ML20039A495
| ML20039A495 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee, Yankee Rowe, Maine Yankee |
| Issue date: | 11/25/1981 |
| From: | Requa G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8112170706 | |
| Download: ML20039A495 (18) | |
Text
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,2 NOV 2 51981 h'h Docket Nos. 50-309
/[e.y 50-271
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LICEllSEES: liaine Yankee Atonic Power Conpany
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kh Yankee Atonic Electric Conpany b.
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Vernont Yankee fluclear Power Corporation 6
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FACILITIES: !!aine Yankee Atomic Power Station
'3 Vemont Yankee Nuclear Power Station Yankee Nuclear Power Station (Yankee Rowe)
SUBJECT:
Suff1ARY OF ttEETIllG HELD Oil OCTOBER 28, 1981 TO DISCUSS A NEW APPROACil TO TECilNICAL SPECIFICATI0!! (TS) REVISIONS TO ALLOW CORE PILOADS UNDER 10 CFR PART 50.59 PROVISIONS llITHOUT PRIOR flRC APPROVAL.
On October 28, 1981, the staff met with fiaine Yankee Vemont Yankee and Yankee Rowe to discuss a new approach to TS revisions that would allow the licensees to perfom core reloads under the 10 CFR Part 50.59 pro-visions without prior NRC approval. fleeting attendees are identified in.
Copies of viewgraphs shown during the course of the necting are contained in Enclosure 2.
The neeting agenda is shown in Enclosure 3.
11eeting highlights are sumarized below.
I.
The Problem Licensees are pemitted to nake changes to the facility under the provisions of 10 CFR 50.59 without prior URC approval pmvided there is no unreviewed safety issue involved and no TS change required. Currently, however, nuncrical values, which Yary by small amounts between fuel cycles, appear in the TS. Because of fuel economics, licensees recalculate thesennumbers for each reload cycle and propose new TS limits which requim NRC staff review. The staff reviews these. changes to see that they still naintain required operating nargins.
mgm II. Licensee's Suqqcsted Approaches x-J A.
Use TSs with Boundinn Numerical Values ao This dictates a margin which is large enough to accomodate various cycles.
o N;j This approach, however, is contrary to fuel economics.
It also requires 8m significant nodifications in nethods to provide operating nargin. Also, there is no guarantee that the bounding nutbers will remain bounding for o-8C future cycles, consequently this defeats a one-tine TS change.
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B.
Re-fomat TSs to remove numerical values for LCO's and LSS's to muu uJ Lud uli.hvui. nuud "vr prior slig.i Qa uwe nt rariau vos i. Lu 9
omcc >....f.9.tIml..nw1.m..aad..aapro. val....... Pac..I xo.ul.d..s.uffidtently..snecify...the..rae.thods..
and conditions under which LCOscad :LSSs'aretto jbe calc 01 ate Ctotassucem
'..."..myt"NUUyMM'YiSFyfli"BiiTMd"in".Etill"f6s0TtT##". duMdFf 55T"VuT.l."
5 OFFICIAL RECORD COPY usc m e - u wea unc ronu sia tio-so> nacu o24o
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,, III. Licensee Recomendation The licensees recommended the second approach above, i.e., reformat the TS to allow cycle dependent variations to be accomodated. This would allow the licensees to recalculate setpoints while naintaining requirdd margins and thus obtain naximum fuel econonics. The TS revisions would identify flRC approved calculational methods, identify the cycle dependent values as coefficients in the methods and identify where the coefficients will be found; such as the Core Performance Analysis Report (CPAR).
The licensees would perform and document safety reviews and issue a CpAR.
An appendix of the CPAR would contain the cycle specific valuas and would be fomally transmitted to the plant (reactor engineer). An annual report required by 10 CFR 50.59 would be issued fomally to the flRC.
Examples of TSs using coefficients rather than specific values were shown in the licensee's presentation (See Enclosure 2, Change to TS 2.1).
IV. f1RC Connents flRC Inspection and Enforcement (I&E) noted that their inspection procedures will require modification because the burden of verification would shift to ILE.
V.
Agreements The flRC staff generally agreed that an approach proposed by the licensees could be developed.
The licensees proposed to submit TS revisions for f1RC staff review in December 1981.
O&nni si~,ned by:
G. Requa, Project !!anager Operating Reactors Branch #3 Division of Licensing
Enclosures:
1.
List of Attendees 2.
Viewgraphs Shown 3.
liecting Agenda cc w/ enclosures:
See next page C
D orb's DL ORB #3:Dipj~ )spRB,#3:Ckr.h?b&
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Nac ronu ats po an Nacu ena OFF 'CIAL RECORD COPY es 2 0 o m -33 m 3
MEETING
SUMMARY
DISTRIBUTION Licensee:
Maine Yankee-Atomic Power Company
- Copies also sent to those people on service (cc) list for subjegt plant (s). --
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Docket File NRC PDR L POR NSIC TERA ORB #3 Rdg
'J01shinski JHeltemes, AE0D.
BGrimes RClark Project Manager Licensing Assistant ACRS (10)
~ Mtg Summary Dist.
NRC Participants O
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Maine Yankee Atomic Power Company 4-cc:
E. W. Thurlow, President Mrs. L. Patricia Doyle, President Maine Yankee Atomic Power Company SAFE POWER FOR MAINE Edison Drive Post Office Box 774 Augusta, Maine 04336 Camden, Maine 04843 Mr. Donald E. Vandenburgh First Selectman of Wiscasset Vice President - Engineering Municipal Building Yankee Atomic Electric Company U. S. Route 1
- 20 Turnpike Road Wiscasset, Maine. 04578 Westboro, Massachusetts 01581 Mr. Gustave A. Linenberger John. A. Ritsher, Esquire Atomic Safety and Licensing Board Ropes & Gray U.S. Nuclear Regulatory Commission 225 Franklin Street Washington, D. C.
20555 Boston, Massachusetts 02110 Davi.d Santee Miller, Esq.
Mr..Rufus E. Brown 213 Morgan Street, N. W.
Deputy Attorney General Washington, D. C.
2Q001 State of Maine Augusta, Maine 04330 Mr. Paul Swetland Resident Inspector / Maine Yankee Mr. Nicholas Barth c/o U.S:N.R.C.
Executive Director P. O. Box E Sheepscot Valley Conservation Wiscasset, Maine 04578 Association, Inc.
.,J P. O. Box 125 Mr. Cha'rles B. Brinkman Alan, Maine 04535 Manager - Washington Nuclear Operations Combustion Engineering Inc.
':iscasset Public Library Association 4853 Cordell Avenue, Suite A-1 High Street Wiscasset, Maine 04578
- Bethesda, Maryland 20014 Mr. John H. Garrity, Director Mr. Torbet H. Macdonald, Jr.
Nuclear Engineering & Licensing Of fice of Energy Resources Maine Yankee Atomic Pcwer Company State House Station #53 Edison Drive Augusta, Maine 04333 Augusta, Maine 04336 Robert M. Lazo, Esq., Chairman Atonic Safety and Licensing Board U.S. Environmental Protection Agency U.S. Nuclear Regulatory Commission-Region I Office Washington, D. C.
20555 ATTN:
Regional Radiation Representative JFK Federal Building Dr. Cadet H. Hand, Jr., Director Boston,. Massachusetts 02203 Bodega Marine Laboratory University of Cal,1fornia Bodega Bay, California 94923 State Planning Officer Executive Department l
189 State Street Augusta, Maine 04330 O
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'r ATTENDANCE LIST FOR NRC STAFF MEETING WITH HAINE YANKEE, VERMONT YANKEE AND YANKEE R0WE CONCERNING' RELOADS USING THE 10 CFR 50.59 OCTOBER 28,198T '
NRC VERMONT YANKEE R. Purple S. J. Jefferson L. Phillips E. Goodwin T. Novak H. J. Richings MAINE YANKEE
'A.
ii. Gill V. Rooney M. A. Whitney G. Lainas R. A. Clark G. Requa C. Liang YANKEE R0WE
- S. Wu J. Voglewede R. H. Groce*
D. Fino G. R. _ Klinger, IE:HQ L. H. Bettenhausen, IE:Rg. I YANKEE ATOMIC ELECTRIC NUCLEAR SERVICES DIVISION P. Bergeron B. Slifer D. Edwards R. J. Cacciapouti A. Husain
- Represents all (3) Yankees l
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' CURRENT SITUATION o
LICENSEES ARE PERMITTED TO MAKE CHANGES TO FACILITY UNDER 10CFR50.59 IF o
NO UNREVIEWED SAFETY QUESTIONS o
N0 TECH. SPEC CHANGES o
T0.0PERATE AT FULL POWER AND STILL MAINTAIN OPERATING MARGIN o
YAEC HAS NEEDED TO MAKE USE OF CYCLE DEPENDENT PARAMETERS o
MAKE MINOR MODIFICATIONS TO TECH. SPECS. TO REFLECT CHANGES FOR CYCLE.
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BASICALLY TWO APPROACHES 1)
USE B0UNDING TECH. SPEC. LIMITS.
A)
REQUIRES SIGNIFICANT MODIFICATION IN METH0DS' TO PROVIDE OPERATING MARGIN
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B)
N0 GUARANTEE THAT TECH. SPECS. WILL BE B0UNDING
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FOR FUTURE CYCLES 2)
RE-FORMAT TECH. SPECS. TO ALLOW CYCLE DEPENDENT-VARIATIONS TO BE ACCOMMODATED WITHOUT NEED FOR FORMAL REVIEW AND APPROVAL.
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PROPOSED' METHOD TO MEET OBJECTIVE o
SUBMITOsETIMEPROPOSEDCHANGEMODIFYINGTECH. SPECS.
O PREPARE CYCLE SPECIFIC CORE PERFORMANCE ANALYSIS REPORT (CPAR) - NO CHANGE o
APPENDIX 0F CPAR WILL INCLUDE CYCLE SPECIFIC VALUES THAT HAVE BEEN IDENTIFIED IN TECH. SPECS. - NO CHANGE o
CPAR WILL BE FORMALLY TRANSMITTED TO PLANT CONTAINING CYCLE SPECIFIC NUMBERS - NO CHANGE b
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BENEFITS I
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REDUCED SCOPE OF REVIEW 0
N0 NEED FOR SER o
ALLOW STAFF RESOURCES TO CONCENTRATE ON MORE IMPORTANT SAFETY ISSUES 4
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o SCHEDULAR %EXIBILITY OF REFUELING SHUTDOWNS 9
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TECHNICAL SPECIFICATIONS WE'D MODIFY TO ENABLE 50.59 RELOADS:
MAINE YANKEX:
1.3 REACTOR - GENERALIZE LIMIT ON ENRICHMENT NOW 3.03 WT%
2.1 LSSS - R.P.S COEFFICIENTS A, B, C ORy AUD Ay FUNCTIONS SYMMETRIC OFFSET TRIP FIGURES 2.1-1A, 2.1-1B, FIGURE 2.1-2 3.10 A. SHUTDOWN MARGIN B. POWER DISTRIBUTION LIMITS LHGR FOR LOCA FIGURES 3.10-1, 2, 3, 4, 5 i
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YANKEE ROWE:
1 2.1.1 SAFETY LIMIT CURVES FIGURES 2.1-1 AND 2.1-2 3/4.1 3.1.1.1 SHUTDOWN MARGIN 3.1.1.4 MODERATOR TEMPERATURE COEFFICIENT
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3.1.3.1 CONTROL ROD OPERABILITY 3.1.3.5 ROD INSERTION LIMIT 3/4.2 3.2.1 POWER DISTRIBUTION lit 1ITS FIGURES 3.2-1, 3.2-2, 3.2-3, AND 3.2-4 i
- 5. 0 DESIGN FEATURES CORE DESCRIPTION 4
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3.ll.A MAPLHGR 3.11.C.1 MCPR OPERATING LIMIT.
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2.1 LIMITING SAFETY SYSTEM SETTING - RELC'CR P.607EC7'05 SiSTEM Applicability: Applies to reactor trip settings and bypasses for the
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instru=ent channels monitoring the process variables which 8
influence the safe operation of the plant.
C'jective:
To provide auto =atic protective action in the event.that the proecss variables approach.a safety.lf=ft.
F F eef fication:
The reactor protective system trip setting li=fts and
.t char.:.els s'all bypasses for tie required operable instru=ent be as follevs:
2.1.1 Core Frotection a)
Variable Nuclear Overpever f_ Q4-10, or 106.5 (whichever is s= aller) for 10fQf100; 120 for Q-110.
where ther=al or nuclear power, whichever is larger.
Q = Fercent b)
Ther=al Margin / Low Pressure 2A Qpx3 + BTc + C, or 1835 psig (whichever is larger) where T
= cold leg te=perature, CF e
A
= 2050.7 3
= 17.9 C
= -10053 Opg3
=Al x QR1 Ay and QR1 are given in Titure 2.1-la and 2.1-lb, respectively.
This trip may be bypassed below 10 percent of rated pcver, c)
The sy=cetric offset trip and pretrip function shall not exceed the li=its shown in Figure 2.1-2, for. three loop operation. This trip may be bypassed below 15 percent of rated power.
d)
Low Reactor Coolant Flev 2,93 percent of 360,000 GPM (3 pump operation)
This, trip may be bypassed below 2 percent of rated power.
2.1-1 An.endment No. 29, }$, $0, $$, Sg JUL 101961 l
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CHANGES TO T.S. 2.1 A
B C
CYCLE 4 2054.7 17.9
-10053 CYCLE 5 2004.3 17.9
-10053 CYCLE 6 2060.7 17.9
'10053
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y 2.1 LIMITING SAFETY SYSTEM SETTING - REACTOR PROTECTION SYSTEM Applic ab ilitg: Applies to reactor trip settings. and bypasses for the l
instrument channels monitoring the process variables which influence the safe operation of the ' plant.
Objective:
To provide automatic protective action in the event that the process variables approach a safety limit, Specification: The reactor protective system trip setting limits and bypasses for the required operable instrument channels shall be as j
follows :
2.1.1 Core Protection a)
Variable Nuclear Overpower dEQ+10, or 106.5 (whichever is sma'11er) for 10di Q f!100; ti20 for Q 610 where
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Q = Percent thermal or nuclear power, whicheverr is larger.
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b)
Thermal Margin / Low Pressure 3 A QDNB + BTe, or 1835 psig (whichever is larger) l where l
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= cold leg temperature, OF e
QDNS = Al x QR1 C, A, and QR1 are functions of the cycle specific A,
B, 1
core loading pa ttern This trip may be bypassed below 10 percent of rated power.
c)
The s ymmetric o ffset trip and pretrip are functions of the cycle specific core loading pattern and shall be set to assure that fuel centerline melt and DNB are net exceeded for anticipated operational occurrances.
d)
Low Reactor Coolant Flow 1
3593 percent of 360,000 GPM (3 pump operation)
This trip may be bypassed below 2 percent of rated power.
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2.1.2 Other Reactor Trips a)
High Pressurizer Pressure f2"'5psig-b)
High Containment Pressure JE5psig c)
Low Steam Generator Pressure 2485psig This trip may be bypassed when the steam generator pressure is less than 100 psig above the setpoint.
d)
Low Steam Generator Water Level At or above the centerline of the fe'edwater ring (5' - 0" below normal wa ter level).
Basis:
A reactor trip at high power level (neutron flux) is provided to prevent' damage to the fuel cladding resulting from those reactivity excursions too rapid to result in a high pressure or thermal margin trip. The prescribed setpoint, with allowance for errors, is conservative relative to the trip point used in the accident analysis.
The high rate-of-changa of power reactor trip does not appear in the specification as this trip is not used in the transient and accident analys is.
This trip provides protection during reactor startup and is s,et at 2.6 decades per minute.
The thermal margin / low pressure trip is provided to prevent the fuel from exceeding thermal criteria (1)
The low setpoint of 1835 psig trips the reactor in the unlikely event of a loss-of-coolant accident.
The coefficients A, B, and C and the functions Al and QR1 are fuel cycle dependent and cust be determined for each fuel loading cycle.
They mus t be determined with methods that have been previously reviewed and found acceptable by the NRC.
ybe symmetric offset trip is provided to assure that excessive axial,
peaking will not cause fuel damage. The symmetric offset is determined from the axially split excore detecotrs. The symmetric off set trip and pre-trip in conjunction with the thermal margin / low pressure trip assures that neither a DNBR of less than 1.3 nor a maximum linear heat rate of more than 21.0 kw/ft will exist as a consequence of axial power i
maldis' tributions. The symmetric offset trip and -pre-trip are fuel cycle dcpendent and must be determined for each fuel loading cycle. They must be determined with methods that have been previously reviewed and found acceptabla by the NRC.
These coefficients are derived from analysis of many axial power shapes j
with allowances - for ins trumenta tion inaccuracies, and the uncertainty l
associated with the excore to incore symmetric offset relationship.
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',l 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 '9UTDOW MARGIN A sufficient SHUTDOW MARGIN ensures that 1) the reactor Can be made sub-critical 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 subcritical to preclude inadvertent criticality in the shutdown condition.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + (0.38)% delta K or R + (0.28)% delta K, as appro-priate. The value of R in units of % delta K is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be positive or zero and must be determined for each fuel loading cycle.
Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOW MARGIN.
The highest worth rod may be determined analytically or by test. The SHUTDOW MARGIN is demonstrated by an insequence control rod withdrawal at t% beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicates that the cargin could be reduced as a function of exposure. Observation of subcritice.lity in this condition assures subcriticality with.the most reactive control rod fully withdrawn.
This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOW MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the a
changes in reactivity can be inferrfd from these_ comparisons of rod patterns.
Since the comparisons are easily done, frequent checks are not an. imposition on normal operations. A 1% change is larger'than is expected for normal i
operation so a, change of this magnitude should be thoroughly evaluated.
A.
change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.
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