ML20141J149

Summary of 860415 Meeting W/Utils in Bethesda,Md Re Util Proposed Tech Spec Mod to Allow for Changing cycle-dependent Parameter Curves W/O Tech Spec Amend.List of Attendees & Viewgraphs Encl

ML20141J149
Person / Time
Site:	Yankee Rowe
Issue date:	04/21/1986
From:	Clifford J Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8604250351
Download: ML20141J149 (35)

Text

.

APR 211986 Docket No.50-029 LICENSEE:

Yankee Atomic Electric Company (YAEC)

FACILITY:

Yankee Nuclear Power Station (Yankee)

SUBJECT:

MEETING

SUMMARY

ON YAEC PROPOSAL FOR MODIFICATION OF TS FOR CORE REL0AD ANALYSES A meeting was held on April 15, 1986 between members of the staff and personnel from YAEC to discuss a YAEC proposal for modification.pf the. Yankee Technical Specifications (TS) to allow for changing cycle dependent parameter

~

curves without a TS Amendment. The meeting was held at the hRC headquarters, Bethesda, Maryland.

A list of attendees is provided in Enclosure 1.

The licensee provided a discussion of their reasons for the proposal, and detailed how they would like to modify the TS. The viewgraphs used in this presentation are provided in Enclosure 2.

Varkups of draft proposed change pages are provided in Enclosure 3.

The licensee's proposal is to remove four cycle dependent parameter curves from the TS, and to modify the TS to reference the curves. The curves themselves would be contained in a separate report. The proposed TS would include a requiren;ent to provide the separate report, a " Core Operational Limits Report," at least 60 days prior to a refueling outage restart. The basic objective of the licensee's proposal is to establish conditions in the TS that would allow the licensee to use approved analysis methods to determine r. ore operating limits, and to implement those limits without amending the TS.

Staff concerns dealt with two basic issues. The first issue is enforceability of the core operating limits report. The general staff consensus, oending confirmation with the legal and inspection staffs, was that the Yankee proposal would be enforceable based on inspection.

The second staff concern dealt with the legal administration of the Yankee proposal. The staff stated that the Yankee request dealt with policy issues that would have to be resolved through internal staff discussions. These issues deal with the applicability of the requirements of 10 CFR 50.59 and 10 CFR 50.90 to the Yankee proposal.

Specifically, the staff needs to resolve whether or not a report that is referenced in the TS becomes a part of the TS that is subject to the amendment requirements of 10 CFR 50.59(c) and 10 CFR 50.90.

8604250351 860421 PDR ADOCK 05000029 P

PDR

APR 211986 The licensee will be informed of the results of these staff deliberations, including any alternatives that may be available to achieve the same objective, b

James W. Clifford, Project Manager Project Directorate #1 Division of PWR Licensing-A Enciosures:

1. List of Attendees

2. Handouts

3. Draft TS Pages cc w/ enclosures:

See next page Office:

PM/P 9#

PD/ PAD #1 -

Surname: JCli, ord/tg Glear V'

Date:

04/I'7/86 04/p/86 E

i f

a Mr. George Papanic, Jr.

Yankee Atomic Electric Company Yankee Nuclear Power Station I

cc:

Mr. James E. Tribble, President Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street i

Boston, Massachusetts 02110 4

Mr. N. N. St. Laurent Plant Superintendent Yankee Atomic Electric Company Star Route Rowe, Massachusetts 01367 Chairman Board of Selectmen i

Town of Rowe Rowe, Massachusetts 01367 i

Resident Inspector Yankee Nuclear Power Station c/o U.S. NRC i

Post Office Box 28 Monroe Bridge, Massachusetts 01350 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health i

150 Tremont Street, 7th Floor j

Boston, Massachusetts 02111 j

l l

APR 211965

. DISTRIBUTION:

' Docket-Files'-

NRC'PDR~

Local POR G. Lear J. Clifford i

0 ELD E. Jordan B. Grimes J. Partlow (Emergency Preparedness only)

ACRS (10)

NRC Participants PDd1 r/f PD#1 s/f f

i i

ENCLOSURE 1 t

i LIST OF ATTENDEES YAEC PROPOSALS FOR TS MODIFICATIONS TO ACCOMODATE OF CORE RELOADS NAME AFFILIATION Jim Clifford NRR/ Yankee PM R. Lobel NRR/RSB-A George Papanic, Jr.

Yankee Robert Harvey Yankee-LOCA Richard Cacciapouti Yankee-Reactor Physics Ausaf Husain Yankee-LOCA Kevin Morrissey Yankee-Reactor Physics 4

Daniel Fieno NRR/RSB-B Cecil Thomas NRR/RSB-B Norm Lauben NRR/RSB-B 1

Marvin Dunenfeld NRR/RSB-A Howard Richings NRR/RSB-BWR Brent Clayton NRR/F0B-PWR-A Robert Perch NRR/F0B-PWR-A Larry Phillips NRR/RSB-BWR Larry Saiyards BG&E/AIF STSI a

d

ENCLOSURE 2 l

WHY e RELOA'DS PERFORMED UNDER 10CFR50.59

• BENEFIT TO YANKEE O REDUCE ADMINISTRATIVE WORK FOR ROUTINE TECH SPEC CHANGES e BENEFIT TO NRC O ELIMINATE REVIEW OF ROUTINE TECHNICAL SPECIFICATION CHANGES t

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HOW

• CORE OPERATING LIMITS REPORT l

i e REPORTS SUBMITTED TO NRC O COLR 60 DAYS PRIOR TO USE O CPAR FOR INFORMATION j

e CONTROLLED BY PLANT i

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• Allowable THER.MA1. Power based on the main coolant purnp c.ombination,in operation.

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NUCLEAR SERVICES DMSION I

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1r DATA REFERENCE MANUAL i

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CONTECS OF JATA REFERE\\CE MAN AL e CORE PARAMETERS e STARTUP PHYSICS DATA e BORON LETDOWN CURVE e SHUTDOWN MARGIN CURVES e CONTROL ROD WORTH CURVES

• MISCELLANEOUS PHYSICS CURVES e LTOP CURVES e CHEMISTRY DATA AND PROCEDURES e SHUTDOWN MARGIN PROCEDURE e CORE OPERAT!NG LIMITS PROCEDURE 4

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~ SIMILAR TO REACTIVITY ANOMALY

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ENFORCEVENT

• LOCA METHODS-APPENDIX K e PHYSICS METHODS APPROVED e PHYSICS VERIFIED BY TESTING e CPAR WILL BE SUBMITTED e COLR WILL BE SUBMITTED e RESIDENT INSPECTOR e i+E AUDITS 1

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l ENCLOSURE 3 l

f 50.59 TS CHANGES PAGE #'S 3/4 1-28 figure reference 3/4 1-29 delete figure 3/4 2-1 figure reference 3/4 2-2 figure references 3/4 2-4 i

3/4 2-5 delete figures 3/4 2-6 3/4 2-7 B 3/4 1-1 figure reference B 3/4 2-1 figure reference B 3/4 2-2 figure reference 5-1 assembly description 6/4, 6-15 add COLR description and retype 6-15 or 6-14A i

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REACTIVITY CONTROL SYST m s

. r" CONTROL ROD INSfRTION LIMITS LIMITING C_0NDITION F_0R OP_ERATION T

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

Reduce THERMAL PC R vithin two.bours to Ines than or equal to that fraction of RATED THIR".AL POWER which is al. loved by the group position using the above figure, or c.

De in at least EOT STANDET within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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.t 0 10 30 50 70 90 CONTROL. ROD CROUP C POSITION (INCHES WITHDRAVN):

• Allowable THERMA 1. Power based on the e.ain coolant pump cornbination in operation.

FICUR. E 3.1-2 YANKEE-ROVE 3/4 1-29 Amendment N6. ,J.g"./.7* f. * - - ....c

3 /4.2 POWER DI3TRIBUTIOS LIMITS 1 FE AK LINEAR HEAT CENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 The peak linear heat generation rate '(LHCR) shall not exceed the 11mits 4f-M ge : 3.2-1 during steady state operation, g 3phd4 m 1._C m C p u m Lt.ad, /% c st APPLIC ABILITY: NODE 1. ACTION: . wd.l n~ (L (w G. Sit T e. lw t *, Lyy With the peak LHCR exceeding the limits -Figure 3.2-1. a. Within 15 tainutes reduce THERMN, POVER to not more than that fraction of the THERMAL POWER allowable for the main coolant pu p combinatien in operation, as expressed below: Liciting LHCR Fraction of THERMAL POWER a, Peak Full Power LHGR b. Within 4 hours reduce the Power Range and Intermediate Power Range Neutron Flux high trip setpoint to f 108% of the fraction of THERMAL POWER allowable for the main coolant pump cochination. SURVEILLA';CE REQUIREMZKIS O p a a-4.2.1.1 The peak LHCR shall be det' ermined to be within the limits ;f Tigure 4v2-1 using incore instrumentation to obtain a power distribution map: Ou. Of. % kM IYt a Prior t.o initial operation above 75% of RATED THERMAL POWER after each fuel loading, and b. At least once per 1,000 ETPH, c. The provisions of Specification 4.0.4 are not applicable. O AmendmentNo.8.f,[.83 YAITKEE-ROVE 3/42-1

r 8 PC%TR DISTRIBUTIOW LIMIIS SURVtf t.t.ANCE REQUIRENENTS (Continued) 4.2.1.2 The below factors shall be included in the calculation of peak full power LHGR: Heat flux power peaking f actor F, sensured using incore instnamentation e. I at a power 1 101.. ?' Z ~" "(' b. The multiplier for xenon redistribution is a function of core lifetime as " ""," fi, elv-a la Figu re4A-3. In addition, if Control Rod Croup C is inserted 5 below)01.:h::i allowable power may not be r* gained until power has been ytTreduced level defined below for at least twenty >four hours with 5 Control Rod Group C between 00 .d "O ir.c.t.;;. R f '7 p -ReduM power - Allowable fraction of full power times., j 4 Gy Op&c 3 maltiplier str= i-Ti un 2.0 4 r fW"' 6A fro,# Exceptions: 1. If the rods are insert E b WeS? !m hes and power does N nat go below the reduced power calculated above, hold at N the lowest attained power level for at least twenty-four Y hours with Control Rod Group C between-49 er.d Winchee-p before returning to allowable power. c .k 2. If the rods are inserted below "I i :t r and seco power is held for more than forty-eight bours, no tWuced power level need be held on the way to the allowable fraction of full power. T c. Shortened stack height factor,1.009. M d. Measurement uncertainty:* t' fd 1. 1.05, when at least 17 incore d'etection system neutron detector thimbles are OPERABI.E. or 2. 1.068, when less than 17 incore detection system neutron detector s s thimbles are OPERABI.E. l Amendment No. M g f !' [ 8P YANKEE-ROWE 3/4 2-2 O 4

O O ALL0HABLE PEAK R00 LHGR VERSUS CYCLE BURNUP E o PRESH FUEL o - EXPOSED FUEL d n = ? 8 S i 4 11 - x r U M s *x s3 $-~ m 4 n' m LD5 "h 1-s 8 N1 ~. e. m s u a ~ fN L u i i -0 9-g~ - dr d E u r S N d s-y t ? 7< 7-g 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 18 17 18 19 20 AVMcGE BURNU? '090/"~.l? = * - -

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s BASES i 3/4.1.1 BORATION CONTR0t s 'w 3/4.1.1.1 and 3/4.1.1.2 SMUTDOWN MARCIM 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 . m L i. 'J acceptable limits, and 3) the reactor will 1,e maintained sufficiently (b .,L suberitical to preclude inadvertent criticality in the shutdown condition. b.ut, Ks t SHUTDOWW MARGIE requirements are a function of the plant operating For critical conditions, minimum shutdown margins are limiteId by the status. Power Dependent Insertion Limits (PDIL) as gir: in "is = 3.1-26 For 4900F the requiremant for a SHUTDOWN MARGIN is established by postulated 1 T yg,ine break considerations with ECCS and NEVs available and covers the a steam l requirements to preclude inadvertent criticality. For 330 $ Tavs < 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 ERVs blocked. With Tavs < 3300F, the reactivity transients resulting from a steam line break cooldown are minimal. 5%Ak/k SHUTDOWW MARCIN (with all rods inserted) provides adequate protection to preclude celticality for all postulated accidents with the reactor vessel head i I in place. To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depiction rate, the predicted relation between fuel burnup and the boron concentration, not easary to maintain adequate control charectoristics, must be adjusted (actualised) to accurately refleet actual cors conditions.. Normally, when fuel power is O reached af ter each refueling, and with the control rod groups in the desired positions, the boron concentretion is mestoured and the predicted steady-state curve is adjusted to this point. As power *operst3cn proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burttup and reactivity is compared with that predicted. This process of normalisation 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 s.ny deviation would be thoroughly investigated and evaluated. t O*,

3/4.7 PO**ER DISTRIBUTION LTMITS ~ B ASEs ' i The spect(icat?,ons of this section previde assurance of fuel integrity during Conditions I. (Ucrual Operation) and II (Incidents of Moderate s < Frequency) events b : (a) saintaining the minim 2n DNBR in the core 1 1.30 / during nomal operation and in short tem transteits, and (b) limiting the fisalon gas release, fuel pellet temperature and i ladding mechanical properties to within assumed design criteria. 3/4.2.1 PEAK LINEAE HEAT CENERATIO5 RATE Limiting the peak Linear Heat Generation Rat i Condition I events provides assvrance that the initial conditihn(!JICR) dur ng s assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 22000F is eat exceeded. When operating at constant power, all rods.ou,, with equilibrium menon, power peaking in ths Tankee Bowe core decreases monotonically as a function of cycle bumup. This has been verified by both calculation and measurement on Yankee cores and is in accord with the expected behivior in a core that does not contain buenable poison. The all-rods-out powsh peaking measured prior to upper bound on all-rods-out power peaking for the re;. loading thus provides exceeding 75% of RATED THERMAL POWER after each fuel mainder of that cycle. Thorwaf ter the measured power peaking shall be checEed every 1000 equivalent full power hours and the latest measured value shall) be used in the computation. The only effects which can increase piaking beyond this value would be control rod insertion and zenon transients and these are accounted for in calculating peak IJsCE. The core in stable with respect to menon, and say menon transients which any be excited are rapidly 3 g The zenon multipliers yi;;;r: 2.2 2 was selected to conservatively account for transients which can result from contro rod action at full power. The multiplier is defined as the retic of the naximum value of Fs due to menon induced top peaked power redistribution and the Fs of the nominal j operating axial shape. This is consistant with the[ methodology used to' derive l the IJaca llaits, wtdch asare generated based on the tsorst top-peaked axial power distribution. The minimum value of the multihlier is unity. 9 YAEXEE-BoWE R3/4.2-1 Amendment No. 88 i O i1 ' 4

pa.e r v - r. n vs iaL6viloti L1 n Lu [ e BASES (Continuso) The limits on power level and control rod position following control red insertion were selected to prevent ' exceeding the z.aximum allowable linear heat O i generation rato limits la-Figure 3r2-1 within the first few hours following return to power after the insertion. With Yankee's highly damped core, the 24 hour hold allows sufficient time for the initial xenon maldistribution to acconunodete itself to the new power distribution. The restriction on control rod location during these 24 hours assures that the return to allowable-fraction of full power will not cause additional redistribution due to rod motion. A.f ter 48 hours at cero power, the average Xenon concentration has decayed to about 20% of the full power concentration. Since the manon concentrations are so low, an increase in power directly to maximamn allowable power creates transient peaking well below the value imposed by the menon redistribution multiplier. Thus, any increase in power peaking due to this operation is below the value accounted for in the calculation of the UiCR. These conclusions are based on plant tests. and on calculations performed with the SIEULATE three dimensional nodal code used in the analysis of Core II (reference cycle) described in Proposed 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-sun-square" of the individual parameters. Tnis method for combining parmaster uncertainties is valid due to the ind2pe'ndence of the paranstors involved. Factor d accounts for uncertainty in the power distribution measurement with the revable incore instrumentation system. Factor e accounts for uncertainty in the calorimetric asasurement for determining core power level. Factor f accounts for uncertainty in O engineering and fabrication tolerances of the fuel. Together these factors, when combined statistically, yield an une'ertainty 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 smaltiplicctively to obtain peak UICR va' lues. 3/4.2.2 and 3/4.2.3 MEAT FLtTI MOT CHANNEL FACTOR AND WUCLEAR ENTHALFY RISE MOT. CHANVEL FACTOR The limita on bact flux and enthalpy hot channel factors ensure that }

1) the design limits on peak local power density and minimum Dust are not exceed 2d and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 22000F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable but will normally only be determined periodically as specified in specification 4.2.2.1 and 4.2.5.1. This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided: YANKEE-ROWE B3/4 2-2 Amendment No. M i 88

3.0 D r.S it.tv F tMT U h t;S 5.1 8ITE EXCt,USTOW AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1. LOW POPUI.ATION ZONE 5.1.2 The low population sone shall be as shown in Figure 5.1-2. SITE BOUNDARY FOR CASEOUS EFFLUESTS 5.1.3 The site boundary for gaseous effluents shall be shown in Figure 5.1-3. BITE BOUNDARY FOR I,IOUID EW LUTETS 5.1.4 The site boundary for liquid effluents st.all be shown in Figure 5.1-4. 5.2 cov7AINMT1rT COWTIGURATIQW 5.2.1 The Epactor Containment Building is a steel spherical shk11 having the following design features: a. Moninal inside diameter = 125 feet. b. Minimum thickness of steel shell = 7/g inches. c. 5et fece volume = 860.000 cubic f9et. EEI.HL.ZEEEEEli AELIEEEEEAHER S.2.1 The resetor containment is designed and shall be maintained for a maximaa internal pressvre of 34.5 pais and a temperature of 249'F. id_REACTQF_CQEZ N 13ENBLIES 5.2.1 The reactor core shall contain 76 fuel assemblies with each fuel assembly containir.g 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 734 kilograms uranium. Reload fuel is sieller in physical design to current fuel and has-s-nominal enrichment c,f 44 weight parcent V-235. A d/ L A g g. M Ame.ndment Nof!I 88 YANKEE-ROWE 5-1 O,

-~ ADMINISTT g VE CONTROLS 6.8.2 Each procedure and r.dninistrative 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 implecentation and reviewed periodically as set forth in administrative procedures. 6.B.3 Procedures that have been developed as a result of changes defined in 10 CTR 50.59(a)(2) shall be independently reviewed to verify that the implementing actions do not constitute an unreviewed safet; question. Those reviews shall be performed by Nuclear Service Division personncI having qualifications at least equivalent to those specified for NSAR Co=::ittee membership in 6.5.2.3. The procedures shall be approved by the Manager of Operations, NSD. 6.8.4 Te=porary 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 1.icense. c. The change is documented, reviewed by PORC and approved by the Plant Superintendent within 14 days of implementation. f.9 REPORTING REQUIREMENTS ( 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, "Reportinf of Operating Information - Appendix A Technical Specifications". yg ROUTINE RIPORTS .9.1 4.Startup Report. 'A s'vas;ary repert of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving and planned increase in power level, (3) installation of fuel that has a different fuel supplier, and (f.) 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 THSR 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 ec,rrective 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. TANKEE-ROVE 6-14 Amendment No. S0 .n,

()6.9.1.1 Core Operational Limits Report. A report providing the following core operational limits shall be provided to the Regional Administrator of the Regional Office of the NRC with a copy to the Director, NRR, Attention: Chief, Core Performance Branch, USNRC at least 60 days prior to each cycle initial criticality unless otherwise approved by the Commission by letter: 1. Power Dependent Insertion Limit 2. Allowable Peak Rod LHGR versus Cycle Burnup 3. Multiplier for Xenon Redistribution versus Cycle Burnup 4. Multiplier for Reduced Power versus Cycle Burnup In addition, in,the event that the limits should change requiring a l new submittal or amended submittal of the Core Operational Limits Report, it shall be submitted at least 60 day. prior to the date the limit would become effective unless otherwis2 approved by the Commission by letter. Any information needed to support the Core Operations 1 j Limits Report will be provided to the NRC upon their request. j l I

~ Yankee Nuclear Power Station Cycle 18 Core Operating Limits Repcrt 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 Yankee Nuclear Power Station, Cycle 18. 1. Figure 1 provides the Power Dependent Insertion Limit. The operating band for 100% allowable power is consistent with the methodology used to derive the Linear Heat Generation Rate (LHGR) and shutdown margin limits. 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 transients which can result from control rod motion' at full power. 4. Figure 4 provides the reduced power multidier. These limits were selected to prevent exceeding the allowable LHGR limits within the first few hours -following return to power af ter control rod insertion. f ..,.,..., _ n-.-y-.

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