ML17138B503

Forwards Request for Addl Info on Reactor Sys for NRC Review of Application for OLs

ML17138B503
Person / Time
Site:	Susquehanna
Issue date:	09/19/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
References
NUDOCS 8010020119
Download: ML17138B503 (12)

Text

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. 95F Mr. Norman W. Curtis Vfce Presfdent - Engfneerfng and Constructfon Pennsylvanfa Power and Lfght Company 2 North Nfnth Street Al'lent', Pennsylvania 13101

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SUBJECT:

SUSQUEHANNA STElN ELECTRIC STATION~ UHITS HOS.

1 AND 2 >> RE(VEST FOR ADDITIONAL INFOlNATION As e result of our revfew of your application for operating licenses for the Susquehanna Steam Electrfc Plant, we find that we need addftfonal information fn the area of Reactor Systems.

The specfffc fnformatfon required fs lfsted fn the Enclosure.

?f you desfre any dfscussfon or clarfffcatfon of the fnformatfon re requested, p'lease contact R. H. Stark, Lfcensfng Prospect

Manager, (301-492-7238).

Sfncerely, Robert L. Tedesco, Assistant 0frector for Licensing Dfvfsfon of Lfcensfng Enclousure:

As stated cc:

See next page DL:...81.......

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MYoungb)ood" RLTedesco.-"-

'FFICE SURNAME DATE$

NRC FORM 318 (9 76) NRCM.0240 U.S. GOVERNMENT PRINTING OFFICE: 1979.289 369 ~

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Hr. Norman W. Curtis Vice President Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsyl vania 18101 CC Hr. Earle M. Mead Project Engineering Manager Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsyl vania 18101 Jay Silberg, Esq.

Shaw, Pi ttman, Potts Trowbridge 1800 M Street, N.

W.

Washington, D. C.

20036 Mr. William E. Barberich, Nuclear Licensing Group'upervisor Pennsylvania Power

& Light. Company 2 North Ninth Street Al I entown, Pennsyl vania 18101 Edward M. >>agel, Esquire General Counsel and Secretary Pennsylvania Power

& Light Company 2 North Ninth Street Al I entown, Pennsyl vani a 18101 Bryan Snapp, Esq.

Pennsylvania Power

& Light Company 2 North,"tinth Street AI I ento>a, Pennsyl vania 18101 Robert M. Gallo Resident Inspector P. 0.

Box 52 Shickshinny, Pennsyl vani a 18655 Susquehanna Environmental Advocates c/o Gera'Id Schultz, Esq.

500 South River Street Wilkes-Barre, PA 18702 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation Bldg. 3500, P. 0.

Box X

Oak Ridge, Tennessee 37830 Nr.

E.

B. Poser Project Engineer Bechtel Power Corporation P, 0.

Box 3965 San Francisco, California 94119 Matias F. Travieso-Diaz, Esq.

Shaw, Pittman, Potts Trowbridge 1800 M Street, N.

W.

Washington, D. C.

20036 Dr.. Judith H. Johnsrud Co-Director Environmental Coalition on Nuclear Power 433 Orlando Avenue State

College, PA

16801, Mr. Thomas H. Gerusky, Director Bureau of Radiation Protection Department of Environmental Resources Commonweal th of Pennsyl vania

,P. 0, Box 2063 Harrisburg, PA 17120 Hs. Colleen L'larsh Box 538A, RD84 Mountain Top, PA 18707 Mrs. Irene Lemanowicz, Chairperson The Citizens Against Nuclear Dangers P. 0.

Box 377 ROD I

Berwick, PA 18503 Mr. J.

M. tlillard Project Manager Mail Code 394 General Electric Comoany 175 Curtner Avenue San Jose, California 95125

0 211

~ 270 (15.2.7.2.1)

I Review of the "loss of all feedwater flow" transient indicates the the feedwater flow decreases to zero in 5 seconds.

For the analyses presented in the FSARs indicated below, the reactor vessel water level decreases to the L3 scran trip setpojyC ay folio~;

FSAR Evsquehanna Fermi-2 Grand Gulf WNP-2 Time at which L3 trip occurs, sec 6.8

7. 36 Vessel ID, in./no. of Rated

Power, fuel assemblies MWt 251/764 3293 251/800 3833 251/764 3323 Based on power level and vessel size only, the L3 setpoint for Susquehanna should be attained at approximately the same time as Fermi-2 and MNP-2.

Explain why there is a difference between the times that the L3 trip is attained for Susquehanna and the other reactors.

Include appropr iate design considerations (differences in piping design, level setpoints, etc) in the response.

'211.

271 (15. 0)

For each transient and accident analyzed in Chapter 15, identify each normally operating system for which credit has been taken.

Q211. 272

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(15.2. 1-2. 3)

Q211. 273 (15.2.9.2. 1)

It is indicated that the "pressure regulator-closed" transient with failure of the backup pressure regulator is less severe than the "turbine trip with bypass" transient in Section 15.2.3.

This agrees with GESSAR 238.

As a result, only a qualitative evaluation of the transient was provided.

However, quantitative results frcm the Grand Gulf FSAR indicate the opposite.

The staff' concern is that quantitative results for this transient may be similar to those for Grand Gulf.

Provide a quantitative analysis of the "pressur e regulator-closed" transient assuning failure of the backup pressure regulator and revise Section 15.2.1.2.3 accordingly.

'Ihis request should be coordinated with the information requested in Q211.160 for the case where the backup pressure regulator regains control.

Revise Table 15.2-15 to indicate the time that suppression pool alarms are received, the Technical Specification limit >s

exceeded, and the maximun value of the suppression pool temper ature is attained.

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~~211. 274

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I Specific input parameters for the models used to evaluate blowdown rate and suppression pool temperature. are shown in Table 15.2-16 along with the analytical results in Figures 15.2-12 and -13.

In connection with this, provide the following information:

a)

Identify the analytical models used to evaluate blowdown rate and suppression pool temperature.

b)

Provide justification that the input, parameters used are con-servative, by reference to approved topical reports, or other documentation.

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.~211. 276

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The response to Q211.122 indicates that studies show use of a 40 F HPCI temperature is conservative.

Provide a reference to these studies.

Q211. 276 (15.5. 1.2. 3)

Q211. 277 (15.0)

Q211.278 (15.6.5.2;.1)

From the discussion of single failures for the "inadvertent tlPCI startup" transient, it is indicated that a single failure of. the pressure regulator or level control will aggravate the transient, resulting in reduced thermal margins.

Provide the MCPR and peak vessel pressure values that result for this event with the most limiting of the above single failures considered in the analysis.

The response to Question 221.3 indicates that 8 x 8 fuel bundles with two w'ater rods will be used at Susquehanna instead of the 8 x 8 fuel bundles with one water rod.

a)

Have the transients and accidents in Chapter 15 been evaluated with 8 x 8 fuel bundles using one or two water ro'de b) If the transients and accidents in Chapter 15 were analyzed with the one water rod fuel bundles, would any significant changes in MCPR, peak vessel

pressure, percent of rods experiencing boiling transition, and the radiological consequences be expected if the two water rods design was used in the analyses?

Discuss any changes to the above event parameters in quantitative terms.

In the description of'event sequences for LOCA inside containment, confirm that the zero reference time for Tables 6.3-1 and 6.2-8 are the same.

Q211. 279 (15.6.5)

'Ihe thermal power of 3493 YWt used in Chapter 15 analyses (Table 15.0-2) is indicated as 104.4$

NBR.

For LOCA calculations inside containment, the thermal power value of 3434 in Tables 6.2-4 and 6.3-2 is indicated as a design overpower of 105$.

Explain the discrepancy in the thermal power values specified.

Q211.280 In connection with parameters and assumptions used for LOCA (15.6.5.3.2) calculations inside containment, provide:

a)

The specific value of KSIV closure time and justification for the value used.

b)

A.tabulation of all permitted axial power shapes addressed by

. LOCA calculations inside containment.

Identify the least favorable axial shape (most conservative) associated with each br eak si.ze and provide justification of its conservatism.

Q211.281 The response to Q2'11.3 is unacceptable.

Address the requirements (4.6) of Standard Review Plan 4.6 with regard to the standby liquid control system and the recirculation flow control system.

211. 282 (5.4.6. 1.2.1)

Q211. 283 (4.6.4.1)

'Ihe text indicates all components of the RCIC system are capable of individual functional testing during normal plant operation.

Table 1.3-8 indicates each component, except the flow controller, is capable of functional testing.

Resolve the discrepancy with respect to functional testing of the RCIC flow controller.

Specify the comnon mode failure probability value for both the control rod drive system (CRDS) and the standby liquid control system (SLCS).

Q211. 284

. (5.4.6)

Q211. 285 (4.6.1.1.

2.4.2.1)

Q211. 286

. (4.6.2.3. 1.2)

Q211 287 (5.4.6.2.2)

Is the 12" exhaust pipe shown in Figure 5.4-9a installed as a

sparger to prevent flow os illations which have been known to damage check valves in the turbine exhaust line of the RCIC system? If not, are there other design features used at Susquehanna to prevent this type of damage?

Resolve the following items relating to filtration of condensate water for the CRD hydraulic system.

a)

The text. description and Figure 4.6-5a indicate that normal filtration of condensate water on the suction side of the CRD water punp is acccmplished by a single 25-micron disposable filter.

Explain Ay no filter'is provided in the bypass line to allow for servicing of the punp suction filter.

b)

Describe provisions in the SSES design and operating procedures to protect CRD hydraulic system ccmponents and instrunents from pluggage due to inadvertent failure of either the punp suction filter or the drive water filters. If none

'xist, prov. ide justification that.inadvertent failur e of either type filter will not cause pluggage and result in failure of the system to perform its function.

Identify the layout studies done to assure that no interference exists which will restrict the passage of control rods and the pre-operational test(s) that. are used to show acceptable per formance.

Section 5.4.6.2 of Regulatory Guide 1.70 requires that significant design paraneters for all components of the RCIC system be identified and that all components be shown on appropriate PAI diagrans.

Design paraneters for only a portion of the RCIC components are included in Section 5.4.6.2.2.2.

Some of the more important ccmponents emitted are the:

a)

Mater leg (jockey) pump b)

Vacuun punp c)

Vacuun tank d)

Condensate punp e)

.Turbine and steam supply drain pots f)

Turbine governing and trip throttle valves g) 'ump suction strainers in the suppression pool Provide the significant design parameters for all RCIC components not 'included already in Section 5.4.6.2.2.2 and verify that each ccmponent can be identified on Figures 5.4-9a,and 5.4-9b.

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+Q211. 288 (5.4.6)

Describe the design features and operating procedures that preclude water hanner effects at the punp dis harge of the RCIC system.

Q211

~ 289 (5.4.6.4)

Section 14.2. 12.5 (P50. 1) does not provide sufficient details of the RCIC pre-operational and initial startup test progran to determine whether the RCIC system meets the requirements of Regulatory Guide 1.68.

Provide this information.

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<211. 290 (15.0)

Q211. 291 (15.6. 5)

Q211. 292 (5.2. 2)

Q211. 293 (5.2.5)

Q211. 294 (5.2.5)

For the major ity of events analyzed in Section 15, the'ecirculation flow control mode (autcmatic or manual) assuned in the analysis is not specified.

Our concern is that the mode selected may not result in the most severe margins on MCPR and peak vessel.pressure.

a)

Specify the recirculation flow control mode assuned for each event analyzed in Section 15.

b)

Specify the change in MCPR and peak vessel pressure for each event if the opposite recirculation flow control mode had been assuned in the analysis.

'Ihe intent of Question 211.264 is to have the applicant provide a list of all plant-specific break sizes and locations analyzed.

In addition to this request, provide the peak cladding temperature and peak local oxidation associated with each plant-specific break size.

Subsection 5.2.2.4.1 of the FSAR states that each safety/relief valve is provided with a device to counteract the effects of backpressure which results in the discharge line when the valve is open and discharging steam.

What type of device is provided'?

Nhat effects would be antici.pated if the device were to fail'

'ubsection 5.2.5.3.2 of the FSAR implies that all identified leakage can be'easured while the reactor is operating.

It is not clear how the base data will be established to permit.

compari:son with the 25 gpn identi.fied leakage limit.

Provide the frequency that these data will be recorded and indicate what procedural guidelines are to be used to record the magnitude of the base identified leakage rate.

It is unclear whether comparative "grab" samples of the continmusly monitored containment atmosphere can and will be taken on a periodic basis.

Resolve this ambiguity. If "grab" samples are not to be taken, justify emission of these comparative data..

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