ML19281A284
| ML19281A284 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/05/1979 |
| From: | Counsil W NORTHEAST UTILITIES |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19281A285 | List: |
| References | |
| NUDOCS 7903090291 | |
| Download: ML19281A284 (28) | |
Text
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a NOIZTHI!A!iT ITrII.lTII!!i P O BOX 270 HARTFORD. CONNECTICUT 06101 (203) 666-6911 k
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March 5, 1979 Docket No. 50-245 Director of Nuclear Reactor Regulation Atta:
Mr. D. L. Ziemann, Chief Operating Reactors Branch #2 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Gentlemen:
Millstone Nuclear Power Station, Unit No. 1 Provisional Operating License No. DPR-21 Reload #6 License Amendment Submittal Pursuant to 10CFR50.90, the holders of Provisional Operating License, DPR-21, hereby propose a change in the next refueled core as described in the attachnent, Reload 6, Supplemental Reload Licensing Submittal.
The reload fuel consists of 148 General Electric 8x8R bundles, with average bundle enrichment of 2.65 w/o, identical to fuel type 8DBR265HSL described in "CE/BWR Ceneric Reload Fuel Application", NEDE-24011-P-A, Revision 0, August, 1978. All 148 bundles will have drilled lower tie plates and finger springs to regulate bypass flow.
Fuel channels of an improved design based on developmental channels will be installed on all new reload fuel.
The channels are similar to those in-stalled during the past few refuelings except for a special heat-treatment which is intended to reduce channel corrosion.
The Technical Specification changes proposed in the attachment to this letter are derived from the reanalysis of certain limiting transients and accidents for the reloaded core configuration as described in the enclosed submittal.
The submittal contains a discussion of operation in a coastdown mode beyond the end of full power life. As the analysis shows, end-of-cycle limits are conservative operating conditions for this mode; and, therefore, no new specifications are required for coastdown to 70% power.
An additional analysis reported in the attached supplement demonstrates the ability to operate at end of cycle with a reduction in feedwater temperature. Thermal limits at the end of cycle conservatively bound operation in this mode.
Also included in this submittal is NEDE-20592-4P, "STR Bundle Submittal, Millstone Unit 1 Segmented Test Rod Bundle", Supplement 4, January, 1979. This supplement updates the original report, submitted October 3, 1974, as part of the Reload No. 2 license amendment submittal, which was supplemented July, 1975, August, 1976, and November, 1977, and describes proposed changes to be made in the bundle composi-tion during the spring, 1979, refueling outage. We have been informed by the General Electric Company that this information is considered proprietary by them, and as such, it is being forwarded under separate cover.
79030902'll
. The present schedule calls for plant shutdown on April 21, 197').
A five-week outage is anticipated.
The Nuclear Review Board has reviewed and approved the Technical Specification changes for this Reload No. 6 License A=endment Submittal.
NNECO has reviewed the above proposed License Amendment pursuant to the requirements of 10CFR170, and has determined that the proposal constitutes a Class IV amendment. Accordingly, enclosed herewith is payment in the amount of $12,300 (twelve thousand three hundred dollars). The basis for this determination is that the proposal involves several changes of the Class III type.
Very truly yours, NORTHEAST NUCLEAR ENERGY C051PANY l
]{}l0L' W. G. Counsil Vice President Attacinents Submittal includes three signed and notarized originals and 37 copies.
STATE OF CONNECTICUT )
)
ss. Berlin 2y 6 /97/
COUNTY OF HARTFORD
)
Then personally appeared before me W. G. Counsil, who being duly sworn, did state that he is Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute snd file the foregoing information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.
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Not"ary Public My Commission Egires March 31, 1031
DESCRIPTION OF TECilNICAL SPECIFICATION CllANCES TiiERMAL LIMITS Maximum Average Planar Linear lieat Generation Rate and Minimum Critical Power
- Ratios, Section Page Figure 3.11.1h 3/4 11-7b Table 3.11.1 3/4 11-10 Bases 2.1.1 B2-1, B2-2 Bases 2.1.2A, B, E, F B2-6, B2-7, B2-8 Bases 3.2 B3/4 2-3 The revised operating limit MAPLHGR's and CPR's are the result of transient analysis performed with the Cycle 7 core.
The limits are supported by the at tached topical reports NEDO-24168 and NEDO-24168-1, Supplemental Reload Licensing Submittals for Millstone 1, Reload 6.
A technical review of these specifications has found them to be acceptable; safety evaluation has also been performed in accordance with 10CFR50.59 and has a
concluded that these changes do not constitute any unreviewed safety questions.
The Millstone Nuclear Review Board has reviewed and approved the proposed changes and has concurred with the above determination.
CONTROL ROD WITHDRAWAL Control Rod Drop Accident 0.013 delta k Sec tio n Page 3.3.B.3 3/4 3-3 Bases 3.3.B.3 B3/4 3-3 B3/4 3-4 The GE Generic Reload Fuel Licensing Topical Report, NEDE-240ll-P-A provides justification fer removing the 0.013 delta k requir'ement from the Control Rod Drop Accident evaluation.
The justification is given on Page 5-47 of NEDE-240ll-P-A and in Section 7-3 of the NRC SER (contained in Appendix C cf the topical).
The topical indicates it is unrealistic to set a specific value of maximum control rod worth, since this is only one of many parameters in the peak fuel enthalpy calculations for a control rod drop accident.
The change does not alter the requirement to satisf y the ultimate 280 cal /gm design limit.
A technical review of these specifications has found them to be acceptable; a safety evaluation has also been performed in accordance with 10CFR50.59 and has concluded that these changes do not constitute any unreviewed safety questions.
The Millstone Nuclear Review Board has reviewed and approved the proposed changes and has concurred with the above determination.
SAFETY / RELIEF VALVES Reactor Coolcat System and Automatic Pressure Relief Subsystem Section Page 2.2.2.B 2-6 Bases 2.2.1 B2-10 Bases 2.2.2 B2-11 4.5.D.l.a 3/4 5-6 3.5.D.2 3/4 5-7 4.5.D.2 3/4 5-7 3.6.E 3/4 6-5 4.6.E 3/4 6-5 3/4 6.E Bases B3/4 6-4, B3/4 6-5 These changes to the Technical Specifications envelop and supersede the changes proposed by NNECO in D. C.
Switzer's letter to G. Lear dated January 27, 1978.
The changes in t ha t previous submittal cre, therefore, withdrawn and replaced with the attached changes.
The previous changes submitted by NNECO on January 27, 1978 provided increased surveillance of the Automatic Pressure Relief (APR) and Safety Relief Valves (SRV) in order to improve valve performance and overall plant safety. The attached changes envelop these and also provide, 1) clarification of the surveillance requirements for the APR valves to delete the requirement to mechanically lift the APR's in Section 4.5.D.l.a since manual actuation is required by Section 4.5.D.l.b (consistent with BWR STS), and 2) staggered SRV setpoints to prevent any potential of repeated simultaneous popping, thereby reducing torus loading, and 3) increased SRV setpoints to improve valve simmer margin.
Our technical review of these changes also considered the following :
(1) The effect on the torus of the proposed change of the SRV setpoints has been evaluated.
A.
The increased setpoints reduce the probability of spurious actuation of the valves, thus, decreasing the clearing load cycles on the torus.
The reduced probability of spurious opening likewise reduces the possibility of a stuck open SRV and any stresses associated with such condition.
B.
As reported in GE report NEDC-21581-P, "Monticello Saf ety Relief Valve Discharge Load Test", the loads or the torus wall for a second actuation (hot pop) of an SRV are greater than for a first actuation of the valve.
Furthermore, as reported in the attachments to GE letter MI-G-179, the ef fect of several valves opening simultaneously is increased loads compared to one valve.
The staggered setpoints assure that for various reac-tor isolation conditions, only one valve would open more than once.
The staggered setpoints, therefore, assure that for the hot pop (controlling loading case), only one valve would open; and there would be no additive effect of the hot pop loads.
C.
The increased setpoints can result in an increase in the reaction load where the ramshead is attached to the torus ring girder.
The magnitude of this load is on the order of 50 kips, and the increase because of a setpoint change would be on the order of 10 percent or 5 kips.
This cbange can then be compared to pool swell down loads and how they have ocen reduced because of recent testing compared to the Short-Term Program.
The STP reference plant net downward pressure was 21 psi (Figure 3-11, NEDC-20989-P, June, 1976) which resulted in an average torus column load of approximately 650 kips (Teledyne report TR-2138 (a), " Plant Unique Analysis", July 26, 1976).
More recent testing reported in NEDM-21688-P, " Preliminary Load Evaluation Report", August, 1977, indicates a maximum downward pressure of less than 12 psi.
Since the short-term criteria on the torus, in particular, the ring girder and columns, was satis-fled for the 21 psi load and since the down load has been signifi-cantly reduced to 11 psi, it obviously follows that any possible increase in the ramshead reaction load is less than the reduction in downward pool and, therefore, of no consequence.
(2) The ef f ect on the main steam and relief-s 'lve piping has been evaluated.
A.
Teledyne Engineering Services (TES) has reviewed the ef fects of resetting the actuation points of the Millstone Unit No. 1 Main Steam Safety Relief Valves. The following setpoints were recommended by TES.
Relief Line MS-8a
- no change (1095 psig)
"alief Line MS-8d
- reset to 1110 psig Relief Line MS-8b, 8c, 8e, and 8f - reset to 1125 psig This review was based on a detailed evaluation of dynamic piping stresses, which were done previously by TES.
B.
The revised main steam relief valve actuation setpoints have been reviewed to assure that the increased dynamic load does not adversely affect the structural integrity of the main steam and relief valve piping system.
All six (6) relief lines and the associated steam lines were i nalyzed ior dynamic loading as a part st the EOC modifi-cations made in 1975.
The analytical technique used to calculate the hydrodynami - loading was state-of-the-art at that time, however, more recent t _chniques indicate the presence of conservatism beyond that ant ic ipt.ted. A margin did exist at that time between the calculated stress and the conservatively based ASME Code limits. A review of this margin and the slightly increased setpoints in light of present-day analytical techniques indicates that the ASME Code limits will not be compromised.
C.
It should be noted here, that the addition of quenchers to the relief lines is planned for a future modification.
The effects that the quenchere have on the pipe stresses and pressure setpoints will be
the topic of a TES analysis scheduled to begin in the immediate future.
Additionally, further analysis may have to be initiated to further increase the simmer margin of the SRV's as recommended in GE SIL No. 196, supplement 3.
The attached NEDO-24168 and NEDO-24168-1, Supplemental Reload Licensing Submittal for Millstone 1, Reload 6, include the staggered and eleva ted SRV setpoints proposed herein.
Thus, the transient and accident response has been evalcated and the results provide assurance that the appropriate limits have not been exceeded with tha revised setpoints.
General Electric evaluations of Millstone 1 submitted to NNECO conclude that any grouping of SRV setpoints which have one valve set at least 10 psi below the other five results in only one valve actuating on the second pop.
Conse-quently, the GE parametric study serves to af firm that the proposed setpoints staggered 1-1-4 at 15 psi margin, will prevent a second actuation of all but one valve.
A technical review of these specifications has found them to be acceptable; a saf ety evaluation has also been performed in accordance with 10CFR50.59 and has concluded that these changes do not constitute any unreviewed safety questions.
The Millstone Nuclear Review Board has reviewed and approved the propoced changes and has concurred with the above determination.
DOCKET No. 50-245 PROPOSED TECIINICAL SPECIFICATION CIIANGES FOR MILLSTONE 1, RELOAD 6 PEBRUARY, 1979
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS
~
2.2.1 REACTOR COOLANT SYSTEM 2.7.2 REACTOR COOLANT SYSTEM Appl ica bil i ty:
Applica bil i ty :
Applies to limits on reactor coolant systen 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 be.ow which the integrity of Objective:
the reactor coolant system is not threatened due to an overpressure condition.
To define the level of the process variables at which automatic protective action is initiated Specification:
to pres ent exceeding the safety limits.
The reactor vessel pressure shall not exceed Speci fi ca tion :
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:
N g__._ p f V a_l v e_s_
Se_tPoint(PSIGl 1
1095 + 1:
1 1110 i 1-4 112511:
/
2-6 Amendment No. 2'0
2.1.1 Bases The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient.
Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the minimum critical power ratio (MCPR) is no less than 1.07.
MCPR > 1.07 represents a conservative margin relative to the conditions required to maintain fuel cladding 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 from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety setting. 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 thermal stresses pay cause gross rather than incremental claddina 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 transition 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 mir.imum critical power ratio (MCPR).
It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variable, i.e., normal plant operation presented on Figure 2.1.2 by the nominal expected flow l
control line.
The Safety Limit (MCPR of 1.07) has sufficient conservatism to assure that in the evcnt of an abnormal operational transient initiated from a normal operating condition more than 99.9% of the fuel rods in the core are expected to avoid boiling transition.
The margin between MCPR of
'.0 (onset of transition boiling) l and the safety limit (MCPR = 1.07) is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as described in Refecence 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. 1/5, 34' B2-1
Because the boiling transitin e vrelation is based on a large quantity of full scale data there is a very l
high confidence that operation of a fuel assembly at the condition of MCPR = 1.07 would not produce boiling transi tion.
However, if boiling transition were to occur, clad perfoi, tion would not be expected. Cladding temperatures would increase to approximately 1100 F which is below the perforation temperature of the cladding material.
This has been verified by tests in the General Electric Test 0.eactor (GETR) where fuel similar in design to fiillstone operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation. Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity. The limit of applicability of the boiling transition correlation is 1400 osia during normal power operation. However, the reactor presture is limited as per Specification 2.2.1.
l In addition to the boiling transition limit (MCPR = 1.07) operation is constrained to a maximum LHGR= 17.5 kW/f t for 7 x 7 and 13.4 kW/ft for 8 x 8.
At 100% power this limit is reached with a maximum total peaktag factor (MTPF) of 3.08 for 7 x 7 fuel and 3.04 for 8 x 8 fuel.
For the case of the t1TPF 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 Speci fica tion 2.1.2. A.l.
At pressures below 800 psia, the core evaluation pressure drop (0 power, O flow) is greater than 4.56 psi.
At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 103 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and gas a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 10 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.
Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 2.1.l A or 2.1.lB will not be exceeded.
Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage.
However, for this specification a Safety Limit Polation will be assumed when a scram is only accomplished by means of a backup feature of the plant design.
The con-cept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.
B2-2 Amendment No. 9, M
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 U1GR 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 initiated 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 APRf1 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 Responst. to Questions A-12, A-14, A-15, and D-3.
For operation in the refuel or startup/ hot standby modes while the reactor is at low pressure, the APRM reduced flux trip scram setting of < 15% of rated power provides adequate thermal margin between the 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 unifom 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 adequate to assure a scram before the power could exceed the safety limit. The APRM reduced trip scram remains active until the mode switch is placed in the run position.
This switch occurs when the reactor pressure is greater than 880 psig.
The IRM trip at <_ 120/125 of full scale remains as a backup feature.
Amendment No. M, 11 82-6
The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.
During steady-state operation with one recirculation pump operat-ing the equalizer lire shall be open.
Analyses of transients from this operating condition are less severe than the same '.ransients from the two pump operation.
B.
APRM Control Rod Block Trips _
Reattor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent 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 rcte, prevents an increase in tw 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 aesumed for the analysis was 3.03 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 2.1.2B if peaking factors greater than 3.08 for 7x7 fuel and 3.04 for 8x8 fuel exist, thus preserving the APRM rod block safety margin.
The APRM rod 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 ba. s for the safety limit is maintained.
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 below the clad melting temperature to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%.
To accomplish this function, the capacity of each emergency core cooling system component was established based on the reactor low low water level.
To lower the setpoint of the low water level scram would require an increase in the capacity of each of the ECCS components.
Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of ECCS capacity requirements.
B2-7 Amendment No. U, 77
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 ECC 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 l
MCPR remains above 1.07 ever during the worst case transient that assumes the turbine bypass is closed.
This scram is bypassed when turbine steam flow is < 45% of rated, as measured by the turbine first stage pressure.
F.
Turbine Control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and sub-sequent failere of the bypass; i.e., it prevents MCPR from becoming less than 1.06 for this transient.
For the load
- ejection 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 belov 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 accommodate the full load rejection capability, this scram trip must be bypassed because it would be actuated and would scram the reactor during load r7jections.
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.
Amendmen' No.p, fg B2-8
2.2.1 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 a t 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, Sectian IIIfor the pressure vessel and isolation condenser and USAS B31.1 Code for the reactor coolant systen piping.
The ASME Boiler and Pres-sure Vessel Code pennits pressure transients up to 10? over design pressure (110! x 1250 = 1375 psig),
and the USASI Code permits pcessure transients up to 20" over the design pressure (1201 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 af 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 comparable 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 turbine bypass system limits the pressure to less than 1085 psig.
The safety / relief valves are set at 1095 psig,1110 psig, and 1125 psig and are sized to keep the reactor coolant system pressure below 1375 psig with no credit taken for the turbine bypass system.
Credit is taken for the neutron flux scram, however.
During operation, reactor pressure is continuously displayed in the control room on a 0-1500 psig pressure recorder.
Amendment No.
B2-10
2.2.2 Bases In compliance with Section III of the ASf1E Boiler and Pressure Vessel Code,1965 Edition, the specified settings of the pressure relieving devices are below 103:e of design pressure. As described in the General 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 i1SIV 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.
Anendment No-B2-11
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 201 rated thenral operable, the following s teps must power, no control rods shall be moved be performed.
unless the rod worth 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 af ter 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 rad block function of the rod per second.
worth minimizer shall be verified by attempting to withdraw ari out-of-sequence control rod be-yond the block point.
(b)
If the rod worth minimizer is inoperable while the reactor is in the startup or run rode below 10 ~ rated thermal power, and a second independent operator or engineer is being used, he shall verify that all rod positions are correct prior to comrencing withdrawal of each rod croup.
i.T:endment No. 22, /0 3/4 3-3
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.
From and after the date that the FWCI subsyste, 3.
When it is detennined that FWCI sub-is made or found to be inoperable for any systen is inoperable, the LPCI subsysten, reason, reactor operation is pennissible only both core spray subsystems, the automatic during the succeedino 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-Su bsys tem, the ccre spray subsystems, LPCI able immediately. The automatic pressure subsystem, and isolation condenser system relief subsystem and notor operated are operable, isolation valves and shell side makeup systen of the isolation condenser shall 4.
If the requirenents of 3.5.C cannot be net, 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 Subsysten shall be performed as follows:
D. Automatic Pressure Relief ( APR]_Sy_bsystems
- 1. During each operating cycle, the follow-1.
Except as specified in 3.5.D.2 and 3 below, ino shall be performed:
the APR subsysten shall be operable when-ever the reactor pressure is greater than
- a. A simulated automatic initiation of 90 psig and irradiated fuel is in the the systen 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 detected by an individual located outside the torus in the vicinity of each relief line.
Amendment No.
3/4 5-6
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2.
From and af ter 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 subsysten is made or relief subsysten is inoperable the found to be inoperable when the reactor is actuation logic of the remaining APR pressurized above 90 psig with irradiated valves and FWCI subsystem shall be fuel in the reactor vessel, reactor opera-demor.strated to be operable immediately tion is pennissible only durino the and daily thereafter.
succeeding seven days unless repairs are made and provided that during such time E. Surveillance of the Isolation Condene.er l
the remaining automatic pressure relief System shall be performed as follows:
valves, FWCI subsysten 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 irradicted 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.
huend.nent No. 76 3/4 5-7
LIMITING CONDITION FOR OPEPATION SURVEILLANCE REQUIREMENT coolant system leakage into the primary con-E.
S_afety and Relief Valves tainment shall not exceed 25 gpm.
If these conditions cannot be met, initiate an orderly 1,
Three of the relief / safety valves top shutdown and have the reactor in the cold works shall be bench checked or replaced shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
with a bench checked top works each re-fueling outage. All six valves top works E.
Safety and Relief Valves shall be checked or replaced every two refueling outages.
The set pressure 1.
During power operation and whenever the shall be adjusted to correspond with a reactor coolant pressure is greater than steam set pressure of:
90 psig, and temperature greater than 320 F, the safety valve function of the No. of Valves Set Point (psic) six relief / safety valves shall be operable.
l (The solenoid activated relief function of 1
1095 + 1:
the relief / safety valves shall be operable l
1110 + 1; as required by Specification 3.5.D) 4 1125{1:
2.
At least one of the relief / safety valves 2.
If Specification 3.6.E.1 is not met, initiate shall be disasserbled and inspected each an orderly shutdown and have the reactor refueling outage.
coolant pressure below 90 psig and terpera-ture below 320 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
During each operating cycle with the reactor at low pressure, each safety valve shall be manually opened until operability has been F.
Structural Integrity _
verified by torus water level instrumenta-tion, or by an audible discharge detected The structural integrity of the prinary system by an individual located outside the torus boundary shall be maintained at the level re-in the vicinity of each discharge.
quired by the original acceptance standards throughout the life of the plant.
F.
Structural Integrity The nr., destructive inspections listed in Table 4.6.1 shall be perforced as specified.
The results obtained from compliance with this specification will be evaluated af ter 5 years and the conclusions of this evaluation will be reviewed with the AEC.
Amendment No. 20 3/4 6-5
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5.
10.
15.
20.
25.
30.
Planar Average Exposure (GWd/t)
Figure 3.11.1 h MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.
Amendmen t no.
PL ANAR AVERAGE EXPOSURE.RE LOAD 6 (8DRB265H and 8DRB265L) 3/4 11 -7 b
TABLE 3.11.1 OPERATING LIMIT MCPR'S FOR CYCLE 7 Core _ Ave _ rage Burn-up Range Operating Lir:it MCPR 7 x 7 Fuel 8 x 8 Fuel BOC7 to EOC7 1.27 1.34 Coastdown beyond EOC7 1.27 1.34 (1007 power to 70? power)
(Restricted to 100:' flow)
End of Cycle is defined as end of full power life for the cycle Amendment No. 2/,, 31, 17 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 performed.
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.
l 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 withdraw 31 is blocked within an adequate margin to fuel damage limits.
This margin varies slightly from reload to reload and, thus, each reload submittal contains on update of the analysis.
Below m 707 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.
The IRM rod block functions assure proper upranging of the IRM system, and reduce the probability of spurious scrams during startup operations.
A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough or the neutron flux is belnw 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.
Amendment No. Y6 B 3/4 2-3
3.
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. 1 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 detennined 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 /qm should a postulated control rod drop accident occur, l
Each core reload will be analyzed to show conformance to the limiting pa r 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.
The maximum possible rod drop velocity (3.11 ft/sec).
e.
f.
The design accident and scram reactivity shape function.
9 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.
(5) To include the power spike effect caused by gaps between fuel pellets.
Amendment No. 22, 40 B 3/4 3-3
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 cuideline 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 tne operating rod power differences.
The RWM provides automatic supervision to assure that out-of-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences.
Reference Section 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 wher 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 af ter 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 207 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 RW'A a va il abil i ty, the RWM is required to be operating during a startup for the withdrawal of a significant number of control rods for any startup.
I Amendment No. 22, M
B 3/4 3-4
For a crack size which gives a leakage rate of 2.5 gpm, 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 containment was established by considering the removal 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 concercial 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 leak _.
The system per-formance 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 experience 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-fied in Section III of the ASME Boiler and Pressure Vessel Code as 11' 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 opcn.
This function 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.
Amendment No.
B 3/4 6-4
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 members,or other signs of possible deterioration.
Testing at low reactor pressure is required during each operating cycle.
It has been demonstrated that the blowdown of the valve to the torus causes a wave action that is detectaie on the torus water level instru-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.
Structural Integrity A preservice inspectica of the components 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 eddition, the facility has been designed such that defects greater than code should not occur throughout plant life.
Infornation 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 confirmation 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 common 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 Amendment 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 wnere 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 assistance from Southwest Research Institute and Teledyne Materials Research.
In early 1965, initial efforts 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.
Af ter considerable effort of all parties, the initial inspectior program for Millstone Unit I was finalized in November 1968; this was one month af ter issuance of the first draf t of the Code for In-Service Inspection of Nuclear Reactor Coolant Systems.
Amendment No.
B 3/4 6-5