ML20211K750

Proposed Changes to Tech Specs to Allow Performance of 10CFR50.59 Reviews for Future Core Reloads

ML20211K750
Person / Time
Site:	Yankee Rowe
Issue date:	06/24/1986
From:	YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20211K736	List: ML20211K732 ML20211K750
References
NUDOCS 8606300244
Download: ML20211K750 (21)

Text

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4 ATTACHNENT A TECHNICAL SPECIFICATION CHANGES l '

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REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5 The control groups shall be limited in physical insertion as provided in Figure 1 of the Core Operating Limits Report. l APPLICABILITY: MODES 1* and 2*#,

ACTION:

With the control groups inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2. either;

a. Restore the control groups to within the limits within two hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.3.5 The position of each control group shall be determined to be within the insertion.11mits at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• See Special Test Exceptions 3.10.2 and 3.10.4.

1. With Kegg 1 1.0.

YANKEE-ROWE 3/4 1-28 Amendment No. 77

s Figure 3.1-2 This Figure Intentionally Blant f

t YANKEE-ROWE 3/4 1-29 Amendment No. 43,69,77

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J/4.2 POWER DTSTRTBIITION LIMITS PEAK LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 The peak linear heat generation rate (LHGR) shall not exceed the limits provided in Figure 2 of the Core Operating Limits Report during steady l clato operation.

_ APPLICABILITY: MODE 1.

ACTION:

With the peak LHGR exceeding the limits provided in Figure 2 of the Core Operating Limits Report:

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a. Within 15 minutes reduce THERMAL POWER to not more than that fraction of the THERMAL POWER allowable for the main coolant pump combination in operation, as expressed below:

Limiting LHGW Fraction of TIIERMAL POWER = Peak Full Power LHGR

b. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reduce the Power Range and Intermediate Power Range Neutron Flux high trip setpoint to 1 108% of the fraction of THERMAL POWER allowable for the main coolant pump combination.

SUPVETtLANCE FSOUIREMENTS 4.2.1.1 The peak LHCR shall be determined to be within the limits provided in Figure 2 of the Core Operating Limits Report using incore instrumentation to obtain a power distribution map:

4. Prior to initial operation above 75% of RATED THERMAL POWER after l

each fuel loadin6, and

b. At least once per 1,000 EFPH.

c. The provisions of Specification 4.0.4 are not applicable.

YANKEE-ROWE 3/4 2-1 Amendment No. 44,54,72,83

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POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.2 The below factors shall be included in the calculation of peak full power LHCR:

a. Heat flux power peaking factor, F(, measured using incore instrumentation at a power g 10%.

b. The multiplier for xenon redistribution is a function of core lifetime as provided in Figure 3 of the Core Operating Limits Report. In addition, if Control Rod Group C is inserted outside the operating band for 100% allowable power, 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 within the operating band for 100% allowabic power.

Reduced Power = Allowable fraction of full power times multiplier provided in Figure 4 of the Core Operating Limits Report.

Exceptions: 1. If the rods are inserted outside the operating band for 100% allowable power and power does not go below the reduced power calculated above, hold at the lowest attained power level for at least twenty-four hours with Control Rod Group C within the operating band for 100% allowable power before returning to allowable power.

2. If the rods are inserted outside the operating band for 100% allowable power and zero power is held for more than forty-eight hours, no reduced power level need be held on the way to the allowable fraction of full power.

c. Shortened stack height factor, 1.009.

d. Measurement uncertainty:*

1. 1.05, when at least 17 incore detection system neutron detector thimbles are OPERABLE, or

2. 1.068, when less than 17 incore detection system neutron detector thimbles are OPERABLE.

YANKEE-ROWE 3/4 2-2 Amendment No. 43,53,72,77,88 l

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Figure 3.2-1 This Figure Intentionally Blank YANKEE-ROWE 3/4 2-4 Amendment No. 82,88

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Figure 3.2-3 This Figure Intentionally Blank YANKEE-ROWE 3/4 2-6 Amendment No. 69,77,88

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Figure 3.2-4 This Figure Intentionally Blank

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YANKEE-ROWE 3/4 2-7 Amendment No. 69,77

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RfACTIVITY CONTROL SYSTEMS.

EADiS 3 / 4 .'1.1' 80 RATION CONTROL $

3/4.1.1.1 md 3/4.1.1. 2 SHUTDOWN MARCIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditio~ns, 2) the reactivity transients associated with postulated accident conditions are controllab.le within acceptable limits, and 3) the reactor will bd 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 Power Dependent Insertion Limits (PDIL) as provided in Figure 1 of the Core Operating Limits Report. For 4900F i T vg, a the requirement for a SHUTDOWN MARCIN is established by postulated steam line break conuiderationsevith ECCS and NRVs available and covers the requirements to preclude inadvertent criticality. For 330 s Tavg < 4900F, the requirement for a SHUTDOWN

, MARGIN is sufficient to preclude inadvertent criticality and covers the requirements of steam line breaks with automatic initiation of ECCS and NRVs blocked. With Tavg < 3300F, the reactivity transients resulting from a steam line break cooldown are minimal. 5% Ak/k SHUTDOWN MARGIN.(with all rods inserted) provides adequate protection to preclude criticality for all postulated accidents for the reactor vessel head in place.

To eliminate possible errors in the calculations of the initial 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, must be adjusted (normalized) to accurately reflect actual core conditions. Normally, when full power is l reached after each refuoling, and with the control rod groups in the desired 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 normalization should be completed after about 10%

of the total core burnup. Thereafter, actual boron concentration can be compared with prediction and the reactivity status of the core can be continuously evaluated, and any deviation would be thoroughly investigated and evaluated, i

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YANKEE-ROWE B3/4 1-1 Amendment No. 82,88 l

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3/4.2 POWER DISTRIBUTION LIMITS BASES 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 1 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

3/4.2.1 PEAK 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 criteria limit of 22000F is not exceeded.

When operating at constant power, all rods out, with equilibrium xenon, 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 after each fuel loading thus provides an upper bound on all-rods-out power peaking for the remainder of that cycle. Thereafter the measured power peaking shall be checked every 1,000 equivalent 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 LHCR.

The core is stable with respect to xenon, and any xenon transients which may be excited are rapidly damped.

The xenon multiplier provided in Figure 3 of the Core Operating Limits Report was selected to conservatively account for transients which can result from control rod motion at full power.

i The multiplier is defined as the ratio of the maximum value of Fz due l

to xenon induced top peaked power redistribution and the Fg of the nominal l

operating axial shape. This is consistent with the methodology used to

derive the LHGR limits, which were generated based on the worst top-peaked l axial power distribution. The minimum value of the multiplier is unity.

i YANKEE-ROWE B3/4 2-1 Amendment No. 88

T 3/4.2 POWER DISTRIBUTION LIMITS BASES (Continued)

The limits on power level and control rod position following control rod insertion were selected to prevent exceeding the maximum allowable linear heat generation rate limits provided in Figure 2 of the Core Operating Limits Report within the first few hours following return to power after the insertion. With Yankee's highly damped core, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hold allows sufficient time for the initial xenon maldistribution to accommodate itself to the new power distribution. The restriction on control rod location during these 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the return to allowable fraction of full power will not cause additional redistribution due to rod motion.

After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at zero power, the average xenon concentration has decayed to about 20% of the full power concentration. S!nce the xenon concentrations are so low, an increase in power directly to maximum allowable power creates transient peaking well below the value imposed by the xenon redistribution multiplier. Thus, any increase in power peaking due to this operation is below the value accounted for in the calculation of the LHGR.

These conclusions are based on plant tests and on calculations performed with the SINULATE three dimensional nodal code used in the analysis of Core XI (reference cycle) described in Proposcd Change No. 115, dated March 29, 1974.

The Factors d, e and f in Specification 4.2.1.2 will be combined statistically as the " root-sum-square" of the individual parameters. This method for combining parameter uncertainties is valid due to the independence of the parameters involved. Factor d accounts for uncertainty in the power distribution neasurement with the movable incore instrumentation system.

Factor e accounts for uncertainty in the calorimetric measurement for determining core power level. Factor f accounts for uncertainty in engineering and fabrication tolerances of the fuel. Together these factors, when combined statistically, yield an uncertainty of 8.5% for less than 17 operating incore thimbles and 7.1% for greater than 17 operating thimbles.

This factor and Factors a, b, e and g will be combined multiplicatively to obtain peak LHGR values.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux and enthalpy hot channel factors ensure that

1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 0 F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable but will normally only

, be determined periodically as specified in Specifications 4.2.2.1 and l 4.2.3.1. This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

YANKEE-ROWE. B3/4 2-2 Amendment No. 43,88 1

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).O DESIGN FEATURES 5.1 SITE.

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 CASEOUS 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 shell having the following design features:

a. Nominal inside diameter = 125 feet,

b. Minimum thickness of steel shall = 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 psig and a temperature of 2490F.

5.3 REACTOR CORE 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 shall have a maximum 1 nominal enrichment of 4.0 weight percent U-235.

i YANKEE-ROWE 5-1 Amendment No. 82,88

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ADMINISTRATIVE CONTROLS 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the Plant Operations Review Committee and approved by the Plant Superintendent prior to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 Procedures that have been developed as a result of changes defined in 10CFR 50.59(a)(2) shall be independently reviewed to verify that the implementing actions do not constitute an unreviewed safety question. Those reviews shall be performed by Nuclear Service Division personnel having qualifications at least equivalent to those specified for NSAR Committee membership in 6.5.2.3. The procedures shall be approved by the Manager of Operations, NSD.

6.8.4 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License.

c. The change is documented, reviewed by PORC and approved by the Plant Superintendent within 14 days of implementation.

6.9 REPORTING REOUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted. The reporting requirements of Specifications 6.9.1, 6.9.2, 6.9.3 and 6.9.4 are in accordance with Revision 4 of Regulatory Guide 1.16. " Reporting of Operating Information - Appendix A Technical Specifications."

ROUTINE REPORTS 6.9.1.1 Core Operating Limits Report. A report providing the following core operational limits shall be provided to the Project Manager of the USNRC NRR, with a copy to the Regional Administrator of the Regional Office of the NRC at least 60 days prior to each cycle initial criticality unless otherwise approved by the Commission by letter:

1. Power Dependent Insertion Limits

2. Allowable Peak Rod LHGR Versus Cycle Burnup

3. Multiplier for Xenon Redistribution Versus Cycle Burnup

4. Multiplier for Reduced Power Versus Cycle Burnup YANKEE-ROWE 6-14 Amendment No. 46,80

I ADMINISTRATIVE CONTROLS (Continued)

In addition, in the event that the limits should change requiring a new submittal or amended submittal of the Core Operating Limits Report, it shall be submitted at least 60 days prior to the date the limits would become effective unless otherwise approved by the Commission by letter. Any information needed to support the Core Operational Limits Report will be provided to the NRC upon their request.

6.9.1.2 Startup Report. A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving any planned increase in power level (3) installation of fuel that has a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FHSR and shall in general include a description of measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be. described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

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YANKEE-ROWE 6-14A Amendment No.

7-ADMINISTRATIVE CONTROLS

d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

e. An evaluation of the change, which shows the expected maximum exposures to member (s) of the public at the site boundary and to the general population that differ from those previously estimated in the license application and amendments thereto;

f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; cnd

h. Documentation of the fact that the change was reviewed and found acceptable by PORC.

2. Shall become effective upon review and acceptance by PORC.

6.17 ANALYTICAL METHODS 6.17.1 The analytical methods used to generate the data presented in the core Operating Limits Report described in Specification 6.9.1.1 were previously reviewed and approved by the NRC. If changes to these methods are deemed necessary, they will be submitted to the NRC for review and approval prior to their use if the change is determined to involve an unreviewed safety question, or if such a change would require the amendment of previously submitted documentation.

YANKEE-ROWE 6-27 Amendment No. 80

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I ATTACHMENT B CORE OPERATING LIMITS REPORT l

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Yankee Nuclear Pcwer Station Core Operating Limits Report Per the requirements of Technical Specification 6.9.1.1, this Core Operating Limits Report has been prepared to provide the necessary limitations on reactor power and control rod position for the operation of the Yankee Nuclear Power Station, Cycle 18.

1. Figure 1 provides the Power Dependent Insertion Limit (PDIL).

The PDIL is based on the requirements of the steam line break transient analysis and is also an important consideration in the analysis of the Loss of Coolant Flow and Boron Dilution transients, since it limits the minimum shutdown margin available.

2. Figure 2 provides the allowable peak rod LHGR. Limiting the peak LHGR during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limits of 10CFR50.46 are not exceeded.

3. Figure 3 provides the xenon redistribution multiplier. The xenon multiplier was selected to conservatively account for xenon redistribution transients which can result from control rod motion at full power.

4. Figure 4 provides the reduced power multiplier. This multiplier was selected to prevent exceeding the allowable LHGR limits within the first few hours following return to power after control rod insertion outside the operating band for 100% allowable power as provided in Figure 1.

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