ML20138C815
| ML20138C815 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 11/27/1985 |
| From: | Lear G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20138C713 | List: |
| References | |
| NUDOCS 8512130096 | |
| Download: ML20138C815 (20) | |
Text
. .__ - -
ne.g% UNITED STATES
[ h* NUCLEAR REGULATORY COMMISSION ,
g a wAsHmorow o.c.nossa
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YANKEE ATOMIC ELECTRIC COMPANY i
DOCKET NO. 50-29 YANKEE NUCLEAR POWER STATION i'.
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AMENDMENT TO FACILITY OPERATING LICENSE
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Amendment No. 88 License No. DPR-3
- ,t ji' 1. The Nuclear Regulatory Commission (the Commission) has found that:
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A. The application for amendment by Yankee Atomic Electric Company
- (the licensee) dated August 30, 1985, complies with the ij- standards and requirements of the Atomic Energy Act of 1954,
'l as amended (the Act), and the Commission's rules and regulations
- l 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; -
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,, C. There is reasonable assurance (i) that the activities authorized
- by this amendment can be conducted without endangering the health a 4
and safety of the public; and (ii) that such activities will be i conducted in compliance with the Commission's regulations; ,
!- D. The issuance of this amendment will not be inimical to the common if . defense and security or to the health and safety of the public; and
- i I! 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.
k 8512130096 851203 j PDR ADOCK 05000029 a
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- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-3 is hereby amended to read as follows:
(2) Technical Specifications
)
The Technical Specifications contained in Appendix A as j revised through Amendment No. 88, are hereby incorporated i in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance. '
f-t FOR THE NUCLEAR REGULATORY COMMISSION A
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! "Geor'g'5: E'. Lea'r. Director l
' fMR Project 31rectorata #1 "Myision of'tH* Licensing-A
Attachment:
Changes to the Technical Specifications
,. Date of Issuance: November 27, 1985 ,
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l ATTACHMENT TO LICENSE AMENDMENT NO.88 ,
FACILITY OPERATING LICENSE NO. DPR-3 DOCKET NO. 50-29 -
l Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3/4 1-2a 3/4 1-2a ,
3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3
>' 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6
, 3/4 2-7 3/4 2-7
+ -
3/4 2-12 3/4 2-12
' i 3/4 2-13 3/4 2-13 3/4 3-14 3/4 3-14 3/4 9-1 3/4 9-1 83/4 1-1 B3/4 1-1 B3/4 2-1 83/4 2-1 B3/4 2-2 -
83/4 2-2 B3/4 2-3 B3/4 2-3 83/4 2-4 B3/4 2-4 I: 5-1 5-1 =
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REACTIVITY CONTROL SYSTEMS i LIMITING CONDITION FOR OPERATION
' 3.1.1.1.2 The shutdown margin requirement is as follows:
4
- 1. 0 490 F 1 Tavs Shutdown Margin 1 6%Ak/k for Tavs = 5160F
- 2. 3300F 1 Tavs < 4900F Shutdown Margin 1 5%Ak/k for Tayg = 3300F I
APPLICABILITY: MODE 3*
I ACTION: I With the SHUTDOWN MARGIN less than required, immediately initiate and continue boration at 1 26 gym of 2200 ppa boron concentration or equivalent until the i
required SHUTDOWN MARGIN is restored.
SURVEILLANCE REQUIREMENTS .
1 j: __ 4.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal 1 to that required: .
- s. When in Mode 3, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the
. _ . O following factors:
I 1. Main Coolant System boron concentration,
! 2. Control rod position, j 3. Main Coolant System average temperature,
- 4. Fuel burnup base'd on gross thermal energy generation.
- 5. Xenon concentration, and
- 6. Samarium concentration.
+
4.1.1.1.2.2' During a reactor startup in which core reactivity or control positions for criticality'are not established, a plot of inverse *
'{-- multiplication rate (or count este) versus reactivity shall be made.
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- see Special Test Rxception 3.10.1.
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. YANKEE-ROWE 3/4 1-2a Am Adment No. 88 i
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POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) s.
4.2.1.2 The below factors shall be included in the calculation of peak full power LHGR:
- a. Heat flux power peaking factor, F , measured using incore instrumentation at a power 1 10%.
- b. The multiplier for xenon redistribution is a function of core lifetime as given in Figure 3.2-3. In addition, if Control Rod Group C is inserted below 80 inches, allowable power may not be regained until power has been at a reduced level defined below for at least twenty-four hours with Control Rod Group C between 80 and 90 inches.
Reduced power = Allowable fraction of full power times multiplier given in Figure 3.2-4.
4 Exceptions: 1. If the rods are inserted below 80 inches and power does i not go below the reduced power calculated above, hold at 6L the lowest attained power level for at least twenty-four hours with Control Rod Group C between 80 and 90 inches I
before returning to allowable power.
. 2. If the rods are inserted below 80 inches and zero power f- is held for more than forty-eight hours, no reduced -
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f power level need be held c;. tha way to the allowable t fraction of full power.
( -
I c. Shortened stack heigh't factor,1.009.
' - d. Measurement uncertainty:*
- 1. 1.05, when at least 17 incore detection system neutron detector l=
+ thimbles are OPERABLE, or
- 2. 1.068, when less than 17 incore detection system neutron detector thimbles are OPERABLE.
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YANKEE-ROWE 3/4 2-2 Amendment No. [ [ J Z 88 1
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[ POWER DISTRIBUTIfB LIMITS L
- SURVEILLANCE REOUIREMENTS (Continued)
- e. Power level uncertainty *,1.03.
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- f. Heat flux engineering factor *, F , 1.04. -
- g. Core average linear heat generation rate at full power.
i *The Factors d, e and f will be combined statistically as the " root-sum-square"
' of the individual parameters.
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ALLOWABLE PEAK ROD LHGR VERSUS CYCLE BURNUP i
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i YANKEE-ROWE 3/4 2-5 Amendment No[f',%/f,88 l
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. Multiplier for Xenon Redistribution as a j
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- POWER DTSTRIBUTION LIMITS
,. DNB PARAMETERS I
LIMITING CONDITION POR OPERATION i
3.2.4 The following DNB related parameters shall be maintained within the l limits shown on Table 3.2-1: l
- a. Highest Operating Loop Cold Leg Temperature
- b. Main Coolant System Pressure
- c. Main Coolant System Total Flow Rate APPLICABILITY: MODE 1 ACTI.ON:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of ,
RATED THERMAL' POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
I -
'I SURVEILLANCE REQUIREMENTS 4.2.4.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.4.2 The Main Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.
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l YANKEE-ROWE 3/4 2-12 Amendment No. 88 l l
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i TABLE 3.2-1
[ -
DNB Paramet rs .
LIMITS i 4 Loops in Operation Parameter 0
Operating Loop Cold Les Temperature 1 520 F,'
Main Coolant System Pressure 1 1950 psig*
Main Coolant System Total Flow Rate 1 38.3 x 106.lb/hr
- Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED THERMAL POWER Per minute or a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER. ,
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YANKEE-ROWE 3/4 2-13 Amendment No. 43, 88 4
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4 TABLE 3.3-3 Engineered Safeguards System Instrumentation Trip S9tonintn
- 1 Functional Unit Trip Setpoint
- 1. SAFETY INJECTION .
- a. Actuation Channel #1
- 1) RPS Low Main Coolant - Loop 1 1 1650 psis Pressure Channel i 2) High Containment Pressure Sensor s 5 psig
- 3) Manual Initiation Not Applicable
- b. Actuation Channel #2 ,
- 1) RPS Low Main Coolant - Loop 2 1 1650 psig ?
Pressure Channel r 2) .High Containment Pressure Sensor s 5 psig
- 3) Not Applicable f Manual Initiation
! 2. CONTAINMENT ISOLATION I
- a. Manual Initiation Not Applicable
! b. Actuation Channel A
- 1) High Containment Pressure Sensor s 5 psig
, - 2) Safety Injection (All Safety Injection, Setpoints) t 5
- c. Actuation Channel B f__
l 1) High Containment Pressure Sensor s 5 psig
- 2) Safety Injection (All Safety Injection Setpoints)
- 3. MAIN STEAM ISOLATION
- a. Low Steam Line Pressure 1 200 psig
- b. Automatic Trip Logic Not Applicable
- c. Manual Initiation Not Applicable
- d. High Containment Pressure s 5 psig Trip-Containment Isolation YANKEE-ROWE 3/4 3-14 Asaendment No. ,5( .f4 88 ,
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REFUELING OPERATIONS
$ REACTIVITY 1,IMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Main Coolant system and the shield tank cavity shall be maintained uniform and sufficient to ensure a k gg of 2
0.93 calculated or less.
APPLICABILITY: MODE 6*
ACTION:
- a. With the boron concentration requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at 1 26 spa of 2200 ppm boron concentration or its
! equivalent until K,gg is reduced to 1 0.93.
, I
.h b. With a significant unexpected increase in the count rate on any channel, or an unexpected increase in the count rate by a factor of two after addition of a new fuel assembly or removal of a control rod,' suspend CORE ALTERATIONS until the situation is reviewed by ;
plant technical supervisory personnel.
"' c. The provisions of Specification 3.0.3 are not applicable.
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'f *The reactor shall be maintained in MODE 6 when the reactor vessel head is j unbolted or removed with fuel in the vessel.
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YANKEE-ROWE 3/4 9-1 AmendmentNo/T,'88 4
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3/4.1 NTIVITY CONTROL BYexure mASES
- 3/4.1.1 BORATION CONTROL 1
3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently '
suberitical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements are a function of the plant operating status. For critical conditions, minimum shutdown margins are limited by the 0 Power Dependent Insertion Limits (PDIL) as given in Figure 3.1-2. For 490 F 1 T,yg, the requirement for a SHUTDOWN MARGIN is established by postulated steam line break considerations with ECCS and NRVs available and covers the 4900F, requirements to preclude inadvertent criticality. For 330 1 Tav
'j therequirementforaSHUTDOWNMARGINissufficienttoprecludekn< advertent
[
criticality and covers the requirements of steam line breaks with automatic
- initiation of ECCS and WRVs blocked. With Tavs < 3300F, the reactivity lly -- transients resulting from a steam line break cooldown are minimal. 5%Ak/k
- SHUTDOWN MARGIN (with all rods inserted) provides adequate protection to l _
preclude criticality for all postulated accidents with the reactor vessel head
.! in place.
I To eliminate possible errors in the calculations of the initisi j; reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, saast be adjusted (normalized) to accurately reflect actual sore conditions. Normally, when fuel power is
- ' reached after each refueling, and with the control rod groups in the desired
- i positions, the boron concentration is measured and the predicted steady-state curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope *
,{= of the curve relating burnup and reactivity is compared with that predicted.
This process of normalisation should be completed af ter about 10% of the total i 'e
! core burnup. Thereaf ter, actual boron concentration can be compared with t - prediction and the reactivity status of the core can be continuously evaluated, and any deviation would be thoroughly investigated and evaluated.
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l YANKEE-ROWE 33/4 1-1 Amendment No. g 88 4
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i 3/4,2 POWER DISTRIBUTION LIMITS BASES T
The specifications of this section provide assurance of fuel integrity during Conditions I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core A 1.30 during normal operation and in short teria transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical
< properties to within assumed design criteria.
3/4.2.1 PEAX LINEAR HEAT GENERATION RATE Limiting the peak Linear Heat Generation Rate (LHGR) during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance critaria limit of 22000 F is not exceeded.
o When operating at constant power, all rods out, with equilibrium menon, ,
- power peaking in the Yankee Rowe core decreases monotonically as a function of
!_~
cycle burnup. This has been verified by both calculation and measurement on Yankee cores and is in, accord with the expected behavior in a core that does
. not contain burnable poison. The all-rods-out power peaking measured prior to exceeding 75% of RATED THERMAL POWER af ter each fuel loading thus provides an f-upper bound on all-rods-out power peaking for the remainder of that cycle.
Thereaf ter the zessured power peaking shall be checked every 1000 equivalent
- j>'
full power hours and the latest measured value shall be used in the computation. The only effects which can increase peaking beyond this value
, would be control rod insertion and xenon transients and these are accounted
> for in calculating peak LHGR.
'! The core is stable with respect to menon, and any menon transients which may be excited are rapidly damped.
't j The menon multiplier in Figure 3.2-3 was selected to conservatively =
account for transients which can result from control rod motion at full power.
3 The multiplier is defined as the ratio of the maximum value of Fg due to menon induced top peaked power redistribution and the Fg of the nominal i
' operating axial shape. This is consistent with the methodology used to derive 1, the LHGR limits; which were generated based on the worst top-peaked axial
'! power distribution. The minimum value of the multiplier is unity.
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l YANKEE-ROWE B3/4 2-1 Amendment No. 88 l
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3/4.'2' poser DISTRIBUTION LIMITS BASES (Continued) ~
l
- s. Control rods in a single group move together with no individual rod insertion differing by more than i 8 inches from any other rod in the group.
- b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.
- c. The control rod insertion limits of specification 3.1.3.5 is maintained.
The relaxstion in F as a function of THERMAL POWER allows e nges in the radial power shape h all permissible rod insertion limits. H will be maintained with its limits provided Conditions a through e above a
. maintained. '
When an gF measurement l's taken, experimental error, engineering g' tolerance and fuel densification must be allowed for. 5% is the appropriate
- allowance for a full core map taken with the incore detector flux mapping system, 4% is the appropriate allowance for engineering tolerance and 3% is
. the appropriate allowance for fuel densification.
When F is measured, experimental error must be allowed for and 5% is theappropbteallowanceforafullcoremaptakenwiththeincoredetection system.
3/4.2.4 DNB PARAMETERS 4
The limits on the DNB related parameters assure that each of the lg parameters are maintained within the normal steady state envelope of operation
- e
- assumed in the transient and accidsnt analyses. The limits are consistent with the accident analysis assumptions and have been analytically demonstrated
- adequately to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The Main Coolant System inlet temperature assumed in the analysis
'- is conservatively 40 F in excess of the limit to allow for uncertainty in plant measurement. The Main Coolant system pressure assumed in the analysis is 1925 psis, conservatively 25 psig less than the limit to allow for l!s uncertainty in plant measurement. The assumed operating deadband of i 50 psig
- is applied to the nominal 2000 psig limit, yielding a minisman operation limit of 1950 psig.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the Main Coolant System total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
I YAMXEE-ROWE B3/4 2-3 Ame8dsent No. O ,88
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t 5.0 DESIGN FEATURES
(
. 5.1 SITJ ,
EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1. .
! LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.
SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be shown in Figure 5.1-3.
SITE BOUNDARY FOR LIOUID EFFLUENTS
- 5.1.4 The site boundary for liquid effluents shall be shown in Figure 5.1-4.
5.2 CONTAINMENT
',' CONFIGURATION 5.2.1 The Reactor Containment Building is a steel spherical shall having the following design features:
'( a. Nominal inside diameter = 125 feet.
! b. Minimum thickness of steel shell = 7/8 inches. ,
- c. Net free volume = 860,000 cubic feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a
= ,
maximum internal pressure of 34.5 psis and a temperature of 2490F.
! 5.3 REACTOR CORE f-
- ' FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 76 fuel assemblies with each fuel assembly containing up to 231 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 91 inches. Each fuel assembly shall contain a maximum total weight of 234 kilograms uranium. Reload fuel is
! similar in physical design to current fuel and has a nominal enrichment of 3.7 I weight percent U-235. -
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YANKEE-ROWE 5-1 Amendment No.g 88 l
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