Text

Provides Response to NRC 990805 RAI Re Amend Request to Increase Fuel Storage Capacity
	ML18017A858
Person / Time
Site:	Harris
Issue date:	09/03/1999
From:	Alexander D; CAROLINA POWER & LIGHT CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18017A859	List: HNP-99-129, Provides Response to NRC 990805 RAI Re Amend Request to Increase Fuel Storage Capacity; ML18017A860; ML18017A861;
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	HNP-99-129, NUDOCS 9909100158
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~ CATEGORY 1 REGULATO Y INFORMATION DISTRIBUTIOh SYSTEM (RIDS)

ACCESSION NBR:9909100158 DOC.DATE: 99/09/03 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH.NAME AUTHOR AFFILIATION ALEXANDER,D.B.

Carolina Power Sc Light Co.

RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)~

SUBJECT:

Forwards addi info as suppl to 981223 request for amend to license NPF-63 to increase fuel storage capacity, as C

requested by 990805 NRC RAI.,

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NOTES:Application for permit renewal filed.

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Carolina Power &Light Company Harris Nuclear Plant P.O. Box 165 New HillNC 27562 SEP 8 1999 SERIAL: HNP-99-129 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEARPOWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION REGARDING AMENDMENTREQUEST TO INCREASE FUEL STORAGE CAPACITY

Dear Sir or Madam:

By letter HNP-98-188, dated December 23, 1998, Carolina Power &Light Company (COL) submitted a license amendment request to increase fuel storage capacity at the Harris Nuclear Plant (HNP) by placing spent fuel pools C 0 D in service. The U. S. Nuclear Regulatory Commission (NRC) issued letters dated March 24, 1999, April29, 1999, and June 16, 1999 requesting additional information regarding our license amendment application. HNP letters HNP-99-069, dated April30, 1999, HNP-99-094, dated June 14, 1999, and HNP-99-112, dated July 23, 1999 provided our respective responses.

By letter dated August 5, 1999, the NRC issued a fourth request for additional information (RAI) regarding our license amendment application to place spent fuel pools C 8c D in service. The Enclosures to this letter provides the HNP response to the NRC staff's August 5, 1999 RAI.

The enclosed information is provided as supplement to our December 23, 1998 amendment request and does not change our initial determination that the proposed license amendment represents a no significant hazards consideration.

Please refer any questions regarding the enclosed information to Mr. Steven Edwards at (919) 362-2498.

Sincerely,~ wW~ j~ D.grg4~r Donna B. Alexander Manager, Regulatory Affairs Harris Nuclear Plant (g

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Document Control Desk SERIAL: HNP-99-129 Page 2 KWS/kws

Enclosures:

1.

HNP Responses to NRC Request For Additional Information (RAI) 2.

Calculation SF-0040, entitled "Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis" (w/o Attachments) 3.

Calculation SF-0041, entitled "Harris Fuel Pool Heatup Calculation" 4.

Attachment Z to Calculation SF-0040 - Evaluation of CCW System LOCA-Containment Sump Recirculation (RHR and SFP) Alignment Thermal Performance Mr. J. B. Brady, NRC Senior Resident Inspector (w/ Enclosure 1)

Mr. Mel Fry, N.C. DEHNR (w/ Enclosure 1)

Mr. R. J. Laufer, NRC Project Manager (w/ all Enclosures)

Mr. L. A. Reyes, NRC Regional Administrator - Region II(w/ Enclosure 1)

Document Control Desk SERIAL: HNP-99-129 Page 3 bc: (all w/ Enclosure 1)

Mr. K. B. Altman Mr. G. E. Attarian Mr. R. H. Bazemore Mr. C. L. Burton Mr. S. R. Carr Mr. J. R. Caves Mr. H. K. Chernoff (RNP)

Mr. B. H. Clark Mr. W. F. Conway Mr. G. W. Davis Mr. W. J. Dorman (BNP)

Mr. R. S. Edwards Mr. R. J. Field Mr. K. N. Harris Ms. L. N. Hartz Mr. W J. Hindman Mr. C. S. Hinnant Mr. W. D. Johnson Mr. G. J. Kline Mr. B. A. Kruse Ms. T. A. Head (PEEcRAS File)

Mr. R. D. Martin Mr. T. C. Morton Mr. J. H. O'eill, Jr.

Mr. J. S. Scarola Mr. J. M. Taylor Nuclear Records Harris Licensing File Files: H-X-0511 H-X-0642

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 1 of 20

..9909100158 SHEARON HARRIS NUCLEARPOWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION REGARDING THE LICENSE AMENDMENTREQUEST TO INCREASE FUEL STORAGE CAPACITY Re nested Information Item 1:

In September 1983, the staff issued NUREG-1038, "Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Units 1 and 2," which included a review of the spent fuel storage facility. The review of the spent fuel storage facility, including the two spent fuel pool cooling systems (SFPCSs) and the four fuel storage pools, was performed in accordance with the applicable sections ofNUREG-0800, "Standard Review Plan." The U. S.

Nuclear Regulatory Commission (NRC) staff's review found the design of the Unit 1 and Unit 2 fuel storage facilities acceptable.

At the time NUREG-1038 was issued, construction of the Unit 2 SFPCS was still ongoing and expected to be completed.

In November 1983, plans to complete Unit 2 were canceled and construction of the partially completed Unit 2 SFPCS was placed on hold.

On December 23, 1998, Carolina Power 0 Light Company (CP&L) requested a license amendment to activate spent fuel pools (SFPs) C and D. The submittal provided information to the staff regarding the activation of the pools; however, no information was provided on the design of the SFPCS supporting SFPs C and D. Given that the SFP C and D SFPCS was never placed in operation, and that significant changes to the design were proposed in the December 23, 1998 submittal, please provide information to show how the portions of the Unit 2 spent fuel storage facility (e.g., SFPs C and D, and the fuel pool cooling and cleanup system - as built), that are not already addressed in the December 23, 1998 submittal, meet the guidance in Regulatory Guide 1.13, "Spent Fuel Storage Facility Design Basis," and NUREG-0800. You may reference the NRC's acceptance of those portions of the fuel storage facilitythat have not changed from the design the staff previously accepted.

Res onse1:

This requested information item is addressed below by a matrix that shows how the portions of the spent fuel storage facilityoriginally intended to support Unit 2 (i.e., SFPs C and D, and the Fuel Pool Cooling and Cleanup System - as built) meet the guidance in Regulatory Guide 1.13, "Spent Fuel Storage Facility Design Basis," and NUREG-0800, Standard Review Plan. The matrix provides a cross-reference listing of the relevant NUREG-0800/NUREG-1038 sections associated with spent fuel storage, fuel pool cooling, and fuel pool area ventilation; identifies the proposed changes to the portions of SFPs C and D and associated Fuel Pool Cooling and Cleanup System previously accepted by the NRC staff (

Reference:

NUREG-1038, "Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Units 1 and 2, dated November 1983); and provides the reason / basis for the proposed changes.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 2 of 20 Reconciliation of SFP Activation Project with SER (NUREG-1038)

SRP/SER Section 9.1.2 Section paragraph number Change from Design As Documented in SER No change from design previously accepted by NRC Staff as documented in the SER.

N/A Reason / Basis for Change 9.1.2 9.1.2 Change with regard to the following:

Infers that two units willbe completed Discussion of storage capacity has been revised per the license amendment request Discussion of rack arrangements has been revised Note that discussion pertaining to maintaining K,tr at or below 0.95 has not been affected.

No change from design previously accepted by NRC Staff as documented in the SER.

Unit 2 was not completed. A single new fuel pool is provided for Unit 1, with the remaining 3 pools representing existing or proposed spent fuel storage capacity. A current description of the completed facility at this point in time is provided in the FSAR, Section 9.1.2.2. Information relative to the proposed number and type of storage locations associated with the proposed configuration is provided in Enclosure 1 of the license amendment request (HNP-98-188, dated 12/23/98).

N/A 9.1.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.2 Assumes that two units are completed and addresses shared portions of facilities, stating that a loss of offsite power willnot impair the ability to safety store spent fuel.

Since Unit 2 was not completed, there are no shared facilities between units. Relative to the proposed change, redundancy is provided so that an accident or loss of power in the operating unit willnot impair the ability to safely store spent fuel in an of the fuel stora e

ools.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 3 of 20 9.1.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.2 10 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.2 12 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.3 No change from design previously accepted by NRC Staff as documented in the SER.

N/A

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 4 of 20 9.1.3 9.1.3 9.1.3 Assumes that two units are completed.

Refers to one fuel pool cooling system being provided "foreach unit".

States that each fuel pool cooling pump is capable of being loaded to onto a separate emergency power supply in case of loss of offsite power, and that each cooling train is a 100% subsystem servicing the new and spent fuel storage pool in that unit.

No change from design previously accepted by NRC Staff as documented in the SER. The equipment in this discussion is the sum of that which would be provided for the entire facility, not just one unit.

Assumes two units were completed.

Describes the facilityas having new storage pools at either end, and the spent fuel pools being connected to the fuel transfer canal "in its unit." Also states "Makeup to the pools may be provided from a seismic Category 1 source (the refueling water storage tank) by means of the fuel pool cooling pumps."

The proposed configuration completes the FPCCS as described in the SER for the two unit site. Two separate fuel pool cooling systems are provided, one for the two pools currently in service, and one for the two additional pools which were originally intended forUnit 2 (i.e., the C and D pools). Consistent with the description in the SER, each FPCCS willcontain two cooling trains; each train including a heat exchanger, strainer, and fuel pool cooling pump, with each pump capable of being manually loaded onto a separate emergency power supply in the event of loss of offsite power. Each cooling train is a 100% subsystem, servicing both oolsinthats stem.

N/A See FSAR Section 9.1.2.2 for a description of the facility. The RWST for Unit 1 is available as a source of makeup water, willbe connected to the fuel pool cooling pumps for both FPCCS, and has been evaluated and found sufficient to perform this function for all four pools in the proposed configuration.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 5 of 20 9.1.3 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.3 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.3 Identifies the commitment by COL to provide two cooling pumps and two heat exchangers for Unit 1 and to provide a similar arrangement for Unit 2.

The proposed configuration is consistent with the commitment made for the Unit 2 design. The new FPCCS willbe completed to essentially the same design as originally proposed to service Unit 2, including two fuel pool cooling pumps and two heat exchangers.

The detailed description of the Unit 2 FPCCS is essentially the same as that provided in FSAR Section 9.1.3.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 6 of 20 9.1.3 10 Discusses transfer of fuel between units with regard to meeting requirements of GDC 5.

Since Unit 2 was not completed, there can be no transfer of fuel between units. Relative to fuel pool cooling capacity, the license amendment request proposes a new Technical Specification to limitfuel pool (C and D) heat loads to no more than 1.0 MBtu/hr. This relatively low heat load limitis determined sufficient to support spent fuel storage needs until analyses associated with Steam Generator Replacement and Power Uprate progress to the point at which the integrated effect on CCW can be quantified. Once this integrated assessment is made, a subsequent license amendment request willbe required to increase the heat load limitto reflect full spent fuel storage capacity in pools C and D.

9.1.3 Assesses temperature ofUnit 2 spent fuel pool with the assumption that this unit was completed and that this pool has the greatest heat load.

This assessment is not valid, because Unit 2 was not completed and pools C and D willonly be used for "colder" spent fuel meeting specific burnup limitations.

Under the license amendment request to place pools C and D in operation, spent fuel storage in the Unit 2 spent fuel pools (i.e., pools C and D) willbe limited to 1.0 MBtu/hr. As a result, the heat load in these pools is bounded by that which might exist in the Unit 1 spent fuel pools. The current temperature limitassociated with the operating Unit 1 (i.e., A and B) pools is 137 'F. That limitis not affected by this license amendment request.

Relative to the Unit 2 (C and D) pools, peak temperatures are anticipated to be well below this value at the maximum allowable pool heat load of 1.0 MBtu/hr.

9.1.3 12 No change from design previously accepted by NRC Staff as documented in the SER.

N/A

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 7 of 20 9.1.3 13 Describes makeup being provided from two Refueling Water Storage Tanks, one from each unit. Staff acknowledges that only one RWST willbe available while Unit 2 is being built, and SER states that the single Refueling Water Storage Tank (RWST) is sufficient. SER states that ESW is available through valved and flanged emergency connections as a backup seismic Category 1 water source.

Since Unit 2 willnot be completed, no separate RWST exists with regard to the Unit 2FPCCS.

The Unit 1 RWST willbe connected to both FPCCS. This single RWST has been evaluated and determined adequate for providing a seismic Category 1

makeup source for all four pools.

Other sources of makeup water are also available, including the seismic Category 1 ESW System, the Demineralized Water System, and the Reactor Makeup Water Storage Tank.

The emergency connections described in the SER willnot be provided on the Unit 2 FPCCS, since there is no ESW supply in the proximity to which they can be connected.

Rather, ESW is available at a location in the Unit 1 RAB, where a cross-tie to both FPCCS can readily be made.

9.1.3 14 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.3 15 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.1.3 16 States that spent fuel pool water willbe sampled weekly.

Chemical impurity limits are to be maintained in accordance with Westinghouse WCAP-7452, Revision 2, 1977.

Fuel pool water chemistry limits are consistent with guidelines and specifications established by the NSSS vendor, fuel manufacturer, and EPRI standards.

Fuel pool water is monitored routinely by chemical and radiochemical analysis of grab samples.

9.1.3 17 No change from design previously accepted by NRC Staff as documented in the SER.

N/A

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 8 of 20 9.1.3 9.4.2 18 Reiterates the commitment to provides two fuel pool cooling pumps and heat exchangers for the Unit 2 FPCCS.

No change from design previously accepted by NRC Staff as documented in the SER.

The proposed configuration of the Unit 2 FPCCS is essentially the same as that for Unit 1, and includes two

separate, 100% subsystems.

Each train includes a fuel pool cooling pump, strainer and heat exchanger.

The fuel pool cooling pump for each train is powered by a separate emergency power supply to provide spent fuel pool cooling capability even in the event of a loss of offsite power. The completed design is essentially the same as that shown in FSAR Figures 9.1.3.1, 9.1.3.2, 9.1.3.3 A 9.1.3.4 as being on "Construction Hold."

N/A 9.4.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 9 of 20 9.4.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.4.2 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.4.2 9.4.2 10 No change from design previously accepted by NRC Staff as documented in the SER, except to note that this section contains a statement that precautions willbe taken during construction ofUnit 2 to protect operating features of the spent fuel pool area ventilatibn s stem.

No change from design previously accepted by NRC Staff as documented in the SER.

Control of all site work activities (including those associated with completion of the Unit 2 spent fuel storage facilities) is controlled under site procedures for work control and screened for potential impact on the operating and licensed portions of the plant.

N/A 9.4.2 12 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.4.2 13 No change from design previously accepted by NRC Staff as documented in the SER.

N/A

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 10 of 20 9.4.2 14 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.4.2 15'o change from design previously accepted by NRC Staff as documented in the SER, except to note that a typo exists. Should reference Position C.4 ofRG 1.13, not RG 1.14.

N/A 9.4.2 16 No change from design previously accepted by NRC Staff as documented in the SER.

N/A 9.4.2 17 No change from design previously accepted by NRC Staff as documented in the SER.

N/A Re uested Information Item 2:

As shown on Table 9.2.2-1 of the SHNPP Final Safety Analysis report (FSAR), each of the two Component Cooling Water (CCW) heat exchangers (HXs) has a design heat transfer rate of 50 MBtu/hr. Table 9.2.1-3 of the SHNPP FSAR shows the maximum service water heat load from the CCW HXfollowing a loss-of-cooling accident (LOCA) to be 273 MBtu/hr. The unreviewed safety question (USQ) analysis in Enclosure 9 of the license amendment appears to indicate that one RHR HX (a single failure assumption) and one CCW train could remove 111.1 MBtu/hr.

Discuss the differences among FSAR Tables 9.2.1-3, 9.2.2-1, and the USQ analysis.

Res onse2:

HNP FSAR Section 9.2.1 addresses Emergency Service Water (ESW) capabilities. HNP FSAR Table 9.2.1-3 shows the maximum ESW system heat loads estimated to exist during post-LOCA conditions. In order to postulate the maximum potential heat load on the ESW system, both RHR loops are considered to be in service under worst case conditions. Since a single train of RHR is analyzed to remove up to 111.1 MBtu/hr in the post-LOCA scenario, the RHR contribution to the CCW portion of the total ESW heat load is 222.2 MBtu/hr. An additional 50.4 MBtu/hr of station auxiliary loads cooled by CCW is added to this value, resulting in the 272.6 MBtu/hr value shown in FSAR Table 9.2.1-3 for the total CCW contribution to ESW heat

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 11 of 20 loads. These cumulative heat loads represent the maximum estimated ESW heat loads existing under post-LOCA conditions, whereas the 111.1 MBtu/hr is the heat removal requirement for a single train of RHR based on containment analyses.

HNP FSAR Section 9.2.2 describes the design and operation of the CCW system.

The 50.5 MBtu/hrvalue shown in FSAR Table 9.2.2-1 is the "design heat transfer rate" for CCW. This value is consistent with design requirements for normal plant operation, and is based on a CCW outlet temperature of 105' (vs. 120' for post-LOCA conditions). Since heat transfer varies with flow rates and inlet conditions, the heat exchanger is capable of a wide range of heat removal rates.

Analyses show that both 50.5 MBtu/hr (normal operation) and 111.1 MBtu/hr (post-LOCA) are within the capability of the CCW heat exchanger under the conditions associated with the scenario being evaluated.

Relative to the USQ analyses, the thermal-hydraulic calculations which support using the Unit 1 CCW system to provide cooling to the C and D spent fuel pools did not change any assumptions regarding maximum sump temperatures or RHR heat removal requirements under post-LOCA containment conditions. The analyses did, however, identify that fluidproperties at the higher RHR temperatures associated with the post-LOCA scenario would result in an increase in heat exchanger heat transfer coefficient (HTC) values over the fixed value currently assumed.

For the purpose of this analysis, the RHR HTC value was therefore allowed to vary as a function of fluid properties in order to ensure that the adequacy of downstream heat sinks was demonstrated under most limitingconditions. The CCW flow rate which corresponds to the requisite 111.1 MBtu/hr heat removal rate was calculated in this manner and used to prescribe the CCW system flow rebalance associated with the additional spent fuel pool heat load.

Re uested Information Item 3:

In Enclosure 6, Section 5.0, "Thermal-Hydraulic Considerations," of the license amendment request, Holtec provides a summary of the methods, models, analyses, and numerical results to demonstrate the compliance of the SHNPP SFPs C and D with the provisions of Section IIIof the NRC OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications (Apri11978). Section 5.3 discusses the bulk pool temperature analysis. Holtec's conclusion is that the cooling water system must meet the design flow versus inlet water temperature specifications shown on Figure 5.3.1 ofLAREnclosure 6. Given that the SFPCS is already designed and constructed, and that the licensee has proposed to limitthe heat load in SFPs C and D to 1.0 MBtu/hr (proposed Technical Specification (TS) 5.6.3.d), provide a thermal-hydraulic analysis using the system parameters for the SFPCS that support SFPs C and D that show the maximum bulk pool temperature for SFPs C and D willnot exceed 137 'F assuming a single active failure.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 12 of 20 Res onse3:

The Holtec scope of supply included a single analysis to support operation with up to fullpool conditions in both the C and D spent fuel pools. The Holtec analysis considers 15.63 MBtu/hr and establishes the system conditions (spent fuel pool cooling flow and temperature) required to maintain the current spent fuel pool limitat that heat load. Section 5.0 of the Holtec report is pertinent to the consideration of forced cooling requirements at 1.0 MBtu/hrin that it includes a review of pool design and layout, and shows that short-circuiting of cooling flowwillnot occur.

Actual requirements for the 1.0 MBtu/hr heat load are established in HNP calculation SF-0040, which considers not only spent fuel pool cooling requirements, but also the cooling requirements of downstream heat sinks (i.e., CCWS, ESWS, and the ultimate heat sink). The body of calculation SF-0040 is provided in Enclosure 2. Note that the same spent fuel pool cooling pumps and piping willbe used for the initial phase at 1.0 MBtu/hr as for the fullpools, such that the same spent fuel pool cooling flow rates willexist for both scenarios.

Calculation SF-0040 demonstrates the adequacy of plant heat sinks to maintain SFPs C and D at or below 137 'F given a single active failure for all plant conditions which require that assumption.

Re nested Information Item 4:

Table 5.2.3 ofEnclosure 6 states that bounding decay heat input from stored fuel in spent fuel pools C and D assumed in the thermal-hydraulic analysis totals 15.63 MBtu/hr. The proposed TS 5.6.3.d limits the heat load in spent fuel pools C and D to 1.0 MBtu/hr. Explain the difference between the maximum heat load requested in the license amendment and the heat load calculated and used in the Enclosure 6 decay heat analysis.

Res onse4:

CP&Lis proceeding with a phased approach to licensing HNP spent fuel pools C and D. The first phase willcomplete the spent fuel pool cooling systems and other supporting systems and provide for a maximum heat load of 1.0 MBtu/hr. This phase is now being reviewed by the NRC. The second phase willassess conditions necessary to utilize C and D pools at fullcapacity and consider the impacts of power uprate and steam generator replacement projects (scheduled for implementation in the fall of year 2001) on plant heat sinks. This phased approach is necessary, because spent fuel generation and storage requirements dictate that construction begin to complete spent fuel storage facilities expeditiously; however, the analyses supporting the power uprate and steam generator replacement projects have not progressed to the point that a comprehensive evaluation can be conducted.

Therefore, based on a review of spent fuel generation and storage plans, an interim heat load of 1.0 MBtu/hr was chosen as a limitwhich supports fuel handling operations at HNP through the year 2001.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 13 of 20 Re uested Information Item 5:

Section 5.4.1 ofEnclosure 6 discusses time-to-boil assuming a complete loss ofcooling to spent fuel pools C and D. The analysis assumes a decay heat load of 15.63 MBtu/hr, which results in a heat up rate of 5.4 'F/hr. Given that the storage pools are limited to 1.0 MBtu/hr by the proposed TS, provide a pool heat up analysis using a decay heat rate of 1.0 MBtu/hr. In addition, discuss the available makeup sources to spent fuel pools C and D and their capacities relative to the calculated boil offrate.

Res onse5:

The time to boil and pool heatup analyses for the 1.0 MBtu/hr scenario are well bounded by the time to boil and pool heatup analyses for the 15.63 MBtu/hr scenario presented in Section 5.4.1 ofEnclosure 6 to the license amendment request.

Analyses specific to 1.0 MBtu/hr have been performed and are documented in HNP Fuel Pool Heatup Calculation SF-0041, provided herein as Enclosure 3. These analyses calculate an estimated pool heatup rate of approximately 0.33 'F

/hr, and conclude that the pools would not heat up to 140 'F until approximately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months into the event, and an additional 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months would be required to reach boiling conditions.

Re uested Information Item 6:

The USQ analysis results of Enclosure 9 indicate that a minimum CCW system (CCWS) flow rate of 4874 gpm at 120'F is required at the beginning of the containment sump recirculation phase of a LOCA and that, assuming a 6% model uncertainty and degraded inservice test (IST) pump performance, the specified CCW flow to the residual heat removal (RHR) HXwould be 5166 gpm, which is less than 5600 gpm in the existing analysis.

This result is based on (1) the RHR HXheat rejection rate of 111.1 MBtu/hr, which is said to be consistent with the existing post-LOCA containment pressure/temperature calculations, and (2) the use of a "dynamic" RHR HXperformance model, in which the tube side inlet temperature is postulated to rise to 244.1'F during the initial phase of sump recirculation, rather than a fixed 139'F assumed in the existing analysis, resulting in an increase of the overall RHR HX heat transfer coefficient (HTC) of approximately 10% due to change in tube side viscosity.

Provide the dynamic RHR HX heat transfer analysis during and subsequent to the recirculation phase of a LOCA. Important parameters to be provided include the time-dependent decay heat rate, the containment sump water temperature, and the HTCs, heat transfer rates and flow rates (both tube and shell sides) of the RHR HX, and CCW HX, etc. Also describe how the effects of HX degradation mechanism such as fouling and tube plugging of the RHR and CCW systems are accounted for in the RHR and CCW HX heat transfer calculations.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 14 of 20

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The RHR heat exchangers provide long-term cooling during the containment sump recirculation phase of a LOCA. This function is accomplished by aligning the RHR system to take reactor coolant from the containment sump, circulating the reactor coolant through the RHR heat exchangers, and then returning the reactor coolant back to the RCS cold legs. Thermal performance of the RHR heat exchangers at HNP has been analyzed using the dynamic RHR heat exchanger performance model and shown to remain comparable to that calculated by the current containment pressure/temperature analyses.

The dynamic RHR heat exchanger performance model, however, yields a slight reduction in heat transfer relative to long-term post LOCA environmental conditions.

An assessment of the dynamic RHR heat exchanger performance model has been completed for long-term containment heat removal and equipment qualification. From this assessment, it is noted that the heat removal rate calculated by using the dynamic RHR heat exchanger performance model is 111.9 MBtu/hr at peak containment sump temperature (occurring at t =

3600 seconds into the event), which is marginally higher than the 111.1 MBtu/hr value obtained using the fixed HTC model. At 10" seconds into the event, the calculated heat removal rate using the dynamic RHR performance model is 92.2 MBtu/hr, still slightly higher than the 92.1 MBtu/hr associated with the fixed HTC model used in the current containment analysis. By 10 seconds into the event, the calculated containment sump temperature has decreased from 244.1 'F to 167.8 'F, and the heat removal rate calculated by using the fixed HTC model has now become slightly higher than that calculated by the dynamic HTC model (42.8 MBtu/hr compared to 41.9 MBtu/ hr, respectively). At 10 seconds into the event (approximately 12 days), calculated containment sump temperature has decreased to 142.7 'F, and the heat removal rate using the fixed HTC model is 20.3 MBtu/hr, compared to 19.7 MBtu/hrcalculated by the dynamic HTC model.

The table shown on the followingpage provides the requested heat exchanger performance parameters.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 15 of 20 Comparison ofRHR Heat Exchanger Performance for Long-Term Post LOCAEnvironmental Qualification 9 t=3600secs.

9 t=10'secs.

8 t=10 secs.

9 t=10 secs.

Fixed HTC Variable HTC Fixed HTC Variable HTC Fixed HTC Variable HTC Fixed HTC Variable HTC No. U-Tubes 592 592 592 592 592 592 592 592 Surface Area (f

4280 4280 4280 4280 4280 4280 4280 4280 UA (BTU/hr-'.635E6 1.758E6 1.635 E6 1.734E6 1.635 E6 1.6GSE6 1.635 E6 1.627E6 Q

(MBTU/hr 111.9 92.1 92.2 42.8 41.9 20.3 19.7 RHR Flow (10 Ibm/hr RHR Inlet Temp (o

CCW Flow

(

m 1.846 244.1 5600 1.846 244.1 4874 1.846 1.846 1.84G 222.9 222.9 1G7.8 5600 4874 SGOO 1.846 167.8 4874 1.846 142.7 1.846 142.7 4874 CCW Inlet Temp (o

120 120 120 120 120 120 120 120 Because the relationship between heat exchanger flow rates and heat transfer is not linear, the analysis summarized above shows that the reduction in CCW flow from 5600 gpm to 4874 gpm yields no reduction in heat transfer at the earlier stages of the event associated with the highest postulated containment sump temperatures.

Comparably, only a minimal reduction in heat transfer (about 3%) occurs much later into the event when containment sump temperatures have decreased significantly. Both the fixed and dynamic HTC models consider design fouling conditions and 0% tube plugging.

Re uested Information Item 7:

The USQ analysis also indicates the need to increase the current minimum required emergency service water (ESW) flow to the CCW HX from 8250 gpm to 8500 gpm, which is said to have been verified to be within the capacity of the current system. WillFSAR Table 9.2.1-1 be revised to require the ESW flow through the CCW HXto be 8500 gpm?

Res onse7:

Yes, FSAR Table 9.2.1-1 willbe revised accordingly to reflect the revised system flow requirements.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 16 of 20 Re uested Information Item 8:

HNP FSAR Sections 9.2.2.2 and 9.2.2.2.2 provide CCW system and component descriptions, including information about the CCWS surge tank. This surge tank accommodates changes in the fluidvolume of the CCWS from thermal expansion and contraction and accommodates water which may leak into the system from cooled components.

The surge tank also provides the CCWS with a limited water supply until a leaking cooling line can be isolated. Discuss the effects of the additional system volume and heat load (from the piping and components added to support SFPs C and D activation) on the capability of the CCWS surge tank to perform its design function.

Res onse8:

Section 9.2.2.2.2 of the HNP FSAR provides a discussion about four specific functions of the Component Cooling Water (CCW) surge tank. The impact on each of these functions by placing spent fuel pools C and D in service is discussed below:

a) Accommodates changes in CCWS water volume due to changes in operating temperature.

To place spent fuel pools C and D in service, the required modifications to the CCW system will add approximately 2000 gallons of water volume to the CCW system, including the shell side volumes of both spent fuel pool heat exchangers.

Assuming this volume of water undergoes a

temperature increase from 60 'F to 105 'F, the incremental volume increase in CCW inventory would be about 15 gallons. Since normal CCW surge tank level is at approximately 1000 gallons and the tank has a 2000 gallon capacity, this represents only about 1.5% of the tank's available surge volume. Even then, the tank is fitted with overflow and overpressure protection, sized to provide adequate relief from comparatively high volume makeup sources.

Based on these considerations, an increase in CCW volume brought about by an abrupt rise in temperatures would be of no consequence to either plant operation or nuclear safety. The same can be said of the impact of abrupt temperature decreases, where the volume of water maintained in the tank and the makeup capability to the system are adequate to compensate for shrinkage. Finally, the CCW surge tank is equipped with high and low level instrumentation, which alerts the operator in the control room to significant changes in level so that any needed corrections can be readily made.

b) Accommodates, for 20 minutes, the maximum flow from either makeup water supply.

The activation of spent fuel pools C and D willnot change makeup water supply capabilities, normal water level, or the capacity of the CCW surge tank. There is no impact with regard to the capability to accommodate makeup flow.

c) A reservoir of water to provide time to locate and terminate a system leak should one develop.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 17 of 20 The design and construction of the piping and equipment being added to support activation of spent fuel pools C and D is similar to that already installed, such that the size and nature of leaks which might occur and the isolation capability of the equipment is consistent with that which already exists. Given these considerations, the adequacy of the reservoir to provide time to locate and terminate a leak is not adversely impacted.

d) Accommodates, for about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the Technical Specification maximum identified reactor coolant leakage of 10 gpm.

As with item b) above, there is no impact regarding the capability to accommodate leakage from the RCS, because the activation of spent fuel pools C and D willnot affect the capacity of the tank or the normal water level.

Re uested Information Item 9:

Willthe SFPCCS and makeup system(s) for SFPs C and D be included in the inservice inspection program or an inspection program similar to those used with systems that support SFPs A and 8?

R~9:

Portions of the SFPCS willbe included in the site ISI/ IST program, consistent with the treatment of equipment and support systems associated with spent fuel pools A and B.

Specifically, the piping within Code boundaries willbe included in the ISI program, and subject to regular inspections per the requirements of that program. In addition, the followingspent fuel pool cooling system components willbe added to the site IST program:

~

spent fuel pool cooling pumps

~

spent fuel pool cooling system relief valves

~

spent fuel pool cooling pump discharge check valves Re uested Information Item 10:, Part 1 of the significant hazards consideration determination discusses the probability or consequences of an accident previously evaluated.

In the fourth paragraph, you allude to the movement of fuel assemblies "... required to be performed to support this activity (e.g., installation ofracks)..." Since SFPs C and D are currently empty, and no reracking of SFPs A and B are included in this licensing amendment, what fuel movements do you anticipate willbe required during the course of the modifications authorized by this license amendment?

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 18 of 20

~Rill:

There is currently no fuel in the C (or D) spent fuel pool, nor willany be installed until such time as the initial installation of racks are completed and approval from the NRC to place the pool in service is obtained. No fuel movement willbe involved with the installation of racks performed in support of this license amendment request.

Re uested Information Item 11:

In Section 4.6, "CCWS Hydraulic Margins," the first paragraph refers to a modification to the CCWS piping to SFP HXs C and D to be able to throttle flow to 2.03% for the Hot Shutdown (350'F) alignment. CP&L staff note that the CCWS valves to these HXs must be heavily throttled and willrequire a suitable sized bypass line with a smaller throttle valve in order to achieve acceptable throttling characteristics.

Willthese modifications be performed as part of the system activation? Ifnot, how willoperators throttle flow to the SFP HXs to meet the design conditions specified in SF-0040 Table 6?

Res onse11:

A 6" bypass line and a 6" throttle valve willbe installed as part of the modifications performed as part of system activation. This arrangement has been sized to provide acceptable throttling characteristics at the relatively low flow rates required to accommodate a 1.0 MBtu/hrheat load.

The requisite throttle position willbe set in an initial flow balance.

Thereafter, system alignment willconsist of opening and closing isolation valves. Itis not anticipated that subsequent adjustments to throttle valve position willbe necessary.

Re uested Information Item 12:

Tables 7a through 7j present the results of a CCWS flow analysis to determine the hydraulic margins for various CCWS lineups. In its summary in Section 4.6, CP&L states that the evaluation of the system thermal analysis results during the "LOCA: Recirculation (RHR and SFP) alignment" (Table 7i) shows that the steady state equilibrium temperature of the fuel pool A/B does not exceed 136 'F even with degraded CCWS pump flow and design fouling of all HXs. Please provide a copy of Attachment (Z), "Containment Sump Recirculation (RHR and SFP) Alignment Thermal Performance," or provide the basis, assumptions, and results for this calculation, including the assumed decay heat load, the duration of time the CCWS system is providing insufficient flow to SFP HXA, and the maximum SFP bulk temperature for all SFPs.

Res onse12:

As requested, a copy of Attachment (Z) to calculation SF-0040 is provided herein as Enclosure 4.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 19 of 20 Re uested Information Item 13:

CPAL's description of the refuel-normal and abnormal fullcore offload analysis results (Tables 7e and 7f) indicate that the SFP HXA (or B) can just accommodate an assumed fullcore offload heat load of 31.7 MBtu/hr at design SFPCS thermal conditions; therefore, the negative CCW flow margin is acceptable under these extreme thermal-hydraulic conditions; however, no basis is provided for this conclusion.

Please provide your justification for concluding that operating the SFPCS with CCWS flow7% less than the minimum flow stated in Tables 7e and 7f assures the design limits for the SFPs are not exceeded.

In addition, in table 7f, the minimum CCWS flow to RHR pump A with a 6% uncertainty is calculated to be 8% less than the minimum required flowrate, yet no justification is given why this deficient condition is acceptable.

Please provide the basis why this condition is acceptable.

Res onse13:

Calculation SF-0040 Attachments (M) and (N) document the SFP Hx A (or B) thermal-hydraulic analysis at the design SFPCS flowof 3750 gpm at 137 'F, a design CCWS supply temperature of 105 'F, minimum ESWS flow and maximum ESWS temperature, a design fouling of 0.0005 hr-ft -F/BTU on both the inside and outside tube surfaces and assuming no tubes plugged. This analysis shows that the SFP heat exchanger(s) can accommodate a heat duty of 31.69 MBtu/hr.

The estimated CCWS supply temperature for this system alignment is 104.7 'F with the CCW heat exchanger operating design fouling factors and ESW flowof 8500 gpm at 95 F.

The assumption of no tubes plugged in the SFP heat exchangers is valid since these heat exchangers are being placed into service for the first time, and this analysis willbe utilized for only a single operating cycle, after which system thermaVhydraulic performance willbe re-evaluated in support of power uprate and steam generator replacement projects.

In addition, current operating practice at HNP is to evaluate spent fuel pool heat loads each cycle, and specify a minimum time prior to offloading fuel to ensure adequacy of the CCWS. In this instance, core offload would not be performed ifSFP heat load exceeded 31.69 Btu/hr.

Relative to the adequacy of CCWS flow to the RHR pumps, Table 7f reflects that the minimum CCWS flow to RHR pump A with a 6% uncertainty is calculated to be 8% less than the minimum design flow rate. This table is applicable to Mode 6, wherein the RHR pumps are used intermittently for volume control rather than heat removal. In this scenario, the spent fuel pool cooling system rejects the heat associated with the offloaded core. For the purposes of SF-0040, it was assumed that both trains of RHR were in operation, even though no design requirement exists for doing so and as it would not be likely as a matter of practicality. This is a conservative approach in that it ensures that flow which might be diverted through the RHR seal coolers is not considered available to other heat loads.

Document Control Desk Enclosure 1 to SERIAL: HNP-99-129 Page 20 of 20 Durmg Mode 6, the CCWS trains are required to be separated to prevent CCW pump run out. In this case, the "B"CCW train, which supplies only the safety related header, has significant flow margin. The "A"CCW train supplies the other safety related header along with the nonessential header, and has slightly less than the 5 gpm design flow to the RHR pump seal cooler (-8%)

under these conditions. Assuming the design RHR pump seal cooler heat load (0.07 MBtu/hr),

this deficiency in flow would be expected to result in a slight temperature increase in the seal water returning from the cooler. However, this maximum heat load is associated with maximum RHR operating temperature of 350 'F, considerably higher than the RCS temperatures that would exist during defueled conditions (well below 200 'F). As RHR temperatures decline, the heat removal requirements on the seal cooler would also diminish. In fact, HNP Operating Procedure OP-111 does not include a requirement for any seal cooling at RHR temperatures below 225 'F.

It is concluded that the 8% flow deficit listed for the RHR pump seal cooler in Table 7f is of no consequence to the performance or reliabilityof the RHR pumps.

Re uested Information Item 14:

not exceed 10

~R14:

On page 28 of SF-0040, you state that the SFPs are conservatively assumed to be at the maximum temperature limitof 105 'F prior to the beginning of the transient. What administrative controls are used at SHNPP to assure that the SFP bulk coolant temperature will 5 'F during normal operation?

HNP Operating Procedure OP-116, Fuel Pool Cooling and Cleanup, includes a fuel pool heat exchanger outlet temperature limitation of 105 'F under normal operating conditions. A control room alarm is provided to alert the operator ifthis value is exceeded, and spent fuel pool temperatures are recorded every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in the reactor operator log books. Operating experience has found that the system is capable of maintaining temperatures well below this value, even under fullcore offload conditions.

HNP-99-129, Provides Response to NRC 990805 RAI Re Amend Request to Increase Fuel Storage Capacity

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HNP-99-129, Provides Response to NRC 990805 RAI Re Amend Request to Increase Fuel Storage Capacity

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Reference:

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