ML20070H243

Forwards Response to NRC Questions on Pressurized Level Limiting Conditions Tech Specs.Table Presenting Summary of Evaluation of Impact of Changing Water Level for Each DBE Encl

ML20070H243
Person / Time
Site:	Calvert Cliffs
Issue date:	12/17/1982
From:	Lippold W BALTIMORE GAS & ELECTRIC CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
NUDOCS 8212230274
Download: ML20070H243 (8)

Text

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e B ALTIMORE G AS AND ELECTRIC COMP ANY P.O. B O X 14 7 5 B A LTIM O R E. M A R YL A N D 21203 NUCLEAR POWER DEPARTMENT COLVERT OLIFF$ NUCLEA R PQwER PL ANT LUSSY, M ARYLAND 20657 December 17, 1982 Office of Nuclear Reactor Regulation U. S. Imclear Regulatory Commission Washington, D. C. 20555 ATTENTION: MR. R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

SUBJECT:

Calvert Cliffs Nuclear Power Plant Units No.1 and 2 Docket Nos. 50-317 and 50-318 Response to NRC Questions on Pressurizer Level LCO Technical Specification

REFERENCE:

(A)

A. E. Lundvall to R. A. Clark letter dated 9-22-82,

" Amendment to Operating License DPR-53 Fifth Cycle License Application."

Gentlemen:

The staff posed a verbal question on our request for modification of the Technical Specification on pressurizer level LCO. The attachment to this letter is our response.

In the course of research to respond to the question an error in Reference (A) was identified. The whole body dose resulting from the Steam Generator Tube Rupture Event in Unit 1, Cycle 5 had been reported erroneously by Combustion Engineering to BG&E. Instead of.08 REM to the whole body, the dose should have been.17 REM to the whole body. The error was subsequently identified and documented by Combustion Engineering during normal QA audit procedures. Since the error was so small and the absolute number was still two orders of magnitude below the 10CFR100 guidelines BG&E was not notified at that time.

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l 9212230274 021217 DR ADOCK 05000

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Office of Nuclear Reactor Regulation December 17, 1982 Page 2 Should you have any questions, please contact us.

Very truly yours, s

T. J. Lippold Nuclear Fuel Management WJL:fld Attachment (40 copies) cc:

J. A. Biddison, Esquire G. F. Trowbridge, Esquire D. H. Jaffe - NRC R. R. Mills - CE R. E. Architzel-NRC/CC i

Office of Nuclear Reactor Regulation December 17, 1982 Page 3 bee R. E. Denton A. E. Lundvall, Jr.

R. C. li. Olson J. A.Tiernan l

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ATTACHMENT glestion Relative to the proposed pressurizer level LCO, document the review process which lead to selection of the most limiting events for further analyses.

Response

The following is the result of a review to select those Design Basis Events which were dadlyzed for the Calvert Cliffs Unit I and Unit 11 pressurizer water level Technical Specification limits modification. The objective of the evaluation was to ensure that pressure limits were not exceeded (as a result of the change in pressurizer high-level Technical Specification) by DBEs which result in increased RCS pressure. Also, DBEs which reduce pressurizer level were evaluated to ensure that ree: tor coolant circulation i

is not adversely effected (as 4 result of the change in pressurizer low-level Technical j

Specification). The evaluation was accomplished by reviewing all DBEs to determine which events result in the largest increase or decrease in pressurizer level /p essure.

These limiting events were then reanalyzed to ensure that the applicable criteria were not exceeded.

l This evalu4 tion shows that the limiting DBEs sensitive to the proposed T.S. change are the Loss of Lo4d, Excess Load and Steam Generator Tube Rupture event. Therefore, these events were analyzed to assure that a change in pressurizer water level does not lead to exceeding the criteria for these transients. The Excess Charging event was determined to be the event having the maximum potential for filling the pressurizer.

Therefore, it was analyzed as part of the process to set the pressurizer water level upper Technical Specification LCO limit.

Table 1 pre:ents a summary of the evaluation of the impact of changing the pressurizer Wdter level for each Design Basis Event. Details of the events reanalyzed are provided in Reference 1.

Reference 1.

Letter, A.

G. Lundvall (BG&E) to R. A. Clark (NRC), "Calvert Cliffs Unit No. I dnd Unit No. 2 Docket Nos. 50-317 and 50-318 Request for Amendment to Pressurizer Level LCO Technical Specification," dated 9/29/82.

TABLE 1 IMPACT OF THE NEW PRESSURIZER WATER LEVEL BAND ON DBE ANALYSES The Impact of the Change in Design Bw.is Event Event Criteria Pressurizer Level on DBE CEA Withdrawal SAFDLs not to be exceeded.

During a CEAW event, the approach to SAFDL is driven by a reactivity and power distribution transient. The event is not sensitive to pressurizer conditions and i

consequently was not reunalyzed. This is due to the fact that for the lirnit4ng CEA withdrawals, the transient is terminated by the Variable High Power Trip before the pressurizer pressure rises significantly. For slow withdrawals, the transient is terminated by the High Pressurizer Pressure Trip. However, the post trip pressure increase is less limiting than in the Loss of Load event.

Boron Dilution To assure sufficient time for There is no impact on the analysis of this event since i

the operator to take action credit is not taken for the pressurizer water volume.

j before shutdown margin is lost.

Excess Load SAFDLs not to be exceeded.

Reanalyzed for new pressurizer level band (see Reference 1).

Loss of Load

1) SAFDLs not to be exceeded, Reanalyzed for new pressurizer level band (see

2) RCS upset pressure limit not Reference 1).

to be exceeded.

TABLEI (Continued)

The Impact of the Change in Design Basis Event Event Criteria Pressurizer Level on DBE Loss of Feedwater

1) SAFDLs not to be exceeded, Loss of Load event is more limiting than Loss of Flow

2) RCS upset pressure limit not Feedwater. Therefore, the impact of the pressurizer to be exceeded.

level change on this event is bounded by the results of the Loss of Load event.

There is no impact on the approach to SAFDLs during this transient since no credit was taken for pressure increase.

Excess Feedwater SAFDLs not to be exceeded.

This event has a similar (but less severe) effect on Heat Removal Due the primary system as does the Excess Load event.

To Feedwater The Excess Load event reanalyzed assumed an increase Malfunction in turbine demand of approximately 45 %

The equivalent turbine demand for the Excess Feedwater Heat Removal event is less. Therefore, approaching the SAFDLs is less severe than for the Excess Load event analyzed and the results of the Excess Load analysis is bounding.

Reactor Coolant SAFDLs not to be exceeded.

SAFDL protection for this event is provided by the System Depressurization TM/LP trip. The TM/LP coefficients are determined by the Excess Load Event.

Since the coefficients established by the Excess Load event are conservative for the RCS Depressurization event, protection against exceeding SAFDLs during this event is assured by the Excess Load analysis.

Loss of Coolaat SAFDLs not to be exceeded.

There is no impact of a change of initial pressurizer Flow water level on the Loss of Coolant Flow event, since credit is not taken for the RCS pressure rise.

TABLE 1 (Continued)

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The Impact of the Change in Design Basis Event Event Criteria Pressurizer Level on DBE

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Loss of Non-

1) SAFDLs not to be exceeded, The Loss of Non-Emergency AC Power transient be-Emergency AC Power

2) Site boundary doses not to haves like a Loss of Flow event. Therefore, the exceed 10CFR100 guideline discussion on the Loss of Flow event also applies here.

limits.

Pressurizer level has no impact on the site boundary dose calculated for this event.

I Control Element SAFDI.s not to be exceeded.

The impact of a change of the initial pressurizer Assembly Drop water level on this analysis is insignificant.

The transient approach to SAFDLs during this transient are driven by the post drop power recovery and redistri-bution. Initial pressurizer level has no effect on LHR and has negligible effect on the transient DNBR.

Asymmetric Steam SAFDLs not to be exceeded.

The limiting ASGT event margin degradation is driven Generator Events by the temperature induced power distribution tilt.

Because of the relatively stable pressurizer conditions through the time of MDNBR, pressurizer level has no

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impact on this event.

CEA Ejection

1) Maintain coolable geometry, The effect of a change in pressurizer water level on i

2) Site boundary doses not to the behavior of this event is negligible. The exceed 10CFR100 guideline transient is driven by the reactivity insertion 1

limits.

and power distribution effects of the ejected rod and is 1

insensitive to pressurizer level.

Steam Line Break

1) Maintain coolable geometry, During the SLB event, the time of maximum degrada-

2) Site boundary doses not to tion in DNBR occurs well after emptying of the exceed 10CFR100 guideline pressurizer. Consequently, there is no effect of a limits.

change in pressurizer water level on the behavior of this event.

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TABLE 1 (Continued)

The impact of the Change in Design Basis Event Event Criteria Pressurizer Level on DBE Steam Generator

1) Site boundary doses not to Reanalyzed for new pressurizer level band (see Tube Rupture exceed 10CFR100 guideline Reference 1).

limits.

Seized Rotor Event

1) Maintain coolable geometry, The effect of a change in pressurizer water level l

2) Site boundary doses not to on the behavior of this event is negligible. Since I

exceed 10FCR100 guideline credit is not taken for an increase in RCS pressure, I

limits.

the relationship between pressurizer level and degradation in DNBR is similar to that of the LOF event.

Loss of Coolant Criteria of !')CFR50.%.

Increasing (decreasing) the initial pressurizer level Accident will result in prolonging (shortening) the blowdown period of a LOCA transient.

It will have r.egligible impact on the refill /reflood hydraulic transient. It is-estimaty that a change in the initial pressurizer level of 200 ft will change the time of end of blowdown by approximately 0.2 sec. The resultant effect of such a small change to the time of end of blowdown on peak cladding temperature (PCT) is insignificant because ::he PCT occurs during reflood (at approximately 250 seconds) and is determined by such reflood parameters as steam cooling heat transfer and thermal radiation.

For this reason it is concluded that the change in pressurizer level will have an insignificant impact on ECCS performance.