ML17055D462

Proposed Tech Specs 2.1.1, Fuel Cladding & 3.1.7, Fuel Rods & Associated Bases,Changing Specs for Reload Fuel Scheduled for Spring 1988

ML17055D462
Person / Time
Site:	Nine Mile Point
Issue date:	12/18/1987
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17055D460	List: ML17055D459 ML17055D461 ML17055D462 ML18038A382 ML18038A758
References
NUDOCS 8712290214
Download: ML17055D462 (28)

Text

ATTACHMENT A NIAGARA MOHAWK POWER CORPORATION LICENSE NO.

DPR-63 DOCKET NO. 50-220 Pro osed Chan es to Technical S ecifications (A

endix A)

This proposal requests that the existing pages of our Technical Specification be replaced with the attached revised pages as shown below.

These pages have been retyped and the marginal markings indicate changes.

Existin Pa e

Revised Pa e

20 63 64a 64b 64c 70 70b 70d 20 63 64a 64b 64c 69al 70 70b 70d The attached revised pages

64b, 64c, 70, 70b and 70d also reflect changes proposed in our August 21, 1987 Technical Specification Amendment Application.

Our application of August 21, 1987, requires approval prior to or concurrent with the application contained herein.

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BASES FOR 2.1.1 FUEL CLADDING SAFETY LIMIT Because the boiling transition correlation is based on a large quantity of full scale data there is a 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'ould 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 operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.

If reactor pressure should ever exceed 1400 psia 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 kN/ft for

8x8, Bx8R, P8x8R and GE8x8EB fuel (Reference 15).

At 100/. power, this limit is reached with a Maximum Total Peaking Factor (MTPF) of 3.02 for 8x8 fuel, 3.00 for 8x8R and P8x8R 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 APRM 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

REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1)

General Electric BWR 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 Water 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.

(ll)

Nine Mile Point Nuclear Power Station Unit 1

Load Line Limit Analysis, NEDO 24012,

Hay, 1977.

I (12)

Licensing Topical Report General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P-A,

May, 1986.

(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 SIL 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation."

(15)

Letter (and attachments) from C.

Thomas (NRC) to J. Charnley (GE) dated Hay 28,

1985, "Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-B, Amendment 10."

20

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.7 FUEL RODS A

1 i cabi 1 it 4.1.7 FUEL RODS A

1 i cabi 1 it The Limiting Conditions for Operation associated with the fuel rods apply to those parameters which monitor the fuel rod operating conditions.

O~b'ective:

The objective of the Limiting Conditions for Operation is to assure the performance of the fuel rods.

The Surveillance Requirements apply to the parameters which monitor the fuel rod operating conditions.

O~bective:

The objective of the Surveillance Requirements is to specify the type and frequency of survei.llance to be applied to the fuel rods'li

~

t'i e ifi a.

Avera e Planar Linear Heat Generation Rate (APLHGR)

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 the 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 10% per hour until APLHGR at all nodes is within the prescribed limits.

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

BOC to EOC 1.40 If at any time during power operation it is 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 10%

per hour until MCPR is within the prescribed 1 imit.

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.

d.

Power Flow Relationshi Durin 0 eration The power/flow relationship shall not exceed the limiting values shown in Figure 3.1.7.aa.

This limit shall be determined to be applicable each operating cycle by analyses performed utilizing the ODYN transient code.

LIMITING CONDITION FOR OPERATION c.

Minimum Critical Power Ratio (MCPR)

During power operation, the MCPR for all 8

x 8 fuel at rated power and flow shall be as shown in the table below:

LIMITING CONDITION FOR OPERATION MCPR C

A tt R

~lite PR'URVEILLANCE RE UIREMENT c.

Minimum Critical Power Ratio (MCPR)

MCPR shall be determined daily during reactor power operation at

> 25% 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.

e.

Partial Loo 0 eration Under partial loop operation, surveillance requirements 4.1.7.a,b,c and d above are applicable.

64a

LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREMENT 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'/ per hour until the power/flow relationship is within the prescribed limits.

e.

Partial Loo 0 eration During power operation, partial loop operation is permitted provided the following conditions are met.

When operating with four recirculation loops in operation and the remaining loop uninsolated, 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 99% of the limiting values shown in

( Figures 3.1.7f and 3.1.7g.

When 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, 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 condi tions are met for the isolate loop.

1.

Suction valve, discharge valve and discharge bypass valve in the isolated loop shall be in the closed position and the associated motor

'reakers shal be locked in the open position.

64b

LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREMENT 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 90% 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 991. of the limiting values shown in Figures 3.1.7f and 3.1.7g.

During 3 loop operation, the limiting MCPR 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.

64c

12 11.8 11.6 11.4 11.2 MAPLH.GR Limits for BD321B (GESX8EB FUEL) 11.84 11.83 11.59 11AO 11.05 10.8 10.6 10.4 10.2 10 9.8 9.6 9A 9.2 10.51 10.40 1080 1076 10A9 9.97 9.40 8.8 8.6 0

10 20 30 40 8.74 50 Average Planar Exposure {GND/ST)

't ~

Figure 3.1.7g Maximum Allowable Average Planar LHGR Applicable to 803218 and Future Reload Fuel as described in Reference 16 69a 1

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 10CFR50, Appendix K.

The 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 (MCPR).

In the Reference 15, analysis a

MCPR value of 1.30 was assumed.

If future transient analyses should yield a MCPR limit below this value, the Reference 15 LOCA analysis MCPR value would become limiting.

The current MCPR limit is

> 1.40.

For fuel bundles analyzed with the Reference 13 LOCA methodology, assume MCPR 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'L power to determine if fuel burnup or control rod movement has caused changes in power distribution.

Minimum Critical Power Ratio (MCPR)

At core thermal power levels less than or equal to 25K, the reactor will be operating at a minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience and thermal-hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin.

Hith this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

During initial startup testing 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 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 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.

For fuel bundles analyzed with the methodology used in Reference 13, MAPLHGR shall be reduced 2X and 4% for 4 and 3

loop operation respectively.

For fuel bundles analyzed with the methodology used in References 15 and 16, MAPLMGR shall be reduced by ll for both 4 and 3 loop operation.

Partial loop operation and its effect on lower plenum flow distribution is summarized in Reference ll.

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.

Hhen the reactor vessel is 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.

Nhen the steam separators and dryer are removed, safety limit 2.1.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.

70b

REFERENCES FOR BASES 3.1.7 AND 4.1.7 FUEL RODS (1) "Fuel Densification Effects on GE Boiling Hater Reactor Fuel," Supplements 6,

7 and 8, NEDM-10735, August 1973.

(2) Supplement 1 to Technical Report on Densifications of GE Reactor

Fuels, December,14, 1974 (USAEC Regulatory Staff).

(3) Communication:

V. A. Moore to I. S. Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.

(4)

"GE Boiling Hater Reactor Generi.c Reload Application for 8 x 8 Fuel," NED0-20360, Supplement 1 to Rev.

1, December 1974.

(5)

GE Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K," NED0-20566.

(6)

GE Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G.L.

Gyorey to Victor Stello, Jr.,

dated December 20, 1974.

(7) "Nine Mile Point Nuclear Power Station Unit 1, Load Line Limit Analysis," NED0-24012.

(8) Licensing Topical Report GE Boiling Hater Reactor Generic Reload Fuel Application, NEDE-24011-P-A, August 1978.

(9) Final Safety Analysis Report, Nine Mile Point Nuclear Station, Niagara Mohawk Power Corporation, June 1967.

(10)

NRC Safety Evaluation, Amendment No.

24 to DPR-63 contained in letter from G. Lear, NRC, to D.

P. Disc dated May 15, 1978.

(11) "Core Flow Distribution in a GE Boiling Hater Reactor as Measured in Quad Cities Unit 1," NED0-10722A.

(12) Nine Mile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal

-(Cycle 6), NED0-24185, April 1979.

(13) Loss-of-Coolant Accident Analysis Report for Nine Mile Point Unit 1 Nuclear Power Station, NED0-24348, Aug.

1981.

(14)

GE Boiling Water Reactor-Extended Load Line Limit Analysis for Nine Mile Point Unit 1 Cycle 9, NEDC-31126, February 1986.

(15) Nine Mile Point Unit 1, Loss-of-Coolant Accident Analysis, NEDC-31446P, June 1987.

(16) Supplement 1 to Nine Mile Point Generating Station Unit 1

SAFER/CORECOOL/GESTR-LOCA Analysis Report NEDC-31446P-l, Class III, September 1987.

70d

ATTACHMENT B NIAGARA MOHANK PONER CORPORATION LICENSE DPR-63 DOCKET NO. 50-220 Su ortin Information and No Si nificant Hazards Consideration Anal sis The proposed revision to Specification 2.1.1 includes changes to the formula contained on Figure 2.1.1 for adjusting the'flow biased scram and APRM rod block setpoints in those cases where the calculated total peaking factor exceeds the Maximum Total Peaking Factor for the fuel type.

(Specifically the maximum Total Peaking Factor of 2.90 for the General Electric fuel bundle type BD 321B.)

This is the new Maximum Total Peaking Factor for the GE 8x8EB fuel to be added to the core during the 1988 Refueling and Maintenance Outage.

The Maximum Total Peaking Factor for the General Electric fuel bundle type BD 321B (GE 8xBEB fuel) was calculated by GE to be 2.90 (see reference 4).

In

addition, the Maximum Total Peaking Factor for each type of GE fuel currently in the Nine Mile Point Unit 1 core has been identified.

Finally, the factors in the formula (equation) contained on Figure 2.1.1 have been clarified.

The proposed changes to Specification 3.1.7 and the addition of Figure 3.1.7g reflects the addition of Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for the General Electric fuel bundle type BD321B.

These limits were calculated using the SAFER/CORECOOL/GESTR-LOCA computer codes and methodology (Reference 1,

copy attached).

In addition, exposure dependent Minimum Critical Power Ratio (MCPR) limits have been eliminated and replaced with one MCPR which is applicable for the entire cycle; an increase in the

'limit is not required at end of cycle.

The MCPR limits were calculated and documented in the Reload 11, Supplemental Reload Licensing Submittal (Reference 2,

copy attached).

Finally, the Bases for Specifications 2.1.1, Fuel Claddin Inte rit and 3.1.7 and 4.1.7, Fuel

Rods, have been updated to reflect the addition of General Electric BD321B fuel during the 1988 Refueling and Maintenance Outage.

The description of the compensatory actions required when the calculated total peaking factor exceeds the maximum total peaking has been revised for clarity.

10 CFR 50.91 requires that at the time a licensee requests an amendment, it must provide to the Commission its analysis, using the standards in 10 CFR 50.92, about the issue of no significant hazards consideration.

Therefore, in accordance with 10 CFR 50.91 and 10 CFR 50.92, the following analysis has been performed.

The o eration of Nine Mile Point Unit 1

.in accordance with the ro osed amendment wi 1 1 not involve a si nificant increase in the robabi lit or conse uence of an accident reviousl evaluated.

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ATTACHMENT B (Continu The methods used to analyze the Loss of Coolant Accident response of the BD321B fuel conform to 10 CFR 50 Appendix K requirements and follow the methodology previously approved by the Nuclear-Regulatory Commission (Reference 3).

MAPLHGR limits for BD321B fuel are illustrated in Figure 3.1.7g.

The peak cladding temperature and maximum oxidation fraction are less than the maximum limits specified in 10 CFR 50.46.

In addition, the limiting transients have been analyzed in Reference 2 (copy attached).

The results of this analysis indicate that if a minimum critical power ratio of 1.37 is maintained throughout the cycle, it wi 11 assure that the minimum critical power ratio will not go below 1.07 during the most limiting transient.

The Technical Specification minimum critical power ratio of 1.40 throughout the entire fuel cycle is above the minimum required critical power ratio of 1.37.

Based on the analyses in References 1

and 2, the proposed amendment will not result in a significant increase in the probability or consequences of an accident previously evaluated.

The o eration of Nine Mile Point Unit 1

in accordance with the ro osed amendment wi 1 1 not create the ossi bi lit of a new or different kind of accident from an accident reviousl evaluated.

The proposed amendment provides the Technical Specification limits for the BD321B fuel.

The new reload fuel has essentially the same mechanical properties as the fuel presently in use.

Therefore, the proposed amendment will not create the possibility of a new or different accident from any previously evaluated.

The o eration of Nine Mile Point Unit 1

in accordance with the ro osed amendment will not involve a si nificant reduction in a mar in of safet Analyses of the Loss of Coolant Accident and limiting transient response of proposed fuel bundle type BD321B have been completed as described in References 1

and 2 (copies attached).

These analyses demonstrate that there is no significant reduction in a margin of safety.

As determined by the analysis

above, the proposed amendment involves no significant hazards consideration.

References 1.

Supplement 1 to Nine Mile Point Nuclear Generating Station Unit 1

SAFER/

CORECOOL/GESTR-LOCA Analysis Report, NEDC-31446P-1, Class III, September 1987.

2.

Supplemental Reload Licensing Submittal for Nine Mile Point Nuclear Power Station Unit 1, Reload ll, Report 23A5862, Revision 0, Class I, October 1987.

3.

Letter, A. C.

Thadani (NRC) to H. C. Pfefferlen (GE),

"Acceptance for Referencing of Licensing Topical Report NEDE-30996-P, Volume II, 'SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet and Non-Jet Pump Plants,'"

May 1987.

4.

Letter from H.

H. Hetzel (General Electric) to D. K. Greene, "GE 8X8EB Maximum Total Peaking Factor",

dated December 15, 1987.

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