ML20054D695

Forwards Info Requested to Facilitate Review of FSAR W/Respect to Generating Ser.Info Includes Steam Generator Inventory for Safety Analyses & Fuel in DNB for Partial Pump Locked Rotor Analysis

ML20054D695
Person / Time
Site:	Midland
Issue date:	04/14/1982
From:	Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Harold Denton Office of Nuclear Reactor Regulation
References
16627, NUDOCS 8204230284
Download: ML20054D695 (9)

Text

a O CODSumBIS Power James W Cook GQ8Q vica r,,,isa., - er.iacts. enzi.a.,i s a.1 Co.struction General offices: 1945 West Pernell Rosa, Jackson, MI 49201 * (517) 788 0453 April 14, 1982 l

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NECEtt'EO 9 liarold R Denton, Director d-4PR 22 gg Office of Nuclear Reactor Regulation y,%r US Nuclear Regulatory Commission E

Washington, DC 20555 7,

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MIDLAND PROJECT MIDLAND DOCKET NO 50-329, 50-330 SAFETY EVALUATION REPORT INFORMATION FILE: 0505.16 SERIAL:

16627 ENCLOSURES:

(1) Steam Generator Inventory for Safety Analyses (2) Fuel in DNB for Partial Pump Locked Rotor Analysis (3) Locked Rotor with LOOP at Turbine Trip (4) Performance Characteristics for Midland Units 1 and 2 (5) B&W Memorandum, llot Shutdown-Reactivity Effects with All Rods Inserted (1) through (5) provide ir ormation requested by the NRC Staff to Enclosures r

facilitate their review of the Midland FSAR with respect to generating the Safety Evaluation Report. This information will be incorporated in a forthcoming FSAR revision and will close out the outstanding items involved.

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CC RJCook, Midland Resident Inspector, w/a Rllernan, USNRC, w/a RWiluston, Washington, w/a DBtliller, Midland, w/a a204286 WTM (PjI E

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CONSUMERS POWER COMPANY Midland Units 1 and 2 Docket No 50-329, 50-330 Letter Serial 16627 Dated April 14, 1982 At the request of the Commission and pursuant to the Atomic Energy Act of 1954, and the Energy Reorganization Act of 1974, as amended and the Commission's Rules and Regulations thereunder, Consumers Power Company submits the following responses to outstanding NRC Staff items:

(1) Steam Generator Inventory for Safety Analyses (2) Fuel in DNB for Partial Pump Locked Rotor Analysis (3) Locked Rotor with LOOP at Turbine Trip (4) Performance Characteristics for Midland Units 1 and 2 (5) B&W Memorandum, Hot Shutdown - Reactivity Effects with All Rods Inserted CONSUMERS POWER COMPANY By

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J Cook, Vice" President Proje s, Engineering and Construction Swor and subscribed befor me this 19L day of April 19Jh3

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Notary Publi Jackson County, Michigan My Commission Expires September 8, 1984 i

miO482-0866d131

Enclosure (1)

STEAM GENERATOR INVENTORY FOR SAFETY ANALYSES The steam generator inventory assumed as an initial condition for a secondary system upset exerts a large influence on the consequences of the transient.

Early steam generator inventory predictions for low and high power 177 operating plants of 55,000 and 62,500 lb, respectively, were used for double-ended steam line break analyses. These inventory predictions are a conserva-tively large estimation of the maximum amount of mass that could be present in the steam generator tube and downcomer region that would still result in a guaranteed minimum superheat production of 35* F during rated power operation with fouled steam generators. These predictions are based on consideration of actual steam generator geometry and the matching of tube region temperature distributions with an Alliance Research Center test steam generator model in order to obtain a profile of fluid density along the steam generator tube region.

These inventory predictions have, over the past several years, been refined to more accurately reflect true generator inventory. Operating experience at ANO-1 has indicated a very slight degradation in steam generator exit steam temperature as a function of fouling which was less than originally predicted.

This results in a lower calculated inventory for fouled conditions than was previously considered necessary.

Steam generator inventory prediction with clean conditions has become neces-sary with the advent of loss of feedwater calculations.

In general, core overheating transients are made more severe in RC pressure response by using a low prediction of secondary inventory.

A conservative approach to steam generator inventory as an initial condition for transient analyses is still ensured by using a nominal value +10% in the direction of known conservatism.

miO482-0866a131

l Enclosure (2)

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FUEL IN DNB FOR PARTIAL PUMP LUCKED ROTOR ANALYSIS The analysis of the locked reactor coolant pump rotor from an initial 2 pump (1/1) operation at 1251 MWt conservatively predicts that 27% of the core will experience a minimum DNBR below 1.3 (BAW-2). A calculation of maximum cladding surface temperature indicates a maximum surface temperature of about 1100* F a few seconds after film boiling is predicted. This surface tempera-ture is considered to be insufficient to initiate appreciable Zr-H 0 reaction 9

or result in clad damage by melting. The maximum centerline fuel temperature is calculated to increase less than 25 F above the nominal of 2250 F for operation at 50% power.

In view of the small change in transient power and the fuel temperature excursion predicted for this transient, the potential for fuel cladding failure due to a rupture is negligible.

Internal fuel pin pressure does not significantly increase as a result of the fuel temperature change. Therefore, rupture at low cladding temperature is precluded.

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miO482-0866a131

Enclosure (3)

LOCKED ROTOR WITH LOOP AT TURBINE TRIP The reactor coolant pump shaft seizure event is a non-mechanistically postu-lated design basis event. Following the seizure of one RC pump rotor, forced RC flow is assumed to continue using the unaffected pump (s). The transient results in reactor trip followed shortly thereafter by turbine trip.

The consequences of a loss of offsite power during the transient are con-sidered to have an inconsequential impact on the film boiling predictions. A loss of offsite power at the initiation of the event would result in an early reactor trip thus improving the DNB response. A loss of forced reactor coolant flow initiated at the time of turbine trip following reactor trip results in approximately an additional 2% flow degradation in the second following initiation. The RC flow degradation in this time after reactor trip has an inconsequential impact on the prediction of the DNB response.

l miO482-0366bl31 t

Enclosure (4)

PERFORMANCE CHARACTERISTICS FOR MIDLAND UNITS 1 AND 2 Parameters 4-Pump 3-Pump 2-Pump Design Power, % of 2452 MWt 100 72 50 Design Overpower, % of 2452 MWt 112 90.2 62.5 System Pressure, psia 2200 2200 2200 DNBR Design Power 2.36 Design Overpower 1.95 1.85 2.49 Minimum DNBR 1.30 1.30 1.30 Critical lleat Flux Correlation BAW-2 BAW-2 BAW-2 Coolant Flow 6

Total Flow Rate (10 lb/hr) 131.2 98.2 65.0 Average Velocity Along Fuel Rods (fps) 15.5 11.6 7.7 Coolant Temperature, F

Vessel Inlet 555.2 555.2 555 Vessel Outlet 602.8 602.8 603 lleat Transfer 2

Total Heat Transfer Surface (ft )

49,130 49,130 49,130 Average Heat Flux (BTU /hr-ft )

166,000 124,500 83,000 Maximum Heat Flux (BTU /hr-ft )

443,000 332,400 221,600 Average Linear Heat Rate (kw/ft) 5.47 4.10 2.73 l

miO482-0866c131

Babc ck&WilCSX Nuciocr Powcr Gsnsraticn Division Enclosure (5)

To J.D. Agar, Project Engineering F.J. Levandoski, NSS Product Line swwr 2ossw.sn

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620-0012,13 0.L. Smith, Operational Analysis O/ h T3.38, T3.81 Hot Shutdown - Reactivity Effects with all Rods Inserted April 12,1982 RCS boration via the Emergency Boration System (EBS) will not be necessary for'the Midland plants for the achievement and maintenance of hot shutdown provided all the control rods are inserted.

This is valid for both the first and second fuel cycles (where later cycles will be analyzed as the necessary nuclear input becomes available). The attached figures provide the necessary justification of the.above statement.

1.

Figure 1 shows that the Hot Full Power (HFP) critical boron concentration.

is in excess of that required for hot shutdown (with no Xenon,1%

subcriticality, all rods inserted) for the first fuel cycle. Thus, no RCS boration will be necessary for the achievement and maintenance of hot shutdown for this first cycle (provided no major injection o?

de-borated water occurs after trip).

2.

Similarly, Figure 2 shows that no RCS boration will be necessary for the second fuel cycle.

Reference:

B&W Document 86-1128114-00 "RCS Boration with all Rods Inserted," N. Smith.

QA' Statement Infomation transmitted in this memo is applicable to the first and second fuel cycle only for Midland 1 and 2 plants.

This infomation, approved ~.

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