Text

Supplement to License Amendment Request
	ML053050300
Person / Time
Site:	Beaver Valley
Issue date:	10/28/2005
From:	Cosgrove T; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-05-168, TAC MC4645, TAC MC4646, TAC MC6725
	Download: ML053050300 (157)
	v • d • e

FENOC Beaver Valley Power Station Route 168' RPO. Box 4 FirstEnergy Nuclear Operating Company Shippingport, M 15077 0004 Thomas S. Cosgrove 724-682-5203 Director, Maintenance October 28, 2005 L-05-168 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Supplement to License Amendment Request Nos.

320 (Unit No. 1 TAC No. MC6725) and 302/173 (Unit No. 1 TAC No. MC4645/Unit No. 2 TAC No. MC4646)

This letter transmits supplements to two license amendment requests that were previously submitted by FirstEnergy Nuclear Operating Company (FENOC). The listed submittals requested amendments to the above licenses in the form of changes to the Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 Technical Specifications.

The license amendment requests being supplemented are:

1. Replacement Steam Generators, submitted by FENOC Letter L-05-069, License Amendment Request 320, dated April 13, 2005

2. Extended Power Uprate, submitted by FENOC Letter L-04-125, License Amendment Requests 302 and 173, dated October 4, 2004 The primary reason for these two supplements is because the changes to Technical Specification 3.5.1, Accumulators, proposed in the extended power uprate submittal require a revision. The revision consists of raising the minimum accumulator cover pressure to a value that is consistent with a revised small break loss of coolant accident analysis. Since the revised analysis is applicable to both the extended power uprate and replacement steam generators submittals, the revision to the Unit 1 Technical Specification is withdrawn from the extended power uprate submittal and added to the replacement steam generators submittal.

Enclosure 1 provides details of these license amendment request (LAR) supplements and an assessment of the proposed changes. The attachments to Enclosure 1 provide the proposed changes for each of the LARs being supplemented and the supporting technical justification.

Letter L-05-069 requested approval of the replacement steam generators LAR by January 2006 in order to support the installation of the BVPS Unit No. 1 replacement steam generators during the 2006 spring outage. The requested amendment shall be implemented prior to the first entry into Mode 4 during plant startup from the IR17 refueling outage planned for the spring of 2006.

Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Requests 320 and 302/173 L-05-168 Page 2 Letter L-04-125 requested approval of the proposed extended power uprate (EPU) amendments by November 2005. However, as documented in the NRC acceptance review letter dated July 19, 2005, NRC approval of the EPU LAR is expected by July 19, 2006. The Unit No. 1 EPU amendment shall be implemented within 120 days following issuance of the amendment.

The Unit No. 2 EPU amendment shall be implemented prior to the first entry into Mode 4 during plant startup from the 2R12 refueling outage planned for the fall of 2006.

The changes proposed in the supplements have been reviewed by the Beaver Valley Power Station review committees. The changes were determined to be safe and do not negate or negatively impact the no significant hazard considerations submitted in FENOC letters L-04-125 or L-05-069.

No new commitments are contained in this submittal. If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager - Licensing, at 330-315-7243.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October A2- 2005.

ncerely, Thomas S. Co ve

Enclosure:

FENOC Evaluation of the Proposed Supplements C: Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Senior Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

L-05-168 Enclosure I FENOC Evaluation of the Proposed Supplements BACKGROUND The NRC review of the replacement steam generator (Reference 1) and extended power uprate (Reference 2) license amendment requests (LAR) has resulted in several requests for additional information (RAI) pertaining to small break loss of coolant accident (SBLOCA) analysis. The SBLOCA analysis supporting these two LARs was performed to be consistent with the NRC approved Westinghouse NOTRUMP-EM methodology. These RAIs requested information that incorporates adjustments to the NRC approved Westinghouse NOTRUMP-EM methodology.

The requested information pertains to the impact of not applying the reactor coolant system (RCS) loop seal clearing restriction and using a smaller mesh (non-integer break) size to determine the limiting size break. A re-analysis of SBLOCA at the extended power uprate power level of 2900 MWt was performed to show that acceptable results, consistent with 10 CFR 50.46, can be obtained using the NRC approved Westinghouse NOTRUMP-EM methodology with adjustments that address the NRC concerns (loop seal clearing and non-integer break sizes). However, changes to input assumptions to support the analysis has resulted in the need to revise the minimum accumulator cover pressure in Technical Specification 3.5.1, Accumulators, for both Beaver Valley Power Station (BVPS) Units I and 2, to support the changes proposed in the replacement steam generator and extended power uprate LARs.

Since the re-analysis of SBLOCA requires a minimum cover pressure that is different than what is proposed in the extended power uprate (EPU) LAR, and the revised SBLOCA analysis is required to support the replacement steam generator (RSG) LAR, supplements to the RSG and EPU LARs are being submitted for NRC approval.

SUPPLEMENT DETAILS Presently the EPU LAR proposes changes to Technical Specification 3.5.1 for both units. This EPU supplement proposes a revision to the Technical Specification 3.5.1 changes proposed in the EPU LAR. Since the revised SBLOCA analysis is applicable to both units, and required for the replacement steam generator project, the revised Technical Specification 3.5.1 markup is added to the changes proposed in the RSG LAR and withdrawn from the EPU LAR.

Presently the RSG LAR, which is applicable to Unit I only, does not propose any change to Technical Specification 3.5.1. Therefore, this RSG supplement incorporates the revised changes to Technical Specification 3.5.1 into the RSG LAR.

CHANGE DESCRIPTION The existing proposed changes to Technical Specification 3.5.1 consists of three parts. The first part consists of changing the limits on accumulator water volume and cover pressure. The second part consists of replacing the word "contained" with "usable" in Surveillance Requirement 4.5.1.a.l. The third part consists of providing the accumulator volumes in percent of indicated level in addition to gallons in the volume portion of the Limiting Condition for Operation (LCO) statement.

L-05-168 Enclosure 1 Page 2 of 5 The revision proposed by these supplements consists of raising the minimum accumulator cover pressure to 611 psig, deleting percent of indicated level in the accumulator volume portion of the LCO statement and inserting "usable" into the same portion of the LCO statement. The new minimum cover pressure is the result of the revised SBLOCA analysis. The percent of indicated level is being removed from the LCO because it needlessly complicates the LCO. This addition was originally proposed as an operator aid, but is unnecessary. Inserting "usable" into the LCO provides consistency with revised Surveillance Requirement 4.5.1.a.l.

Proposed Technical Specification Bases changes for both units were provided for information in the EPU LAR. These changes state that the Technical Specification limits for usable accumulator water volume, boron concentration and minimum cover pressure are analysis values and that the maximum cover pressure limit preserves accumulator integrity. A revision to these Bases changes, reflecting the proposed revisions to the Technical Specification, are included with these supplements. As with the Technical Specification changes, the Unit I Bases changes are moved to the RSG LAR and withdrawn from the EPU LAR.

Attachments A-i and A-2 contain the markups to Technical Specification 3.5.1 for Units I (RSG) and 2 (EPU), respectively. Attachments B-I and B-2 contain the markups to the applicable Technical Specification Bases section for Units 1 (RSG) and 2 (EPU), respectively.

The proposed changes to the Technical Specifications and Bases have been prepared electronically. Deletions are shown with a strike-through and insertions are shown double-underlined. This presentation allows the reviewer to readily identify the information that has been deleted and added. To meet format requirements the Index, Technical Specifications and Bases pages will be revised and repaginated as necessary to reflect the changes being proposed by these supplements.

Attachment C contains a revision to Section 5.2.2 of Enclosure 2 of the RSG submittal (Reference 1) that provides the revised SBLOCA analysis. The revised analysis supports the changes proposed to the accumulator volume and the cover pressure, including the new minimum pressure value. The revisions to Section 5.2.2 of Enclosure 2 of the RSG submittal are highlighted for ease of identification.

Attachment D contains a revision to Section 5.2.2 of Enclosure 2 of the EPU submittal (Reference 2) that provides the revised SBLOCA analysis. The revised analysis supports the changes proposed to the accumulator volume and the cover pressure, including the new minimum pressure value. The revisions to Section 5.2.2 of Enclosure 2 of the EPU submittal are highlighted for ease of identification.

TECHNICAL EVALUATION The primary reason for these LAR supplements is to change the minimum accumulator cover pressure. The change is being made to reflect the revised SBLOCA analysis that does not take credit for RCS loop seal clearing of intact loops and uses non-integer sizes to determine the limiting size break. The revised SBLOCA analysis establishes the minimum cover pressure needed to ensure that the analysis will meet the peak cladding temperature (PCT) and maximum cladding oxidation limits of 10 CFR 50.46.

L-05-168 Enclosure I Page 3 of 5 Section 5.2.1 of Enclosure 2 of the RSG and EPU submittals (References 1 and 2), and the revised Section 5.2.2 of Enclosure 2 of the RSG and EPU submittals, Attachments C and D respectively, provide the technical justification for the proposed change to the accumulator limits on water volume and cover pressure. These changes are consistent with analysis inputs and provide the necessary operating margin at the current power level with the replacement steam generators and at the EPU power level. Changing the LCO statement and Surveillance Requirement 4.5.1.a.1 to address the usable volume is consistent with the analysis volume limits.

RSG SUPPLEMENTAL NO SIGNIFICANT HAZARDS CONSIDERATION The changes to the replacement steam generator license amendment request (Reference 1) being proposed with this supplement consists of:

1. changing the limits on the accumulator water volume and cover pressure,

2. replacing the word "contained" with "usable" in the volume surveillance requirements, and

3. inserting the word "usable" in the accumulator volume portion of the Limiting Condition for Operation (LCO) statement.

The analysis supporting these changes was conducted at extended power uprate (EPU) conditions of 2900 MWt and thus bounds operation at the current power level with the replacement steam generators.

FirstEnergy Nuclear Operating Company (FENOC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed supplemental changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.

A comprehensive analytical effort has been performed to incorporate the proposed supplemental changes into the design and analysis basis for Beaver Valley Power Station (BVPS) Unit No. 1. Analyses and evaluations have been performed for the nuclear steam supply systems (NSSS) and balance of plant (BOP) systems and components, including the nuclear fuel. These comprehensive analytical efforts, which include the proposed changes to the accumulator volume and pressure limits, demonstrate that BVPS Unit No. 1 meets applicable design and licensing requirements. The analyses are conservative and bounding with respect to operation with replacement steam generators at the current power level. The safety and radiological dose consequence analyses confirm that safety analysis and dose consequence analysis acceptance criteria will be satisfied with the proposed supplemental changes.

The accumulators are not initiators of any design basis accident or event, and therefore the proposed supplemental changes will not increase the probability of any accident previously evaluated. The probability of any evaluated accident or event is independent of the supplemental changes being proposed. The proposed supplemental changes will

L-05-168 Enclosure I Page 4 of 5 not adversely affect accident initiators or precursors. They will not alter or prevent the accumulators from performing their intended function within the applicable acceptance limits.

The proposed supplemental changes were evaluated for their effect on accident dose consequences. The updated dose consequence analyses demonstrate compliance with the limits set forth for alternative source term (AST) applications in 10 CFR 50.67, as supplemented by Regulatory Guide 1.183 or 10 CFR 100.

Therefore, none of the proposed changes involve a significant increase in the probability of an accident previously evaluated, and the dose consequences remain within the allowable limits set forth for AST applications in 10 CFR 50.67, as supplemented by Regulatory Guide 1.183 or 10 CFR 100.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The proposed supplemental changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

No new accident scenarios, failure mechanisms or single failures are introduced as a result of the proposed supplemental changes. All systems, structures and components previously required for the mitigation of an event remain capable of fulfilling their intended design function. The proposed supplemental changes will not have an adverse effect on the accumulators and will not challenge their performance or integrity.

Therefore, the proposed supplemental changes will not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed supplemental changes will not involve a significant reduction in a margin of safety.

The proposed supplemental changes will not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed supplemental changes to the accumulator Technical Specification limits are being made to provide adequate margin such that BVPS Unit No. 1 can be operated in a safe manner with the replacement steam generators at either the current or EPU power levels. The revised small break loss of coolant accident analysis establishes the minimum cover pressure needed to ensure that the analysis will meet the peak cladding temperature and maximum cladding oxidation limits of 10 CFR 50.46 and thereby does not involve a significant reduction in a margin of safety. These supplemental changes will not adversely impact plant safety because they will not adversely affect the ability of the accumulators to perform their function.

Therefore, the proposed supplemental changes do not involve a significant reduction in a margin of safety.

L-05-168 Enclosure 1 Page 5 of 5 EPU SUPPLEMENTAL NO SIGNIFICANT HAZARDS CONSIDERATION The no significant hazard considerations submitted with the EPU LAR (Reference 2) remains valid for the extended power uprate supplement changes.

Based on the above, FENOC concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

CONCLUSIONS The impact of all the changes to Technical Specification 3.5.1 for BVPS Unit No. I has been evaluated in the no significant hazard considerations provided in this supplement to the RSG LAR. Thus, the changes proposed by this RSG supplement augment the no significant hazard considerations submitted in Reference 1.

The impact of all the changes to Technical Specification 3.5.1 for both BVPS units has been evaluated in the no significant hazard considerations submitted with the EPU LAR. Thus, the change to the minimum cover pressure being proposed by this EPU supplement does not negate or negatively impact the no significant hazard considerations submitted in Reference 2.

ENCLOSURE 1 REFERENCES

1. FENOC Letter L-05-069, License Amendment Request 320, dated April 13, 2005

2. FENOC Letter L-04-125, License Amendment Requests 302 and 173, dated October 4, 2004.

ATTACHMENTS A-1 Supplement to LAR 320 Technical Specifications (Unit 1)

A-2 Supplement to LAR 173 Technical Specifications (Unit 2)

B-I Supplement to LAR 320 Technical Specification Bases (Unit I)

B-2 Supplement to LAR 173 Technical Specification Bases (Unit 2)

C Revised Replacement Steam Generator Submittal (L-05-069) Enclosure 2 Section 5.2.2 D Revised Extended Power Uprate Submittal (L-04-125) Enclosure 2 Section 5.2.2

Attachment A-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Changes License Amendment Request No. 320 Supplement The following is a list of the affected pages:

Page 3/45-1 I

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a. The isolation valve open,

b. Between 7-6464-681 and 781626A5 gallons of usable borated water,

c. Between 2300 and 2600 ppm of boron, and

d. A nitrogen cover-pressure of between 6b5611 and 66_8_5 psig.

APPLICABILITY: MODES 1, 2 and 3.*

ACTION:

a. With one accumulator inoperable due to boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

b. With one accumulator inoperable for reasons other than Action a, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With either Action a or b not being completed within the specified completion time, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to

< 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1. Verifying the centaiinedusable borated water volume and nitrogen cover-pressure in the tanks are within limits, and

2. Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

BEAVER VALLEY - UNIT 1 3/4 5-1 Amendment No. 2-5a

Attachment A-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Changes License Amendment Request No. 173 Supplement The following is a list of the affected pages:

Page 3/45-1

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

a. The isolation valve open,

b. Between 7536859-8-galLons and !78-2"019 gallons of usable borated water,

c. Between 2300 and 2600 ppm of boron, and

d. A nitrogen cover-pressure of between &8-5--1J and E656 BS psig.

APPLICABILITY: MODES 1, 2 and 3.*

ACTION:

a. With one accumulator inoperable due to boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

b. With one accumulator inoperable for reasons other than Action a, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With either Action a or b not being completed within the specified completion time, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to

< 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1. Verifying the entainedusable borated water volume and nitrogen cover-pressure in the tanks are within limits, and

2. Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

BEAVER VALLEY - UNIT 2 3/4 5-1 Amendment No. 3 l

Attachment B-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Bases Changes License Amendment Request No. 320 Supplement The following is a list of the affected pages:

l Page l B 3/45-1

I1 j- - -I - - 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ProvidedforInformation Only.

BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each of the RCS accumulators ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the accident analysis are met. _The-specified--echnical Specification Yalues_,for-us ableaccumulator volnme-,boron-concentration-and-minimum pes s ureare-analysisvalus TheTechnicalSpecification-maximum pressure-As.below-the-maximum-analysis-pressure-so-that-the-relief valve-is-not-challenged-at-the-maximum-Technical-Specification pressure. -Themaximumni-trogen-coverePrsuresu Ilimit-pre-vents acCU d pip Vlve acuimn and ut-im nresere accum lator int grity The smal 1bhrak LENA analy ia performedLat the~minimum nxzgen covegrpre ure.-s-ince-sensitivity analseshave demonstrated hat highex itrogen-come= pressureesnults =irn a computed peak cladding temperature benefit-. The___rechnical Specification-values-for-pressure-and volume-do-not account__for instrumentation-uncertainty.-

If the boron concentration of one accumulator is not within limits (Action a), it must be returned to within the limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In this condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core.

In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large main steam line break. Thus, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to return the boron concentration to within limits.

If one accumulator is inoperable for a reason other than boron concentration (Action b), it must be returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In this condition the required contents of two accumulators cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur under these conditions, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The completion time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable

accumulator to OPERABLE status is justified by WCAP-15049-A, Revision 1, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times", dated April 1999.

If the accumulator cannot be returned to OPERABLE status within the associated completion time (Action c), the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the reactor coolant system pressure reduced to < 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The BEAVER VALLEY - UNIT 1 B 3/4 5-1 Change No. 1-D20 I

Attachment B-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Bases Changes License Amendment Request No. 173 Supplement The following is a list of the affected pages:

l Page I B3/45-1

I P., -

ll Provided for Information Only.

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each of the RCS accumulators ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the accident analysis are met. -The-specified-TechnicalSpecification values-for-usable-accumulatorvDolumes boronconcentration-and-minimum pressure-are-analysis-values_ TheTechnical-Specification-maximum pressure-is-below-the-maximum-analysis-pressure-so-that-the--relief yalve ilsnot-challenged~at-the-maximum-Technical-Specification r sres The _ maximum limitsure1i ipregvyed accumulator relief_ valve actuation-and ultima t-e-P-y==preservsc; aumulato-i rity-The-small breakL anaysissperformea iOC the-minimum-nit en vepressr sincisensitin yanaLyseshave demonstrated-that 3hihernitrogen coverpressureresults in-a compiuted peak _claddinqg temperaturebenefi-t. The Technical Specification values-for__pressure-and-volume-do-not-account-for instiumentation-uncextainty-If the boron concentration of one accumulator is not within limits (Action a), it must be returned to within the limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In this condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core.

In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large main steam line break. Thus, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to return the boron concentration to within limits.

If one accumulator is inoperable for a reason other than boron concentration (Action b), it must be returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In this condition the required contents of two accumulators cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur under these conditions, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt actions will be taken to return the inoperable accumulator to OPERABLE status. The completion time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable accumulator to OPERABLE status is justified by WCAP-15049-A, Revision

1, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times", dated April 1999.

If the accumulator cannot be returned to OPERABLE status within the associated completion time (Action c), the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the reactor coolant system pressure reduced to < 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The BEAVER VALLEY - UNIT 2 B 3/4 5-1 Change No. 2-O13.QI I

Attachment C Revised Replacement Steam Generator Submittal (L-05-069) Enclosure 2 Section 5.2.2

Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 5

5.1 5.2 5.2.1 5.2.2 Small Break LOCA 5.2.2.1 Introduction This section contains information regarding the Small Break Loss-of-Coolant Accident (SBLOCA) analysis performed in support of the EPU for BVPS-1 (with Model 54F replacement steam generators) at the NSSS power level of 2910 MWt (2900 MWt reactor power). The purpose of analyzing the Small Break LOCA is to demonstrate conformance with the 10 CFR 50.46 (Reference 1) requirements for the conditions associated with the EPU. Important input assumptions, as well as analytical models and analysis methodology for the Small Break LOCA are contained in subsequent sections. Analysis results are provided in the form of tables and figures, as well as a more detailed description of the limiting transient. The analysis has shown that no design or regulatory limit related to the Small Break LOCA would be exceeded due to the EPU power and associated plant parameters.

5.2.2.2 Input Parameters and Assumptions The important plant conditions and features for BVPS-1 are listed in Table 5.2.2-l A. Several additional considerations that are not identified in Table 5.2.2-lA are discussed below.

Figure 5.2.2-1 depicts the hot rod axial power shape modeled in the Small Break LOCA analysis. This shape was chosen because it represents a distribution with power concentrated in the upper regions of the core (the axial offset is +13%). Such a distribution is limiting for Small Break LOCA since it minimizes coolant swell while maximizing vapor superheating and fuel rod heat generation at the uncovered elevations. The chosen power shape has been conservatively scaled to a standard 2-line segment K(Z) envelope for BVPS-I based on the peaking factors shown in Table 5.2.2-lA.

Figures 5.2.2-2 and 5.2.2-3 provide the SI flowve'rsus pressure curves modeled in the'Small Brecak LOCA analysis. -Figure 5.2.2-2 shows the flows from one High Head Safety Injection (HHSI) pump, where the faulted loop injects to RCS pressure. Figure 5.2.2-3 shows flows from one HHSI pumip' and one Low Head Safet* Injection (LHSI) pump, where the faulted loop injects into'containment.

5.2.2.3 Description ofAnalyses and Evaluations Analytical Model The requirements for an acceptable ECCS evaluation model are presented in Appendix K of J0 CFR 50.

For LOCAs due to Small Breaks, less than 1 square foot in area, the Westinghouse NOTRUMP Small Break LOCA Emergency Core Cooling System (ECCS) Evaluation Model (References 2, 3, and 4) is 5-7

I I 111ll-FENOC Enclosure IAttachment C REPLACEMENT STEAM GENERATOR used. The Westinghouse NOTRUMP Small Break LOCA ECCS Evaluation Model was developed to determine the RCS response to design basis Small Break LOCAs, and to address NRC concerns expressed in NUREG-0611 (Reference 5).

The Westinghouse Small Break LOCA ECCS Evaluation Model consists of the NOTRUMP and LOCTA-IV computer codes. The NOTRUMP code is employed to calculate the transient depressurization of the Reactor Coolant System (RCS), as well as to describe the mass and energy release of the fluid flow through the break. Among the features of the NOTRUMP code are: calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, regime-dependent drift flux calculations in multiple-stacked fluid nodes and regime-dependent heat transfer correlations. These features provide NOTRUMP with the capability to accurately calculate the mass and energy distribution throughout the RCS during the course of a Small Break LOCA.

The RCS model is nodalized into volumes interconnected by flow paths. The broken loop is modeled explicitly, while the intact loops are lumped together into a second loop. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum. The multi-node capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a calculation of the behavior of the loop seal during a Small Break LOCA. The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations.

Fuel cladding thermal analyses are performed with a version of the LOCTA-IV code (Reference 2) using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions. The LOCTA-IV code models the hot rod and the average hot assembly rod, assuming a conservative power distribution that is skewed to the top of the core. Figure 5.2.2-4 illustrates the code interface for the Small Break Model.

Analysis The EPU Small Break LOCA analysisconsdered nine differen b esf BVPS-1 a- indicated by the results inTable' 5.2.2-6AK! A break spectrum of 1.5-, 2-, 2.25-, 2.5-; 2.75-, 3-, 3.25-, 4-. and 6-inch breaks was considered. For BVPS-l,'the 2.75-inch break was found to be limiting for PCT and the 2.5-inch break was found to be limiting for oxidation; For BVPS- I, the 1.5-inch case was found to be non-limiting in NOTRUMP and therefore PCT information was not calculated.

The most limiting single active failure used for a Small Break LOCA is that of an emergency power train failure which results in the loss of one complete train of ECCS components. In addition, a Loss-of-Offsite Power (LOOP) is postulated to occur coincident with reactor trip. This means that credit may be taken for at most one high head safety injection (HHSI) pump. In the analysis for BVPS-I, one HHSI pump is modeled. The Small Break LOCA analysis performed for both units models the ECCS flow as being delivered to both the intact and broken loops at the RCS backliriessure for breaks smaller than the cold leg HHSI nozzle (1.5-inch inch breaks) and at containment pressure for breaks greater than' the cold leg HHSI nobzle (6-inch breaks). These Si flows are illustrated in Figure 5.2.2-2 and 5.2.2-3 for each scenario. Note that for the 6-inch breaks, no SI is assumed in the faulted loop because the break is postulated along the SI line. The LOOP and the failure of a diesel generator to start as the limiting single 5-8

__ Enclosure I Attachment C REPLACEMENT STEAM GENERATOR failure for Small Break LOCA is part of the NRC approved methodology and does not change as a result of the EPU conditions. The single failure assumption is extremely limiting due to the fact that one train of SI, one motor driven auxiliary feedwater (AFW) pump, and power to the reactor coolant pumps (RCPs) are all modeled to be lost. Any other active single failure would not result in a more limiting scenario since increased SI flow would improve the overall transient results.

Prior to break initiation, the plant is in a full power (100.6%) equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. Other initial plant conditions used in the analysis are given in Table 5.2.2-IA. Subsequent to the break opening, a period of reactor coolant system blowdown ensues in which the heat from fission product decay, the hot reactor internals, and the reactor vessel continues to be transferred to the RCS fluid. The heat transfer between the RCS and the secondary system may be in either direction and is a function of the relative temperatures of the primary and secondary conditions. In the case of continuous heat addition to the secondary during a period of quasi-equilibrium, an increase in the secondary system pressure results in steam relief via the steam generator safety valves.

When a Small Break LOCA occurs, depressurization of the RCS causes fluid to flow into the loops from the pressurizer resulting in a pressure and level decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low-pressure reactor trip setpoint, conservatively modeled as 1935 psia, is reached. LOOP is postulated to occur coincident with reactor trip. A safety injection signal is generated when the pressurizer low-pressure safety injection setpoint, conservatively modeled as 1745 psia for BVPS-I, is reached. Safety injection flow is delayed 27 seconds after the occurrence of the low-pressure condition. This delay accounts for signal processing, diesel generator start up and emergency power bus loading consistent with the loss-of-offsite power coincident with reactor trip, as well as the pump acceleration and valve delays.

The following countermeasures limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection supplement void formation in causing a rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay. No credit is taken in the Small Break LOCA analysis for the boron content of the injection water. In addition, credit is taken in the Small Break LOCA analysis for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the reactor trip signal, considering the most reactive RCCA is stuck in the full out position. A rod drop time of 2.7 seconds was used while also considering an additional 2 seconds for the signal processing delay time. Therefore, a total delay time of 4.7 seconds from the time of reactor trip signal to full rod insertion was used in the Small Break LOCA analysis.

2. Injection of borated water provides sufficient flooding of the core to prevent excessive cladding temperatures.

During the earlier part of the Small Break transient (prior to the postulated loss-of-offsite power coincident with reactor trip), the loss of flow through the break is not sufficient to overcome the positive core flow maintained by the reactor coolant pumps. During this period, upward flow through the core is maintained. However, following the reactor coolant pump trip (due to a LOOP) and subsequent pump coastdown, a period of core uncovery occurs. Ultimately, the Small Break transient analysis is terminated 5-9

I- l IIl ._

FENOC Enclosure 1Attachment C REPLACEMENT STEAM GENERATOR when the top of the core is recovered or the'core mixing level is increasing, and ECCS flow provided to thee RCS exceeds the break flow ratei The core heat transfer mechanisms associated with the Small Break transient include the break itself, the injected ECCS water, and the heat transferred from the RCS to the steam generator secondary side. Main Feedwater (MFW) is conservatively isolated in 10 seconds for BVPS-I (consisting of a 3 second signal delay time and a 7 second main feedwater isolation valve stroke time) following the generation of the pressurizer low-pressure SI signal. Additional makeup water is also provided to the secondary using the auxiliary feedwater (AFW) system. An AFW actuation signal is derived from the pressurizer low-pressure SI signal, resulting in the delivery of AFW system flow 60 seconds after the generation of the SI signal. The heat transferred to the secondary side of the steam generator aids in the reduction of the RCS pressure.

Should the RCS depressurizetoapproximately 625 psia (accumulator minimum pressure), the cold leg accumulators begin to inject borated water into the reactor coolant loops.

5.2.2.4 Acceptance Criteria and Results The acceptance criteria for the LOCA are described in 10 CFR 50.46 (Reference 1) as follows:

1. The calculated maximum fuel element cladding temperature shall not exceed 2200'F.

2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Criteria I through 3 are explicitly covered by the Small Break LOCA analysis at EPU conditions.

For criterion 4, the appropriate core geometry was modeled in the analysis. The results based on this geometry satisfy the Peak Clad Temperature (PCT) criterion of 10 CFR 50.46 and consequently, demonstrate that the core remains amenable to cooling.

For criterion 5, Long-Term Core Cooling (LTCC) considerations are not directly applicable to the Small Break LOCA transient analysis addressed in this section, but are assessed by Sections 5.2.3 and 5.2.4 as part of the evaluation of ECCS performance.

5-10

iENQC Enclosure IAttachment C REPLACEMENT STEAM GENERATOR The acceptance criteria were established to provide a significant margin in ECCS performance following a LOCA.

In order to :deterjmine at produced the'most limiting Small Break LOCA case' (as determined by the highest calculated peak cladding temperature), nine break cases were examined for BVPS-I.' These cases were investigated to caPture the most severe postulated Small Break LOCA event.

The following discussions provide insight into the analyzed rconditioiis.

Limiting Temperature Conditions The RCS temperature analyzed was based ona nominal vessel average temperature of 580.00 F.-'Howevier, the analysis is applicable over the range of 566.2- 580.00 F. The analysis supports a+/-'.F T,-'.

uncertainty. The'analysis showed that the 2.75-inch break case is the limiting PCT case for BVPS-l.,The limiting case transient is discussed below.

Limiting Break Case The results of Reference 6 demonstrate that the cold leg break location is limiting with respect to postuated cold leg, hot leg and pump suction leg break locations. The PCT results are shown in Tables 5.2.2-4A and 5.2.2-5A."Inherent in the Small Break analysis are several input parameters (see Section 5.2.2.2 and Table 5.2.2-lA), while Table '5.2.2-6A provides the key transient event times.

For the EPU Small Break LOCA analysis, th.e limiting PCT case for BVPS-1 was the 2.75-inch break case. A summary of the transient response'for'the limiting PCT case is shown in Figures'5.2.2-5A through Figure 5.2.2-15A. '-These figures present the response of the following parameters.

RCS Pressure

Core Mixture Level

Core Exit Vapor Temperature

Broken Loop and Intact Loop Secondary Pressure

Break Vapor Flow Rate

Break Liquid Flow Rate

Broken Loop and Intact Loop Accumulator Flow Rate

Broken Loop and Intact Loop Pumped Safety Injection Flow Rate

Peak Clad Temperature

Hot Spot Fluid Temperature

Rod Film Heat Transfer Coefficient Upon initiation of the limiting 2.75-inch break for BVPS-I there is an initial rapid depressurization of the RCS followedbyan intermediateequi'ibrium at pproximately 1150 psia (see Figure -5.2.2-SA).- The limiting 2.75-inch break depressurizes to the' accumulator injection' setpoint of 625 psia at approximately 1438 seconds for BVPS-I (see Figure 5.2.2-I lA). During the initial period of the Small Break transient, the effect of the break flow rate is not sufficieint to overcome the flow rate maintained by the reactor coolant pumps as 'they coast down: As suchiinoimial upward flow is maintained through the core and core heat isadequately removed. Following reactor trip, the removal of the heat generated as a result of fission products decay is accomplished via a two-phase mixture level covering the core. 'The core mixture level 5-11

.- I t ills FENOC Enclosure IAttachment C REPLACEMENT STEAM GENERATOR and peak clad temperature transient plots for the limiting break calculations are illustrated in Figures 5.2.2-6A and 5.2.2-13A; resp-ctiveiy. These figures show that the peakd6ad temperature occurs near the time when the core is most deeply uncovered and the top ofhe core is being cooled by stearm Thstime is characterized by the highest vapor superheating~above 'th'e mixture level (refer to Fig'urQ 5.2.2-7A). For BVPS-l the limiting PCT time-in-lifewas determined to be 8,000 MWD/MTU.

A comparison of the flow provided by the' safety injection system to the intact nd brokeni loops can be found in Figure 5.2.2-12A. The cold leg break vapor and liquid mass fow rates are provided in Figures 5.2.2-9A and 5.2.2-10A, respectively. Figures 5.2.2-14A and 5.2.i-15A provide additional information oni the fluid temiperature at the hot spot and hot rod surface heat transfer coefficienttat the hot spot, respectively. Figure 5.2.2-8A depicts the secondaiy side pressure for both the inttci and bioken loops foi the limiting PCT break case.

Total Oxidation For the EPU Small Break LOCA analysis, the naximum local oxidation'case' for BVPS4I was the 2.5-inch break case. The maximum local'transien't oxidation is 11.07% for BVPS-' at 20,000 MWDIMTU. The limiting transient oxidation occurs at the burst elevation and inicludes both outside'and post-rupture inside oxidatio i Pre-existing (iire-triinsient) oxidation was "also considered and the sum of the pre-transient and transient' oxidation remains below- 17% at all times in lifeUfor all fuel resident iin the core.

Additional Break Cases Studies documented in Reference 6 have deterimined that the limiting PCT Small Break transient occurs for breaks of less than I 0-inches in diameter in the cold leg.' For BVPS-I, the limiting PCT is captured by the 1.5- 2-, 2.25-, 2.5-, 2.75-, 3-, 3.25-, 4- and 6-inch break spectrum.- The beginning-of-life (BOL) results of these break spectrum cases are given in Table 5.2.2-4A. Figures 5.2.21 6A through 5.2.2-36A address the'non-limiting BOL cases-(2-, 2.25-, 2.5., 3- '3.25 and 6-inch) analyzed for BVPS-I. The 1.5-inch case for BVPS-1 produced only minimal core utncovery a'Ad therefore PCT information wasnot calculated. Note that plots for the 1.5-inch' case are not included here. The-plots for each of the additional'non-limiting break cases include:

1. RCS Pressure

2. Core Mixture Level

3. Peak Clad Temperature For BVPS-1, the PCTs for each of the additional breaks consideredare shown in Table 5.2.2-4A and are less thian the limiting 2.75-inch break case.' The PCT wVasnot caulated for the i.5-inch'case due to the minimal core uncovery.

Transient Termination Then 10 CFR 50.46 criteria contin ue id be satisfied beyond the end of the calculated transient due to the piresence of some' 6ri all oft he following onditio ns:

5-12

%NQ9 Enclosure IAttachment C REPLACEMENT STEAM GENERATOR

1. The RCS pressure is gradually decreasing.

2. The net mass inventory is increasing.

3. The core mixture level is recovered, or recovering due to increasing mass inventory.

4. As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel cladding temperatures will continue to decline indicating that the temperature excursion is terminated.

ZIRLO/Zirc-4 Cladding At the time at which this analysis is implemented, no new Zirc4 fuel is expected to be inserted into the core. All of the Zirc-4 fuel will be burned for at least one cycle, and ZIRLOTh fuel will be implemented at non-EPU conditions at least one reload cycle before the EPU is implemented. Therefore, the ZIRLOThm fuel is considered limiting with a PCT of 1895.00 F, at 8,000 MWD/MTU burnup for BVPS-l. The fuel temperatures/pressures used in these calculations were based on NRC approved fuel performance code PAD 4.0 (Reference 7) which addresses all the helium release related issues. This ahalysis hasbeen performed using the rnost limiting tasrperatuieip essure 'calculated for 17x17 noiFBA' RFA'fuel. he non-IFBA fuel bounds IFBA fuel for Small Break LOCA'ainalyses. Note that the effect of ainnular peilets, which are only'present in the IFBA fuel, was considered in the ana'ysis.

5.2.2.5 Conclusions The Small Break LOCA analysis considered 'a break spectrum of 1.5-, 2-,2.25-,2.5-,-2.75-, 3-, 3.25-,4-,

and 6-inch diameters for BVPS-1. 'For BVPS l, peak cladding temperature of 1895TF was calculated at the lini"ting time-in-life of 8,500 MWD/MTU 'for the 2.75-inch case an'd a maximum transiaht oxidation of 11.07% was calculated at the limiting time-in-life of 20,000 MWD/MTU for the 2.5-inch case.

The analysis presented in this section shows that the accumulator and safety injection subsystems'6f the Emergency Core' Cooling System, together with the beat removal capability of the steam generator, provide sufficient core heat removal capability to maintain the calculated peak cladding' temperature for Small Break LOCA below the required limit of i1 CFR '50.46.- Furthermore, the analysis shows that the local cladding oxidation and core wide average oxidation, including consideration of pre-existing and post-LOCA oxidation, and outsid6'and pst-rupture pcladding inside oxidation, areless than the 10 CFR 50.46 (Reference 1) limits.

Table 5.2.2-7 provides a resuts summary for the BVPS-lBSBLOCA EPU analysis. Results include PCT, m aximum local oxidation and total hydrogen'ge'neration.

The results and conclusions' of the analysis-performed for Small Break LOCA for the reactor po we~r of 2900 MWt (2910 MWt NSSS power) bound and support operation at the current reactor power of 2689 MWt (2697 MWt NSSS power), thus supporting the staged implemenitation' of EPU at BVPS-1.

5.2.2.6 References 1.' "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume'39, Number 3, January 1974, as 'amended in Federal Register, Volume 53, September 1988.

5-13

I, IflL FENOC Enclosure IAttachment C REPLACEMENT STEAM GENERATOR

2. Meyer, P. Elk "NOTRUMP -A Nodalisie t Small Break and General Network Code,'"

WCAP-10079-P-AJ(proprietary) and WCAP-10080-NP-A (non-proprietary), August 1985,

3. Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (proprietary) and WCAP-10081-NP-A (non-proprietary),

August 1985.

4. Thompson, C. D. et " endum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," WCAP-lI054-P[-A; Addendum 2, Rev. I (proprietary), July 1997.

5. - "GenericEvaluation of Feedwater.Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plant,," NUREG-061 1, January 1980.

6. Rupprecht, S. D. et al., "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP- II145-P-A (proprietary), October 1986.

7. Slagle, NV. H., (ed.) et al., "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," WCAP-15063-P-A, Revision 1, July 2000.

5-14

FENOC zr- Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Table 5.2.2-IA BVPS-I Input Parameters Used In the Small Break LOCA Analysis Input Parameter Value Core Rated Thermal Power-i 00% 2900 Calorimetric Uncertainty, % 0.6 Fuel Type 17 X 17 Robust Fuel Assembly (RFA)

Total Core Peaking Factor, FQ 2.40 Hot Channel Enthalpy Rise Factor, Fm1 1.62 Hot Assembly Average Power Factor, P1Y 1.42 Maximum Axial Offset, % +13 Initial RCS Loop Flow, gpmnoop 82,840 Initial Vessel T..s, 0F Max: 580.0 Min: 566.2 Initial Pressurizer Pressure (plus uncertainties), psia 2300 Reactor Coolant Pump Type Model 93A with Weir Pressurizer Low-Pressure Reactor Trip Setpoint, psia 1935 Reactor Trip Signal Delay Time, seconds 2.0 Rod Drop Delay Time, seconds 2.7 Auxiliary Feedwater Temperature (Maximum), 'F120F Number of AFW Pumps Available Following a LOOP I Motor Driven AFW Flow (Minimum) to all 3 Steam Generators, gpm 294 (98 gpmISG

3) at 1107 psig AFW Flow Delay Time (Maximum), seconds 60 AFW Actuation Signal Pressurizer Low-Pressure Safety Injection Isolation of Steam Line Signal Pressurizer Low-Pressure Reactor Trip/LOOP Steam Generator Type Model 54F Maximum AFW Piping Purge Volume, ft3 168 Steam Generator Tube Plugging (Maximum), % 10 Maximum MFW Isolation Signal Delay Time, seconds 3 MFW Control Valve Isolation Ramp Time, seconds 7 MFW Isolation Signal Pressurizer Low-Pressure Safety Injection Steam Generator Secondary Water Mass, IbmISG 99,930 Containment Spray Flowrate for 2 Pumps, gpm 4983 (plus 981 gpm to account for flow to the sump)

RWST Deliverable Volume (Minimum), gallons 317,000 5-15

- I L- - - _-_ Il lIL FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Table 5.2.2-1A (continued)

BVPS-1 Input Parameters Used in the Small Break LOCA Analysis Input Parameter Value SI Temp at Cold Leg Recirculation Time (Maximum), IF 190 ECCS Configuration 1 HHSI pump; faulted line injects to RCS pressure (1 .S-inch -4inch breaks) 1 HHSI pump, I LHSI pump, no ECCS in the faulted loop (6-inch bireak) lECCS Water Temperature (Maximum), °F65 Pressurizer Low-Pressure Safety Injection Setpoint, psia 1745 SI Flow Delay Time, seconds 27 ECCS Flow vs. Pressure See Table' 5.2.2-2iand 5.2.2-3 Initial Accumulator Water/Gas Temperature, IF 105 3

Initial Nominal Accumulator Water Volume, 957 Minimum Accumulator Pressure, psia 625 5-16

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Table 5.2.2-2 Safety Injection Flows Used in the Small Break LOCA Analysis (l HHSI pump, faulted loop liniect t ps .5-iich

-c4-inch breaks for BVPS-1)

RCS Pressure (psia) Intact Loop (Ibmlsec) Broken Loop (ibm/sec) 314.7 37.59 20.28 414.7 36.63 19.79 514.7 35.56 19.17 614.7 34.45 18.61 714.7 33.42 18.06 814.7 32.34 17.50 914.7 31.25 16.88 1014.7 30.14 16.25 1114.7 29.03 15.70 1214.7 27.92 15.07 1314.7 26.67 14.45 1414.7 25.28 13.61 1514.7 23.85 12.92 1614.7 22.43 12.08 1714.7 20.97 11.39 1814.7 19.50 10.56 5-17

FENOC

.__._% Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Table 5.2.2-3 Safei fnjeCtioi

^ Iit askstfobs flows Used in.bthe in _,.e's..................

.& " . i,__]~

Small i;hi _,

I ,,, ,A Break'LOCKrAnulyiis (illHS! pminp, I LISI pump, n-O ECCS in the fauited loop becausethe break is postulated along the HHSI line- 6-inch breaks for BVPS-1) lRCS Pressure jpsia) BVPS-1 Intact Loop (lbrnsec) 14.7 336.34 24.7 313.32 34.7 290.45 64.7 216.34 104.7 109.7 1i4.7 45.12 119.7 34.73 164.7 214.7 34.73 314.7 32.43 4i4.7 30.14 514.7 27.78 614.7 25.42 714.7 22.92 814.7 20.42 914.7 17.78 1014.7 15.00 1114.7 12.22 1214.7 9.31 1314.7 639 1414.7 3.06 1514.7 0.0 5-18

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR iTable 52.2-4A BV-PS1ISBLOCTA BOL Results Break Siie (inj 2 225 i.5 2.75 3 325 4 P )723 1804.8 1i93.2 1839.2 i77.6 i617.0 1334.2 267.9 POTTime(s) 31603 2i47.0 2209.8 1743.7 b1386.4 l2i5.8 780.2 i2092 PCT Elevation i2 12t 12 12 .11.75 11.5 11.25 1i.5 Max;Local ZrO'2 (%) 3.15 3.6 5.04 4.28 2.87 1.19 0.i4 0.13 Max. Local Zr2 Elev.(ft 12 12 12 12 11.75 11.5 11.25 115

^Core-Widc Av.'ZrOk (%) 0.4 048 0.O64 0.56 0.4 0.18 0.02 0.02 Table 5.2.2-5A BVPS-1'SBLOCTA Limitinig Results from the 25-inch (Traisient Oildation) and 2.75-inch (PCI) Time-in-Life Study Break Siz 2.5 2.75 Tire-in-Life (MWD/MTU) 20,000 8,000 PF)) 1i796 1895.0 PCTine(s) 2212.9 1723.7 PCTEevaion (ftj 12 12 Hot Rod Buirst Time (s) ;1770.2 1721.7 Hot Rod BurstElevation(fl) ,11.75 '12 Max. Local Transient ZrO 2 (%) 11.07 8.82 Max.Local Transient 7rO2 Elev. (fi) ,11.75 12 Core-Wide Avg. ZrO2 (%j 0.62 0.52 5-19

I ULE...

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Table 5.2.2-6A BVPS-1 NOTRUMP Results 1.5- 2.25- 2'5- 2.75- 3.25-Event Time (seci Inch 2-Inch inch nh Inch 4-iUiinc 6inch lBreak Initiation 0 0 0 0 0 0 0 0 0 lReactor Trip Signal 54.1 29.0 22.5 17.9 14.6 12.3 10.6 7.3 4.A S-Signal 75.9 42.4 33.9 27.9 23.8 20.8 18.6 14.4 i0.2 lS Flow Delivered 102.9 69A. 60.9 54.9 50.8 47.8 45.6 4i.4 37.2 t

Loop Seal Cdearing ? 1840 930 725 656 484 414 358 241 62 lCore Uncovery (3) 1020 813 658 672 526 362 233 121 AccumulatorInjec-tion N/A 4017 2378 1821 1438 1138 996 637 291 RWST Volume Delivered 3033 3025 3017 3011 3006 3001 2998 2992 N/A PCT Time (BOL) . 3160.3 2417.0 2209.8 1743.7 1386.4 1215.8 780.2 2209.2 Core Recovery N/A (2) (2) (2) (2) (2) (2) (2) (2)

Notes:

(1) Loop seal cleanng is defined as break vapor flow> Ilb/s.

(2) For the cases where core recovery is>TMAX basis for transient termination can be concluded based on some or all of the following: (1) The RCS system pressure is decreasing which will increase SI flow, (2)Total RCS system mass is increasing due to SI flow exceeding break flow, and (3) Core mixture level has begun to inc'rease and is'expected to continue for the remainder of the accident!

(3) It has been judged that no core uncovezryof any consequence will take place and the l.5-inch'case is'non-limniting. Tlherefore-no PCT calculations were performed.

5-20

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Table 5.2.2-7 BVPS-1 ,SBLOCA Results Summiay Peak Ciadding Temperatuiie (0F) 189 Maximum i Apl T iientOCidatioin(%i.) i i:07 Ttal Hydrogen Generation (%) <io 5-21

. - : I - _ _ _ _ _ _ __ _ _ _ __ _ _ _ l l FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 14 12

^ 10 0S 8 n

C) 1 6 4

2 Elevation (ft)

Figure 5.2.2-1 Small Break Hot Rod Power Shape 5-22

FENOC Enclosure IAttachment C REPLACEMENT STEAM GENERATOR L

Lumped Intact Loop Injected Flow

--- - Broken Loop In jected Flow 40-35 -

E

-o c 3:

0 U) v)

° 20-S. -

15 - S.

S.

.5 S.

.5 10 200 400 600 860 10oo 1200 1400 1600 1800 2000 Pressure (psio) s* - Figure 5.2.2-2 Small Break LOCA Safety Injection Flows (1 1H1SI pump, faulted loop injects to RCS pressure -11.5-inch 4-inch breaks for BVPS-1) 5-23

__ _ __ _ __ _ __ __ _ _ __ _ -- __ _ __ __ _ __ __ t jI .

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

- Un i t 1 Lumped I n I oc t L oop I n j ec t e d F I ow 350

()

E

-o 0

>2 U)

C)

Pressure (psia)

Figure 5.2.2-3 Small Break LOCA Safety Injection Flows (1 IIHISI pump, 1 LIISI pump, no ECCS in the faulted loop because the break is postulated along the HIISI line inch break for BVPS-1) 5-24

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR CORE PRESSURE, CORE FLOW, MIXTURE LEVEL, AND FUEL ROD POWER N HISTORY O<TIME<CORE COVERED 0 L T 0 4r, R C U T M A P

Figure 5.2.24 Code Interface Description for Small Break Model 5-25

_ 'li FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 2500 2000 o 1500 v) 0_

U)

U) 1000 500 0

Time (s)

Figure 5.2.2-5A BVPS-1 2.75-inch Break RCS Pressure 5-26

FENOC

_. _ Enclosure 1Attachment C REPLACEMENT STEAM GENERATOR

_- --- TOP OF CORE = 21.783 ft 35 -

30 l

_25 20

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BVPS-1 2.75-inch Break Core Mixture Level 5-27

. ___1111 FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 1400 1200 UA-E U)

Figure 5.2.2-7A BVPS-1 2.75-inch Break Core Exit Vapor Temperature 5-28

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Broken Loop

- - -- Intact Loop 1200-1100- if

.2 1000 qa, U)

=3 (I

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990 800 -

700 .- . .

0 100 2000 3000 4000 5000 6000 lime (s)

Figure 5.2.2-8A BVPS-1 2.75-inch Break Brokn Loop and Intaict Loop Secondary Pressure 5-29

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Figure 5.2.2-9A BVPS-1 2.75-inch Break Break Vapor Flow Rate 5-30

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 1400 1200 1000 U)

E 800 0

600 U) 400 200 0

Time (s)

Figure 5.2.2-iDA BVPS-1 2.75-inch Break Break Liquid Flow Rate 5-31

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Time (s)

Figure 5.2.2-11A BVPS-l 2.75-inch Break Broken Loop and Intact Loop Accumulator Flow Rate 5-32

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR Broken Loop

- - - -Intoct Loop 40 30 -

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Figure 5.2.2-12A BVPS-1 2.75-inch Break B,rok cn. . L.. . ..a..- -. a , P. . . . . S ..e. . j.c ... o F w Ra.- te Brken Loop and4 Intact Loop Pumnped'Safety Injection Flow Rate 5-33

_____IIIIL FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 2000 1800 1600 1400

' 1200 w

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- 1000 800 600 400 Time (s)

Figure 5.2.2-13A BVPS-1 2.75-inch Break Peak Clad Temperature 5-34

FENOC Enclosure 1Attachment C REPLACEMENT STEAM GENERATOR 1800 1600 1400 L_

1200 a)

I-C:I 10X0 800 600 400

_Figure 5.2.2-14A BVPS-l 2.75-inch Break iiots pot Fluid Temperature 5-35

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U-66000-r-

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2000 -

______*1-0 0 1000 2000 3000 4000 5000 6000 Time (s)

Figure 5.2.2-15A BVPS-1 2.75-inch Break Rod Film Heat Transfer Coefficient 5-36

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR w 1500 U) ca Ai Figure 5.2.2-16A BVPS-I 2-inch Break RCS Pressure 5-37

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

- -- - TOP OF CORE = 21.783 It 35 30-

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.25 20 -

15 I-. I . . I . I . -- - -

Time (s)

Figure 5.2.2-17A BVPS-1 2-inch Break Core Mixture Level 5-38

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

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_..,.,i, _ 5.2.2-i8A Figure BVPS-I 2-inch Break Peak Clad Temperature 5-39

_____11L FENOC

_ Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 0.

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=3 Time (s)

Figure 5.2.2-19A BVPS-1 2.25-inch Break RCS Pressure 540

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

- -- - TOP OF CORE = 21.783 ft 35 -

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w25-20-15- ,... .... .I..

Figure 5.2.2-20A BVPS-1 2.25-inch Break Core Mixture Level 541

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1800 -

1600 -

II nvJl AA-

-s 1200 a,

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I 1000 800 600 400 0 10000 20 3000 4000 5000 6000 7000 lime (s)

Figure 5.2.2-21A BVPS-1 2.25-inch Break Peak Clad Temperature 542

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 2500 -

2000 -

o2 1500 -

Cn U)

" 1M000 -

500 -

. I I I . . I . I I I. . .

0 0 idoo 2000 3000 4000 5000 6000 Time (s)

Figure 5.2.2-22A BVPS-i 2.5-inch Break RCS Pressure 5-43

_ _ _ _ . -- - __ IlRJL FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

- TOP OF CORE - 21.783 It 35 z

a> 25

.0_

Time (s)

Figure 5.2.2-23A BVPS-1 2.5-inch Brcak Core Mixture Level 5-44

FENOC Enclosure 1Attachment C REPLACEMENT STEAM GENERATOR 1800 1600 1400 Lo 1200 C)

C~-1000 E1 800 600 400 Time (s)

Figure 522-24A BVPS-1 72.5inch Break Peak Clad Temperature 545

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 2500 2000 1500 a)

U)

V) 1000 500 0

Time (s)

Figure 5.2.2-25A BVPS-I 3-inch Break RCS Pressure 5-46

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

-- -- TOP OF CORE = 21.783 ft 35

-iD a.)

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x Figure 5.2 .2-26A BVPS-I 3-inch Break Core Mixture Level 5-47

_ _ __ _ _ _ _ _ _ _ _ _ _ _ _l u L FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 1800 1600 1400 UW 1200 w

a)

E 1000 800 600 400 Time (s)

Figure 5.2.2-27A BVPS-I 3-inch Break Peak Clad Temperature 548

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 2500 -

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o 1500 -

0n 0-L..

U)

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CD 1000 -

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Figure 5.2.2-28A BVPS-1-3.25-inch Break RCS Pressure 549

111.

FENOCft Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

--- - TOP Of CORE = 21.783 ft 35 -

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-ZZ- -J

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-;~20 15 10 Figure 5.2.2-29A BVPS-I 3.25-inch Break Core Mixture Level 5-50

FENOC Enclosure 1 Attachment C REPLACEMENT STEAM GENERATOR 1800 M-Eu w

Time (s)

Figure 5.2.2-30A BVPS-1 3.25-inch Break Peak Clad Temperature 5-51

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. _a 1500

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Time (s)

Figure 5.2.2-31A BVPS-I 4-inch Break RCS Pressure 5-52

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

-- -- TOP OF CORE = 21.783 ft 35 -

30-U)

-J 10 -

15-10 I I Figure 5.2.2-32A BVPS-1 4-inch Break Core Mixture Level I.

5-53

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P ci, E 800 600 400 Time (s)

Figure 5.2.2-33A BVPS-I 4-inch Break Peak Clad Temperature 5-54

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR 2500 2000 c 1500

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O 1000 0-500 Time (s)

Figure 5.2.2-34A BVPS1 6-inch Break

.RCS Pressure 5-55

FENOC Enclosure I Attachment C REPLACEMENT STEAM GENERATOR

-- -- TOP OF CORE = 21.783 1 35 -

30

= 25-

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Time (s)

Figure 5.2.2-35A BVPS-1 6-inch Break Core Mixture Level 5-56

FENOC Enclosure 1 Attachment C REPLACEMENT STEAM GENERATOR 1400 1200 1000 U-800 a.)

co E

CD 600 400 200 Time (s)

Figure 5.2.2-36A BVPSI 6-inch Break Peak Clad Temperature 5-57

Attachment D Revised Extended Power Uprate Submittal (I.04-125) Enclosure 2 Section 5.2.2

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 5

5.1 5.2 5.2.1 5.2.2 Small Break LOCA 5.2.2.1 Introduction This section contains information regarding the Small Break Loss-of-Coolant Accident (SBLOCA) analyses performed in support of the EPU for BVPS-I (with Model 54F replacement steam generators) and for BVPS-2 at the NSSS power level of 2910 MWt (2900 MWt reactor power). The purpose of analyzing the Small Break LOCA is to demonstrate conformance with the 10 CFR 50.46 (Reference 1) requirements for the conditions associated with the EPU. Important input assumptions, as well as analytical models and analysis methodology for the Small Break LOCA are contained in' subsequent sections. Analysis results are provided in the form of tables and figures, as well as a more detailed description of the limiting transient. The analysis has shown that no design or regulatory limit related to the Small Break LOCA would be exceeded due to the EPU power and associated plant parameters.

5.2.2.2 Input Parameters and Assumptions The important plant conditions and features for BVPS-I and BVPS-2 are listed in Table 5.2.2-lA and Table 5.2.2-lB, respectively. Several additional considerations that are not identified in Table 5.2.2-IA or Table 5.2.2-lB are discussed below.

Figure 5.2.2-1 depicts the hot rod axial power shape modeled in the Small Break LOCA analyses. This shape was chosen because it represents a distribution with power concentrated in the upper regions of the core (the axial offset is +13%). Such a distribution is limiting for Small Break LOCA since it minimizes coolant swell while maximizing vapor superheating and fuel rod heat generation at the uncovered elevations. The chosen power shape has been conservatively scaled to a standard 2-line segment K(Z) envelope for BVPS- I and BVPS-2 based on the peaking factors shown in Table 5.2.2-1 A or 5.2.2-1 B.

Figures 5.2.2-2 and 5.2.2-3 'provide the SI flowversus-pressure curves modeled in the Small Break LOCA analyses 'Figure 5.2.2-2 shows'the flows fro'mii one High Head Safety Injection (HHSI p'ump,-where the faulted loop injects to RCS pressure., Figure 5.2.2-3 shows flows from 'one'HHSI pump' and 6one'Low Head Safety Injection'(LHSI) pump, where the faulted loop injects into containment.

5.2.2.3 Description of Analyses and Evaluations Analytical Model The requirements for an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50.

For LOCAs due to Small Breaks, less than I square foot in area, the Westinghouse NOTRUMP Small 5-9

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Break LOCA Emergency Core Cooling System (ECCS) Evaluation Model (References 2, 3, and 4) is used. The Westinghouse NOTRUMP Small Break LOCA ECCS Evaluation Model was developed to determine the RCS response to design basis Small Break LOCAs, and to address NRC concerns expressed in NUREG-0611 (Reference 5).

The Westinghouse Small Break LOCA ECCS Evaluation Model consists of the NOTRUMP and LOCTA-IV computer codes. The NOTRUMP code is employed to calculate the transient depressurization of the Reactor Coolant System (RCS), as well as to describe the mass and energy release of the fluid flow through the break. Among the features of the NOTRUMP code are: calculation of thermal non-equilibrium in all fluid volumes, flow regime-depcndent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, regime-dependent drift flux calculations in multiplc-stacked fluid nodes and regime-dependent heat transfer correlations. These features provide NOTRUMP with the capability to accurately calculate the mass and energy distribution throughout the RCS during the course of a Small Break LOCA.

The RCS model is nodalized into volumes interconnected by flow paths. The broken loop is modeled explicitly, while the intact loops are lumped together into a second loop. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum. The multi-node capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a calculation of the behavior of the loop seal during a Small Break LOCA. The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations.

Fuel cladding thermal analyses are performed with a version of the LOCTA-IV code (Reference 2) using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions. The LOCTA-IV code models the hot rod and the average hot assembly rod, assuming a conservative power distribution that is skewed to the top of the core. Figure 5.2.24 illustrates the code interface for the Small Break Model.

Analysis The EPU Small Break LOCA analyses considered nine different break cases each for BVPS- I and BVPS-2 as indicated by the results in Tables 5.2.2-6A and 5.2.2-6B, respectively. A break spectrum of 1.5-, 2-,

2.25-, 2.5-, 2.75-, 3-, 3.25-, 4-and 6-inch breaks was considered. For BVPS-I, the 2.75-inch break was found to be limiting for PCT and the 2.5-inch break was found to be limiting for oxidation. For BVPS-2, the 3-inch case was found to be limiting for PCT and the 2.5-inch break case was found to be limiting for oxidation. For both units, the 1.5-inch case was found to be non-limiting in NOTRUMP and therefore PCT information was not calculated.

The most limiting single active failure used for a Small Break LOCA is that of an emergency power train failure which results in the loss of one complete train of ECCS components. In addition, a Loss-of-Offsite Power (LOOP) is postulated to occur coincident with reactor trip. This means that credit may be taken for at most one high head safety injection (IIHSI) pump. In the analyses for BVPS-I and BVPS-2, one HHISI pump is modeled. The Small Break LOCA analysis performed for both units models the ECCS flow as being delivered to both the intact and broken loops at the RCS backpressure for breaks smaller than the cold leg HHSI nozzle (1.5-inch inch breaks) and at containment pressure for breaks greater 5-sl

FENQC Enclosure IAttachment D EXTENDED POWER UPRATE than the cold leg HHSI nozzle (6-inch breaks).'Thes 'SI flows are illustrated in Figuie 5.2.2-2 and 5.2.2-3 for each scenario. Note that for the 6-inch breaks, no SI is assumed in the faulted loop because the breaklis postulated along the SI line. The LOOP and the failure of a diesel generator to start as the limiting single failure for Small Break LOCA is part of the NRC approved methodology and does not change as a result of the EPU conditions. The single failure assumption is extremely limiting due to the fact that one train of SI, one motor driven auxiliary feedwater (AFW) pump, and power to the reactor coolant pumps (RCPs) are all modeled to be lost. Any other active single failure would not result in a more limiting scenario since increased SI flow would improve the overall transient results.

Prior to break initiation, the plant is in a full power (100.6%) equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. Other initial plant conditions used in the analysis are given in Table 5.2.2-1A or Table 5.2.2-1B. 'Subsequent to the break opening, a period of reactor coolant system blowdown ensues in which the heat from fission product decay, the hot reactor internals, and the reactor vessel continues to be transferred to the RCS fluid. The heat transfer between the RCS and the secondary system may be in either direction and is a function of the relative temperatures of the primary and secondary conditions. In the case of continuous heat addition to the secondary during a period of quasi-equilibrium, an increase in the secondary system pressure results in steam relief via the steam generator safety valves.

When a Small Break LOCA occurs, depressurization of the RCS causes fluid to flow into the loops from the pressurizer resulting in a pressure and level decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low-pressure reactor trip setpoint, conservatively modeled as 1935 psia, is reached. LOOP is postulated to occur coincident with reactor trip. A safety injection signal is generated when the pressurizer low-pressure safety injection setpoint, conservatively modeled as 1745 psia for BVPS-I and 1760 psia for BVPS-2, is reached. Safety injection flow is delayed 27 seconds after the occurrence of the low-pressure condition. This delay accounts for signal processing, diesel generator start up and emergency power bus loading consistent with the loss-of-offsite power coincident with reactor trip, as well as the pump acceleration and valve delays.

The following countermeasures limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection supplement void formation in causing a rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay. No credit is taken in the Small Break LOCA analysis for the boron content of the injection water. In addition, credit is taken in the Small Break LOCA analysis for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the reactor trip signal, considering the most reactive RCCA is stuck in the full out position. A rod drop time of 2.7 seconds was used while also considering an additional 2 seconds' for the signal processing delay time. Therefore, a total delay time of 4.7 seconds from the time of reactor trip signal to full rod insertion was used in the Small Break LOCA analysis.

2. Injection of borated water provides sufficient flooding of the core to prevent excessive cladding temperatures.

During the earlier part of the Small Break transient (prior to the postulated loss-of-offsite power coincident with reactor trip), the loss of flow through the break is not sufficient to overcome the positive core flow maintained by the reactor coolant pumps. During this period, upward flow through the core is 5-11

1111V F- -QC Enclosure IAttachment D EXTENDED POWER UPRATE maintained. However, following the reactor coolant pump trip (due to a LOOP) and subsequent pump coastdown, a period of core uncovery occurs. Ultimately, the Small Break transient analysis is terminated when the top of the core is recovered or the core mixing level is increasing, and ECCS flow provided to the RCS exceeds the break flow rate.

The core heat transfer mechanisms associated with the Small Break transient include the break itself, the injected ECCS water, and the heat transferred from the RCS to the steam generator secondary side. Main Feedwater (MFW) is conservatively isolated in 10 seconds for BVPS- I (consisting of a 3 second signal delay time and a 7 second main feedwater isolation valve stroke time) and 7 seconds for BVPS-2 (consisting of a 2 second signal delay time and a 5 second main feedwater isolation valve stroke time) following the generation of the pressurizer low-pressure SI signal. Additional makeup water is also provided to the secondary using the auxiliary feedwater (AFW) system. An AFW actuation signal is derived from the pressurizer low-pressure SI signal, resulting in the delivery of AFW system flow 60 seconds after the generation of the SI signal. The heat transferred to the secondary side of the steam generator aids in the reduction of the RCS pressure.

Should the RCS depressurize to approximately 625 psia (accumulator minimum pressure), the cold leg accumulators begin to inject borated water into the reactor coolant loops.

5.2.2.4 Acceptance Criteria and Results The acceptance criteria for the LOCA arc described in 10 CFR 50.46 (Reference I) as follows:

I. The calculated maximum fuel element cladding temperature shall not exceed 2200'F.

2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Criteria I through 3 are explicitly covered by the Small Break LOCA analysis at EPU conditions.

For criterion 4, the appropriate core geometry was modeled in the analysis. The results based on this geometry satisfy the Peak Clad Temperature (PCT) criterion of 10 CFR 50.46 and consequently, demonstrate that the core remains amenable to cooling.

5-12

__ Enclosure I Attachment D EXTENDED POWER UPRATE For criterion 5, Long-Term Core Cooling (LTCC) considerations are not directly applicable to the Small Break LOCA transient analysis addressed in this section, but are assessed by Sections 5.2.3 and 5.2.4 as part of the evaluation of ECCS performance.

The acceptance criteria were established to provide a significant margin in ECCS performance following a LOCA.

In order to determine the conditions that produced the most limiting Small Break LOCA case (as deterinied by the highest calculated peak cladding temnperature), nine break case's were examined for BVPS-1 and BVPS-2.`These cases were investigated to capture the nmo'sfevere postulated Small Break LOCA event. The following discussions provide insight into the analyzed conditions.

Limiting Temperature Conditions TheRCS temperature analyzed was based 'on'a nomal vessel average teperature of 580.0F. However, the analysis is applicable over the range of 566.2 - 580.0 0F. -The analysis supports a'+/-40F TiVE-uncertainty. The analysis showed thatfthe 2.75-inch break case is the limiting PCT cas'e for BVPS-l and the 3-inch break case is the limiting PCT case for BVPS-2. -The limiting case transients are discussed below.

Limiting Break Case The results of Reference 6 demonstrate that the cold leg break location is limiting with respect to postulated cold leg, hot leg and pump suction leg break locations. The PCT results are'shdwn in Table 5.2.2-A and 5.2.2-SA or Table 5.2.24B and Table'5.2.2-5B. Inherent in the Srriall Breakalysis are several input parameters (see Section'5.2.2.2 and Table 5.2.2-lA or Table 5.2.2-1 B), while Table 5.2.2-6A&or Table 5.2.2-6B provide thekey transient event times.

For the EPU Snail Break LOCA analysis, the limiting PCT case for BVPS-I was'the' 2.75-inch break case and the limiting PCT case for BVPS-2 w~as th'e 3-inch break case. 'A summary of the transient response for the limiting PCT case is shown in Figures 5.2.2-SA or Figure 5.2.2-SB through Figure

-15A orFige 5.2.2-'15B. Thesefigures present the response of the following parameters.

RCS Pressure

Core Mixture Level

Core Exit Vapor Temperature

Broken Loop and Intact Loop Secondary Pressure

Break Vapor Flow Rate

Break Liquid Flow Rate

Broken Loop and Intact Loop Accumulator Flow Rate

Broken Loop and Intact Loop Pumped Safety Injection Flow Rate

Peak Clad Temperature

Hot Spot Fluid Temperature

Rod Film Heat Transfer Coefficient 5-13

__ _ __ _ - . l ljI I FENOC Enclosure IAttachment D EXTENDED POWER UPRATE Upon initiation of the limiting 2.75-inch break for BVPS-I and the limiting 3-inch break for BVPS-2, there is an initial rapid depressurization of the RCS followed by an intermediate equilibrium at approximately 1150 psia (see Figure 5.2.2-SA or Figure 5.2.2-5B). The limiting 2.75-inch break depressurizes to the accumulator injection setpoint of 625 psia at approximately 1438 seconds for BVPS-1 (see Figure 5.2.2-11 A). The limiting 3-inch break depressurizes to the accumulator injection setpoint of 625 psia at approximately 1082 seconds for BVPS-2 (see Figure 5.2.2-11 B). During the initial period of the Small Break transient, the effect of the break flow rate is not sufficient to overcome the flow rate maintained by the reactor coolant pumps as they coast down. As such, normal upward flow is maintained through the core and core heat is adequately removed. Following reactor trip, the removal of the heat generated as a result of fission products decay is accomplished via a two-phase mixture level covering the core. The core mixture level and peak clad temperature transient plots for the limiting break calculations are illustrated in Figures 5.2.2-6A or 5.2.2-6B and 5.2.2-13A or 5.2.2-13B, respectively. These figures show that the peak clad temperature occurs near the time when the core is most deeply uncovered and the top of the core is being cooled by steam. This time is characterized by the highest vapor superheating above the mixture level (refer to Figure 5.2.2-7A or 5.2.2-7B). For BVPS-I the limiting PCT time-in-life was determined to be 8,000 MWD/MTU. For BVPS-2 the limiting PCT time-in-life was determined t6 be 6,500 MWD/MTU.

A comparison of the flow provided by the safety injection system to the intact and broken loops can be found in Figure 5.2.2-12A or 5.2.2-12B. The cold leg break vapor and liquid mass flow rates are provided in Figures 5.2.2-9A or 5.2.2-9B and 5.2.2-1OAor 5.2.2-LOB, respectively. Figures 5.2.2-14Aor 5.2.2-14B and 5.2.2-15A or 5.2.2-1SB provide additional information on the fluid temperature at the hot spot and hot rod surface heat transfer coefficient at the hot spot, respectively. Figure 5.2.2-8A or 5.2.2-8B depicts the secondary side pressure for both the intact and broken loops for the limiting PCT break case.

Total Oxidation For the EPU Small Break LOCA analysis, the maximum local oxidation case for BVPS-1 and BVPS-2 was the 2.5-inch break case. The maximum local transient oxidation is 11.07% for BVPS-I at 20,000 MWD/MTU and 13.42% for BVPS-2 at 15,000 MWD/MTU. The limiting transient oxidation occurs at the burst elevation and includes both outside and post-rupture inside oxidation. Pre-existing (pre-transient) oxidation was also considered and the sum of the pre-transient and transient oxidation remains below 17% at all times in life, for all fuel resident in the core.

Additional Break Cases Studies documented in Reference 6 have determined that the limiting PCT Small Break transient occurs for breaks of less than 10-inches in diameter in the cold leg. For BVPS-I and BVPS-2, the limiting PCT is captured by the 1.5- 2-, 2.25-, 2.5-, 2.75-, 3-, 3.25-, 4- and 6-inch break spectrum. The beginning-of-life (BOL) results of these break spectrum cases are given in Table 5.2.2-4A or Table 5.2.24B. Figures 5.2.2-16A through 5.2.2-36A address the non-limiting BOL cases (2--,2.25-, 2.5-,3-,3.25-,4-and 6-inch) analyzed for BVPS-I. Figures 5.2.2-16B through 5.2.2-36B address the non-limiting BOL cases (2-,

2.25-, 2.5-, 2.75-, 3.25-, 4- and 6-inch) analyzed for BVPS-2. The 1.5-inch cases for BVPS-l and BVPS-2 produced only minimal core uncovery and therefore PCT information was not calculated. Note that plots for the 1.5-inch cases are not included here. The plots for each of the additional non-limiting break cases include:

5-14

FENOC Enclosure IAttachment D EXTENDED POWER UPRATE

1. RCS Pressure

2. Core Mixture Level

3. Peak Clad Temperature For BVPS-1 I the PCTs for each of the additional breaks considered 'ar'eshown in Table 5.2.2-4A"and PC s are less than the limiting 2.75-inch break case. For BVPS-2, the PCTs for the additional breaks are shown in Table 5;2.2-4B and are 'esIthan the limiting 3-mich break case. The PCT was not calculated for either BVPS-l or BVPS:2 for the 1.5-inch case due to the minimial core uncovery.

Transient Termination The 10 CFR 50.46 criteria-continue to be' satisfied beyond the end of the calculated transient due to the presence of some or'all of the following conditions:

1. The RCS pressure is gradually decreasing.

2. The net mass inventory is increasing.

3. The core mixture level is recovered, or recovering due to increasing mass inventory.

4. As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel cladding temperatures will continue to decline indicating that the temperature excursion is terminated.

ZIRLO/Zirc4 Cladding At the time at which this analysis is implemented, no new Zirc-4 fuel is expected to be inserted into the core. All of the Zirc-4 fuel will be burned for at least one cycle, and ZIRLOTM fuel will be implemented at non-EPU conditions at least one reload cycle before the EPU is implemented. Therefore, the ZIRLOTM fuel is considered limiting with a PCT of ,1895.00F at 8,000 MWD/MTU burnup for BVPS-I and with a PCT of 1917.;1F at'6,500 MWD/MTU) for BVPS-2. The fuel temperatures/pressures used in these calculations were based on NRC approved fuel performance code PAD 4.0 (Reference 7) which addresses all the helium release related issues. This anaIysis has been 'performed using the most limiting temperature/pressure as calculated for 17x17 non.-IFBA RFA fuel.' The non-IFBA fuel bounds IFBA'fuel for Small Break LOCA an'aiyses. Note that the effect of annular pellets, which are only'present in the IFBA fuel, was considered in-the analyses.

5.2.2.5 Conclusions The Small BreakLOCA analyses 'considered a break spectrum of 1.5-, 2-,2.25-,'2.5-,'2.75-,:3-,3.25-, 4-and 6-inch diaiiieters for BVPS-1 and BVPS-2. 'For BVPS-l ,'a peak cladding teimperatuie of 1895 0 F was calculated at the limiting time-in-life of 8500 MWD/MTU for the 2.75-inch case and a maximuin transient oxidation of 11.07% was calculated at the limiting time-in-life of 20,000 MWD MTU for the 2.5-inch case. ForBVPS-2, a peak cladding temperature of 1917 0F was calculated at the limiting time-in-life of 6500 MWD/MTU for the 3-inch case and a maxim'um transient oxidation of 13.42% was calculated at the limiting timei-in-life of 15,000 MWD/MTU for the 2.5-inch case.

The analyses presented in this section show that the accumulator and safety injection subsystems of the Emergency Core Cooling System, together with the heat removal capability of the steam'generator, 5-15

FENOC Enclosure IAttachment D EXTENDED POWER UPRATE provide sufficient core heat removal capability to maintain the calculated peak cladding temperatures for Small Break LOCA below the required limit of 10 CFR 50.46. Furthermore, the analyses show that the local cladding oxidation and core wide average oxidation, including consideration of pre-existing and post-LOCA oxidation, and cladding outside and post-rupture inside oxidation, are less than the 10 CFR 50.46 (Reference 1) limits.

Table 5.2.2-7 provides a results summary for the BVPS-I and BVPS-2 SBLOCA EPU analyses. Results include PCT, maximum local oxidation and total hydrogen generation.

The results and conclusions of the analyses performed for Small Break LOCA for the reactor power of 2900 MWt (2910 MWt NSSS power) bound and support operation at the current reactor power of 2689 MWt (2697 MWt NSSS power), thus supporting the staged implementation of EPU at BVPS-I and BVPS-2.

5.2.2.6 References

1. "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974, as amended in Federal Register, Volume 53, September 1988.

2. Meyer, P. E., "NOTRUMP - A Nodal Transient Small Break and General Network Code,"

WCAP-10079-P-A, (proprietary) and WCAP-10080-NP-A (non-proprietary), August 1985.

3. Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (proprietary) and WCAP-10081-NP-A (non-proprietary),

August 1985.

4. Thompson, C. D. et al., Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," WCAP-10054-P-A, Addendum 2, Rev. I (proprietary), July 1997.

5. "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plant," NUREG-061 1, January 1980.

6. Rupprecht, S. D. et al., "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP- 1145-P-A (proprietary), October 1986.

7. Slagle, W. H., (ed.) et al., "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," WCAP-15063-P-A, Revision 1, July 2000.

5-16

FE O% Enclosure I Attachment D EXTENDED POWER UPRATE Table 5.2.2-IA BVPS-1 Input Parameters Used In the Small Break LOCA Analysis Input Parameter Value Core Rated Thermal Power-1 00% 2900 Calorimetric Uncertainty, % 0.6 Fuel Type 17 X 17 Robust Fuel Assembly (RFA)

Total Core Peaking Factor, FQ 2.40 Hot Channel Enthalpy Rise Factor, FAX 1.62 Hot Assembly Average Power Factor, P1 AI.A2 Maximum Axial Offset, % +13 Initial RCS Loop Flow, gpm/loop 82,840 Initial Vessel Tags, OF Max: 580.0 Min: 566.2 Initial Pressurizer Pressure (plus uncertainties), psia 2300 Reactor Coolant Pump Type Model 93A with Weir Pressurizer Low-Pressure Reactor Trip Setpoint, psia 1935 Reactor Trip Signal Delay Time, seconds 2.0 Rod Drop Delay Time, seconds 2.7 Auxiliary Feedwater Temperature (Maximum), IF 120 Number of AFW Pumps Available Following a LOOP I Motor Driven AFW Flow (Minimum) to all 3 Steam Generators, gpm 294 (98 gpm/SG

3) at 1107 psig AFWV Flow Delay Time (Maximum), seconds 60 AFW Actuation Signal Pressurizer Low-Pressure Safety Injection Steam Generator Type Model 54F 3

Maximum AFW Piping Purge Volume, ft 168 Steam Generator Tube Plugging (Maximum), % 10 Maximum MFW Isolation Signal Delay Time, seconds 3 MFW Control Valve Isolation Ramp Time, seconds 7 MFW Isolation Signal Pressurizer Low-Pressure Safety Injection Isolation of Steam Line Signal Pressurizer Low-Pressure Reactor Trip/LOOP Steam Generator Secondary Water Mass, lbm/SG 99,930 Containment Spray Flowrate for 2 Pumps, gpm 4983 (plus 981 gpm to account for flow to the sump)

RWST Deliverable Volume (Minimum), gallons 317,000 5-17

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Table 5.2.2-1A (continued)

BVPS-I Input Parameters Used in the Small Break LOCA Analysis Input Parameter Value SI Temp at Cold Leg Recirculation Time (Maximum), 'F 190 ECCS Conf iguration 1 HHSI pump, faulted line injects to RCS pressure (1.5-inch inch breaks)

I HHSI pump, I LHSI pump, no ECCS in the faulted loop (6-inch break)

ECCS Water Temperature (Maximum), 'F 65 Pressurizer Low-Pressure Safety Injection Setpoint, psia 1745 S1 Flow Delay Time, seconds 27 ECCS Flow vs. Pressure See Tables 5.2.2-2 and 5.2.2-3 Initial Accumulator Water/Gas Temperature, 'F 105 Initial Nominal Accumulator Water Volume, 113 957 Minimum Accumulator Pressure, psia 625 5-18

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Table 5.2.2-lB BVPS-2 Input Parameters Used in the Small Break LOCA Analysis Input Parameter Value Core Rated Thermal Power-I 00% 2900 Calorimetric Uncertainty, % 0.6 Fuel Type 17 X 17 Robust Fuel Assembly (RFA)

Total Core Peaking Factor, FQ 2.40 Hot Channel Enthalpy Rise Factor, FAI 1.62 Hot Assembly Average Power Factor, P1A 1.42 Maximum Axial Offset, % +13 Initial RCS Loop Flow, gpm/loop 82,840 Initial Vessel T*vg, OF Max: 580.0 Min: 566.2 Initial Pressurizer Pressure (plus uncertainties), psia 2300 Reactor Coolant Pump Type Model 93A with Weir Pressurizer Low-Pressure Reactor Trip Setpoint, psia 1935 Reactor Trip Signal Delay Time, seconds 2.0 Rod Drop Delay Time, seconds 2.7 0

Auxiliary Feedwater Temperature (Maximum), F 120 Number of AFW Pumps Available Following a LOOP I Motor Driven AFW Flow (Minimum) to all 3 Steam Generators, gpm 294 (98 gpm/SG *3) at 1107 psig AFW Flow Delay Time (Maximum), seconds 60 AFW Actuation Signal Pressurizer Low-Pressure Safety Injection Steam Generator Type Model 5 IM 3

Maximum AFW Piping Purge Volume, ft 125.7 Steam Generator Tube Plugging (Maximum), % 22 Maximum MFW Isolation Signal Delay Time, seconds 2 MFW Control Valve Isolation Ramp Time, seconds 5 MFIV Isolation Signal Pressurizer Low-Pressure Safety Injection Isolation of Steam Line Signal Pressurizer Low-Pressure Reactor Trip/LOOP Steam Generator Secondary Water Mass, lbm/SG 99,500 Containment Spray Flowrate for 2 Pumps, gpm 4450 RWST Deliverable Volume (Minimum), gallons 403,000 5-19

- ho -

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Table 5.2.2-1B (continued)

BVPS-2 Input Parameters Used in the Small Break LOCA Analysis Input Parameter Value Si Temp at Cold Leg Recirculation Time 212 (Maximum), 0F ECCS Configuration I HHSI pump, faulted line injects to RCS pressure (1.5-inch inch breaks)

I HHSI pump, I LHSI pump, no ECCS in the faulted loop (6-inch break)

ECCS Water Temperature (Maximum), 'F 65 Pressurizer Low-Pressure Safety Injection 1760 Setpoint, psia Si Flow Delay Time, seconds 27 ECCS Flow vs. Pressure See Tables 5.2.2-2 and 5.2.2-3 Initial Accumulator Water/Gas Temperature, 'F 105 3

Initial Nominal Accumulator Water Volume, 997 Minimum Accumulator Pressure, psia 625 5-20

FENOC Enclosure 1Attachment D EXTENDED POWER UPRATE Table 5.2.2-2 Safety Injection Flows Used in the Small Break LOCA Analysis

( pumpfaulte loopi toRCSpr -. 5in inch breas for BVPS1 and WVPSr2)

RCS Pressure (psia) Intact Loop (Ibm/sec) Broken Loop (ibm/sec) 314.7 37.59 20.28 414.7 36.63 19.79 514.7 35.56 19.17 614.7 34.45 18.61 714.7 33.42 18.06 814.7 32.34 17.50 914.7 31.25 16.88 1014.7 30.14 16.25 1114.7 29.03 15.70 1214.7 27.92 15.07 1314.7 26.67 14.45 1414.7 25.28 13.61 1514.7 23.85 12.92 1614.7 22.43 12.08 1714.7 20.97 11.39 1814.7 19.50 10.56 5-21

111' FENOC Enclosure IAttachment D EXTENDED POWER UPRATE Table 5.22-3 Safety Injection Flows Used in the Small Break LOCA Analysis (1 IIISI pump, 1 LHSI pump, no ECCS in the faulted loop because the break is postulated along the HHSI line inch breaks for BVPS-1 and BVPS-2)

RCS Pressure (psia) BVPS-1 Intact Loop (Ibmlsec) BVPS-2 Intact Loop (Ibmlsec) 14.7 336.34 376.26 24.7 313.32 353.74 34.7 290.45 330.07 64.7 216.34 250.07 104.7 60.64 109.7 34.35 114.7 45.12 34.35 119.7 34.73 164.7 34.35 214.7 34.73 34.35 314.7 32.43 32.23 414.7 30.14 30.14 514.7 27.78 27.99 614.7 25.42 25.7 714.7 22.92 23.41 814.7 20.42 20.97 914.7 17.78 18.61 1014.7 15.00 16.11 1114.7 12.22 13.47 1214.7 9.31 10.83 1314.7 6.39 7.92 1414.7 3.06 5.0 1514.7 0.0 5-22

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Jiable 5.2.24A BVPS1,SBLOCTA BOL Results Break Size'(in) 2 .2i5 Y_ 2.75 3 3.25 i 6 0

_CI' 1 F~ ~723.i 1804; 78 8391i032 7 1267.9 PC r Time (s) 3160.3 2417i.0 229.8 i7ii3.i 86.4 1215.8 780.2 2209.2 PCTElevaton^(fl) 12 P 12 12 11.75 11.5 11.25 1.5 Max. Local ZrO2 d(%) 315 3.6 <i04 4.28 2.87 1.19 0.14 0.3 Max'.l~r 2o02tiev.(ftl) I 12 ' 5 1.75 111.5 1125 i.5

-;rJW4de1Avg. bb oii8 0.5 0.4 01A8 '7 0.0o ' 0.2 41able 5.2.2-5A_

BVPS-1 SBLOCTA Limiting Results from the 2.5-inch (Transient O)xidation) snd 2.75-inch (PCI) Tinie-inLife Study Brek Sizc 2.5 2.75 Time-in-Life WD/MTU) 20,000 8,000 PC__96(OF) 1895.0 PCI'Tine (s) 2212.9 1723.7 PCT,Elevation (ii) 112 :12 HiiroiR B

-rs' Tii5-e() 1770.2 1721.7 Hot Rod Burst Eievation (ii) 11.75 12 Max.'Locai Transient ZrO2 1L7

-. 8.82 Mx. Local TigtZO 2 EI2 E?(fl) ii.75 'i2 Core-Wide 'g' ZrO2 t(%) ' 0.62 0.52 5-23

________ _______i I FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Table 5.2.241B BVPS-2 SBLOCTA BOL Results Break Size (in) 2 2.25 2.5 2.75 3 3.25 4 6 PCT (°F) 1752.9 1846.2 1838.8 1829.3 1852.6 1711.8 1455.7 899.9 PCT Time (s) 3220.2 2317.6 2118.9 1680.8 1329.4 1151.1 723.5 3274.6 PCT Elevation (fl) 12 12 12 12 12 11.75 11.5 11.25 Max. Local ZrO2 (%) 3.83 4.3 5.95 4.74 3.74 1.9 0.37 0.01 Max. Local ZrO2 Elev. (Q) 12 12 12 12 11.75 11.75 11.25 11.25 Core-Wide Avg. ZrO2 (%) 0.48 0.56 0.74 0.62 0.52 0.29 0.06 0.00 Table 5.2.2-5B BVPS-2 SBLOCTAL Limiting Results from the 2.5-inch (Transient Oxidation) and 3-inch (PCT) Time-in-Life Study Break Size 2.5 3 Time-in-Life (MWDID/MTU) 15,000 6,500 PCT(°F) 1845.1 1917.1 PCT Time (s) 2118.9 1316.6 PCT Elevation (fl) 12 12 Hot Rod Burst Time (s) 1725.2 1314.4 Hot Rod Burst Elevation (fl) 11.75 12 Max. Local Transient ZrO 2 (%) 13A2 7.79 Max. Local Transient ZrO2 Elev. (fl) 11.75 12 Core-Wide Avg. ZrO2 (%) 0.77 0.49 5-24

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

_Table 5.2.2-6A BVPS-1 NOTRUMP Resuits 152 -2i 25- i.75. _ .5 Evwent Time (sec) inch 2-inch inch inch inch 3-inch inch 4-nin3 6-indi Break Initiation 0 D 0 0 0 0 0 0 Reactor Trip Signal 54.1 29.0 22.5 P.9 5i4.6 123 10.6 7.3 4t S-Signal 75.9 42.4 33.9 27.9 23.8 20.8 18.6 '14.4 10 2 Sl FlowDelivered i02.9 694 60 .9 5.9 50.8 47.8 45.6 4.4 3i.2 Lo6p Sealcleariig? 1840 930 725 656 h84 4i4 358 241 62 CorUncovery t3) 020 813 '658 672 526 362 233 T121 uiiiatoirinjecti6o N/A 017 2378 82i 1438 1138 996 637 291 RWSTVoluine Delivered 3033 3025 MO 7 3011 3006 3001 2998 2992 -N/A PC3 Tine (BOI; . 31603 2417.0 2209.8 1743 i 1386.i 1215.8 780.2 22092 Core Recovery N/A (2) jj2f) (2) z2 z2) (2) ____

Notes:

(1) Loop seal clearing is defined s break vapor flow >1 Ib/s.

6 the cases where core recovery is > TM For asis for transient terminationcani be conciuded based on sorme or all of the following: (Ij The RCS system pressure is decreasirg which will increase SI flow, (2) Total RCS system mnass is increasinFlue to SI flow exceeding break flow, and (3) ECore mixture level has begun to increase'and is expected to continue for the remainder of the accident.

3 it has judged that n core any conuenc wll tae ko pace and the .5-inch case isnon-litg. Thefore o PCT calculations were performed.

5-25

I I'l L FENOC Enclosure IAttachment D EXTENDED POWER UPRATE Table 5.2.2-6B BVPS-2 NOTRUMP Results 2.25- 2.5- 2.75- 3.25-Event.Time (sec) inch 2-inch Inch inch inch 3-inch inch 4-Inch 6-inch Break Initiation 0 0 0 0 0 0 0 0 0 Reactor Trip Signal 57.7 30.6 23.8 18.9 15.3 12.6 10.9 7.5 4.4 S-Signal 74.7 42.1 33.8 28.1 24.1 20.9 18.9 15.3 11.1 SI Flow Delivered 101.7 69.1 60.8 55.1 51.1 47.9 45.9 42.3 38.1 Loop Seal Clearing(;) 1887 910 720 549 452 372 384 207 110 Core Uncovery (3) 1169 882 689 577 611 560 280 252 Accumulator Injection N/A 3494 2278 1714 1340 1082 915 557 261 RWST Volume Delivered 5076.7 5039.8 5025.5 5014.3 5004.7 4998.0 4991.8 N/A N/A PCT Time - 3220.2 2317.6 2118.9 1680.8 1329.4 1151.1 723.5 3274.6 Core Recovery N/A 5687 5599 (2) (2) (2) (2) (2) 3564 Notes:

(1) Loop seal clearing is defined as break vapor flow > 1 lb/s.

(2) For the cases where core recovery is > TMAX, basis for transient termination can be concluded based on some or all ofthe following: (1) The RCS system pressure is decreasing which will increase SI flow, (2) Total RCS system mass is increasing due to SI flow exceeding break flow, and (3) Core mixture level has begun to increase and is expected to continue for the remainder of the accident.

(3) It has been judged that no core uncovery of any consequence will take place and the 1.5-inch case is non-limiting. Therefore no PCT calculations were performed.

5-26

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE lTable 5.2.2-7 BVPS-1 adBVPS :2 SBLOCA Results Summary

.BVPS-i BV.PS-2 0

Peak Cladding Teeipature ( F) 1895 1917 Maxirnm Loca Tranient Oxidation 11iI.07 13:42 lTotal Hydroigehn Generation (%) -__

i__%o 5-27

FENOC t _a Enclosure I Attachment D EXTENDED POWER UPRATE 14 12 10

-Zl 0

Q.)

0~

- 6 4

2 Elevation (ft)

Figure 5.2.2-1 Small Break Hot Rod Power Shape 5-28

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE Lumped Intoct Loop Injected Flow

broroken Loop Injec led Flow 40 35

-i 30 E

0 25 3-0 cn o 20-

.1 15 - "I II I - I I - . I . I I . I-10 -

200 400 600 800 1000 12iOO 1400 1600 1800 2000 Pressure (psia)

IFigure 5.2.2-2

. . Small Break LOCA Safety Injection Flows (1 HHSI pump,ufaulted loop injects to RCS pressure - 1.5-inch inch breaks for BVPS-1 and BVPS-2) 5-29

- - l -- __ I 1_

11_

FENOC Enclosure I Attachment D EXTIENDED POWER UPRATE Un i t I L ump e d I n t ac I L o op I n j e c I e d F I ow Un i t 2 L umped I n tac t L oop I n j ec t ed F I o w 40 300 E

c 200-V)l U) 100-0 200 400 600 800 1000 1200 1400 1600 Pressure (psia)

Figure 5.2.2-3 Small Break LOCA Safety Injection Flows (1 1111SI pump, 1 LIISI pump, no ECCS in the faulted loop because the break is postulated along the 1111SI line inch breaks for BVPS-1 and BVPS-2) 5-30

FENOC

-h

_ _ Enclosure I Attachment D EXTENDED POWER UPRATE CORE PRESSURE, CORE FLOW, MIXTURE LEVEL, AND FUEL ROD POWER N HISTORY O<TIME<CORE COVERED 0 L T 0 R C U T M A P

Fig-ui're 5.2.24 Code Interface Description for Small Break Model 5-31

-- __ __ 1j2-FENOC Enclosure I- Attachment D EXTENDED POWER UPRATE

= =- - ----- -- -

2500 2000 U 1500 v')

0-Q>

V)

U) m 1000 500 0

Time (s)

Figure 5.2.2-5A BVPS-1 2.75-inch Break RCS Pressure 5-32

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500-2000 -

-o 1500-en U) m1000-500-0* I.. . . .. . . .. . . .. . ...

0 Idoo 2000 3000 4000 50 Time (s)

- Figure 5.22-5B BV`PS-2 3-inch Break RCS Pressure 5-33

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

- - -- TOP OF CORE = :1.783 fI 35 -

30

-25 X,

20-

,x I Time (s)

Figure 5.2.2-6A BVPS-I 2.75-inch Break Core Mixture Level 5-34

FENOC Enclosure IAttachment D EXTENDED POWER UPRATE

- - -- TOP OF CORE = 21.7862 ft 35 x- 25

-J5 VX20 Time (s)

Figure 5.2.2-6B BVPS-2 3-inch Break Core Mixture Level 5-35

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1400 1200 I t- 1000 0~

I--

E w~ 800 600 400 Figure 5.2.2-7A BVPS-1 2.75-inch Break Core Exit Vapor Temperature 5-36

FENOC Enclosure 1 Attachment D EXTENDED POWER UPRATE

_s L-0Z) 0-E CU.

Figure 5.2.2-7B

. BVPS-2 3-inch Break Core Exit Vapor Temperature 5-37

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE UB r oke n Loop

--- -EI n ac IL oop 1200 -

1100 - L

.2 U')

1000-0)

U)

U) 0n 800 -

700 -

6 0doo 2000 3000 4000 5000 6000 Time (s)

Figure 5.2.2-8A BVPS-1 2.75-inch Break Broken Loop and Intact Loop Secondary Pressure 5-38

FENOC Enclosure 1 Attachment D EXTENDED POWER UPRATE sBroken Loop

--- - Intact Loop 1200 1100

.2 1000 En CD U')

Co 900' 0-800 700 -

1........I.... .... 1. ...

6 1000 2000 3000 4000 5660 Time (s)

Figure 5.2.i-8B BVIPS-2 3-Inch Break Broken Loop and Intact Loop Secondary Pressure 5-39

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE llU .

80-E

-60 60-

-)

Ci)

Cn 40- 0f 20 -

0 0 1000 2000 3000 4000 500 6000 Time (s)

Figure 5.2.2-9A BVPS-1 2.75-inch Break Break Vapor Flow Rate 5-40

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 120 100 80 E

a>

cU)

En 640 20 o0 Time (s)

Figure 5.2.2-9B BVPS-2 3-inch Break Break YVap or Flow Rate 541

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1400 1200 1000 tn E

-o 800 W

C) 3 0 600 is M°E 400 200 0

Time (s)

Figure 5.2.2-1OA BVPS-I 2.75-inch Break Break Liquid Flow Rate 542

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1600

(-I E

-o C) 3-0 U) 0)

Figure 5.2.2-1OB BVPS-2 3-inch Break Break Liquid Flow Rate 543

- - _ _ _ _ _ _ _ _ -- I i L FENOC Enclosure IAttachment D EXTENDED POWER UPRATE

- ~ B r ooke n L oop

___-- I n Iact I L oop 300 250 200 E

-)

-2 0 150 0

Lo Cn C 1)

C I 00 Ii 50

_ _11 I__

0 0 idoo 2000 3000 4000 5000 6C0O Time (s)

Figure 5.2.2-1lA BVPS-I 2.75-inch Break Broken Loop and Intact Loop Accumulator Flow Rate 544

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE E

Broken Loop

_--- I n t ac t L oop 400 300 -

U)

E

-o p 200-I

.I 0

I U-I I

0 I I

I 100 - I I II

~IiA 0- __

I. ..

II . . . . i . . .

. I A. . ,. I I I I -1 1000 2000 3000 4000 5000 Time (s)

Figure 5.2.2-11B BVPS-2 3-inch Break Broken Loop and Intact Loop Accumulator Flow Rate 545

-- -- IM-FENOC Enclosure I Attachment D EXTENDED POWER UPRATE IEr olken 1 oop

_ -- -I n t a c tL oop 40 -

t/

30 -

n) I/

I/

E I/

0 20 -

- I IX I

0 U)

In U')

Mn 10 -

0 0 10I00 2000 3000 4000 5000 6000 Time (s)

Figure 5.2.2-12A BVPS-1 2.75-inch Break Broken Loop and Intact Loop Pumped Safety Injection Flow Rate 546

FENOC Enclosure IAttachment D EXTENDED POWER UPRATE

-~ Broken Loop

--- - I n t a c tL o o p 40 -

,f A//

30 - I/

I E Ir - _'-1

-M co 0

10 -

U ,

0 1000 2000 30.0 4000 51D0 Time (s)

- Figure 5.2.2-12B BIOPSY 3-ln h Break Broken Loop and Intact Loop Pumped Safety Injection Flow Rate 547

ail FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2000 1800 1600

- 1400

-1 a) 03 E

1000 800 600 400 Time (s)

Figure 5.2.2-13A BVPS-1 2.75-inch Break Peak Clad Temperature 548

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE GALA JIX),

1800 -

1600 -

1400-

- 1200-ca.

E w

1000-

!I 800 -

600 -

400 0 IObo 2000 300O 4000 5000 Time (s)

Figure 52.2-B BVPS-2 3-inch Break Peak Clad Temperature 549

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1800 1600 1400 1200

=3 a) 800 600 400 Time (s)

Figure 5.2.2-14A BVPS-1 2.75-inch Break Hot Spot Fluid Temperature 5-50

FENOC

._ Enclosure I Attachment D EXTENDED POWER UPRATE 2000 1800 -

1600 U-1400 1200 Q

0 E

co 1000 800 600 Figure 5.2.2-14B BVPS-2 3-inch Break HotSpot, Hopot'Fluid ep- eratu

- Temperature 5-51

-- -- ~ .1 FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 10000 -

8000 -

I-r-

a 6000-cn 0

C-)

a3 4000-u',

C I-2000 -

0 .1 U 1000 2000 3000 4000 5000 6000 Time (s)

Figure 5.2.2-15A BVPS-1 2.75-inch Break Rod Film Ileat Transfer Coefficient 5-52

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 10000 -

8000 -

1L c-6000 -

I-L..

4000 -

I 2000 -

0-1

-2000 U 1000 2000 3000 4000 5000 Time (s)

Figure 5.2.2-15B BVPS-2 3-inch Break Rod Film Heat Transfer Coefficient 5-53

FENOC

. A..

_ Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000 cn as 1500 V)

U, 1000 500 Time (s)

Figure 5.2.2-16A BVPS-1 2-inch Break RCS Pressure 5-54

FENOC

.~ _ Enclosure I Attachment D EXTENDED POWER UPRATE 2500-2000 -

0 e> 1500-u)

I-1000 -

500 -+

1000 2000 3A00 4000 5000 6000 Time (s)

Figure 5.2.2-16B BVPS-2 2-inch Break RCS Pressure 5-55

lilts I

FENOC Enclosure I Attachment D EXTENDED PONVER UPRATE TOP OF CORE = 21.783 It 35 -

30 -

"25 20-15 . . . '

Time (s)

Figure S.2.2:17A BVPS-1 2-inch Break Core Mixture Level 5-56

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.7862 ft 35 Z-a)

-a-' 25 cu x

Time (s)

. I.. . -. . - .....

1.

Figure 5.2.2-17B BVPS-2 2-inch Break Core Mixture Level 5-57

- I tIJIHI FENOC Enclosure I Attachment D EXTENDED POWVER UPRATE 1800 U-0W 0-E Time (s)

Figure 5.2.2-18A BIvPSY- 2-inch Break Peak Clad Temperature 5-58

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1800 1600 1400 U-1200 CD L.-

0 C1-1000 800 600 400 Time (s)

Figure 5.2.2-18B BVPS-2 2-inch Break Peak Clad Temperature 5-59

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 -

2000 -

U,

.(nC w 1500-CI, 1000 -

. . I . . . . . . . . . . . .

500 - .

0 1000 2000 300 4000 5000 6000 Time (s)

Figure 5.2.2-19A BVPS-1 2.25-inch Break RCS Pressure 5-60

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000 1o

.2_

0-a) 1500 en LI) 0-1000 500 Time (s)

Figure 5.2.2-19B BVPS-2 2.25-inch Break

-RCS Pressure 5-61

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP or CORE = 21.783 It 35 -

30 wt250 15-Time (s)

Figure 5.2.2-20A BVPS-1 2.25-inch Break Core Mixture Level 5-62

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.7862 ft 35 -

30 -

w25

.X _ _ 1--- ----------

20 -

15-Time (s)

Figure 5.2.2-20B BVPS-2 2.25inch Brea-k Core Mixture Level 5-63

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2000 -

1800 -

1600 -

_ 1400-U-

- 1200 -

0~M-a, E

H- 1000-800 -

600 -

I .. I. . . . . I II 400 -t l 1000 2000 3000 4000 5000 6000 7000 Time (s)

Figure 5.2.2-21A BVPS-1 2.25-inch Break Peak Clad Temperature 5-64

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2000 1800 1600

- 1400 LL 0,

-c 0

1200 E

'- 1000 800 600 400 Figure 5.2.2-21B BVPS-2 2.25-inch Break Peak Clad Teiperature 5-65

I i I FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 -

2000 -

o 1500-0 5 Oa)

3 U,)

cl)

"1000 -

500 -

. . I I I . I 0

1doo 2000 3000 4000 5000 6000 Time (s)

Figure 5.2.2-22A BVPS-I 2.5-Inch Break RCS Pressure 5-66

FENOC

_ _ Enclosure I Attachment D EXTENDED POWER UPRATE O

U2 1500 Cn a_

U)

En C 1000 Time (s)

Figure 5`.2.-i2B BVPS-2 2.5-Inch Break RCS Pressure 5-67

~i 1lulL FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.783 ft 35 -

30 51 a.,t

-III 20 15 . r I I Time (s)

Figure 5.2.2-23A BVPS-1 2.5-inch Break Core Mixture Level 5-68

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.7862 ft 35 -

30

,25 I20- - l\

20 15- . I I . I I I . ....

Figure 5.2.2-23B BVPS-2 2.5-inch Break Itore Mixture Level 5-69

FENOC Enclosure I Attachment D EXTENDED PONVER UPRATE 1800 L-C__

E a)

Time (s)

Figure 5.2.2-24A BVPS-I 2.5-infch Break Peak Clad Temperature 5-70

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2000 0,

L-0~

E w.

Time (s)

Figure 5.2.2-24B BVPS2 2.5-inch Break Peak Clad Temperature 5-71

I ~lilt FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000

.2 1500 (n

a) en (n

a,

_ 1000 500 0

Time (s)

Figure 5.2.2-25A BVPS-1 3-inch Break RCS Pressure 5-72

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000 -

-o 1500-0L to C>

Ln L 1000-500 -

0-0 Idoo 2d00 3000 4000 5000 6000 Time (s)

Figure 5.2.2-25B BVPS-2 2.75-inch Break RCS Pressure 5-73

FENOC ZZ-. Enclosure I Attachment D EXTENDED POWER UPRATE

TOP or CORE = 21.783 ft 35 -

30 I I--__

-J

-- 25 x 20-5-

10-Time (s)

Figure 5.2.2-26A BVPS-1 3-inch Break Core Mixture Level 5-74

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.7862 ft 35 -

30-15.

15 _ Ia Time (s)

- Figure 5.2.2-26B BVPS-2 2.75-inch Break Core Mixture Level 5-75

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1800 1600 1400 U-1200 w

0 a>

Ea 1000 800 600 400 Time (s)

Figure 5.2.2-27A BVPS-I 3-inch Break Peak Clad Temperature 5-76

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 0~

E wi Figure 5.2.2-27B BVPS-2 2.75-inch Break Peak Clad Temperature 5-77

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE o 1500 v)

W C,,

C 1000 Figure 5.2.2-28A BVPS-1 3.25-inch Break RCS Pressure 5-78

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 0~

'V 0-a) 0>

Time (s)

Figure '5.2.2-28B BVPSM2 3.25-inch Break RCS Pressure 5-79

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.783 ft 35 x 25 Time (s)

Figure 5.2.2-29A BVPS-I 3.25-inch Break Core Mixture Level 5-80

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

- - -- TOP Or CORE = 21.7862 ft 35 -

30-

~- 25- _ _ _ _ _ _ _

x 20-] -

15 -

10 - - - - - - - - - - - - -

.. ... ^ , - a -

Figure 5.2.2-29B BVPS-2 3.25-inch Break Core Mixture Level 5-81

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1800 1600 1400

_s 1200 a) 0 1000 800 600 400 Time (s)

Figure 5.2.2-30A BVPS-1: 3.25-inch Break Peak Clad Temperature 5-82

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 0-a-,

E a,

Time (s)

Figue 5.2.2-3DB BVPS-2 3.25-Inch Break Peak Clad Temperature 5-83

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000

- 1500 0-L.

CD U) an w 1000 cnr 500 0

Figure 5.2.2-31A BVPS-1 4-inch Break RCS Pressure 5-84

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000

.2 1500 U)

U)

" 1000 500 0

Time (s)

Figure S.2.2-31B BVPS-2 4-inch Break

-RCS Pressure 5-85

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.783 ft 35

= 25 a.)

-J a.)

x 20 Time (s)

Figure 5.2.2-32A BVPS-I 4-inch Break Core Mixture Level 5-86

_.E.O_ Enclosure I Attachment D EXTENDED POWER UPRATE Enclosure I Attachment D EXTENDED POWER UPRATE

- - -- TOP OF CORE = 21.7862 ft 35 25

-5

-J a,

x 20 Time (s)

Figure 5.2.2-32B BVPS-;2 4-inch Break Core Mixture Level 5-87

- .I,,,.

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1400 1200 1000 Q) 0~

E E, 800 600 400 Time (s)

Figure 5.2.2-33A BVPS-I 4-inch Break Peak Clad Temperature 5-88

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 1o a.

0~

E a,

Figure 5.2.2-33B BVPS-2 4-inch Break Peak Clad Temperature 5-89

,i'll FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000 o

.2 1500 U)

-C,)

Q>

CL 1000 500 0

Time (s)

Figure 5.2.2-34A BVPS-I 6inch Break RCS Pressure 5-90

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 2500 2000 o 1500 L.2 U)

U)

" 1000 500 500 0

Time (s)

Figure 5.2.2-34B BVPS-2 6inch Break RCS Pressure 5-91

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

-- -- TOP OF CORE = 21.783 ft

~- 25 z

-J a.,

_ 20 Figure 5.2.2-35A BVPS-1 6-inch Break Core Mfixture Level 5-92

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE

TOP OF CORE = 21.7862 ft 35-30-

-^25-a, 120-150-10II I I I a Time (s)

Fuigre 5.2.2-35B BVPS-i -CinchBreak Core Mixture Level 5-93

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE L-0 0~

Ew Figure 5.2.2-36A BVPS-1 6Linch Break Peak Clad Temperature 5-94

FENOC Enclosure I Attachment D EXTENDED POWER UPRATE 900 800 700 Lo D 600 0-E CU-500 400 300 Time (s)

Figure 5.2.2-36B BVPS-2 6inch Break Peak Clad Temperature 5-95

v t e Letter Sequence Supplement
TAC:MC4645, (Approved, Closed) TAC:MC4646, (Approved, Closed) TAC:MC6725, (Approved, Closed)
Initiation Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting Questions 11 March 2005 5 May 2005 2 August 2005 30 September 2005 1 November 2005 Answers 6 September 2005 Results Approval, Approval, Approval Other: ML060330262, ML061770605
MONTHYEAR2005-06-28 [Table View]

L-05-168, Supplement to License Amendment Request

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L-05-168, Supplement to License Amendment Request

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