ML20028A287
| ML20028A287 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/12/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028A285 | List: |
| References | |
| NUDOCS 8211180051 | |
| Download: ML20028A287 (35) | |
Text
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UNITED STATES l
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NUCLEAR REGULATORY COMMISSION l
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WASHINGTON, D. C. 20556
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THE CONNECTICUT LIGHT AND POWER COMPANY THE HARTFORD ELECTRIC LIGHT COMPANY WE51ERN MASSACHUSETT5 ELEGIRIC COMPANY AND NORTHEAST NUCLEA G ERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 87 License No. DPR-21
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1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by the Connecticut Light and Power
. Company, the Hartford Electric Light Company, Western Massachusetts Electric Company and Northeast Nuclear Energy Company.(the licensees) dated October 15, 1982, as supplemented November 2, 1982,
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complies with the standards and requirements of the Atomic Energy Act of 1954, as ar.. ended (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 Connission; C.
There is reasonable assurance (1) that the, activities authoriied by the 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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PDR ADOCK 05000245 e
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Accordingly, the.llcense is amended by changes to the Technical Specifications,as indicated in the attachment-to this license amendment and Par.agraph 3.B of Provisional Operating License No. DPR-21 is hereby amended to read as follows:
B.
Technical Specifications
^
The Technical Specifications contained in Appendix A as revised through Amendment No. 87, are hereby incorporated in the license. Northeast Nuclear Energy Company _shall operate the facility in accor-
' dance with the Technical Specifications.
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3.
Thi's license amendment isleffective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
y k' h D l
Dennis M. CrutetMkkf Operating Reactors 1 Branch #5 l
Division of Licensing Atta'chment:
Changes't'o the Technical Specifications Date of Issuance:
November 12, 1902 e
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. ATTACHMENT T0 LICENSE AMENDMENT N0. 87 PROVISIONAL OPERATING LICENSE N0. DPR-21 DOCKET NO. 50-245 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages contain the captioned amendment number and vertical lines indicating the areas of change.
PAGES a
2-1 through 2-3 2-5 through 2-7 B 2-4 8 2-9 3/4 2-2 3/4 2-5 (also replaces 3/4 2-5a).
e..,.
3/4 3-7 3/4 7-5 3/4 11-2 through 3/4 11-11 3/4 12-10 3/4 12-11 3/4 12-11a B 3/4 2-2 B 3/4 3-5 B 3/4 11-1 through 8 3/4 11-3 6-21
SAfLTY LIMITS LIMITING SATETY SYSTEM SETTINGS 2.1.1 full CLADDING INTEGRITY 2.1. 2 FUEL Cl. ADDING INTEGRITY Apt cal illty,:
Appilcabil l_ty,:
li Applies to the interrelated variables associated Applies to trip settings of the instruments and with fuel thermal behavior.
devices which are provided to prevent the reactnr system safety limits from being exceeded.
Obj ec t_.t ve :
To establish limits below which the integrity of 0_hjective:
1 the fuel cladding is preserved.
To define the level of the. process variables at which automatic protective action is initiated Syeci fica tion _:
to prevent the safety limits from being exceeded.
A.
Men the reactor pressure is greater than 800 psia and the core flow is greater than 10% of Specificatidn:
rated design, a minimum critical power ratio (MCPR) less thani.07 shall constitute a The limiting safety system settings shall be as violation of the fuel cladding integrity specified below:
safety limit.
A.
Neutron Flux Scram D.
E en the reactor pressure is less than or equal to 800 psia or reactor flow is less than 101 1.
APRM Flux Scram of design, the reactor thennal power transfe'rred Trip Setting (Run Mode) to the coolant shall not exceed 25% of rated.
a.
Een the Mode Switch is in the C.
1.
To assure that the Limiting Safety System RUN position, the APRM flux Settings established in Specifications scram trip setting shall be as 2.1.2A and 2.1.2 Bare not exceeded, each shown on Figure 2.1.2 and shall required scram shal.1 be initiated by its be:
i primary source signal. The Safety Limit shall be assumed to be exceeded when scram S < 0.58 W + 62 is accomplished by a means other than the Primary Source Signal.
2-1 Amendment No. (, M,##,B 7
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LlHITING SAFETY SYSTEM SETTINGS SAFETY LIMITS v
where:
2.
When the process computer is out of
. service, this safety limit shall be S = Setting in percent of rated assumed to be exceeded if the neutron thermal power (2011 MWt) flux exceeds the scram setting established by Specification 2.1.2A W = Total recirculation flow and a control rod scram does not in percent of design
- occur, See Note (1)
D.
Whenver the reactor is in the cold shutdown The trip setting shall not
~
condition with irradiated fuel in the reactor exceed 90 percent of rated vessel, the water level shall not be less power during generator than that corresponding to 12 inches above load rejections from an the top of the active fuel when it is seated initial generator power in the core. This level shall be continuously greater than 307 MWe.
monitored.
The APRM scram setdown shall be 90% of rated within 30 seconds after initiation of full load rejection.
b.
In the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
5 S (0.58 W + 62 )[h3
- where, FRP = fraction of rated thermal power (2011 MWL)
Note (1) Design flow to be defined as the recirculation 6
flow (not to exceed 33.48 x 10 lbs/hr.) needed to achieve 100% core flow.
Amendment No. 16, #6, 16, 8 7 2-3 r
LIMITING SAFETY SYSTEM SETTINGS SAFETY LlHITS for no continatinn of loop recircul..-
j tion flow rate and core thermal pau-r shall the APRM flux scram trip setting be allowed to exceed 120'. of rated thermal power.
2.
APRM Reduced flux Trip Setting (Refuel or Startup/ilot Standby Model When the mode switch is in the refuel or Start Up/ 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.
B.,
1.
APRM Rod Block Trip Setting The APRM rod block trip setting a.
shall be as shown in Figure 2.1.2 and shall be:
(Run Mode)
S ul f6 0.58W 1-50 f
where:
SR8 = 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 pouic density (MFLPO) grerter than ie fraction of rated power (FRP), the setting shall be modified as follows:
1 d
Amendment No. JS. 34. $54 76,6 7 2-5 c
w_--______-
SAFETY LIMilS LlHlilNG SAFETY SYSTEM SETTINGS SR8 I (0.58W + 50 ) [H D
where:
FRP = fraction of rated ther. pial f
power (2011 MWt)
MFLPD = maximum fraction of liniiting power density where the limiting power density is 13.4 KW/f t for 8x8, 8x8R and P8x8R fuel.
The ratto of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
c.
During power ascensions with power levels less than or equal to 90t, APRM Rod Block Trip Setting adjust-ments may be made as described below, provided that the change in scram setting adjustment is less than 10% and a notice of the adjust-ment is posted on the reactor con-trol panel:
The APRM meter indication is adjusted by:
ARPM = (NFLPD) P FRP where:
APRM = APRM Meter Indication P
= % Core Thermal Power Amendment No. J$ $$' 78,6 7
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. r 2-6
O SAFETY LIMIT L IMi rillG 5,AFETY SYSTEM SETTitlGS B.
'2.
The APRM rod lilock trip setting Inr 5
the refuel and startup/hnt standby,
node.shall be less than or equal to 12% rated thernal power.
C.
The reactor Low Water Level Scram trip setting shall oe greater than or equal to 127 inches above the top of the active fuel.
1 D.
The Reactor Low Low Water Level ECCS Initiation trip point shall not be 9reater than 83 inches nor less than 79 inches.
E.
The turbine Stop Valve Scram trip set. ting 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
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selected bypass valves to start opening within 260 milliseconds.
The maximum setting of the time delay relays which bypass this scram shall be 260 milliseconds.
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l G.
The Main Steam Isolation Valve Closure Scram trip settipgs shall be less than or equal to ten percent valve closure from full open.
H.
The Main Steam Line Low Pressure trip which initiates main steam line isolation valve closure shall be greater than or equal to 825 psig.
l 4
Amendment No.16, 34. M,8 7 4
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l 2.1.2 Bases:
The transients expected during operation.of the Mil stone Unit 1 have been analyzed up to the thermal powir' condition of 2011 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.3.1 (2).
In addition, 2011 MWt is the licensed maximum steady-state power level i
of Millstone Unit 1.
This maximum steady-state power will never be knowingly exceeded.
Conservatism was incorporated by conservatively estimating the controlling factors such as void reactivi%
i l
coefficient, control rod scram worth, scram delay time, peaking factors, axial power shapes, et'c.
Thesc 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 substan-tiated in operation as a conservative tool for the evaluation of reactor dynamics performance.
Comparisons I
1 have been made showing results obtained from a General Electric boiling water reactor and the predictions made by the model. The comparisons and results are summarized in Reference 1.
l The void reactivity coefficient utilized in the analysis is conservatively estimated to be about 25% larger than the most negative value expected to occur during the core lifetime.
The scram worth used has been derated to be equivalent to the scram worth bf 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 l
Technical Specifications. The effect of scram worth, scram delay time and rod insertion rate, all conserva-tively applied, are of greatest significance in the early portion of the negative reactivity insertion. The l
rapid insertion of negative reactivity strongly turns the transient and the stated 5% and 20% insertion times conservatively accomplished this desired initial effect. "The time for 50% and 90% insertion are given to assure proper completion of the insertion stroke, to further assure the expected performance in the earlier portion of the transient, and to 'stablish the ultimate fully shutdown steady-state condition.
e 8
I For analyses of the thermal consequences of the transiente. MCPRs specified in Section 3.11.C are conservatively assumed to exist prior to initiation of the transients.
1 (1) Linford, R. B. " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor,"
i NEDO-10802.
(2)
" Extended Load Line Limit Analysis, Millstone Point Nuclear Power Station, Unit 1" NEDO-24366 and NEDO 24366-1.
I I
e B2-4 Amendment No.), Jg, /f3 g 7 I
..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 resultinf rapid cooldown of the vessel. Advantage was taken of, the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.
Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram.
Thus, the combina-tion of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.
In addition, the isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.
With the scrams set at 10% valve closure, there is no increase in neutron flux during valve closure.
H.
Main Steam Line Low Pressure Initiates Main Steam Isolation Valve Closure The -low pressure isolation at 825 psig was provided to give protection against fast reactor depressuriza-l tfon and the resulting rapid cooldown of the vns."
Mvantage wa.s taken of the scram feature which octurs when the main steam line isolation valves err.!amd to rrovide for reactor shutdown so that opera-tion at pressures lower than those specified in the 'lermae hyo,,ulic safety limit does not occur, although operation at a pressure lower than e2s c<
..9ul.4 wessarily constitute an unsafe condition.
e mendmentf!a.[/ b 't B2-9 3
1
I Al!L L 1.2.1 INSlRUMLNTArlOrt TitAT INITI ATLS l'RIMARY CCfifAlaMt HI ISOLATION FU I
Minimum Number of Operable instrument Action [1]
Tr_ip, Lei /e1 Sett ip3 Channel > Per Trip i
Instruments A
l Sfstem (1)
_127 inches above top of active fuel A
Reactor Low water 79 (+4-0) inclies above top of active fuel 2
Reactor low Low Water A
l 2
2 (4)
High Drywell Pressure 1 2 psig B
120% of rated steam flow HighFlowMainSteamlitie 1
2 (2) (5)
B 2 of 4 in each of High Temperature Main Steamline Tunnel 1 200*F 2 subchannels 8
2 High Radiation Main 7 times normal rated power background Steamline Tunnel 1
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8 l
l 2
Low Pressure Main
> 825 psig Steamlines C
l 164 inches > trip setting (water differential l
29 High Flow Isolation 150 inches.
onsteamlik)j*isetting(waterdifferential Condenser Line 44 inches > tri i
l on water side) 35 inches.
i d trip systems for each l
Whenever primary containment integrity is required, there shall be two op i
(1)
If If the first column cannot be met for one of the trip systems, that trip system shall be tr l
(2) Per each steamline.
l h ll be taken:
the first column cannot be met for both trip systems, the appropriate actions listed be ow s (3) Action:
Initiate an orderly shutdown and have. reactor"in cold shutdown condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Initiate an orderly load reduction and have reactor in Hot Standby within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- I A.
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B.
Close isolation valves in isolation condenser system.
i C.
PS-1621, A through D -
May be bjpassed when necessary by closing the manual instrument isolation valve for line (4) during purging for containment inerting or deinerting.
Minimum number of operable instrument channels per trip system requirement does not hav l
(5) if both containment isolation valves in the line are closed.
3/4 2-2 AmendmentNo./,M,bi(
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TABLE 3.2.3 s
INSn JMENTATION THAT INTIATES ROD BLOCK
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Minimum Number of Operable Inst.
Channels per Instrument Trip Level Setting Trip System (1) 1 (7)
APRM Upscale (Flow Biased)
See Specification 2.1.28 APRM Downscale
> 3/125 Full Scale 1 (7) 1 (6)
Rod Block lionitor Upscale (Flow Biased) 1 58W + 50 (2) l 1 (6)
Rod Block Monitor Downscale
> 3/125 Full Scale 3'
IRM Downscale (3)
> 3/125 Full Scale 3
IRM Upscale i 108/125 Full Scale I
e lower cap to j
1 Scram Discharge Volume - Water Level High 1g neige.}
1 Scram Discharge Volume - Scram Trip Bypassed N/A (1)
For the Startup/ Hot Standby and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks; IRM downscale are not operable in the RUN position and APRM downscale need not be operable in the Startup/ Hot Standby mode.
If the first cohann cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped.
If the first column cannot be met for both trip systems, the systems shall be tripped.
(2) W is the recirculation flow required to achieve rated core flow expressed in percent.
(3)
IRM downscale may be bypassed when it.is on its lowest range.
(4) This function may be bypassed.when the count rate is > 100 cps or when all IRM range switches are above Position 2.
'(5) One of these trips may be bypassed. The SRM function may be bypassed in the higher IRM ranges when the IRM upscale r@d block is operable.
(6) The trip may be bypassed when the reactor power is < 30% of rated. An R8M channel will be considered inoperable if there are less than half the total number of normal inputs from any LPRM level.
' (7) There must be a total of at least four (4) operable or operating APRM channels.
AmendmentNo./%,' 87 3/42-5
ALLOWABLE COMBIrlATI0fiS OF TOTAL CORE FLOW AtlD POWER LEVEL FIGURE 3.3.1 l'
140 t,
100% POWER UNE 120 100% INTERCEPT POINT (100187)
APRM ROD BLOCK LINE (1W100) 10.55W + 50%)
L (100/100l
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.20 Amendment No. 52,8 7
.I COPI FL W (5)
Ref. NEDO 24366 Sept. 1981 3/4 3-7 I
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 8
3.
Primary containment integrity as defined in Section 1.0 shall be maintained at all 3.
The primary containment integrity l
times when the reactor is critical or when
~
shall be demonstrated as follows:
the reactor water' temperature is above Integrated Primary Containment 212*F and fuel is in the reactor vessel a.
except while performing low power physics Leak Test (IPCLT).
test at atmospheric pressure during or after refueling at power levels not to exceed 5 he(t).
(1)
Integrated leak rate tests shall be performed prior to initial unit operation at the test pressure of 43 psig. Pt(43),toobtain measured leak rate L (43).
(2) Subsequent leak rate tests shall be performed without preliminary leak detection surveys or leak repairs immediately prior to or during the test at an initial pressure of approximately 43 psig.
Amendment No. 51,8 7
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Amendment No, I6, 14, 49, 61, 18)6
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FUEL. IYPE P8DRB283.
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT B.
Linear Heat Generation Rate (LHGR)
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B.
During steady state power operation, the lined'r Linear Heat Generation Rate (LHGRT '
i heat generation rate (LilGR) of any rod in any The LHGR shal fuel assembly at any axial location shall not be checked daily during reactor operation exceed the maximum allowable LilGR of 13.4 KW/f t' at > 25% rated thermal power.
for 8 x 8 fuel bundles.
(
t f.
i i
d I
During power operation, the *LHGR i shall not exceed the limiting vaiue.
If at any time during operation it is determined by normal survalliance 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 LilGR is not returned to within the prescribed limits within two (2) hours, the reactor ghall 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 reacL6r operation is i
within the prescribed limits.
i I
~
Amendment No. 4, 16, 28 b 7 j
i LIMITING CONillTION FOR OPERATI0tt SURVEILLANCE REQUIREMENT C.
Itinlesw Tritical Power Ratio (MCPJR h
C.
Minimdie Critical Power Ratio (l'CPR) nuring power operation,*HCPR shall be as shown in
- 1. hCPR shall be determined daily during
" fable 3.11.1 If at any time during operation reactor power operation st > 254 rated It is determined by notwal surveillance that the theswal power and following any change in limiting value for MCPR is being exceeded, action power level or distrileution that would cause operation with a Ilmiting control rod pattern shall f.e initiated within 15 minutes to restore operation to within the prescribed limits.
If as descrified in the bases for specification the steady st.ste MCPR is not returned to within
'3.3.D.S.
tiee prescrified limits within two (2) hours, the reactor shall be brought to the Cnid Shutdown
- 2. Utilization of Option B Operating limit condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillant.e and MCPR values requires the scram time testing i
corresponding action shall continue until reactor.
of 15 or more control rods on a rotating i
operation is within the prescribed Iimits, basis every 120 operating days.
i For core flows other than rated the MCPR's in Table 3.11.1 shall be multi-led by K. where g
K is as shown in figure 3.11.2.
g D.
If any of the limiting values identified in Specifications 3.ll.A. 3, or C. ard exceeded, even if corrective action is taltest, as pre-scribed. a Reportable Occurrence riiport shall be sulamitted.
1 a
0 4mendmentNo.7.Jf,/ho7 4
374 yi,,
TABLE 3.11.1 OPERATING LIMIT MCPR'S FOR CYCLE 9 a
(OPTION G)
BOC9 TO EOC 9 EOC9 TO 70% COASTDOWN FUEL TYPE 1.40 1.40 8x8 1.40 1.40 8 x 8R 1.43 1.43 P8 x 8R OPERATING LIMIT MCPR'S FOR CYCLE 9 (OPTION A) i BOC9 TO EOC 9 EOC9 TO 70% COASTDOWN FUEL TYPE 1.45 1.45 8x8 1
1.45 1.45 8 x 8R 1.48 1.48 P8 x 8R 9
I 4
Amendment No. 28, $(, 47, $l, 73;8 7 3/4 11-10 I
00 1
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0 1
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SURVEILLANCE REQUIREMENTS
- LIMITING cot *DITION FOR OPERATION E.
Fire Detection Instrumentation Fire Detection Instrumentation.
e 4
1.
The fire detection instruments listed in 1.
The minimum required fire derection instrumenta-Table 3.12.2 shall be demonstrated OPERABLE tion for each fire detection zone shoUn in Table.
3.12.2 shall be OPERABLE whenever equipment in at least once per 6 months be performance of an INSTRUMENT FUNCTIONAL TEST with the that fire detection zone is required to be exception that the functional test may Ol'ERABLE.
consist of injecting a simulated electr.ical 2.
With less than the minimum required nuimber of the signal into the measurement channel rather fire detection instrument (s) shown in Table than the instrument. Due to the inaccessability 3.12.2 OPERABLE:
of the fire detectors located in the condenser bay, a sample consisting of 1/3 of the detectors Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a watch. patrol per channel will be tested during every refuel a.
to inspect the zone (s) with the inop-outage. The sample test cycle will be completed
.erable instrument (s) at least once per every third refueling outage.
hour, and 2.
The non-supervised circuits between the above b.
Restore the inoperable instrument (s) to required detection instruments and the control OPERABLE status within 24 days or, in room shall be demonstrated OPERABLE at least lieu of any other report required by once per 31 days, per approved procedures.
Specification 6.9.1, prepare and sub-mit a Special Report to the Comunission
.... pursuant to Specification 6.9.2 within- -
the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restor-ing the instrument (s).to OPERABLE status.
e e
b7 Amendment No. $$,
\\
TABLE 3.12.2 FIRE DETECTION INSTRUMENTS NOTE: Notwo(2)adjacentdetectors inoperable.
HEAT SMOKE MINIMUM MINIMUM TOTAL INSTRUMENTS TOTAL INSTRUMENTS INSTRUMENT LOCATION AVAILABLE REQUIRED AVAILABLE REQUIRED 1.
Cable Vault 15 12 2.
H Seal Oil Unit 1
1 2
3.
Condenser Bay 36 25 4.
Diesel Generator Room 6
4 l'
5.
D/G Fuel Oil Day Tank Room 1
1 l
6.
Gas Turbine Enclosure 3
2 7.
RX Bldg (1-FDS-1) a.
R-2a RX Bldg 14'6" 7
5
&kl. Corner Including Crd Bank b.
R-2B RX Bldg 14'6" N.E. Corner 4
3 c.
R-2C RX Bldg 14'6" S.E. Corner 3
2 d.
R-2D RX Bldg 14'6" S.W. Corner 3
2 including C.R.D. Bank
- ~
1 1
e.
R-3 Tip Room 1
1 f.
R-5 Shutdown Cooling Pump Room 8.
RX Bldg (42' elev) 1-FDS-2 3
2 a.
R-17 Clean-up Pump Room b.
R-18 Shut Down Heat Exchanger Room 1
1 c.
R-19 RX Bldg 42' N.W. Corner to SE Corner 9
7 j
9.
RX Bldg (Elev. 65' and 82') 1-FDS-3) a.
R-12 RX Bldg Elev. 65' 9
7 except gated area along North Wall b.
R-13 RX Bldg (Elev.82") West Side 3
2 c.
R-14 RX Bldg Elev. 82' East Side 2
1 Amendment No. ##, 83,B 7
-(
3/4 12-11
TABLE 3.12.2 FIRE DETECTION INSTRUMENTS NOTE: Notwo(2) adjacent detectors inoperable.
HEAT SM0KE MINIMUM MINIMUM TOTAL INSTRUMENTS TOTAL INSTRUMENTS INSTRUMENT LOCATION AVAILABLE REQUIRED AVAILABLE REQUIRED
- 10. Turbine Bldg (14'-6") 1-FDS-4 a.
T-5A Condensate Pumps and Condensate Booster Pumps 6
4 b.
Condensate Demin Panel, RX Feed Pumps 6
4 c.
T-SC TBCCW to Stator Cooling Unit 15 12 d.
T-6 Decon Room 1
1 e.
T-12 Chem Lab 3
2 f.
T-17 Mezzanine Above MCC D-2 (l A-2) 3 2
- 11. Rad. Waste 1-FDS-5 a.
RW-A Radwaste Storage Bldg.
3 2
b.
RW-B Liquid Raduaste Bldg.
6 4
- 12. Turb. Bldg. (Elev. 36'-6". 7' 34'-6") 1-FDS-6)
- a. "T-10A Make Up Demin Storage Tk, Make Up Demin
1 b.
T-19B 4KV Bus 14F and 480V Bus 12F c.
T-19C 4KV Bus 14A, 4KV Bus 14C and 480V 3
2
~
d.
T-19D 4KV Bus 14B & 4KV Bus 14D e.
T-19E 4KV Bus 14G, DC Buses 101A & 101B 8
6 480 V Buses 12C & 12D, Iso Phase Bus Duct -
f.
T-15A H &V Equipment Rm. Elev. 54'6' 2
1
~
13.
Control Room T-21 (Elev. 36',-6")
9 7
14.
Screen House SH 7
5
- 14. Battery Rooms, Station Nc. 2 a.
Battery Rm.1 (Sta. #2) 1 1
b.
Battery Rm. lA (Sta. #2) 1 1
Amendment No. U; 5 7 4
3/4 12-11a
The high drywell pressure instrumentation is a back-up to the wat er leal Instrumentation and in adds tionet ur the break s..
Initiating ECCS it causes isolattun of Group 2 isolation valves.
h l ts' tion will initiate ECCS operation at about the same time as the low low water level instrumentatio given above are applicable here also.'
Group 2 actustion also initiates the $8GT5.
valves, and reactor building ventilation isolation valves.
ltd pressure activates only these valves because high drywely pressure could occur as the result of non-safety re causes such as not purging the dryweIl air during startup.
The low low water level instrumentation initiatrs tions and only the valves in Group 2 Tre required to close.
l protection for the full spectrum of loss of coolant accidents and causa a trip of all primary system isolation Venturis are provided in the main steamlines as a means of measuring steam flow and also Ilmiting the los Inventory from the vessel during a steamline break accident.
The primary function of the lastrumentation it, to tion is provided which causes a trip of Group 1 isolation valves.
For the worst case accident, nin steamline detect a break in the main steamline, thus Group 1 valves are closed.
d main steamline closure, limit the mass inventory loss such that fuel is not unc less The main steam-than 1500*F and release of radioactivity to the environs is well below 10 CFR 100 guideline values.
line high flow break detection is a one out of two twice logic for each individual steamline, four detectors per When a steamline is isolated by closing both main steam isolation valves the d
operable instrument channels per trip system requirements are not required to be met because the p line for a total of 16 detectors.
by the ranaining operable logic in the in-service steamlines provides complete recognition of the steam measurements required'for correct protcctive action.
Trips Temperature monitoring instrumentatton is provided in the main steamline tunnel to detect leaks in this srea.
p:
Its setting of are provided on this in:trumentation and when exceeded cause closure of Grtg 1 isolation valves.
200"F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire sne For large breaks, it is back-up to high steam flow instrumentasion Jiscussed above, and for small d
with the resultant small release of radioactivity, gives isolation before the gulaelines of 10 CFR 100 are exce we.
of breaks.
This instru-High radiation monitors in the main steamline tunnel have been provided to ditect gross fuel failure.
mentation causes closure of Group 1 valves, the only valves required to close to prevent further release to the With the established setting of seven times normal background, and main steamline isolation valve closure, fission product release is Ifatted so that 10 CFR 100 guideline values are not exceeded for the mo environment.
failure mechanism postulated (control rod drop accident).
9 Al Pressure instrumentation Is provided which trips when main steamline pressure at the turbine drops be In the " Refuel," " Shutdown," and trip of this instrumentation results in closure of Group 1 isolation valves.This function is provided primarily to provide protection "Startup/ Hot Standby" mode this trip function is bypassed.
With the trip agafnst a pressure regulator malfunction which would cause the control and/or bypas 1500*F; thus, there is no release of fission products other than those in the-reactor water.
8 3/4 2-2 Amendment No. /, [g, 8 7
.j i
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- -. ~. -
1 s
4.
The Source Dange Monitor (SRH) system perrnrms n3 automatic safety system functica; l.e., it has no scraa function.
It does provide the operator with a visual Indication of neutren level. This is needed fcr knowledgeable and ef ficient reactor star tup at Idw neutron levels. The requirement of at least 3 counts-per second assures that adequate monitoring capability is ava!1able. One operable SRM channel would be
. adequate to monitor the approach to criticalityrusing homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRit's ar) provided as an added conservatism.
5.
The Rod Block Monitor (RBH) is designed to automatically preve'nt 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 he hypassed from the console for maintenance and/or testing. ! ripping 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 restric-tions with one channel out of service conservatively assure that fuel damage will not occur due to rad withdrawal errors when this condition exists. During reactor operation with certain Ifmiting contrni rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with MCPR's less than 1.07 DurIng use of such patterns,1t is judged thet testing of the RDM systen pr1or to withdrawal.of such rods to assure Itis operability will assure that improper withdrawal does not occur.
It is the responsibl11ty of the Reactor Engineer to identify these Ilmiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of in-operable control rods in other than limiting patterns.
C.
Scram Insertion Times T'he control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; 1.e., to prevent the MCPR from becoming less than Lo'7. The limiting power transient is that.
resulting from a generator load rejection coincident 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 protection, and MCPR rehains greater than 1.06.
Amendment 21 shows the control rod scram reactivf ty insertion data used in analyzing the transients.
The limit on the number and pattern of rods permitted to have long scram times fs spectfled to assure that the reactivity insertion rate effects of rods of long scram times are minfelred. Grouping of long scram time rods is prevented by not allowing more than one control' rod in any group of four control rods te have long insertion times. The minimum amount of reactivity to be inscrted dut'ing a scram is controlled by permitting no more than
)
10% of the operable rods to have long scram, times.
In the analyttal treatment of the transient. 290 milli-1 seconds are allowed between a neutron sensor reaching the scram point and the start of motfon 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 solenold deenergizes. Approntmately 120 al11tseconds later, the control rod motion is estimated to actually begin. However. 200 afillseconds is conservatively assumed for this time Interval in the transient analyses and this is also included in the allowable scram insertfon times of Specification 3.J.C.
The time to deenergize the pliot valve scram solenold 15 measured during the calibration tests required by Specification 4.1.
Amendment No. (, M,0 7 4
gg 7j4 3,5
l 3.11 and 4.11, Bases e
i A.
Average Planar Linear Heat Generation Rate (APLHGR)(
This specification assures that the peak cladding temperature following the postulated design basis loss-of.
coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.
l The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent Since expected local variations in secondarily on the rod to rod power distribution within an assembly.' elated peak clad temperature by less than i 20*F power distribution within a fuel assembly affect the c -
relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation l
The rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit.
l i
limiting value for AplHGR 11 shown in Figure,3.11.1.
Conservative LOCA calculations predict that nucleate boiling will be maintained for several seconds follesing This res' lts in early removal of significant amounts of stored energy which. if present later in the transient, when heat transfer coefficients are cor.siderably lower, would result in higher peali a design basis LOCA.
u As core flow is reduced below about 901, the time of onset of boiling transition enkes I
cladding temperature.
The approved ECCS evaluation model a sudden change from greater than about 5 seconds to less than I second.
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 The effect is to significantly reduce the energy calculated to be removed from the fuel during This results in an increase in calculated peak clad temperature of about 100'F which can be offset l
temperature.
For flows less than 90% of rated, a 51 reduction in the MAPLHGR limits in Figure blowdown.
by a 51 reduction in MAPLHGR.3.11.1. derived for 1001 flow will assure that the plant is operated in compliance to lower ficus.
B.
Lirear Heat Gene *ation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linser heat The UlGR shall be checked daily during reactor operation at),25% power to determine if fuel generation rate.
1 B 3/4 11-1 AmGndment No. i,16,18, M. Ap, y, ff, 8 7
l burnup, or control rod movenent has caused changes in power distribution. Fcr LHGR to b2 a llat ting valu2 below 25% rated thermal power, the MTPF would have to be greater than 10'which is precluded by a consider-~'
able margin when employing any permissible control rod pattern.
C.
The steady state value for MCPR was selected to p-ovide a margi:
3 accomodate transients and uncertainties.
in monitoring the core operating state as well asluncertainties in the critical power correlation itself.
This value ensures that; s
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.
At core thermal power levels 1 251, the reactor will be operating at minimum recirculation pump speed, and moderator void content will be very small. For 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 eequire-ments. 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 251 rated i
themal 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 in accordance with GE letter A. D. Vaughn'to P. A. Blasioli, July 9, 1982 Proposed Technical Specification Changes for Millstone Unit 1.
D.
Reporting 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 1
I plant is determined to be exceeding them. It is a requirement, as stated in Specifications 3.ll.A, B, and C that if at any time during power operation.it is determined that the limiting values for MAPLHCR, LHGR, or MCPR are exceeded, action is then initiated to restore operation to within the prescribed l
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 l
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.
I
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Amendment No. 4, 76, 28, 49 8 7 B 3/4 11-2 3
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9 A
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e l ADMIN!$TRATIVE CONTROLS I
~
s THIRTY. DAY %'RITTEN REPORTS (Continued) i ompleted copy of a licensee event report form. Information provided on che licensee event report foria shall be supplemented, as needed, by addi-ttional narrative material to provide complete explanation of the circus.
stances surrounding the even.t.,.
Reactor' protectior system. or engineered safety features instrument a.
settings which are found As be'less conservative than those estab-11shed by the techniga). specifications but which do not prevent the fulfil.1 ment of the functional requirements of affected systems b.
Conditions leading to operation in a degraded mode permitted by
.a limiting condition for operation or plant shutdown required by
- s' limiting condition for operation.
Observed inadequacies in the implementation of administrative or c.
procedural controls'which threaten to cause red 9etion of degree of. redundancy provided in reactor protection systems or engineered safety. features systems'.
e: d.
Abnormal degradation of systems other than those specified in 6.9.1 8.c. above, designed to contain radioactive material resulting from th'e fission process.
SPECIAL REPORTS Special reports sha.11 be submitted to the Director of the Office of 6.9.2 Inspection and Enforcement Regional Office within the time period specified.
l for each report. These reports shall be submitter. covering the activities identified below pursuant to the requirements of the appitcable reference l
specification:
9 f
. In-service Inspection Results. Specification 4.6.F.
a.
b.
Primary Containment Leak Rate Test Results. SpeO fication 4.7.A.3
- d.
Materials Radiation Surveillance Specimen E:: amination and Results.
Specification 4.6.8.3.
Fire detection instrumentation. Specification (3.12.E.2).
e.
.f.
Fire suppression systems. Specifications (3.12.A.2. 312.t.2, 3.12.C.2 and'3.12.C.4).
1
/ Amendment No.15. H. M3 87 i
6,
I
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