ML19261D920
| ML19261D920 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/29/1979 |
| From: | Vollmer R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19261D915 | List: |
| References | |
| NUDOCS 7906290018 | |
| Download: ML19261D920 (34) | |
Text
e R R80p UNITED STATES
[
f,j NUCLEAR REGULATORY COMMISSION E
WASHINGTON. D. C. 30GGS g
\\...../
CONNECTICUT LIGHT AND POWER COMPANY THE HARTFORD ELECTRIC LIGHT COMPANY WESTERN MAT 5ACHUSETTS ELECTRIC COMPANY NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 61 License No. DPR-21 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for emendment by Connecticut Light ano Power Company, The Hartford Electric Light Company, Western Massachusetts Electric Company, and Northeast Nuclear Energy Company (the licensees), dated March 5,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
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 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 common 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 Commission's regulations and all applicable requirements have been satisfied.
2306 A0 79062900/8i 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 Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendnent No. 61, are hereby incorporated in the license.
Northeast Nuclear Energy Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0fEISSION f
RiE ard H. Vollmer, Assistant Director for Systems & Projects Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 29,1979 230(> 3ll
ATTACHMENT TO LICENSE AMENDMENT NO. 61 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 include the captioned amendment number and contain vertical lines indicating the areas of change.
Overleaf pages* are included for document completeness.
REMOVE INSERT 2-6 2-6 B2-1 -- B2-2 B2-1 -- B2-2 B2-6 B2-6 B2-7 -- B2-8 B2-7 -- B2-8 B2-10 B2-10 B2-11 B2-ll 3/4 3-3 3/4 3-3 3/4 5-6 3/4 5-6 3/4 5-7 3/4 5-7 3/4 6-5 3/4 6-5 3/4 ll-7b 3/4 11-10 3/4 11-10 B3/4 2-3 B3/4 2-3 B3/4 3-3 --
B3/4 3-3 --
B3/4 3-4 B3/4 3-4 B3/4 6-4 B3/4 6-4 B3/4 6-5 B3/4 6-5
- 2-5, B2-5, B2-9, B2-12 (intentionally Blank), 3/4 3-4, 3/4 J-5, 3/4 5-8, 3/4 6-6, 3/4 11-9, B3/4 2-4, B3/4 6-3, and B3/4 6-6.
2306 J12
p SAFETY LIMIT LIMITING SAFETY SYSTEM SETTINGS A
S B'-(0.65 W + 42%)
g 7 pp where:
bT A = 3.08 for 7x7 fuel 6
9
= 3.04 for 8x8 fuel Cr9 MTPF = The value of the existing maximum total g
peaking factor.
(<
3 0.2. The APRM rod block trip setting for the r efuel mO, and startup/ hot standby mode, shall be less y
than or equal to 12% rated thermal power.
A C.
The reactor Low Water Level Scram trip settinq C-g_;
shall be greater than or equal to 127 inches above the top of the active fuel.
D.
The Reactor Low Low Water Level ECCS Initia-tion 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 g
u shall trip upon actuation of the acceleration o
relay in conjunction with failure of seier.tei!.
m
^
bypass valves to start openinq within 260 milliseconds.
6 The maximum setting of the time delay relays which bypass, this scram shall be 260 milliset i,r.ds G.
The Main Steam Isolation Valve Clm,us e ' t r.c I
trip settings shall be less than or eq.a! to ten percent valve closure from tul! opan.
n t. No. Ifr 34 2-5 H.
The Main Steam Line Low Pressure t rip which initiates main steam line iso!.it ion v.il.. i !osure.
shall be greater than or equal
'n
- n,,. i q.
m SAFETY LIMITS LIMITING SAFETY SYSTEN SETTINGS 2.2.1 REACTOR COOLANT SYSTEM 2.2.2 REACTOR COOLANT SYSTEM App 1tcability:
Applicabili ty:
Applies to limits on reactor coolant system pressure.
Applies to trip settings of the instruments and devices which are provided to prevent exceeding Objective:
the reactor coolant system safety limits.
To establish a limit below which the integrity of Objective:
the reactor coolant system is not threatened due to an overpr. essure condition.
To define the level of the process variables at which automatic protective action is initiated Specification:
to prevent exceeding the safety limits.
The reactor vessel pressure shall not exceed Specification:
1325 psig at any time when irradiated fuel is present in the reactor vessel.
A.
Reactor Coolant High Pressure Scram Trip Setting shall be < 1085 psig.
B.
The safety valve function settings of the six dual purpose relief / safety valves shall correspond with a steam pressure of:
No. of Valves Set Point (PSIG) rs) l 1095 + 1%
tid 1
1110 i 1%
CD 4
1125 i 1%
C7N Q-
.s>
2-6 Amendment No. 76, 61
2.1.1 Bases
. The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnomal 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 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 th,is source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result feom thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking the thermally caused cladding perforations signal a threshold, beyond which still greater themal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding 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 transi-tion, is not a directly observable parameter in an operating reactor. Therefore, the margin to boiling t4nsition 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 ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR).
It is assumed that the plant operation is controlled to the nominal protective setpoints via the g
instrumented variable, i.e., normal pla,nt operation presented on Figure 2.1.2 by the nominal expected flow u
c-->
l control line. The Safety Limit (MCPR of 1.07) has sufficient conservatism to assure that in the event of an abnomal operational transient initiated from a normal operating condition more than 99.9% of the fuel rods in ch 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
l uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation
~
U 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 Application, NEDO 10958.
Amendment No. Yg, W, 61 82-1
Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR = 1.07 woUld not produce boiling l
transition.
However, if boiling transition were to occur, clad perforation would not be expected. Clad [fingtemp 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 P.eactor (GETR) where fuel similar in design to Millstone operated abon the critical heat flux for a significant periud of tfme (30 mtnotes) without clad Thus, although it is not required to establish the safety limit, additional margin exists perforation.
The limit of applicability between the safety limit and the actual occurrence of loss of cladding integrity.
However, the reactor of the boiling transition correlation is 1400 psia during normal power operation.
presture is limited as per Specification 2.2.1.
In addition to the boiling transition limit (MCPR = 1.07) operation is constrained to a maximum LHGR= 17.5 l
kW/ft for 7 x 7 and 13.4 kW/ft for 8 x 8.
At 100% power this limit is reached with a maximum total peaking factor (MTPF) of 3.08 for 7 x 7 fuel and 3.04 for 8 x 8 fuel. For the case of the P1TPF exceeding these values operation is permitted only at less than 100% of rated themal power and 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.
Since At low pow:r and all flows this pressure differential is maintained in the bypass region of the core.
the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power Analyses show that with a flow of 28 x 103 lbs/hr bundle and all flows will always be greater than 4.56 psi.
Thus, the flow, bundle pressure drop is nearly independent of bundle power and gas a value of 3.5 psi.
bundle flow with a 4.56 psi driving head will be greater than 28 x 10 lbs/hr irrespective of total core flow Full scale ATLAS test data taken and independent of bundle power for the range of bundle powers of concern.
at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is With the design peaking factors this ccrresponds to a core themal power of more approximately 3.35 MWt.
Thus, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than than 50%.
10% is conservative.
Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Scram times are checks.d periodically to Safety Limit of Specification 2.1.1A or 2.1.18 will not be exceeded.The thermal power transient resulting when a y
assure the insertion times are adequate.
other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine u
stop valves) does not necessarily cause fuel damage.
However, for this specification a Safety Limit violation ca The con-will be assumed when a scram is only accomplished by means of a backup feature of the plant design.
CA cept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive
{
plant safety analysis.
~
Amendment No. f M, 61 O
Y
- ' i@
3
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,'This choice of using conservative values of controlling parameters and initiating transients at the licensed m.iximum.teady. Late power level produces more pessimi,L it answers than would resul t by using ryp-ti d values of control parameters and analyzing at higher power levels.
Steady-state operation without forced recirculation will not be permitted, except during startup t.. <. t i ng.
Ihr analysis to support operation at various power and flow relationships has considered operation uith either one or two recirculation pumps.
In sununary:
1-The licensed maximum steady-state power level is 2011 MWt.
2-The abnormal operational transients were analyzed to a thermal power level of 2011 MWt.
3-Analyses of transients employ adequately conservative values of the controlling reactor p,.c ceters.
4-The analytical procedures now used result in a more logical answer than the alternative method of assuminn a higher starting power in conjunction with the expected values for the parameters.
A.
Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken durinq steady-state conditions, reads percent of rated thermal power.
Since fission chambers provie the basic input signals, the APRM system responds directly to average neutron flux.
During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.
Therefore, during transients induced by disturbances and with APRM scram settings as specified by Section 2.1.2A, the thermal power of the fuel vill be less than that indicated by the neutron flux at the scram setting.
Analyse:s demonstrate tha t, evm. wi th a y
fixed 120% scram trip setting, none of the postulated transients result in violation of the fuel sa fety ua limit and there is a substantial margin from fuel damage.
Therefore use of a flow-biasbd m.re,m providev ch even additional margin.
Increasing the APRM scram setting would decrease the margin present before the thermal hyd aulic safety v-
~
limit is reached.
The APRM scram setting was determined by an analysis of margins required to provide N
a reasonable range for maneuvering during operation.
A reduction in this operating margin..oald increa:.
the frequency of spurious scrams which have an adverse affect on reactor safety.
Thus, the speci fied All' I
" *ndmen t No. 16 n2-s
h
/
setting was selected to provide adequate margin from the thermal-hydraulic safety limit and allow operating margin to minimize the frequency of unnecessary scrams.
The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MTPF and reactor core thermal power. The scram setting is adjusted in accordance with the formula given in Specification 2.1.2A.1, when the MTPF is greater than 3.08 for 7x7 fuel and 3.04 for 8x8 fuel.
l Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.07 when the transient is init;iated from MCPR's specified in Section 3.ll.C.
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 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 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. Reference Amendment 16 Response to Questions A-12. A-14, A-15, and 0-3.
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 themal margin between the maximum power and the safety limit, 25% of rated power. The margin is adequate to 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 coeffecients 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 power rise.
In an assumed uniform rod withdrawal approach to the scram level, the APRM system would be more than c
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 6
pressure is greate* than 880 psig.
00 The IRM trip at <_ 120/125 of full scale remains as a backup feature.
Amendment No. M, #, 61 82-6
The analysis to support operation at various power and flow relationships has considerei operation with either one or two recirculation pumps. During steady-state operation with one recirculclion puo operat-ing the equalizer line shall be open. Analyses of transients from this operating condition are hs ;
severe than the same transients from the two pump operation.
1 APRM Control Rod Block Trips B.
Reactor power level nay be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant l
recirculation flow rate and thus to protect against a condition of a MCPR < l.07.
This rod block setpoint, which is automatically varied with 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 re-circulation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship. Therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting.
The total peaking factor assumed for the analysis was 3.08 for 7x7 and 3.04 for 8x8 fuel. The actual power distri-bution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram setting, the APRM rod block setting is adjusted downward according to the equation included in Specification 2.1.28 if peaking factors greater than 3.03 for 7x7 fuel and 3.04 for 8x8 fuel exist; thus preserving the APRM rod block safety margin.
The APRM red block setpoint is reduced to < 12% of rated thermal power with the mode switch in refuel or Startup/ Hot 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 N
bases for the safety limit is maintained.
LN D.
Reactor Low Low Water Level ECCS Initiation Trip 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 and to limit fuel clad temperature to well v
below the clad melting temperature to assure that core geometry remains intact and to limit any clad m
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 setp-:nt 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 not be set lower because of ECCS capacity requirements.
B2-7 Amendment No.1), 71, 61
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
[
MCPR 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 The turbine control valve fest closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting fr&4 fast closure of the turbine control valves due to a load rejection and sub-sequent failure of the bypass; i.e., it prevents MCPR 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 MCPR. This trip is bypassed below a generator output of'307 MWe l
because, below this power level, the MCPR is greater than 1.07 throughout the transient without the scram.
In order to accennodate 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. After 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 Amendment 16.
u CD Ch N
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i.2.1 Bases:
The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products.
It is essential that the integrity of this system be protected by establishing a pres-sure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.
The pressure safety limit of 1325 psig as measured by the vessel steam space pressure indicator is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value is derived from the design pressures of the reactor pressure vessel, coolant system piping and isolation condenser.
The respective design pressures are 1250 psig at 575 F,1175 psig at 564*F, and 1250 psig at 575*F. The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes:
ASME Boiler and Pressure Vessel Code Section IIIfor the pressure vessel and isolation condenser and USAS B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pres-sure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 psig),
and the USASI Code permits pressure transients up to 20% over the design pressure (120% x 1175 = 1410 psig).
The Safety Limit pressure of 1375 psig is referenced to the lowest elevation of the primary coolant system.
The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is more than a factor of 1.5 below the yield strength of 43,300 psi at 575*F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield strength.
The relationships of stress levels to yield strength are conparable for the isolation condenser and primary system piping and provide a similar margin of protection at the established safety pressure limit.
The normal operating pressure of the reactor coolant system is 1035 psig. For the turbine trip or loss of electrical load transients the turbine trip scram or generator load rejection scram, together with the ps) turbine bypass system limits the pressure to less than 1085 psig. The safety / relief valves are set at (sa 1095 psig,1110 psig, and 1125 psig and are sized to keep the reactor coolant system pressure below 1375 c:3 psig with no credit taken for the turbine bypass system.
Credit is taken for the neutron flux scram, C7s however.
p(,
During operation, reactor pressure is continuously displayed in the control room on a 0-1500 psig pressure recorder.
ps, Amendment No. 61 B2-10
2.2.2 Bases In compliance with Section III of the ASME Boiler and Pressure Vessel Code,1965 Edition, the specified settings of the pressure relieving devices are below 103% of design pressure. As described in the General l
Electric Topical Report, NEDE-240ll-P-A, Generic Reload Fuel Application (Section 5.3), the most severe isolation event with indirect scram has been evaluated. The most severe isolation is the MSIV closure from steady-state operation at 2011 MW. The evaluation assures that the sizing and settings of the t
pressure relieving devices is 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.
4 Nuo Ch Nu Amendment No. 61 82-11
,i i
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.B Control Rod Withdrawal 4.3.B Control Rod Withdrawal 3.
Whenever the reactor is in the startup
- 3. (a) To consider the rod worth minimizer or run mode below 20% rated themal operable, the following steps must power, no control rods shall be moved be performed:
unless the rod wor',n minimizer is operable or a second independent (i) The control rod withdrawal operator or engineer verifies that sequence for the rod worth the operator at the reactor console minimizer computer shall be is following the control rod pro-verified as correct.
l gram. The second operator may be used as a substitute for an inoper-(ii) The rod worth minimizer compute able rod worth minimizer during line diagnostic test shall be a startup only if the rod worth successfully completed.
minimizer fails after withdrawal of at least twelve control rods.
(iii) Proper annunciation of the select error of at least one 4.
Control rods shall not be withdrawn out-of-sequence control rod in for startup or refueling unless at each fully inserted group shall least two source range channels be verified.
have an observed count rate equal to or greater than three counts '
(iv) The rod block function of the rod per second.
worth minimizer shall be verified by attempting to withdraw an out-of-sequence control rod be-yond the block point.
Nu (b)
If the rod worth minimizer is inoperable CD while the reactor is in the startup or run mode below 10% rated themal power, and a second independent operator or
{
engineer is being used, he shall verify that all rod positions are correct prior g
to commencing withdrawal of each rod group.
Amendment No. 22, 40, 61 3/4 3-3
=. _
- b
.;'a LIM {TINGCONDITIONFOROPERATION SURVEILLANCE REQUIREMENT 5.
During operation with limiting control 4.
Prior to control rod withdrawal for rod patterns, as detenuined by the startup or during ref ueling; verify l
reactor engineer, either:
that at least two source range channels have an observed count rate of at a.
Both RBM channels shall be operable; least three counts per second.
or 5.
When a limiting control rod pattern b.
Control rod withdrawal shall be exists, an instrument functional test blocked; or of the RBM shall be perfoemed prior to withdrawal of the designated c.
The operating power level shall be rod (s) and daily thereafter.
limited so that the MCPR will remain above 1.06 assuming a single
~
C.
Scram Insertion Times error that results in complete withdrawal of any single oper6ble During each operating cycle, each operable control rod.
control rod shall be subjected to scram time tests from the fully withdrawn position.
C.
Scram Insertion Times If testing is not acconplished during reactor power operation, the measured 1.
The average scram insertion time, based scram insertion times shall be extrapolated on the deenergization of the scram pilot to the reactor power operation condition valve solenoids as time zero, of all utilizing previously detennined correlations.
operable control rods in the reactor power operation condition shall be no greater than:
% Inserted from Average Scram Fully Withdrawn Insertion Times (sec)
Nu 5
0.375 O
20 0.900 50 2.000 90 3.500 N
0%
a Amendment No. J6, 47 3/4 3-4 I
- s%
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LJNITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2.
The average of the scram insertion times D.
Control Rod Accumulators for the three fastest control rods of all groups of four control. rods in a two by Once a shift, check the status in the two array shall be no greater than:
control room of the pressure and level alarms for each accumulator.
% Inserted From Average Scram Fully Withdrawn Insertion Times (sec) 5 0.398 20 0.954 50 2.120 90 3.800 3.
The maximum scram insertion time for 90%
insertion of any operable control rod shall not exceed 7.00 seconds.
D.
Control Rod Accumulators At all reactor operating pressures, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a:
1.
2.
Directional control valve electrically disarmed while in a non-fully inserted
, rsa position.
LeJ c:3 3.
Scram insertion greater than maximum pennissible insertion time.
p()
If a control rod with an inoperable accumulator sa is inserted " full-in" and its directional control l
valves are electrically disarmed, it shall not be considered to have an inoperable accumulator.
am n.inent No. 4 flovember 1,1974 3/4 3-5
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.
From and after the date that the FWCI subsystem 3.
When it is determined that FWCI sub-is made or found to be inoperable for any system is inoperable, the LPCI subsystem, reason, reactor operation is permissible only both core spray subsystems, the automatic during the succeeding seven days unless such pressure relief subsystems and the motor subsystem is sooner made operable, provided operated isolation valves and shell side that during such seven days all active com-makeup system for the isolation condenser ponents of the Automatic Pressure Relief system shall be demonstrated to be oper-Subsystem, the core spray subsystems, LPCI able immediately. The automatic pressure subsystem, and isolation condenser system relief subsystem and motor operated are operable.
isolation valves and shell side makeup system of the isolatian condenser shall 4.
If the requirements of 3.5.C cannot be met, be demonstrated to be operable daily an orderly shutdown shall be initiated and thereafter.
the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. Surveillance of the Automic Pressure Relief Subsystem shall be performed as follows:
D. Automatic Pressure Relief (APR) Sybsystems
- 1. During each operating cycle, the follow-1.
Except as specified in 3.5.D.2 and 3 below, ing shall be performed:
the APR subsystem shall be operable when-ever the reactor pressure is greater than
- a. A -imulated automatic initiation of i
90 psig and irradiated fuel is in the the system throughout its operating reactor vessel, sequence but excludes actual valve opening, and
- b. With the reactor at low pressure, each relief valve shall be manually opened until valve operability has been veri-fied by torus water level instrumenta-tion, or by an audible discharge r\\)
detected by an individual located l'd outside the torus in the vicinity of f(
each relief line.
u rNJ CD Amendment No. 61 3/4 5-6
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2.
From and after the date that one of the 2.
When it is determined that one safety!
three relief / safety valves of the auto-relief valve of the automatic pressure matic pressure relief subsystem is made or relief subsystem is inoperable the
- -- found to be inoperable when the reactor is actuation logic of the remaining APR l'
pressurized above 90 psig with irradiated valves and FWCI subsystem shall be fuel in the reactor vessel, reactor opera-demonstrated to be operable immediately tion is permissible only durino the and daily thereafter.
succeeding seven days unless repairs are made and provided that during such time E. Surveillance of the Isolation Condenser l
the remaining automatic pressure relief System shall be performed as follows:
valves, FWCI subsystem and gas turbine generator are operable.
1.
Isolation Condender System Testing:
3.
If the requirements of 3.5.D cannot be
- a. The shell side water level and met, an orderly reactor shutdown shall temperature shall be checked be initiated and the reactor shall be daily.
in a cold shutdown condition within 24 hou rs.
E. Isolation Condenser System 1.
Whenever the' reactor pressure is greater than 90 psig and irradiated fuel is in the reactor vessel, the isolation con-denser shall be operable except as specified in 3.5.E.2 and the shell' side water level shall be greater than 66 inches.
Nuo Ch rs) w Amendment No.15, 61 3/4 5-7
LiffilINfi CONDITION FOR OPERATION SURVEILLkNCEREQUIREMENT 2.
From and after the datd that the isola-h.
Simulated automatic actuation and tion concenser system is made or found to be inoperable for any reason, full functional system testing shall be power reactor operation is permissible perfonned during each refueling only during the succeeding 15 days unless outage or whenever major repairs such systen is sooner made operable,
'are cumoleted on the system.
provided that during such 15 days all The system heat removal capability c.
active canponents of the FWCI and gas turbine generator are operable (except shall be determined once every as stated in 3.9.8.2).
five years.
d, Calibrate vent line radiation 3.
If the requironents of 3.5.E cannot be met, an orderly shutdown shall be initiated monitors quarterly.
and the reactor shall be in a cold 2.
When it is determined that the isolation shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
condenser system is inoperable, the F.
Minimum Core and Containment Cooling System FWCl subsystem and gas turbine generator X[aliihil[ty, shall be demonstrated to be operable immediately and daily thereafter.
1.
Except as specified in 3.5.F.2, 3.5.F.3, F.
Surveillance of Core and Containment Cocling 3.5.r.7 and 3.5.F.8 below, both emer-System gency power sources shall be operabic whenever irradiated fuel is in the 1.
The surveillance requirements for normal reactor.
operation are in Section 3.9.
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TABLE 4.6.1 IN-SERVICE INSPECTION REQUIREMENTS FOR MILLSTONE UNIT 1 Examination Inspection During Ca_tegory, Examination Area Method 10-Year Interval Extent of Examinations (see Note 1)
A Longitudinal and circum-Note: Plant design ferential shell welds in does not provide core region capability to inspect B
Longitudinal and circum-Volumetric At or near end of 10-Top 10 feet of vertical vessel welds ferential welds in shell year interval (for 10%
accessible from I.D.
100% of each (other than those of of each longitudinal vertical weld in this region will be Categories A & C) and and 5% circumferential inspected during interval. 10% of each meridional and circum-length of each seam) longitudinal weld and 5% of the circum-ferential seam welds in ferential weld in the vessel closure head bottom head and closure will be inspected during interval.
head (other than those Note: Bottom head closure welds not of Category C).
applicable with present plant design.
C Vessel-to-flange and Volumetric Cumulative 100% cover-1/3 of vessel-to-flange and head-to-head-to-flange -
age by end of 10 years flange circumferential weld area every circumferential welds of each circumferential third refueling seam D (1)
Primary nozzle-to-vessel Volumetric Cumulative 100% cover-(1) Nozzle Welds welds and nozzle-to-age of nozzle-to-shell isolation Cond.
vessel inside radiosed weld. Core spray and Outlet (2) 1/5 years section feedwater inner radii Core Spray Inlet (2) 1/5 years will be inspected at Steam (4) 2/5 years high strain points.
Feedwater (4) 2/5 years Main steam and isolation Recirc. in (10) if condenser will be possible*
1/ year inspected 100%
Recirc. out (2) if possible*
1/5 years CRD retbrn 1/10 years
- See 4.6 Bases D (2)
Nozzle-to-vessel head Volumetric Cumulative 100% cover-(2) Nozzle-to-Vessel Head Welds welds and nozzle-to-head age of nozzle-to-shell Head Instrumenta-
'inside radiused section weld and 100% of inner tion (2) 1/5 years N
radius section of the 6"
nozzle-to-shell juncture llead Spray Inlet (1) 1/10 years os uu 3/4 6-6 U
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Planar Average Exposure (GWd/t) d Figure 3.11.1h MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.
PLANAR AVERAGE EXPOSURE. RELOAD Amendment no. 61 6 (8DRB265H and 8DRB265L) 3/4 11 -7b
- I -
ji.
i g
s LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C.
Hinimum Critical Power Ratio (MCPR)
C.
Minimum Critical Power Ratio (MCPR)
During power operation, MCPR shall be as shown in NCPR shall be detennined daily during Table 3.11.1.
If at any time during operation reactor power operation at > 25% rated it is determined by normal surveillance that the thennal power and following any change in limiting value for MCPR is being exceeded, action power level or distribution that would cause shall be initiated within 15 minutes to restore operation with a limiting control rod pattern operation to within the prescribed limits.
If as described in the bases for specification the s teady state MCPR is not returned to within
'3.3.B.S.
the prescribed limits within two (2) hnurs, 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, for core flows other than rated the MCPR's in Table 3.11.1 shall be multiplied by Kf, where K
is as shown in figure 3.11.2.
g D.
If any of the limiting values identified in Speci fica tions 3.ll.A. 8, or C. are exceeded, even if corrective action is taken, as pre-scribed, a Reportable Occurrence report shall be submitted.
Nu CD Ch
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nan.Ln.nt No. it, Jil, 211 3/4 11-9 e
TABLE 3.11.1 OPERATING LIMIT MCPR'S FOR CYCLE 7 Core Average Burn-up Rance Operating Limit MCPR 7 x 7 Fuel 8 x 8 Fuel 80C7 to E0C7 1.27 1.34 Coastdown beyond E0C7 1.27 1.34 (100% power to 70% power) l (Restricted to 100% flow)
End of Cycle is defined as end of full power life for the cycle Nu O
.,us t.P Amendment No. 28, 3#, 97, 61 3/4 11-10
Two sensors on the isolation condenser supply and return lines are provided to detect line failure and actuate isolation action. The sensors on the supply and return sides are arranged in a 1 out of 2. logic and to meet the single failure criteria, all sensors and instrumentation are required to be operable. The trip settings of 127 inches of water and 79 inches of water and valve closure times are such as to prevent core uncovery or exceeding site limits.
The instrumentation which initiates ECCS action is arranged in a dual bus system. As for other vital instrumenta-tion arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being perfonned.
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not I
decrease to < l.07.
The trip logic for this function is 1 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 RBM may be reduced by one for a short period of time to allow for maintenance testing and calibration.
I The APRM rod block trip is flow biased and prevents significant approach to MCPR=1.07 especially during operation at reduced flow. The APRM provides gross core 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 fuel 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 fecm reload to reload and, thus, each reload submtttal contains on update of the analysts. Below s 70%
l power, the withdrawal of single control rod results in MCPR > 1.07 without rod block action, thus requiring the RBM system 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 RBM response will be adequate to prevent rod withdrawal errors.
Nu The IRM rod block functions assure proper upranging of the IRM system, and reduce the probability of spurious scrams cD during startup operations.
ch A dwycale indication on an APRM or IRM is an indication the instrument has fatled or 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.
Amer.dment No. M, 61
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-63.
The peak fuel enthalpy content of 280 cal /gm is below the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental data as is discussed in refer-
^'
ence 1.
~~
Since Millstone Unit No. I has referenced the report, "GE/BWR Generic Reload Application for 8 x 8 Fuel, Rev.1, Supplement 4 (NED0-20360)," the assumptions regarding the control Rod Drop Accident are applica-ble to Millstone Unit No.1.
By using the analytical models described in this report coupled with conservative or worst-case input parameters, it has been determined that for power levels less than 20%
of rated power, the specified limit on in-sequence control rod or control rod segment worths will limit the peak fuel enthalpy content to less than 280 cal /gm. Above 20% power even single operator errors cannot result in out-of-sequence control rod worths which are sufficient to reach a peak fuel enthalpy content of 280 cal /gm should a postulated control rod drop accident occur.
Each core reload will be analyzed to show conformance to the limiting pa{gr eters.
A startup inter-assembly local power peaking factor. of 1.30 or less.
a.
b.
An end of cycle delayed neutron fraction of 0.005.
A beginning of life Doppler reactivity feedback.
c.
d.
The Technical Specification rod scram insertion rate.
s The maximum possible rod drop velocity (3.11 ft/sec).
e.
f.
The design accident and scram reactivity shape function.
g.
The moderator temperature at which criticality occurs.
(3) Stirn, R. C., Paone, C.
J., and Haun, J. M., " Rod Drop Accident Analysis of Large Boiling Water Reactor Addendum No. 2 Exposed Cores," Supplement 2 - NED0-10527 January 1975.
g (5) To include the power spike effect caused by gaps between fuel pellets.
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It is recognized that these bounds are conservative with respect to expected operating conditions.
If any one of the above conditions is not satisfied, a more detailed calculation will be done to show com-pliance with the 280 cal /gm design limit.
Should a control rod drop accident result in a peak fuel energy content of 280 cal /gm, less than 660 (7 x 7) fuel rods are conservatively estimated to perforate.
This would result in offsite doses twice that previously reported in the FSAR, but still well below the guideline values of 10 CFR 100.
For 8 x 8 fuel, less than 850 rods are conservatively estimated to perforate, which has nearly the same consequences as for the 7 x 7 fuel case because of the operating rod power differences.
The RWM provides automatic supervision to assure that out-cf-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences.
Reference Section 7-9 FSAR.
It serves as an independent backup of the normal withdrawal procedure followed by the opera-tor.
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 sequenc-ing 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 ro'ds for any startup.
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ihmever, there are various conditions under which the dissolved oxygen content of the reactor coolant water could be higher than 0.2-0.3 ppm, such as refueling and reactor startup.
During these periods with steaming rates less than 100,000 pounds per hour, a more restrictive limit of 0.1 ppm has been established to assure
,th'e chloride-oxygen combinations are maintained at conservative levels. At steaming rates of at least 100,000 pounds per hour, boiling occurs causing deaeration of the reactor water, thus maintaining oxygen
. concentration at low levels.
When conductivity is in its proper normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range. When conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values.
This would not necessarily be the case. Conductivity could be high due to the presence of a neutral salt; e.g., Na250 ;,
which would not have an effect on pH or chloride.
In such a case, high conductivity alone is not a cause for shutdown.
In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.
In the case of BWRs, however, where no additives are used and where neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water.
Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include operation of the reactor cleanup system, reducing the input of impurities and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to re-establish the purity of the reactor coolant.
During startup periods, which are in the category of less than 100,000 pounds per hour, conductivity may exceed 2 pmho/cm because of the initial evolution of gases and the initial addition of dissolved metals.
During this period of time, when the conductivity exceeds 2 pmho (other than short-term spikes), samples will be taken to assure that the chloride concentration is less than 0.1 ppm.
N The conductivity at the reactor coolant is continuously monitored.
The samples of the coolant which are u taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate O to assure accurate readings of the monitors.
If conductivity is within its normal range, chlorides and other m impurities will also be within their normal ranges. The reactor coolant samples will also be used to deter-mine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in y the' chloride ton content.
Isotopic, analyses to' determine major contributors to activity can be performed by a a gamma scan.
D.
Coolant Leakage The 2.5 gpm limit for leaks from unidentified sources was established by assumintf the leakage was from the primary system. Tests demonstrate that a relationship exists between the size of a crack and the probability l
that a crack will propagate.
B 3/4 6-3
For a crack size which gives a leakage rate of 2.5 gpri, the probability of rapid propagation is less than 10-5 A leakage rate of 2.5 gpm is detectable and measurable.
The 25 gpm limit on total leakage to the containnent was established by considering the :enoval capabilities of the pumps.
The capacity of either of the drywell sump pumps is 50 gpm and the capacity of either of the drywell equipment drain tank pumps is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
The performance of the reactor coolant leak detection system will be evaluated during the first year of commercial operation and the conclusions of this evaluation will be reported to the AEC.
The main steam line tunnel leakage detection system is capable of detecting small leaks.
The system per-fonnance will be evaluated during the first five years of plant operation and the conclusions of the evalu-ation will be reported to the AEC.
E.
Safety and Relief Valves Present dxperience with the new safety / relief valves indicates that a testing of at least 50% of the safety l
valves per refueling outage is adequate to detect failures or deterioration. The tolerance value is speci-fled in Section III of the ASME Boiler and Pressure Vessel Code as +1% of design pressure.
An analysis has been performed which shows that with all safety valves set 1% higher the reactor coolant pressure safety limit of 1375 psig is not exceeded.
The relief / safety valves have two functions; i.e., power relief or self-actuated by high pressure. The solenoid actuated function (automatic pressure relief) in which external instrumentation signals of coinci-dent high drywell pressure and low-low water level initiate the valves to open.
This functio'n is discussed in Specification 3.5.D.
In addition, the valves can be operated manually.
The safety function is performed by the same relief / safety valve with a pilot valve causing main valve opera-tion.
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When the setpoint is being bench checked, it is prudent to disassemble one of the relief / safety valves to examine for crud buildup, bending of certain actuator memoers,ur otner signs of possible deterioration.
Testing at low reactor pressure is required during each operating cycle.
It has been demonstrated that the i blowoown of the valve to the torus causes a wave action that is detectable on the torus water level instru-7 mentation.
The discharge of a safety valve is audible to an individual located outside the torus in the vicinity of the line, as experienced at other BWR's.
F[I Structural Integrity A preservice inspection of the component 3 listed in Table 4.6.1 will be conducted to establish a reference base for later inspections.
Construction oriented nondestructive testing is being conducted as systems are fabricated to assure freedom from defects greater than code allowance.
In addition, therfacility has been designed such that defects greater than code should not occur throughout plant life.
Infomation concerning the structural integrity of the reactor pressure vessel can be found in Appendix E to the FSAR. This Appendix contains documentation of design, fabrication, inspection, analysis and testing of this pressure vessel.
Design confimation and construction adequacy will be demonstrated during the plant startup and power ascen-sion test program. As part of this program, cold and hot vibration tests on certain reactor vessel internals will be performed. The tests, described in Amendments 17 and 18, are designed to obtain data on the unique design features of Millstone Unit 1 as compared to Dresden Unit 2 design. Thus, the basis for the Millstone vibration test program is predicted on obtaining necessary data to confirm comon design features shared with earlier BWR plants such as Dresden Unit 2.
In the event that data from these earlier plants are not available before routine power operation of Millstone Unit 1, the matter will be reviewed as indicated in Amendnent No. 23.
In order to monitor the integrity of the primary pressure boundaries throughout plant life, the inspection program stated in Table 4.6.1 was developed.
This program was developed using the ASME Code for In-Service Inspection as a basis and provides for exclusion of certain inspection parameters where current technology does not provide a means of inspecting or where equipment access and radiation hazards are of greatest significance. The initial inspection program was developed by Northeast Utilities Service Company with N
assistance from Southwest Research Institute and Teledyne Materials Research.
In early 1966, initial efforts W
were made to establish a nondestructive testing program for reactor vessel surveillance. Shortly after this
[
program initiation, the services of Southwest Research Institute and Lessels Associates, now Teledyne Mater-ials Research, were retained to aid in this program development. After considerable effort of all parties, v
the initial inspection program for Millstone Unit I was finalized in November 1968; this was one month after A
issuance of the first draf t of the Code for In-Service Inspection of Nuclear Reactor Coolant. Systems.
N Amendment No. 61 B 3/4 6-5
The initial inspection program took into account the mechanisms which might lead to a failure in the reactor pressure vessel and placed major inspection emphasis on the high-strain areas as determined by a general design evaluation and experience with similar systems. The intent of the inspection program was to detect flaws significant to possible brittle fracture and gross leakage.
The major premises of the program were:
1.
Selected areas of high strain would be inspected at periodic intervals 2.
Flaw initiation or growth would mean that additional areas would be inspected.
3.
Results of each inspection would be reviewed to determine need for further inspection.
4.
If required by analysis of Items 2 and 3, additional areas would be inspected.
~
Since formulating the initial program, additional information and experience has become available. The major information available is the current draft of the American Society of Mechanical Engineers Code for In-Service Inspection.
The experience available has been ot.tained from the operation of light-water moderated reactors around the world.
The revised in-service inspection program presented at this time is based on a thorough evaluation of present technology and state-of-the-art inspection techniques. The program will be continually reevaluated as technology expands in the field of nondestructive inspection and equipment development. After five years, an evaluation of the program will be presented to the AEC.
The interest of Northeast Utilities Service Company in the development of techniques for nondestructive testing of nuclear pressure boundaries is indicated by their participation in an Edison Electric Institute sponsored project.
This project is primarily aimed at developing new techniques for continuous in-service inspection monitoring, namely, the accoustic emission and accoustic spectrometer techniques.
The EEI program is also devoting smie funding to the improved conventional ultrasonic inspection techniques.
Ihe current Millstone Unit 1 in-service inspection program complies with the _ intent of the October 1968 draft of the Code for in-Service Inspection cf Nuclear Reactor Coolant Systems.
However, there are some g exception. to li teral compliance wi th this code.
These exceptions or exclusions are necessary because plant tja design doe. not provide the capability to inspect certain items. Millstone Unit I was designed by General c) Electric in mid-1965. Changes in design have been made in several areas to permit more compatibility with C7s the current inspection philosophy.
It is not possible, however, to make all changes that might be desired to insure. literal compliance with all areas of the current inspection code.
The areas of exclusion and reasons I[;forthisexclusionarediscussedbelow.
Category designations refer to Table 4.6.1.
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