ML19305D272

Proposed Revision to Tech Specs 4.2.4.2.5 & 5.2.17.5 Applying Exposure Limitation Only to Fuel Assemblies in Core Interior & Deleting Exposure Limit on Peripheral Assemblies. Supporting Documentation Encl

ML19305D272
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Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	04/01/1980
From:	DAIRYLAND POWER COOPERATIVE
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ML19305D270	List: ML19305D267 ML19305D272
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Download: ML19305D272 (11)

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ATTACHMENT 1 TO DPC LETTER, LAC-6846, DATED APRIL 1, ~1980 l

Proposed Technical Specification wording and the revised bases for LACBWR Technical Specifications 4.2.4.2.5 and 5.2.17.5.

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800.4140 34 V f

- 32cn -

POWER DISTRIBUTION LIMITS MAXIMUM AVERAGE FUEL ASSEMBLY EXPOSURE LIMITING CONDITION FOR OPERATION 4.2.4.2.5 The maximum average exposure of any fuel assembly not on the periphery of the core shall be limited to 15,600 MWD /MTU.

APPLICABILITY:

OPERATIONAL CONDITION 1.

ACTION:

With the maximum average fuel assembly exposure of any non-peripheral assembly greater than 15,600 MWD /MTU, be in at least HOT SHUTDOWN with the main steam line isolation valve closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENT 5.2.17.5 The maximum average exposure of each fuel assembly not on the periphery of the core shall be determined to be less than 15,600 MWD /MTU by calculation at least once per 31 EFPD.

Amendment No.

- 32gg -

POWER DISTRIBUTION LIMITS BAS S FOR SECTIONS 4.2.4.2 and 5.2.17 LINEAR HEAT GENERATION RATE - (Continued)

For Type I and Type II (A-C) fuel, the original design LINEAR HEAT GENERATION RATE specified by the fuel manuf acturer was conserva-cively reduced to 11.94 kw/ft to account for the effects of dens-ification, power spikes and manufacturing factors.

For Type III (ENC) fuel, the design LINEAR HEAT GENERATION RATE of 11.52 kw/ft i

is also calculated with design conservatisms that are larger than the calculated axial densification effects plus manufacturing tolerances and power spike effects, Reference 6.

The daily requirement for surveillance of the core LHGR above 25%

of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The surveillance of core LHGR after power increases > 15% of RATED THERMAL POWER will assure that significant increases in LHGR are determined.

4.2.4.2.5 and 5.2.17.5 Maximum Average Fuel Assembly Exposure Fuel cladding integrity is a function of many parameters including fuel exposure, pellet clad interaction, THERMAL POWER, rate of change in power density, coolant chemistry, etc.

Therefore, limiting fuel exposure to 15,600 MWD /MTU in the non-peripheral fuel assemblies which experience higher than average power densities and rates of change of power will give additional assurance that the condition of the fuel during operation will be satisfactory.

It is not necessary to limit exposure in the peripheral core locations since operating experience at LACBWR has shown that the 28 peripheral fuel assemblies have a much lower rate of failure than the 44 interior fuel assemblies.

This trend has been attributed to the lower power density at these locations, and the minimal effects of control rod movements which cause local power peaking in the fuel rods near the tips of the control rods.

The outer control rods are fully withdrawn at the beginning of cycle (BOC) and remain withdrawn during normal cycle operations.

Minor clad defects that may occur in the peripheral core positions would be expected to develop very slowly, and the consequences of such failures would be minimal.

During previous operation with A-C fuel, a number of fuel assemblies have exceeded 15,600 MWD /MTU without any indication of failure and at the end of Cycle 3, @OC-3),

four assemblies had exceeded 18,000 MWD /MTU without failure.

The average exposure of the 25 assemblies discharged at EOC-3 was 15,530 MWD /MTU and the peak exposure was 21,532 MWD /MTU.

The average exposure of the 32 assemblies discharged at EOC-4 was 16,459 MWD /MTU.

It is expected that the new improved EXXON fuel will give even better service.

Amendment No.

- 32hh -

POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4. 2. 4. 2 and 5. 2.17 Maximum Average Fuel Assembly Exposure - (Continued)

Pellet-clad interaction is a well known and documented contrib-uting factor to fuel rod failures.

The presence of pellet cladding interaction has been identified in post-irradiation examinations of fuel rods removed from LACBWR fuel assemblies.

Fuel rods removed from fuel assemblies with average exposure up to 14,700 MWD /MTU have been examined.

The strength, ductility, and condition of the cladding in these rods was found to be adequate as determined by mechanical tests.

The examination further confirmed that power history of the rods is of prime importance, though not the only factor in contributing to fuel rod failure.

A limit of 15,600 MWD /MTU fuel element average exposure is con-sistent with the results obtained from examinations conducted on fuel assemblies with sLmilar exposure history.

During future operation the rate of withdrawal of control rods when the THERMAL POWER is above 25% of RATED THERMAL POWER will be reduced from that experienced during operation prior to Cycle-5 which will also significantly reduce the stresses in the fuel clad.

Additional surveillance and limitations on coolant and off-gas activity will assure that operation does not continue with grossly failed fuel.

i

References:

1.

" Technical Evaluation Adequacy of La Crosse Boiling Water Reactor Emergency Core Cooling System", Report SS-942, Gulf United Nuclear Corporation, May 31, 1972.

2.

" Review of Densification Effects in La Crosse Boiling Water Reactor", Report SS-1085, Gulf United Nuclear Corporation, May 15, 1973.

3.

NRC Safety Evaluation Report, Letter, Reid to Madgett, dated August 12, 1976.

4.

"ECCS Analysis for Type II and Type III Fuels for the La Crosse Boiling Water Reactor", Exxon Nuclear Company, Inc., XN-NF-77-7, March 1977.

5.

" Transient Analysis for LACBWR Reload Fuel", Response to Question 4, Nuclear Energy Services, Inc., Report 81A0025, February 18, 1977.

Amendment No.

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4

- 3211 -

POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 References - (Continued) 6.

" Description of Exxon Type III Nuclear Fuel for Batch 1 Reload in the LACBWR", Dairyland Power Cooperative, LAC-3929, May 17, 1976.

7.

Exxon Nuclear Co. Letter, J. A. White to C. W. Angle,

Subject:

MAPLHGR Limits for Type I (Allis-Chalmers) Fuel, dated June 22, 1977.

8.

DPC Letter, LAC-6846, Linder to Ziemann, dated April 1, 1980.

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(Next page is page 33)

Amendment No.

ATTACHMENT 2 TO DPC LETTER, LAC-6846, DATED APRIL 1, 1980 Analysis and Results of a LACBWR Turbine Trip Without Bypass Transient Starting Conditions and Assumptions Events which result in a turbine trip include the turbine over-speeding to 110% of rated speed, a lubricating oil malfunction, low condenser vacuum, opening of the generator output breaker by a complete generator load rejection, or manual initiation.

The starting conditions and assumptions for this transient have been chosen such that a conservative estimate of the transient results.

The assumptions are as follows:

(1)

The reactor is initially operating at 102% of rated power.

(2)

Recirculation flow is at 100% of its rated value.

(3)

The turbine trip results in an instantaneous and total loss of steam flow to the turbine.

(4)

No credit is taken for the reactor partial scram (13 control rods) or the recirculation pump run back to 80% of full flow caused by the turbine stop valve clovtre.

(5)

No credit is taken for a reactor scram on high Power to Plow ratio at 115% of rated power.

(6)

The reactor is operating at the end of the fuel cycle.

(7)

Operation of the shutdown condenser is initiated when reactor pressure is greater than 1325 psig.

(8)

The reactor is assumed to scram on high power (120% of rated power) and a run back of the recirculation pumps to 80% of full flow is initiated at the same time.

It is conservatively assumed that negative reactivity in-sertion by control rods is delayed-an additional 1.5 seconds (See Scram Reactivity Curve in Attachment 3) and that actual reduction of recirculation flow is delayed for 0.5 seconds after initiation of the scram signal.

l

ATTACHMENT 2 - (Continued)

(9)

No credit is taken for Turbine Bypass steam flow.

Results and Consequences Figures 1, 2, 3, and 4 present curves of power, heatflux, recirc-i ulation flow and pressure versus time respectively.

Steam flow is instantaneously stopped at time zero due to the turbine trip which closes the turbine stop valve.

The cut-off of steam flow results in a rapid increase in reactor pressure and a reduction in core voids.

The reduction in moderator voids adds reactivity to the core causing a rapid increase in power.

At 0.0012 seconds, power exceeds 120% of rated power and a reactor scram is initiated.

The rapid increase in power is limited by doppler feedback and the peak reactor power for the transient, 260.6%, occurs at 0.043 seconds.

At 0.051 seconds, core recirculation flow begins to decrease which reduces core reactivity.

The decreasing recircul-ation flow and doppler feedback reduce reactor power to approxi-mately 113% at 1.5 seconds.

At 1.501 seconds, negative reactivity from control rod insertion becomes effective, and, at 3~.001 seconds, the control rods are completely inserted.

At approximately 3.3 seconds, operation of the shutdown condenser is initiated by high reactor pressure (1325 psig).

The peak reactor pressure during the transient, 1349 psig, occurs at 8.5 seconds.

For comparision purposes, the ACPR's that would result from this transient were determined for the same core conditions (which bound Cycle-6 conditions) that were used for the basis of the ACPR's reported in DPC Letter, LAC-4654, Madgett to Reid, dated April 27, 1977.

Table 1 of LAC-4654 is reproduced on the following page with the expected CPR's from this transient added.

As can be seen from the table, the ACPR's resulting from this transient are not as limiting as those resulting from a Control Rod Withdrawal transient, or an Increase in Feed Water Flow transient.

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P TABLE 1 INITIAL AND TRANSIENT CPR'S F

A-C Ass'y 6,6 XN Ass'y 6,4 Initial Transient Initial Transient CPR MCPR CPR MCPR Turbine Trip w/ Bypass ***

1.59 1.45 1.73 1.57 Turbine Trip w/o Bypass 1.59 1.49 1.73 1.61 MSIV Closure 1.59 1.51 1.73 1.64 Increase in FW Flow 1.59 1.39 1.73 1.51 Accidental Opening of 1.59

>1.59 1.73

>1.73 Turbine Bypass Valve Loss of Recirculation Flow 1.59 1.58 1.73

>1.73 Isolated Loop Startup 3.10*

1.53 3.25*

1.66 Rod Withdrawal 1.59 1.38 1.94**

1.67

• This transient starts from 52% Reactor Power with One Recirculation Loop Operational.

• • No Xenon.

• • • This transient assumed that the reactor did not scram.

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o ATTACHMENT 3 TO DPC LETTER, LAC-6846, DATED APRIL 1, 1980 Scram Reactivity Curve The scram reactivity curve used in the transient analyses presented in this letter and reproduced here is the scram reactivity curve used in the transient analyses for LACBWR reload fuel and shown in Figure 3.4-2 of NES 81A0025, LAC-4523, February 25, 1977.

In this figure, the scram reactivity used in the analysis is compared with that cal-culated for Cycle 5.*

The curve used in the analysis is seen to be very conservative compared with calculated values.

Basically, the conservatism consists of reducing the total worth of the rods by 20% and neglecting the contribution of the strongest rod ($3.92).

Since the worth of the control rods in Cycle 6 is well within 20% of the worth in Cycle 5, the scram worth during Cycle 6 operation is bounded by the scram reactivity used in the analyses submitted in February 1977 for Cycle 5.

• For the calculated scram reactivity curve for EOC-5, a delay of 0.5 sec. between the initiation of the scram signal and the start of control rod insertion was conservatively assumed.

Amendment No. 26 to the LACBWR Safeguards Report (SAR) states that "...The time required to initiate control rod motion due to a high flux condition is 0.068 sec."

During testing of control rod scram times the maximum total scram time, from initiation signal to full in, is routinely less than 2.3 seconds and the total scram time for the majority of the rods is % 2.0 seconds.

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