ML20217C731

Forwards Response to 970725 RAI Re License Amend Request Revising RCS Flow Requirements to Allow Increased SG Tube Plugging.Ts Page from 970609 License Amend Request to Convert to Improved Ts,Encl

ML20217C731
Person / Time
Site:	Calvert Cliffs
Issue date:	09/29/1997
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217C733	List: ML20217C731 ML20217C736
References
TAC-M97855, TAC-M97856, NUDOCS 9710020046
Download: ML20217C731 (5)

Text

"

CHART 1R II. CCUSE Ildtimore Gas t.nd Electric Company

? Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calven Cliffs Parkw ay Lusby, Maryland 20657 410 495-4455 leptember 29,1997 l

l l

U. S. Nuclear Regulatory Commission Washington,DC 20555 A'ITENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Responce to the July 25, 1997, Request for Additional Informaticit: License Amendrnent Request; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (TAC Nos. M97855 and M97856)

REFERENCES:

(a) Le;':r fran Mr. C. II. Cruse (DGE) to NRC Document Control Desk, dated January 31, 1997, License Amendment Request; Change to Reactor Coolant System Flow Requirements to Al'ow Increased Steam Generator Tube Plugging (b) Letter from Mr. A. W. Drnmerick (NRC) to Mr. C.11. Cruse (DGE),

dated July 25, 1997, Request for Additional Information - Proposed '

Technical (Specification) Changes to Reactor Coolant System Flow I!mit [], Calvert Cliffs Nuclear Power llaut, Unit Nos. I and 2 (TAC Nos. M97855 and M97856)

(c) Letter from Mr. C 11. Cruse (BGE) to NRC Document Control Desk,

) dated June 9,1997, Revision I to the License Amendment Request to Convert to the Improved Technical Specifications (TAC Nos. M97363 and M97364)

By letter dated January 31,199* l Reference a), Baltimore Gas and Electric Company (BGE) submitted a license amendment request _to the Nuclear Regulatory Commiwlon (NRC) to support operation of 1

Calvert Cliffs Units I and 2 with up to 2500 steam generator tubes plugged in each steam generator. By letter dated July 25,1997 (Reference b), the NRC requested additional information regarding the license amendment request. This letter provides BGE's response to References (b). In addition, this letter im%mits a marked-up page from the recently revised Improved Technical Specification (ITS) license az ..dment request (Reference c), which was not included in Referene: (a) for the easons explained below.

One of the proposed changes in Reference (aT is a revision to the current Technical Specification Table 4.7-1 to change the maximum allowable lift settings for Main Steam Safety Valves RV-3996/4004, RV-3997/4005, RV-399814]6, and RV-3999/4007 from 1065 psig to 1050 psig. This change is 9710020046 970929 ADOCK 0500 7 DR l

3,; /, :..

=.7 . Document Control Desk' 1 September 29,19972 -

- Page 2 j 4 necessary to support the seanalysis~of the Loss of Load and Loss of Feedwater Events, which' credit a =

more restrictive lift setting range for the highest set valves, in Reference (a), it w.:s noted that this

- change would not affect ITS, as the.lTS submittal proposed relocating this information from the Technical Specifications to the Inservice Testing Program. By Reference (c), BGE revised the ITS submittal. 'Ihe revised submittal proposes retaining the main steam safety valve lift settings in the ITS as

- indicated in Table 3.7.1-2. To accommodate this change, ITS Table 3.7.12 is marked up indicating the revised lift settings for the four highest set valves. This change is consistent with those proposed for the current Technical Specifications in Reference (a).

l Attachment (1) t; this letter provides BGE's response to_ the questions posed in References (b).

Attachment (2) is marked-up ITS Teble 3.7.12. The information contained in Attachments (1) and (2) -

does not change the No Significant llazards Determination presented in Reference (a). Should you have further questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours,-

/

- STATE OF MARYLAND ~

TO WIT:

L COUNTY OF CAlVERT -:

h -

- 1, Charles H. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Compary (BGE), and that I am duly authorized to execute and file this L cense Amendment Request on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my -

personal knowledge, they are based.upon information provided by other BGE employees and/or consultants. Such information has been reviewed in accordance with company pmetice and I believe it to be reliable.

/ne & -Lv

/

J is A 11 of 9 WITNESS my Hand and Notarial Scal: A b 7 Lib

- Notary Public

. My Commission Expires: b k Date CHC/GT/dtm Attachments . _ (1) Response to NRC Request for AdditionalInformation dated July 25,1997

.(2) Marked-up Improved Technical Specificatica Page cc: R. S. Fleichman, Esquire H. J. Miller, NRC J. E. Silberg, Esquire Resident inspector, NRC Director, Project Directorate 1-1, URC R.1. McLean, DNR A. W. Dromerick, NRC J. II. Walter, PSC r

ATTACIIMENT (1) 9 BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC REQUEST FOR ADDITIONAL INF01GIATION DATED JULY 25,1997 r- ___

Calvert CliJs Nuclear Power Plant Units 1 & 2 September 29,1997

,. ,s. .

ATTACllMENT (1) I IIALTIMOIG GAS AND ELECTRIC COMPANY

, RESPONSE TO NRC REQUEST FOR- ADDITIONAL INFORMATION DATED JULY 25,1997 NRC Ouestion Address concerns associated with reactor coolant pump loop-seal clearing and break orientation, and compliance with the requirements of 10 CFR $0.46(b), including concerns regarding metaVwater reaction, and long-term cooling. A non-proprietary submittal by Framatome Technologies incorporated (FTI) provided additional information to describe the concern and associatedphenomena.

The small-break loss-of-coolant accident (SBLOCA) scenario involving the principal safety concern involves long term cooling conditions in which the brevk si:e and orientation would be such that the primary system would not depressuri:e, limiting the emergency core cooling system injectionflow rate to that only capable ofmatching the core bolloffrate due to decay heat. Also in this scenario a column of water in the reactor coolant system pump suction loop seal would inhibit the vent path to the break and exert enough additionalpressure to the steam space above the core that the level in the core would be depressed below the top of the core. Should this condition exist as an equilibrium condition, core uncovery would be indefinite, since the attendant s'ecay heat rate would be small and virtually constant.

Under these ci?cimtstances, the cladding oxidation criteria of10 CFR $0.46 could be violated.

To address this concern in the near term, the licensee shouldprovide information to assure that the Calvert Chfs , bnts, in their present configurations (including plant design, technical specifications, procedures, analyses) will operate such that the criteria of10 CFR $0.46 will not be violatedfor it:

SDLOCA analyses in consideranan of the scenario (s) of concern. To confirm its near term assessment and address longer term concerns associated with this SBLOCA scenario, and the capability of this model to quantify the scenarlofs) of concern for ongoing operation andfuture con]Igurations, the licensee should describe its action plan to update its licensing basis SBLOCA model with 'toding and correlations to explicitly simulate the phenomenafor the scenarios) andplant configurations ofconcern as required by 10 CFR $0.46.

Information supplied by ABB/CE in support ofitsforthcoming updated SBLOCA evaluation model (S2M) confirons the potentialfor the scenario (s) ofconcernfor some plant designs, and addresses, both near term and long term, the associated concernsfor all des'gns licensed with the model. This model awaits NRC apt oval; however, vendor assistance may be available to respond to this questionfor the model used in the Calve-t Chffs SBLOCA analyses.

BGE Iksponic Asea Brown Boveri - Combustion Engineering (ABB-CE) is the Nuclear Steam Supply System vendor for Calvert Cliffs. ABB-CE also currently manufactures the fuel for Calvert Cliffs and provides our Emergency Core Cooling System performance evaluation. Reference (1) is the recent ABB CE response ta NRC concerning the Small Dreak Loss-oLCcvant Accident (SB LOCA) scenario discussed above for all ABB-CE plantr.

As discussed in Reference (1), Calvert Cliffs is a " Class A" ABB-CE plant. This is a plant for which the Reactor Coolant System (RCS) loop seal elevation is higher than the top of the active core. Also as discussed in Reference (l), since the Calvert Cliffs loop seal elevation is above the top of the core, Calvert Cliffs will not experience hydrostatically-induced core uncovery due to loop seal clearing and/or refilling behavior. The RCS hydrostatic pressure balance during the SB LOCA scenario of concern is a function of the loop seal elevation. This pressure balance controls the potential for core uncovery. A 1

ATTACllMENT (1)

HALTIMORE GAS AND ELECTRIC COMPANY

• . RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION DATED JULY 25,1997 plant with a loop seal elevation higher than the top of the core is not susceptible to loop seal clearing induced core uncovery during a SB LOCA. Therefore, the core collapsed liquid level during the time of highest differential pressure will be abave the top of the core. As a result, the current Emergency Core Cooling System performance evaluation for Calvert Cliffs is not affected by the above loop seal clearing -

concern, and the criteria of 10 CFR 50.46 will not be violated.

11.e Calvert Cliffs Emergency Opuating Procedures are based on ABB CE guidance provided in CEN 152," Combustion Engineering Emergency Procedure Guidelines." This guidance, with respect to operator response to the SB LOCA scenario of concern, is described in more detail in Reference (1). The Calvert Cliffs optimal procedure for LOCA directs RCS cooldown and depressurization, based on a reasonable time frame for operator actions, within 20 minutes of implementing the LOCA procedure.

This rapid, controlled RCS depressurization will significantly increase the liigh Pressure Safety injection delivery to the RCS and reduce the inventory loss through the break. Although, as discussed above, the loop seal elevation at Calvert Cliffs will prevent the core collapsed liquid level from falling below the top of the core (aner the initial blowdown and recovery) without credit for operator action, the forced cooldown implemented by the operators will avert the conditions that could lead to sustained core uncovery.

As discussed in Reference (1), the ABB-CE SB LOCA evaluation model (the model currently used for Calvert Cliffs) conservatively simulates the phenomena for this scenario and plant configurations of concern as required by 10 CFR 50.46. Under conservative analysis conditions, tha calculations presented in Section 5 of Enclosure (1) to Reference (l' demonstrate that the ABB-CE SB LOCA evaluation model adequately calculates loop seal clearing and the hydrostatic pressure distributions that may act to uncover the core during the clearing process including the SB LOCA scenarios of concern. Furthermore, the results of these (Section 5) calculations show that the SB LOCA evaluation model adequately evaluates the impact of break orientation by virtue ofits elevation dependent modeling capabilities and separated tiow methodologies. Therefore, the current Calvert Cliffs SB LOCA evaluation model uses the noding and correlations that are able to quantify the scenario (s) of concern for ongoing operation and futurr configurations. As a result, an updating of the Calvert C!iffs licensing basis SB LOCA movcl is not required to address this concern.

REFERENCE (1): ABB-CE letter to NRC, LD-97-010, " Additional Response to NRC Request for Additional Information Regarding CENPD-137, Supplement 2 P," dated April 3,1997 2