ML19296B221
| ML19296B221 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/23/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19296B219 | List: |
| References | |
| NUDOCS 8002200285 | |
| Download: ML19296B221 (17) | |
Text
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[pa arog UNITED STATES ec g
NUCLEAR REGULATORY COMMISSION g.'
- C WASHINGTON, D. C. 20555 t;m+....f J
CONNECT (CUT LIGHT AND POWER C0tiPANY THE HARTFORD ELECTRIC LIGHT COMPANY WESTERN MASSACHUSETTS ELECTRIC COMPANY NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL OPEPATING LICENSE Amendment No. 66 License No. DPR-21 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Connecticut Light and Power Company, The Hartford Electric Light Comoany, Western Massachusetts Electric Company, Northeast Nuclear Energy Company (the licensees) dated April 9,1979, as supplenented September 24, 1979, complies with the standards and require-ments 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 reasonable assurance (i) that the activities autho-rized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such ac-tivities 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 M the Comission's regulations and all applicable requireran",s have been satisfied.
8002200 g
. 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 Operating License No.
DPR-21 is hereby amended to read as follows:
B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendments No. 66, are hereby incorporated in the license.
Northeast Nuclear Energy Company shall operate the facility in accor-dance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
' d.
Wo m Dennis L. Ziema
, Chief Operating Reactors Branch #2 Division of Ooerating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: Janua ry 23, 1980 9
ATTACHMENT TO LICENSE AMENDMENT NO. 66 PROVISIONAL 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 nmnber and contain vertical lines indicatir.g the area of change.
PAGES 2-1 2-3 2-4 2-5 2-6 2-7 2-8 B 2-2 B 2-6 B 2-7 Overleaf pages 2-2, B 2-1, B2-5, and B 2-8 are included for document completeness.
There is no change on this page; the provisions have merely been repositioned.
Cnly involves a page number change.
SAFELY LittITS LlillTitlG SAFETY SYSTEll SETTitlGS 2.1.1 futL CLADI)lt4G IllTEGRITY 2.1.2 fuel CLADDitlG IllTEGRITY Applicabil i ty:
Applicability:
Applies to tlie interrelated variables associated Applies to trip settings of the instruments and with fuel thennal behavior.
devices which are provided to prevent the reactor system safety limits from being Ghjective:
exceeded.
To establish limits below which the integrity of Objective:
the fuel cladding is preserved.
To define the level of the process variables at Specification:
which automatic protective action is initiated to prevent the safety limits from being exceeded.
A.
When the reactor pressure is greater than 800 psia and the core flow is greater than 10% of Specification:
rated design, a minimum critical power ratio (MCPR) less thanl.07 shall constitute a The limiting safety system settings shall be as violation of the fuel cladding integrity specif.ed below:
safety limit.
A.
Heutron Flux Scram B.
When the reactor pressure is less than or equal to 800 psia or reactor flow is less than 10%
1.
ApHM Flux Scram of design, the reactor thermal power transferred Trip Setting Run Mode) to the coolant shall not exceed 25% of rated.
a.
When the Mode Switch is in the l
C.
1.
To assure that the Limiting Safety System Rufi position, the APRM flux Settings established in Specifications scram trip setting shall be as 2.1.2A and 2.1.2 Bare not exceeded, cach shown on Figure 2.1.2 and shall required scram shall be initiated by its be:
prinury source sigml.
The Safety Limit shall be assumed to be exceeded when scram S < 0.65 H + 55%
is accomplishea by a means other than the Primary Source Signal.
Amendment tio. d, Jp, 66 2-1
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SAlLlY LIMITS LIMilitiG SAFETY SYSTD1 SETTINGS for no combination et loop recircula-tion flow rate and core thennal power shall the APRM flux scram trip setting be allowed to exceco 120% of rated thermal power.
2.
APRM Reduced Flux Trip Setting Thefuel or Startup/li5t Stantlby~ Mode),
When the mode switch is in the refuel or Start Up/ilot Standby position, the APRM scram shall be seldown to less than or equal to 15% of rated thennal power. lhe IRM scram trip setting shall not exceed 120/125 of full scale.
B.
1.
APRM Rod Block Trip Setting a.
The APRM rod block trip setting l
shall be as shown in Figure 2.1.2 and shall be:
(Run Mode)
SRn< 0.65 W 6 42%
where:
SRB = Rod block setting in percent of rated thennal power (2011 Mut).
W
= Total recirculation flow in percent of design (29.7 x 10 lixn/hr) b.
In the event of operation with a maximum total peaking factor (MTPF) greater than the design value of A, the setting shall be modified as follows, except l
Amendment No.,J$. 3;, 66 2-5
SAFETY LIMITS LIMITitlG SAFETY SYSTEM SETTitlGS I
as specified in Paragrapli 2.1.2.B.1.c:
Sgg <_ (0.65 W + 42%)
pp where:
A = 3.08 for 7x7 fuel
= 3.04 for 8x8 fuel
= 3.00 for 8x8R fuel MTPF = The value of the existing IllaXillium total peaking factor, c.
During power ascensions with power levels less than or equal to 90% APRM Rod Block Trip Setting adjustments kly be made as described below, provided that the change in scram setting adjust-ment is less than 10% and a notice of the adjustment is posted on the reactor control panel:
The APRM meter indicalion is adjusted by:
APRM =
P where:
APRM = APRM Meter Indication P
= % Core Thennal Power Amendment tio. )6', 66 26
SAFETY LIMIT LIHliif1G SAFETY SYSTEM SET 11tlGS B.
'2.
Ilie ApHM rod block trip setting for the refuel and startup/ hot stantiby mode,shall be less than or equal to 12% rated thernal power.
C.
The reactor Low Water Level Scram trip' setting shall be greater than or equal to 12/ inches above the top of the active fuel.
D.
The Reactor Low Low Water Level ECCS Initiation trip point shall not be greater than 83 inches nor less than 79 inches.
E.
The turbine Stop Valve Scram trip setting shall be less than or equal to ten percent valve closure from full open.
F.
The Turbine Control Valve Fast Closure Scram shall trip upon actuation of the accelera-tion relay in conjunction with failure of selected bypass valves to start opening within 260 milliseconds.
The noximum setting of the time delay relays which bypass this scram shall be 260 milliseconds.
G.
The Main Steam Isolation Valve Closure Scram trip settings shall be less than or equal to ten percent valve closure from full open.
II.
The Main Steam Line Low pressure trip which initiates main steam line isolation valve closure shall be greater than or equal to 880 psig.
2-7 Amendment flo. M, 34, 66
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2.1.1 Bases The fuel cladding integrity limit is set such that no calculated fuel damage would occur 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 1.07.
MCpR > 1.07 represents a conservative margin relative to the conditions required to maintain fuel cladding The fuel cladding is one of the physical barriers which separate radioactive materials from the integrity. The integrity of this cladding barrier is related to its relative freedom from perforations or environs.
cracking. Although some corrosion nr use related cracking may occur during the life of the cladding, fission product migration from th,is source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above While fission product migration frun cladding design conditions and the protection system safety settings.
perforation is just as measurable as that from use related cracking the thermally caused cladding perforations signal a threshold, beyond which still greater thennal s.resses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with margir. 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.
Ilowever, the existence of critical power, or boiling transi-tion, is not a directly observable parameter in an operating reactor.
Therefore, the margin to boi;ing tfDnsition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.
The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the ratto of the bundle power which would produce onset of transition boiling divided by the The minimum value of this ratio for any bundle in the core is the minimum critical power actual hundle power.
ratio (MCPR).
It is assumed that the plant opera. tion is controlled to the nominal protective setpoints via the instrumented variable, i.e., normal plant operation presented on~ Figure 2.1.2 by the nominal expected flow control line.
The Safety limit (MCPR of 1.07) has sufficient conservatism to assure that in the event of an l
abnormal. operational transient initiated from a nonnal 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 = 1.07) is derived from a detailed statistical analysis considering all of the l
uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as described in Reference 1.
The uncertainties employed in deriving the safety limit are provided at the beginning of each fuel cycle.
1.
General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Appifcation, NE0010958.
B2-1 Amendment No. Ig g 6gg
Iterause the halling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of PCPR = 1.07 would not produce bofitng transition.
Ilowever. If bolling transition were to occur, clad perforation wouid not be expected. Cladling teroperatures would increase to approximately 1100*F which is below the perforation tenperature of the cladding material.
This has been verified by tests in the General Electric Test Peactor (GETR) where fuel similar in design to Hillstone operated above the critical heat flux for a significant period of time (30 minutes) without clad pe r fora t ion.
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 appilcability of the bo#1Ing transition correlation Is 1400 psia during normal power operation. Ilowever, the reactor pressure is ilmited as per Specification 2.2.1.
In addition to the bofitng transition Ifmit (MCPR = 1.07) operation is constrained to a maximum LilGR= 17.5 kW/f t for 7 x 7 and 13.4 kW/f t for 8 x 8.
At 100% power this Ilmit is reached with a maximum total peaking factor (HIPF) of 3.08 for 7 x 7 fuel, 3.01 for 8 x 8 fuel, and 3.00 for u x UR fuel, for tne case of the MIPI exceeding these values,~ operation is periiitted only at less titan 100% of rated thermal power ar.d only with reduced APRM scram settings as required by Specification 2.1.2.A.I.
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 dif ferential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentfally all elevation head, the core pressure drop at low power and all flows wl11 always be greater than 4.56 psi. Analyses show that with a flow of 28 x 103 lbsihr bundle flow, bundle pressure drop is nearly independent of bundle power and gas a value of 3.5 pst. Thus, the bundle flow with a 4.56 pst 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 assenbly critical 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 core themal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.
Plant safety W1yses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 2.1.1A or 2.1.1B will not be exceeded.
Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure.of the main turbine stop valves) does not, necessarily cause fuel dai..ge.
Ilowever, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The con-cept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.
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setting was selected to provide adequate margin from the thermal-hydraulic safety limit and allow operating margin to minimize the frequency of unnecessary scrams s The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of HTPF and reactor core thermal power. The scram setting is adjusted in accordance with the formula given in Specification 2.1.2A.1, when the HTPF is greater than 3.08 for 7x7 fuel, 3.01 for lixtl fuel, and 3.00 for 8xfiR fuel.
Analyses of the Ilmiting transients show that no scram adjustment is required to assure HCPR > 1.07 when the transient is initiated from HCPR's specified in Section 3.11.C.
In order to assure adequate core margin during full load rejections % the event of failure of the select rod insert, it is necessary to reduce the april scram trip setting ts 90% 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 90% of rated has been estab11shed to provide substantial margin during such an occurrence. The trip setdown is delayed to prevent scram during the inititi portion of the transient. The specified maximum setdown delay of 30 seconds is conservative because the cold feedwater transient does not produce significant lacreases in reactor power before approximately 60 seconds following the_ load rejection. Reference Amendinent 16 Response to Questions A-12, A-14, A-15, and D-3.
for operation in the refuel or startup/ hot standby modes while the reactor is at low pressure, the APRM reduced flux trip scram setting of < 15% of rated power provides adequate thermal margin between the
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maximum power and the safety ilmit, 25% of rated power.
The margin is adequate to acconsnodate 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 coef fectents a " small.and controi rod patterns are constrained to be unifonn by operating procedures backed up by the rod worth minimizer.' blorth of individual n ods 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 power 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 Ilmit. The APRM reduced trip scram remains active untti the mode switch is placed in the run position.
This switch occurs when the reactor pressure is greater than 800 psig.
The IRH trip at < 120/125 of full scale remains as a backup feature.
Amendment flo. H #, py, 66 B2-6
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The design of the ECCS components to meet the above criteria was dependent 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 not 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 during normally expected transients.
E.
Turbine Stop 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 < 10% of valve closure the resultant increase in surface heat flux is limited such that
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HCPR remains above 1.07 even 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 Thc turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutrun flux resulting from fast closure of the turbine control valves due to a load rejection and sub-sequent failure of the bypass; i.e., it prevents HCPR from becoming less than 1.06 for this transient.
For the load rejection from 100% power, the heat flux increases to only 106.5% of its rated power value which results in only a small decrease in HCPR.
This trip is bypassed below a generator output of'307 HWe because, below this. power level, the MCPR is greater than 1.07 throughout the transient without the scram.
g In order to accomodate 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 260 millisec following initiation of load rejection.
Af ter 260 millisec, the trip is bypassed providing the. bypass valves have opened.
If the bypass valves have not opened after 260 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 during the worst transient even if this trip fails.
There are many other trip functions which protect the system during such transients.
Reference Response D-3 of Amend. int 16.
B2-8 Arendment No. J, if, 61
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