ML20099H354

Forwards Reactor Vessel Integrity Calculation Summary to Address 920703 Small Break Loss of Coolant Event,Which Verifies That No Adverse Effects Resulted from Event

ML20099H354
Person / Time
Site:	Fort Calhoun
Issue date:	08/14/1992
From:	Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-92-285R, NUDOCS 9208190005
Download: ML20099H354 (8)

Text

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s Omaha Public Power District 444 South 16th Street Mall Onnha Nebraska 68102-2247 402/636 2000 August 14, 1992 LIC-92-285R U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Mai' Station PI-137 Washin5 ton, DC 20555 C

Reference:

Docket No. 50-285 Centlemen:

SUBJECT:

Transmittal of Reactor Vessel Integrity Calculatior. Summary for Fort Calhoun Station i

On July 3,1992, Omaha Public Power District (OPPD) experienced a small break loss of coolant event at Fort Calhoun Station Unit No.1. OPPD has performed an analysis of the reactor vessel which verifies that no adverse effects resulted from this event. During the week of July 27, 1992, Dr. J. T. Larkins of the NRC requested that OPPD submit the attached analysis titled " Fort Calhoun Station Reactor Vessel Integrity Calculation Summary."

If you should have any questions, please contact me.

Sincerely, th

W. G. Gates Divisicn Manager Nuclear Operations WGG/3el Attachment c

LeBoeuf, Lamb, Leiby & MacRae J. L. Milhoan, NRC Regional Administrator, Region IV R. P. Hullikin, NRC Senior Resident inspector S. D. Bloom, NRC Acting Project Manager l

J. T. Larkins, NRC Director, Project Directora^e lV-1 bg 9208190005 920814 PDR ADOCK 0500028S R(

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U. S. Nuclear Regulatory Commission LIC-92-285R Attachment Page 1 Fort Calhoun Station Reactor Vessel Int.qgrity Calculation Summary tgg.ust 14. 191, With respect to the July 3,1092, small break loss of coolant event at fort Calhoun Station (FCS), Omaha Public Power District (OPPD) has performed an evaluation related to the establishment of natural circulation and a conservative analysis for stagnant reactor vessel downcomer flow utilizing the REMIX code.

Based upon computer simulations of natural circulation events at other Combustion Engineering (CE) designed plants, flow into the downtomer decreased to about 3%

of full flow at its lowest point following the trip of the final two reactor coolant pumps (at 23:49 on July 3,1992). The ability of fCS to transition into natural circulation following a pump trip is one of the fundamental design bases of the CE reactor coolant-system (RCS). To stagnate flow and prevent the plant from going into natural circu141on during the time frame immediately following a pump i. rip is considered improbable.

Subcooled natural circulation is wified by the operator per the appropriate floating steps in Emergency Operawg Procedure E0P-20, Functional Recovery Procedure, as indicated by: (1) AT (cold leg and hot leg) is less than or equal to 50*F, (2) the difference between core exit thermocouples (CET) and RCS T,,

g is hss than or equal to 10*F, (3) RCS Tso and T lowering and (4) RCS subcooling is greater,than or,,i, temperatures are stable or equal to 20*F. By confirming that all.of the above requirements are met, the operator is assured that natural circulation is present in the RCS.

The operator logged confirmation of natural circulation during the transient at 00:04 on July 4, 1992.

Natural circulation most likely existed earlier due to the -fixed geometry of the RCS, i.e., the position of the core relative to the steam gcnsrator, and the inventory of relatively cold water that exists above the.

core in the cold legs and in the steam generator U-tubes. However, some time is required for development of a stable core AT and subsequent verification of the natural circulation criteria by the operator using E0P-20. Natural circulatton

_provides sufficient mixing to preclude flow - stratification in the cold leg /downcomer regions. Based on data from tht event, OPPD and CE qualitatively concluded that natural circolation was established.

However, since a detailed computer simulation for the July 3 event is not available, a conservative

! assumption of flow stagnation was made to ensurt.1 bounding lower temperature was obtained.

CE performed an evaluation of the thermal-hydraulic transients resulting from the injection of_ high pressure safety injection (HPSI) flow into a -RCS cold leg L

during a period of postulated flow stagnation in the RCS loop.

The resulting-thermal = stratification in the cold leg and the vessel downcomer is of importance for pressurized thermal shock (PTS) related evaluations.

The FCS safety injection (SI) nozzle is at a 75* angle, re11tive to the center line of the cold leg, for_ injection into the cold leg which enhances mixing with the RCS fluid.

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4 U. S. Nuclear Regulatory Commission LIC-92-285R Attachment Page 2 Backaround Info.tmation on REMIX:

In general, transient system codes such as RELAP or RETRAN provide bulk coolant temperatures and not local temperature distributions which may result from thermal stratification phenomena.

These codes assume uniform mixing in each node.

This assumption is valid as lon as forced or natural circulation is present in the loop and downcomer.

Ty ically, natural circulation flows are several times the HPSI flows,tratification would be obtained only duri e flow which woul result in good mixing of the loop flow with the HPSI flow.

Flow s stagnation.

The regional mixing model of the REMIX code calculates the effect of HPSI flow.

stratification in the cold leg and the downcomer during loop flow stagnation.

The models of the REMIX code have been successfully used to interpret test data h(Reference 2) within the parameters experienced at FCS.

The use of REMIX at igher system pressure (1000-1250 psi) and lower HPSI flow rates (ive bounds of 200 gpm total) experienced during the FCS-transient is within the conservat validity of the REMIX code.

Physical' Model of REMIX Code:

The physical situation, modelled under assumed stagnation conditions, is depicted in Figure 1.

The relevant portions of the RCS which participate in the regional mixing include all fluid particles that can reach the injection nozzle through a sequence of horizontal and upward vertical translation.

These include the loop-seal, cold leg, downcomer and the lower plenum volumes.

i Safety injection flow enters the system through the SI line and an equivalent flow rate exits through the reactor core, as dictated by continuity considerations. - For all practical Si rates, a stratified. cold leg configuration is obtained as-shown in figure 1.

A " cold stream" originates with-the injected stream, continues towards both ends of the cold leg and decays away as the resulting plumes fall into the downcomer and pump / loop-seal regions. A " hot stream" flows counter to the " cold stream" sup)1 yin 0 the flow necessary for mixing (entrainment) he HPSI-flow with the "hotat the mixin

' in rigure 1 (MR-1.to MR 5.

Significant mixin of t stream"' occurs right at th)e point of injection MR 1), in MR-3 and MR-5 mixing occurs because of transitions from horizonta layers into falling plumes.

Negligible mixing occurs in MR-2, the interface between the hot and the cold streams in the cold. leg. MR-4 is the region where the downcomer plume finally decays.

F from a practical standpoint the minimum (centerline)f the downcomer plume which,dow is of significance. It is governed by the strength o in turn, is dictated by the extent of the cold leg stratification.

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i U.'S. Nuclear Regulatory Commission LIC-92-285R Attachment Page 3 Event Soecific Input for tittfgilt Code A period of no flow conditicas in the cold legs and downcomer was conservatively assumed to have taken place immediately following the end of coastdown of the second pair of reactor coolant pumps, RC-3A and RC-3C (23:55 on July 3,1992.

This period was assumed to last until natural circulation flow was operational)y l

confirmed at 00:04 on July 4, 1992.

Review of plant data from the Qualified Safety Parameter Display System (QSPDS) with respect -to cold leg temperature transients, reactor coolant pump (RCP) operation and safety injection cycles indicated that cold leg 1A and the reactor vessel downcomer region below the reactor vessel inlet nozzle 1A are the most limiting with respect to the coolant temperature. Assuming the lowest initial coolant -tem downcomer. perature results in-a conservative final fluid temperature in the Accordingly, Cold Leg 1A tem)erature (QSPDS channel-Tall 20) data was used to identify the appropriato init< al coolant tem >erature as input to the REMIX code.

The identified temperature value, 494'F, is ;he lowest-cold leg temperature just 3rior to and during the assumed stagnation period.

For most of this period, 10 wever, the coolant temperature was around 520*F.

Summary of REMIX Results:

=The transient was simulated using the REMIX computer code (Reference 1).

The results for the cases are summarized in Table 1. All cases conservatively assume nine minutes of flow stagnation and a constant average safety injection rate into the stagnated cold ieg, even though actual event data quanti (tativelySI) flow indicates the establishment of natural circulation. The initial loop temperature utiliz6d is 494'F; the ' I flow temperature utilized is 80'F, the safety injection and refueling water tank temperature noted in the control room log.

Case 1 utilizes the lower plenum volume as a ) art of the total mixing volume in the REMIX model.. For additional conservatism ;his mixing volume was modified for Case 2.

Case 2. assumes the lower plenum volume does not participate in the regional mixing process. This assumption is conservative and results in a lower downcomer plume temperature.

Table 1 lists the calculated downcomer plume centerline temperatures. The plume centerline-temperature is at 10.4 feet below the cold leg centerline which is at the top of the most limiting 3-410 reactor vessel (longitudinal) weld. Other key RCS parameters from the event.are also shown in Table 1.

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i U. S. Nuclear Regulatory Commission LIC-92-285R Attachment Page 4 ASME Section XJ Annendix E:

Appendix E of the ASME Code Section XI provides acceptance criteria and guidance for performing an engineering evaluation of the effects of an out-of-limit condition on the structural integrity of the reactor vessel beltline region.

Showing compliance with either Paragraphs E-1200 or E-1300 assures that the beltline region has adequate structural integrity for the unit to return to service.

For thermal transients where AT /At 110 F/hr, Paragraph E-1200 states adequate c

structural integrity of the reactor vessel beltline region is assured if the following criteria are satisfied throughout the event:

1)

T - RT,y > 55'F and c

2)

Maximum pressure does not exceed design pressure.

The minimum temperature conservatively calculated for T in the belt'ine region e

is 363 F as shown in Table 1.

The limiting RT value used in this analysis is 242 F for the 3-410 weld. Therefore, T -RTug-idl F, which is much greater than c

55 F.

The maximum calculated pressure for the transient, including uncertainties, was 2472 psi which was less than the design pressure of 2500 psi.

Thus, it was safe for the reactor vessel to return to and be in service.

To provide an additional measure of assurance that operation of the FCS reactor vessel is safe, the more rigorous approach of evaluating the vessel condition as specified in paragraph E-1300 was evaluated. The coolant temperature, calculated by REMIX, is input in the form of a curve describing the temperature response of the coolant at the wetted surface of the beltline region during the event. The temperature response is used in a heat transfer analysis to provide a detailed temperature profile of the metal temperature through the vessel wall for all time points in the event. The temperature profile is then used to calculate K, for i

the time points. The K result is combined with other mechanical loads, K, and i

i K,, which are also calc,ulated in this process. This sum is compared to Ki, in tbo below listed relationship, to establish the margin to crack initiation throughout the transient.

1.4(K, + K ) + KgsK, i

ig i

if the equation is satisfied for all time points, the criterion is met and the vessel is safe.

The minimum temperature value corresponding to the plume centerline temperature and satisfying this equation is 135 F.

Therefore, since 363 F is much greater that 135 F, the requirements of paragraph E-1300 are also met.

I U.'S. Nuclear Regulatory Commission LIC 92 285R Attachment Page 5 Additional conservatism was also incorporated into the evaluation.

The heat transfer coefficient used to determine the jem vessel wall was assumed to be 1000 BTV/hr f t F.perature distribution through the A more typical value for natural circulation flow would be 300 BTV/hr ft'F.

The temperature profile used the lowest temperature on the centerline of the plume on the entire vessel; the crack depth used in the analysis is 28% greater than the crack depth required to be used-in Appendix E.

The results from the recently completed March 1992 vessel 100% inservice inspection show that an indication even close to the 1.00 inch crack depth required by Appendix E does not exist in the FCS vessel. A smaller crack would increase the margin in E-1300.

References:

1.-

K. Jyer, it.P. Nourbaksh, T.G. Theofanous, " REMIX: A Computer Program for Temperature Transient Due to High Pressure Injection After Interruption of Natural Circulation", NUREG/CR-3701 R2, May 1986, 2.

T.G. Theofnaous, and H. Yan, "A Unified Interpretation of One Fif th to full Scale Thermal Mixing Experiments Related to Pressurized Thermal Shock", NUREG/CR-5677, April 1991.

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U. $. Nuclear Regulatory Commission LIC 92 285R Attachment Page 6 TABLE 1

[alculated Temperatures Nine Minutes After Loop Staanalian Lower Plenum S1 Flow Plume Center Case No, In Model Rate, GPM Line Temp *,*F 1

Yes 100 386 2

No 100 363

• This is the downcomer plume centerline temperature at 10.4 Ft. below the cold leg centerline which is-the top of the limiting longitudinal seam weld

-3 410 The results Itemperatures listed above are for the analysis assumption of nine minutes of. slagnation in )the loop and are based on the following data:

Initial loop temperature - 494* F SI flow temperature-80*F System pressure - 1000 psia

- SI injection in two loops only

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