ML20195J396

Application for Amend to License DPR-46,revising Containment Overpressure Contribution to ECCS Pump,Revising NPSH Requirement post-LOCA & Revising Encl USAR Section 5.2.6. Proprietary GE Repts Encl.Proprietary Info Withheld

ML20195J396
Person / Time
Site:	Cooper
Issue date:	06/15/1999
From:	Swailes J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20137T792	List: ML20195J396
References
NLS990052, NUDOCS 9906210003
Download: ML20195J396 (40)

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N Nebraska Public Power District Nebraska's Energy Leader NLS990052 June 15,1999 U.S. Nuclear Regulatory Commission Attention: Document Control Desk-Washington, D.C. 20555-0001 Gentlemen:

Subject:

Proposed License Amendment Containment Overpressure Contribution to ECCS Pump NPSH Requirement Post LOCA Cooper Nuclear Station, NRC Docket 50-298, DPR-46

References:

1. Letter NLS980201 to USNRC from J. H. Swailes (NPPD) dated December 23,1998, " Response to NRC Generic Letter 97-04 Request for Additional Information."

2. Letter to G. R. Horn from NRC dated November 24,1998,"CNS Response to Request for Additional Information Penaining to Generic Letter 97-04."

In accordance with the provisions specified in 10 CFR 50.90 and 10 CFR 50.4, the Nebraska Public Power District (District) requests that the NRC review and approve a Cooper Nuclear Station (CNS) License Amendment. The proposed License Amendment would allow:

I (1) reliance on a slightly larger amount of containment overpressure for Residual Heat j Remeval (RHR) and Core Spray (CS) pump operation during worst case long term Loss of Coolant Accident (LOCA) conditions (i.e., greater than 102 seconds) while still maintaining f

original license margins of 3 and 6 psi, respectively, for the difference between minimum available containment pressure and the pressure required for minimum pump Net Positive Suction Head (NPSH), ol (2) reliance on a small amount of containment overpressure for CS pump runout during worst case short term LOCA conditions (i.e., less than 10 minutes) while still maintaining an adequate pressure margin of at least 5 psi; and, (3) the use of ANS 5.1 decay heat model in the USAR Section 5.2.6 as currently presented  !

based on analysisjustifying the use of this model as summarized in the attached Safety Evaluation.

Cooper Nudear Station C [u L T (2 E fl C l' P.o. Box 98/ Brownville, NE 683210098 l

[ g y g0f Telephone: (402) 825-38n / Fax: (402) 825 52n 20 ('

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9906210003 990615 9 "" ppd co*

PDR ADOCK 05000298 3 P PDR .2

s NLS990052 Page 2 of 5 During initial plant licensing, the Atomic Energy Commission (AEC) staff questioned CNS compliance with the requirements of Safety Guide (SG) 1. In particular, the AEC questioned the need for reliance on containment overpressure to ensure adequate low pressure ECCS pump NPSil. In response, CNS cited a conservative calculation that determined the minimum available containment pressure and indicated, that although the SG 1 requirement was not met I

for all low pressure ECCS pumps, there was sufficient margin to ensme adequate ECCS pump NPSil. In the Safety Evaluation Report (SER) for initial plant operation, the AEC granted an exception to the SG 1 requirement on the basis of calculated margin and conservatisms assumed in the analysis. The SER indicated that at least a 3 psi margin exists between minimum containment pressure and the pressure required for NPS11 for the RliR pumps and 6 psi margin for the CS pumps.

11istorically, CNS has interpreted SER approval on the basis of maintaining the 3 and 6 psi margins between minimum available containment pressure and required pump NPSil. With the exception of CS pump runout conditions during the first 10 minutes of a worst case LOCA, these margins have always been maintained. The CS runout flow was recently re-evaluated and determined to be higher than previously considered, llowever, even at the higher runout flow there is still a margin of at least 5 psi between the minimum available containment overpressure, and the required overpressure to assure adequate NPS11 for the CS pumps.

l By Reference 2, the NRC clearly established their interpretation of the original SER indicating that the District was allowed to rely on a slight amount of containment overpressure for long  ;

term RIiR pump operation, i.e., "on the order of 1 psi." The letter also indicated that the allowed I reliance was for long term conditions, did not apply to CS pumps, and that short term requirements had not been addressed. Some concerns with the apparent use of different calculation methods and analysis assumptions were also noted. In Reference 1, the District committed to submitting a license amendment to address the specific concerns.

The Attachments to this letter, the No Significant liazards Consideration Evaluation (Attachment 1) and the Safety Evaluation (Attachment 2), demonstrate that there are no significant hazards or other safe:y concerns associated with the License Amendment request as described above. As demonstrated in the Attachments, adequate CS and RiiR pump margins are maintained with all the following considerations:

large break (worst case) LOCA conditions, Service Water temperature of 90"F and initial torus water temperature of 95 F, R11R heat exchanger ef71ciency that accounts for projected end of plant life performance, ANS 5.1 decay heat model (with conservatisms equivalent to a 2 sigma uncertainty),

design basis ECCS strainer debris loading: and,

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NLS990052 Page 3 of 5 1 other conservative assumptions consistent with the original license analysis, such as assumed primary containment leakage of 5% per day, which is well in excess of the 0.635% per day limit allowed by CNS Technical Specifications.

Therefore, the District requests that the NRC approve this request as shown in Attachment 3 for addition to the USAR by February 15,2000.

This proposed change has been reviewed by the necessary Safety Review Committees and the District has concluded that the proposed change does not involve a significant hazards consideration.

By copy of this letter and attachments the appropriate State of Nebraska official is being notified in accordance with 10 CFR 50.91(b)(1). Copies to the Region IV Office and the CNS Resident Inspector are also being sent in accordance with 10 CFR 50.4(b)(2). .

In accordance with the provisions of 10 CFR 2.790, the reports, GE-NE-E1200141-04 and GE- 1 NE-T2300769-00-01, enclosed herewith contain proprietary information and should be withheld from public disclosure. Attached is General Electric's affidavit attesting to the proprietary nature of the information contained in the reports.

Should you have any questions concerning this matter, please contact Mr. Guy Cesare, Nuclear Licensing and Safety Manager at (402) 825-5433.

I Sincerely, John 11. Swailes Vice President ofNuclear Energy

/rar Attachments i cc: Regional Administrator w/ attachment USNRC - Region IV Senior Project Manager 10 copies w/ attachment USNRC - NRR Project Directorate IV-1 i

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NLS990052 Page 4 of 5 i Senior Resident Inspector w/ attachment  !

USNRC Environmental Health Division-Program Manager w/ attachment Nebraska Department of Health NPG Distribution w/o attachment l

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NLS990052 Page 5 of 5

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i STATE OF NEBRASKA ) l

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NEMAHA COUNTY )

Michael F. Peckham, being first duly sworn, deposes and says that he is an authorized i representative of the Nebraska Public Power District, a public corporation and political  !

subdivision of the State of Nebraska; that he is duly authorized to submit this correspondence on i behalf of Nebraska Public Power District; and that the statements contained herein are true to the j best of his knowledge and belief. 1 i

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t Michael F. Peckham Subscribed in my presence and sworn to before me this l[ day of Jw ,1999.

I l GBERAL N01ARY-State of Nekaska  !

I LUANN BRAY d  ! @s My Comm. Eg May 11,2002 NOTARY PUBLIC I

Attachment 1 to NL.S990052 Page1of4 LICENSE AMENDMENT USAR REVISION TO SIIOW TIIE LATEST ECCS NPSH ANALYSIS COOPER NUCLEAR STATION NRC DOCKET NO. 50-298, LICENSE DPR-46

1.0 INTRODUCTION

The Nebraska Public Power District (District) requests that the Nuclear Regulatory Commission

(.NRC) approve a License Amendment to the Cooper Nuclear Station (CNS) design basis. The purpose of the requested License Amendment is to revise the Updated Safety Analysis Report (USAR) to incorporate the latest analysis to demonstrate adequate Net Positive Suction Head (NPSH) for the low pressure Emergency Core Cooling System (ECCS) pumps following a large break Loss of Coolant Accident (LOCA).

2.0 NECESSITY FOR LICENSE AMENDMENT During initial plant licensing, the Atomic Energy Commission (AEC) staff questioned CNS's compliance with the requirements of Safety Guide (SG) 1. In particular, the AEC questioned the CNS position on the requirement that no credit be assumed from containment overpressure, to assure adequate NPSH for the ECCS pumps during LOCA conditions. In response to this question, CNS cited a conservative calculation, which determined the minimum containment pressure following a large break LOCA. This calculation indicated that, although CNS did not meet the SG 1 requirement, there was sufficient margin between the calculated containment overpressure and the overpressure required to assure adequate ECCS pump NPSH.

In the Safety Evaluation Report (SER) for the initial plant license, the AEC granted an exception to SG 1 on the use of containment overpressure to assure adequate NPSH for the low pressure ECCS pumps. As stated in the SER, this exemption was granted based on the conservatisms inherent in the minimum containment pressure calculation and the margins between the containment overpressure and the amount of overpressure required to assure adequate ECCS pump NPSH.

The following factors, which can affect the available NPSH for the ECCS pumps, have changed since the time ofinitial licensing:

The maximum suppression pool temperature, allowed by Technical Specifications (TS) during normal plant operation, has been increased from 90 F to 95 F.

The maximum Service Water (SW) temperature, allowed by TS during nonnal plant operation has been increased from 85 F to 90 F. (Note: Service Water temperature limit was

Attachment 1 to NLS990052 Page 2 of 4 not in the CNS Technical Specifications at the time of original licensing. The 90 F limit was included in the " Improved Technical Specifications"(ITS) as part of Amendment 178.)

The Residual Heat Removal (RHR) system neat exchanger minimum performance criteria (i.e., minimum heat removal factor "K") has been Acreased to provide increased tube plugging margin.

The decay heat model has been changed from the May-Witt to the ANS 5.1-1979 model.

l The ECCS suction strainers have been modified and the allowable debris loading and thus I the pressure drop across the strainers has been changed to be in conformance with NRC Bulletin 96-03.

l The net effect of the above changes has been to increase the required contribution from the calculated containment overpressure to assure adequate NPSH for the low pressure ECCS pumps I following a large break LOCA. However, the margin between available and required i containment overpressure established in the original SER has been maintained. The proposed i License Amendment, and the attached Safety Evaluation, demonstrates that this increased reliance on containment overpressure does not pose a significant safety concern, and sufficient margin between minimum available and required pressure remains.

3.0 DESCRIPTION

OF USAR CHANGES The USAR marked up pages required to dese.ibe the latest FCCS pump NPSH analysis in post LOCA conditions are shown in Attaciment 5. These include:

1. New Figure VI-5-15 showing the containment response and required containment overpressure for RHR and Core Spray (CS) pump NPSH under LOCA conditions.

2. Clarification of assumptions in Chapter VI and XIV to discuss 2-sigma decay heat uncertainty when using ANS 5.1-1979.

3. Add assumption to Chapter VI discussing the inclusion of fibrous and debris loading in determining NPSH requirements.

4. Remove discussion of"special" Case E in Chapter XIV that was performed with lower RHR and CS flows of 6,900 and 4,250 gpm, respectively. This case has not been evaluated with a 2-sigma decay heat uncertainty.

4.0 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of no significant hazards posed by the issuance of the amendment.

4 Attachment I to NLS990052 Page 3 of 4 This evaluation is to be performed with respect to the criteria given in 10 CFR 50.92 (c). The following analysis meets those requirements.

Evaluation of this Amendment with Resnect to 10 CFR 50.92 The enclosed Updated Safety Analysis Report (USAR) changes, necessary to implement the proposed License Amendment, are judged to involve no significant hazards based on the following:

1. Does not involve a sigm'ficant increase in the probability or consequences ofan acci .91 previously evaluated The proposed change does not involve an increase in the probability of an accident previotely evaluated in the USAR. There are no changes being proposed to the maintenance, operation, or design of plant systems or equipment postulated to initiate accidents or transients.

1 The proposed change does not involve an increase in the consequences of an accident l previously evaluated in the USAR. This conclusion is based on the conclusions of the safety j evaluation (Attachment 2). This safety evaluation demonstrates that the containment

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overpressure is sufficiently conservative, and that the calculated margins between the available containment overpressure and the overpressure required to assure adequate low pressure ECCS pump NPSH are such that ECCS pump operation, as credited in the CNS accideat analysis, remains unchanged.

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2. Does not create the possibilityfor a new or different kind ofaccidentfrom any accident previously evaluated The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the USAR. The proposed license amendment does not introduce any new equipment or hardware changes. The attached safety evaluation demonstrates that the only equipment affected by this License Amendment are the low pressure ECCS pumps and that these will retain their ability to function following a LOCA.

3. Does not create a sigmficant reduction in the margin ofsafety The proposed activity does not involve a significant reduction in a margin of safety. The safety evaluation (Attachment 2) demonstrates that, although there is an increased reliance on containment overpressure to assure adequate low pressure ECCS pump NPSH, there remains sufficient margin to provide confidence that the ECCS pumps will operate as required.

Sufficient margin is demonstrated with the added conservatism of a 2-sigma (2 standard deviation) uncertainty in the decay heat model, increased suction strainer debris loading, increased RHR heat exchanger tube plugging margin, and increases in SW and Suppression

Attachment 1 to NLS990052 Page 4 of 4 Pool temperatures. The minimum margin available between available overpressure and required overpressure is at least 5 psi for CS (just prior to 10 minutes) and at least 3 psi for RHR (well after 10 minutes).

5.0 ENVIRONMENTAL IMPACT EVALUATION 10 CFR 51.22(c)(9) provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazard consideration, (2) result in a significant change in the types or significant increase in the amount of any effluents that may be released offsite, or (3) result in an increase in individual or cumulative occupational radiation exposure. The District has reviewed the proposed license amendment and concludes that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the proposed license changes. The basis for this determination is as follows:

1. The proposed license amendment does not involve significant hazards as described previously in the No Significant Hazards Consideration Evaluation.

2. As discussed in the No Significant Hazards Consideration Evaluation, this proposed change does not result in an increase in radiological doses for any Design Basis Accident as there is no increase in accident consequences. This proposed license amendment does not result in a change in the types or amounts of any effluents that may be released offsite. The proposed license amendment does not introduce any new equipment, nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform. The District has concluded that there will not be an increase in the types or amounts of any effluents that may be released offsite and these changes do not involve irreversible environmental conseqcences beyond those already associated with normal operation.

6.0 CONCLUSION

The District has evaluated the proposed License Amendment, and the associated changes to the ECCS pump NPSH analysis. The conclusion of this evaluation is that there are no significant safety concerns associated with this proposed License Amendment. Therefore, for the reasons detailed above, the District requests NRC approval of this proposed License Amendment.

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, i Attachment 2 to NLS990052 i Page1 of11  !

SAFETY EVALUATION {

PROPOSED LICENSE AMENI' MENT ECCS NPSli ANALYSIS 1

1.0 PURPOSE-  !

The 'atent of this Safety Evaluation is to evaluate the safety significance of the proposed License Amendment and the associated USAR changes.

2.0 SYSTEMS AND ACCIDENTS AFFECTED 2.1 Systems Affected The systems affected by this License Amendment are the Primary Containment (PC), Residual 11 eat Removal (RHR) system and the Core Spray (CS) system.

l 2.2 Accidents Affected i The accident analysis potentially affected by this change is the Loss of Coolant Accident (LOCA), 1 3.0 EVALUATION 3.1 History 3.1.1 Initial Plant Licensing Phase Safety Guide (SG) I requires that no credit be taken for calculated increases in containment overpressure, i.e., any pressure greater than initial containment pressure, when calculating the  !

available NPSH for the low pressure ECCS pumps following a LOCA. In Final Safety Analysis Report Question 6.4 (FSAR Q&A) the Atomic Energy Commission (AEC) requested that CNS outline its conformance to SG 1. CNS responded by citing a calculation, which had been performed to determine the minimum containment overpressure following a LOCA. The l following conservative assumptions were made to minimize the available containment overpressure:

• _ An initial drywell temperature of 150 F, which minimizes the initial non-condensible gas mass and thus, minimizes the long-term containment pressure.

1 Attachment 2 l

to NLS990052 l Page 2 of l1 An initial drywell pressure of 0 psig since a higher initial containment pressure will result in a higher containment pressure following the LOCA.

q A post accident containment leak rate of 5% per day which is almost a factor of ten higher than allowed by Technical Specifications and 10 CFR 50 Appendix J testing.

To maximize the suppression pool temperature following the LOCA, the following assumptions were included.

No credit for heat loss through containment walls.

• A power level based on 105% steam flow,2486 MWt.

An initial suppression pool temperature of 90 F.

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A Service Water (SW) temperature of 85 F.

l The performance capabilities of the Residual Heat Removal (RHR) heat exchanger were not  !

explicitly stated in the Q&A. The analysis assumptions also did not explicitly state whether all I of the water in the vessel was assumed to be at saturated conditions.

A loss of off-site power (LOOP) and the failure of one Diesel Generator (DG) were assumed at the beginning of the event. This later resulted in the availability of only one RHR suppression pool cooling loop with one SW and one RHRSW Booster Pump.

The decay heat model used in this analysis was the May-Witt decay heat curve.

The calculation indicated that CNS did not conform to the requirement of SG 1 since a contribution is required from containment overpressure to assure adequate low pressure ECCS NPSH in a worst case long term post LOCA condition. The response to FSAR Q&A 6.4 indicated that, at the time when the pool temperature is at its maximum, the RHR pumps required a contribution of approximately 1.9 psi from the containment overpressure to provide adequate NPSH. The CS pumps had adequate NPSH without any contribution from containment overpressure. The calculation also indicated that the containment pressure at the time of the maximum pool temperature was sufficient to assure that there was a minimum of a 3 psi margin above that required to assure adequate ECCS pump NPSH. The calculation indicated a maximum suppression pool temperature of 192 F.

The AEC, in their Safety Evaluation Report (SER) dated 2/14/73, accepted this non-confomaance with the SG 1 requirements based on a 3 psi margin for the RHR pumps and a 6 psi margin for

r-Attachment 2 to NLS990052 4

Page 3 of 11 the CS pumps and the conservatisms inherent in the minimum available containment pressure calculation. As discussed later in this submittal, it was this " margin" that CNS considered as the design and licensing basis for NPSH analysis. Therefore, changes made st CNS since the original license have ensured this margin was maintained.

3.1.2 Suppression Pool Technical Specification Change There were no plant changes that affected the above position on the contribution of containment overpressure to ECCS NPSH until a 1983 Technical Specification (TS) change, TS Amendment

82. This TS change requested that the maximum allowed suppression pool temperature during normal operation be raised from 90 F to 95 F. The supporting NPSH analysis for the TS change, based on GE report NEDC-24360-P, indicated that there was an increase in the maximum pool temperature from 180 F to 184 F following a large break LOCA. This maximum pool temperature is less than the post-LOCA temperature of 192 F calculated for the analysis cited in the Q&A 6.4 response. Since the maximum initial suppression pool temperature in the TS Amendment analysis had been increased to 95 F, compared to the initial temperature of 90 F for the Q&A 6.4 analysis, a higher, not lower, peak pool temperature would be expected post LOCA. The analysis also reported using a Service Water temperature of 84 F (versus 85 F used in the initial licensing phase).

This apparent discrepancy was noted during the recent NPSH investigations and the District requested that General Electric (GE) investigate this apparent discrepancy. The result of this investigation was that in the analysis for the 1983 TS change, two SW pumps and two RHRSW

. Booster Pumps were assumed to be operating. This contrasts with the analysis for the Q&A 6.4 response, which assumed that the RHR cooling loop has only one RHR, one SW and one RHRSW Booster pump operating. With only one DG operating, the DG loading calculations indicate that there is only sufficient capacity to operate one SW pump and one RHRSW Booster 1 Pump. Thus the appropriate case, with only one SW pump and one RHRSW Booster Pump, was not used to support the TS Amendment. Later analyses have demonstrated that there is adequate NPSH with an initial suppression pool temperature of 95 F, as discussed below. This discrepancy has not been identified as having a significant implication for public heaah and safety as recent analyses demonstrated adequatc margin is maintained to demonstrate pump operability.

3.1.3 Maximum SW Temperature Change The next change which could have affected low pressure ECCS pump NPSH, was an increase in the maximum SW temperature from 85 F to 90 F. This change was performed under a 10 CFR 50.59 change and, based on the evaluation, prior NRC approval was not required. The analysis tojustify this change was based on a GE Report, EAS-053-0889, dated August 1989.

O Attachment 2 to NLS990052 Page 4 of 11 This GE report did contain an analysis of the effect of the increase in SW temperature on the available low pressure ECCS NPSH. This analysis used the same parameters as the original licensing analysis except that the SW temperature was increased from 85 F to 90 F and the initial pool temperature was increased from 90 F to 95 F. This analysis used the conservative May-Witt decay heat curve. The calculated maximum suppression pool temperature was 196 F and the analysis also indicated that there was approximately 3.3 psi margin remaining to the limiting ECCS pump NPSH requirements. This was the basis for CNS to conclude that the plant was within the design and licensing basis with respect to the ECCS NPSH issue. The 196 F peak pool temperature was also evaluated against, and found to be within, equipment operating limits.

3.1.4 Analysis to Increase RHR Heat Exchanger Tube Plugging Margin In 1993 a concern was raised that the 4% tube plugging margin for the RHR heat exchangers would not be adequate for the operating cycle beginning in 1994. An evaluation by Dominion Engineering in 1993 projected the tube plugging margin to reach 16 % (best estimate) by the year 2009. An analysis was subsequently performed by GE, detailed in report GENE-637-045-1293 and GENE-673-020-0993, which analyzed the plant .-sponse to accidents, with a higher tube plugging margin, and included a calculation of the available ECCS pump NPSH. This NPSH calculation used the same parameters as the above initial licensing analysis with the following exceptions:

The ANS 5.1-1979 decay heat curve was used instead of the May-Witt model. The May-Witt curve is more conservative, i.e., it predicts a higher decay heat production than the ANS 5.1 standard. Therefore, use of the ANS standard would support a higher tube plugging margin. A 2-sigma correction was not applied to the ANS 5.1 decay heat curve. However, the conservative default values for delayed neutron capture in ANS 5.1-1979 were included in this analysis.

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A power level of 102%,2429 MWt, in accordance with Reg. Guide 1.49.

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A tube plugging margin of 23% was used, which leads to a decrease in the heat exchanger heat removal capability, represented by a single numerical value, K. The K factor is defined ,

l to be the total heat exchanger heat removal rate divided by the difference between the two heat exchanger inlet temperatures (i.e., suppression pool and service water temperatures). A ,

conservative K factor of 177 Btu /sec- F, corresponding to the 23% plugging margin at the l design fouling was used.

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Attachment 2 j

to NLS990052 1 Page 5 of 11 The following conservative assumptions, in addition to those assumed in the initial licensing analysis, were used:

A conservative quantity of feedwater in the system that was at a temperature greater than the bulk suppression pool temperature was added to the vessel.

100% of the ECCS pump work, converted to heat, was included.

The calculated maximum suppression pool temperature in a post LOCA condition was 195.9 F, which is nearly identical to the conclusions reached for the 1989 analysis in support of a SW temperature increase from 85 F to 90 F. Therefore, the K factor of 177 Btu /sec- F offset the )

J difference between the May-Witt and the ANS 5.1-1979 (without 2-sigma) decay heat models and resulted in the same maximum pool temperature of 196 F as the 1989 analysis.

The RHR and CS NPSH margin.c were shown to be greater than 3 and 6 psi, respectively. The ('

results of this analysis were approved by CNS under 10 CFR 50.59 and incorporated into the USAR.

3.1.5 Generic Letter 97-04 and Responses On October 7,1997, the NRC issued Generic Letter (GL) 97-04, " Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps." In this Generic Letter the NRC requested that licensees submit information necessary to confirm the I adequacy of the NPSH available for the ECCS and containment heat removal pumps. At CNS,  ;

the RHR pumps perform both the ECCS and containment heat removal functions, in letter NLS970226 dated 1/5/98, CNS responded to the specific questions with the conclusion that, I based on the 1993 analysis, CNS was within its licensing basis with respect to the ECCS pump NPSH concern. In a letter dated 8/14/98, the NRC noted that they had reviewed the CNS l

response to GL 97-04 and had a concern that CNS may not be within its licensing basis because the contribution now required from contairunent overpressure to assure adequate ECCS NPSH was greater than that previously reviewed and approved by the NRC. CNS responded that, although it is true that a larger contribution is presently required for adequate NPSH, the margins )

of 3 and 6 psi stated in the AEC SER have been maintained and, therefore, CNS is within its design and licensing basis.

In response to this CNS position on this issue, the NRC issued a letter dated 11/24/98, in which they noted that they did not agree with the CNS conclusion that the plant was within its licensing basis with respect to the ECCS NPSH concem. The NRC detailed in an attachment to this letter their reasons for their disagreement, which can be summarized as follows:

Attachment 2 to NLS990052 Page 6 of 11 Although the AEC SER only noted the containment overpressure margins, the response to the FSAR Q&A 6.4 did provide a series of curves showing the actual magnitude of the contribution required. Thus, this amount of containment overpressure contribution to ECCS NPS11 is part of the CNS licensing basis.

The use of the ANS 5.1 decay heat curve (without a 2-sigma uncertainty), was a change from the May-Witt model reviewed by the NRC.

The SW temperature assumed in the 1993 analysis was 90 F versus the 84 F value used in the last analysis reviewed by the NRC (TS Amendment 82). The reculting peak pool temperature for the 1993 analysis is 196'F versus 184 F. The NRC stated that was indicative of different contaimnent models being used to analyze the maximum pool temperature.

The power level used in the TS Amendment 82 analysis was 104% versus 102% used 6 the 1993 analysis.

Based on the above, the NRC concluded that CNS is outside its licensindasis with respect to containment overpressure contribution to ECCS NPSli. In response, letter NLS980201, CNS committed to submit a License Amendment by 4/30/99 which would address the above issues and in addition, would address the short term CS NPSli issue (see below) and the effect of the debris loading required by NRC Bulletin 96-03.

3.1.6 Short Tenn CS Pump NPSH Concern The District has determined that the previously calculated maximum CS pump runout flow of 6100 gpm, based ou pre-operational test data, was in error due to an inadequate flow meter used in the pre-operational testing. The correct maximum possible CS pump runout, determined in District calculation NEDC 94-142, Rev 3, is 6500 gpm. In the event of a LOCA, this flow could continue for 10 minutes into the event since no credit can be assumed for operator action for the first 10 minutes of the event. There is not sufficient NPSIl for the CS pumps for a flow rate of 6500 gpm without assuming some credit for containment overpressure. Previous submittals to the NRC had never taken credit for containment overpressure in the first 10 minutes of the event.

An analysis of the required and available overpressure is discussed in section 3.4.

3.1.7 Debris Loading on ECCS Strainers NRC Bulletin 96-03 required that a revised ECCS strainer loading, in a post LOCA condition, be included in the calculation for ECCS pump NPSli. CNS installed new strainers to meet these requirements through the 50.59 process on the basis that the 3 and 6 psi margins were maintained, however added containment overpressure was required. This increased reliance on l

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4 Attachment 2 to NLS990052 Page 7 of 11 containment overpressure is included in this Licensing Amendment and discussed further in Section 3.3.

3.2 Effect of 2-Sigma Decav Heat Uncertainty on Long Term NPSH Requirements To demonstrate that the analysis performed in 1993, and presently described in the USAR, assures adequate ECCS pump NPSH following a large break LOCA, the District contracted GE to perform a sensitivity calculation, which includes the following assumptions:

An initial drywell temperature of 150 F, which is a conservative assumption since it will result in a lower maximum containment pressure. This is the same conservative assumption used in the initial licensing analysis (reviewed by the AEC), and also used in the 1993 analysis.

An initial drywell pressure of 0 psig since a higher initial containment pressure will result in a higher containment pressure following the LOCA. This is the same conservative assumption used in the initial licensing analysis, and also used in the 1993 analysis.

A post accident containment leak rate of 5% per day which is almost a factor of ten higher )

than allowed by Appendix J testing. L,owering the non condensibles in the containment will also reduce the peak containment pressure post LOCA. This is the same conservative l

assumption used in the initial licensing analysis, and also used in the 1993 analysis.

Yo maximize the suppression pool temperature following the LOCA, an initial suppression pool temperature of 95 F is assumed and a SW temperature of 90 F, both of these values are the maximum allowed by Technical Specifications and the normal values are generally ,

lower. This differs from the initial licensing analysis, which assumed a maximum pool l

temperature of 90 F and a maximum SW temperature of 85 F. These are the same  !

assumptions used in the 1993 analysis.

f The present design fouling factor and tube plugging margin for the RHR heat exchanger was also assumed, which results in a "K" factor of 177 Btu /sec- F. This is the same assumption used in the 1993 analysis. Although not explicitly stated in the initial licensing analysis, the original GE process diagrams suggest the K factor was 228 Btu /sec *F.

The loss of off-site power (LOOP) and the failure of one DG was assumed at the beginning j of the event. This results in the availability of only one RHR suppression pool cooling loop l with one SW and one RHRSW Booster pump. This is the same conservative assumption {

used in the initial licensing analysis, and also used in the 1993 analyats. 1 s

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Attachment 2 to NLS990052 1

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= ' The ANS 5.1-1979 decay heat model was used with a 2-sigma uncertainty. The delayed neutron capture ("G" factor) was adjusted to envelope the current and projected CNS cycle.

This is a different decay heat model and delayed neutron capture model from that used in either the initial licensing analysis and the 1993 analysis. A discussion of this revised decay heat model is presented in Attachment 4.

The water in the vessel was modeled to reflect the subcooling present during normal power operation. The 1993 analysis assumed that all of the water in the vessel was in a saturated condition. 1 i

The feedwater flow into the vessel was adjusted to more accurately reflect the expected t .-

conditions. This results in a lower amount of feedwater flow into the vessel and also a lower maximum suppression pool temperature. This revised model is based on the fact that the feedwater and condensate pumps will not be in operation following a LOCA/ LOOP and only  !

the water with a temperature above 212*F would be able to flow into the vessel with some of the water remaining in the feedwater piping and heaters. The initial licensing analysis does not discuss and apparently did not include the effects of feedwater addition. The 1993 analysis included an overly conservative assumption regarding the quantity and temperature

. of feedwater.

The results of this analysis indicate a peak suppression pool temperature of 196.7*F. There is i less than a 1*F difference between the latest GE analysis, using 2-sigma uncertainty in the decay l heat and other more realistic but still conservative assumptions, and the 1993 GE analysis. This I indicates that the more realistic assumptions regarding the delayed neutron effect, the amount of subcooling in the vessel and the amount of feedwater flowing into the vessel offset the addition of the 2-sigma uncertainty.to the ANS 5.1 decay heat model. Therefore, the 1993 analysis l provides an acceptable calculation of containment pressures and temperatures, which are used in the current NPSH analysis. The GE report discussing this sensitivity case is included as Attachment 5.

Note that the 1993 analysis also evaluated the plant response, with a reduced heat exchanger performance, to Appendix R and Anticipated Transient Without Scram (ATWS) events.

However, the decay heat models used in these analyses were the same as used in previous analyses (e.g.,~ANS 5.1 - 1979 for Appendix R and a realistic value based on the May-Witt

' Correlation for the ATWS analysis). Therefore, the additional conservatism associated with "2 sigma" is not applicable, or required, for these special events.

E

4 Attachment 2 .

to NLS990052 ' 1 Page 9 of 11 3.3 Effect of New Debris Londinn on ECCS Pumn NPSH The debris loading on the ECCS suction strainers is based on GE topical report NEDC-32721P,

" Application Methodology for GE Stacked Disc ECCS Suction Strainer," dated November 1997, which includes a portion of the Boiling Water Reactor Owners Group (BWROG) methodology presented in NEDO-32686-A. - CNS has further reviewed the analysis to ensure the concerns of the NRC's SER on the BWROG methodology have been appropriately incorporated. The CNS

>- ECCS sucticn strainers were sized based on the following:

Utilizing the " Zone ofInfluence" method described in GE document NEDO-32686-A, a 50%

destruction factor and 100% transport factor resulted in the worst case zone generating j 91.2 lbs, of fibrous material and 20 lbs. of calcium silicate in the suppression pool. The 50% )

destruction factor is conservatively higher than the 28% value approved by the NRC.

. Other debris assumed available for loading on the strainers includes (in lbs.) 550 for sludge, ,

150 for di't/ dust,500 for paint chips, and 50 for rust. These are in accordance with the I BWROG methodology, or for the case of sludge and paint chips, are conservatively higher ]

based on CNS evaluations. i Reflective Metal Insulation (RMI) was conservatively excluded from the mix in accordance with the BWROG testing, which indicated that with CNS type of strainers (stacked disc) the head loss across the strainer would be reduced when RMI was included. Also, the original sizing calculations by the strainer vendor show that RMI loading alone is not limiting with l respect to strainer head loss.

l Run-out flows of 6,500 gpm for CS and 9,240 gpm for RHR were assumed for the first 10 minutes of the accident. Design (throttled) flows of 4,750 gpm and 7,700 gpm for CS and RHR, respectively, were assumed for the remainder of the event (Note: A second RHR pump would also be operating during the first 10 minutes. The debris loading analysis assumes a low flow of 6,500 gpm for this pump, to conservatively maximize the debris loading on the other 2 operating pumps).

The cumulative effects of debris loading and changing suppression pool temperature during a DBA LOCA were analyzed utilizing the 1993 containment NPSH analysis (the basis for utilizing the 1993 analysis was discussed previously in this License Amendment). The attached proposed USAR curves show the containment response and pressure required to meet ECCS NPSH requirements. The minimum margin between available overpressure and required overpressure is at least 5 psi for CS (just prior to 10 minutes) and at least 3 psi for RHR (well after 10 minutes).

t

' {' .

l 9

.. . ]

L j

' Attachment 2 to NLS990052 I Page 10 of11 Additional analyses were performed utilizing the same 1993 containment response profiles, but

' with higher CS and RHR flows to address potential instrument uncertainties. Other sensitivity cases were performed to ensure an early transfer to containment cooling would not cause any adverse debris loading conditions. In all cases there was adequate overpressure available to {

ensure reliable pump operation, but with reduced margins (greater than 1 psi margin with RHR  !

flow at 9,000 gpm'and greater than 4 psi margin with CS flow at 5,450 gpm).

3.4 Short Term Effect on'CS Pumn NPSH Due to Revised Maximum Flow As discussed above, the high run-out flow calculated for the CS pump, which is assumed to I occur for up to 10 minutes, results in a need for crediting containment overpressure during the initial stages of the Design Basis Accident LOCA. However, this run-out flow can only be achieved after the vessel has sufficiently depressurized and at this point in the event there will be adequate containment overpressure available to ensure reliable pump operation. NPPD calculation NEDC 97-044, Rev.1, determines the contribution required from containment overpressure in the first ten minutes to assure adequate CS pump NPSH in the full runout condition. This calculation is based on the 1993 containment response and shows there is at least a 5 psi margin between the required overpressure and the available overpressure during the first 10 minutesc In addition the pump vendor has stated in a letter to CNS that the CS pump can survive for at least 10 to 15 minutes without adequate NPSH, 3.5 Undate of TS Amendment 82 As discussed above, the TS Amendment 82 assumed an incorrect model for the RHR suppression pool cooling mode (i.e.,2 SW and RHRSW Booster Pumps instead of 1). However, the 1993 containment analysis was performed with the correct ECCS pump combinations. The more recent NPSH calculations, which include debris loading, have demonstrated there is acceptable margin.- Therefore, the acceptability of a 95*F pool temperature limit and the impact on ECCS pump NPSH have been properly evaluated. It is also noted that the " worst case" transient / Safety Relief Valve discharge event was also re-evaluated in 1993 and determined to be acceptable, i using the same conservative assumptions for feedwater, decay heat, and reactor vessel temperature as discussed above.

4.0 OTHER FACTORS THAT COULD BE AFFECTED BY PROPOSED CHANGE Since there are no other hardware or operating procedures affected by these proposed changes, the above analysis has adequately addressed any safety concerns.

i

O Attachment 2 .

to NLS990052 Page 11 of11 5.0

SUMMARY

There is no significant effect on the health and safety to the public as a result of this proposed change. Conservative assumptions are utilized to calculate the containment overpressure available for ECCS pump NPSH requirements. The strainer head loss due to debris loading is conservatively estimated and the resulting required overpressure is met by the containment response. There is adequate margin available to address potential uncertainties and to ensure the ECCS pumps will be capable of performing their required safety function.

6.0 REFERENCES

1. FSAR Q&A 6.4

2. GE Report NEDC-24360-P, dated August 1981, Cooper Nuclear Station Suppression Pool Temperature Response

3. GE Report EAS-053-0889, dated August 1989, Analysis of Plant Operation with Higher Service Water Temperature for Cooper Nuclear Station

4. NEDC 94-034, Review of GE Nuclear Analysis GENE 673-020-0993 and GENE 637-045-1293, Supporting the Increase of the RHR Heat Exchanger Tube Plugging Margin (contains Attachment 5)

5. NEDC 99-009, Decay Heat Evaluation, Review of GE Report GE-NE-E1200141-04 (contains Attachment 4)

6. NEDC 97- 042, Review of GE Report GENE-E12-00147-02, Debris Loads Report for Sizing of Cooper RHR and Core Spray Pump Suction Strainers

7. NEDC 97-044, Revision 1, NPSH Margins for the RHR and CS Pumps

8. GE Letter N&SA 99-116, dated March 30,1999, Review of Supporting Analysis for Suppression Pool Operating Temperature Increase Documented in NEDC-24360-P

9. USAR Chapter XIV, Station Safety Analysis

10. USAR Chapter VI, Core Standby Cooling Systems

11. Dominion Engineering Report DEI-354, Rev. O, dated January 18,1993, Evaluation of Tube Degradation in the Cooper Nuclear Station RHR Heat Exchangers

12. Letter NLS970226 to USNRC fromG. R. Horn (NPPD) dated January 5,1998, " Response to NRC Generic Letter 97-04"

13. Letter to G. R. Horn from NRC dated November 24,1998, "CNS Response to Request for Additional Information Pertaining to Generic Letter 97-04"

14. Letter NLS980201 to USNRC from J. H. Swailes (NPPD) dated December 23,1998," Response to NRC Generic Letter 97-04 Request for Additional Information" ,

15. GE topical report NEDC-32721P, " Application Methodology for GE Stacked Dise ECCS Suction Strainer," dated November 1997

16. NEDC 94-142, Revision 3, Core Spray Flows with Minimum Flow Valve Open i 1

e Attachment 3 USAR Changes for 2-sigma: to uts990052 Page I of10 i

Insert "A" for Chanter VI:

15. Fibrous and miscellaneous containment debris loading are included in accordance with NRC Bulletin 96-03 and BWROG methodologies (reference 55).

(Add Reference 55 to Chapter VI- NPPD calculation NEDC 97-044)

Insert "B" for Chanter VI:

Separate analysis were performed to verify the assumptions concerning heat input to containment (i.e. the ANS 5.1-1979 decay heat model, feedwater addition, etc.) for the DBA LOCA demonstrated a level of conservatism equivalent to two (2) standard deviations (2-sigma). Cooper Nuclear Station has committed to include the ANS 5.1-1979 decay heat model with 2-sigma in any future analysis of design basis accidents.

(reference 56,57)

(Add Reference 56 to Chapter VI- NPPD calculation NEDC 94-034)

(Add Reference 57 to Chapter VI- NPPD calculation NEDC 99-009)

Insert "C" For Chanter VI:

New Figure VI-5-15, Available and Required Overpressure for ECCS NPSH (see next page)

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~ Nebraska Public Power District Note: Core Spray pumps initially require Cooper Nuclear Station Containment overpressure due to high run-out flows. Only the Norst case"is presented here Updated Safety Analysis Report (USAR) 4 (i.e.Just prior to 10 minutes when operator acuan is assumed to throtee Core Spray and Minimum Containment Pressure RHR flow). See Referenca 55 for addiuonal For Operation of ECCS Pumps informauon- Figure VI-515

Ahclunem 3 j USAR Changes for 2-sigma: to NLS990052 I

%3 M0 Insert "D" for Chapter XIV:

Separate analysis were performed to verify the assumptions concerning heat input to r:ontainment (i.e. the ANS 5.1-1979 decay heat model, feedwater addition, etc.) for the DBA LOCA demonstrated a level of conservatism equivalent to two (2) standard  ;

deviations (2-tigma). Cooper Nuclear Station has committed to include the ANS 5.1-1979 decay heat model with 2-sigma in any future analysis of design basis accidents.

(reference 69,70)

(Add Reference 69 to Chapter XIV - NPPD calculation NEDC 94-034)

(Add Reference 70 to Chapter XIV - NPPD calculation NEDC 99-009) 4 l

l O

I I

3

,. , USAR

, Attachment 3 Thus, neither the i to NLS990052 independent of any external signal. dent environment in the containment affects the operability of the Page 4 oM0 go, the accident. It is concluded that safety design basis 9 is satisfied.

Using the suppression pool as the source of water for the LPCI cubsystem establishes a closed loop for recirculation of LPCI water escaping from tha break. The LPCI and appropriate portions of the reactor recirculation loops Are designed as Class I (see Appendix C) so that they meet design basis 8.

5.3 CSCs Pumns NPSH l The entire speccrum of possible operating modes of the CSCS has been cxamined for adequacy with regard to NPSH at the various pu nps. Adequate NPSH is eveilable to the CSCS pumps for all the various modes of operation.

The most limiting of all the various modes occurs during the long term transient following a design basis LOCA when one Core Spray and one RHR pump will be running continuously. Figure VI-5-15 is a plot of both the minimum ,

containment pressure required for the Core Spray and RHR pumps to have adequate NPSH and a plot of the minimum containment pressure thatgwau)c ptpuayly3 occur.

At all times there would be at le.ast a 3 pai margin. Ls c.fw(,4,J 4.

In order to demonstrate the margin inherent in Figure VI-5-15, the following is a list of the major assumptions used to calculate the suppression pool temperature and the minimum containment pressure following the design basis LOCA. "

1. The reactor is initially at 102 percent of rated thermal power per Regulatory Guide 1.49. The ANS 5.1 decay heat model is used assuming an -

cxposure of 25,700 MWD /st (amount of energy generated per unit metric-ton fuel mass), which represents a high fuel burnup, and, therefore, a high decay power condition.

2. offsite power is assumed lost at the initiation of the accident cnd is not restored during the entire event.

3. only one onsite diesel generator is available during the entire svent. Consequently, only one pump each for the Core Spray (4,720 gpm) ,

LPCI/ Containment Cooling (7, YOO gpm), and RHR Service Water Booster System (4,000 gpm) is assumed to be available.

4. Tre power required to operate the core spray pump and LPCI/ Containment cooling pump is added to containment heat load by increasing water temperature at the pump discharge accordingly.

5. At the initiation of the accident, the suppression pool has the minimum water volume of 87,650 ft and the maximum temperature of 95*F.

8

6. The service water temperature remains at 90*F throughout the Gvent. .

I

7. During the event, the portion of feedwater in the feedwater cystem that is higher in temperature than peak pool temperature is assumed to continue to return to the ' reactor vessel. -

8. No heat loss from the prj. mary containment to the reactor ,

building airspace is assumed.

9. The minimum thermal capability specified in Table IV-8-1 is used for the RHR heat exchangers.

l VI-5-16 08/15/98

.. , USAR

. Attachment 3

10. Containment cooling by the PHR heat exchi to NLS990052 :ed at C00 seconds into the event. Page 5 ofl0

11. The drywell bulk temperature is assumed to be 150*F together with 100 percent humidity prior to the accident. Normal operating conditions would be 135*F with 20 percent humidity.

12. The initial containment pressure is O psig. Normal operating pressure could be as high as 1.5 psig. There are no circumstances under which a cub-atmospheric pressure could exist in the containment.

13. A containment gas leakage rate of 5 percent per day is =tssumed.

This is substantially higher than the allowed leakage rate of 0.635 percent per day at 58 psig (see Section V-2, " Primary Containment System").

34. The discharge from the RHR heat exchanger always returns to the suppression pool via the drywell and torus sprays to minimize the containment I pressure by spraying cold water into the containment air space.

E 4 The result of assumptions 1 through 9 is to maximize the peak suppression pool temperature. With no offsite power and with one diesel-generator 8 out of service the pool will be cooled by one RHR heat exchanger with one RHR and one SW pump. This, together with the maximum Service Water temperature, results in a peak pool temperature of 196*F. The suppression pool is assumed to be the 1{

a only heat sink even though the metal structures within the containment are T ) capable of storing considerable energy. No credit is taken for any heat losses b/ from the containment other than the energy being removed by the RHR heat I exchanger.

Assumptions 11, 12, 13, result in the minimum possible quantity of non-condensible gases being present in the containment during the transient; l which in turn results in the minimum possible pressure. Assumption 14 gives the j minimum containment gas temperature and thus also minimized the pressure. 1 The combination of maximum fluid temperature and minimum containment pressure calculated with the above assumptions are the most severe conditions for which adequate NPSH must be shown to exist. Figure VI-5-15 shows that adequate I

NPSH would indeed exist even under these very degraded circumstances.

It can be seen that there is a period during which the containment pressure must be in excess of atmospheric pressure if adequate NPSH is to be provided. The design basis for the CSCS for CNS did not include the requirements l that they be functional with no containment back prensure.

These pumping systems were designed long before Safety Guide No. 1 was proposed, and due to back pressure dependence do not strictly conform with Safety Guide No. 1. However, adequate NPSH is provided for all conditions. The dependence on back pressure is justified by the conservative models used to calculate the minimum back pressure. In the incredibly remote event that all the unlikely circumstances assumed above do indeed occur simultaneously for some reason and there is an accompanying loss of back pressure, the plant operator would always have recourse to the RHR/ Service Water intertie.

Separate analyses were performed to demonstrate that local suppression pool tenperatures are maintained below 200*F to ensure stable steam condensation.H'"""

d VI-5-1*7 08/15/90 l

77 USAR

' Attachment 3 to NLS990052 Letter, R. L. Gridley (GE) to D. G. Eisenhut (

45. Page 6 oM0 Low-Core Flow Effects on LOCA Analysis for (

Revision 2," May 8, 1978.

46. Letter, D. G. Eisenhut (NRC) to R. L. Gridley (GE), " Safety Evaluation Report on Revision of Previously Imposed MAPLHGR

-(ECCS-LOCA) Restriction for BWRs at Less Than Rated Core Flow," May 19,1978.

47. NRC Safety Evaluation of the CNS Response to the Station Blackout Rule dated August 22, 1991.

48 NRC Supplemental Safety Evaluation of the CNS Response to the Station Blackout Rule dated June 30, 1992.

49. NRC Supplemental Safety Evaluation and Closcout of Staff Review of the CNS Response to the Station Blackout Rule dated November 19, 1992.

50. Enercon Services Report NPP1-PR-01, " Station Blackout Coping Assessment for Cooper Nuclear Station," Rev. 2, June 1993.

51. NEDC 91-157.

52. DC 95-036.

53. General Electric Letter TCL88032, September 15, 19 8 8, T . C . Le e (G.E.) To G.R. Smith (NPPD), HPCI System Start Time Surveillance Testing Criteria.

54. NEDC 92-100.

5(. Moc 97-o'H E6. NEcc 99 o3 y j

51 NEcc 49-009 i

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VI-7-3 12/08/98

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• COOPER NUCLEAR STATION j UPDATED SAFhTY ANALYSIS REPORT (US i Minimum Containment Pressure l for Operation of ECCS Pumps

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' Figure VI-5-15 07/22/94

USAR Attachment 3 Case E - Operation of one RHRS cooling loop with 1 to NLS990052 R Service Water Booster pump, 1 service water pump, an( Page 8 of 10 t sxchanger - with containment spray.

The major assumptions used in the DBA-IDCA containment analysis were as follows:

1. The reactor is initially at 102 percent of rated thermal power per Regulatory Guide 1.49. The ANS 5.1 decay heat model is used assuming an exposure of 25,700 MWD /st (amount of energy generated per unit metric-ton fuel mass), which represents a high fuel burnup, and, therefore, a high decay power condition.

2. Offsite power is assumed lost at the initiation of the accident and is not restored during the entire event.

3. The power required to operate core spray pump (s) a r.

LPCI/ Containment Cooling pump (s) is added to containment heat load by increasing water temperature at the pump discharge accordingly.

4. At the initiation of the accident, the suppression pool has the j minimum water volume of 87,650 ft'.

5. The service water temperature remains at 90'F throughout the event.

6. During the event, the portion of feedwater in the feedwater system that is higher in temperature than peak pool temperature is assumed to continue to return to the reactor vessel.

7. No heat loss from the primary containment to the reactor building airspace is assumed.

8. The minimum thermal capabilities specified in General Electric Drawina 729E211BB are used for the RHR heat exchangers.

9. The initial air space pressure is 0.75 psig in both the drywell and suppression chamber.

i

10. The initial air space relative humidity is 20 percent for the drywell and 100 percent for the suppression chamber.

11. Initially, the drywell temperature is 135'F, while the suppression pool temperature is at its maximum limit of 95'F.

The containment responses for the five DBA cases are evaluated with the above assunptions, utilizing the SHEX computer code and an analytical model based on NEDO-10320-and NEDO-20533. Case E, the most limiting case, was reanalyzed to evaluate the impact of the ECCS parameter relaxations. A ,

description of the long-term response model is previded in Section 8.4.2. The values of key parameters used as input for DBA-LOCA containment analysis are given in Table XIV-6-3c. The calculated long-term pressure and temperature responses for the five cases are shown in Figureo XIV-6-5 through XIV-6-9.

The initial pressure response of the containment (the first 600 seconds after break) is essentially the same for each of the five cases.

During the long term containment response (after depressurization of the reactor vessel is couplete) the suppression pool is assumed to be the only heat sink in the containment system. The effects of decay energy, stored energy, and energy from the metal-water reaction on the suppression pool temperature are considered, bs*N t r(, XIV-6-24 10/15/98

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• Attachment 3

53. Exten('ed Load Line Limit and ARTS Improven to NLS990052 Lyses for Cooper Nuclear Stat. ion Cycle 14, General Electric C' Page 10 of 10 392P, Rev. 1, May 1991.

54. Letter from J.M. Pilant (NPPD) to T.A. Ippolito (NRC),

" Supplemental Reload Licensing Submittal and Proposed Technical Specifications for Cooper Nuclear Station Reload 4, Cycle 5, NRC Docket No. 50-298, DPR Attachment 2 Change in Main Steam Line Low Pressure Isolation Settings,"

January 31, 1979. 1 l

55. " Evaluation of ATWS Performance at Cooper Nuclear Station,"

MDE 270-1285, General Electric Company, December 1985.

56. Letter from J.S. Charnley (GE) to C.O. Thomas (NRC), Licensing ,

Credit for Banked Position Withdrawal Sequences on Group Notch Plants, Many 10, 1985.

57. Letter from R.E. Engel (GE) to D.M. Vassallo (NRC), Elimination of Control Rod Drop Accident Analysis for Banked Position Withdrawal Sequence Plants, February 24, 1982.

58. Fuel Densification Effects on General Electric Boiling Water Reactor Fuel, August 1973, (NEDM-10735, Supplement 6).

59. Safety Evaluation Report for Cooper Nuclear Station (including Supplements 1 and 2).

60. Letter from G. R. Horn (NPPD) to USNRC, dated January 13, 1993,

" Response to Request for Additional Information Related to Proposed Change No. 100 to Technical Specifications, ' Elimination of Main Steam Line Radiation i Monitor Scram and Isolation Functions,' (TAC No. M83768), Cooper Nuclear Station, i l

NRC Docket No. 50-298, DPR-46." l

61. U.S. NRC Standard Review Plan, Section 15.4.9, NUREG-0800, July 1981.

62. NEDC 94-070 " Control Room Operator Dose Due to Emergency Bypass Filter Shine."

63. NEDC 94-071 " Control Room Operator Dose Due to Inleakage to Control Room."

64. NEDC 94-072 " Control Room Operator Dose Due to Reactor j Building, 10" Core Spray Line and LOCA Cloud." l

65. NEDC 94-157 " Control Room Operator Dose Due to Main Steam Line Break."

66. NEDC 94-176.

67. Supplemental Reload Licensing Report for Cooper Ntamr Station Reload% 15, Cycle 16, General Electric Company, 23A7199, Rev. O, Februory 1993.

68. General Electric Report NEDC-32688P, "Emergene,f Cort. Cooling System Parameter Relaxations for Cooper Nuclear Station, ECCS Non-IOC7. malysis,"

December 1996.

d. # Ecc 94- Od XIV-11-4 08/15/98 p9 v

i

[ .

{

Attachment 4 to NLS990052 consisting of GE Affidavit, Pages 1 through 4 and GE Report GE-10E-E1200141-04, Decay Heat Evaluation for Cooper Nuclear Station, dated April 1999, Pages 1 through 24

L General Electric Company AFFIDAVIT 1, David J. Robare, being duly sworn, depose and state as follows:

(1) I am Technical Account Manager, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which ,

is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in a proprietary GE report GE-NE-E1200141-04, entitIed, Decay Heat Evaluation for Cooper Nuclear Station, dated April 1999. The proprietary information is delineated by bars marked in the margin adjacent to the specific proprietary material.

(3) In making this application for withholding of proprietary information of which it is -

the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 1

. USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information",

and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerav Project v. Nuclear Regulatory Commission.

975F2d871 (DC_ Cir.1992), and Public Citizen Hea th Research Group v. FDA.

704F2dl280 (DC Cir.1983).

(4) Some examples of categories of infomiation which fit into the definition of l proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting l data and analyses, where prevention ofits use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; Affidavit Page 1

1

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers; l

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held.

The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to

- regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclor,ure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical models, methods and processes, including  ;

computer codes, which GE has developed, discussed with the NRC, and applies in the Containment analyses for the BWR.

Affidavit Page 2

]

r a

The development and approval of the containment computer code was achieved at a significant cost, on the order of several million dollars, to GE.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

(9) Public disclosure of the information sought to be withheld is likely to cause '

substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commer/al value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a  ;

substantial investment of time and money by GE. l l

The precise value of the expenise to devise an evaluation process and apply the j correct analytical methodology is difficult to quantify, but it clearly is substantial.  !

l GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same l or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opponunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

Affidavit Page 3

/

STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

David J. Robare, being duly swom, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at San Jose, California, this 3 day of 3\ltn 1999.  :

David J. Robare General Electric Company R

Subscribed and sworn before me this 8 day of CF#6 1999.

/

W W/

Notary Public, State ofCalifonda 4

ANNAHANUN L a-Comminion # 1184507 I WPubic-Collfornla 3antaCien County J

WhBe*esJun1P,2002- -..._,f-Affidavit Page 4

,i L&--

r

)

l Attachment 5 to NLS990052 consisting of )

GE Affidavit, Pages 1 through 4 mid GE Report GE-NE-T2300769-00-01-R1, Cooper Nuclear Station Containment Analysis with ANS 5.1 + 20 Decay Heat, dated June 1999, Pages 1 through 24

.. . . - . )

r- .

I J,,'

General Electric Company AFFIDAVIT I, David J. Robare, being duly sworn, depose and state as follows:

(1) I am Technical Account Manager, General Electric Company ("GE") and have been )

! delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

i (2) The information sought to be withheld is contained in a proprietary GE report GE-l NE-T2300769-00-01-R1, entitled, Cooper Nuclear Station Containment Analysis a with ANS 5.1 + 2cr, dated June 1999. The proprietary information is delineated by bars marked in the margin adjacent to the specific proprietary material.

l (3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and l 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information",

and some portions also qualify under the narrower definition of" trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerav Pro _iect v. Nuclear Regulatory Commission.

975F2d871 (DC Cir.1992), and Public Citizen Health Research Groun v. FDA.

I 704F2d1280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supportmg l data and analyses, where prevention ofits use by General Electric's competitors l without license from General Electric constitutes a competitive economic l l advantage over other companies; I <

b. Information which, if used by a competitor, would reduce his expenditure of I resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; l

Affidavit Page 1

e

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held.

The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treaPnent of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordai.ce with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed, discussed with the NRC, and applies in the Containment analyses for the BWR.

Affidevit Page 2

l

,e l The & ' elopment and approval of the containment computer code was achieved at a significant cost, on the order of several million dollars, to GE.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

l (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability ofprofit-making opportunities. The information is part of GE's comprehensive BWR l safety and technology base, and its commercial value extends beyond the original I development cost. The value of the technology base goes beyond the extensive i l

physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the  !

technology base includes the value derived from providing analyses done with 1 NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.  ;

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this informahon to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

Affidavit Page 3

)

p i

STATE OF CALIFORNIA )

) ss: i COUNTY OF SANTA CLARA )

David J. Robare, being duly sworn, deposes and says: -

l That he has read the foregoing aflidavit and the matters stated therein are tme and correct to the best of his knowledge, information, and belief. 4 l

Executed at San Jose, Califomia, this 8 D day of &WE 1999.

1 David J. Robare General Electric Company Subscribed and sworn before me day ofthis (77dvf 8 7F 1999.

1

, J fA/Y Notary Public, State of California ANNA HANUN p j Commission # 1184507 Notory Public-CoCfomic l [

] Santo Clara County 9  ;

1 My Comm.E@esJun 19,2002 1 Affidavit Page 4