ML20101U395

10CFR50.46 Annual Rept - ECCS Evaluation Model Revisions, for June 1991 to June 1992

ML20101U395
Person / Time
Site:	Callaway
Issue date:	06/30/1992
From:	Schnell D UNION ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ULNRC-2664, NUDOCS 9207220414
Download: ML20101U395 (12)

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1901 Otoute;ta An've Po:1 O'1rw F::t 149 St iout h% cud &31f t; 314SS4lE50 4

" "",9;N"~"'L E.LECTR,lC

l. INION J uly 16, 1992 Uh

?E U.S. Nuclear Regulatory Ccmmission

- ATTN: Document Control Desk

- Mail Station P1-137 Washington, D.C. 20555 Gentlemen: ULNRC-2664 DOCKET NUMBER 50-483 CALLAWAY PLANT 10CFR50.46 ANNUAL REPORT-ECCS EVALUATION MO_REL REVISIONS i

References:

1. ULNRC-2439 dated 7-19-91

2. ULNRC-2373 dated 2-28-91

3. ULNRC-2141 dated 1-19-90

4. ULNRC-2535 dated 12-18-91 Attachment I to this letter describes changes to Westinghouse ECCS Evaluation Models, 10CFR50.59 safety evaluations, and LOCA-related margin allocations which have been implemented for Callaway for the time period from-June l 1991 to June 1992. Attachment 2 prevides s an ECCS Evaluation Model Margin Assessment which accounts for the peak' cladding temperature (PCT) changes resulting from the resolution of the issues described in Attachment 1 as they apply to Callaway.

References 1-3 above transmitted prior 10CFR50.46 reports.

Attachment 1 describes the resolution of those issues which have been implemented for Callaway. In some cases this results in peak cladding temperature (PCT) margin allocations. The margin allocations for Callaway are identified in Attachment 2. Sinco the PCT values determined in the large and small break LOCA analyses of record, when combined with all permanent and temporary PCT margin allocations, remain less than the 2200*F r'gulatory limit, no reanalyses will be performed.

Should you have any questions regarding this letter, please contact us.

Very truly yo +3, FT

': i G 0 u d g ,

9207220414 920630 Donald F. Schnell gDR ADOCK 05000483 PDR GGY/kea l Attachments . $ \;

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cc: T. A. Baxter, Esq.

Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.

Washington, D.C. 20037 Dr. J. O. Cermak CFA, Inc.

18225-A Flower Hill Way Gaithersburg, MD 20879-5334 R. C. Knop

-Chief, Reactor Project Branch 1 -

U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road .

Glen Ellyn, Illinois 60137 Bruce Bartlett Callaway Resident Office U.S. Nuclear Regulatory Commission RR#1 Steedman, Missouri 65077 L. R. Wharton (2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852 .4 Manager, Electric Department -

Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102

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1 ULNRC 2664 4 i

1 ATTACHMENT ONE >

i L CHANGES AFFECTING CALLAWAY LARGE AND l f

I SMALL BREAK LOCA PCT VALUES i ir

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, 1. HOL ROD INTERNAL PRESSURF ASSUMPTION

.ga Break LOCA (LBLOCA) analyses are performed at near ff oeginning of life (BOL) fuel rod conditions, which have been shown to be limiting in sensitivity studies. The fuel rod f*??"

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performance utilized in the analyses corresponds to rod

. internal pressure (RIP) calculations using NRC approved PAD 74P models which provide a value for RIP that contains some

'M conservatism in its calculation. Higher RIP typically results in ea*11er and greater rod burst and blockage and ultimately a PCI penalty. Recently, questi ns were raised icerning the calculation of BOL RIP unce. ainties which

,g ntribute to ti.^ upper bcund BOL RIP utilized in the LBLOCA p welysis. Evaluation of the issue determined that a jpg sunding 65 pa3 increase to the already conPervative upper

, aound BOL RIP applies. Sensitivities to BOL RIP r/e been quantified which indicate tnat a penalty of 2 F should be

_1sessed for LBLOCA analyses using the 1981 ECCS Evaluation

.odel (EM) with BASE, as used at Callaway.

2. IFJA NON-LIMITING FUEL ROD ASSUMPTICE Recent revisions to the cladding swelling and rupture models in the Westinghouse ECCS ems raised questions whether higher PCTs could result for fuel assemblies with Integral Fuel .

Burnable Absorbers (IFBAs). Previous sensitivity studies py, demonstrating that IFBA fuel was not limiting were N'

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questioned due to the unknown impact on those studies by the '

re risions to the modeling of cladding swelling and rupture behavi "cace some Callaway fuel assemblies contain IFBAs, -ed'..onal analyses ~4re perfacmed for the Cycle 6 core recencly loaded. Tha 1FBA fuel rods were shown, by analysis, to be lens limiting than the non-IFBA fuel assumed in the 2.BLOCA analysis documented in the Callaway FSAR.

This is due to an inherent power density reduction caused by the neutron poisoning and flux depression of th. absorver.

IFLA power density reduction input to the ECCS Evaluation Model has been verified to be conservative relative to the Callaway Cycle 6 core. The IFBA futt is bounded by the non-IFBA fuel assumed in the LBLOCA analysis documented in the FSAR. Thus, operation of Cycle 6 with IFBA fuel meets the requirements of 10CFR50.46 and Appendix K to 10CFR50 and no PCT penalty ic assessed for either LBLOCA or SBLOC'.

3. STEAM GENERATOR FLOW AREA - SEISMIC /LOCA TUBE COMA"0E The issue of steam generator tube collapse under LOCA, seismic, or combined seismic and LOCA loads has been re-evaluated for Callaway due to the recults of structural integricy analyses for several steam generator types.

Callaway had been asseesed a PCT penalty of 0 F in the 1991 10CFR50.46 report (Reference 1) which is now increased to id.6 F. However, a plant-specific analysis has recently been completed for Wolf Creek with Model F steam generators

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which, taking credit for " Leak Before Break" (LBB) methodology, has been shown to result in no steam generator tube collapse for the double-ended cold leg guillotine (DECLG) break. Similar results would be expected for Callaway.

Originally, the structural integrity analysis results for Callaway from WCAP-10043 (submitted by SLNRC 82-0047 dated 12-3-82) showed tnat 0% tube collaps esulted from seismic forces (SSB) and 6.2% from SSE plus Lv 1 for a double-ended pump suction (DEPS) break at the steam generator outlet. Po data was available for the DECLG breaks but it wac postulated that lower loads would result with only minor llk tube collapse expected. At the time it was believed that only the DEPS break resulted in tube collapse due to the _

c7.ose proximity to the steam generators. Further, the PCT penalty associated with the 6.2% area reduction for the DEPS b_eak was bounded by the PCT margin (approximately 700*F) between it and the DECLG break. Therefore, no penalty was assessed in the 1991 10CFR50.46 report (Reference 1).

Recently, analyses have been performed for several steam generator types which show that the loads are not significantly reduced for the DECLG break and tube collapse may occur. However, the DEPS break remains the more limiting break location with respect to tube collapse. As a

. result, the 6.2% flow area reduction from WCAP-10043 will be

1. 1. used as a conservative upper bound for the DECLG break.

With respect to PCT, conservative sensitivities of 3*F per percent steam generator tube plugging (%SGTP) have been documented for the BASH Evaluation Model. This results in a LBLOCA PCT penalty of 18.6 F, based on the 6.2% flow area reduction from SSE plus LOCA loads for a DEPS break. .

4. ECCS FLOW TECH. SPEC. CHANGE The limiting break for the Callaway licensing basis LBLOCA analysis of record is that with a discharge coefficient ;Co) of C.6 (minimum safeguards assumptions) with a PCT of 2014"F. Assessments for changes made to the 1981 Evaluation Model with DASH, as well as other changes per Attachment 2, result in PCT penalties af 51.7 F , 203F of which were reported previously in References 1 and 3, for a cumulative PCT of 2065.7'F.

Minimum and maximum safeguards assumptions were examined with respect to the revised CCP and SIP flows discussed in Reference 4. No changes to the RHR flow were made. For the case of maximum safeguards, total flow decreases by a negligible amount such that the current ana' lysis remains limiting. For the case c minimum safeguards, reductions in flow result in a 7.5 F Ptr penalty. This penalty is offset by taking credit for RHR delivery against a containment backpressure of 2." psig, which providos a 7.5*F PCT benefit, instead of 0.0 psig as assumed in the analysis of record. This results in a 7.5*F LBLOCA PCT penalty being offset by a 7.5*F PCT benefit for a net change of 0*F (absolute change in ?Cf of 15*F) .

3 The current limicing treak size for the Callaway licensing-basis SBLOCA analysis is the 4-inch equivalent diameter with a PCT of 1528'F. Assessments for changes made to the NOTRUMP Evaluation ModC result in PCT penalties of 306.1 i, 306*F of which were reported previously in References 1 and 3, for a cumulative PCT of 1934.1*F.

The method of evaluation for the safety in a ' tion (SI) shortfall from Reference 4 involves the calculation of the integrated SI reduction from SI initiation to PCT time based on .avised flow rates. Using the revised SI flow rate at

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PCT time, the time to compensate for the lower integrated flow was determined. This time is assumed to coincide with an additional period of fuel cladding heat up.

The revised flows exceed the currently analyzed flows over an intermediate pressure range only, while experiencing a decrease at both the high and low pressures. As a result, the Callaway SBLOCA analyses were evaluated to ensure no shift in the break spectrum.

SI flow rates were higher for the major portion of the

c. limiting 4-inch diameter break transient, with an overall increase in integrated SI. Thel fore, no PCT penalty was assessed fo the 4-inch break.

SI reductions at high pressures tend to shift the limiting break to smaller sizes since these breaks depressurize more _

slowly. As a result, the 3-inch break was also examined since the flow rates at the higher pressures were being adversely affected. The 3-inch break demonstrated a similar increase in integrated SI but to a slightly lesser extent such that no PCT penalty was assessed and no shift in the

' break spectrum is anticipated.

5. CONTAINMENT PURGE EVALUATION A 10'F LBLOCA PCT penalty is assigned due to two containment mini-purge lines being in operation (for up to 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year per Technical Specification 3.6.1.7.b), each with

,e 18,-inch butterfly isolation valves inside and outside f' containment. These valves allow a volume of the post-LOCA atmosphere to escape resulting in a lower containment pressure transient. Lower containment pressure adversely affects the core flooding rate calculation for LBLOCA resulting in increased PCT. Th^ effect of these mini-purge valves was not incorporated into the current licensing basis LBLOCA analysis of record. Based on a pressure reduction of l

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- approximately'O.2-psi,-sensitivities were used to assign a 10*F' penalty for this-issue.

6..' CYCLE 6 FUEL RECONSTITUTION

' In: order to determine-the effect of reconstitution of

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assembly G87 in the. Cycle 6 core on LBLOCA PCT, the increase

'in; core average power-due to the presence of two non-power producing stainless steel' filler rods must be considered.

This results in_a very slight increase in peak Kw/ft and Ltotal'Fq .. Using Westinghouse' internal sensitivities for-F q, a PCT' increase of less tLan 0.1*F would' result. For the purpose of_ reporting:Tnd tracking, this value will be conservativelycrounded up-to 0.1*F. In addition, a

- reduction in power =of an assembly reconstituted with filler rods.would increase the water density up_the channel and result inLincreased cross. flow from neighboring channels and

- assemblies. This effect on PCT has been determined tc be less than a 1*F-increase. 'This.will be conservatively <

treated as.a l'F increase. Thes3-increases (1.1*F), when

.added to the licensing basis PCT of 2064.6*F, result in a O

= LBLOCA PCT of 2065.7*F,'which continues to meet the

, acceptance criteria of 10CFR50.46.

The effect:of reconstitution of assembly G87 on the SBLOCA PCT was-obtained by determining the increase-in the clad heat-up rate-during core uncovering caused by the increase in peak-power of the hot rod. This was determined to be an increase-of less than 0.1*F. For the purpose of reporting

- and1 tracking, this'value will be rounded up to 0.1 F.. This increase,3whenLadded~to-the licensing basis PCT of 1834.0*F, results.in a SBLOCA PCT of 1834.1*F,.which continues to meet the acceptanco1 criteria of-10CFR50.46.

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ATTACHMENT TWO

-ECCS EVALUATION MODEL h

8 MARGIN ASSESSMENT FOR CALLAWAY ,

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LARGE BREAK LOCA A'.' t 'AL13IS OF RECORD

- PCT = 2014*F B. 1989 LOCA MODEL' ASSESSMENTS + 10*F

-(refer'to ULNRC-2141 dated 1-19-90)

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C. 1990'LOCA MODEL ASSESSMENTS + 0*F (refer to ULNRC-2373 dated 2-28-91) 1 D' . 1991 LOCA MODEL ASSESSMENTS + 1**F J: (refer to'ULNRC-2439 dated 7-19-91.)

E. CURRENT LOCA MODEL ASSESSMENTS - JUNE 1992 1.-BOL ROD INTERNAL PRESSURE ASSUMPTION + 2F l -(refer to Item 11 of Attachment 1) 1

-2. IFBA NON-LIMITING FUEL ROD ASSUMPTION + 0F '

. (refer to Item 2 of Attachment 1)

F. OTHER LOCA-RELATED' MARGIN ALLCCATIONS - JUNE 1992

1. SG FLOW AREA - SEISMIC /LOCA TUBE COLLAPSE + 18.6 F

-(refer.to. Item 3 of Attachment 1)

G. 10CFR50.59 SAFETY EVALUATIONS - JUNE 1992

1. ECCS FLOW TECH. SPEC. CHANGE + 0*F q (refer to Item 4 of Attachment 1) i

a. Reduction in SI and CCP flow +7.5*F

11. Increased RHR, taking credit for -7.5'F h 2.7;psig containment backpressure: 0*F

'2.-. CONTAINMENT PURGE EVALUATION

+ 10 F (refer to Item 5 of Attachment 1)

3. CYCLE-6 FUEL RECONSTITUTION + 1.1*F

- (refer to Item 6 of Attachment 1)

H. CURRENT LOCA~MODEL ISSUES - JUNE 1992

1. POWER DISTRIBUTION ASSUMPTION + 0*F LICENSING BASIS PCT +. MARGIN ALLOCATIONS = 2065.7'F NOTES: i

1. The 19911 assessments-included penalties of +10 F for fuel rod model; revisions and 0 F for burst and blockage. The O'F and +100 F penalties reported in 1991 for SG flow gb . area and LBLOCA power distribution assumption, respectively, are revised in this report per Items F.1 and H.~1'above.

1

. -LARGE BREAK LOCA NOTES (cont.)

2.-This penalty applies only as long as reconstituted fuel assembly G87 is in the core.

3. This is a Cycle 6 assessment only. The Westinghouse Power Shape Sensitivity Model (PSSM), discussed in WCAP-12935

-(May 1991), Westinghouse ECCS Evaluation Model: Revised Large Break LOCA Power Distribution Methodology, was used-to ensure that the chopped cosine power distribution remains limiting for Cycle 6. The PSSM, currently under -

NRC review, will also be used to msure that the chopped

-cosine power distribution remains limiting for future reloads, i

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SMALL BREAK LOCA A. ANALYSIS OF RECORD PCT = 1528'F B. 1989-LOCA MODEL ASSESSMENTS +229 F (refer to ULNRC-2141 dated 1-19-90)

C. 1990-LOCA MODEL ASSESSMENTS + 0*F (refer to ULNRC-2373 dated 2-28-91) 1 D.'1991 LOCA MODEL ASSESSMENTS + 77*F #

(refer-to ULNRC-'2439 dated 7-19-91)

E. CURRENT LOCA MODEL ASSESSMENTS - JUNE 1992

1. IFBA NON-LIMITING FUEL ROD ASSUMPTION + 0'F (refer to Item 2 of-Attachment 1)

F. 10CFR50.59 SAFETY EVALUATIONS - JUNE 1992

1. ECCS FLOW TECH. SPEC. CHANGE + 0'F

-(refer to Item 4 of Attachment 1)

2. CYCLE 6: FUEL RECONSTITUTION- +0.1*F

!(refer-to Item 6 of Attachment 1)

LICENSING-BASIS PCT + MARGIN ALLOCATION = 1834.1*F NOTES:

1. The 1991 assessments included penalties of +37'F for fuel

-rod:model revisions, O F for NOTRUMPfcode solution convergence,_.+40*F~for SBLOCA: rod internal pressure assumption, and O'F;for SBLOCA broken loop SI flow assumption. The.SBLOCA rod internal-pressure' penalty of

+40'F was composed of two individual,' aspects._For SBLOCA analyses the limiting rod internal pressure (RIP) assumption depends on whether burst is predicted to-occur. A-higher RIP may' lead to a higher _ calculated PCT if burst io predicted to occur. Conversely, a lower RIP

+ rmly decrease -cladding creep -(rod -swell) away from the 1 fuel pellets when the fuel _ rod internal pressure is greater than the RCS pressure._Therefore, a lower RIP

-could then result in a. higher calculated PCT, since thu

-cladding would be closer.to the fuel pellet, for an analysis that did_not predict fuel rod burst. Rod burst c is-not predicted to occur in the Callaway SBLOCA analysis of record - (see FSAR Table 15.6-15) . A 20*F PCT penalty was assessed in:1991 to account for this-effect. This issue-also involve 9 an error in the cladding strain model assumed in the small break clad heatup calculation for which another +20*F PCT penalty was assessed in 1991, for a total _of +40*F as reported in ULNRC-2439. Since the

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SMALL BREAK LOCA NOTES (cont.):

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1. time of'the 1991 report, a related issue, BOL Rod Internal Prescure Uncertainty, was opened for non-IFBA

- fuell' rods. Using a conservative combination of BOL uncertainties results in an estimated decrease of up to 65-psi =in'the predicted BOL RIP.2 Based on sensitivity-analyses, a PCT penalty of +20*F was assessed. Final resolution of the rod internal pressure issue, as-

-reported herein, incorporates the rod internal pressure portion of the original-issue but not the cladding strain model error. As-such, the original +40*F PCT penalty was

. reduced.to +20*F with +20*F being reallocated for the ,

l uncertainty issue for a total of +40 F. Therefore, the L . total penalty reported in 1991 stays the same.

.2. This penalty applies-only as long as reconstituted fuel assembly G87 is in.the core.

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