ML19269E791

Amend 16 to License DPR-45 Adding Core Power Distribution Limits for Allis-Chalmers Type I Fuel Assemblies & Changing Limits for Offgas Emission Rates to Account for Cleanup of Tramp Activity

ML19269E791
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	05/25/1979
From:	Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML19269E789	List: ML19269E788 ML19269E791 ML19269E794 ML19269E801
References
NUDOCS 7906290639
Download: ML19269E791 (18)

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E WASHINGTON, D. C. 20555

%...../ DAIRYLAND POWER COOPERATIVE DOCKET N0. 50-409 LA CROSSE BOILING WATER REACTOR AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.16 License No. DPR-45 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Dairyland Power Cooperative (the licensee) dated February 26, 1979, as supplemented by letters dated May 9 and 23,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (1) 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 comon 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 Comission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Provisional Operating License No. DPR-45 is hereby amended to read as follows: 2143 220 7906290637,

s (2) Technical Specifications The Technical Specifications contained in Appendix A issued October 21, 1969, with Authorization No. DPRA-6, as revised through Amendment No.16, are hereby incorporated in the license. The 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 Dennis L. Zieman Chief Operating Reactors Branch #2 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: May 25, 1979 2143 221

ATTACHMENT TO LICENSE AMENDHENT N0.16 LA CROSSE BOILING WATER REACTOR (LACBWR) PROVISIONAL OPERATING LICENSE NO. DPR-45 Revise Appendix A by replacing the following pages with the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the areas of change. Pages 27z 30a 32a 32b 32f 32v 32y 32z 32bb 32cc 32dd 32ee 32ff 3299 32hh 2143 222

- 27z - SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS REACTORCOOLANTSYSTEMPRESSUR( 4.0.2.1.3 The reactor coolant system pressure at the top of the reactor vessel shall not exceed 1540_psig. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. ACTION: With the reactor coolant system pressure at the top of reactor vessel above 1540 psig, be in at least HOT SHUTD0'.:N with the reactor coolant system pressure 1 1540 psig within 2 hours. REACTOR VESSEL _yATER LEVEL 4.0.2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5 , ACTION: With the reactor water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the reactor vessel water level, after depressurizina the reactor vessel, if required. 300R ]G NAL 2143 223 Amendment No. JJ,16

- 30a - 4.2.2.4.d The reactor vessel ternperature, when the core is critical (except for low power physics tests or when the reactor vessel is vented), shall be no less than the minimum temperature above RTNDT shown by curve 63 on Figure 5. 4.2.2.4.e The Forced Circulation loops shall not be pressur-ized ur.less their temperature is above 70 F and shall not be 0 pressurized above 280 psig unic.;s their temperature is at least 1300F. 'Ihe mexi=um pressure scetir.;s fer the reactor vessel and coolant ' 4.2.2.5 systen relief vc1vcs shall not L:<: erd the linitati.nc of'the MF.E Loiler and Pressure Vessel Cede, Sc: tion VIII, cnd nucicar code cases applicabic as of June 1962. 4.2.2.6 At least one forced circulation loop shall be non isolated, i.e. suction and discharge valves open, anytime the shutdown condenser is operating, or the Reactor Building Main Steam Isolation Valve (MSIV) and the Turbine Building MSIV are non-isolated. 4.2.2.7 The suction, discharae, and discharge bypass valves of the forced circulation pumps shall operate as described in Sec. 2.3.3.4. l l 2143 224 Amendment No. S, JJ,16

- 3Za - RE ACTOR COOLANT SYSTEM _ 4.2.2.22 ACTIVITY LIMITING CONDITION FOR OPERATION I 4.2.2.22 The activity of the: Reactor coolant shall be limited to: a. 0.2 pCi/ gram 00SE EQUIVALENT l-131, 1. 1 100/T pCi/ gram, and 2. 1 3'. 1 5.0 x 10 Ci/ gram gross alpha activity. -6 Of f-gas emission, measured at,the 150 cu. f t. of f-gas holdup 750 Ci/d (P/P ) b. tank effluent monitor, shall be limited to 1 F T (1-0.018t) (P/PF), where: +A (P/P ) = f raction of RATED THERMAL P0h'ER between 33 1. r and 165 MWt, = tramp activity, Ci/d, not to exceed 1500, at THERMAL POWER (P ) es determined upon initial power escalation A 2. 7 i f after each refueling. t = days after the determination of AT with the limitation l that t <_ 50. ty values shall be nomalized for correlation with Acti 4. previous cycles de to changes in monitoring, samplinn and/or anclysis methods. OPERATIONAL CONDITIONS 1, 2, 3 and 4. APPLICABILITY: ACTI0t{: In OPERATIONAL CONDITION 1, 2 or 3, with the activity of the i a. i-reactor coolant-4.0 uCi/ gram, >0.2 uCi/ gram DOSE EQUIVALENT I-131 but 1 to OPERATIONAL CONDITION 1, 2, 3 or 4 is penr.itted 1. entry and operation may continue for up to 48 hours provided that operation under these conditions shall not exceed Should [ 800 hours in any consecutive 12-month period. the total operating time at a primary coolant specific activity >0.2 uCi/ pram DOSE EQUIVALEtlT I-131 exceed 500 hours in any consecutive six-month period, the licensee shall repori the number of hours of operation above this limit to ;he NRC within 30 days. 2143 225 0.2 uCi/ gram DOSE EQUIVALENT I-131 for more than 48 2. hours during one continuous time inter al or > 4.0 uCi/gra-F n. [ be in at least HOT SHUT 00WN with the main steam line [ isolation valve closed within 12 hours. ' Amendment No. 71,16

- 32b - ~ REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued): > 100/E pCiTg m or > 5.0 x 10-6 Ci/ gram gross alpha 3. activity, be in at least HOT SHUTDOWN with the main steam line isolation valve closed within 12 hours and in COLD ' SHUTDOWN within the next 24 hours. In OPERATIONAL CONDITION 1, with off-gas activity: b. 1000 Ci/ day >750 Ci/ day ((P/Pp)18t)T((1-0.018t) (P/PF) but iP/PF). POW +A 1. (P/PF) + AT 1-0.0 for up to 48 hours provided that operation under t' lese conditions shall not exceed 800 hours in any consecutive 12-month period. Should the total operating time at an l off-gas activity > 750 Ci/ day P/Pp + AT (1-0.018t) (P/PF) exceed 500 hours in any consecutive six-month period, the licensee shall report the number of hours of operation above this limit to the NRC within 30 days. T (1-0.018t) (P/PF) for more than l f > 750 Ci/ day (P/P ) + A F 2. 48 hours during one continuous time interval or > 1000 C1/dsy be in at least HOT SHUTDOWN with the main steam line isolation valve closed within 12 hours and in COLD SHUTDOWN within the next 24 hours. ' ~ ~ = = ~ With the reactor required to be shutdown as a result of ACTION a.3 or b.2 above, obtain Comission approval prior to increasing c. l reactor coolant system temperature above 212 F. In OPERATIONAL CONDITION 1, with lHERMAL POWER increased by d. more than 25% of RATED THERMAL POWER since the last performance of Table 4.2.2.22-1, Item 4, perform Table 4.2.2.22-1, Items ic and 4c. In OPERATIONAL CONDITION 1 or 2, with off-gas activity increased by more than 1000 pCi/sec within one hour, perform Table e. Off-gas activity values shall be 4.2.2.22-1, Item 1c and 4c. normalized for correlation with previous cycles due to changes in monitoring, sampling and/or analysis methods. In OPERATIONAL CONDITION 1, 2, 3 or 4 with the activity of the -2 f. primary coolant > 0.2 pCi/ gram DOSE EQUIVALENT I-131, or > 100/E pCi/ gram, or > 5.0 x 10 " pCi/ gram gross,ajpha activity, or with the off-gas activity > 750 Ci/ day (P/P )+A (1-0.018t) (P/Pp, l F T perform the sampling and anaTysis requirements of Items Ib, 4b and 6b of Table 4.2.2.22-1. 2143 226 A REPORTABLE OCCUT:RENCE report shall be prepared and submitted to the Commission pursuant to Specification 3.9.2. Inis report shall contain the results of the activity analyses and the time duration when the activity exceeded each limit together with the b iti i1forpation. v 81 1 Amendment No. JJ,16

4 s - 3.2f - REACTOR C00LM;T SYSTEM 4.2.2.22 and 5.2.16 Activity The limitations on the specific activity of the primary coolant ensure that the 2 hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operction will not exceed small fractions of the dose guidelines of 10 CFR 100. Permitting operation to continue for limited time periods with higher specific activity leveis accomodates short term iodine spikes which may be cssociated with power level changes, and is based on the fact that a steam line failure during these short time periods is considerably less likely. Operation at the higher activity levels, therefore, is restricted to a small fraction o'f the unit's total operating tin:2. The upper limit of coolant iodine concentration during short term iodine spikes ensures that the thyroid dose from a steam line failure will not exceed 30 CFR Part 100 dose guidelines. Information obtained cn iodine spiking will be used to assess the A reduction in frequency parameters associated t.ith spiking phenonenc. of isotopic analysis following pnv.er changes may be permissible if justified by the data obtained. 3 Closing the main steam line isolation valves prevents thc release of + The activity to the environs should the steam line rupture occur. surveillance requiremants provide adequate assu:ance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. The limitation on off-gas emission and the limitation on gross alpha activity are established to ensure fuel integrity. The numerical limits were selected using extensive data collected during fuel I The off-gas activity limitation formula has been . cycle 4 and 5. derived to relate off-gas production and power level and conservatively precludes reduced power operation without also reducing the limiting condition for operation off-gas c.~ivity. The fonnula contains a 1 factor AT (1-0.018t) P/PF which accounts for base level uranium tramp e -' activity in the coolant and on the core surfaces at the beginning of the cycle. This tramp activity term is reduced to 0.1 of the BOC valve within 50 days to account for cleanup of the system during s operation. Factor) is limited The attitity o'flhe off-gas emission (including the At to ensure that the potential radiological consequences of an off-gas system failure resulting in the release of the entire off-gas system

volume will not exceed small fractions of the dose avidelines of 10 CFR 100.

Operation at higher levels of off-gas enissicn for s'. ort periods is r This restricted to a smell fraction of the unit's Lotal opereting time. upper lir,it permits normal transients as raignt be expected during periods of power change. aus 227 ~ P00R OR8 W Amendment No. 17,16

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-32z-POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 4.2.4.2.4 The LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the design LHGR of 11.94 kw/ft for Type I or Type II (A-C) fuel rods and 11.52 kw/ft for Type III (ENC) fuel rods during steady-state operation. APPLICABILITY: OPERATIONAL COFIITION 1.* ACTION: With the LHGR of any fuel rod exceeding the limit, initiate adjustment within 30 minutes so that the LHGR is below the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMElfrS 5.2.17.4 LHGR's shall be determined to be equal to or less than the limit by verifying that each control rod is within the control rod pattern and withdrawal sequence requirements during operation at > 25% RATED THERMAL POWER: At least once per 24 hours during steady state a. operation, and b. Each time power level of the reactor has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established. 25% of RATED THERMAL POWER POLR(%Igljv],u{ Amendment No. JJ, " 2143 230

-32bb-4.2.4.2 POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 4.2.4.2.1 and 5.2.17.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The specifications of these sections assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident (LOCA) will not exceed 2300 F in compliance with the limits established by the Interim Acceptance Criteria, June 1971 as applied to LACBWR stainless steel clad fuel, Reference 1, 2, 3, 4 and 7. The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat gener-ation rate of all the rods of a fuel assembly at any axial loca-tion and is dependent only secondarily on the rod to rod power distribution within an assembly. The PCT is conservatively calcul-ated assuming the reactor is operating at 102% full rated power and the axial times radial peaking factor is at a maximum value of 2.43 from 0-24 GWD/MTU maximum average planar exposure for any Type I, Type II or Type III fuel assembly. The factor is then reduced to 1.75 at 30 GWD/MTU exposure to prevent calculated failure of internal rods. Operation with peaking factors below these values at RATED THERMAL POWER will ensure that peak cladding temperature during a LOCA will not exceed the limit of 23000F for stainless steel fuel. The corresponding limiting Maximum AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) va?.ue for Type III is shown in Figure 4.2.4.2.1-1 and that for Tyra I and Type II fuel in Figure 4.2.4.2.1-2. The calculational procedures used to establish the MAPLHGR limits shown in these two figures are based on the loss-of-coolant acci-dent analyses performed by Gulf United Nuclear Corporation and Exxon Nuclear Company using approved calculational models consis-tent with the requirements of ECCS criteria as it applies to the LACBWR fuel. The effects of the following parameters were in-cluded: (1) Radial conduction within each fuel rod, (2) Rod and canister convection, (3) Rod thermal radiation among the rods and canister, (4) Gap conductance with exposure and densification, (5) Rod ballooning, and other mechanical and neutronic parameters. The peak cladding temperatures achieved during a postulated LOCA Type II and Type III fuel are shown in Bases Figure for Type I, 4.2.4.2.1-1. Operation at RATED THERMAL POWER with fuel assembly AVERAGE PLANAR HEAT GENERATION DATES below the MAPLHGR limits of Figures 4. 2. 4. 2.1-1 and 4. 2. 4. 2.1-2 will ensure that the PCT's will not exceed the 2300 F limit. 0 A list of the significant input parameters to tne loss-of-coolant accident analysis is presented in Reference 4. 2143 231 i y

i D DD 2400 ~ O n NC3 o' 2300 N ~ D TYPE 'P, j -s y ' ~ ~ \\ 2200 N D-e g 2100 \\ d ~ E \\ ~ 5 \\ E' \\ F 2000 ~ Type fil \\ F (ENC) h q l 1900 G a 3 1800 n. 1700 N N 5 10 15 20 25 30 { E, AVERAGE PLANAR EXPOSURE (GWD/MTU) u = PEAK CLADDING TEMPERATURE VS. AVERAGE PLANAR EXPOSURE N y* AT MAPLMGR LIMITS OF FIGURES 4.2.4.2.1.-1 AND 4.2.4.2.1-2. u N Figure 84.2.4.2.1-1

t -32dd-POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 The daily requirement for surveillance of the core APLHGR 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 APLHGR are determined. 4.1.4.1.1 and 5.2.17.2 THERMAL POWER-RECIRCULATION FLOW RELATIONSHIP The THERMAL POWER-recirculation flow limiting condition for steady-state operation has been conservatively set below the LSSS curve presented in Figure 4.0.2.2.1-1 of Specification 4.0.2.2.1. This specification also requires that the reactor not be operated in a steady-state condition above the RATED THERMAL POWER, 165 MWth, authorized in the NRC license for the facility. The limiting minimum flow setpoint of 30% of rated recirculation flow is sub-stantially above the natural circulation flow and the flow at which hydraulic instability occurs. A ratio of > 1.6 exists be-tween the low-flow scram setpoints and the instability-natural circulation flow. Therefore, adequate protection of the core against flow and core instability exists over the full power range of anticipated reactor operation as limited by the THERMAL POWER-recirculation flow operating region. The low flow limitation requires the establishment of a minimum flow cf 30% of rated recirculation flow before reactor startup. Operation in a steady state condition < 6 8. 3 % o f RATED THERMAL POWER at the minimum 30% of RATED RECIRCULATION FLOW assures that the CPR remains above the minimum allowable value of 1.32 during an abnormal reactor transient (recirculation flow speedup is the most limiting). The steady state limiting CPR values corresponding to other operating conditions bound by the Power-Flow LSSS curve reported in Fig. 4.0.2.2.1-1 are defined in Fig. 4.2.4.2.3-1. The daily requirement for surveillance of the power to recirculation flow relation is sufficient since this relation shifts very slowly when there have not been signi'icant power, flow or control rod changes. The surveillance of this relation after power increases 15% of RATED THERMAL POWER will assure that significant changes In the relation are determined. 4.2.4.2.3 and 5.2.17.3 MINIMUM CRITICAL POWER RATIO The required operating limit MINIMUM CRITICAL POWER RATIO (MCPR) at steady-state operating conditions as established in Specification 4.2.4.2.3 is derived from an analysis of abnormal operational trans-ients with the transient CRITICAL POWER RATIO > 1.32. The CPR 2143 233 klendment No. JJ,16

-32ee-POWER DISTRIBUTION LIMITS 9 BASES FOR SECTIONS 4.2.4.2 and 5.2.17 MINIMUM CRITICAL POWER RATIO - (Continued) criterion of 1.32 was established, Reference 5, based on the XN-2 predicted power to the measured critical power, assuring better than 99% confidence, a 99% probability of avoiding boiling trans-ition. For any abnormal operating transient analysis, evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the MCPR limit of 1.32 at any time during the transient assuming Limiting Safety System Settings given in Specification 4.0.2.2. To assure that MCPR limit of 1.32 is not exceeded duriag any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO. The type of transients evaluated with pressure increase, moderator temperature decrease, coolant inventory decrease, core coolant flow increase and decrease, and positive reactivity insertion. The limiting transient which determines the required steady-state limit during operation at RATED THERMAL POWER is the rod with-drawal transient. A minimum transient CPR of 1.32 is caused by steady state CPR's of 1.73, 1.49 and 1.50 for Type I, II, and III fuel respectively. A steady state MCPR of 1.73 has been accordingly established for Type I fuel at RATED THERMAL POWER and RATED RECIRCULATION FLOW conditions to ensure no penetration of the mir.. i - allowed CPR value of 1.32. The corresponding steady state MCPR ter Types II and III have been maintained at the Cycle 5 value of 1.59 for conservatism and continuity, as Cycle 6 maximum penking factors are lower than these for Cycle 5. For core flow less than rated recirculation flow, the most 4.miting transient is the recirculation two-pump speedup caused by a failure of the motor-speed control system. The MCPR limit for steady state operation at flows less than rated recirculation flow are shown in Figure 4.2.4.2.3-1. MCPR values were calculated using a flow control line that corres-ponds to the Limiting Power-Flow line of Figure 4.0.2.2.1-1 which intersects 116.7% power at a maximum of 110% rated recirculation flow. The recirculation pumps are operated in the manual mode only, and the maximum core flow is limited by pump scoop tube travel and pump capacity. 2143 234 P 0 m gj rJ Q

]u Amendment No. 77,16

-32ff-POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 MINIMUM CRITICAL POWER RATIO - (Continued) The limiting bundle relative power was adjusted until the MCPR was equal to 1.32 at maximum flow. For additional conservatism, the transient was assumed to terminate at 120% of RATED THERMAL POWER rather than expected 116.7% power intercept point. Using this relative bundle power as a basis, the MCPR's were calculated at different flows. The calcu)ated MCPR's for Type I, Type II and Type III fuel were then used to establish the curves shown in Figure 4.2.4.2.3-1. These curves represent the minimum allowable operating MCPR for the most limitinJ fuel assemblies over the full core flow range of permissible operation. The nominal expected flow control line falls below the Limiting Power-Flow line and intersects the 100% of RATED THERMAL POWER-100% of rated recirculation flow intercept point. The terminal MCPR during a postulated transient from a normal low flow starting condition therefore would result in a MCPR > 1.32. In addition, an automatic reacter trip would be expected to occur due to power-but flow and/or 120% overpower trip signals during the transient, these were conservatively ignored for the analyses. The MCPR limit curves shown in Figure 4.2.4.2.3-1 are conservative and operation with greater MCPR values will assure that the MCPR limits will not be penetrated for the most severe operational transient. At THERMAL POWER less than or equal to 15% of the RATED THERMAL POWER, the moderator void content will be small even at minimum recirculation flow. For all designated control rod patterns under excess of require-these conditions, the resulting MCPR value is _a ments by a considerable margin. With the J-s void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for surveillance of the core MCPR is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The surveillance of core MCPR after power increases > 15% of RATED THERMAL POWER will assure that significant reductions in MCPR are determined. 4.2.4.2.4 and 5.2.17.4 LINEAR HEAT GENERATION RATE The LINEAR HEAT GENERATION RATE (LHGR) specifications for Type I, Type II and Type III fuel assure that during steady-state operation, the peak LINEAR HEAT GENERATION RATE in any fuel red is less than the design LINEAR HEAT GENERATION RATE. The specifications also assure sufficient margin to accommodate maximum centerline fuel temperatures less than the melting point during operational 2143 235 transients.P00RBRGgi Miendment No. JJ,16

-32gg-POWER DISTRIBUTION LIMITS BASES 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 manufacturer was conserva-tively reduced to 11.94 kw/ft to account for the effects of densifi-cation, power spikes and manufacturing factors. For Type III (ENC) fuel, the design LINEAR HEAT GENERATION RATE of 11.52 kw/ft is also calculated with design conservatisms that are larger than the calculated axial densification effects plus manufacturing toler-ances 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. 4.2.3.2.5 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 will give additional assurance that the condition of any fuel assembly during operation will be satisfactory. The opera'ing history of the LACBWR indicates that a limit of 15,000 MWD /MTU for the maximum average exposure of any fuel assembly is very conservative. During previous operation, a number of fuel assemblies have exceeded 15,000 MWD /MTU without any indication of failure and at the end of Cycle 3, EOC-3, four assemblies had ex-ceeded 18,000 MWD /MTU v'thout 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. Gross fuel failure involving observ-eble loss of fuel or clad has only occurred in three fuel assemblies, all during fuel Cycle 4, and all three of these assemblies were over 17,400 MND/MTU exposure. The average exposure of the 32 assemblies discharged at EOC-4 was 16,459 MWD /MTU. Pellet-clad interaction is a well known and documented contributing factor to fuel rod failures. The presence of pellet cladding interaction has also been identified in post-irradiation examinations of #uel rodL removed from LACBWR fuel elements. Fuel rods removed fr(n fuel elements with average exposure up to 14,700 MWD /MTU have been exanined. 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 im-portance, though not the only factor in contributing to fuel rod 2143 236 Amendment No. JJ,16

-32hh-POWER DISTRIBUTION LIMITS BASES FOR tECTIONS 4.2.4.2 and 5.2.17 MAXIMUM AVERAGE FUEL ASSEMBLY EXPOSURE - (Continued) failure. A limit of 15,000 MWD /MTU fuel element average exposure is consistent with the results obtained from examinations con-ducted on fuel elements with similar 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 prior operation 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.

References:

1. " Technical Evaluation Adequacy of La Crosse Boiling Water Reactor Emergency Core Cooling System", Report SS-942, i Gulf United Nuclear Corporation, May 31, 1972. 2. " Review of Densification Effects in La Crosse Boiling Water Reactor", Report SS-1085, Gulf United Nuclear Corpor-ation, 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. 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. (Next page is page 33) Amen &1ent No. U,16}}