ML20116D892

Evaluation of Dhrs

ML20116D892
Person / Time
Site:	FitzPatrick
Issue date:	07/16/1996
From:	Herrmann T, Ruddy D, Vehstedt K POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20116D890	List: JAFP-96-0291, Informs of Util Intent to Install Alternate Dhrs at Plant in Advance of Refueling Outage 12,scheduled to Commence in Oct 1996.Info Encl Provides Details of Sys Design & Safety Considerations Pertinent to Sys Installation & Testing ML20116D892 ML20116D898 ML20116D902
References
JAF-RPT-DHR-024, JAF-RPT-DHR-024-R13, JAF-RPT-DHR-24, JAF-RPT-DHR-24-R13, NUDOCS 9608050004
Download: ML20116D892 (33)

Text

s James A. FitzPatrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No. JAF RPT-DHR-02413 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT EVALUATION OF THE DECAY HEAT REMOVAL SYSTEM Report no JAF-RFT-DHR-02413 REVISION DATE PREPARED BY; REVIEWED BY:

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K. J Vehstedt T. J. Herrmann 7/

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's APPROVED:

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'D. A. FVddy /

Director, Design Engineering

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James A. Fit 7 Patrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No. J AF.RPT-DHR-02413 1

EXECUTIVE

SUMMARY

This study report is prepared in accordance with Design Control Manual 7, " Preparation of Technical l

Stud es and Reports" to document design bases supporting analyses, safety considerations and l

operational limitations associated with insta!!at;on of a Decay Heat Removal (DHR) system at the J A.

FitzPatnck Nuclear Power Plant.

l The resu.tc of this study are.

l i

(1)

Installation and operation of the DHR will constitute a significant enhancement in decay removal capability. particularly dunng refueling outages l

i (2) installation and operation of the DHR will improve the ability to control refueling cavity and spent fuel pool water temperature during refueling operations.

(3)

Installation and operation of the system will eliminate current restrictions on fuel movement which are tied to existing spent fuel pool decay heat removal capacity.

(4)

There are no unreviewed safety questions associated with the DHR. Therefore, system design, installation and operation can be evaluated pursuant to 10 CFR 50 59

)

(5)

Installation and operation of the DHR will provide greater flexibility in outage planning

)

and has the potential to reduce refueling outage length.

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l 1

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Page 2 of 33

James A. Fitz?atrick Nuclear Power P: ant Evaluation of the Decay Heat Removal System Report No. JAF.RPT DHR-02413 TABLE OF CONTENTS Section Subject 1

Introduction 2

Design Bases Considerations 2.1 Heat Removal Capability 22 Interactions with existing Plant Systems 3

Description of the Decay Heat Removal System 3.1 Mechanical Features 3.1.1 Normal Operating configuration 3.1.2 Other Operating Configurations 3.1.3 Primary - to - Secondary Leakage Considerations 3.1.4 Location of Major Components 3.1.5 System "Layup" Considerations 3.2 Instrumentation and Control Features 3.3 Electrical Features 3.4 Civil / Structural Features 4

Engineering and Design issues 4.1 Single Failure Considerations and Overa!! Heat Remova! Capability 4.1.1 Phase 1: From Plant Shutdown to Removal of the SFP Gates 4.1.2 Phase 2: SFP Gates Removed and the Total Decay Heat Load Exceeds DHR Capability in the Nominal Configuration 4.1.3 Phase 3: SFP Gates Removed and the Total Decay Heat Load is Less Than DHR Capability in the Nominal Configuration 4.2 Seismic Design 4.3 System Power Supply 4.4 Electncal Separation 4.5 Operational Radiation Fields 46 Fire Protection issues 4.7 Wind Loadings on Exterior Components 4.8 Internal Flooding 4.9 Heavy Loads 4.10 Decay Heat Load Calculations 5

Nuclear Safety Considerations 5.1 Maintenance Of Spent Fuel Pool Water inventory 5.2 Natura! Circulation Cooling of Fuel Assembhes in the Reactor Pressure Vessel Page 3 of 33

James A. Fit: Patrick Nuc! car Power Plant Evaluation of the Decay Heat Rcrnoval System Report No. J AF RPT DHR-02413 53 Potential for Excessive Cooling of the Spent Fuel Pool 54 Technical Specification and Design Bases Document Review 55 UFSAR Review 6

System Installation and Modification Acceptance Testing 7

Initla! Thermal Performance Testing and Verification of Natural Circulation 7.1 Thermal Performance Testing Dunng RO12 7.2 Verification of Natural Circulation Cooling During RO12 8

System Operation 8.1 DHR Operation with the Plant on line 82 DHR Operation Dunng Refueling Outages - SFP Gates Installed 8.3 DHR Operation Dunng Refueling Outages - SFP Gates Removed 9

Additional Supporting Analyses 9.1 Evaluation of Single Active failures of DHR Components. and Prospective Mitigating Actions with the System Operating in the Nominal Heat Removal Configuration 9.1.1 Postulated f etive Failure of a Mechanical Component 9.1.2 Postulated Active Failure of an Electrical, Instrument, or Control Component 91.3 Evaluation of Prospective Mitigating Actions 9.1.4 Time Frame for the Implementation c' Mitigating Actions 9.2 Estimated time to Boiling Given A Loss of all Decay Heat Removal 9.3 Radiological Consequences of Boiling 9.4 DHR System Reliability Assessment j

9.5 Effects of Elevated Water Temperature on SFP Components 10 Conclusions and Recommendations 1

List of Tables 1

Decay Heat loads 2

Decay Heat Removal Capabilities 3

Dose Consequences Following a Postulated Loss of Decay Heat Removal List of Figures 1

DHR System Flow Diagram - Sheet 1 of 2 2

DHR System Flow Diagram - Sheet 2 of 2 3

DHR System Electrical One-line Diagram 4

Decay Heat Versus Time After Shutdown, Core and Spent Fuel Pool Page 4 of 33

James A. Fit 2Patr,ck Nuclear Power P' ant Evaluation of the Decay Heat Removal System Report No. JAF RPT DHR-02413 List of References 1

Design Control Manual 7. " Preparation of Technical Studies and Reports" Rev:sion 0 l

2 Code of Federal Regulations. Title 10 Part 50. subpad 59. " Changes tests and expenments" 3

IE Bulletin 80-10, " Contamination of Non-radioactive Systems and Resulting Potential for Unmonitored and UncontroUed Release to the Environment" USNRC 4

NYPA Calculation no. JAF-CALC-MISC-02244, " Assessment of the Combined Decay Heat Load of the Reactor Core and Spent Fuel Pool during Refueling Outages",

Revision 0 5

NRC Letter, B. C. McCabe to R. E. Beedle (NYPA), Issuance of Amendment for James A. FitzPatrick Nuclear Power Plant (TAC No. M76937)", dated 31 december 1991 (Note. This is Technical Specification Amendment no 175 and contains the NRC SER for the most recent spent fuel pool storage expansion).

6 Safety Evaluetion for the James A. FitzPatrick Nuclear Power Plant, Docket No. 50-333, U. S. Atomic Energy Commission, Directorate of Licensing, Issued 20 november 1972 7

JAF Nuclear Safety Evaluation no. JAF-SE-90-042, " General Heavy Load Handling System Requirements to Meet NUREG-0612 Cnteria", Revision 0.

8 JAF Maintenance Procedure MP-088 01, " Heavy Load Handling", Revision 12.

9 NUREG-0800, USNRC Standard Review Plac.. Section 9 2.5. " Ultimate Heat Sink".

Revision 2, dated july 1981 (Note Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long-Term Cooling" is appended to this section of the SRP).

10 Significant Operating Experience Report 85-1, " Reactor Cavity Seal Fadure", INPO, January 1985 (and NYPA's evaluation thereof).

11 General Electric Nuclear Energy Letter, F. Paradiso/H. Choe to K. A. Phy (NYPA),

" Decay Heat Removal System Project, Natural Circulation Heat Transport Calculatior.

Results", dated 18 april 1996 12 NYPA Calculation no. JAF-CALC-DHR-02380, " Alternate Decay Heat Removat System Thermal-Hydraulic Analysis", Revision 0 13 General Electric Nuclear Energy Letter W. H. Brown to D. Lindsey (NYPA), "ADHR Questions", dated November 15,1995 14 NYPA Des;gn Basis Document for the Residual Heat Removal System l

15 NUREG 1433, " Standard Technical Specifications for General Electric Plants, BWR/4",

USNRC, April 1995 i

i 16 NYPA Calculation no, JAF-CALC-MISC-02373, " Decay Heat Management Calculation With New SFP Volume". Revision 1 Page 5 of 33 l

James A. F,tzPatrick Nuclear Pcoer F! ant Evaluation of the Decay Heat Removal System Report No. J AF RPT-DHR-02413 17 NYPA Calculation No. JAF-CALC-RAD-00053. " Radiological Impact of Postu!ated Failure of the Alternate Decay Heat Removal System and Loss of Spent Fuel Coohng".

Revision 0 18 Technical Information Document (TID) 14844, " Calculation of Distance Factors for Power and Test Reactor Sites" US Atomic Energy Commission. dated 23 march 1962 19 NYPA Memorandum, JAG-93-285, J. A. Gray, Jr. to H Salmon, " Licensing Basis for Spent Fuel Pool Cooling", dated 19 november 1993 (plus attachment 1, " Loss of Spent Fue! Pool Cooling During design Basis Accident", and attachment 2, " Licensing Basis for Expansion of Spent Fuel Pool Storage Capacity, Analyzed Accidents and Assumptions for Spent Fuel Pool Cooling System")

20 NYPA Report No. JAF ANAL-MISC-02372, " Alternate Decay Heat Removal System",

Revision 0 21 NUREG-0800, USNRC Standard Review Plan, Section 9.1.3. " Spent Fuel Pool Coohng and Cleanup System", Revision 1. dated july 1981 22 JAF Administrative Procedure AP 10 09, " Outage Risk Assessment". Revision 3 Page 6 of 33

James A, Fitdatrick Nuclear Power P! ant

,4 Evaluation of the Decay Heat Removal System Report Nc, [[::JAF-RPT-DHR|JAF-RPT-DHR]] 02413 1,

introduction Mod'fication F1-95-121, currently under development. will install a Decay Heat Removal (DHR) l System in the facihty. The DHR will take suction from and discharge to the scent fuel pool (SFP) As detaled in the ensuing sections of this study the system is designea as a non-safety system, is physically indeper dent of existing plant equipment to the maximum extent and is pnmarily intended to enhance existing decay heat removal capabilities during refueling outages with the ultimate goal of enhandng outage perfmm;,, ice F3ysical installation of the modification is expected to take place prior to reiueling outage 12 (RO12) during which time the plant is expected to be in operation. Modification acceptance testing 9 alf.o expecbd to be completed prior to RO12.

This study is being prepared in accordance with the requirements of DCM 7 (Reference 1) for the purpose of documenting key system design features, consolidating supporting technical information (in advance of issuance of the modification package), and evaluating whether this modification can be performed under the provisions of 10 CFR 50 59 (Reference 2).

l 2.

DesLqn Basis Considerations The three overriding design bases considerations for the DHR are-(1) the decay heat removal capability of the DHR must equal or exceed the combined decay heat load of irradiated fuel in the SFP and the Reactor Pressure Vessel (RPV) approximately 4.5 days post-shutdown while remaining tolerant of a wide spectrum of postulated component failures, (2) to the extent feasible, the DHR shall be mechanically and electrically independent of existing plant systems. and (3) the system must not adversely affect any safe shutdown function of existing plant systems i

2.1 Heat Removal Capability The DHR system is shown schematically in figure nos 1 and 2. (Note' The figures are not design drawings and are included for information only). The nominal configuration of the system will involve operation of one primary side pump, one heat exchanger, one secondary side pump, and one set of cooling towers. The system is designed such that the system heat removal capacity in that configuration will be approximately 30 X 10 BTU /HR given an outside 5

ambient wet bu;b temperature of 73 *F.

The combined RPV and SFP decay heat loads have been conservatively calculated as a function of time post-shutdown and 30 X 10' BTU /HR corresponds to approximately 4.5 days under worst case (maximum heat load) conditions 2.2 Interactions with Existina Plant Systems The DHR is designed to have minimum interface with existing plant systems; in particular systems important to plant safety. With the exception of penetrations through the Reactor Building pressure boundary, connections to the condensate transfer and demineralized water systems (for primary and secondary loop fiD and makeup, respective!y), use of existing cable trays, #commoning" of DHR annunciator inputs into the existing Fuel Pool Cooling & Cleanup (FPCC) control panel, and the obvious interface with the SFP. the DHR is mechanically and electrically separated from existing systems Details of the design follow The DHR system has oeen designated as (new) system number 32 and the component names used it the Page 7 of 33

James A. :: 2 Patrick %c: ear Power Mant Evaluation of the Decay Heat Removat System Report No. JAF-RPT-DHR-02413 l

l following tert ref ect that designation.

I I

3.

Description of the Decay Heat Removal Syste_m l

31 Mechanical Features r

3 1.1 Normal Operatina Configuration.

As shown in figure nos 1 and 2, the DHR system is comprised of a primary loop which pumps the water from the SFP through heat exchanger (s) and returns it to the SFP, and a secondary loop which removes the heat from the primary loop heat exchangers via cooling towers The primary loop includes two 100% pumps. two heat exchangers. and appropnate isolation valves and monitoring instrumentation The normal or nominal operating configuration of the system will be to have one pomary pump in service 32P-1A or 32P-18, one of the two plate and frame heat exchangers 32E-3A or 32E-3B in service, and one of the two seconoary pumps 32P-2A or 32P-2B in service discharging to one set of cooling towers Cooling tower fan operation would be dictated by the heat load on the system A single strainer. 32STR-1 ',ith an automatic backwash feature, is located in the primary loop pump discharge common Icader upstream of the heat exchangers to filter any solid particles in the SFP water. The straiier is provided for ALARA purposes to minimize contamination of the heat exchangers and is 70t required for systern operation.

3 1.2 Other Operating _C_onfigurations it will be possible to operate the system with two secondary pumps in service and both heat exchangers in service. In that maximum heat removal configuration, the system is designed to remove 45 X 10' BTU /HR at a wet bulb temperature of 73 F. Intermediate configurations are 0

also permitted, to allow system heat removal to most close!y match system heat load 3.1.3 Primary _to SecondaE_L_cakage Considerations The design specifically utilizes plcte and frame heat exchangers to eliminate the potential for primary to secondary leakage. In addition, the secondary side of the DHR system will be maintained at a higher operating pressure than the primary side operating pressure (approximately 15 psid) through use of an automatic pressure control va!ve. 32PCV-100, in the common secondary side header on the outlet side of the heat exchangers The design includes controls which will automatically trip an operating primary side pump if the secondary to primary pressure differential where to decrease below 10 psid. These features of the DHR are consistent with the guidance contained in NRC IE Bulletin 80-10. reference 3 Based on those features, there is no need for installation of a radiation monitor in the DHR secondary loop 31.4 Location of Malor Components The primary side components (pumps, strainer, heat exchangers and a control panet) are skid-mounted and will be located on Elevation 326'-9" inside the Reactor Building. The secondary side components (pumps and cooling towers) will be located on the roof of the Railroad and Truck Bay and Standby Gas Treatment Building at El. 293'-0.

Exterior portions of the secondary loop piping, and the demineralized water makeup line to the cooling tower basins, will be insulated and heat-traced to provide freeze protection. In addition, immersion heaters will be provided to prevent freezing of the cooling tower basin inventory.

Page 6 of 33

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Jarnes A. Fit 2 Patrick Nuclear Power Pl ant Evaluation of the Decay Heat Removal System l.

Report No. JAF-RPT-DHR-02413 i

31.5 S._ystem "Layup" Considerations.

It is expected the DHR system will only be ahgned for service immediately onor to dunng ano perhaps immediately after scheduled refueling outages although there is a potential the system could be used during non-outage conditions to supplernent or subst;tute for FPCC.

Tnerefore. the design incorporates the ability to physically remove the suction and discharge pipe spools in the vicinity of the SFP and for their storage at Elevation 369-6"inside the Reactor Building in addition, the design provides for ability to isolate and drain the exterior secondary loop components using manual isolation valves to be installed inboard of the Reactor Building penetrations.

l 32 Instrumentation and Control Features A DHR control panel will be installed in the Reactor Building on elevation 326' 9"in the l

immediate vicinity of the primary side pumps and the heat exchangers. This panel will contain equipment status lights. indicators, and alarms sufficient so an operator can monitor and control system operation Local pressure and temperature indicators will be provided in the l

primary and secondary loops at locations appropriate for monitoring system operation.

l Sensed differential across the strainer will also be used to initiate the automatic backwash feature i

Differential pressure between the primary and secondary sides of the DHR heat exchangers will also be monitored and indicated at the control panet Sensed primary to secondary differential will also be used to position 32PCV-100 and if the sensed differential decreases l

below 10 psid, trip an operating primary pump (s). A common trouble alarm will be provided for the DHR system at the FPC local panel on Elevation 326'-9" which will also alarm the FPC annunciator in the Control Room. Tne common trouble alarm will be actuated on any of the j

following conditions:

1) low flow in the primary system l

2) low flow in the secondary system 3) primary loop pump trip 4) coolng tower fan high vibration j

5) cooling tower basin low water level 1

6) transformer winding temperature - high 3.3 Electrical Features The DHR system will be powered from a reliable, offsite power source (13 2 KV switchgear J02) which will feed into a 1500 KVA transformer and a 480V motor control center (MCC). The 13.2 KV source is independent of the existing safety related and non-safety related power supplies to the power block. Figure 3, attached, provides the DHR system electrical one-line diagram. (Note: Figure 3 is not a design drawing and is provided for information only).

i Dunng refueling outages DHR reliability will be enhanced through the use of a portable diesel generator which will be directly connected to the DHR 480V MCC. The portable diesel generator will be sized to start and carry DHR loads for operation in the nominal configuration l

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Page 9 of 33 l

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James A. FitzPatrick Nuclear Power Plant l

Evaluation of the Decay Heat Rernoval System l

Report No, JAF-RPT-DHR-02413 and would be used in the event of loss of the normal 13 2 KV supply during DHR operation Transfer from the 13 2 KV source to the diesel generator would be accomplished manua!!y 34 CivillStructural Features The secondary side DHR piping will penetrate the south wall of the Reactor Building at approximate elevation 310' 5" and connect to the equipment to be located on the roof of the Railroad and Truck Bay and Standby Gas Treatment Building The design of the penetrations.

including classification of associated piping and supports, will be consistent with existing des:gn standards for Reactor Building penetrations, (Refer to UFSAR Table nos 5 3-1 and 7 2-2 for tabulations of existing RB penetrations and associated environmental carameters for eacn penetration, respectively). The penetrations will be classified as OA Category I including the face plates, face plate anchorages to the Reactor Building wall, and the welos between the face plates and the process piping consistent with the classification of the Reactor Buildiag itself. Also. a missile shield is being provided to protect the penetrations from tornado-generated missiles DHR piping and equipment within the Reactor Building will be supported to meet the QA Category ll/l seismic design criteria, that is, a failure of a DHR component or pipe during a postulated design basis earthquake will not prevent a safety-related component or pipe in the vicinity from performing its required safety-related function. The remainder of the piping, components, equipment, and supports is classified as OA Category ll/Ill.

4 Engineerina and Design _ Issues 4.1 Single Failure Considerations and Ovcrall Plant Decay _ Heat Removal Capability The total decay heat load, RPV plus SFP, is shown in Table 1. Details of how that table was generated are contained in section 4.10, below. As shown in Table 2, the heat removal capacity of the DHR is significantly greater than all existing plant systems which can remove heat from either the SFP or the RPV (during refueling outages). The ensuing discussion compares the pre-and post-DHR decay heat removal capabilities during three distinct phases of a typical refueling outage (assuming a full core offload was performed) with corresponding differing decay heat load conditions The conscauences of postulated single failures or natural phenomena (i e., design basis earthquake or high wind event) on overall plant decay heat l

removal capability dunng each phase are also evaluated.

4.1.1 Phase 1: From Plant Shutdown to Removal _of the SFP Gates Pnor to floodup of the reactor cavity and removal of the SFP gates, the maximum decay heat load which could be placed on the DHR would be the SFP decay heat load Per reference 4 the typical (pre-outage) SFP heat load is less than 2 X 10' BTU /HR. While the system is capable of cooling the SFP (only), that configuration ;s not a design consideration for the 5

system given its nominal heat removal capacity of 30 X 10 BTU /HR.

Refer to Section 8.1 of this report for evaluation of DHR operation in lieu of FPCC. Although this condition (Phase 1) is essentially not applicable to the DHR, it is instructive to review existing plant decay heat removal capabilities during this phase of a RO.

1 4.1.1.1 SFP Coolina The existing plant (1esign provides two means of removing decay heat from fuel in the SFP The normal means is via operation of the non-safety related Fuel Pool Cooling and Cleanuo (FPCC) system As delineated in the UFSAR. the fuel cool cooling neat removal functior' is a Page 10 of 33

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James A. FitzPatrick Nuclear Power P! ant Evaluation of the Decay Heat Removat System l

Report No. J AF-RPT-DHR-02413 l

l power generation objective and design basis; that is it is not a safety-related function The removal capabihties of the FPCC are approximate;y 6 3 and 10 X 10' BTU /HR in the nomina' and maximum configurations, respectively. The FPCC is not powered off of the plant safeguards buses and would be unavailable following a postulated loss of offsite power l

(LOOP). The FPCC is not seismically cesigned and would not be available following a l

postulated design basis seismic event l

Under accident conditions and in the initial period after a full core offload the safety related Residual Heat Removal (RHR) system, operating in the fuel pool assist configuration, can be l

used to supplement or substitute for FPCC operation. RHR system components are powered from safety related sources and would be available following a postutated LOOP. The RHR system is seismically designed and would be available following a postulated seismic event.

However, the RHR fuel pool assist piping ties into the FPCC piping. which, like the DHR, is supported to meet the QA Category ll/Ill seismic design criteria Hence, RHR fuel pool cooling assist capability is not assured following a design basis earthquake.

Prior to installation of the DHR, a postulated loss of SFP cooling in this configuration would have been mitigated by; (a) restoration of FPC cooling or (b) initiation of RHR fuel pool assist if FPCC had been lost due to a seismic event. SFP decay heat removal capability would have l

been lost and the pool temperature would increase accordingly. If repairs could not be affected. the pool would eventually boil SFP inventory would be ma;ntained using existing systems, with the ultimate capability of adding lake water to the pool via a fire hose l

Installation of the DHR in no way detracts from the ability to either restore FPC or initiate RHR l

fuel pool assist. If available, the DHR could be used to cool the SFP and obviate the need for l

initiation of RHR fuel pool assist.

l 4.1.12 RPV Cooling Prior to installation of the DHR, a postulated loss of SDC could only be mitigated (directly) by restoration of SDC. While such efforts were taking place, RPV and reactor cavity water temperatures would increase, and unless SDC were restored, the reactor cavity water would eventually boil. Installation of the DHR has no bearing on that scenario as it can not remove heat from the RPV until the reactor cavity has been flooded and the SFP gates removed 4.1.2 Phase 2: SFP Gates Removed and the Total Decay Heat Load Exceeds DHR Capabilityln i

the Nominal Confinuration l

Dunng this phase the decay heat load is assumed to exceed the rejection capacity of DHR l

operating in the nominal heat removal configuration Conservative!y assuming the DHR were 0

l operating at its design condition, i e, at a wet bu!b temperature of 73 F, the decay heat load during this phase must exceed 30 X 10'3 BTU /HR. Based on reference 4, such a heat load conservatively corresponds to approximately 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> (4.5 days) post-shutdown. As the plant is currently prevented from initiation of core offload for the first 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, post-shutdown, and is limited to transfer of four assemblies per hour from the RPV to the SFP, due to SFP heat removal capability, in the pre-DHR configuration it is highly unlikely a significant amount of fuel (and hence decay heat) would have been transferred to the SFP dunng this phase of the outage 4.1.2.1 SFP Coo.l.in.g

~

Prior to installation of the DHR. the decay heat load in the SFP was limited (based on the aforementioned limitations on fuel transfer) such that the peak SFP temperature following a l

Page 11 of 33

James A. FitzPatrick Nuclear Pcher Plant j

Evaluation of the Decay Heat Removal System Report No. J AF-RPT-DHR-02413 2

core off;oad and a postulated single failure in the FPCC would rema;n less than 150 F. Refer to reference 5 for details of that evaluation. Under those conditions, poo! heatup was lim.ted by operation of RHR fuel pool assist and FPCC (s;ngle pump operation) During phase 2 of the RO i e,during the initial stages of fuel transfer and prior to compietion of a core officad the decay heat toad in the SFP would remain within the capabilities of FPCC.

If FPCC were not available, RHR fuel pool assist would be initiated Installation of the DHR in no way detracts from the ability to either restore FPCC following a postulated failure or initiate RHR fuel pool assist With DHR available. the postulated loss of FPCC under these conuit:ons could be mitigated by simply operating the DHR system in its nominal configuration Operation of the DHR under these conditions would limit SFP heatup and obviate the need to initiate RHR fuel pool assist.

4.1.2.2 RPV Co_oling Pnor to installation of the DHR, a postulated loss of SDC could only be mitigated (d;rectly) by restoration of SDC lt should be noted the SDC configuration is not single failure proof. While such efforts were taking place, RPV and reactor cavity water temperatures would continue to increase, and unless SDC were restored, the reactor cavity would eventually boil. With the DHR installed, and operating in its maximum configuration, natural circulation cooling of fuel assemblies in the RPV would be established ar* '

'V and SFP heatup precluded Under such conditions, DHR operation in the nominal t

..guration would significantly decrease the net heat addition to the combined RPV and SFP water volumes, establish natural circulation cooling of fuel assemblies in the RPV, and effectively extend the period of time available for restoration of SDC.

4.1.3 Phase 3: SFP Gates Removed and the Total Decay Heat Removal Load is Lese Than DHR Capability _1.n the Nominal Confiauration Dunng this phase the decay heat load is within the rejection capacity of the DHR in the nominal configuration. As noted above, this would occur approximately 4.5 days post-0 shutdown if the DHR where operating at a wet bulb temperature of 73 F. At this point in a RO the amount of fuel transferred from the RPV to the SFP could vary widely. The aforementioned limitations on fuel movement, see section 4.12 above, effectively limit the rate of decay heat transfer to the SFPin the pre-DHR plant configuration. with the majonty of heat remaining in tne RPV.

4.1.31 SFP Coolina SFP cooling issues during phase 3 are identical to those described for phase 2 above with the exception a full core officad would be completed during this phase and the peak SFP water C

temperature following a postulated FPCC failure would be less than 150 F. The existing limitations on fuel movement are important in the pre-DHR configuration as those limitations are required to ensure adequate SFP cooling (the delay in offload initiation and the slow transfer rate serve to both minimize the SFP heat load directly and to maximize the amount of heat removal by SDC) In the post-DHR configuration the extent to which fuel has been transferred to the SFP is irrelevant as the DHR heat removal capability is independent of the location of the fuel. Under these heat loads, the DHR would be operating in the nominal configuration. Given a postulated failure of a DHR component, its redundant counterpart could be placed in service or the " failed" component by-passed and nominal heat removal capability restored SFP heatup would be limited to the short perio f time needed to " realign" the l

DHR.

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Page 12 of 33 i

e James A. Fit 2 Patrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No. JAF RPT-DHR 02413 4.1.3 2 RPV_ Cooling RPV cooling issues dunng phase 3 are identical to those discussed for phase 2 above, with the exception that the decay heat load would have decreased in proportion to the extent of completion of the officad. The heat load in this configuration would have continued to be picked up by SDC, until core officad was completed, and a postulated loss of SDC could only be mitigated through restoration of SDC. Post DHR installaton, a postulated loss of SDC could be mitigated either through restoration of SDC or operation of the DHR in either the nominal or maximum configurations. Given the vast heat removal capacity of the DHR, it would be possible for the plant to withstand the theoretical failures of SDC, FPCC, and a DHR component failure in this configuration while maintaining adequate decay heat removal capability.

As detailed in section 5.2 below. the DHR system is capable of removing decay heat from both the SFP and the RPV. Thus, the ensuing evaluation is valid regardless of the location of the fuel and independent of the status of core officad or reload activities.

4.2 Seismic Design The DHR is not designed to remain functional following a design basis earthquake DHR equipment and piping will be supported such that their (functional) failure dunng a design bases carthquake will not adversely effect safety-related equipment. That being the case instaliation and operation at the time of a postulated earthquake would have no adverse effect on plant safety.

S stem Power Supply 4.3 J

Electric power for the DHR will be provided from a reliable offsite source (13.2 KV). An underground feeder, consisting of a conduit encased in concrete, will be installed from the

" grey shed" in the yard to a 13.2 KV disconnect switch and to a 13.2 KV/4SOV transformer.

The transformer will be located in the yard east of the Railroad and Truck Bay and Gas Treatment Building. Adjacent to the transformer will be a 480V motor control center (MCC).

The MCC will distribute power to all DHR components A system one-line diagram is contained in figure 3.

The DHR electrical supply and distribution system is physically separate from existing safety related and non-safety related electric equipment in the power block For scheduled refueling outages, DHR reliability will be enhanced through use of a portable diesel generator which will be directly connected to the DHR 480V MCC. The portable diesel generator will be sized to start and carry DHR loads for operation in the nominal configuration i

and would be used in the event of loss of the normal 13 2KV supply. Transfer from the 13.2 KV source to the diesel generator would be accomplished manually.

4.4 Electrical Separatic3 l

All cabling associated with the DHR will be in accordance with applicable plant standards and procedures. All routing will be in accordance with existing plant procedures and standards to ensure the appropriate separation criteria are met.

1 Page 13 of 33 i

i

James A. FitzPatrick Nuclear Powe-P! ant Evaluation of the Decay Heat Removal System Report No. JAF-RPT-DHR-02413 Operational _ Radiation Fields 4.5 e

Radiation fields in the vicinity of DHR components and piping are expected to be the same as the fields exper;enced outside the FPCC heat exchanger and pumps rooms with FPCC in operation The pnmary difference between the FPCC ano the DHR in terms of operational radiation fields could be associated with the lack of shielding of the DHR onmary pumps strainer, and heat exchangers (vice the existing shielding of the analogous FPCC components) Radiation fields in the vicinity of the pumps and heat exchangers will be monitored during DHR operation and corrective actions initiated if necessary The exact nature of any corrective actions would be dictated by RES personnel in con _iunction with Engineenng 4.6 Fire Protection Issues Installation of the DHR will result in addit;onal combustible loading in the affected plant areas Fire Protection review of those loadings, as well as review of the DHR for Appendix R conciderations will bc performed as part of the Modification process.

4.7 Wind Loadings on Exterior Components _

Secondary side components to be located outdoors will be evaluated for 90 mph wind conditions. Those wind conditions are cited in the NRC SERT for the facility, reference 6.

4.8 Internal Flooding installation of the DHR will result in new energized piping routes in the Reactor Building.

Review of those routes for potential flooding, equipment qualification effects. and area accessibility will be performed as part of the Modification process 4.9 Heavy Loads Installation of the DHR is expected to take place while the unit is on line and it is imperative all equipment and piping rigging and lifts be performed in accordance with existing plant procedures and standards for movement of heavy loads. Of particular import are rigging and lifts internal to the Reactor Building and the SFP. All rigging and lifting will be performed in accordance with reference nos 7 and 8.

4.10 Decay _H_ eat Load _ Calculations 1

RPV and SFP decay heat loads throughout the life of the plant have been calculated using J

conservative methodologies and assumptions. Using the methodology of Branch Technical Position (BTP) ASB 9-2 (Reference 9), an estimate of the end of life decay heat loads versus 1

time after shutdown is provided in Calculation JAF-CALC-MISC-02244 (Reference 4). These heat loads are quite conservative based on the following assumptions used:

1)

Operation at the power uprate level of 2536 MWt is assumed (versus the present operating power level of 2436 MWt). The worst case heat input case considered corresponds to a high-energy 24 month.;ycle fuel load in the RPV (or the SFP).

2)

Tne calculation conservatively conside s an SFP capacity of 2661 bundles or sixty four (64) assemblies above the current licensed ccpacity of 2797 assemblies The full core offload is 560 oundles Page 14 cf 33 l

I James A. FitzPatrick Nuclear Power Plant l*

Evaluation of the Decay Heat Removal System t -

Report No. J AF-RPT-DHR 02413 1

3)

An end of life, full spent fuel pool less a full core offload capability (2861 - 560 =

2301), is assumed (Note if the SFP and RPV were both assumed full, the SFP heat load would not significantly increase, since the 560 empty spaces in the SFP would be filled with the oldest fuel in inventory. Using the oldest fuel would contribute less than 15% to the SFP heat load )

4)

The energy of the fuel batches discharged to the poolis maximized Lifetime plant capacity between the end of Cycle 12 and the end of life is assumed to be 90%. Fuel cycles are assumed to be 24 months with a 40-day refuehng, and 95% capacity factor from startup to shutdown.

The above assumptions were used to generate the decay heat versus time curve used to size the DHR system (nominal configuration). An additional margin of 10% was not added since the results are used to size the DHR heat removal capacity, ano an additional 10% over sizing I

is deemed not necessary. (Note: The ASB BTP specifies apphcation of margin to the oecay heat curve when used as input to safety-related analyses. i.e.,10 CFR 50 46 ECCS performance evaluations)

Note that, throughout this report, DHR heat removal capacity is expressed in terms of approximate number of days, or hours. post-shutdown. These times are Dased on l

conservative analysis of the worst-case design heat load condition and are provided for I

comparative purposes only. The actual time equivalents (as experienced in the plant) would be less than those predicted by such calculations, as can be seen from the conservative assumptions presented above to derive the decay heat values The calculated values of the SFP plus RPV decay heat, versus time after shutdown, for the time periods relevant to this evaluation, are given in Table 1.

Based on Calculation JAF-CALC-MISC-02244, the DHR system is sized to provide a design 8

l maximum heat removal capability of 45 X 10 BTU /HR when using one primary pump through both heat exchangers and using both secondary side pumps and both pair of single-celled cooling towers. In the design normal! nominal heat removal configuration, the DHR system 5

heat removal capability is 30 X 10 BTU /HR. These capabilities far exceed the capabilities of the existing plant systems / configurations capable of cooling both the SFP and the RPV (with the exception of RHR SDC). Thus, installation and operation of the DHR system represents a significant enhancement in the decay heat removal capability of the facility.

l The relative decay heat removal capabilities of existing systems and the DHR are summarized in Table 2. The detailed inputs to the respective heat removal capacity calculations are not identical. since these capacities were generated by different analyses and use different initial l

conditions and assumptions in all cases. however, the design inputs for the DHR sizing are the most conservative.

Figure 4, attached, presents the decay heat loads. and the heat removal capabilities available from the DHR system, the RHR SDC or RHR assist modes, and the FPC system, versus time after shutdown.

l The only plant feature capable of removing a greater amount of decay heat is RHR SDC. As can be seen on Figure 4, installation of the DHR system does not eliminate the need for SDC operation at the beginning of an outage.

1 Page 15 of 33 l

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James A. Fit 2 Patrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No. [[::JAF-RPT-DHR|JAF-RPT-DHR]] 02413 5

_ Nuclear S_afety Considerations 5.1 Maintenance _of SFP Water Level Followinla_P_ostulated Breach of System Pressure Bounda ry.

The DHR system suction and discharge piping in the vicinity of the SFP include holes to preclude siphoning of the SFP in the event of a breech in DHR primary loop pressure boundary. Those holes are functionally equiva!cnt to the vacuum breakers provided on the existing FPCC. The elevation of the holes in the DHR sparger pipes will be such that adequate inventory would be maintained in the SFP following a postulated breech of DHR pnmary loop pressure boundary.

The potential for loss of refueling cavity inventory due to postulated failure of the of the reactor cavity seal has previously been evaluated for JAF as part of our response to INPO Significant Operating Experience Report (SOER) 85-1 (reference 10). Installation and operation of the DHR has no bearing on the SOER evaluation in terms of the design of the cavity seal and the potential for catastrophic failure As detailed in the SOER response, a design review was performed for JAF which concluded gross failure of the seal was not a credible event. The review went on to address associated issues such as the installation of nozzle dams. etc The results of those reviews are not affected by DHR operation.

5.2 Natural Circulation Cooling of Fuel Assemblics in_the RPV Thermal hydraulic calculations performed by General Electric (Reference 11) and NYPA Calculation JAF-CALC-RHR-02380 (Reference 12) confirm the ability of the DHR to remove the decay heat from fuel assemblies located in the RPV. that is, prior to core offload and subsequent to reload. The calculations assume the RPV head removed. the reactor cavity to be flooded up. and the SFP gates to be removed. Under such a condition. i e., prior to core offload, decay heat from the fuel in the RPV would oe removed 'uy natural circulation. The analyses also confirm natural circulation cooling would continue throughout the officad process These calculations confirm the establishment of natural circulation currents in the RPV and reactor cavity whenever the DHR is operated with the SFP gates removed and the cavity flooded to refueling level The analyses considered the following factors-1) operation of the DHR in its design maximum heat removal mode or its nominal heat removal mode 2) the time after shutdown at which the core is offloaded (ihe calculations assume offload into the SFP commences at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and is complete at 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> post-shutdown) 3)

the location of the DHR discharge pipe in the SFP, including the theoretical case j

where the DHR suction and discharge pipes are in close proximity to each other and near the surface of the pool.

For these analyses, the initial bulk SFP water temperature is assumed to be at 114 F. The results of those calculations show an initial increase in RPV water temperature, essentially core average temperature, upon termination of RHR SDC and forced flow through the core i

The magnitude of temperature increase is conservatively predicted to be on the order of 15-

)

20*F. Dunng this penod of temperature increase, the water volumes in the RPV. the i

reactor cavity, and the SFP are approaching a quasi-steady state equilibrium

]

condition. Approximately one nour after SDC termination. natural circulation flow Page 16 of 33 l

James A. FitzPatrick Nuclear Power P: ant Evaluation of the Decay Heat RemovaI System Report No. JAF RPT-DHR-02413 patterns are estabiished and RPV water temperature begins to decay exponentially and approaches the SFP water temperature The peak core exit temperature predicted by either the GE or NYPA calculations is 138'F. The peak water surface temperature is predicted for the theoretical case where the DHR suction and discharge pipes are in close proximity to each other and near the pool surface The peak value for that case is 133*F. The maximum predicted bulk surface temperature of the SFP/RPV/ reactor cavity water volume for all other cases analyzed is 118*F.

Sensitivity analyses indicate the surface temperature of the combined SFP. reactor cavity water volume is relatively insensitive to; 1) the fraction of fuel offloaded from the RPV, 5

2)

DHR operation at design maximum heat removal capacity (45 X 10 BTU /hr) versus nominal heat removal capacity (single train operation rernoving up to 30 X 10' BTU /hr), and 3) the approximate time, post-shutdown. of DHR initiation. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> vice 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> assuming no other method of decay heat removal Details of natural circulation verification to be performed dunng RO12 are given in section 71 l

below.

General Electric has analyzed the heat transport from the core (RPV) to the SFP for Plant Hatch, Hope Creek, and the J.A. Fitzpatrick plants (Reference 13). A review of the geometry l

of the RPV/SFP/ reactor cavity at these plants was performed. The parameters included the distance between the core centerline to fuel pool gate, the elevation of the bottom of the canal versus the RPV flange, the elevation difference of the flange versus the refueling floor deck, the elevation difference between the bottom of the canal and the bottom of the pool, and the fuel pool gate width. The review concluded that there were no significant differences between the three plants that would affect natural circulation flow between the core and the refueling pool. Therefore, there is no technical requirement to test the natural circu!ation capability; that l

is, one is merely proving that hot water rises and cold water sinks.

l 5.3 Potential for Excessive Cglina of the SFP l

The heat removal capability of the DHR is vastly greater than the decay heat load of the SFP l

during non-refueling outage conditions and nominal heat removal capacity exceeds the combined RPV and SFP decay heat loads approximateiy 4.5 days post-shutdown. As the DHR is a manually controlled system, the design of which does not include automatic temperature control features, the potential does exist for excessive cooling of both the RPV and the SFP.

The lower temperature limit on SFP temperature is a plant safety limit derived from criticality analyses of spent fuel stored in specific rack designs. The lower temperature limit for the RPV is based on metal brittle fracture analyses.

Operation of the DHR is acceptable provided the appropriate administrative controls and procedures are in place to ensure plant operators are aware of the potential for overcooling and have been properly trained in mitigating actions. Under any condition or configuration.

manualinp of operating pomary side pumps at the DHR control panel would terminate any possible cool down.

j Page 17 of 33

l James A. Fit 2 Patrick Nuciear Power Pl ant Evaluation of the Decay Heat Removal System Report No. JAF-RPT DHR 02413 l

54 Technical Specification and Design _ Da.s.es_ Document Revievi The following pertinent Technical Specification sections were reviewed to perform this report 3/4.0 Applicability tand Bases!

3/4.2 instrumentation (and Bases) 3/4.7 Containment Systems tano Bases) 3/4.9 Auxiliary Electrical Systems (and Bases) 3/4.10 C Spent Fuel Storage Pool Water Levei (and Bases) 3/4.11 Additional Safety Related Plant Capabilities (and Bases) 6.0 Administrative Controls. including Sections 6.5. Review and Audit, and 6.8. Procedures and Programs Based on review of the above sections, it has been determined a change to the Technical Specifications is_not required.

Installation of the DHR does require a change to the Design Basis Documents. The Design Basis Document for the Residual Heat Removal System,010 (Reference 14), requires a discussion of the use of the DHR system.

5.5.

UFSAR RevicM UFSAR Sections 9.3 and 9.4 (as a minimum) will be revised to clarify and update the decay heat loads assumed, the various heat removal assumptions, and the various modes of operation which are available to the operators with the addition of the DHR system.

Clarification is required due to the addition of the DHR system. but also due to the various decay heat removat assumptions used in previous submittal, the number of fue! assemblies assumed in the SFP, the discussions of RHR assist as the only method, etc. Also. the results of the thermal analyses which support installation and operation of the DHR system shcw DHR l

operation to be independent of fuel location. Mcnce. the UFSAR will be revised to reflect the fact existing limitations on fuel movement, a minimum delay of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> post-shutdown prior to initiation of fuel movement and a maximum fuel transfer rate of four assemblies per hour into l

the SFP. would not be applicable when the DHR was available and operating. Said limitations would remain valid if the DHR were not available during a refueling outage.

A draft UFSAR revision will be prepared as part of Modification Package number F1-95-121, l

" Decay Heat Renioval System, Project CJ3110."

1 The following UFSAR Sections were reviewed to perform this SE, and to determine whether a change to the UFSAR is required:

j Section 1.2 Definitions Section 1.3 Methods of Technical Presentation Section 1.4 Classification of BWR Systems Cnteria, and Requirements for Safety l

Eva!uation Section 7.6 Refueling Interlocks Section 7.12 Process Radiation Monitoring System Section 7.13 Area Radiation Monitonng System Section 8.6 Emergency AC Power System Section 9 3 Spent Fuel Storage Section 9 4 Fuel Pool Cooling and Cleanup System Section 9 9 Heating Ventilation. and Air Conditioning Systems. including Section 9.9 3.3 Reactor Building Ventilation System Page 18 of 33

James A. FitzPatrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No. J AF-RPI-DHR-02413 Section 12.

Classification of Structures and Equipment Section 12.3 Description of Pnncipal Structures Section 12 4 Structural Loading Conditions Section 12.6 Analysis of Spent Fuel Storage Pool Section 13.8 Plant Procedures Section 14.5 Analysis of Abnormal Operational Transients and Reactor Vessel Overpressure, including Section 14.5 8. Event Resulting in A Core Coolant Temperature Increase (loss of RHR shutdown cooling)

Section 14.6 Analysis of Design Basis Accidents Chapter 16 Appendices' Section 16 5, Pressure Integrity of Piping and Equipment Pressure Parts, and Section 16 6. Conformance to AEC Design Cntena Chapter 17 Quality Assurance Program. Appendix 17.2B, Conformance with NRC Regulatory Guides: and Appendix 17.2C. Plant Administrative Procecures Genera: List 6.

S_ystem Installation and Modification Acceptance Testing As mentioned above, the system is expected to be installed and tested prior to RO12 and the unit is expected to be on line dunng that period. The appropnate administrative controls and procedures are in place to ensure system installation does not adversely effect plant operation or plant safety.

Acceptance testing will be performed with the system isolated from the SFP, i.e., spool pieces will be used local to the SFP to provide a primary side flow path which does not involve circulation of radioactive fluid to and from the SFP.

7.

Initial Thermal Performance Testing of the DHR System Decay Heat Removal Capability Modificationacceptance testing of the DHR will be performed prior to RO12 However, it is not meaningful to perform a thermal performance test of the DHR with the system aligned for SFP cooling (only), i.e., prior to reactor cavity floodup and removal of the SFP gates, due to the limited heat load Therefore. performance testing of the system will be performed during RO12 7.1 Thermal Performance Testinn During RO12 The DHR system will be placed in service and FPCC and SDC secured after the SFP gates are removed With SDC secured, but available. the ability of the DHR to remove the combined RPV and SFP decay heat load will be functiona!!y venfied by operating the system and venfying a negative trend in DHR pomary side suction temperature. The combined SFP ana RPV heat loads at that point are expected to be within the capability of the DHR ooerating in the nominal heat removal configuration.

Detailed engineering data obtained during that (and subsequent periods) will be used to evaluate DHR thermal performance However, the functional test of the DHR will be simply to maintain or decrease pool water temperature when RHR and FPC are not in service.

I Page 19 of 33

Jarnes A. FitzPatrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No, JAF-RPT-DHR-02413 7.2 Verification of Natural Circulation During_ Initial System _ Operation _(RO12)

The heat removal capability of the DHR is independent of location of the fuel and hence is not chronolog;cally tied to the initiation of core officad acavities However. it is expected - but not required - that both the maximum and nominal heat removal configurations of the DHR system will be tested prior to snitiation of offload Assuming such to be the case, the estaolishment of natural circulation cooling of fuel in the RPV will be verified by observation of a decreasing trend in SFP water temperature with RHR SDC not in operation 8.

System Operation 8.1 DHR Operation with the Unit on line The DHR can be used in lieu of FPCC to cool the SFP with the unit on line; the sole safety concern being the potential for overcooling which is addressed in section 5 3 above 82 DHR Operation Durinn Refueling _Outa2es - SFP Gates Installed System operation in this configuration is acceptable. At the start of the outage, this operating condition is identical to use of the DHR in lieu of FPCC with the unit on line During ROs (when the SFP gates are installed) there are two possible situations during which the DHR could be used.

The first condition would correspond to a hypothetical situation wherein core officad had been completed and for some reason the SFP gates were installed, i.e., to facilitate in vessel inspections or repairs. The DHR could be operated in the nominal configuration in lieu of FPCC and RHR fuel pool assist.

The second condition would occur at the end of each outage after core reload had been completed and in advance of vessel reassembly activities The DHR could be operated in this configuration in lieu of FPCC and RHR fuel pool assist J

8.3 DHR Operation Durina RefuetLng Outages - SFP Gates Removed.

Operation of the DHR in this configuration is the design configuration for the system. As detailed in section 4.1 above, operation of the DHR in the nominal configuration (beginning approximately 4.5 days post-shutdown) eliminates the need for FPCC and RHR (SDC and fuel pool assist) operation while maintaining full redundancy for decay heat removal The existing JAF Technical Specifications do not require RHR SDC operability when the reactor cavity is flooded However, the Standard Technical Specifications ( STS), reference 15, would require one train of SDC to be available and operating as long as irradiated fuel were in the RPV. The JAF Tennical Specifications do not mention the DHR. However, the l

" REQUIRED ACTION" statement of STS 3 9.8 states. " Verify an alternate method of decay heat removal is available" if RHR SDC were not available. The bases for that STS section reads in part as follows: "..the volume of water above the RPV flange provides adequate heat removal capability to remove decay heat from the core" and goes on to recognize the ability of other plant systems, i.e., Reactor Water Cleanup. to handle the RPV decay heat at some point in the outage.

Based on figure 4, the total (SFP and RPV) decay heat load would decrease Delow the capacity of the DHR in the nominal configuration (at the design wet bulb temperature) approximately 4.5 days post-shutdown. From that point on in an outage. assuming the cavity Page 20 of 33

James A. FitzPatrick Nuclear Power Plant Evaluation of the Decay Heat Rernoval System o

Report No J AF-RPT-DHR-02413 flooded and the SFP gates removed, the DHR would be single (active) failure proof, with operator action, for decay heat removal regardless of the status of refueling activities The unavailability of RHR SDC at that point in an outage wouio, under the most conservative interpretation, be tantamount to being in tne STS " ACTION" statement. In reality, DHR availability at that time would provide redundant means of RPV (and SFP) decay heat removal.

It is more reasonable to consider RHR SDC and the DHR (nominal configuration) to be redundant means of decay heat removal when flooded up with the SFP gates removed By administrative!y requiring SDC to remain operable until the decay heat load was within the capability of DHR in the nominal configuration, plant management will ensure the decay heat removal function was single (active) failure proof throughout the outage. During the initial part of an outage, i e., beginning approximately 1.5 days post-shutdown until approximately 4.5 days post-shutdown, RHR SDC and DHR (maximum heat removal configuration) would be considered redundant (provided the cavity were flooded and the SFP gates removed). Once the total decay heat load has been shown to be within the capability of DHR in the nominal configuration, RHR SDC and DHR (with all redundant equipment operable) should be considered as three methods of decay heat removal Thus removai of SDC from service under the-e conditions would be acceptable as the decay heat rernoval function would rematn single (active) failure proof. It should be noted the detailed analyses presented in section 5 2 of this report venfy the ability to ' monitor RPV temperature' (in the STS sense of the phrase) using temperature instrumentation on the pnmary side inlet to the DHR heat exchanger.

Additional measures to maintain key spare DHR components onsite would provide an even greater measure of security and further enhance the " reliability" of the decay heat removal function and reduce the potential for loss of decay heat removal events.

9.

Additional Supportina Analy_ses 9.1 Evaluation of Postulated Single Active Failures of DHR Components and Prospective u

Mitinating Actions with the System Operatina in the Nominal heat Removal configuration As noted above, the DHR is designed to withstand a single active failure while maintaining, 5

with operator action, a heat removal capability of 30 X 10 BTU /HR. An assessment of those potential failures, and associated mitigating actions, follows 9.1.1 Postulated Active Failure of a Mechanical Component The normal configuration of the DHR system, a singie primary pump and heat exchanger with neat removed by a single secondary pump and single, two-cell, cooling tower package would be in service Therefore, single active failure of any single mechanical component in the primary or secondary side is mitigated by operation of the backup component. Moreover, if one cell, basin, or fan of a cooling tower package fails then the backup components using one of the two backup tower cells may be placed into service. In addition, the design allows for the operation of either primary or secondary pump operating "through" either heat exchanger.

l These design features provide improved capability and flexibility in comparison to the FPCC system, wherein a failure of one 50% capacity heat exchanger results in immediate loss of 50% of the heat removal capacity.

The primary side strainer is the single primary side component which does not have a backup This strainer is provided for ALARA purposes, that is, to preclude excessive buildup of radioactive particles in the heat exchangers. A loss of the strainer by plugging or fouling is mitigated by bypassing the strainer and continuing w th the DHR heat removal function The Page 21 of 33

l James A. Fit 2Patrich Nuclear Power Pl ant Evaluation of the Decay Heat Removal System o

l-Report No. JAF RPT-OHR-02413 l

normal purity and clarity of the SFP water. from continuing use of the SFP filtration system prior to the use of the DHR system, proviaes assurance that the biocmage of the strainer.s a low probabil,ty event. Moreover, it provides assurance tnat when the OHR system is olacec into service the SFP water will not contain significant auantities of particles which could result in a bypass of the strainer.

The sing!e secondary side component which does not have a backup is the pressure control valve. 32PCV-100. which maintains the secondary side heat exchanger exit pressure greater than the primary side heat exchanger inlet pressure In the event of a failure, the PCV would be isolated and the manual bypass valve opened PCV failure might require the operator to restart a seccndary side pump Assuming the initially operating secondary pump tripped, the j

operator would then start a primary side pump (since the low secondary-to-primary pWd would trip the primary pump). These actions restore the decay heat removal capability of the DHR system.

Based on the above, the provision of a backup component to mitigate the failure of a single mechanical component, (or manual action to bypass a failed component) provides a reliable l

means of removing decay heat from the SFP and/or the RPV. From a mechanical / hydraulic / heat removal perspective, therefore, the DHR system can supplement or replace the FPCC system and provide a system that is more tolerant of a single mechanical failure 9.1 2.

Postulated Active Failure of an_Flectrical Instrumentation and Control Comp _onent m

The DHR system power is normally supplied from a plant 13 2 KV switchgear J02 to a 13 2 KV-480V transformer outdoor power center. The power center contains an oil-filled transformer to step down the voltage and a motor control center (MCC) which distributes the j

power to the motors. heaters, and other loads The MCC contains fused combination starters I

and contactors and a 120/208V distribution transformer and panel. The equipment is NEMA grade, sized to handle the full operating current of the loads and available short circuit of the power system. Cables are per ICEA standards, installed in tray and conduit. The equipment is grounded to the plant grounding system per NEC and IEEE standards. Lighting is provided at the power center and at the package cochng tower areas The DHR system electrical power supplies and distribution system are independent of existing plant systems, both safety-related and non-safety related. Thus, a failure of a DHR electrical system or component cannot affect safety-related electrical equipment or componenty During refueling outages, a truck-mounted i

diese! generator is provided near the DHR system 13 2KV-480V transformer to supply emergency power should the normal power supply fail. Thus, for a loss of offsite power, the DHR system is provided with its own dedicated backup power supply, which is also available in the event of a failure of the dedicated 13.2kV power supply. As indicated on the one-line l

diagram, fused contactors are provided to each individual load such that failure of one contactor or load does not affect the other loads. Therefore, the loss of one load is enveloped by the considerations of single active fa: lure of mechanical components discussed above.

l The DHR system will be manually controlled locally from the control panel Differential pressure instrumentation across the primary loop strainer is provided to initiate the automatic backwash feature. Differential pressure between the pomary and secondary sides of the DHR heat exchangers is also monitored and features provided in the design automatically trip the j

running DHR primary pump (s)if the sensed pressure differential oecreases below 10 psid A cornmon trouble alarm is provided for the DHR system at the FPC local panel on elevation 326'-9" which also alarms the FPC annunciator in the Control Room. Tne common trouble alarm indicates any of the follow:ng conditions' i

l Page 22 of 33 l

l

i

(

4 James A. Fit 2 Patrick Nuclear Power Plant j'

Evaluation of the Decay Heat Removal System l

Report No. J AF-RPT-DHR-02413 1) low flow in the pnmary system 2) low flow in the secondary system 3) primary loop pump trip 4) cooling tower fan high vibration 5) cooling tower basin low water level 6) transformer winding temperature - high A single failure of an I&C component therefore does not preclude the DHR heat removal function, and sufficient redundancy and alarms are provided to give ample time to the operators to effect required repairs 9.1.3 Evaluation of Potential Mitigating Actions The mitigating actions necessary to restore DHR operation in the nominal configuration following a postulated single active failure or loss of normal power supply are straightforward l

and, with the exception of use of a backup source of secondary side makeup, involve only DHR components. Adequate diagnostic information is made available to the plant operators, and the appropriate DHR operating procedure (s) will be in place, to ensure that the DHR system nominal heat removal operation can be restored in a timely manner.

As an additional measure, NYPA will maintain selected spare DHR system components available, including spare primary and secondary side pumps, on site during refueling outages.

In doing so, and in having the corresponding implementing procedures in place, the time to effect repair and restore DHR redundancy is minimized.

9.14 Time Frame for the implementation of Mitigatin1 ctions A

Early in an outage, when the combined RPV and SFP decay heat load might exceed the l

removal capability of a single DHR train NYPA will require a RHR SDC train remain OPERABLE Under those conditions, a postulated DHR failure (of either of the two trains or loss of power supply) could be mitigated by initiation of SDC and, if FPCC is unavailable RHR fuel pool assist Such actions are already defined in plant procedures and can be j

accomphshed relatively quickly.

Although not a licensing basis consideration, if RHR is unavailable in the condition where the decay heat load is less than or equal to the capacity of a single DHR train, the parameter chosen herein to assess recovery action is the time before boiling occurs in either the SFP or the combined SFP/ reactor cavity volume. That time is conservatively calculated as l

approximately 8.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (per reference 16). Any of the mitigating actions described above (following a postulated single active failure of a DHR component or loss of the normal DHR power supply) can easily be accomplished in significantly less time and 30 X 10* BTU /hr heat I

removal capability restored 9.2 Estimated Time to Boiling Calculation JAF CALC-MISC-02373 (Reference 16) was performed to determine the minimum time to reach boiling conditions in the SFP assuming a full core offload is initiated 2 5 days into the outage and completed by day 5 after shutdown (that is. tull core offload in approximately l

l Page 23 of 33

James A, FitzPatrich Nuclear Powcr Plant Evaluation of the Decay Heat Removal System Report No. JAF RPT-OHR-02413 2 5 days) This conservative assumption is made to maximize the SFP decay heat load at the point in time DHR (and any other methods of oecay heat removal) are assumed to oecome unava:!able The core offload beginning at 2 5 cays. and the resultant ncrease in the decay heat of the SFP are shown graphica!!y on Figure 4 The reactor cavity is assumed flooded up and the SFP/RPV! reactor cavity water volumes are interconnected since the fuel pool gates are removed Under these conditions the calculated worst-case minimum time to boil in the SFP is approximately 8.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> The maximum SFP heatup rate as fuel is being moved into the SFP over 2.5 days is 12.1'F/hr which is concurrent with the minimum time to boil (on day 5 after shutdown). The SFP boiloff rate is calculated to be about 28.444 pounds per hour, and the SFP makeup rate under these conditions is calculated to be about 59 gpm required-Besides being based on the overly conservative decay heat loads previously described, this calculation does not account for any heat absorption into the SFP steel liner or concrete walls, nor does it account for any other heat sinks which would delay the SFP water heatup to the boiling point.

This time to boil (even though it is pessimistic because of the layer upon layer of conservative assumptions used throughout the supporting calculations and the assumptions made for the time to boil calculation) is of interest because it is a very long time for accident mitigation purposes A time frame of eight or more hours allows ample time to recognize the problem.

plan mitigative measures, and carry out those plans Ample water supplies exist on the nuclear plant site, and Lake Ontano is an ultimate water source if need be. The 200.000 gallon Condensate Storage Tanks, whien are seismic Class I, will survive an earthauake without inventory loss, and one tank provides ample capacity for fuel pool makeup The Fire Protection Systern, with two diesel-driven fire pumps. is another source of water supply and/or pumping capability, notwithstanding the numerous pumps in the plant. Fuel pool boiling is therefore not a significant safety concern, and in fact. as the next section demonstrates. is not a concem for the public health and safety.

9.3 Radiological _Consequylces of a Protracted Loss of Decay _ Heat Removal The licensing basis for JAFNPP does not require performance of either a SFP analysis of the time to reach boiling, or a radiological assessment of the results of a protracted loss of SFP cooling. Nevertheless, these analyses have been performed in support of the DHR system installation and operation. Once again, very conservative assumptions are made; in this case, it is assumed that makeup water is supplied to the SFP to maintain the SFP boiloff, but restoration of SFP cooling is not assumed, i e., no credit was taken for heat removal due to feed and bleed (steaming) of the pool.

Conservative calculations were performed (Reference 17) to determine site boundary, low population zone (LPZ) and control room doses following a loss of all decay heat removal The calculation model considers the combined SFP/RPV/ reactor cavity water volumes physically interconnected via the refueling canal (flooded up) w:th the fuel pool gates removed R2ference 18 was utilized for the distance factors used in the analyses. The analyses do not take credit j

for the time delay (at least 81 hours9.375e-4 days <br />0.0225 hours <br />1.339286e-4 weeks <br />3.08205e-5 months <br />) between the postulated loss of cooling and the onset of boiling. Also, the analyses take no credit for hold-up or retention of radioactive airborne geses J

in the Reactor Building. The net effects of these two assumptions are to assume a radioactive release outside of the Reactor Building immediately uron loss of cooling.

As shown in Table 3, the worst case conservative estimate of the control room dose is only about 15% of the applicable limit. All of the other calculated doses are much lower percentages of their associated regulatory limits Page 24 of 33

3 e

James A. FitiPatrick Nucicar Power Plant Evaluation of the Decay Heat Removal System Report No. JAF-RPT-DHR-02413 Additional analyses were performed to determine thyroid. whole body, and skin doses on the refueling level of the reactor building. These calculations assume a 70,000 scfm air change at the refueling level (corresponding to reactor building ventilation operation). The calculated thyroid dose rate could be as high as 1 Rem /hr, and could necessitate the use of breathing apparatus during the hypothetical recovery actions on the refueling level of the reactor building Hov,ever, these estimated doses are well above what would be expected were " realistic" assumptions to be used (Reference 19) 9.4 DHR System Reliability. Assessment _

In addition to the above failure modes and effects analyses. a reliability assessment (Reference 20) was performed for the DHR system (in its nominal heat removal capability configuration) to pinpoint any potential vulnerabilities and to plan contingency measures. The results of the reliability assessment indicate that the DHR system design is nignly rehable The highest contributors to total system unreliability are associated with the concurrent failure of the pumps (either both primary side or both secondary side). or with the failure of the pressure control valve,32PCV-100. These vulnerabilities are compensated for by provision of duplicates of the major components primary pump, secondary pump, and heat exchanger, or

]

manual bypass of the strainer or PCV. The selected spares are maintained onsite and 1

therefore the time to effect the repair to restore the system to its two-loop configuration is minimized. In the event that the offsite power supply to the DHR system is not available. the DHR electrical power system would be provided by a backup truck-mounted diesel generator during refueling outages The truck-mounted diesel generator will be hard-wired into the DHR system MCC 9,5 Effects of Elevated Water Temperature on SFP Comp _onents_

i The effects of elevated SFP water temperature were evaluated as part of the previous SFP expansion (rerack). The results of these evaluations are summarized in the NRC Safety j

Evaluation (Reference 5) for Operating License Amendment No.175 Those results are unaffected by the installation of the DHR system, with one positive exception: the determination of the maximum pool water temperature given a single active failure in the FPCC system l

Two cases were considered for the rerack; FPCC failure after a partial (approximately one-third) core offload, and FPC failure after a full core offload. The peak SFP water temperatures calculated for those cases are less than 140 F and 150 F respectively, and consider RHR operation in the fuel pool assist mode.

There is no directly analogous full core offload evaluation with the DHR system installed and operating in its normal heat removal capacity (in lieu of the FPCC system), as the heatup of the SFP is mitigated through operation of the backup DHR component or train. Assuming that it takes up to two hours to effect the backup, pool heatup would be a maximum of approximately 24 F (from an initial assumed SFP water temperature of 114 F). Note that this maintains the water temperature below the 140 F guideline of Standard Review Plan Section 91.3, reference 21. Note too, the heat load analysis performed in support of this DHR evaluation conservatively assumes that core offload is initiated 2.5 days after shutdown and completed within approximately 2.5 days (whereas the previous SFP expansion analyses assumed a 4-day delay after shutdown prior to fuel movement, and limited the transfer rate of assemblies to four per hour). Hence, it is obvious that the installation and operation of the DHR system is beneficial in limiting peak SFP water temperatures following a postulated single active failure. Since the peak SFP water temperatute for the DHR case is less tnan the peak SFP water temperature for the FPCC case. the previous equipment evaluations are bounding Page 25 of 33

James A. FitzPatrich Nuclear Power Plant Evaluation of the Decay Heat Removal System l

Report No. JAF-RPT-OHR-02413 and need not be re-cerformed for this evaluation 10 C o n c l u s i o n s_a n d_R_e c o m m e n d a tio n s The DHR will provide the ability to remove the decay heat from the RPV and the SFP independently of the RHR system and the FPCC system. Provision of the DHR system to augment or replace existing decay heat removal capabilities results in a significant increase in flexibihty and added assurance decay heat can be removed for various transients and accidents The installation and operation of the DHR will 1)

Provide enhanced SFP cooling capability during normal plant operation, for example, I

when the FPC is unavailable because of test or maintenance.

2)

Provide enhanced SFP cooling capability during refueling outages.

3)

Provide an independent means of removing decay heat from fuel elements in the RPV.

l provided the reactor cavity is flooded and the SFP gates removed 4)

Reduce the need for RHR operation in the fuel pool cooling assist mode RHR operation in this configuration necessitates throttling of the RHR service water (RHRSW) valve on the outlet of the RHR heat exchanger in service Throttling of the l

heat exchanger outlet valve is thought to be a contributing factor in the I

erosion / corrosion observed in the downstream RHRSW piping.

l Use of RHR in the fuel pool cooling assist mode has historically resulted in decreased j

l clarity of SFP and reactor cavity water. Operation of the DHR prior to and during refueling would eliminate that concern.

l l

5)

Reduce the need for RHR operation in the SDC mode during refueling outages (after approximately 4.5 days).

6)

Provides for additional flexibility in terms of outage planning consistent with ALARA and Shutdown Risk Management principles (Reference 22).

Based on the reviews summarized in this report, instal!ation and operation of the DHR would not l

constitute an unreviewed safety question. Therefore, it is recommended Nuclear Safety Evaluations be developed pursuant to 10 CFR 50.59, for the installation, testing, and operation of the DHR.

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James A. FitzPatrick Nuclear Power Plan; I*

Evaluation of the Decay Heat Removal System Report No. JAF-RPT-DHR-02413 l

l TABLE 1 l

OECAY HEAT LOADS Approximate Decay Heat, 1

Days After BTUlbr Shutdown 1

49.77 X 10' 1.5 44.69 X 10' 5

2 41.00 X 10 3

35.48 X 10' 4

31.57 X 10' 5

4.5 30.04 X 10 5

28.73 X 10' 5

10 21.82 X 10 15 18.87 X 10' 20 16.93 X 10' 30 14.22 X 10' 40 12.40 X 10 5

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James A. FitzPatrick Nuclear Power Plant Evaluation of the Decay Heat Removal System Report No. J AF-RPT-DHR-02413 TABLE 4-2 DECAY HEAT REMOVAL CAPABILITIES Method of Decay Heat Removal 10' BTU /hr DHR System (Maximum) 45 DHR System (Nominal) 30 RHR Assist + FPC (1 pump,1 HX) 24 FPC (2 pump,2 HXs) 10 FPC (1 pump,2 HX) 6.3 1

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Page 28 of 33 l

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James A. FitzPatrick Nuclear Power P ant Evaluation of the Decay Heat Removal System Report No. JAF RPT-DHR 02413 T_ABLE 3 Dose Consequences Following A Postulated Loss of Decay Heat Removal Receptor Thyroid Whole Body Skin Dose, Remarks Location Dose, Rem Dose, Rem Rem Site Boundary 6.2 X 10-3 5.2 X 10 1.5 X 10' 4

(Dose Limitw)

(300)

(25)

(N/A)

(From 10 CFR Part 100)

Low Population 2.2 X 10 2 1.3 X 10' 3.6 X 10' Zone (Dose Limits )

(300)

(25)

(N/A)

(From 10 CFR Part 100)

Control Room 4.5 5 4 X 10

7.9 X 10 (Dose Limits )

(30)

(5)

(30)

(Thyroid, Skin from SRP 6.4; WB from 10 CFR 50 App A, GDC 19)

The regulatory dose limits for a design basis Loss of Coolant Accident (LOCA) are given in parentheses. The exposure intervals assumed are two hours at the site boundary, and 30 days at the LPZ and in the Control Room. Thyroid doses are based on the dose conversion factors in TID-14844 (Reference 19).

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Page 29 of 33

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