ML17053C991
| ML17053C991 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/24/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17053C992 | List: |
| References | |
| NUDOCS 8201150051 | |
| Download: ML17053C991 (24) | |
Text
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0'NITEDSTATES NUCLEAR R EGULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION UNIT NO.
1 ".
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
47 License No. DPR-63 l.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated October 26, 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 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; C.
Ther e is reasonable assurance
( i) that the activities authorized by this amendment can be conducted without endang'ering 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 amendment and paragraph 2.C.(2) of Facility License No. DPR-63 is hereby amended to read as follows:
(2)
Technical.S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 47, are hereby incorporated in the license.
The licensee shall operate the facility in.
accordance with the Technical Specifications.
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PDR '~ADOCK',05000220,
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 24, 1981.
Thomas
. I po ito, Chief Operating Reactors Branch 82 Division of Licensing
0
ATTAGHMENT TO LICENSE AMENDMENT NO. 4",
FACILITY OPERATING LICENSE NO.
'DOCKET NO. 50-220 Revise Appendix A pages
-as fo)lows:
Remove 64b 64c 65 66 67 68 70 70b 70d Insert 64b 64c 65 66 67 68 70 70b 70d
LINITIHG CONDITION FOR OPERATION If at arp time during power operation it is determined by normal sur veillance that the limiting
'alue for the power/flow'elationship is being
- exceeded, action shall be initiated within 15 minutes to restore operation to within the prescrjbed limits. If the power/flow relationship is not returned to within the prescribed limits within two
( 2) hours, reactor power reductions shall be
~ initiated at a rate not less than 10X per hour untjl the power/fl ow relationship is within the prescribeg limits.
e.
Partial Loo 0 eration During power operation, partial loop operation js permjtted provided the fpllowing conditions are met, When operating wifh fout recirculation loops jn operation and the remaining loop unisolated, the reactor m@r operate at ]00 percent of full licensed power level jn accordance with Figure 3.1,7aa and ap APLHGR not to exceed 98 percent of the limiting values shown in Figures 3.1. 7a, 3.1, 7b, 3.1',7c, 3.1.7d, and 3.1.7e.
Mhen operating with four recirculation loops in operation and one 1pop isolated, the reactor mW operate at 100 percgnt of fu)l licensed power in accordance with Figure 3.1.7aa and an APLHGR not to.
exceed 98 percent of )he limiting values shown in Figures 3.1.7a, 3.1.7b, 3.1. 7c;. 3.1.7d, and 3.1.7e, provided the -following conditions are met for the isolated loop.
1.
Suction valye, discharge valve and discharge bypass valve in the isolated loop shall be in the closed position and the associated motor breakers shall be locked iq the open position.
SURYEILLAHCE RE UIREHEHT
~
~
0'mendment No. PY, 47 64b
LIllITINGCONDITION FOR OPERATION 2.
Associated pump motor circuit breaker shall be opened and the breaker
- remoVed, If these conditions are not met, core power sha]1 be restricted to 90.5 percent of full licensed
- power, When operating y(jth three recirculation loops in
.operation and the two reqajning loops isolated or unisolated, the reactor pay operate at 90 percent of
.full licensed power jn accordance with Figure 3,1,7aa and an APLHGR not to exceed 96 percent of the limiting values shown jn Figures 3.1.7a, 3'-1.74, 3,1.7c, 3.1.7d, and 4,1,7e, During 3 loop operatjon,. the lipiting HCPR shall be increased by O.Ol, Power operation is not permitted wjth less t$aq three
'ecjrculatjon loops jn operation.
If at any time during power operation it is determinned by norma] survej))ance that the ljmjtjng value for APLHGR under one and two isolated loop operation is Peing exceed at any node in the cope, action shall be initjated within. 15 minutes to restore operation to within the prescribed limits,.
If the APL)lGR at all,nodes in the core is not.
returned to wjthjn the prescribed limits for one and two isolated loop operation within two (2) hours, reactor power reduction shall be jnjtiated at a rate not less than 10 percent per hour until APLHGR at a])
nodes is >cithin the prescribed limits.
SURVEILLANCE RE UIREMENT
'mendment No. 47 64c
0
(
NINE MILE POINT UNIT l
)0 CX O
- 8. 58 8.57 8.67 8.73
- 8. /f 8.64 8.60 8.58 8.50
>C 7
0
~ 5 10 20 25 30 35 40 AVERAGE PLAl,'AR EXPOSURE
{GHD/ST)
F jgure 3.1.7a ffAXIHUNALL014ABLE AVERAGE PLANAR LHGR APPLICABLE TO 8I)8250 FUEL AS DESCRIBED Amendment No. g', + 4j IN REFERE(CE 8.
l 65!
0 I-
NINE MILE POINT-UNIT 1
. 10 S 62:
- 8. 58 S. 66
- 8. 74 8.70 8.65
- 8. 60 8.56 8.48 8.48 0
5 35 30 10 15 20-
'5 AVERAGE PLAN'AR EXPOSURE (GHD/ST) figure 3,1.7b Maximum Allowable Average Planar LHGR Applicable to SDB274L and 8DB274H Fue] as described
$ n Reference 8.
e 40 Amendment No. Q, Pl 47 66
NINE HILE POINT UNIT 1 j0 9.26
.24 9.23 9,22 9.20 9.16 9.13 9.12
- 8. 83
- 8. 55 0
Amendment No. /, /VI 47 10 35 15
'20 25,.
30 AVERAGE. PLANAR EXPOSURE (GWD/ST)
Fjgure 3. 1.7c Haxjmum Allowab]e Average Planar LHGR Appljcable to 8DNB277
, Fuel as descrjbed in Reference 8.
45 I
67
NgNE MILE POINT-UNIT 1 10 9.30
.24 9,23 9.20 9.17
- 9. 06 8.90
- 8. 72
- 8. 46 Cej CD 4X Lai Amendment No. g; p(
47 10 15 20.
25 r 30 35 AVERAGE. PLANAR EXPOSURE (GWD/ST)
Figure 3. 1.7d Maximum Allowable Average Planar LHGR Applicable to P8DNB277 and Future Reload Fuel as described in Reference 8.
40 45 68 I
BASES FOR 3.1;7 AHD 4;1.7 FUEL RODS Avera e Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature and the peak local cladding oxidation following the pnstulated desjgn basis loss-of-coolant accident.will not exceed the limits specffjed fn 10CFR60, Appendix K.
The peak cladding temperature following a postulated loss-of-coolant accident fs primarily a functfon of the average heat generation rate of all the rods of a fuel assemhly at any axial location and is only dependent secondarily on the rod-to-rod power dfstrfputfon within an assembly.
Since expected lncal varfatfons in power distribution within a fuel assenbly affect the ca]culated peak clad temperature hy less than + 20 F re1atfve to the peak temperature for a typical fuel design, the limit on the. average linear heat generation rate 'fs sufficient to assure that calculated temperatures are within the 10CFR50, Appendix K limit.
The lfnftfng value for APLHGR fs shown jn Figure 3.1.7.
These l
curves are based on calculations using the mogels described ip References 1, 2, 3, 6, 6 and 13.
The Reference 13 LOCA analysis is sensitive to minimum critical pover ratio IHCPR).
In that analysis HCPR values of 1.30 for 5 loop operation and 1,36 fnr 4 and 3 loop operation, were assumede If future transient analyses should yield a MCPR limit be]ow either'f these values the Reference 13 LOCA analysis MCPR value would become limiting.
The current MCPR lfmjt js ].38.
Linear Heat Generatfon Rate (LHGR)
Thf s specffjcatfon assures that the lineat heat generation rate fn any rod fs less than the design linear heat generation even ff fpe] pellet densjffcatfnn fs pnstulated (Reference 12).
The LHGR shall he checked daily during reactor operation at ~ 26K power to determjne if fuel turnup ot control rod movement has caused changes fn power distribution.
Minimum Critical Power Ratfn (MCPR-)
At core thermal power levels less than or eaual to 26K, the reactor will he operating at a minimum recirculation pump speed and the mnderator void content will he very small.
For all designated control rod patterns which may be employed at this point, operating plant experience and therrlal-hydraulic analysis indicated that the resulting MCPR value js fn excess nf requfrements by a cnnsfderahle margin.
llfth this low void content, any inadvertent cnre flow increase would only. place operation fn a more conservative mnde relative tn MCPR.
During initial startup testing Amendment No. ~, jVI '7 70
BASES FOR 3.1.7 AND 4'1.7 FUEL RODS Partial Loo 0
eration'pe requirements of Specjficatjon 3.1.7e for. partial loop operation jq which the idle loop is isolated precludes the inadvertent startup of a recirculation pump with a cold leg.
However, if these conditions cannot be met, power level is restricted to 90e5 percent power based on current transient analysis (Reference
-9).
For three loop operation power',
)evel js restricted to 90 percent power based on the Reference 13 LOCA analysis.
I s
The results of. the ECCS calculation are affected by one or more recirculation loops Peing unisolated and out of
- service, This js due to the fact that credit is taken for extended nucleate boiling caused by flow coastdown in the unbroken loops.
The reduced core floral coastdown following the break results in higher peak clad temperature due to an earlier boiling transition time, The results of the ECCS calculations are also affected by one or more recirculation loops being isolated and out of service.
The mass of water in the isolated loops unavailable during blowdown results iri a earlier uncovery tjpe for the hot node.
This results jn an increase in the peak clad temperature.
To assure peak c] ad temperatures remain belou 2200op, ana1ysis has shove that the limiting average planar linear ~
heat generation rate for each fuel type shall be reduced 2 percent and 4 percent for 4 and 3 loop operation respectively (Reference 13).
Partial loop operation and its effect on lower plenum flo1v distribution is summarized jn Reference ll.
Since the lower plenum hydraulic design in a non-jet pump reactor js virtually identical to a jet pump reactor, application of these results is justified.
Additionally, non-jet pump. plants contain a cylindrical baffle plate which surrounds the guide tubes and distributes t11e impinging water jet and forces f)oy in a cjrcumferentjal direction around the outside of the baffle.
\\
Recirculation Loo s Requiring the suction and discharge for at leasg two (2) recjrculaj:ion loops to be full open assures that an adequate flow path exjsts from the annular region between the pressure vessel wall and the core shroud, to the core region.
This provides fop communication between those areas thus asqurjng that reactor water level instrument readings are indicative of the water level i.n the core region, 1<hen the reactor vesse] js'flooded to the level of the main steam line nozzle, communication between the core region and annulus exists above the core to ensure that indicative water level monitoring in the core region exists.
When the steam separators and dryer are. removed, safety limit 2.l.ld and e requires water level to be higher than 9 feet below minimum normal. water leve) (Elevation 302'9"),
This level is above the core shroud elevation which would ensure connnunjcation between the core region and annulus thus ensuring indicative water level monitoring in the core region.
Therefore, maintaining a recirculation'loop jq the full open position in these two instances are not necessary to ensure indicative water level monjtorjng.
70b Amendment No.~
47
REFERENCES FOR BASES 3.1.7 AND 4.1.7 FUEL RODS (1)
"Fuel Densification Effects on General Electric Boiljng 1later Reactor Fuel", Supplements 6,
7 and 8, NEDM-10735, August 1973.
(2)
Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (USAEC Regulatory StaffI.
I I
.1 (3)
Conmunicafion; II,A, t<oore to I.S, Mitchel], "Modjfied GE godel for Fuel Densification", Docket 50-321, March 27, 1974
~
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(4)
"-General Elecjric Boiling Matey Reactor Generic Reload Application for 8 x 8 Fuel", NED0-20360, Supplement 1
to gevjsion 1,'ecember 197$.
I (5)
"General Electric Company Analytical Model for Loss of Coolant Analysis ip Accordance with 10CFR50 Appendix KN NED0-20566.
(6)
General Electric Refill Reflood Calculation (Suppleme'nt to SAFE Code Description) transmitted to the USAEC by letter, G.L. Gyorey to Victor Stello Jr.,
dated December 20, 1974.
(7)
"Nipe Mile pojnt tluclear Power Stytjon Unit 1, Load Ljne LinitAnalysis", NED0-24012.
(8)
Licensing Topical Report Genera] Electric Boiljng Hater Reactor Generic Reload Fuel Application, NEDE-2401]-P-A, Augu'st, 1978.
(9)
Final Safety Analysjs RePort, Nine Mile Point Nuclear Station,.Niagara Mohawk PoWer Corporation, June 1967..
(10)
NRC Safety Evaluation, Amendment No.
24 to.DPR-63 contained in a letter from George Lear, NRC, to D.P. Disc dated May 15, 1978.
P
(]1) "Core Flow Distrjbution in a General Electric Poiljng pater Reactor as Measured in squad Cities Unit 1",
HED0-10722A.
(12) Nine Mile point Nuclear Power Statjon Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6), HE00-24185, April 1979.
(13) Loss of Coolant Accident Analysis Report for Nine Hjle Point Unit One Nuclear Power Station, NED0-24348, August 1981'0d Amendment No. 47
0