NUREG-0630, Forwards Revised SER Input for Facility

Forwards Revised SER Input for Facility

ML20213E139
Person / Time
Site:	Columbia
Issue date:	03/31/1982
From:	Rubenstein L Office of Nuclear Reactor Regulation
To:	Tedesco R Office of Nuclear Reactor Regulation
References
CON-WNP-0505, CON-WNP-505, RTR-NUREG-0630, RTR-NUREG-630, TASK-1.C.1, TASK-1.C.8, TASK-1.C.9, TASK-1.D.1, TASK-2.F.2, TASK-2.K.3.27, TASK-TM IEB-79-26, NUDOCS 8205120109
Download: ML20213E139 (47)

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/ n Wa 3 /st / 83-f1EMORN1DUtt FOR: R. L. Tedesco, Assistant Director for Licensing, DL FR0!1: L. S. Rubenstein. Assistant Director for Core & Plant Systems, DS[

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SUBJECT:

REVI SED SER I!!PUT FOR WNP-2

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/ N Plant flame: 'UN Docket tiumber:

Wr1P-2 50-397

/L L:/ m td Licensing Stage: Operating License M aC

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Responsible Branch: Licensing Branch flo. 2 G

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Projectfianager: Nu :. .i DSI Review Branch:

R. Auluck '( g .'" i .. . ,d Core Perfomance Branch V.N /'&

Review Status: Two License Conditions in Section 4.7%, i Two License Conditions in Section%4-Four Confimatory Issues in Section 4.2 As a result of the lawyer's review of our previous SER input, we were asked by the project manager to make a large number of editorial changes and to reclassify some of the issues. We have completed that work and have given those changes directly to the project manager to avoid further delays in his SER. Because the changes were rather extensive, we are providing this memorandum to confim the revision.

The license conditions and confimatory issues are now identified as follows:

License Conditions

1. Tlie need for periodic measurements of channel box deflections must be resolved prior to startup of the second cycle of operation.

2. The effects of high burnup fission gas release nust be resolved prior to startup of the second cycle of operation.

3. Operation beyond Cycle 1 is not pemitted until a stability analysis is provided and approved for the additional cycles of operation.

4. A final report to analyze ICC instrumentation requirements should be subnitted by July 1982 for NRC review and approval.

Withdrawn,_I a ssue

1. On reviewing the SER, I personally changed an issue dealing with waterside corrosion from an open to a closed status, thus withdrawing its write-up from this SER. The basis for this change is provided in a memorandum, L. Rubenstein to C. Derlinger, dated 3/30/82 (copy attached).

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' R, L. Tedesco Co_nfirma_ tory I ssues

1. Fuel rod mechanical fracturing.

2. Fuel assembly structural damage from external forces.

3. Fuel rod bowing.

4. Overheating of gadolinia fuel pellets.

L. S. Rubenstein, Assistant Director for Core and Plant Systems Division of Systems Integration

Enclosure:

Revised SER Input cc w/o enclosure:

R. I'.attson D. Eisenhut A. Schwencer R. Auluck DI STRI BUTI ON: ,

Docket Files CPB RDG.

LRubenstein

• CBerlinger i LPhillips g M DFieno RMeyer 3 3fk

• SE E P 1005 CONCURRENCE.

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1 f tEff0RAf'Dul' FOR: R. L. Tedesco, Assistant Director for Licensinc, PL FR0ft: s

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'for Pubenstein, and Plant Systens, Assistant DSIDirector [/

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SUBJECT:

REVISLO SER If!PUT FOR !!i!P-2 '

N Plant flane:  !!flP-2 Docket flumber:

Licensing Stage: \x 50-397 Operating License Responsible Branch:

Project ftanager: \x Licensino Branch flo. 2 R. Auluck DSI Peview Branch: \ Core Perfomance Branch Review Status: NTwo License Conditions in Section 4.2 Two License Conditions in Section 4.4 On'e. 0 pen Issue in Section 4.2 Four onfimatory Issues in Section 4.2 i g As a result of the lawyer's review of our previous SER input, we were asked by the project manacer to nake a laroe n'onber of ' editorial chances and to reclassify sone of the issues. !!e havel completed that work and have given those changes directly to the project r/ananer to avoid further delays in his SEP. Because the chances were rather extensive, we ares,providing this renorandun to confim the revision. /

/ '

The license conditions, open issue, and confimatory issues 'are now identified as follous: /

/

License Conditions '

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1. The need for periodic neasurenents of channel box deflections \ st nu be resolved prior to startup of the second cycle of operation. N N

2. The effects of hig' h burnup fission gas release nust he resolved prior \

to startup of the second, cycle of operation. x

3. Operation beyond Cycle l is not pemitted until a stability analysis is provided and approved for the additional cycles of operation.

4. A final report to analyze ICC instrunentation reouirenents should be submitted by July 1982 for f!RC review and approval.

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Open Issun

1. Fuelgod waterside corrosion.

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'N Confimatory Issues l

1. Fuel rod nechanical fracturino, t I

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2. Fuel assenbly s'tructural danaqe fron externa /

l forces.

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3. Fuel rod bowing.

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4. Overheating of gadolinie fuel pellets. /

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'N L. S.Iruhenstein, Assistant Director

\ fod Core and Plant Systens

' Division of Systens Inteoration

Enclosure:

Pevised SER Input \/

cc: w/o enclosure 'y R. l'attson s D. Eisenhut \

A. Schwencer \

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4 REACTOR '

4.1 General The reactor design for WNP-2 is similar to tbst reviewed and founc ticceptable by the NRC staff for other BWR/4/5/6 pl6nts, such as LaSaile, Shorehan, Susquehanna, Zimmer, Fermi 2,~ Grand Gulf, and Clinton., And, as indicated in the WNP-2 FSAR, Chapter 4 was prepared Osing General E3ectric's BWR/4 and BWR/S fuel design Topical Report (NEDE-20944-P). Additional information of a generic nature has been provided in the General Electric relcad fuel desiga fopical Report (NEDE-24011) and in various documents as discus <:ed below. Therefore, our WNP-2 safety evaluation relies heavily on. previous staff reviews, with appropriate modification as necessitated by the generation of new information.

The r'e view was performed in accordance with the requirements gf the latest version of the Standard Review Plan (NUREG-0800).

4.2 Fuel System Desi6n

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The objectives of this fuel system safety review as described in Section 4.2 of the Standard Review Plan are to provide assurance that (a) the fool syst,em is not damaged as a result of normal operation and anticipated operational occur-rences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always chin-tained. A "not damaged" fuel system is defined as meaning that fuel'.ods 60 not fail, that fuel system dimensions remain within operatior.al toletandes, and that functional capabilities are not reduced below those assumed in the safety

. analysis. Objective (a) above implements General Design Criterion 10 (10 CFR 50, Appendix A), and the design limits that accomplish this are called Specified Acceptable Fuel Design Limits (SAFDLs). " Fuel rod failure" geens that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached. Fuel rod failures r;:ust be accounted for in the dose analysis required by 10 CFR 100 for postulated accidents.

"Coolability," which is sometimes termed "coolable geometry,d means, in general, that the fuel assembly retains its rod-bundle geometrical configuration with adequate coolant channeling to permit removal of residual heat even after a severe accident. The general requirements to maintain control rod insertability and core coolability appear repeatedly in the General Design Criteria (e.g., GDC 27 and 35). Specific coolability requiremente fer the loss-of-coolant accidents are given in 10 CFR 50 Section 50.46.

To assure that the above stated objectives are met, the following areas are examined: (a) design bases, (b) description and design drawings, (c) design evaluation, and (d) testing, inspection, and surveillance plans. In assessing the adequacy of the design, operating experience, prototype testing, and analytical predictions are compared with acceptance criteria for fuel system damage, fuel rod failure, and fuel coolability.

Because the format of the WNP-2 FSAR and its support documents (NEDE-24011-P-A, NEDE-20944-P, and NEDE-20944-1P) does not correspond with the format of SRP muum os

e, t .

Section 4.2, our technical review focused to a large extent on clarifying the design bases and limits and determining the analytical, experimental, and operational support for those bases and limits. Our review was complicated by the fact that the FSAR itself cor.tained little more than an outline that corresponds to the format used in a referenced topical report on SWR /4 anc BWR/5 fuel designs (NEDE-20944-P and NEDE-20944-1P), whereas that report has been partially superseded as a reference document by the General Electric generic reload report (NEDE-24011-P-A). Therefore, as an aid in conducting the review, GE provided a guide or "roadmap" (Engel, August 11, 1981) that was used to cross reference the support documents with SRP Section 4.2. This SER section will follow the format of the SRP.

4.2.1 Design Bases Design bases for the safety analysis address fuel system damage mechanisms and suggest limiting values for important parameters such that damage will be limited to acceptable levels. For convenience, the acceptance criteria for these design limits are grouped into three categories in the SRP: (a) fuel system damage criteria, which are most applicable to normal operation, including anticipated operational occurrences (A00s), (b) fuel rod failure criteria, which apply!to normal operation, A00s, and accidents, and (c) fuel coolability criteria,twhich apply to accidents.

4.2.1.1 Fuel System Damage Criteria

' The following paragraphs discuss the NRC staff's evaluation of the design bases and corresponding design limits for the damage mechanisms listed in the SRP. -

These design limits along with certain criteria that define failure (see Section 4.2.1.2 of this SER) constitute the SAFDLs required by GDC 10. The design limits in this section should not be exceeded during normal operation including anticipated cpe' rational occurrences.

(a) Cladding Design Stress As stated in Section 2.2 of NEDE-24011,* "the fuel assembly must be designed to ensure that possible fuel damage would not result in the release of radioactive materials in excess of applicable regulations."

Also in keeping with the GDC 10 SAFDLs, fuel damage criteria should assure that fue.1 system dimensions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analyses.

While GE does not provide a des 41n basis statement for fuel rod cladding stress in NEDE-24011, it is clear froin the above quoted statement and its relationship to the GDC and SRP requirements that a design basis is implied and may be simply stated as follows: the fuel syst'em shall not be damaged due to cladding stresses. To satisfy that design basis, GE employs the strength theory, terminology, and stress categories presented in the ASME Boiler and Pressure Vessel Code,Section III, as a guide for determining fuel rod cladding stress

^Unless otherwise specified, reference to NEDE-24011 is intended to apply to the latest approved version of the repart, NEDE-24011-P-A-2.

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1 limits. Since, as indicated in SRP subsection 4.2.II.A.1, stress limits that are obtained by methods similar to those given in Section III to the ASME code j are acceptable, the GE cladding stress limits are acceptable.

i (b) Strain Fatigue i

As indicated in Table 2-13 and Table 2-13a of NEDE-24011, the strain fatigue

! criteria described in SRP Section 4.2, viz., a safety factor of 2 on stress  ;

amplitude or a safety factor of 20 on the number of cycles, are satisfied and i

are, therefore, acceptable. In addition, where stress rupture and fatigue

gycling are both significant, GE applies the following limiting condition:

I actual time at stress + actual number of cycles at stress { 1.0 l allowable time at stress allowable cycles at stress

This design limit is more conservative than the SRP and is also acceptable.

(c) Fretting Wear .

l Although no design bases or limits are presented for fretting wear in NEDE 14011, it is evident from the discussion presented in Section 2.6.3 that fretting wear is considered in the design analysis. Confirmation of this is provided in a GE letter (Engel, August 11,1981), which also indicates that -

, analytical criteria for the fuel assembly mechanical design are provided in '

4 subsections 2.5.1 and 2.5.1.1 of NEDE-24011. While these criteria do not i

address fretting wear either, assurance is provided that the general

! requirement (discussed in SER Section 4.2.1.1(a)) concerning reductjen of f functional capability is met. Since the SRP does not provide nur,erical i acceptance criteria for fretting wear, and since fretting wear is addressed in i the design analysis, the NRC staff concludes that the intent of the SRP has i been adequately met.

(d) Exterr:al Corrosion and Crud Buildup With respect to external corrosion and crud buildup, no design bases are l presented in the FSAR or the generic report, NEDE-24011. But, as with fretting  ;

wear, corrosion and crud effects are considered in the design analyses, and j thiy is consistent with the SRP guicelines.  ;

4 (e) Dimensional Chances ,

i Fuel assembly components such as the fuel rods and channel boxes may undergo

! various types cf dimensional changes such as fuel rod bowing, axial growth, and l'

channel box deflection. Such phenomena are related to neutron fluence, fuel  !

burnup, and assembly core residence time and must be accounted for in the fuel design analyses to estcblish operational tolerances and to assure that all i effects are accommodated in th,e thermal and mechanical design.  ;

Fuel ro.d boving is a phenocenon that alters the pitch dimensions between .  ;

adjacent fuel rods and thus affects local nuclear power peaking and heat  !

l transfer to the coolant. GE has established (NEDE-24011-P-A) a "deflectior)"  !

limit corresponding to about half the nominal spacing to ensure that the  !

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rod-to rod and rod-to-ch.annel clearances are sufficient to allow free passage of coolant water to all heat transfer surfaces. Although the SRP does not require specific numerical limits for rod bowing, the proposed deflection limit is consistent with the intent of the SRP and is, therefore, acceptable.

With regard to channel box deflection, GE has issued a generic report on channel design and deflection (NEDE-21354-P) and has responded to NRC questions in two supplements. The GE report documents (a) the fuel channel description, (b) the design bases and analyses, and (c) the creep deflection phenomenon. In the design basis section of the generic report, the design basis for channel structural adequacy is discussed in terms of material properties such as yield strength, stress-rupture strength, and operational duty including normal maneuvers, transients, and accidents. The discussion meets the guidance of the SRP with regard to design bases and is, therefore,' acceptable.

. (f) Fuel and Poison Rod Pressures Section 4.2 of the SRP identifies excessive fuel rod internal pressure as a potential _ fuel system damage mechanism. In this sense, damage is defined as an increased potential for elevated temperatures within the rod as well as an increased potential fpr cladding failures. Because traditional analytical methods for fuel performance analysis do not adequately treat the effects of net outward stress on the cladding and because these effects (e.g., unstable high fuel temperatures and ballooning during DNB events) might be important, the Standard Review Plan calls for rod pressures to remain below nominal system pressure during normal operation unless otherwise justified. Although the applicant does not present a rod pressure limit, it will be seen in Section -

4.2.3.1(g) of this safety evaluation report that the WNP-2 fuel design meets the Standard Review Plan acceptance criterion.

(g) Fuel Assembly _ Liftoff The SRP calls for the fuel assembly holddown capability (gravity and springs) to exceed worst-case hydraulic loads for normal operation, which includes anticipated operational occurrences, Although the applicant does not discuss such a criterion, it will be seen in Section 4.2.3.1(g) of this safety evalua-tion report that the WNP-2 fuel design meets this Standard Review Plan acceptance criterion.

(h), Control Material Leaching '

The SRP and GDCs require that control rod reactivity be maintained. Control rod reactivity can sometimes be lost by leaching of certain poison materials if the control rod cladding is breached, as indeed has been observed (Eisenhut, October 22, 1979). GE has postulated (NEDE-24226-P) that breaching by cracking is caused by stress corrosion resulting from solidification (sintering) of the Boron Carbide (B 4 C) particles in the rods followed by swelling (due to helium and lithium) of the sintered 8 4C. The stress in the tubes caused by the B4 C is believed to accelerate the intergranular corrosion that proceeds from the outside surface of the rod cladding Prior to the discovery of the B C4 leaching loss mechanism, the previously aefined life-!imiting parameter for the , control blades was loss of boron by depletion. The current criterion defining end of control blade life is a loss

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of total reactivity equal to 10 percent of initial control blade worth. If total boron-10 loss by depletion and leaching is considered, GE has determined (NEDE-24226-P) that a 10 percent reduction in worth occurs when the total boron-10 reduction, averaged over the top quarter of the blade, reaches 34 percent. Based on our review of the available data and discussions with GE personnel, we have determined that the 10 percent reduction-in worth criterion is appropriate. The implementation of actions to ensure that the above cri-terion is satisfied is discussed in NRC Office of Inspection and Enforcement Bulletin No. 79-26.

4.2.1.2 Fuel Rod Failure Criteria The NRC staff's evaluation of fuel rod failure thresholds for the failure mechanisms listed in the SRP is presented in the following paragraphs. When these failure thresholds are applied to normal or transient operation, they are used as limits (and hence SAFDLs), since fuel failures under those conditions should not occur (according to the traditional conservative interpretation of GDC 10). When these thresholds are applied to accident analyses, the number of fuel failures must be determined for input to the radiological dose calcula-tions required by 10 CFR 100. The basis or reason for establishing these i

failurethresholdsisjthuspredetermined,andonlythethresholdvaluesare reviewed below.

l (a) Internal Hydriding To prevent internal hydriding, GE specifies a fabrication limit on moisture content. Based on the reported H2 O hydride formation threshold value, the H 2O specification limit for 8x8 fuel (set at a value less than half the hydride formation threshold value) was established on the basis of the previous limit for 7x7 fuel and information obtained from Joon, 1972. To include the total fuel moisture content, GE converts the total fuel rod moisture limit to a total i . fuel column hydrogen limit. The GE limits for moisture and hydrogen are well l below those specified in SRP Section 4.2, and the current ASTM specification

(C776-76, Part 45) for a UO2 pellet-equivalent limit of 2 ppm hydrogen from all

. sources is also satisfied. The GE limits for fuel moisture and hydrogen l content to preclude hydriding failures are, therefore, acceptable.

1 (b) Claddino Collapse If axial gaps in the fuel pellet column were to occur due to densification, the cladding would have the potential of collapsing into a gap (i.e., flattening).

Because of the large local strains that would result from collapse, collapsed i cladding is assumed to be failed. In order to define a collapse criterion to reflect the operational conditions of the reactor, GE has elected (NEDE-20606-P-A) to adopt a collapse criterion that is related to an assumed pressure increase during a turbine trip without bypass; that is, if the fuel rod can sustain, without collapse, an instantaneous increase in the hot system pressure of a given magnitude, it is considered safe against collapse during normal operation, including A00s. The maximum ovality which precedes this collapse-safe transient is defined as the design limit ovality. NEDE-20606-P-A, which l contains these limits and definitions, has been reviewed and approved (Butler,

April 1975). The NRC staff, therefore, concludes that the cladding collapse design basis and limit have been adequately addressed for the WNP-2 fuel i design. ,

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e (c) Overheating of Cladding As indicated in SRP Section 4.2.II.A.2, it has been traditional practice to assume that failures will not occur if the thermal margin criterion is satis-fied. This is a conservative assumption for events that.cause failures by high For BWR fuel, thermal margin is stated in temperature cladding mechanisms.

terms of the minimum value of the critical power ratio (MCPR), which corres-As indicated in Sec-ponds to the most limiting fuel assembly in the core. tion 5 of NEDE-2401 by selecting an MCPR based on a statistical analysis as follows:

" Moderate frequency transients caused by a single operator error or equipment malfunction shall be limited such that, considering uncertainties in manu-facturing and monitoring the core operating state, more than 99.9 percent of the fuel rods would be expected to avoid boiling transition."

Both the normal operation and transient thermal limits in terms of MCPR are derived from this basis, which is described fully in NEDE-10958-P-A and NED0-10958-A.. The design basis and associated limits are consistent with the thermal margin guidelines of SRP Section 4.2.II. A.2 and are thus acceptable from the standpoint /of the fuel mechanical design. The review of thermal /

hydraulic design methods is described in Section 4.4 of this safety evaluation report. .

(d) Overheating of Fuel Pellets Although it is stated in SRP Section 4.2.II. A.2.e that it has been traditional practice to assume that failure will occur if fuel pellet cen fuel. For BWRs, a limited amount of calculated UO2 melting has been permitted for certain events such as rod withdrawal as long as See the melting is not suf-paragraph 4.2.1.2f ficient to cause 1 percent cladding plastic strain.

below for a discussion of the 1 percent strain limit. It should be noted that fuel melting is not expected to occur for normal steady-state full power opera-tion (see Section 2.4.2.5 of NEDE-24011).

(e) Excessive Fuel Enthalpy For a severe reactivity initiated accident (RIA) in a BWR at zero or low power,

. fuel failure is assumed in the'SRP to occur if the radially averaged The 170 calfuel /g rod enthalpy is greater than 170 cal /g at any axial location.

enthalpy criterion, developed from SPERT tests (Grund et al., August 1969), is nrimarily intended to address cladding overheating effects, but it also indirectly addresses pellet / cladding interactions of the type associated with severe RIAs.

As indicated in the Engel letter of August 11,1981 and in NED0-10527, GE uses 170 cal /g as a cladding failure threshold and thus satisfies the SRP.

(f) Pellet / Cladding Interaction Fuel failures due to pellet / cladding interaction (PCI) have been encountered in PCI generally occurs during a power Although in-operating BWR fuel (NEDE-24343-P).

crease as the fuel pellet expands and exerts stresses on the cladding.

the exact mechanisms that contribute' to PCI damage have not been established rs

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beyond a doubt, operating experience indicates that irradiated Zircaloy cannot readily accommodate stresses or strains of this kind, particularly when the Zircaloy has been exposed to certain embrittling (stress-corrosion) fission product species such as iodine or cadmium.

As indicated in SRP Section 4.2.II.A.2.g, there are no generally applicable criteria for PCI failure, but two acceptance criteria of limited application are (a) 1 percent cladding strain and (b) centerline melting. The centerline melting criterion was discussed in paragraph 4.2.1.2d and found to be acceptable.

General Electric employs a 1 percent plastic strain " safety" limit (i.e.,

design limit) for cladding strain based on data from fuel rods operated in BWRs (NEDD-10505). As noted in NEDE-24011, none of the data reported. fell below the 1 percent plastic strain value, but a statistical distribution fit to the available data indicated that the 1 percent plastic strain value was approx-imately the 95 percent point in the total population. The distribution implied, therefore, that there was a small (less than 5 percent) probability that some cladding segments may have plastic elongations less than 1 percent at failure.

Although a 1 percent plastic strain limit will not necessarily preclude all types of PCI failures (Tokar, November 14, 1979), that strain criterion coin-cides with the only other licensing criterion for PCI in the SRP and is, therefore, acceptable l (g) Cladding Rupture

• Zircaloy cladding will rupture (burst) under certain combinations of tempera-ture, heating rate, and stress during a LOCA. While Appendix K to 10 CFR 50 requires that the incidence of rupture during a LOCA not be underestimated, there are no design limits required for cladding rupture. A rupture-temperature correlation (NEDO-20566) is used in the LOCA emergency core cooling system (ECCS) analysis, and that correlation is evaluated in Section 4.2.3.2 of this safety evaluation report.

(h) Fuel Rod Mechanical Fracturing The term " mechanical fracture" refers to a cladding defect that is caused by an externally applied force such as a hydraulic load or a load derived from core plate motion.. These loads are bounded by the loads of a LOCA and safe-shutdown earthquake (SSE), and the mechanical fracturing analysis is usually done as a part of the seismic-and-LOCA loads analysis (see Section 4.2.3.3(d) of this safety evaluation report). Since that analysis has not been completed for WNP-2, it is not clear what design limit will be used for the mechanical fracturing analysis. The NRC staff will report on this issue in a supplement to the safety evaluation report.

4.2.1.3 Fuel Coolability Criteria For major accidents in which severe fuel damage might occur, core coolability must be maintained as required by several GDCs (e.g., GDC 27 and 35). The following paragraphs discuss the staff's evaluation of limits that will assure that coolability is maintained for the severe damage mechanisms listed in Section 4.2 of the SRP.

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(a)

FragmentationkEmbrittledCladding i l To meet the requirebents of 10 CFR 50.46 as i ittemperature relates and to claddin ment for a LOCA, acceptance criteria of 2200 F on peak cladd ngThese c 17 percent on maxirqumeladding oxidation must be met.

by the applicant (see NEDE-20566A).

Violent Expulsion or Fuel (b) d drop, In a severe reactivity initiated accident (RIA) such lt in fuel as a BWR con melt-t into the large an,d rapid deposition of energy in the fuel can resu ing, fragmentation, and violent dispersal of fuelhdroplets fuel or the primary coolant.

can be sufficient to destroy the cladding and ro.d-bundle To meet the guidelinesgeomet and to produce pressure pulses in the primary system.idespread fragmentation an

.of the SRP as it relates to the prevention of w i hin the persal of fuel and the avoidance of pressurel/gpulse should generation be w reactor vessel, a radially averaged enthalpy limit of 280 caand Section 5 de' sign limit.

observed. As indicated in NED0-10527 employs this 280 cal /g criterion as a control rod drop accident

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(c) _Claddinn Ballooning Zircaloy cladding will balloon (swell) under While Appendixcertain K to 10 CFRcombination timated, 50 there ture, heating rate, and stress during a LOCA. d

' requires that the degree of swelling during a LOCA not bel un is, eresA are no design limits required for cladding swelling.is aluation-used in and that correlation is evaluated in Section 4.2.3.3 of this safety e (NED0-20566) report.

Fuel Assembly Structural Damace From External Forces ld (d)

Earthquakes an'd postulated pipe breaks in the reactor thatcoolant result in external forces on the fuel assembly. h n require:

i Aopendix A state that fuel system ccolabilit design limits during these low probability accidents.

pleted .this analysis for WNP-2, it is not cleardixwhat A. The the NRC exact this will be, but they must follow the guideline's of SR safety evaluation.

4.2.2 Descriotio'n and Desicn Drawings 011.

d in The WNP-2 fuel assembly design is descri d NEDE-20944-1P).

ided in the GE the BWR/4 and BWR/S fuel design topical ieport (NEDE-20944-P While each parameter listed in SRP subsection 4.2.2 disthus not satisfy provfficie i

topical reports, enough information is provided in s the intent of the SRP.

4-8 _ _ _

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• 4.2.3 Desian Evaluation Design bases and limits were presented and discussed in SER Section 4.2.1. In this section we review GE methods of demonstrating that the WNP-2 fuel desigr, meets the design criteria that have been established. This SER subsection will, therefore, correspond to subsection 4.2.1 of the SER point by point. The methods of demonstrating that the design criteria have been met include opera-ting experience, prototype testing, and analytical predictions.

4.2.3.1 Fuel Svstem Damage Evaluation The following paragraphs discuss the NRC staff's evaluation of the ability of the WNP-2 fuel to meet the fuel system damage criteria described in Sec-tion 4.2.1.1. Those criteria apply only to normal operation and anticipated transients.

(a) Claddino Desian Stress A description of the fuel' rod stress analysis is provided in subsection 2.5.3.1.2 of NEDE-24011. Detailed evaluations of fuel assembly component structural integrity are presented in NEDE-23542-P along with descriptions of the methods used. As indicated in Section 2.3.1 of NEDE-20944-P, LJditional information regarding the fuel thermal and mechanical analyses is presented in NED0-20360. .

The results of these analyses show that the WNP-2 fuel assembly design meets the stress criteria established by GE. The results of these analyses shows that the WNP-2 fuel assembly design meets the stress criteria established by GE. The NRC staff has not audited these results because problems with design stress during normal operation are not expected based on general LWR fuel operating experience.

(b) Strain Fatigue

. According to information provided in subsection 2.5.3.1.4 of NEDE-24011, GE's fuel design fatigue analysis utilizes the linear cumulative damage rule; i.e.,

Miner's hypothesis (Miner, 1945). In the analysis it is assumed that the area most subject to fatigue damage is the intersection of a fuel rod tube and an end plug at the weld. That intersection is a circumferential notch. Following Peterson's fatigue' theory (O'Donnell and Purdy, May 1964), fatigue failure at a l sharp notch is assumed to be caused by the action cf stress at a finite dis-

'l tance below the material surface. The method to calculate the stress at a small distance below the surface of a notch is provided in a reference (Winnie and Wundt, November 1958). Using that method, and taking the depth below the notch from a reference (O'Donnell and Langer, 1964), GE has calculated (NEDE-24011-P-A) that the cumulative fatigue damage is less than the allowable damage (see SER Section 4.2.1.1.c for fatigue limits). Therefore, based on the t6ct that well established standard theory and methods were used to determine that the cumulative fatigue damage in GE pre pressurized 8x8 retrofit (P8x8R) fuel will not exceed acceptable design limits, the NRC staff concludes that the strain fatigue analysis for WNP-2 fuel is acceptable and that there is reason-able assurance that strain fatigue will not be a problem at WNP-2.

.- , , . . . . - . . . . - . _ - . . - . . . , = - - -

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(c) Fretting Wear In the GE BWR P8x8R fuel design, individual rods in the fuel assembly are held in position by spacers located at intervals along the length of the fuel rod, and springs are provided in each spacer cell so that the fuel rod is restrained to avoid excessive vibration. Various in pile and out of pile tests are described in Section 2.6.3 of NEDE-24011 along with the results of a continuing fuel surveillance program that has utilized nondestructive methods including eddy currents to locate discontinuities in the cladding and detailed visual examinations to characterize the nature of defects. As reported in NEDE-24011, no significant fretting wear has been observed. GE concludes (NEDE-24011-P-A) that significant fretting wear is avoided through the use of an active spring force to eliminate any clearance that would otherwise exist between the spacer structure and the fuel rods. Based on the results'of these fuel test and surveillance programs, we conclude that reasonable assurance has been provided that WNP-2 fuel rods and spacers will perform adequately with respect to fretting wear.

In addition to the fuel rods and spacers, there is another fuel system component whose functionality must be assured as an objective of the review of BWR fuel system fretting concerns; viz., the fuel assembly channel box. The fuel

" channel" that encloses the fuel bundle performs three functions: (1) the channel provides a barrier to separate two parallel flow paths (one to cool the fuel bundle and the other to cool the bypass region between channels); (2) the channel guides the control blade and provides a bearing surface for it; and (3) the channel provides rigidity for the fuel bundle. Thus, the potential for cracks or holes in a " channel" or channel " box" is of concern since it would -

allow part of the cooling water that normally flows through a fuel bundle to flow out of the cracks or holes and bypass the fuel rods. Such a change in flow pattern would reduce the safety margin for fuel thermal performance and would lead to fuel overhe'ating and damage in the event of some anticipated operating transients and postulated accidents. Significant channel box cracking and wear could also adversely affect mechanical strength.

In the mid 1970s, channel box wear and cracking was observed. The wear was located adjacent to incore neutron monitor and startup source locations. It was postulated (NEDC-21084), and later confirmed by out of-reactor testing, that the wear was caused by vibration of nearby incore tubes due primarily to a high-velocity jet.of water flowing through the bypass flow holes in the lower cor,e plate. To eliminate significant vibration of instrument and source tubes and the resultant wear on channel box corners, WNP-2 will incorporate modifica-tions similar to those described (NEDE-21156) for BWRs currently in operation.

Those modifications involve the elimination of the bypass holes in the lower core plate and addition of two holes in the lower tie plate of each assembly to provide an alternate flow path. This design modification has been determined to have n'egligible adverse effects on th'e mechanical, thermal, and nuclear per-formance of the channel boxes, as is discussed in our generic safety evaluation on this subject (Eisenhut, March 2, 1976). Because channel box wear has been observed (Engel, October 15, 1977) to have been significantly reduced in opera-ting BWRs following the design modification, we conclude that there is reason-able assurance that channel box wear and cracking will not be a problem in

.WNP-2.

(d) External Corrosion and Crud Buildup Waterside Corrosion See Memorandum, L. S. Rubenstein to C. Berlinger, dated 3/30/82, 1

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WNP 2 WR 4_11

~

' 4 Crud Buildup The builoup of a corrosion film and a crud layer on the outer surface of a fuel rod during irradiation causes gradual flow reductions and impedes heat transfer to the coolant. The effects of crud buildup on flow are discussed in Sec-tion 4.4.5 of this SER. As indicated in Section 2.4.2.2 of NEDE-24011-P-A, GE calculates the cladding surface temperature using the cladding surface heat flux at a given axial position of a fuel element i~n conjunction with an overall cladding-to-coolant film coefficient that is taken to represent the combined effects of crud and oxide resistances and a liquid film resistance based on the Jens-Lottes (Jens and Lottes, May 1951) wall superheat equation. The impact of high cladding temperature, such as degreased yield strength and reduced clad-ding thickness due to oxidation, was considered in GE's design evaluation

~(NEDE-24011-P-A). GE's methods for analyzing the effects of oxidation and crud on fuel cladding temperatures were reviewed and approved in connection with NEDE-24011-P-A, and the NRC staff, therefore, finds that approach acceptable for WNP-2. ,

(e) Dimensional Changes -

Fuel Rod Bowing -

GE has stated (NEDE-21660-P) that BWR fuel operating experience, testing, and analysis indicate that there is no significant problem with rod bowing even at small rod-to-rod and rod-to-channel clearances. GE noted that (a) no gross bowing has been observed (excluding the rod bowing-related failures in an early design); (b) a very low frequency of minor bowing has been observed; (c) mechan-ical analysis indicates deflections within design bases; and (d) thermal-hydraulic testing has shown that small rod-to rod and rod-to-channel clearances pose no significant problem. The staff 'is reviewing a GE generic topical report (NEDE-24289-P) that is intended to update the GE rod bowing experience and document the overall GE ro.d bowing safety analyses. The NRC staff's review is scheduled for completion by mid 1982. Based on GE's statements that there is no significant problem and the NRC' staff's belief that' rod bowing effects in BWRs are smaller than in PWRs, where the effects have been compensated for with available margins, fuel rod bowing is considered to be a confirmatory issue and will be reported on in a supplement to this safety evaluation report when the generic rod bowing review is complete.

l Channel Box Deflection '

Fuel channel design and deflections under operating conditions are discussed in the BWR/4-5 fuel topical report (NEDE-20944-P), but that discussion is out of date and has been superseded by~a generic report (NEDE-21354-P). The design bases and analyses portions of NEDE-21354 document the channel box development

.and are, therefore, largely of histcrical interest since this design has been in use for some time, however, the creep, deflection phenomenon is relevant to WNP-2 operation. Creep deflection results in the bulging out of the channel

. . . . . . - . . . . . . . - - . . - . - . . - L t .

box in a way that reduces the size of the gap provided for control blade insertion. This dimensional change, if it were large, could interfere with control blade insertion, so channels are discharged before this deflection becomes excessive.

In NEDE-21354, GE describes a channel lifetime prediction method and a bickup

~

recommendation for periodic channel measurements that consist of settling friction tests. The settling friction tests would provide an exact profile of control rod drive friction versus position by measuring the hydraulic pressure under the drive piston as the drive " settles" to a latch position. The NRC staff believes that the friction settling tests or an acceptable alternative (such as channel deflection measurements) should be performed.

Because channel box deflection is a function of service in reactor, there is no need to reach resolution on this issue prior to issuing a license. However, resolution of this issue prior to startup of the second cycle of operation will be made a license condition. This issue has been resolved to NRC staff's satisfaction for Zimmer and 'other recent BWR operating license applications through acceptance of a channel box surveillance plan proposed by those appli-cants (Rubenstein, September 18,1981) and it should be possible to do the same for WNP-2 if a similar plan is proposed for this plant. WNP-2 has been notified (Eisenhut, November 3,1981) of the acceptable resolution of this issue on other plants and has been requested to commit to the acceptable plan. .

(f) Fuel and Poison Rod Pressure The applicant has stated (Bouchey, July 21, 1981) that the internal pressure is used in conjunction with other loads on the fuel rod cladding in the calcula-tion of cladding stresses. The results of such calculations, as provided in NEDE-24011, show that the calculated cladding stresses can be accommodated.

The NRC staff has examined these calculations and agrees that the acceptance criterion for cladding stress (see SRP Section 4.2.II. A.1.a) has been met, but that criterion is not the same as the one for rod internal pressure (see SRP Section 4.2.II.A.1.f). Because the criterion for rod internal pressure involves more than the cladding mechanical limits, the staff considers the GE analysis described in NEDE-24011 to be insufficient for meeting the rod internal pressure criterion. The NRC staff further notes that the rod internal pressures used in the GE cladding stress analysis are well in excess of the nominal coolant system pressure, but because the calculation of these rod internal pressures involves a number of conservative. assumptions, the NRC staff regards them as i excessively conservative for this purpose.

In less conservative analyses such as those previously cited (Sherwood, December 22, 1976), fuel rod internal pressure was shown to remain below system pressure for rod peak burnups below 40,000 mwd /t. This conclusion remains unchanged for l

the newer prepressurized fuel design as well (NEDO-23786-1). Therefore, on the l

basis of these calculations, the NRC staff concludes that the rod internal pressure criterion has been met for WNP-2.

(g) Fuel Assembly Liftoff

ct The potential for BWR fuel assembly liftoff is addressed in .ee General Electric )(

letter (Gridley, July 11, 1977), which covers accident conditions as well as normal operation. Section 4.2.II. A.1.g of the SRP deals with normal operation f - -

- - . - - - . - . . . . . . . . . . . . . . . - . .~

only, however, and so in this SER normal operation liftoff is separated from the combined seismic-and-LOCA liftoff concern, wnich is addressed in SER Section 4.2.3.3.d. As reported by GE in the letter of July 11, 1977, for normal operation and " abnormal transients" (of whicn a stcam-line break LOCA was considered bounding), the minimum net downward force on a fuel bundle, considering only the fluid lifting forces and bundle weight, is approximately 300 lbs. The maximun upward frictional force that can be exerted by a control rod upon a fuel bundle during end-of-life channel scram is approximately 100 lbs. Based on this information, the NRC staff concludes that there is reasonable assurance that WNP-2 fuel assemblies will have sufficient margin to preclude liftoff during normal operation and abnormal transients (up to a steam-line break LOCA).

(h) Control Material Leaching -

. The loss of boron carbide (B4C) by leaching from cracked control blade tubing is addressed in NRC IE Bulletin No. 79-26, Revision 1, which requires operating BWRs to perform various actions including, but not limited to, shutdown-margin tests. Since WNP-2 will be subject to the IE Bulletin when the plant starts operation, assurance will be maintained that control blade reactivity will not be significantly degraded by boron carbide leaching. Subsequent to the issuance of IE Bulletin 79-26,EGE performed analyses and postirradiation examinations (NEDE-24226-P) and developed a boron depletion model (NEDE-24325-P). This .

model supports GE's claim that the amount of boron loss can be accurately determined analytically and that potential control blade degradation due to this mechanism will not significantly affect plant operation. Further generic review of this matter may thus lead to a suspension or revision of IE Bul-letin 79-26 requirements.

4.2.3.2 Fuel Rod Failure Evaluation The following paragraphs discuss the staff's evaluation of (a) the ability of the WNP-2 fuel to operate without failure during normal operation and antici-pated transients, and (b) the accounting for fuel rod failures in the applicant's accident analysis. The fuel rod failure criteria described in Section 4.2.1.2 were used for this evaluation.

(a) Internal Hydriding GE employs a hot vacuum outgassing (drying) method for removal of the moisture con'tamination of loaded fuel rods to assure that the level of moisture is well below the limits discussed in SER Section 4.2.1.2.a. And, to prevent the' introduction of other hydrogenous impurities such as oils, plastics, etc.,

various procedural controls are utilized during manufacturing (NEDE-24011-P-A).

If even in the face of those manufacturing controls, moisture or hydrogenous impurities are still inadvertently introduced into a fuel rod during manufac-ture, further assurance against chemical attack is provided by the use of a hydrogen getter material placed in the upper plenum of the fuel rod. The incorporation of a hydrogen getter in GE fuel rods is described in NED0-20922.

GE calculations, based on an assumed initial rod moisture content equal to the specification limit, indicate that no cladding damage will occur during the active hydrogen / moisture gettering period (NEDE-24011-P-A). Field experience (NEDE-24343-P) has confirmed that hydrid,ing is not an active failure mechanism for fuel manufactured since mid-1972. On th'e basis of the information presented, t _. _

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' we conclude that adequate assurance has been provided that fuel and burnable poison rod hydriding will not be a problem at WNP-2.

(b) Cladding Collapse As stated in Section 2.5.3.1.1 of NEDE-24011, GE has performed a cladding'

. (long-term creep) collapse analysis. Based on the results of that analysis, which was performed with an approved model (NEDE-20606-P-A), cladding collapse was not calculated to occur for either 8x8 or 8x8R fuel rods. Consequently, P8x8R (pre pressurized) fuel, the type to be used in WNP-2, will have an even greater margin to collapse. The NRC staff, therefore, concludes that there is reasonable assurance that long-term creep collapse will not occur in WNP-2 fuel during normal operation.

(c) Overheating of Cladding The methods employed to show that the MCPR design basis will be met are re-viewed as part of a thermal-hydraulics review in Section 4.4 of this safety evaluation report. Analyses that show that the MCPR design basis is met for normal operation and anticipated transients are discussed in Chapter 15 of the WNP-2 FSAR. Other sections of FSAR Chapter 15 discuss accidents where the MCPR criterion is used to define failure in the accident analyses. All of the MCPR related analyses are discussed in Chapter 15 of this safety evaluation report.

(d) Overheating of Fuel Pellets Fuel melting temperature is discussed in Section 2.4.2.5 of NEDE-24011 as a function of exposure (burnup) and gadolinia content (of burnable poison rods).

GE states in that report that fuel melting is not expected to occur during normal operation, and that statement is based on fuel temperature calculations performed with a model described in the proprietary supplement to Ammendment 14 of GESSAR. While limited melting during certain events such as an uncontrolled

' control rod withdrawal is permissible, such melting is not predicted to occur.

We have reviewed the U02 properties (thermal conductivity and melting point) that are important in reaching this conclusion and agree that UO2 melting will not be a problem at WNP-2 during normal operation and anticipated transients.

However, the effects of gadolinia concentration on thermal conductivity and melting temperature are addressed in a GE topical report on gadolinia fuel properties (NEDE-20943-P), and our review of this subject has not been completed.

We are, therefore, unable to comment on that aspect of the fuel melting issue at this time. The absence of significant melting in gadolinia-bearing fuel pellets in WNP-2 will be addressed in a supplement to this safety evaluation report and is considered to be a confirmatory issue.

(e) Excessive Fuel Enthalpy Large fuel enthalpies occur only in the postulated control rod drop accident.

GE's analysis of the control rod drop accident is described in Section 5.5.1 of NEDE-24011 and in NEDO-10527. The analysis of this event is reviewed in Section 15.4.9 of this safety evaluation report.

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l (f) Pellet /Claddina Interaction The two PCI criteria in current use in licensing of BWR fuel (limited fuel melting and 1 percent cladding plastic strain) are easily satisfied for normal operation and anticipated transients. Fuel melting is addressed in SER sub-section 4.2.3.2.d and, expect for uncertainties with regard to gadolinia fuel, the fuel melting criterion is met. The model used by GE in the evaluation of the 1 percent plastic strain limit is described in detail in the proprietary supplement to Amendment 4 of GESSAR. As stated in NEDE-24011, dimensions used in conjunction with the model for that evaluation are the most limiting combina-tion of tolerances. These results show that the WNP-2 fuel meets the 1 percent strain. limit. The NRC staff has not audited these results because the tradi-tional 1 percent strain limit is not believed to be precise.

Based on results of developmental investigations and feedback from production

' fuel experience, operating restrictions known as Pre-conditioning Interim Operating Management Recommendations (PCIOMRs) were issued by GE to the BWR operators (NEDS-10456-PC) and will be used at WNP-2. These restrictions are necessary to reduce the. incidence of PCI failures even though the above criteria are met. PCIOMRs have generally been effective in reducing PCI failures that result from operation,al power changes, but they would not prevent PCI failures during unexpected trahsients and accidents.

The staff is continuing to assess the potential for PCI failure during power-increasing transients and accidents, and new techniques are being developed to analyze the potential for PCI failure.s (NUREG/CR-1163) and determine whether or not further staff actions are required. ,

In conclusion, (a) the applicant has met NRC's licensing criteria for normal operation and anticipated transients, (b) the applicant will impose operating restrictions to reduce the potential for PCI, and (c) the NRC staff is studying

.the need for new licensing requirements in this area. There are presently no other PCI licensing requirements that must be met for WNP-2.

(g) Claddina Rupture (Bursting)

The staff has been generically evaluating three fuel materials models-that are used in the ECCS analysis. Those models predict cladding rupture temperature, cladding burst strain (ballooning), and fuel assembly flow blockage (used only in PWR analyses). The staff has (a) discussed its evaluation with vendors and other industry representatives (Denise, November 20, 1979), (b) published NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis," and'(c) required licensees to confirm that their operating reactors would continue to be in conformance with 10 CFR Part 50.46 if the NUREG-0630 models were sub-stituted for the present materials models in their ECCS evaluations and certain other compensatory model changes were al-lowed (Eisenhut, November 9, 1979; and Denton, November 26,1979). This latter requirement was made a condition of approval to the GE ECCS evaluation model (Tedesco, february 4,1981).

A generic sensitivity study of fuel rod cladding ballooning and rupture phe-nonema during a LOCA was submitted (Buchholz, May 15, 1981) in response to this condition of approval. As reported in the generic study, GE assessed the BWR

'ECCS sensitivity to rupture temperature by using three rupture temperature models: (1) the GE CHASTE model, (2) th'e NUREG-0630 model, and (3) a proposed gg. 3_gg g-w

e,. . ' .

GE model termed the' adjusted model. For the WNP-2 type of 8x8 two-water rod fuel design, GE found that the use of the adjusted model, which is a combination oftheCHASTEandNUREGmodels,gaveamaximumimpactonPCTof{10F.

In a recent letter (Bouchey, January 19, 1982) the applicant agreed to use the calculated WNP-2 LOCA peak cladding temperature margin (i.e., 2200 F - 1960 F =

240 F) solely to offset the cladding rupture and cladding ballooning (see Section 4.2.3.3.c) uncertainties. Inasmuch as those combined uncertainties appear to be much less than the 240 F available margin, the NRC staff concludes that cladding rupture has been adequately accounted for in the WNP-2 LOCA analysis.

(h) Fuel Rod Mechanical Fracturing The mechanical fracturing analysis is usually done as a part of the seismic-and-LOCA loads analyses (see Section 4.2.3.3(d) of this safety evaluation report).

Since that analysis has not been completed for WNP-2, the information on mechanical fracturing is not'available. We will report on this issue in a supplement to the safety evaluation report.

4.2.3.3 Fuel Coolability Evaluation f

The following paragraphs discuss the staff's evaluation of the ability of the .

WNP-2 fuel to meet the fuel coolability criteria described in Section 4.2.1.3.

Those criteria apply to postulated accidents.

(a) Fragmentation of Embrittled Cladding The primary degrading effect of cladding oxidation is embrittlement. Such embrittled cladding will have reduced ductility and resistance to fragmen-tation. The most severe manifestation of such embrittlement occurs during a LOCA. The overall effects of a LOCA on peak cladding temperature, oxidation, and embrittlement are analyzed in NEDE-20566 and are reviewed in Section 6.3.3 of this safety evaluation report.

One of the most significant analytical methods that is used to provide input to the analysis in Section 15.6.5 is the steady-state fuel performance code, which is reviewed in Section 4.2. This code provides fuel pellet temperatures (stored energy) and fuel rod gas inventories for the ECCS evaluation model as prescribed by Appendix K to 10 CFR 50.46. Initial conditions for the LOCA have been calculated for the WNP-2 fuel with an approved version of the General Electric analytical model, GEGAP III (NEDC-20181). The model incorporates time-dependent fuel densification, time-dependent gap closure, and cladding creep-down for the calculation of gap conductance.

In 1976 the NRC staff questioned (Ross, November 23, 1976) the validity of fission gas release calculations in most fuel performance codes including GEGAP-III for burnups greater than 20,000 mwd /tU. General Electric was informed of this concern and was provided with a method (NUREG-0418) of correcting gas release calculations for burnups greater than 20,000 Kdd/tu. Although a reanalysis was not performed specifically for the WNP-2 fuel, an 8x8 reanalysis (Sherwood December 22, 1976) was performed for early reflooding plants and reportedly bounds the WNP-2-case. While the fuel rod internal pressure was shown to remain below system pressure for rod peak burnups below 40,000 mwd /t,

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I the generic reanalysis d.id result in higher initial stored energy and rupture pressure in the LOCA analysis. Under LOCA conditions, the higher fission gas release results in a maximum increase of 85 F in calculated peak cladding temperature (PCT) at end of-life ((33,000 mwd /tU planar average exposure).

This increase in PCT has not been accounted for in the WNP-2 ECCS analysis.

In recent submittals (Engel, May 6 and May 28, 1981) the General Electric Company requested that credit for calculated peak cladding temperature margin as well as credit for recently approved ECCS evaluation model changes be used to offset any operating penalties due to high burnup fission gas . release.

While we have found this proposal acceptable in other cases, we note that the applicant (Bouchey, January 19, 1982) has relinquished credit for calculated peak cladding temperature to satisfy another staff concern (see Cladding Rupture Section). As a result, we conclude that the General Electric request is not applicable to WNP-2. However, we find the fission gas effects to be adequately analyzed for early-in-life operation, and the licenseg will be S conditioned to require resolution of this issue prior to startup of the second cycle of operation.

'(b) Violent Expulsion of Fuel i

The analysis, which demonstrates that the acceptance criteria are met, is evaluated in Section 15.4.9 of this safety evaluation report. .

(c) Cladding Balloonino As discussed previously in Section 4.2.3.2(g) of this safety evaluation report, -

GE has submitted a generic sensitivity study (Buchholz, May 15, 1981) in response to a condition in the approval of the General Electric ECCS Evaluation Model.

With regard to the BWR ECCS sensitivity to burst strain (ballooning), the' General Electric submittal assessed the impact of using a burst strain model that bounds the burst strain model given in NUREG-0630. However, prior to performing this comparison, the bounding strain mcdel was appreciably reduced by axially averaging the cladding strain. Two reduction factors were used to effect this averaging process; one was 2.8 (for fuel bundle interior rods) and the other was 4.1 (for fuel bundle peripheral rods). While some averaging is appropriate for the whole-bundle analysis performed for WNP-2, there is still some uncertainty as to the appropriateness of the factors used by General Electric.

Notwithstanding the possibility of a future increase in calculated PCT result-ing from a modification in the reauction factors for averaging the cladding strain, the calculated LOCA PCT margin available (Bouchey, January 19, 1982) is adequate to offset this concern as well as the cladding rupture concern expres-sed previously in Section 4.2.3.2g. Therefore, the NRC staff concludes that cladding ballooning has been adequately accounted for in the WNP-2 LOCA analy-sis.

The overall impact of cladding ballooning on the WNP-2 loss-of-coolant accident

. is evaluated in Section 15.6.5 and is not reviewed further in this section.

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(d) Fuel Assembly Structural Damage From External Forces An analysis must be provided by the applicant that shows that the WNP-2 fuel meets the structural requirements (including liftoff) of Appendix A to Sec-tion 4.2 of the SRP (NUREG-0800). Since generic analytical methods presented in NEDE-21175 have been approved.by the NRC staff (Parr, May 17, 1979) and since favorable sample results were also presented in NEDE-21175, we consider this issue to be confirmatory for WNP-2. We will report on this issue in a supplement to the safety evaluation report.

4.2.4 Testing, Inspection, and Surveillance Plans 4.2.4.1 Testing and Inspection As described in SRP Section 4.2, fuel testing and inspection plans should include verification of-significant fuel design parameters. While details of the manufacture's testing and. inspection programs should be documented in quality control reports, the programs for onsite inspection of new fuel and control assemblies after they have been delivered to the plant should also be described in the SAR (or in supporting documents).

The GE generic reload report (NEDE-24011) and the BWR 4/5 topical report (NEDE-20944) contain descriptions of (1) the type of quality control inspec-tions performed during manufacturing and (2) the plan for onsite inspection and testing of fuel assembles. The applicant has stated that the inspection plan outlined in Table 2-6 of NEDE-20944 will be used as a guide. We conclude, on the basis of the information provided in these referenced reports, that the new-fuel testing and inspection program for the WNP-2 fuel design is acceptable.

4.2.4.2 On-Line Fuel System Monitoring

.BWRs such as WNP-2 have two independent radiation detection systems that are directly capable of sensing fission product releases from failed fuel rods.

The main steam line radiation (MSLR) monitors are used to detect high radiation levels resulting from gross fuel failure. At WNP-2 four redundant MSLR monitors will be used, each of which provides a signal to trip the turbine and initiate reactor scram independent of operator action. Because the MSLR monitors are

located relatively close to the reactor core, they are capable of sensing gross i

fission product releases relatively quickly (i.e., in a few seconds) while the off gas system radiation (OGSR) monitors are capable of detecting low-level l

emissions of noble gases, which could indicate the occurrence of minor fuel damage, in two to three minutes, the time required for the activity to travel from the core to the detectors. The OGSR monitors are set to sound alarms that would initiate operator action (see Chapter 11 of the WNP-2 FSAR for details).

l The NRC staff has reviewed the BWR activity monitors (NUREG-0401) and concludes that the combination of (a) both radiation detection systems and (b) implemen-l tation of applicable Technical Specifications on Limiting Safety System instru-mentation setpoints and specific activity (Technical Specification Section 2.2 and Technical Specification 3/4.4.5, respectively) provide reasonable assurance of the adequacy of the on-line fuel system monitoring at WNP-2.

s. .

t 4.2.4.3 Postirradiation Surveillance In recent submittals (Bouchey, January 6,1982 and January 19, 1982 the applicant has committed to establish a routine fuel inspection progr)am to The program will involve visual examination of selected Typically, asse visual examination concentrating will bundles.

on the lead be made of five to ten percent of the discharged fuel, Visual examination will include but not Additional fuel inspection will be performed, depending on th operational monitoring including coolant activity and the visual fuel inspec-tion.

On-the basis of the above commitment, we conclude that the issue of postirradiation surveillance has been adequately addressed for WNP-2. ~

4.2.5 Mechanical Design Evaluation Findings The WNP-2 fuel system design has been reviewed in accordance with SRP Section 4.2. The staff concludes that, although most of the objectives of the fuel system safety review have been met, there are several con-firmatory issues and one open issue to be resolved before the review can be completed. The confinnatory issues are the following:

and 4.2.3.2(8)), (2)(1) fuel rod mechanical fracturing (SER Sections 4.2.1.2(8)

Sections 4.2.1.3(4) and 4.2.3.3(4)), (3) fuel assembly structural damage from and (4) overheating of gadolinia fuel pellets (SER Section 4.2.3.2(4)). The open fue issue is f_uel rod waterside corrosion (SER Section 4.2.3.1(4 .

The staff has also determined that license conditions are needed for the fo)llowing issues:

burnup fis,sion gas release on LOCA analysis (SER Secti When the confirmatory and open issues are resolved, the staff will be able to conclude that the WNP-2 fuel system has been designed so that (1) it will not be damaged as a result of normal operation and anticipated operational occur-rences, (2) fuel damage during postulated accidents would will always be maintained even after postulated acciden related requirements of the following regulations:

and 35; and 10 CFR 50, Appendix K. 10 CFR 50.46; GDC 10, 27, factors: ._

This conclusion is based on two primary (1)

Withlheexceptionoftheremainingconfirmatoryandopenissues,the applicant has provided sufficient evidence that the design objectives will be met based on operating experience, prototype testing, and analytical predictions.

Those analytical predictions dealing with control rod drop and fuel densification have been performed in accordance with Regulatory Guide 1.77 and methods that the staff has reviewed and found to be acceptable alternatives to NRC Regulatory Guide 1.126.

2.

.The applicant has provided for testing and inspection of new fuel to ensure that it is within design tolerances at the time of core loading.

On-line fuel failure detection equipment will be in place to provide warning of cladding perforations during plant operation. The applicant i

has made a~ commitment to perform po. stir. radiation surveillance to detect anoma21 a

& s or confirm that the fuel has performed as expected.

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Ihe staff conc]udes that the applicant has described methods of adequately .

predicting fuel rod failures during postulated accidents so that radioactivity releases are not underestimated, thereby meeting the related requirements of 10 CEE Part 100. In neeting those requirements, the applicant has (a) used the fission product release assumptions of NBC Regulatory Guide 1.3 and 1.25, and (b) performed the ana]ysis for fuel roa failures for the rod drop accident in

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accordance with the guidelines of NBC Regulatory Guide 1.ZZ.

On the basis of this review, we conclude that, with the above exceptions, a))

the requirements of the applicable regulations have been met and guidelines of the applicable Begulatory Guides and current regulatory positions have been followed for WNP-2 fuel design.

4.3 Nuclear Desian Ihe applicant has referenced the General Electric Licensing Iopical Report NEDE-20944-P (non proprietary version NED0-20944) for al) portions of Section 4.3 e'xcept that pertaining to irradiation of the reactor pressure vessel. This report, entitled "BWR/4 and BWE/5 fuel Design," has been reviewed and approved for such reference (Parr, September 30, 1977). Ihe staff finds the use of this reportforWNP-2acce{ table.

4.3.1 Vessel Irradiation -

The neutron fluence at the inside surface of the pressure vessel has been calculated by means of a one-dimensional transport theory (discrete ordinates) code in an infinite cylinder geometry. The gross radial power distribution was used in the distributed source option for the code. The actual flux values were those at the core height having the axial peak flux when the reactor is operating at full power. The core, core shroud, and down-comer region were modeled along with the pressure vesse). An azimuthal peaking factor is calcu-lated by a two-dimensional transport theory method. The vessel flux was obtained by forming the product of the one-dimensional flux, the azimuthal factor and an additional factor of 1.5, which was included to ensure conserva-tism. This flux is multiplied by a factor of 1 x 109 (equivalent to 40 years at a capacity factor of 0.8) to obtain the value of vessel fluence. This is 1.4 x 1018 neutrons per square centimeter for neutrons having energy greater than 109 electron-volts.

The staff concludes that the analysis of the pressure vessel fluence is accep-table. This conclusion is based on the fact that state of-the-art analysis methods are used and that a large conservatism factor is applied to the calcu-lated result.

4.3.2 Evaluation Finding The WNP-2 nuclear design was reviewed according to Section 4.3 of the Standard Review Plan (NUREG-0800).

The applicant has described the computer programs and calculational techniques used to predict the nuclear characteristics of the rractor design and has provided examples to demonstrate the ability of these methods to predict experimental results. The staff concludes that the information presented adequately demonstrates the ability.of these analyses to predict reactivity and physics characteristics of the WNP-2 plant.

J s .

Io allow for changes of reactivity due to reactor heatup, changes in operating conditions, fuel burnup, and fission product buildup, a significant amount of excess reactivity is designed into the core. Ihe applicant has provided sub-stantial information relating to core reactivity requirements for the first cycle and has shown that means have been incorporated into the design to control excess reactivity at all times. The applicant has shown that suf-ficient control rod worth is available to shut down the reactor with at least a 1.0 percent Ak/k subcritical margin in the cold condition at any time during the cycle with the highest worth control rod stuck in the fully withdrawn position.

On the~ basis of its review, the staff concludes that the applicant's assessment of reactivity control requirements over the first core cycle is suitably conservative and that adequate negative worth has been provided by the control system to assure shutdown capability. Reactivity control requirements will be reviewed for additional cycles as this information becomes available.

The staff concludes that the nuclear design is acceptable and meets the require-ments of 10 CFR 50, Appendix A, General Design Criteria 10, 11, 12, 13, 20, 25, 26, 27, and 28. This conclusion is based on the following:

b

1. The applicant has met the requirements of GDC 11 with respect to prompt inherent nuclear feedback characteristics in .the power operating range by: -

a. calculating a' negative Doppler coefficient of reactivity, and

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b. using calculational methods that have been found acceptable. -

The staff has reviewed the Doppler reactivity coefficients in this case and found them to be. suitably conservative.

2. The applicant has met the requirements of GDC 12 with respect to power oscillations which could result in conditions exceeding specified accept-able fuel design limits by:

a. showing that such power oscillations are not possible and/or can be easily detected and thereby remedied, and

b. using calculational methods that have been found acceptable.

9 The' staff has reviewed the analysis of these power oscillations in this case and found them to be suitably conservative.

3. The applicant has met the requirements of GDC 13 with respect to provision of instrumentation and controls to monitor variables and systems that can affect the fission process by:

a, providing instrumentation and systems to monitor the core power distribution, control rod positions and patterns, and other process variables such as pressure, and

b. providing suitable alarms and/or control room indications for these monitored variables. . .

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4. The applicant has met the requirements of GDC 26 with respect to provisions of two independent reactivity control systems of different designs by:

a. having a system that can reliably control anticipated operational occurrences,

b. having a system that can hold the core subcritical under cold condi-tions, and

c. having a system that can control planned, normal power changes.

5. The applicant has met the requirements of GDC 27 with respect to reac-tivity control systems that have a combined capability in conjunction with poison addition by the standby liquid control system of reliably control-ling reactivity changes under postulated accident conditions by:

a. providing a movable control rod system and a liquid poison system, ,

and

b. performing calculations to demonstrate that the core has sufficient shutdown mafgin with the highest worth stuck rod.

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6. The applicant has meet the requirements of GDC 28 with respect to postu- -

lated reactivity accidents by:

a. meeting the fuel enthalpy limit of 280 cal /gm,

b. meeting the criteria on the capability to cool the core, and

c. using calculational methods that have been found acceptable for reactivity insertion accidents.

7. The applicant has met the requirements of GDC 10, 20, and 15 with respect to specified acceptable fuel design limits by providing analyses demon-strating:

a. that during normal operation, including the effects of anticipated operational occurrences, fuel design criteria will be met.

b. that the automatic initiation of the reactivity control system assures that fuel design criteria are not exceeded as a result of anticipated operational occurrences and assures that systems and components important to safety will operate automatically under acci-dent conditions, and

c. that no single malfunction of the reactivity control system causes violation of the fuel design limits.

4.4 Thermal and Hydraulic Desian The scope of the review included the design criteria, implementation of design criteria as presented by the final core design, and the analyses of core

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thermal-hydraulic performance. The acceptance criteria used as the basis for the staff's evaluation are set forth in Chapter 4.4 of the Standard Review Plan (NUREG-0800).

4.4.1 Thermal Hydraulic Design Bases The thermal-hydraulic safety design bases for WNP-2 can be summarized as follows:

1. Fuel failure due to cladding overheating should not occur as a result of moderate frequency transients. Specifically, the minimum critical power ratio (MCPR) operating limit is specified such that at least 99.9 percent of the fuel rods in the core are not expected. to experience boiling transition during the most severe moderate frequency transient.

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2. Core and fuel thermal-hydraulic design limits for steady-state operation, i.e., MCPR and LHGR limits, should be defined to provide margin between steady-state operating conditions and any operating condition that could result in fuel failure. Specifically, they should assure that no fuel failure results puring the worst moderate frequency transient even with uncertainties ta, ken into account.

3. No undamped oscillations or other hydraulic instabilities should occur ~

either for normal operation or for the most severe moderate frequency transient.

The applicant's thermal-hydraulic design bases are consistent with the Standard '

Review Plan and are, therefore, acceptable.

4.4.2 Thermal-Hydraulic Analysis Basis The General Electric Thermal Analysis Basis (GETAB) described in NED0-10958 was used for WNP-2. The figure of merit chosen for reactor design and operation is the critical power ratio, the ratio of the critical bundle power to the opera-ting bundle power. This method has been previcusly reviewed (Butler, October 21, 1974) and found acceptable.

Critical power tests have been run on prototyp' al 8x8 fuel bundles with two water rods. Test data for cosine axial heat flux shapes indicate that the water rods do not affect the GEXL capability of predicting the bundle critical

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power performance for bundle radial peaking patterns typical of 8x8 retrofit fuel. The staff has previously found that the GEXL data base, which includes top and bottom peak axial heat flux distributions combined with the two-water rod data, demonstrates the adequacy of the GEXL correlation to predict critical power in both 8x8 and 8x8 retrofit bundles. The staff has previously concluded that the GEXL correlation is acceptable for both 8x8 and 8x8 retrofit fuel application (Tedesco, April 16, 1981).

4.4.3 Thermal-Hydraulic Analysis Methods The MCPR limit originally proposed was based upon calculations using-the REDY model described in NED0-10802. The results from the tests performed at Peach

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Bottom-2 revealed that in certain cases the results predicted by REDY are nonconservative. Therefore, the staff required the applicant to use ODYN methods to analyze the following transients: (1) feedwater controller failure at maximum demand with and without bypass, (2) generator. load rejection,.and (3) turbine trip with and without bypass. The staff reviewed the applicant's reanalysis results, which are included in a WPPSS letter dated January 11, 1982. These results indicate that the operating limit MCPR is 1.24 as determined by using the maximum calculated transient MCPR, which was for the rod withdrawal error events. Based on its review, the staff finds that the results of the analyses using the approved ODYN methods do not violate the safety limit MCPR of 1.06. Therefore, the staff concludes that the proposed operating limit MCPR value of 1.24 is acceptable.

4.4.4 Thermal-Hydraulic Stability Recent BWR fuel design changes which affect stability include decreasing the rod s.ize and increasing the gap conductance because of pre pressurization. As a consequence, the maximum decay ratio for most BWRs increases and becomes larger than 0.5, which is the original GE design criterion for BWR stability.

Therefore, GE now proposes a decay ratio of 1.0 for their criterion. Ths staff has not agreed that the proposed criterion of a 1.0 decay ratio calculated by using the FABLE code is acceptable.

To further evaluate this criterion and other stability criteria, the staff is .

performing a generic study of the hydrodynamic stability characteristics of light water reactors under normal operation, anticipated transients, and accident conditions. The results of this study will be applied to the staff's review and acceptance of stability analyses and analytical methods now in use by the reactor vendors.

The WNP-2 stability analysis resulted in a maximum decay ratio of 0.7 for the

'end of the first cycle. The staff has approved for operation the Susquehanna core design (NUREG-0776), which has a calculated maximum decay ratio value of 0.7 for the initial cycle. Since Susquehanna and WNP-2 have similar core designs, the staff concludes that WNP-2 core design stability is acceptable for Cycle 1. However, in order to provide additional margin for stability, natural circulation operation of WNP-2 will be prohibited until the staff's review of these conditions is completed. Any action resulting from the staff's study will be applied to WNP-2.

The WNP-2 operating license will be conditioned to require that a new stability analysis be submitted and approved prior to second-cycle operation. Also, since no analysis has been presented for minimum critical power ratio limits or stability characteristics for single loop operation, the staff will require by Technical Specifications that single loop operation not be permitted until supporting analyses are provided and approved.

4.4.5 Crud Deposition Crud deposition causes gradual flow reductions in some light water reactor cores. However, measurement of core flow by jet pump pressure drop and core 4

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plate pressure drop will provide adequate indication of such flow reductions, if they should occur. Technical Specifications will require that the core flow be checked at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to detect flow reduction.

For pressure drop considerations in design analyses, it is assumea that a conservatively large amount of crud is deposited on the fuel rods and the fuel rod spacers. This is reflected in decreased flow area, increased friction factors, and increased spacer loss coefficients. The effect of this crud deposit is to increase the core pressure drop by approximately 1.7 psi thereby resulting in additional thermal margin. Therefore, the staff concludes that the assumptions regarding crud deposition used in the design analysis in conjunction with the required flow monitoring are acceptable.

4.4.6 Loose Parts Monitorino System

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WNP-2 will have a loose parts monitoring system (LPMS) operational at the time of initial reactor startup testing. This system provides.the capability to detect, alarm, and record acoustic signals generated when loose parts within

.the reactor coolant system impact other reactor coolant system components. The system has been installed to meet the operability requirements of Regulatory Guide 1.133. The appiicant also agrees to evaluate the system to address conformance with the Regulatory Guide 1.133, Revision 1 (Ma The conformance evaluation should emphasize the following areas:y 1981). '

l. A description and evaluation of diagnostic procedures used to confirm the presence of a loose part. .

2. A description of how the operators will be trained in the purpose and implementation of the system.

The staff will review the applicant's conformance~ evaluation report when it becomes available, consistent with the staff's plans for review of the opera-ting plants. Any action resulting from the staff's review will be applied at the time. On this basis, the staff finds the LPMS acceptable for an operating license.

4.4.7 TMI-2 Action Plan Item II.F.2 Requirements ,

A clarification of requirements for inadequate core cooling (ICC) instrumenta-tion, which is required to be installed and operational prior to fuel loading, was provided in a letter to all operating nuclear power plants (Denton, October 30, 1979) and in Section II.F.2 of NUREG-0737, " Clarification of TMI Action Plan Requirement." The requirements specified in NUREG-0737 are the basis for the staff's review of the applicant's submittals in response to Section II.F.2 of the TMI Action Plan.

Inadequate Core Coolino Detection System The BWR Owners' Group, of which WPSS is a member, t.ransmitted a letter to the NRC (Waters, January 31, 1981) with attachments. Unambiguous procedures to

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recognize the approach to ICC are given in one of those attachments, "BWR Emergency Procedure Guidelines, Revision 1 (January 30, 1981)." These pro-cedures rely primarily on decreasing reactor vessel water level as an indication of the approach to ICC. The water level instruments were evaluated by the BWR Owners' Group and documented in NED0-24708. The wide range instruments offscale

, low are presumed by the operator to indicate the approach to ICC. Fuel-zone level measurement is an alternate method for detection of ICC.

Adequate core cooling is assured whenever the reactor is shut down and one or more of the following conditions exist: (1) the active fuel is covered with liquid or a two phase mixture; (2) ECCS flow is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembly; and (3) steam flow is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembly.

One core spray system operating alone is sufficient to cool the core. The emergency guidelines and procedures specify the minimum core water level that is sufficient to remove decay heat by steam cooling alone.

Otherthanwaterleve),thereisnoinstrumentationavailablewhichwill provide an unambiguous indication of approach to ICC. The source range monitors will show changes in w' ater level in the core, but they may be misleading as to whether the level is increasing or decreasing. Incore thermocouples, if they '

were available, might indicate bundle uncovery, but due to spray from-the spray system above the core, they might indicate saturated or subcooled conditions even when the core is partially uncovered. However, under core uncovery conditions without core spray available, incore thermocouples could indicate coolant superheat associated with excessive fuel cladding temperature.

The principal method of confirming adequate core cooling, namely, the reactor pressure vessel water level instrumentation, has been shown through analysis

'and experience to be adequate to assure detection of approach to ICC.

Reactor vessel water level is measured by differential pressure devices.

Condensing chambers connected to the steam space in the reactor vessel furnish saturated water to the reference leg. Pressure taps located at different levels in the water space of the reactor vessel are used as the variable leg sensing taps for the water level instruments.

' Because of the presence of voids in the fluid inside the shroud (core area),

the swollen level inside the shroud will be higher than the level outside the shroud until ECCS operation commences. Thus the level measurement in the annulus region is a conservative measurement of the two phase level inside the shroud prior to ECCS injection. ECCS injection, which injects directly inside the shroud, will subcool the liquid inside the shroud, leading to the possibili-ty that the water level inside the shroud could be somewhat lower than the level outside the shroud. Since the core is adequately cooled whenever the ECCS is injecting, this level discrepancy is inconsequential.

i Evaluation Analyses presented by the BWR Owners' Group show that the level instrumentation is adequate for predicting the approach to ICC and for providing a basis for

operator action. The staff has found these analyses to be acceptable to support emergency procedure guidelines for ICC.

In an earlier submittal (June 30,1980), the BWR Owners' Group provided a draft of the generic guidelines for BWR's emergency procedures. The guidelines were developed to comply with Task Action Plan Item I.C.1(3) as clarified by NUREG-0737 and to incorporate the requirements for short-term reanalysis of small break loss-of-coolant accidents and ICC (Task Action Plan Items I.C.1(1) and I.C.1(2)). The staff subsequently indicated that the generic guidelines prepared by GE and the BWR Owners' Group were acceptable for trial implementa-tion at WNP-2 (Eisenhut, October 21, 1981). Additional information, requested by the staff, was submitted by the Owners' Group on January 31, 1981. This additional information is still under review prior to the staff's making a final conclusion on the acceptability of the guidelines for implementation on all boiling water reactors.

In Amendment 17 to the FSAR (dated July 1981), the applicant committed to implement a program of emergency operating procedures based on the BWR Owners' Group Guidelines, when approved by the staff, in accordance with the schedule of NUREG-0737. In addition to the BWR Owners' Group efforts, the staff has prepared draft guidelines for longterm up grading of emergency operating procedures (NUREG-0799) in accordance with Task Action Plan Item I.C.9. These guidelines, as revised during the resolution of public comments, also should be '

used in preparation of the WNP-2 emergency operating procedures. -Therefore,-

the staff does not plan to conduct a pilot monitoring review of selected emergency operating procedures in accordance with TMI Task Action Plan Item I.C.8 ,

for this plant. However, the staff will review the applicant's submittal for compliance with the final staff guidelines resulting from resolution of public comments on NUREG-0799. The final staff guidelines will include consideration of the BWR Owners' Group . technical guidelines. The staff's review of the emergency operating procedures submitted by the applicant will be completed prior to issuance of the operating license and will be addressed in Sec-tion 13.5.2 (Operating and Maintenance Procedures) of a supplemental safety evaluation report.

The human factors evaluation of reactor water-level indication will be addressed by the NRC staff specifically as TMI Task Item II.K.3.27. The human factors analysis of the types and locations of displays and alarms in the control room is reported as part of Item I.D.1; .

The applicant had adopted the BWR Owner's Group position that no additional instrumentation was needed to monitor ICC beyond that which already exists in their plant. Since Re~ v ision 2 (December 1980) to Regulatory Guide 1.97 requires BWRs to have incore thermocouples, the WNP-2 applicant has been requested to fully address the broader question of ICC and justify their position regarding incore thermocouple requirements as stipulated in Regulatory Guide 1.97 and in Item (4) of the documentation required by NUREG-0737,Section II.F.2.

Recently, the NRC and representatives from General Electric and the BWR Owner's Group met on December 17, 1981 and January 26, 1982 to discuss the NRC require-ments as specified in SECY 81-582 (October 7, 1981),and the Owners' Group position. As a result of these meetings, agreement has been reached to broaden

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• the issue from the specific requirements for incore thermocouples to that of monitoring ICC. The Owners' Group has agreed to actively participate in the analysis of ICC instrumentation requirements and will be submitting a final report for NRC review in July 1982. The applicant has committed to participate in this study. It is the staff's expectation that the conclusions reached from

. that evaluation will be applied to WNP-2.

The staff has reviewed WNP-2's response to TMI Task Item II.F.2 requirements, which was provided in Amendment 17 to the FSAR of WNP-2 dated July 1981. The staff has concluded that the WNP-2 plant is in partial compliance with TMI Task Item II.F.2. Based on the results of its review, the staff has concluded that the existing level instrumentation and the BWR Owners' Group emergency opera-ting procedure guidelines, which will be based on the ICC instrumentation requirements to be developed and implemented upon completion of the staff's review and approval of .the applicant's report to be submitted in July 1982, will satisfy the requirements of TMI Task Item II.F.2. Accordingly, the operating license of WNP-2 will be conditioned for the submittal of this report by July 1982 and to require conformance with any II.F.2 requirements which result from the staff's evaluation of that report.

4.4.8 Thermal-HydraubicComparison A summary of the thermal-hydraulic parameters for WNP-2 is given in Table 4.4.1.

A comparison with the parameters for the Hatch-2 core design is-given for reference. This core design was previously approved in th,e Safety Evaluation Report issued in June 1978 for Hatch-2 (NUREG-0411), which is now an operating The primary difference in core design between WNP-2 and Hatch-2 is

. reactor.

size. Hatch-2 is a BWR/4 core for which recirculation flow is controlled by the recirculation pump speed. WNP-2 is a BWR/5 core for which recirculation flow is controlled by the recirculation flow control valve position. Both use the improved 8x8R fuel assemblies.

The comparability of WNP-2 design with Hatch-2 supports the conclusion that the WNP-2 thermal-hydraulic design is acceptable.

4.4.9 Evaluation Findings -

The WNP-2 thermal-hydraulic design has been reviewed according to Section 4.4 of the Standard Review Plan (NUREG-0800). The scope of review included the design criteria, implementation of the design criteria as presented by the final '

core design, and the steady-state analysis of the core thermal-hydraulic performance. The review concentrated on the differences between the proposed core design (and criteria) and those designs and criteria that have been

• previously reviewed and found acceptable by the staff. It was found that all such differences were satisfactorily justified by the applicant. The appli-cant's thermal-hydraulic analyses were performed using analytical methods and correlations that have been previously reviewed by the staff and found accept- I able. However, the operating license should be restricted with the following -

conditions:

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Table 4.4.1 Thermal-hydraulic design comparison Design Parameter Hatch-2 WNP22 Design thermal output (MWt) 2,436 3,323 Final feedwater temperature (FFWT) (*F) 420 420 Steam flow rate at FFWT (106 lbs/hr) 10.47 14.30 Core coolant flow rate (106 lbs/hr) 77.0 108.5 Feedwater flow rate (106 lb/hr) 10.44 14.2 Steam pressure, nominal in steam dome (psia) 1,020 1,020 Steam pressure, nominal core design (psia) -

1,035 1,035 Average power density (kW/ liter) 49.15 49.15 Maximum linear thermal output (kW/ft) 13.4 13.4' Average linear thermal output (kW/ft) 5.38 5.4 .

Core total heat transfer area (ft2 ) 54,879 74,871 Fuel type .

P8x8R P8x8R Water rods per bundle 2 2 Core inlet enthalpy a,t FFWT (Btu /lb) 526.9 527.6 Core maximum exit void within assemblies (%) 76.3 76.0 Core average void, active coolant (%) 42.2 41.8 Active coolant flow area per assembly (in2 ) 15.82 15.824 Core average inlet velocity (ft/sec) 6. 6 -

6.88 Total core pressure drop (psia) 23.9 24.74 Core support plate pressure drop (psia) 19.46 20.32 Average orifice pressure drop (psia)

Central region 8.0 6.03 Peripheral region 16.52 16.54 Number of fuel rods per bunale 62 62 Rod outside diameter (in.)

Fuel rod 0.483 .0.483 Water rod 0.591 0.591 Active fuel length (in.) 150 150 Rod pitch (in.) 0.640 0.640 1

1. Single loop operation is not permitted unless supporting analyses are provided and approved.

2. Operation beyond Cycle 1 is not permitted until a stability analysis is provided and approved for the additional cycles of operation.

3. The natural circulation operation m' ode is not permitted.

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4. The' core flow should be checked at least once per day. 1

5. A final report to analyze ICC instrumentation requirements should be submitted by July 1982 for NRC review and approval.. .

. The above restrictions should be incorporated into the proposed Technical Specifications, except for Items 2 and 5 which should be incorporated as license conditions.

With the exceptions noted above, the staff concludes that the thermal-hydraulic design.of the core conforms to the requirements of General Design Criterion 10 of Appendix A to 10 CFR 50 and the guidance of NRC Regulatory Guide 1.133 and Section 4.4 of the Standard Review Plan and is, therefore, acceptable.

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15.4 Rod Withdrawal Events 15.4.1 Continuous Rod Withdrawal Durin_g React 6r Startuo Discussien The rod sequence control system and rod worth minimizer are each designed te enforce a particular rod withdrawal sequence during startup and low power operation.

Follcwing that sequsnce would limit the amount of reactivity that cou.ld be inserted in one withdrawal action to an amount that would preclude any violation of fuel thermal limits (the programmed withdrawal sequence constitutes normal operation during startup). The probability of a failure in these systems that would permit the continuous withdrawal of a high worth rod is low. Nevertheless, the consequences of such an event have Deen calculated.

The caiculation was performed generically by the vendori General Electric

&cepany, and is reported in the La Salle County Station Final Safety Analysis Report (Docket No. 50-341, Section 15.4.1). The calculation was performed in two steps. First a detailed analysis, including three-dimensional effects, was performed for a rod worth (1.6 percent reactivity change) somewhat above the maximum anticipated worth, and then a point kinetics calculation was used to extrapolate the results to red worths to be expected for out of sequence VNP 2 SER  %-

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• rods. The calculation is performed at 1 percent of full power because calculations nave shown that this power level produces the maximum consequences. Transient termination is assumed to occur by means of the APRM 15 percent power level scram or the degraded (worst bypass condition) IRM 4

scram. The withdrawal speed is assumed to be the maximum attainable and.. rod worths up to 2.5 percent reactivity change were analyzed. In no case was a

, peak enthalpy greater than 60 calories per gram encountered. The staff concludes that the analysis of this event is acceptable because a conservative analysis has been performec, conservative rod worth values are analyzed, and the consequences show a large cargin to the acceptance criterion of 170 calories per gram.

1 Evaluation Findinos The possibilities for single failure of the reactor control system which could result in uncontrolled withdrawal of control rods under low power startup conditions have been reviewed. The scope of the review has included inves.tigations of initial conditions and control rod reactivity worths, and the course of the resulting transient. The methods used to determine the peak fuel rod response and the inputs to that analysis have been examined.

The acceptable fuel design limit for this event, required by General Design Criterion 10, is defi'ned to be a peak fuel enthalpy of 170 calories per gram.

This is a departure from SRP Section 15.4.1 where this limit is defined in terms of critical power ratio. However, the power transiant resulting from this event is very narrov (200 millisecond full width at half maximum) so that the maximum fuel enthalpy value of 170 calories per gram, which is used for the d

rod drop analysis (see SRP Section 15.4.9), is an appropriate measure of fuel duty for this event also.

The staff therefore ccncludes that the requirements of GDC 10, 20, and 25 have been met. This conclus}Dn is based on the following. The applicant has met the requirement of GDC 10 that the specified acceptable fuel design limits are not exceeded, GDC 20 that reactivity control systems are automatically initiated so that specified acceptable fuel design limits are not exceeded, and GDC 25 l

that single malfunctions cf the reactivity control system will not cause the

' spe' cified etceptable fuei design limits to be exceeded. These requirements have been met by comparing the resulting extreme operating conditions for the fuel (i.e., fuel duty) with the acceptance criterion (peak fuel enthalpy) to assure tnat fuel rod failure will be precluded for this event. The basis for acceptance in the staff review is that the applicant's analyses of the maximum low power condition have been confirmed, that the analytical methods and input data are conservative, and that crecified acceptable fuel design limits will not be exceeded.

15.4.2 Rod Witharawal Error at Power Discussion .

Above a preset pcwer level (approximately 15 percent of full power) the rod withdrawal sequence is no longer enforced by the rod sequence control system or the rod worth minimizer. Instead the core is protected against exceeding

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fuel damage limits by the rod block monitor. When a Pod is' selected for withdrawal, the nearest four strings of local power range monitors are also WNP 2 SER 15-18  :

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selected. The outputs from these monitors serve as inputs to the rod block monitor, the output of which is the average of the detector signals. This output is then input to a trip circuit which is adjusted to block the rod withdrawal before fuel damage limits are exceeded. An analysis is performed to establish the trip setpoint required to accomplish the proper rod block.

The analysis is performed in a conservative manner, by assuming the highest worth rod in a pattern to be continuously withdrawn at its maximum speed. A rod pattern is selected which tends to maximize the consequences, although such a pattern would be prohibited during normal operation. The ccre is assumed to be operating at rated conditions. The two local detector strings having the highest readings are assumed to be inoperable. The rod to be withdrawn is assumed to be fully inserted prior to its withdrawal. The calculation is performed with the BWR Simulator Code (NED0-20953A-) which has been reviewed and approved (O. Parr, dated September 22,1976). This three-dimensional code

~ is suitable since the power rise is slow enough to permi.t the assumption that the neutron and thermal powers can be calculated by time-independent methods.

The core is assumed to be xenon-free for this calculation in order to maximize the reactivity controlled by control rods. The calculation consists of a

. number of " snapshots" of the core power distribution as the control rod is withdrawn.

The results of the calculations are examined and curves of linear heat generation rates and critical power ratio are drawn as a function of '

withdrawal distance for the rod. Such curves are drawn for assemblies containing the highest linear heat generation rate and lowest critical power ratio during the transient. The curves are used to obtain the maximum travel for the rod which will not violate established heat generation rate or .

critical power ratio limits. The more limiting of the two maximum travel distances is then chosen.

.The calculations are also used to obtain the responses of the local detectors wihch provide inputs to the rod block monitor. These are combined in the appropriate manner to obtain the output of the rod block monitor as a function of rod withdrawal. Appropriate assumptions regarding inoperable local detectors are made. The results of the calculation are plotted and the rod block monitor trip is set to the value obtained for the maximum permitted travel. On the basis that the calculational method used is an approved one and that conservative input assumptions are made, the staff concludes that the analysis of this event is acceptable. .

Evaluation Findinas The possibilities for single failure of the reactor control system which could result in uncontrolled withdrawal of control rods under power operation conditions have been reviewed. The scope of the review has included investigations of initial conditions and control rod reactivity worths, and the course of the resulting transient. The methods used to determine the peak fuel rod response and the inputs to that analysis have been examined. The staff concludes that the requirements of GDC 10, 20, and 25 have been met.

The. applicant has met the requirement of GDC 10 that the specified acceptable fuel design limits are not exceeded for the anticipated transient; of GDC 20 that the reactivity control system is automatically actuated to prevent

exceeding the specified acceptable design limits; and of GDC 25 that single WNP 2 SER --

15-19

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malfunctions in the reactivity control system will not cause specified acceptable fuel design limits to be exceeded. These requirements have been met by comparing the resulting extreme operating conditions and response of the fuel (i.e., fuel duty) with the acceptance criteria for fuel damage (boiling transition and 1 percent plastic strain in the cladding) to assure that fuel rod failure will be precluded for this event. The basis for

. acceptance in the staff review is that the applicant's choice of maximum transients for single error control rod malfunctions has been confirmed, that the analytical methods and input data are conservative, and that specified acceptable fuel design limits will not be exceeded.

15.4.3 Operation With an Imoroperly Loaded Fuel Assembly Discussion Strict administrative controls in the form of previously approved established procedures and inspection of the loaded core are followed during loading to prevent operation with an improperly loaded fuel assembly. Nevertheless, an analysis of the consequences of a loading error has been performed. Two types of loading error are considered- placement of a fuel assembly in an improper location, and misorieptation of a bundle in its proper location. The enrichment distributibn is symmetric and the water gap surrounding the fuel assembly is uniform for the WNP-2 lattice. Thus, the only effect of misorientation in the first cycle is on the fuel assembly local power peaking ~

distribution which must be accounted for in the thermal-hydraulic analysis.

The misoriented bundle is' tilted from the v.ertical and the resultant change is less than 0.05 in the critical power ratio. Normal operation with a misoriented bundle will not cause violation of any fuel damage limits.

The most limiting misloading error occurs at beginning of cycle when a low-enriched bundle is misloaded into the high enrichment location of a four-bundle array surrounding a local power range monitoring detector. The

' reading in this monitor will be reduced and it will be assumed that the readings in the three mirror image four-bundle arrays (which are not instrumented) are the same. If the instrumented four-bundle array is placed on limits, the three mirror image arrays will exceed limits. The misloading error with the maximum calculated consequences for the linear heat generation rate results in an increase of this quantity from the operating limit of 13.4 kilowatts per foot to 14.92 kilowatts per foot in the unmonitored assemblies. This is far below the approximately 20 kilowatts per foot required to produce a 1 percent plastic strain in the cladding. The misloading event with the most serious calculated consequences for critical power ratio results in a minimum critical power ratio reduction of approximately 10.8 percent in the most affected bundle. This amount of reduction would yield an MCPR of 1.08 if the initial operating MCPR were 1.20. The description given above applies only to the first cycle of operation. The fuel misloading event is reanalyzed for each succeeding cycle as part of the cycle loading design.

l Evaluation Findinas The applicant has evaluated the consequences of the misloading of a fuel assembly and has concluded that the most serious misloading would not result in the violation of fuel thermal limits (linear heat generation rate or WNP 2 SER 15-20

s minimum critical power ratio) when the reactor is operating at Technical Specification limits on those quantities. This satisfies the guidance of SRP Section 15.4.7 that' normal operation with any misloading event that cannot be detected by the instrumentation provided in the core not produce offsite consequences greater than a small fraction of 10-CFR 100 guidelines.

15.4.4 The Rod Drop Accident Discussion The postulated rod drop accident occurs when a rod which has been stuck in the upper portion of the-core becomes disconnected from the rod drive, the drive is subsequently withdrawn, and the rod becomes unstuck and falls rapidly onto the drive. This results in a power excursion which could, under certain circumstances, result in local fuel damag~e.

~

The consequences of.the rod drop accident depend chiefly on the amount of reactivity inserted into the core by the dropping rod and on the initial thermal-hydraulic conditions of the core. Dependence on rod drop speed, Doppler feedback coefficient, the shape of the scram curve, and the scram speed is less pronounced. The analysis of the rod drop accident has been performed on a generib basis by GE and is reported in NED0-10527, " Rod Drop Accident Analysis for Large Boiling Water Reactors," and Supplements 1 and 2 to that report. The calculation is performed under the following conservative ~

assumptions:

(1) no thermal-hydraulic feedback is assumed; ,

(2) the least negative Doppler coefficient which'is anticipated is used; (3) the rod drop speed i.s assumed to be that measured for the rod design plus three standard deviations; (4) the scram speed is the Technical Speci.fication value; and (5) the shape of the scram curve is assumed to be that which starts with all rods out of the core. This configuration results in the longest delay before significar,t reactivity is inserted into the core.

In addition, the calcula~tional model contains conservatism. .For example, the axial flux shape is assumed to remain constant throughout the excursion. This means that the energy deposition in the hot pellet is maximized. The enthalpy rise in the hot pellet is plotted as a function of the worth of the dropped rod in NED0-10527, Supplement 1.' For the design calculation described above, a rod worth of approximately 1.4 percent reactivity change is required to produce an enthalpy rise of 280 calories. per gram, which is our acceptance criterion.

To assess the extent of the conservatism in the assumption of no thermal-hydraulic feedback in the design calculations, the staff consultant, Brookhaven National Laboratory, performed a series of calculations which included this effect. The results are reported in _BNL-NUREG-28109, " Thermal-Hydraulic Effects on Center Rod Drop Accidents in a Boiling Water Reactor."

These results show that if thermal-hydraulic feedback is included in the WNP 2 SER 15-21

calculations, the resulting enthalpy rise is less than 140 calories per gram for a rod worth of 1.4 percent reactivity change. Thus it may be concluded that a large conservatism factor exists in the design calculations.

The staff has compared the characteristics of the WNP-2 reactor to that used in the generic rod drop accident analysis. The scram reactivity shape for WNP-2 is conservative with respect to the generic study as is the Doppler coefficient of reactivity. The rod drop speed and scram speed are the same as or slightly conservative with respect to the values used in the generic analysis. The staff thus concludes that the generic analysis is applicable to

. the WNP-2 reactors.

The WNP-2 reactor is provided with a rod worth minimizer and a rod sequence control system to monitor and enforce the sequence of rod withdrawals in the operational range from cold startup to approximately 25 percent of full power. In particular, the banked position withdrawal sequence is enforced.

This sequence has been described in a Topical Report, NEDO-21231, " Banked Posit. ion Withdrawal . Sequence. NEDO-21231 contains a generic analysis of potential dropped rod worths and has been reviewed and approved by the staff (O. Parr, January 18, 1978). In the generic analysis, the feel loading pattern was chosen sof as to enhance rod worths compared to what would be' spected in a real case. The values obtained for potential dropped rod worths in the generic analysis were 0.62 percent reactivity change for first cycle rods in the first 50 percent withdrawn with smaller values for succeeding ^

cycles. During withdrawal.s of the second 50 percent of the rods the maximum worth obtained was 0.75 percent reactivity worth during the first cycle increasing to 0.83 percent in the equilibrium cycle. For comparison the maximum worth calculated for withdrawal of the second 50 percent of rods in the first cycle of WNP-2 was 0.47 percent reactivity change.

Using the 0.83 percent value, the design calculated value ~of the peak enthalpy from NE00-01527 is 135 calories per gram. The calculation including thermal-

' hydraulic feedback would predict less than 75 calories per gram. Neither of these peak enthalpies is expected to produce cladding failure nor a signifi-cant pressure pulse. Nevertheless, for purposes of evaluating environmental consequences, it is assumed that 770 fuel rods suffer cladding failure.

Evaluation Findinos The staff has evaluated the applicant's analysis of the assumed control rod drop accident and finds the assumptions, calculational techniques, and consequences acceptable. Since the calculations predict fuel enthalpies less than 280 calories per gram, prompt fuel rupture with consequent rapid heat transfer to the coolant from finely' dispersed molten U02 was assumed not to occur. The applicant asserts that this is a highly local event with no significant change in core temperature or pressure. We concur with this conclusion and further conclude that Service Limit C (as defined in Section III of the ASME Boiler and Pressure Vessel Code) is not violated. The staff

~

believes that the calculations contain sufficient conservatism, both in the initial assumptions and in the analytical models to ensure that primary system integrity will be maintained.

(

The staff further concludes that the requirements of GDC 28 that the potential i

amount and rate of reactivity. increase be limited in postulated accidents to WNP 2 SER 15 '

- - 6 s: . i +,

i preclude greater than limited local yielding in the pressure boundary and preclude significant impairment of the ability to cool the reactor core has been met.

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l' _ E P 2 SER 15-23_---- _ _.. _ _._.. ._ __ _ -. _ . . _ . _ . . _

APPENDIX B BIBLIOGRAPHY Reactor NUREG-0371, " Task Action Plans for Generic Activities," November 1978.

NUREG-0401, B. L. Siegel and H. H. Hagen, "Euel failure Detection in Operating Reactors," March 1978.

NUREG-0418, R. O. Meyer, C. E. Beyer, and J. C. Voglewede, " Fission Gas Release from Euel at High Burnup," March 1978.

NUREG-0630, D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," Apri) 1980.

NUREG-0633, M. D. Houston, "Euel Performance Annual Report," December 1979.

NUREG/CR-1163, P. J. Pankaskie, P. G. Heasier, and J. C. Wood, "PCI Euel Failure Analysis," December 1979.

GESSAR, " General Electric Standard Safety Analysis Report," Docket No.

STN-50-447, May 1974.

NEDC-20181, "GEGAP-III: A Mode) for the Prediction of Pellet-Cladding Thermal Conductance in BWR Fuel Rods," November 1973.

l NEDC-21084, (Proprietary) " Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibration," November 1975.

NEDC-10802, " Analytical Methods of Plant Iransient Evaluation for the General Electric Boiling Water Reactor," February 1973.

NEDC-10958-P-A, " General Electric BWR Ihcrmal Analysis Basis (GEIAB): Data, Correlation and Design Application," January 1977.

NEDE-20566 " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K, Volume 1, January 1976.

NEDE-20606-P-A, " Creep Collapse Analysis of BWR Euel Using SAFE-COLAPS Model,"

August 1976.

NEDE-20943-P, "Urania-Gadolinia Nuclear fuel Physical and Irradiation Characteristics and Material Properties," January 1977.

L e e-A

NEDE-20944-P, R. J. Grossenbacher, K. W. Holtzclaw, J. C. Rawlings, and H. S.

Stier, "BWR/4 and BWR/5 Euel Design," September 1976.

NEDE-20944-P, "BWR/4 and BWR/5 Euel Design Amendment 1," January 1977.

NEDE-21156, " Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration," January 1976.

NEDE-21175, "BWR Euel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings," November 1976; Amendment 1, April 1977; Amendment 2, September 1977 NEDE-21354-P, "BWR Fuel Channel Mechanical Design and Deflection," September 1976.

NEDE-21660-P, R. B. Elkins, " Experience with BWR Euel Through December 1976,"

July 1977.

NEDE-23542-P, W. G. Jameson, Jr. , "BWR 2-5 Euel Assembly Evaluation of Shipping and Handling Loadings," March 1977.

NEDE-23758, "GESTR, A Model for the Prediction of GE BWR Fuel Rod Ihermal/

Mechanical Performance," March 1978.

NEDE-24011-P-A, " General Electric Generic Reload fuel Application," May 1977.

NEDE-24226-P, K. W. Brayman and K. W. Cook, " Evaluation of Contro) Blade Lifetime With Potential Loss of B 4C," December 1979.

NEDE-24226-1P, " Evaluation of Control Blade Lifetime With Potential Loss of B4 C," Supplement 1, March 1981.

NEDE-24289-P, R. J. Williams, " Assessment of Euel Rod Bowing in GE Boiling Water Reactors," August 1980.

NEDE-24343-P, R. J. Williams, " Experience with BWR Euel Ihrough January 1981,"

May 1981.

NEDE-10505, H. E. Williamson and D. C. Ditmore, " Experience with BWR Euel Through September 1971," May 1972.

NEDE-10527, J. Poane, et al., " Rod Drop Accident Analysis for Large BWR's,"

March 1972; Supplement 1, July 1972; Supplement 2, January 1973.

NED0-20360-1P, " General Electric Boiling Water Reactor Generic Reload Applica-tion for 8x8 Fuel," March 25, 1976.

NED0-20922, R. B. Elkins and H. E. Williamson, " Experience with BWR Euel Through September 1974," June 1974.

NE00-23786-1, " Fuel Rod Prepressurization - Amendment 1," May 1978.

NED0-24609, " Boiling Water Reactor Euel Rod Performance Evaluation Program,"

Eebruary 1979.

WNP 2 SER B-5

r q.. .

3 ,,

NEDS-10456-PC, H. E. Williamson, " Interim Op

'i for Euel Preconditioning," Rev.1, June 1975erating Manageme .

' W. E. Bailey, J. S. Armijo, J. Jacobson formance," Proceedings of ANS To d

Performance in Portland, Oregon,pical Meeting on Light W

. April 29 to May 2,1979, pp 1-10.

-i G. D. Bouchey (WPPSS) letter to A., July Schwencer 21, 1981.

(NRC)

<9 j) Performance Branch Concerns, January 19, 1982. , Related

" Nuclear Durin Project No. 2 Cor one Call January 6, 1982,"

I 1 and ODYN Analysis," JanuaryG. D. Bouchey (WPPSS) 11, 1982. ,

J " Response to RSB Questions 1 R. H. Buchholz 1 Clad Swelling an(d Rupture ,Model," 1981.

May

, 15GE) letter to L.

] " General Electric Euel

) W.

LicensingR. Butler (NRC)

Iopical Report letter to I." Acceptance NE00-10958," , Stewartfor(GE) Referencing i October 1 , 1974.

1 U W. R. Butler (NRC) letter to , I.April Stewart 1975. (GE)

J. S. Charnley (GE) letter to J. D. Coffman (NRC)

December 11, 1979 ,

meeting on Vermont Yankee Fuel " D" Presentation Slides --

ecember 11, 1979.

W. F. Calbert (Detroit Ed.) letter to L William F. Colbert (Detroit Ed.) letter to L. L. Kintner (NRC), May 21, 198 L. DelGeorge (Comm. Ed . L. Kintner (NRC), June 4, 1981.

with LRG working paper) response , 1980.

dated December , 1981, 2 letter R. P. Denise (NRC)

Cladding November Rupture 20, 1979.

memorandum Temperature, Cladding ,

" Summary to R.

Strain J.a Mattson MinutesdofAMeeting on n

ssembly Flow B)ockage,"

H. Denton (NRC) letter to All Operating Nuclear P Lessons Learned Short Ierm Requirements ,

, 1979.

" October 30ower Pla Evaluation Mode)s,"26, NovemberH.

1979. ,

R. Denton (NRC) memorand

" Potential Deficiencies in ECCS Significant In-Core Vibration," .

Marchof 2,

" Modification 1976D. G.

E]iminate D. G. Eisenhut (NRC Reactor Regulation,)"Information Memorandum Nomemorandum Control Blades," October 22, 1979. .

18 -- New Eailure Mode for BWR D. G. Eisenhut (NRC) letter to a)) Operating Light Water B 1979.

eactors, November 9, WNP 2 SER 6-6 N e - eb - w= %

s**eee- w.j.m<.e*

e4 e

e e g y

D. G. Eisenhut (NRC) Jetter to S. 'I. Rogers (BWR Owners Group), October 21, 1981.

R. E. Engel (GE) Jetter to R. Baer (NRC), October 15, 1977.

R. E. Engel October (GE) Jetter to M. Iokar (NRC), " Corrosion Product Contro),"

3, 1980.

R. E. Engel (GE) letter to R. O. Meyer (NRC), August 11, 1981.

E. Garza.o))i, R. Von Jan, and H. Stehle, "Ihe Main Causes of Euel Element Eai]uresErlangen, Review, in WaterGermany Cooled (1978).

Power Reactors," invited paper to IAEA Atomic Energy GE 1979.Projects Division Memorandum, " Vermont Yankee Euel Eailure Status," April R. L. Gridley (GE) letter to 0. G. Eisenhut (NRC), " Evaluation of Potential Eue] Bundle Lift at Operating Reactors," July 11, 1977 J. G. Grund, et al., " Subassembly Iest Program Outline for FY 1969 and EY 1970," INEL Report IN-1313, August 1969.

W. A. Jens and P. A. Lottes, " Analysis of Heat Iransfer, Burnout, Pressure Drop, and Density Data for High Pressure Water," USAEC Report 4627, May 1981.

K. Joon, " Primary Hydride Eailure of Zircaloy Clad Fuel Rods," American Nuclear Society Transactions, Vol. 15, No. 1, June 1972. .

M. A. Miner, " Cumulative Damag'e in Eatigue," Journal of Applied Mechanics, Iransactions of the ASME, 67, 1945.

N. C. Moseley Blades," (NRC) memorandum to B. H. Grier, " Boron Loss from BWR Control November 19, 1979.

W. J. O'Donnell and C. M. Purdy, "The Fatigue Strength of Members Containing Cracks," Journa) of Engineering for Industry, May 1964.

W. J. O'Donnell and 8. F. Langer, "Eatigue Design Basis for Zircaloy Components," Nuclear Science and Engineering Vol. 20, (1964), pp 1-12.

Olan D. Parr (NRC), letter to G. G. Sherwood (GE), May 17, 1979.

D. E. Ross (NRC) letter to G. Sherwood (GE), November 23, 1976.

L. S. Rubenstein (NRC) Memorandum to R. L. Tedesco (NRC), " Resolution of Channel Box Deflection Issue for Near-Ierm BWR OLs," September 18, 1981.

" Standard Specification for Sintered Uranium Dioxide Pellets," ASIM Standard C776-76, Part 45, 1977.

G. Sherwood (GE) letter to D. F. Ross (NRC), December 22, 1976.

R. L. Smith (VYNPC) letter to USNRC, NRR, November 21, 1980.

WNP 2 SER B-7 A

.s -

. . .,, t as .

R. L. Smith (VYNPC) letter to USNRC, NRR, Eebruary 5, 1981.

R. L. Tedesco (NRC) letter to G. Sherwood (GE), Eebruary 4,1981.

R. L. Iedesco (NRC) letter to G. Sherwood (GE), " Conditions Bemoval from Acceptance 1981.

for Referencing of Licensing Topical Report NEDE-24011P," April 16, H. Uhlig, Inc., New"York Corrosion (1971).and Corrosion Control, Second Edition, J. Wiley and Sons, D. B. Waters January (BWR Owners Group) letter to D. G. Eisenhut (NRC), BWROG (8117),

31, 1981. '

O. H. Winne and B. M. Wundt, " Application of the Griffith-Irvin Theory of Crack Propagation to the Bursting Behavior of Disks, Including Analytical and Experi-mental Studies," Transactions of the ASME, Vol. 80, November 1958.

1 i

I WNP 2 SER 8-8