ML17055D812
| ML17055D812 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/19/1988 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17055D811 | List: |
| References | |
| NUDOCS 8805040045 | |
| Download: ML17055D812 (30) | |
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0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No.
DPR-63 1.
The Nuclear Regulatory Commission {the Commission) has found that-.
A.
The applications for amendment by Niagara Mohawk Power Corporation of New York, Inc. (the licensee) dated August 21, September 14, December 17, and December 18, 1987; and as supplemented March 9, 1988, complv 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.
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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-63 is hereby amended to read as follows:
(
BSOS04<<'I5 BgOOOmO pDR ADOCN; 05 pDR p
(P.)
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No.
97'are hereby incorporated into this license. 'he licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert A. Capra, Director Project Directorate I-1 Division of Reactor Projects, I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 19, 1988
I
,R
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
97 TO FACILITY OPERATINA LICENSE NO.
DPR-63 DOCKET NO. 50-220 Revise Appendix A as follows:
Remove Pa es Insert Pa es 20 63 64a 64b 64c 69a 70 70b 70d 20 63 64a 64b 64c 69a 69al 70 70b 70d
l4 *-\\
'I'W
FIGURE 2.1.1 FLOMl BIASED SCRAM AND APRM ROD BLOCK.
g
%2D NZES:
1.
Rated Power is 1850 M 2.
Design Flow is 67.5 x 10 lb/hr 3.
Calculated Total Peaking Factor ( MFPF 4.
Core Pressure is>>800 psia SCRAM C) 100 OC 60 CO ROD BLOCK For Calculated Total Peaking Factors
> MEPF MFPP S
= ~~ xS 0
In cases where for a short period the total peaking factor (PKFL) exceeds the maxim'otal peaking factor (tHPF), rather than adjusting the APRH setpoints, the APRH gain may be adjusted so that the APRH readings are greater than or equal to core power x PKFL/Mls provided that the adjusted APRN reading does not exceed 100% of rated t~ power and a notice of adjustment is posted on the reactor contml panel.
3.02 for 8 x 8 Fuel 3.00 for 8 x 8R and P8 x 8R Fuel 2.90 for GE 8 x 8EB Fuel Sn The new Scram and Rod Block Lines PKFL
~
Calculated Total Peaking Factor S
~
Scram and Rod Block Lines Shown Above 0
10 20 30 40 50 60 70 80 90 t00
")10
'120 REC) RCULATION FLOW, PERCENT OF DESlQN Amendment No; Pg, Q, N >>
97 8
l
BASES FOR 2.1.1 FUEL CLADDING SAFETY LIHIT Because the boiling transition correlation is based on a large quantity of full scale data there is' very high confidence that operation of a fuel assembly at the condition of the SLCPR would not produce boiling transition.
- Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity.
However, if boiling transition were to occur, clad perforation would not be expected.
Cladding temperatures would increase to approximately 1100 F which is below the perforation temperature of the cladding material.
This has been verified by tests in the General Electric Test Reactor (GETR) where similar fuel operateg above the critical heat flux for a significant period of time (30 minutes) without clad perforation.
If reactor pressure should ever exceed 1400 psi a during normal power operating (the limit of applicability of the boiling transition correlation) it would be assumed that the fuel cladding integrity safety limit has been violated.
In addition to the boiling transition limit SLCPR operation is constrained to a maximum LHGR of 13.4 kH/ft for
- 8x8, 8x8R, PBx8R and GE8xBEB fuel (Reference 15).
At 100K power, this limit is reached with a Haximum Total Peaking Factor (HTPF) of 3.02 for 8x8 fuel, 3.00 for Bx8R and PBx8R fuel, and 2.90 for GE8x8EB fuel.
During steady-state operation where the total peaking factor is above 2.90, the equation in Figure 2.1.1 will be used to adjust the flow biased scram and APRH rod block set points.
At pressure equal to or below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56
- psi, At low power and all core flows, this pressure differential is maintained in the bypass region of the core.
Since the pressure drop in the bypass region is essentially all elevation
- head, the core pressure drop at low powers and all flows will always be greater than 4.56 psi.
Analyses show that with a bundle flow of 28xl03 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
Therefore, due to the 4.56 psi driving head, the bundle flow will be greater than 28xl03 lb/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel, assembly critical power at 28xl03 lb/hr Amendment No. 5, 37, O'Y~ 97
REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1)
General Electric BHR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO-10958 and NEDE-10958.
(2)
Linford, R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Hater Reactor,"
NED0-10801, February 1973.
(3)
- FSAR, Volume II, Appendix E.
(4)
- FSAR, Second Supplement.
(5)
- FSAR, Volume II, Appendix E..
(6)
- FSAR, Second Supplement-.
(7)
- Letters, Peter A. Morris, Director of Reactor Licensing, USAEC, to John E.
Logan, Vice-President, Jersey Central Power and Light Company, dated November 22, 1967 and January 9,
1968.
(8)
Technical Supplement to Petition to Increase Power Level, dated April 1970.
(9)
- Letter, T. J.
- Brosnan, Niagara Mohawk-Power Corporation, to Peter A. Morris, Division of Reactor Licensing, USAEC, dated February.28, 1972.
( 10)
Letter, Philip D. Raymond, Niagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973.
(11)
Nine Mlle Point Nuclear Power Station Unit 1
Load Line Limit Analysis, NEDD 24012,
- May, 1977.
(12)
Licensing Topical Report General Electric Boiling Hater Reactor Generic Reload Fuel Application, NEDE-24011-P-A,
- May, 19&6.
(13)
Nine Mile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6), NED0-24185, April 1979.
( 14)
General Electric'IL 299 "High Drywell Temperature Effect on Reactor Vessel Hater Level Instrumentation."
(15)
Letter (and attachments) from C.
Thomas (NRC) to J. Charnley (GE) dated May 28,
- 1985, "Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-B, Amendment 10."
Amendment No. Q 20
4
LIHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREHENT 3.1.7 FUEL RODS 4.1.7 FUEL RODS
~A i icabal it:
The Limiting Conditions for Operation associated with the fuel rods apply to those parameters whi ch monitor the fuel rod operating conditions.
~ob '
cti ve:
The objective of the Limiting Condi tions for Operation is to assure the performance of the fuel rods.
~Ail billet The Surveillance Requirements apply to the parameters which moni tor the fuel rod operating condi tions.
O~bectiye:
The objective of the Surveillance Requirements i s to. specify the type and frequency of surveillance to be applied to the fuel rods.
I e~ifi ti a.
Avera e Planar I inear Heat Generation Rate (APLHGR) as if) ti a.
Avera e Planar Linear Heat Generation Rate (APLHGR)
During power operation, the APLHGR for each type of fuel as a function of average planar exposure shall not exceed the limiting value shown in Figures 3.1.7a, 3.1.7b, 3.1.7c, 3.1.7d, 3.1.7e, 3.1.7f and 3.1.7g.
If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded at any node in the core, action shall be initiated within 15 minutes to restore operation to within'he prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10K per hour until APLHGR at all nodes is within the prescribed limits.
Amendment No. gg, Q The APLHGR for each type of fuel as a
function of average planar exposure shall be determined daily during reactor operation at
>25 percent rated thermal power.
63
C.
LIMITING CONDITION FOR OPFRATION Minimum Critical Power Ratio (HCPR)
During power operation, the HCPR for all 8
x 8
fuel at rated power and flow shall be as shown in Lhe table below:
LIMITING CONDITION FOR OPERATION HCPR.
C R
I t
1 E
~LE ~1tl MCPR'URVEILLANCE RE UIREMENT c.
Minimum Critical Power Ratio (HCPR)
HCPR shall be determined daily dur'ing reactor power operation at
> 25K rated thermal power..
d.
Power Flow Relationshi Compliance with the power flow relationship in Section 3.1.7.d shall be determined daily during reactor operation.
tt 1~EM Under partial loop operation, surveillance requirements 4.1.7.a,b,c and d above are applicable.
d.
If at any time during power operation it is
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~
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determined by normal surveillance that the above limit is no longer met, action shall be initiated within 15 minutes to restore operation to within the prescribed limit. If all the operating MCPRs are not returned to within the prescribed limit within two (2) hours, reactor power reductions shall be initiated at a rate not less than lOX per hour until MCPR is within the prescribed
( limit.
For core flows other than rated the MCPR limit shall be the limit identified above times Kf where Kf is as shown in Figure 3.1.7-1.
Power Flow Relationshi Durin 0 eration The power/flow relationship shall not exceed the limiting values shown in Figure 3.1.7.aa.
'his limit shal 1
be determined to be applicable each operating cycle by analyses performed utilizing the ODYN transient code.
Amendment No. 5V 64a
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREHENT If at any time during power operation, it is determined by normal surveillance that the limiting value for the power/flow relationship is being
- exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed 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 10'L per hour until the power/flow relationship is within the prescribed limits'.
Partial Loo 0 eration During power operation, partial loop operation is permitted provided the following conditions are met.
Nhen operating with four recirculation loops in operation and the remaining loop unisolated, the reactor may operate at 100 percent of full licensed power level in accordance with Figure 3.1
~ 7aa and an 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 and an APLHGR not to exceed 99K of the limiting values shown in Figures 3.1.7f and 3.1.7g.
Hhen operating with four recirculation loops in operation and one loop isolated, the reactor may operate at 100 percent of full licensed power in accordance with Figure 3.1.7aa and an 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 and an APLHGR not to exceed 99'/ of the limiting values shown in Figures 3.1.7f and 3.1.7g, provided the following conditions are met for the isolated loop.
1.
Suction valve, discharge valve and discharge bypass valve in the isolated loop shall be in the closed position and the associated motor breakers shall be locked in the open position.
Amendment No. gg, 4'/.
97 64b
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREHENT 2.
Associated pump motor circuit breaker shall be opened and the breaker removed.
If these conditions are not met, core power shall be restricted to 90.5 percent of full licensed power.
Hhen operating with three recirculation loops in operation and the two remaining loops isolated or unisolated, the reactor may operate at 90X of full licensed power in accordance with Figure 3.1.7aa and an APLHGR not to exceed 96 percent of the limiting values shown in Figures 3.1.7a, 3.1.7b, 3.1.7c, 3.1.7d, and 3.1.7e and an APLHGR not to exceed 99'X of the limiting values shown in Figures 3.1.7f and 3.1.7g.
During 3 loop operation, the limiting HCPR shall be increased by 0.01.
Power operation is not permitted with less than three recirculation loops in operation.
If at any time during power operation, it is determined by normal surveillance that the limiting value for APLHGR under one and two isolated loop operation is.being exceeded at any node in the core, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits for one and two isolated loop operation within two (2) hours, reactor power. reduction shall be initiated at a rate not less than 10 percent per hour until APLHGR at all nodes is within the prescribed limits.
Amendment No. 8'i 97 64c
MAPLHGR LIMITS FOR P8DRB299 12 10 9
8 7
C7 Q
8 CD.
a 5
cD E
a 1 0.8 10.9 10.9 10.9/1 0.7
~ >0.<
9.6 9 5 9.3 9 2 9.0 LEGEND MAPLHGR 5
10 15 20 25 30 35 40 45 AVERAGE PLANAR EXPOSURE (GWD/ST)
Figure 3.1.7f Maximum Allowable Average Planar LHGR Applicable to PBDRB299 and Future Reload Fuel as described in Reference l5-Amendment No. Q, pp, 97 ega
, e
12 11.8 11.6 11A 11.2 MAPLHGR Limits for BD321B (GE8XSEB "FUEL) 11.84 11.83 11.59 11.40 11.05 K
L'0c0 lLI Ql0 L0 E3 E
X0 10.8 10.6 10A 10.2 10 9.8 9.6 9.2 10.51 10.40 10.76 10A9 9.97 9.40 8.8 8.6 0
10 20 30 I
l 40 8.74 50 Average Ptanar Exposure (GWD/ST)
Figure 3.1.7g Maximum Allowable Average Planar LHGR Applicable to BD321B and Future Reload Fuel as described in Reference 16 Amendment No. 97 69a 1
0
BASES FOR 3.1.7 AND 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 postulated design basis loss-of-coolant accident will not exceed the limits specified in
- 10CFRSO, Appendix K.
'he peak cladding =temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than
+ 20 F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the
- 10CFR50, Appendix K limit.
The limiting value for APLHGR is shown in Figures
- 3. 1. 7a-g.
These curves are based on calculations using the models described in References 1, 2, 3, 5, 6, 13, 15 and 16.
The Reference 13 and 15 LOCA analyses are sensitive to minimum critical power ratio (HCPR).
In the Reference 15, analysis a
HCPR value of 1.30 was assumed.
If future transient analyses should yield a HCPR limit below this value, the Reference 15 LOCA analysis HCPR value would become limiting.
The current HCPR limit is
> l.40.
For fuel bundles analyzed with the Reference 13 LOCA methodology, assume HCPR values of 1.30 and 1.36 for five recirculation loop and less than five loop operation respectively.
Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation even if fuel pellet densification is postulated (Reference 12).
The LHGR shall be checked daily during reactor operation at
> 25'X power to determine if fuel burnup or control rod movement has caused changes in power distribution.
Hinimum Critical Power Ratio (HCPR)
At core thermal power levels less than or equal to 25'X, the reactor will be operating at a minimum recirculation pump speed and the moderai.or void content will be very small'.
For all deslgpated control rod patterns which may be employed at this point, operating plant experience and thermal-hydraulic. analysis indicated that the resulting HCPR value is in excess of requirements by a considerable margin.
Hith this Iow void content, any inadvertent core flow increase would only place operation in a more conservative
'mode relative to HCPR.
During initial startup testing Amendment No. N 70
BASES FOR 3.1.7 AND 4.1.7 FUEL RODS Partial Loo 0 eration The requirements of Specification 3.1.7e for partial loop operation in 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 90.5 percent power based on current transient analysis (Reference 9)
~
For three loop l operation, power level is restricted to 90 percent power based on the Reference 13 and 15 LOCA analyses.
The results of the ECCS calculation are affected by one or more recirculation loops being unisolated and out of service.
This is due to the fact that credit is taken for extended nucleate boiling caused by flow coastdown in the unbroken loops.
The reduced core flow 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 in an earlier uncovery time for the hot node.
This results is an increase in the peak clad temperature.
l For fuel bundles analyzed with the methodology used in Reference 13, HAPLHGR shall be reduced 2'L and 4X for 4 and 3
loop operation respectively.
For fuel bundles analyzed with the methodology used in References 15 and 16, MAPLHGR shall be reduced by
- 11. for both 4 and 3 loop operation.
Partial loop operation and its effect on lower plenum flow distribution is summarized in Reference 11.
Since the lower plenum hydraulic design in a non-jet pump reactor is 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 the impinging water jet and forces flow in a circumferential direction around the outside of the baffle.
Recirculation Loo s
Requiring the suction and discharge for at least two (2) recirculation loops to be fully open assures that an adequate flow path exists from the annular region between the pressure vessel wall and the core shroud, to the core region.
This provides for communication between those areas, thus assuring that reactor water level instrument readings are indicative of the water level in the core region.
Nhen the reactor vessel is flooded to the level of the main steam line nozzle, communication between the core region and annul us exists above the core to ensure that i ndi cati ve water level monitoring in the core region exists.
Hhen 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 level (Elevation 302'9").
This level is above the core shroud elevation which would ensure communication between the core region and annulus thus ensuring indicative water level monitoring in the core region.
Therefore, maintaining a recirculation loop in the full open position in these two instances are not necessary to ensure indicative water level monitoring.
Amendment No. g9, 4'7',
97 70b
J
> ~
(2)
(3)
REFERENCES FOR BASES 3.1.7 ANO 4.1.7 FUEL RODS "Fuel Densification Effects on GE Boiling Water Reactor Fuel," Supplements 6,
7 and 8, NEDM-10735, August 1973.
Supplement 1 to Technical Report on Oensi fications of GE Reactor
- Fuels, December 14, 1974 (USAEC Regulatory Staff).
Communication:
V. A. Moore to I.
S. Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, Harch 27, 1974.
(4)
"GE Boiling Water Reactor Generic Reload Application for 8 x 8 Fuel," NED0-20360, Supplement December 1974.
1 to Rev.
1,-
(5)
GE Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix (6)
GE Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC Gyorey to Victor Stello, Jr.,
dated December 20, 1974.
"Nine Mile Point Nuclear Power Station Unit 1, Load Line Limit Analysis," NED0-24012.
K," NED0-20566.
by fetter, G.L.
(8)
(9)
(10)
(11)
(12)
(13)
(14) s (15)
(16)
Licensing Topical Report GE Boi ling Water Reactor Generic Reload Fuel Application, NEDE-24011-P-A, August 1978.
Final Safety Analysis Report, Nine Nile Point Nuclear Station, Niagara Mohawk Power Corporation, June 1967.
NRC Safety Evaluation, Amendment No. 24 to DPR-63 contained in letter from G. Lear, NRC, to D.
P. Disc dated Hay 15, 1978.
"Core Flow Distribution in a GE Boiling Water Reactor as Measured in Quad Cities Unit 1," NEO0-10722A.
Nine Mile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysts, License Amendment Submittal (Cycle 6), NED0-24185, April 1979.
Loss-of-Coolant Accident Analysis Report for Nine Hile Point Unit 1 Nuclear Power Station, NED0-24348, Aug.
1981.
GE Boiling Water Reactor Extended Load Line Limit Analysis for Nine Mile Point Unit 1 Cycle 9, NEOC-3l126, February 1986.
Nine Hile Point Unit 1, Loss-of-Coolant Accident Analysis, NEDC-31446P, June 1987.
Supplement 1 to Nine Hile Point Generating Station Unit 1
SAFER/CORECOOL/GESTR-LOCA Analysis Report NEDC-31446P-1, Class III, September 1987.
Amendment No. gj, 4i 70d