]
E26 Amendment No. 1, 29
2.1.2 FUEL CLADDING INTEGRITY USES E.
APRM Control Rod Block Trios Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod bicek to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus to protect against a condition of MCPR <
1.04.
This rod block setpoint, which is automatically varied with l
recirculation flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The specified flow variable setpoint provides substantial margin from fuel damage, assuming steady-state operation at the setpoint, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship. The l
actual power distribution in the core is established by specified control rod sequences and is monitored continuously 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 2.1.2.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, thus reducing the slope and intercept point of the flow referenced rod block curve by the reciprocal of the APRM gain change.
The APRM rod block setpoint is reduced to < 12% of rated thermal power with the mode switch in REFUEL or STARTUP/ ROT STANDBY position.
C.
Reactor low Water level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained.
D.
Reactor low Low Water level ECCS Initiation Trio Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the decay heat associated with the loss-of-coolant accident; to limit fuel clad temperature to well below the clad melting temperature; to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%. To accomplish this function, the capacity of each emergency core cooling system component was established based on the reactor low low water level. To lower the setpoint of the low water level scram would require an increase in the capacity of each of the ECCS components. Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will r.ot be set lower because of ECCS capacity requirements.
l Millstone Unit 1 B 2-7 Amendment No. J, 29 I
i
2. L. 2 FUEL CLADDING INTEGRITY E SES The design of the ECCS components to meet the above criteria was depen-dent on three previously set parameters: the maximum break size, the low water level scram setpoint, and the ECCS initiation setpoint. To lower the setpoint for initiation of the ECCS would prevent the ECCS components from meeting their design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or wn:,5 normally expected transients.
E.
Turbine Stoo Valve Scram The turbine stop valve scram, like the load rejection scram, anticipates the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valves and failure of the bypass. With a scram setting of < 10% of valve closure, the resultant increase in surface heat flux is limited such that MCPR remains above 1.04 even l
during the worst case transient that assumes the turbine bypass is closed.
This scram is bypassed when turbine steam flow is < 45% of rated, as measured by the turbine first stage pressure.
F.
Turbine Control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and-subse-quent failure of the bypass; i.e., it prevents MCPR from becoming less than 1.04 for this transient. For the load rejection from 100% power with operable bypass valves, the heat flux increases to only 106.5% of its rated power value, which results in only a small decrease in MCPR.
This trip is bypassed below a generator output of 307 MWe because, below this power level, the MCPR is greater than 1.04 throughout the transient l
without the scram, b order to accommodate the full load rejection capability, this scram trip must be bypassed because it would be actuated and would scram the reactor during load rejections. This trip is automatically bypassed for a maximum of 280 millisec following initiation of load rejection. After 280 milliset, the trip is bypassed providing the bypass valves have opened.
If the bypass valves have not opened after 280 millisec, the bypass is removed and the trip is returned to the active condition. This bypass does not adversely affect plant safety because the primary system pressure is within limits and MCPR remains above 1.04 during the worst l
transient even if the trip fails. There are many other trip functions which protect the system during such transients.
G.
Main Steam Line Isolation Valve Closure Scram The low pressure isolation of the main steam lines at 825 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the stram feature which occurs when the main steam line isolation valves are
- closed, I'.illstone Unit 1 B 2-8 Amendment No. J, 29 j.
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2.2.2 REACTOR COOLANT SYSTEM l
E'. E5
]n compliance with Section III of the ASME Boiler and Pressure Vessel Code, 1955 Edition, the specified settings of the pressure relieving devices are oelow 103% of design pressure. As described in the General Electric Topical Report, NEDE-240ll-P-A, General Electric Standard Application for Reactor fuel, the most severe isolation event with indirect scram has been evaluated.
1 The most severe isolation is the MSIV closure from steady-state operation at i
2011 MWt. The evaluation anures that the sizing and settings of the pressure relieving devices are adequate to assure that the peak allowable pressure of 110% of vessel design pressure is not exceeded.
Evaluations indicate that a total of six dual purpose safety / relief valves set at the specified pressures maintain the peak pressure during the transient well within the code allowable and safety limit pressure. Results are presented in the Supplemental Reload Licensing Submittal.
Millstone Ur.n ]
B 2-11 Amendment No.1, 29
L1!'! TING CONDITION FOR OPERATION 3.? RU:ilVITY CONTROL E.
Control Rod Withdrawal 1.
Each control rod shall be coupled to its drive or completely inserted and the control rod directional cont,ol valves disarmed electrically. However, for purposes of removal of a control rod drive, as many as one drive in each quadrant may be uncoupled from its control rod so long as the reactor is in the shutdown or. refuel condition and Specification 3.3.A.1 is met.
2.
The control rod drive housing support system shall be in place during power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met.
3.
Whenever the reactor is in the STARTUP or RUN mode below 20% rated thermal power, no control rods shall be moved unless the rod worth minimizer is operable or a second independent operator, or engineer, verifies that the operator at the reactor console is following the control rod program. The second operator may be used as a substitute fc an inoperable rod worth minimizer during a startup only if the rod worth minimizer fails after withdrawal of at least twelve control rods.
4.
Control rods shall nbt be withdrawn for startup or refueling unless at least two source range channels have an observed count rate equal to or greater than three' counts per second.
5.
During operation with limiting control rod patterns, as determined by the reacter engineer, either:
a.
Both RBM channels shall bo operable; or i
b.
Control rod withdrawal shall be blocked; or c.
The operating power level shall be limited so that the MCPR will remain equal to or greater than 1.04 assuming a single l
error that results in complete withdrawal of any single operable control rod.
j i
l'illstone Ur.it 1 3/4 3-3 Amendment No. J, 29
LIMITING CONDITION FOR OPERATION 3.E PT:t2;Y SYSTEM BOUNDARY G.
Jet Pumns 1.
Whenever the reactor is in the STARTUP/ HOT STANDBY or RUN modes, all jet pumps shall be intact and all operating jet pumps shall be operable.
If it is determined that a jet pump is inoperable, an orderly shutdown shall be initiated and the reactor shall be in a COLD SHUTDOWN or REFUEL CONDITION within.24 hourt.
2.
Flow indication from each of the twenty jet pumps shall be verified i
prior to initiation of reactor startup from a cold shutdown condition.
3.
The indicated core flow is the sum of the fl.v indication from each of the twenty jet pumps.
If flow indication failure occurs for two or more jet pumps, immediate corrective action shall be taken.
If flow indication cannot be obtained for at least nineteen jet pumps, an orderly shutdown shall be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor shall be in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURt.'EILL AN:E RE0UIREMENT 4.6 PRIMARY SYSTEM BOUNDARY G.
Jet Puros 1.
Whenever there is a recirculation flow with the reactor in the STARTUP/ HOT STANDBY or RUN modes, jet pump integrity and operability shall be checked daily by verifying that the following two conditions do not occur:
a.
The recirculation pump flow differs by more than'10% from the established speed-flow characteristics; or b.
The indicated total core flow is more than 10% greater than the core flow value derived from established power core l
flow relationships.
1 2.
Additionally, when operating with one recirculation pump with the equalizer valves closed, the diffuser to lower plenum differential pressure. shall be checked daily, and the differential pressure of any jet pump in the idle loop shall not vary by more than 10%
from established patterns.
3.
The baseline data required to evaluate the conditions in Specification 4.6 G.1 and 4.6.G.2 will be acquired each operating l
cycie.
Millstone Unit 1 3/4 6-13 Amendment No. J, 7, 29
l t
LIMITING CONDITION FOR OPERATION
^ '
3.11 REACTOR FUEL ASSEMBLY Anolicability l
The Limiting Conditions for Operation associated with the fuel rods apply to those parameters which monitor the fuel rod operating conditions.
i l
l-Oriective The Objective of the Limiting Conditions for Operation is to assure the performance cf the fuel rods.
Specifications A.
Averace Planar Linear Heat Generation Rate (APLHGR) 1.
During power operation, the APLHGR for each type of fuel as a function of axial location and average planar exposure shall not exceed the limits specified in the Supplemental Reload Licensing Submittal.
2.
If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR specified in the' Supplemental Reload Licensing Submittal is being exceeded, l
action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLHGR 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 />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
SURVEILLAN:E REQUIREMENT 4.11 REACTOR FUEL ASSEMBLY ADolic a bilit y The Surveillance Requirements apply to the parameters which monitor the fuel rod operating conditions.
Obiective The Objective of Surveillance Requirements is to specify the type and frequency of surveillance to be applied to the fuel rods.
Specifications A.
Averace Planar Linear Heat Generation Rate (APLHGR)
The APLHGR for each type of fuel, as a function of average planar exposure shall be determined daily during reactor operation at 1 25%
RATED THERMAL POWER.
l l
Millstone Ur.it 1 3/4 11-1 Amendment No. E, 29 i
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1
. LIMITING CONDITION FOR OPERATION
- .11 REACTOR FUEL ASSEMBLY E.
Linear Heat Generation Rate (LHGR)
During stady state power operation, the linear heat generation rate (LHGR) of any rod in any fuel assembly at any axial location shall not exceed the maximum allowable LHGR specified in the Supplemental Reload 1
Licensing Submittal.
During power operation, the LHGR shall not exceed the limiting value.
If at any time during operation it is determined, by normal surveillance, that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to COLD SHUTDOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
SURVEILLANCE RE0VIREMENTS 4.11 REACTOR FUEL ASSEMBLY B.
Linear Heat Generation Rate (LHGR)
The LHGR shall be checked daily during reactor operation at 25%
RATED THERMAL POWER.
i
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Pillstone Unit 1 3/4 11-2 Amendment No. E 29
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l LIMITING CONDITION FOR OPERATION l
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L.11 REACTO. FUEL ASSEMBLY C.
Minimum Critical Power Ratio (MCPR)
Duri:ig power operation,. the MCPR shall be equal to or greater than the MCPR limit specified in the Supplemental Reload Licensing
- Submittal multiplied by the K[, for core flows other than rated, where k is as shown in Figur 3.11.1.
g If at any time during operation it is determined by normal l
surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to 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 />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
I D.
If any of the limiting values identified in Specification 3.ll.A, B, or C, are exceeded, even if corrective action is taken, as prescribed, a Reportable Event report shall be submitted.
l SURVEILLANCE RE0UIREMENTS 4.11 REACTOR FUEL ASSEMBLY C.
Minimum Critical Power Ratio (MCPR) 1.
MCPR shall be determined daily during reactor power operation at > 25% RATED THERMAL POWER and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for specification 3.3.B.5.
l 2.
Utilization of Option B Operating limit MCPR values requires the scram time testing of 15 or more control rods on'a rotating basis every 120 operating days.
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2 3,2 PROTECTIVE INSTRUMENTAL 10t1 E SES The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to s 1.04.
The trip logic for I
tris function is I out of n; e.g., any trip on one of the six APRM's, eight IRM's, or four SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the IRM and REM may be reduced by one for a short period of time to allow for I
maintenance testing and calibration.
The APRM rod block trip is flow biased and provides gross core l
protection, i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that fuel damage limits are not exceeded.
The RBM provides local protection of the core, i.e., the prevention of fuei damage in a local region of the core, for a single rod withdrawal error.
The trip point is flow biased. The worst case single control rod withdrawal error has been analyzed for the initial core and also prior to each reload; the results show that, with specified trip settings, rod withdrawal is blocked within an adequate margin to fuel damage limits. This margin varies slightly from reload to reload and, thus, each reload submittal contains an update of the analysis.
Below -70% power, the withdrawal of a single control rod results in MCPR > 1.04 without rod block action, thus requiring the RBM system I
to be operable above 30% of rated power is conservative. Requiring at least half of the normal LPRM inputs from each level to be operable assures that the RE" response will be adequate to prevent rod withdrawal errors.
The IRM rod block functions assure proper upranging of the IRM system, and reduce the probability of spurious scrams during startup operations.
A downscale indication of an APRM or IRM is an indication the instrument has failed, the instrument is not sensitive enough, or the neutron flux is below the instrument response threshold.
In these cases the instrument will not respond to changes in control rod motion and thus control rod motion is prevented, The downscale trips are set at 3/125 of full scale.
To prevent excessive fuel clad temperature for the small pipe break, the FWCl or Isolation Condenser systems must function, since for these breaks reactor pressure does not decrease rapidly enough to allow either core spray or LpC1 to operate in time. The automatic pressure relief function and Isolation Condenser system are provided as back-ups to the FWCI in the event the FWCl does not operate. The arrangements of the tripping contacts are such as to provide these fur:tions when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criterion is met.
Ref. Section VI-2.0 FSAR.
The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration. and also minimizes the risk of inadvertent operation; i.e., only one instrumer*. thannel out of service.
Millstone Unit 1 B 3/4 2-4 Amendment No. 29
{t-l-
3.3 REACTIVITY CONTROL BASES The RWM provides automatic supervision to assure that out-of-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences.
Reference Section Vll.10 FSAR.
It serves as an independent backup of the normal withdrawal procedure followed by the operator.
In the event that the RWM is out of service when required, a second independent operator or engineer can manually fulfill the operator-follower control rod pattern conformance function of the RWM.
In this case, procedural control is exercised by verifying all control rod positions after the withdrawal of each group, prior to proceeding to the next group. Allowing substitution of a second independent operator or engineer in case of RWM inoperability recognizes the capability to adequately monitor proper rod sequencing in an alternate manner without unduly restricting plant operations. Above 20% power, there is no requirement that the RWM be operable since the control rod drop accident with out-of-sequence rods will result in a peak fuel energy content of less than 280 cal /gm.
To assure high RWM availability, the RWM is required to be operating during a startup for the withdrawal of a significant number of control rods for any startup.
4.
The Source Range Monitor (SRM) system performs no automatic safety system function; i.e., it has no scram function.
It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The requirement of at least 3 counts per second assures that adequate monitoring capability is available. One operable SRM channel would be adequate to monitor the approch to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's is provided as an acded conservatism.
5.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for i
maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.
This system backs up the operator who withdraws control rods according to a written sequence. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists. 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 a MCPR less than 1.04.
During use of such patterns, l
it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.
It is the responsibility of the Reactor Mill:. tone Unit ]
B 3/4 3-4 Amendment No. 29
3.3 REACT]VITY. CONTROL BASES 1
Engineer to identify these limiting patterns and the designated rods, either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other
{
than limiting patterns.
j C.
Scram insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than 1.04.
The limiting power transient is that resulting l
from a generator load rejation coircident with failure of the turbine bypass system. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification, provide the required 3
protection, and MCPR remains equal to or greater than 1.04.
Amendment 21 l
to the Provisional Operating License shows the control rod scram reactivity insertion data used in analyzing the transients. The limit on the number and pattern of rods permitted to have long scram times is specified to assure that the reactivity insertion rate effects of rods of long scram times are minimized. Grouping of long scram time rods is prevented by not allowing more than one control rod in any group of four control rods to have long insertion times. The minimum amount of reactivity to be inserted during a scram is controlled by permitting no more than 10% of the operable rods to have long scram times.
In the analytical treatment of the transient, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. Approximately the first 90 milliseconds of each of these time intervals result from the sensor and circuit delays; at this point, the pilot scram solenoid deenergizes. Approximately 120 milliseconds later, the control rod motion is estimated to actually begin. However, 200 milliseconds is conservatively assumed for this time interval in the transient analyses and this is also included in the allowable scram insertion times of Specification 3-3.C.
The time to deenergize the pilot valve scram solenoid is measured during the calibration tests required by Specification 4.1.
The scram times for all control rods will be determined at the time of each refueling outage. The weekly control rod exercise test serves as a periodic check against deterioration of the control rod system and also verifies the ability of the control rod drive to scram; since, if a rod can be moved with drive pressure, it will scram because of higher pressure applied during scram. The frequency of exercising the control rods under the conditions of three or more control rods out of service provides even further assurance of the reliability of the remaining control rods.
Millstor,( Unit 1 B 3/4 3-5 Amendment No. 29
' 3.6 and 4.6 PRIMARY SYSTEM B0UNDARY pr A break in' a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared with previous operation.
The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.
I The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Sections 4.6.G.1 and 2.
Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps. This bypass flow is reverse with respect to normal jet pump flow. The indicated total core flow is a summation of the flow indications twenty individual jet pumps. The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow. Thus, the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump. Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during a shutdown period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.
A nozzle riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any sut,stantial stress in the jet pump body makes failure impossible.
H.
Recirculation Pumo Flow Mismatch The LPCI loop selection logic is described in the FSAR, Section 6.2.4.2.
For some limited, low probability accidents, with the recirculation loop operating with.orge speed differences, it is possible for the logic to
{
select the wrong loop for injection.
For these limited conditions the core spray itself is adequate to prevent l
fuel temperatures from exceeding allowable limits. However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.
The analyses for Quad Cities indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 1 55..
Below 80% power the loop select logic would not be expected to function at a speed differential of 20%. This specification provides a margin of 5!, in pump speed differential before a problem could arise.
If the reactor is operating on one pump, the loop select logic trips tnat pump before making the loop selection.
Millstone Unit 1 B 3/4 6-6 Amendment No. 7, 29 l
3.11 REACTOR FUEL ASSEMBLY BASES A.
Averaae Planar Linear Heat Generation Rate (APLHGR) l This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will I
not exceed the limits specified in 10CFR50.46 and that the fuel design I
analysis limits specified in NEDE-240ll-P-A (Reference 1) will not be l
exceeded.
g Mechanical Design Analysis: NRC approved methods (specified in Reference
- 1) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 1.
No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis APLHGR limit.
LOCA Analysis: A LOCA analysis is performed in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power (APLHGR) limits comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.
The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of toe average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than i 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit.
j
. Conservative LOCA calculations predict that nucleate boiling will be maintained for several seconds following a design basis LOCA. This results in early removal of significant amounts of stored energy which, if present later in the transient, when heat transfer coefficients are considerably lower, would result in higher peak cladding temperature. As core flow is reduced below about 90%, the time of onset of boiling transition makes a sudden change"from greater than about 5 seconds to less than I second. The approved ECCS evaluation model requires that at the first onset of local boiling transition, the severely reduced heat transfer coefficients must be applied to the affected planar area of the bundle, and thus exaggerates the calculated peak clad temperature. The effect is to significantly reduce the energy calculated to be removed from the fuel during blowdown. This results in an increase in calculated peak clad temperature of about 100'F which can be offset by a 5% reduction in MAPLHGR.
For flows less than 90% of rated, a 5% reduction in the MAPLHGR limits derived for 100% flow will assure that the plant is operated in l
compliance to 10 CFR 50.46 at those lower flows.
(1) NEDE-240ll-P. A., " General Electric Standard Application for Reactor Fuel," l latest approval version.
Millstone Unit 1 E,3/4 11-1 Amendment No. 29
. 3.11 REACTOR FUEL' ASSEMBLY E'SES The limiting values for APLHGR are specified in the Supplemental Reload Licensing Submittal.
E.
Linear Heat Generati n Rate (LHGR)
Q This specification assures that the linear heat gener6 tion rate in any rod is less than tha design linear heat generation rate. The LHGR shall be checked daily during reactor operation at > 25% power to determine if fuel burnup, or control rod movement, has caused changes in power distri-bution.
For LHGR to be a limiting value below 25% RATED THERMAL POWER, the MTPF would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.
I l
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Millstone Unit 1 B 3/4 11-la Amendment No. 29 L
---_____-----_--___m
1 2.11 REACTOR FUEL ASSEMBLY EASES 1
C.
The required operating limit MCPR's at steady state operating conditions i
are derived from an established fuel cladding integrity Safety Limit MCPR approved by the NRC and an analysis of abnormal operational transients.
For any abnormal operating transient analysis evaluation with the initial condition cf the reactor being at the steady state operating limit, it is l
required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient.
The r.teady state value for MCPR was selected to provide a margin to accomodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself.
This value ensures that:
1.
For the initial conditions of the LOCA analysis, a MCPR of 1.18 is satisfied.
For the low flow ECCS analysis, an initial MCPR of 1.24 is assumed, and 2.
For any of the special transients, or disturbances, caused by single operator error or single equipment malfunction the value of MCPR is conservatively assumed to exist prior to the initiation of the transient or disturbance.
To assun that the fuel cladding integrity Safety Limit is not exceeded I
during any anticipated abnormal operational _ transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR). The type of transients eval-uated are defined in the Supplemental Reload Licensing Submittal.
l At core thermal power levels 125%, the reactor will be operating at minimum recirculation pump speed, and moderator void content will be very small.
Fo? all designated control rod patterns which may be employed at this power, thermal hydraulic analysis indicates that the resultant MCPR value is in excess of requirements. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculation of MCPR at greater than 25% RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The use of the Option B operating limit MCPR requires additional SCRAM time testing and verification as described in the letter from R. C. Tedesco (NRC) to G. G. Sherwood (GE) " Acceptance for Referencing General Electric Licensing Topical Report NED0-24154/NEDE-24154P,"
February 4, 1981.
I Details on how evaluations are performed, on the methods used, and how the MCPR limit is adjusted for operation at less than than rated power and flow conditions are given in Reference 1.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a therm 61 limit.
Millstone Unit 1 B 3/4 11-2 Amendment No. E. 29 i
i e
3.11 REACTOR FUEL ASSEMBLY EtsiS D.
Reportino Requirements The LCO's associated with monitoring the fuel rod operating conditions are required to be met at all times or corrected to within the limiting values of MAPLHGR, LHGR, and MCPR within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the time the plant is determined to be exceeding them.
It is a requirement, as staied in Specifications 3.11.A, B, and C, that if at any time during power opera-tion it is determined that the limiting values for MAPLHGR, LHGR, or MCPR are exceeded, action is then initiated to restore operation to within the prescribed limits. This action is to be initiated within 15 minutes if normal surveillance indicates that an operating limit has been reached.
Each event involving operation beyond a specified limit shall be logged and a reportable occurrence issued.
It must be recognized that there is always an action which would return any of the parameters (MAPLHGR, LHGR, or MCPR) to within prescribed limits, namely power reduction. Under most circumstances, this will not be the only alternative, l
Millstone Unit 1 B 3/4 Il-2a Amendment No. A 29 j
1 ADMINISTRATIVE CONTROLS j
i The report shall include that information delineated in the REMODCM.
Any changes to the REMODCM shall be submitted in the Semiannual Radioactive Effluent Relcase Report.
Supplemental Reload Licensina Submittal 6,9.1.9 Core power distribution limits (MCPR, LHGR, and APLHGR) shall be established and documented in the Supplement Reload Licensing Submittal before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine these limits shall be those previously reviewed and approved by the NRC in NEDE-24011, General Electric Standard Application for Reactor Fuel, latest approved revision.
The core power distribution limits shall be' determined so that all applicable limits
.(e.g.,
fuel thermal-mechanical limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analyses limits) of the safety analysis are met.
The Supplemental Reload Licensing Submittal, including any mid-cycle revisions or supplements thereto, l
shall be submitted upon issuance.
Special Reports l
E.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission Document Control Desk, Washington, D.C.
20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report.
These reports shall be submitted covering the activities identified below pursuant to the requirement of the applicable reference specification:
a.
In-service Inspection Results, Specification 4.6.F.
b.
Primary Containment Leak Rate Test
- Results, Specifica-tion 4.7.A.3.
c.
Materials Radiation Surveillance Specimen Examination and Results, Specification 4.6.B.3.
d.
Fire detection instrumentation, Specification (3.12.E.2) e.
Fire suppression systems, Specifications (3.12. A.2, 3.12.B.2, and 3.12.C.2).
f.
Radiological Effluent Reports required by Specifications in 3.8.C.2, 3.8.D.2, 3.8.0.3, and 3.8.D.4.
Millstone Unit 1 6-19 Amendment No. J, 27, 29
l ADM]NISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
- a. Records and logs of facility operation covering time interval at l
each power level.
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety,
- c. All REPORTABLE EVENTS.
- d. Records of surveillance activities, inspections and calibrations required by these technical specifications.
- e. Records of reactor tests and experiments.
- f. Records of changes made to operating procedures.
- g. Records of radioactive shipments.
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