ML20071J000
| ML20071J000 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/20/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Boston Edison Co |
| Shared Package | |
| ML20071J001 | List: |
| References | |
| DPR-35-A-054 NUDOCS 8204260047 | |
| Download: ML20071J000 (11) | |
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j UNITED STATES
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NUCLEAR REGULATORY COMMISSION 8^l)*(
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WASHINGTON, D. C. 20555
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BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM HUCLEAR POWER STATION AMENDWNT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. DPR-35 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Boston Edison Company (the licensee) dated September 22, 1981, as supplemented by letters dated October 5 and December 2,1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chaoter I; 8.
The facility will operate in confornity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license unendment and paragraph 3.B of Facility Operating License No. DPR-35 is hereby amended to read as follows:
B.-
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 54, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
820426000
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- 3. - This license amendment is effective as of the date of its issuance.
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FOR THE NUCLEAR REGULATORY COMMISSION l.
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'Wh" Domenic B. Vassallo, Chief Operating Reactors Branch #2 l
Division of Licensing 3
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Attachment:
Changes to the Technical I
Specifications
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Date of Issuance: March 20,1982 i
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1 ATTACHMENT TO LICENSE AMENDMENT NO. 54 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 l
l Replace the following pages of the Appendix "A" Technical Specifications l
- with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Insert 205B 205B 205B-1 2058-2 l
205C 205C 205C-2 205C-2 205C-3 205C-3 205D 205D 205E-6 L
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o SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION C.
Plnimum Critien1 Power Ratio (MCPR)
C.
Minimum Critical Power Ratio (MCPR) 1.
MCPR shall be determined daily during 1.
During power operation MCPR shall be reactor power operation at )> 25% rated 25 the MCPR operating limit specified in 3.11.C.2.
If at any time during thermal power and following any change operation it is determined by normal in power level or distribution that would cause operation with a limiting surveillance that the limiting value control rod pattern as described in for MCPR is being exceeded, action the bases for Specification 3.3.B.5.
shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady 2.
The value of 7 fin Specification 3.11.C.2. shall be equal to 1.0 state MCPR is not returned to with-unless determined f rom the result in the prescribed limits within two of surveillance testing of Specif1-(2) hours, the reactor shall be cation 4. 3.C as follows :
brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveil-a) 17' is defined as.
lance and corresponding action shall continue until reactor operation is a:=[Ive
_B within the prescribed limits.
1.275
'B For core flows other than rated the MCPR limits shall be the limits identified above times Kf where Kg is as shown in b)
The average scram time to the 30% insertion position is deter-Figure 3.11-8 mined as follows:
As an alternative method providing n
equivalent thermal-hydraulic protection j[ N li i
core flows other than rated, the cal-1[ ave =i"I at culated MCPR may be divided by Kf, where j(( 3 n
is as shown in Figure 3.11-8.
1 Kf i=1 2.
The operating limit MCPR values as a function of 2 Pare given b Table 3.11-1 where: n = number of surveillance where'tf is given by specification tests performed to date in the cy cle.
4.11. C. 2.
205B Amendment No. JZ 54 em
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SURVEILLANCE REQUIREMENTS LI!!ITING CONDITIONS FOR OPERATION N = number of active control rods g
measured in the ith surveillance i
test.
t 17 f = average scram time to the 307:
insertion position of all rods measured in the ith surveill6n-a test.
c.)
The adiusted analysis mean scran time (I'B) is calculated as follows:
B */c + 1.65 [
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II N i=1 17h ere :
4 = mean of the distribution for f.
average scram insertion time to the 30% position 0.945 sec i = total number of active control rod N
measured in specification 4.3.C (7 = standard deviation of the distributien for average scram insertion time to the 30% position, 0.064 sec.
4 Amendment No. 54 205B-1 a,.o
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TABLE 3.11-1 OPERATING LIMIT MCPR VALUES i
MCPR Operating Limit II' 8x8 P8x8R 6.0 1.32 1.35 0 to.1 1.32 1.36
.1 to.2 1.33 1.36
.2 to.3 1.33 1.36
.3 to.4 1.34 1.37
.4 to.5 1.34 1.37
.5 to.6 1.35 1.38~
.6 to.7 1.35 1.38
.7 to.8 1.36 1.39
.8 to.9 1.36 1.39
.9 to 1.0 1.37 1.40 Q
1 Amendment No. 54 205B-2 1
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BASIS Aversee Planar Linear Rest Ceneration Rate (APMCR)_
3.11A This specifications assures that the peak cladding te=perature f ollowing the postulated design basis loss-of-coolant accident i
will not exceed the li=it specified in the 10 CFR 50 Appendix I.
The peak cladding te=perature (PCI) following a postulatedlos heat gezieration rate of all the rods of a fuel.asse=bly at any secondarily on the rod axial location and is only dependent, The peak clad to rod power distribution withis an asse=bly.
te=perature is calculated assu=ing a MCR for the highest powered rod which is equal to or less than the design MCR.
This LHCR times 1.02 is used in the heat-up code along with the exposure dependert steady state gap conductance and rod-to-rod local peaking factors. The li=iting value for APGCR is this LECR of the highest powered rod divided by its local peaking factor.
The calculational procedure used to establish the APLEOR li=it for each fuel type is based on a loss-of-coolant accident analysis.
The energency core cooling syste= (ECCS) evaluation models which are e= ployed to deter =ine the effects of the loss of coolant accident (LOCA) in accordance with 10CTR50 and Appendix K are The sedels are identified as IA=.!,
discussed in Reference 1.
SCAT, SAFE, RE7LOOD, and GASTE. 'The LAM 3 Code calculates the short ter= blevdawn response and core flov, w'.ich are input into the SCAT code to calculate blevdown heat transfer coef ficients.
The SAFE code is used to deter =ine longer tern syste= responseiihere appropriate, and flows from the various ECC syste=s.
output of SATI is used in the RE7100D code to calculate liquid The results of these codes are used in the CHASTE code levels.
average pla=ar to calculate fuel clad te=peratures and maxiarm::
linear heat generation rates O'.APLBGR) for each fuel type.
The significant plant input parameters and the MAPLEGR's for fuel types calculated by the above procedure are the present The curves in Figures 3.11-1 included in Reference 2.
through 3.11-6 were developed assuming no core spray heat transfer credit in the IDCA analysis.
l Amendment No. Y 54 205C j
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REDENCES General Electric BWR Ceneric Reload Fuel Application NEDE-2I.011-P.
1.
Less of Coolant Accident Analysis Report for Pilgrim Nuclear Power 2.
1 Station, NEDO-21696, August 1977 as amended.
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- Amahbhent No.)T 54 205c-2 m
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MDQt?M CRInCAL POWER ItAMO (MM) 3.11C Operatina Limit NC7R For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state it is required that the resulting MCFR does not i
operating limit, decrease below the Safety Limit MC71 at a=y t1=e during the l
transient assuming instru=ent trip setting given in Specification De difference between the specified Operating Linit MCFR in Specification 3.11C and the Safety Limit MCPR in Specificacian 1.1A defines the largest reduction in eritical power ratio (CPR) permitted during any anticipated abnornal operating transient.
To ensure that this reduction is not exceeded, the most li-iting transients are analized for each relcad and fuel type (8x8 and PSxSR) to deter =ine that transient which yields the largest value This value, when added to the Saferf Limit MCPR must of A C7R.
be less than the mini =us operating 11 sit M.CPR's of Specification ne reruit of this evaluation is doctz:ented in the
" Supple = ental Reload Licensing Subtittal" f or the current reload.
3.11.C.
ne evaluation of a given tranatent begins with the syste= gut para =eters shown in Tables 5-4, 5-6 and 5-8 of NDE-24011-F Supplemented by reload unique inputs gives in the currentnese values are input Supple = ental Reload Licensing Submittal,to a GE core dyna =ic behavior in NDo-10802(2). The transient code used for all pressurisation The MCPR eventh is described in NEDE-24154-P (Ref erence 5).
analysis for pressurization events is done in accordance with the procedures given in Reference 6.
Amendment No. pf 54 205C-3 0
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r REFERENCE 5_
l General Electric B W Generic Reload Yuel Application, ND E-240ll-P.
1.
- 1. 3. Linford, Analytical Methods of Flant Transient Evaluations 2.
for the GE BWK, Februa:7 1973 (NDO-10802).
General Electric Conpany Analytical Model for Iass-of-Coolast 3.
Analysis in Accordance with 10 CTR 50 Appe=diz K. NDE-20566 (Draf t). August 1974 Letter from J. E. Howard, Boston Edison Company to D. L. Zima-t 4.
USNRC, dated October 31, 1973.
5.
Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, October 1978 (NEDE-24154-P).
69 Letter, R. P. Denise (NRC) to G. G. Sherwood (GE), January 23, 1980*
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/cend=ent No. J2' 54 205D 8
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a FIGURE 3.11-6 MAXIMUM AVERAGE PLAMAR LINEAR HEAT GEMERATioM FATE VERSUS FLANAR AVERAGE EXPOSURE FUEL TYPE P,83RB 265 H i
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Flow % 90f ypygn derti 7
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10 4 9.3 9.9 1w S.8 9.8 gI as SS
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5,000 10,000 15,000 20,000.
25p00 3Q000 PLANAR AVERAGE EXPOSURE (MWdh)
Amendment No. 54 205E-6 a
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