ML20116D902
| ML20116D902 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/16/1996 |
| From: | Herrmann T, Vehstedt K POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20116D890 | List: |
| References | |
| JAF-SE-96-042, JAF-SE-96-042-R00, JAF-SE-96-42, JAF-SE-96-42-R, NUDOCS 9608050009 | |
| Download: ML20116D902 (18) | |
Text
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hM O -iP3 O -JAF 10 CFR 50.59 NUCLEAR SAFETY EVALUATION Fcrm Number: JAF-SE-96-042 Revision:
0 Activity: a Modification O Procedure O Test O Experiment O Other Activity Number: nod fMF/-95L/2)
Title:
Use of the Decay Heat Removal System in Various Plant Modes and Confiaurations A. The proposed activity:
1.
Odoes adoes not increase the probability of occurrence of an accident evaluated l
In the safety analysis report.
i 2.
Odoes adoes not increase the consequences of an accident evaluated previously j
in the safety analysis report.
3.
Odoes adoes not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the safety analysis report.
4.
Odoes adoes not increase the consequence of a malfunction of equipment important to safety evaluated previously in the safety analysis report.
5.
Odoes adoes not create the possibility of an accident of a different type than any evaluated previously in the safety analysis report 6.
Odoes adoes not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the safety analysis report.
7.
Odoes adoes not reduce the margin of safety as defined in the basis of any Technical Specification.
8.
Odoes adoes not involve an unreviewed safety question based on questions 1 through 7.
9.
Odoes udoes not degrade the Security Plan, Quality Assurance Program or the Fire Protection System.
B.The proposed activity:
1.
edoes Odoes not require a change to the Final Safety Analysis Report as indicated in Section 3 of this Nuclear Safety Evaluation (NSE).
~
2.
udoes Odoes not require action tracking of the items indicated in Section 5 of this NSE.
C. This proposed activity:
1.
Odoes adoes not require a change to the Technical Specifications.
2.
Odoes adoes not require an Environmental Impact Evaluation.
a: U 3.
sdoes Odoes not require a change to Design Basis Documents.
2:
Date: M Prepared by: K. J. Vehstedt oe o
Reviewed by: T. J. Herrmann w-Date:
/ f 17 6 Approval % Dis proval O PORC Mtg. hd ~ 067 Date:
Recommended:
244$/ 3 7 )/4.CT7dd Date:
n.
Approved by:
,8 -
,7 /6,/P Resident Mgr. or Designee Distribution: SRC Coordinator, JAF Site Eng. MgrilP3 Technical Services Mgr (annual 50.59 report) RMS-JAF/IP3 NYPA FORM MCM-4, ATTACHMENT 4.3 Page 1 of 1
NEW YORK POWER AUTHORITY t.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
NUCLEAR SAFETY EVALUATION JAF-SE-96-042, REV. O I
'i 1.
Purpose:
This Safety Evaluation (SE) evaluates the initial connection of the Decay Heat Removal (DHR)
System to the Spent Fuel Pool (SFP), functional testing (heat removal capability) of the system i
during refueling outage 12 (RO12), and subsequent operation thereafter in various plant modes and configurations.
Issues related to the physical installation of the system and modification acceptance testing (without circulation to/from the SFP) are addressed in JAF-SE-96-039 (Reference 1) and are not repeated herein. Detailed discussions of the DHR design and key supporting analyses are contained in Reference 2 and are not repeated herein.
2.
Description of Procosed Activities As detailed in Reference 2, the DHR is designed as a non-safety related system, is physically independent of existing plant systems to the maximum extent, and is primarily intended to enhance decay heat removal capabilities during refueling outages with the ultimate goal of enhancing outage performance. The key points of the design and analyses of the DHR are i
l summarized below.
l
l.
(1) the decay heat removal capability of the system must equal or exceed the combined decay heat load of irradiated fuel in the SFP and the 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 l
l (3) the system must not adversely affect any safe shutdown function of existing plant systems.
Design basis consideration (1) above is satisfied by sizing the DHR to have a heat rejection capability of 30 X 10' BTU /HR at a wet bulb temperature of 73 F when operating in the l
nominal configuration. When operating in the nominal configuration, one of the two 100%
primary pumps will take suction from the SFP, pump through either of two plate and frame heat exchangers, and discharge back to the SFP. On the secondary side, one of two secondary pumps takes water from either of two pairs of single-cell standard industrial cooling tower basins, sends it through the primary heat exchangers and then bbck to the spray nozzles on the package cooling tower. The line is provided with a connection for emergency makeup to the towers by the fire water system. Figures 1 and 2, attached, depict the DHR primary and secondary loops. Figure 3, attached, provides the DHR system electrical one-line diagram. (Note: Figure nos 1,2, and 3 are not design drawings and are provided for
[
Information only).
j.
Design basis consideration (2) above is satisfied by the lack of any connections between the j
DHR and the existing plant electrical systems (other than commoning of the DHR trouble alarms into the existing FPCC trouble alarm), by limiting mechanical interfaces to fill and l
Page 1 of 17
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION JAF-SE-96-042, REV. 0 I
make-up provisions, and through provision of anti-siphon protection on the suction and discharge spargers in the SFP. While both the FPCC and the DHR take suction from and discharge to the SFP, the FPCC via overflow from the pool to the skimmer surge tanks, the two systems do not share any piping or supports. There are no connections between the DHR and the RHR systemi nor are there connections between the DHR and the FPCC.
Design basis consideration (3) above is satisfied through the physical independence of the DHR (from existing plant systems), the seismic supporting of the system, and the various ar d/ses described herein.
For scheculed refueling outages, DHR reliability will be enhanced through the use of a portable diesel gen trator which is 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 and woul'J be used in the event of loss of the normal 13.2 KV supply. Transfer from the 13.2 Kv source to the diesel generator would be accomplished manually.
3.
SAR Review:
3.A.
A change to the Updated Final Safety Analysis Report (UFSAR) la required.
The following UFSAR Sections were reviewed to perform' this SE, and to determine whether a change to the UFSAR is required:
t Section 1.2 Definitions Section 1.3 Methods of Technical Presentation Section 1.4 Classification of BWR Systems Criteria, and Requirements for Safety Evaluation Section 7.6 Refueling Interlocks l
Section 7.12 Process Radiation Monitoring System i
Section 7.13 Area Radiation Monitoring System l
Section 8.6 Emergency AC Power System l
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 l
Section 12.2 Classification of Structures and Equipment Section 12.3 Description of Principal 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 l
Overpressure; including Section 14.5.8, Event Resulting In A Core l_
Coolant Temperature _ increase (loss of RHR shutdown cooling) l 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 Criteria 4
Chapter 17 Quality Assurance Program: Appendix 17.28, Conformance with NRC 4
Regulatory Guides; and Appendix 17.2C, Plant Administrative Procedures General List 4
i Page 2 of 17
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION JAF-SE-96-042, REV. O 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. Also, the results of the thermal analyses which support installation and operation of the DHR system show DHR operation to be independent of fuel location.
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 4 assemblies per hour into the SFP, would not be applicable when the DHR was available. Said limitations would remain valid if the DHR were not available during a refueling outage.
A draft SAR revision will be prepared as part of Modification Package number F1-95-121,
" Decay Heat Removal System, Project CJ3110."
3.B The modification does not require a change to the Technical Specifications.
The following pertinent Technical Specification sections were reviewed to perform this SE.
j 3/4.0 Applicability (and Bases) i 3/4.2 Instrumentation (and Bases) 3/4.5 Core and Containment Cooling Systems (and Bases) 3/4.7 Containment Systems (and Bases) 3/4.9 Auxiliary Electrical Systems (and Bases) 3/4.10.C Spent Fuel Storage Pool Water Level (and Bases) 3/4.11 Additional Safety Related Plant Capabilities (and Bases) 6.0 Administrative Controls, including Sections 6.5, Review and Audit, and i
6.8, Procedures and Programs l
3.C.
The modification does require a change to the Design Basis Documents.
l l
The Design Basis Document for the Residual Heat Removal System,010 (Reference 9),
requires a discussion of the use of the DHR system, in conjunction with one train of the RHR system, or use of the DHR system by itself, as described in this SE.
4 Review and Analysis The following engineering and design issues have been considered in the DHR design:
(1) single failure considerations and overall heat removal capability (2) seismic design (3) system power supply (4) electrical separation (5) operational radiation fields (6) fire protection issues i
(7) wind loadings on exterior components (8) intemal (Reactor Building) flooding (9) heavy loads (10) decay heat load variations Details of the evaluations of those issues are contained in section 4 of reference 2 and modification package F195-121.
Page 3 of 17 l
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4 NEW YORK POWER AUTHORITY JAMES A.' FITZPATRICK NUCLEAR POWER PLAN 1 NUCLEAR SAFETY EVALUATION
.t l
JAF-SE-96-042, REV, O l
The following nuclear safety issues were evaluated in conjunction with DHR operation:
(1) maintenance of SFP water inventory, including the potential for drain-down of the reactor cavity during refueling (2) natural circulation cooling of fuel assemblies in the RPV
(
(3) potential for overcooling of the SFP Summaries of the evaluations performed for those issues follow in sections 4.1 through 4.3 below. Details of those evaluations are contained in section 5 of reference 2.
The large heat rejection capacity of the DHR, coupled with the ability of the system to remove decay heat from both the SFP and the RPV, obviate the need for RHR operation once the reactor cavity is flooded, the SFP gates removed, and the total (SFP plus RPV) heat load verified to be within the nominal DHR capability. As JAF Technical Specifications do not i
require RHR operability when the reactor cavity is flooded and the SFP gates are removed, i
there is literally no need for Technical Specification revision to allow removal of RHR from service under the specified conditions. However, the Standard Technical Specifications (STS),
reference 3, would require one " train" of RHR SDC remain operable as long as irradiated fuel was in the RPV. In consideration of the STS requirements, this SE will conservatively treat the j
issue of removal of (all) RHR from service under specified conditions as a nuclear safety j
consideration. Refer to section 4.4 below for that evaluation.
4.1 Maintenance of SFP Waterinventorv i
Like the FPCC system, the DHR is not designed to remain functional following a postulated passive failure. However, maintenance of SFP water level following an earthquake or breach of either the FPCC or DHR primary side pressure boundaries is a safety-related function for decay heat removal of spent fuel in the pool. The DHR design considored two potential mechanisms for SFP inventory reduction; primary to secondary leakage and catastrophic failure of the DHR pressure boundary and subsequent SFP drawdown. Specific features have been incorporated into the design to preclude SFP level reduction to unacceptable levels.
The DHR system suction and discharge spargers include holes to preclude siphoning of the SFP in the event of a breech in DHR primary loop pressure boundary. The holes are functionally equivalent to the vacuum breakers provided on the FPCC system and have the added benefit of having no moving parts, Plate and frame type heat exchangers were deliberately selected for the DHR to preclude i
primary-to-secondary leakage. In addition, the secondary side will be maintained at a minimum of 10 psi greater than the primary side operating pressure. The design also incorporates circuitry to automatically trip an operating primary side pump if the sensed secondary to-primary differential pressure were to fall below approximately 10 psid These features of the DHR system are consistent with the guidance contained in NRC IE Bullettin 80-10, reference 4. Based on the design characteristics of the plate and frame heat exchangers and the secondary-to-primary differential pressure considerations, leakage from the primary side into the secondary side is not considered to be a credible event. Therefore, there is no i
need for the installation of a radiation monitor in the DHR secondary loop.
i i
The potential for loss of refueling cav;ty inventory due to postulated failure of the reactor cavity seal has been previously evaluated for JAF as part of our response to INPO Significant
. Operating Experience Report (SOER) 85-01 (reference 5). Installation and operation of the
}
Page 4 of 17
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NEW YORK POiNER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION JAF-SE-96-042 REV. O DHR has no bearing on the SOER evaluation in terms of the design of the cavity seal and the potential for catastrophic seal 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.
4.2 Natural Circulation Coolina of Fuel Assemblies in the RPV As detailed in section 5.2 of reference 2, the DHR has been shown to be capable of removing decay heat from both the RPV and the SFP so long as the reactor cavity is flooded and the SFP gates removed. The ability to remove RPV decay heat under those circumstances has been verified by JAF-specific analysis and previous testing at the Hatch Plant. To ensure relevance of the Hatch data to JAF, a comparison was of the JAF dimensions and geometries to those of Hatch was performed.
The thermal-hydraulic calculations summarized in reference 2 confirm the operation of the DHR, in either the nominal or maximum heat removal configuration, would effectively remove decay heat from irradiated fuel in the RPV. Based on the work detailed in reference 2, the
" operating scenario" can be summarized as follows. With the SFP gates removed, DHR is initiated and RHR SDC and FPCC are terminated. As the DHR is removing heat from the SFP, the average core exit water temperature would increase due to the loss of forced flow through the core. The analyses predict a core exit temperature increase on the order of 15-20 F over approximately a one hour period. During that time, the water volumes in the RPV, the reactor cavity, and the SFP would be approaching a quasi-steady state equilibrium condition.
The (direct) cooling of the SFP water with the DHR would result in the establishment of convective currents from the RPV to the SFP. After approximately one hour, the convective currents would be fully established. The thermal-hydraulic analyses predict a rapid exponential decrease in core exit temperature once the convective currents become fully established.
From that point forward, the combined RPV/SFP/ reactor cavity water volumes are for all intents one homogeneous volume.
General Electric has evaluated the heat transport from the core (RPV) to the SFP for Plant Hatch, Hope Creek, and J.A. FitzPatrick (reference 6). A review of the geometry 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.
4.3 Potential for Overcoolina of the SFP The maximum heat rejection capability of the DHR is vastly greater than the decay heat load of the SFP prior to core offload (approximately 45 X 10' BTU /HR at the design condition vice 8
approximately 2 X 10 BTU /HR). Even after the core u offloaded into the pool, the heat removal capability of the DHR in the nominal configuration will be greater than the decay load (approximately 30 X 10' BTU /HR at the design condition vice approximately 27 X 10' BTU /HR, assuming core offload is completed within 5.5 days post-shutdown). The actual heat removal capability of the DHR will be a function of the outside air temperature and the further the actual air temperature is from the design wet bulb temperature of 73 F the greater the difference between the actual heat removal capability and the design value. As the DHR is a l
l Page 5 of 17
NEW YORK POWER AUTHORIT(
JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION l
l-JAF-SE-96-042 REV, O l
f manually operated system, the potential does exist for excessivo cooling of the SFP, The lower temperature limit is a plant safety limit derived from criticality analyses of spent fuel stored in the specific rack designs.
By design, temperature control during DHR operation is primarily accomplished through control of the cooling tower fans. Fan speed and status (on or off) are automatically controlled based on the temperature of the water on the cold side of the tower basins. In any case, any potential cool down transient could be mitigated by simply shutting off any operating primary side pump (s). Therefore, operation of the DHR for SFP cooling only is acceptable, provided the appropriate administrative controls and procedures are in place and the operators trained in their use.
i 4.4 Removal of fall) RHR From Service Under Specified Conditions The relative decay heat removal capabilities of existing systems and the DHR are summarized in the table below (Note: The table is extracted from reference 2).
TABLE 4-2 l
DECAY HEAT REMOVAL CAPABILITIES l
Method of Decay Heat Removal 10' BTUlhr l
DHP System (Maximum) 45 DHR System (Nominal) 30 RHR Assist + FPC (1 pump, i HX) 24 i
FPC (2 pump,2 HXs) 10 FPC (1 pump,2 HX) 6.3 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 FPCC system, as a function of time after shutdown.
The existing JAF Technical Specifications do not require RHR SDC operability when the reactor cavity ic flooded. However, the STS would require one train of SDC tn be operable as long as irradiated fuel was in the RPV. JAF Technical Specifications do not mention the DHR. However, the " REQUIRED ACTION" statement for STS 3.9.8 states: " Verify an alt 6 mate 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 l
flange provides adequate 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 an outage.
l Based on figure 4, the total decay heat load (SFP and RPV) would decrease below 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 i
flooded and the gates removed, the DHR would be single (active) failure proof, with operator action, for the removal of all decay heat regardless of the status cf refueling activities. The unavailability of RHR SDC at that point in time in an outage, would under the most conservative interpretation be tantamount to being in the " ACTION" statement of the STS. In l
reality, the DHR availability at that time would provide redundant means of RPV (and SFP)
Page 6 of 17 l
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION j
i JAF-SE-96-042, REV. O t
! (-
decay heat removal.. It is more reasonable to consider the RHR SDC and the DHR (nominal configuration) to be redundant means of decay heat removal when flooded up with the SFP j
gates removed. By administratively requiring SDC to remain operable until the decay heat i
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.
Thus, during the initial part of an outage RHR SDC and DHR would be considered redundant methods of decay heat removal and both shall be operable. 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_th.rga methods of decay r
heat removal. Thus, removal of RHR SDC from service under those conditions would be acceptable as the decay heat removal function would remain single (active) failure proof. It should be noted the detailed analyses discussed in reference 2, verify the ability to ' monitor
' RPV temperature * (in the STS sense of the phrase) using temperature instrumentation on the
(
primary side inlet to the DHR heat exchanger, Additional measures to maintain key spare DHR components on site would provide an even greater measure of security, further enhance the " reliability" of the decay heat removal function, and reduce the potential for loss of decay heat removal events.
Therefore, this SE requires at least one SDC train be OPERABLE until such time as the nominal heat removal capability alignment of the DHR system has been shown to be capable of removing the combined RPV and SFP decay heat loads. Prior to removal of SDC from service, the availability of redundant DHR components, including the backup portable power supply, shall be verified. The DHR system, in its nominal heat removal configuration, is therefore capable of being used as a substitute for the RHR system or as a substitute for the FPCC system.
~
Future maintenance' of an RHR SDC train in the standby (operable or available) mode prior to l
use of the DHR system will be determined by prudent outage risk management practices, per reference nos. 7 and 8, recognizing there are physical tasks which have to be performed which j
would render SDC unavailable.
j 4.5 Initial Thermal Performance Testina of the DHR System Decay Heat Removal Capability Modification acceptance testing of the DHR will be performed prior to RO12 as evaluated in JAF-SE-96-039. 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. Therefore, thermal performance testing of the system will be performed 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 verified by operating the system and verifying a negative trend in DHR primary side suction temperature. The combined SFP and RPV heat loads at that point are expected to be within the capability of the DHR operating in the nominal heat removal configuration. Whether DHR is in the nominal or maximum heat removal configuration j
at the time SDC is secured has no bearing on this evaluation.
Detailed engineering data obtained during that (and subsequent periods) will be used to evaluate DHR thermal performance. However, verification of the DHR's ability to handle the i
l then existing decay heat load will be determined simply by observing the system's ability to Page 7 of 17
NEW YORK POWER AUTHORITY l
JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
NUCLEAR SAFETY EVALUATION l
l JAF-SE-96-042, REV. O i I maintain or decrease pool water temperature when RHR and FPCC are not in service, 1
4.6 Effect of DHR Operation on the Probability of Transients or Accidents Applicable to This IE The following anticipated operational transients or postulated design basis accidents previously evaluated in the SAR are considered applicable to this evaluation:
1.
a loss of offsite power during the use of the DHR system, including a loss of the DHR power supply, or 2.
a fuel handling accident (drop of a fuel assembly or heavy load) during the use of the DHR system, or 3.
occurrence of a natural phenomenon, such as a tomado, during the use of the DHR system which results in loss of DHR or with the DHR idle, or 4.
a seismic event, which potentially could result in complete loss of reactor vessel and SFP cooling by the DHR system (with potential SFP boiling) during the use of the DHR system, or 5.
The effect of DHR operation on the probability and consequences of the above transients and accidents is discussed below. Detailed comparisons of the pre-and post-DHR configurations were reviewed in section 4.1 of reference 2 for conditions of varying total decay heat load.
4.6.1.
Loss of Offsite Power: Probabilltv and Consecuences As discussed above, the DHR system electrical power supplies and distribution system are independent of existing plant systems. Thus, a failure of a DHR electncal system or component cannot affect safety-related electncal equipment or components. Moreover, should a loss of offsite power occur while the DHR system is operating during a refueling outage, the provision of a backup diesel generator to power the DHR system components ensures that decay heat removal will resume in the brief amount of time to repower the DHR system via the backup diesel generator, This is an improvement over the present design of the FPCC, which would lose power when offsite power is lost and use the RHR assist mode if SFP cooling is required to be restored before offsite power is restored. If a loss of offsite power were to occur during plant operation when the DHR system is being used (for example, as a substitute for the FPCC) and no portable diesel generator is provided, the consequences are unchanged from this event occurring without the DHR system in place. That is, if a loss of offsite power occurs and SFP cooling is lost, in order to provide SFP cooling the plant would have to shut down and the RHR system would have to be used in the fuel pool assist mode. This scenario l~
would also apply if the DHR system is being used in place of the FPC without the backup diesel generator. Therefore, the provision of the DHR system, and the operation of the DHR system, does not increase the consequences of a loss of offsite power (or the loss of offsite power to the DHR system), but instead improves the restoration of SFP cooling during refueling outages.
Page 8 of 17
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFET( EVALUATION l
JAF-SE-96-042, REV, O i
4.6.2 Fuel Handlina Accident: Probability and Consecuences The DHR system plays no role in preventing or mitigating a fuel handling accident since it is a system for recirculating the SFP water and removing decay heat from the SFP water. As noted in reference 1, during the installation and removal of the DHR system pipe spool pieces, which provide SFP suction and discharge for the DHR system, the load path will conform with existing plant procedures goveming load handling over or around the SFP, Thus, the installation or removal of the DHR system pipe spools is not postulated to cause a fuel handling accident. Should a fuel handling accident occur by some other means during the operation of the DHR system, the consequences therefrom are not changed from those previously analyzed. Therefore, the provision of the DHR system, and the operation of the l
DHR system, does not increase the consequences of a fuel handling accident.
4.6.3 Natural Phenomena: Probability and Consecuences l
The primary side components (pumps, strainer, heat exchangers and a control panel) of the DHR system are skid-mounted and located on Elevation 326'-9" inside the seismic QA l
Category l Reactor Building. The secondary side components (pumps and cooling towers) are located on the roof of the Railroad and Truck Bay and Standby Gas Treatment Building at El.
l 293'-0," and have been evaluated for 90-mph wind loads. The primary side components are therefore protected from natural phenomena by virtue of their location inside the Reactor Building. The secondary side components are mounted on the roof, well above any flood level, with the Reactor Building wall affording protection from heavy winds or a potential tomado-bome missile on the north side of the secondary side components. Since these components are able to withstand a 90-mph wind load, and this wind speed has a recurrence interval of once per 100 years, the secondary side components are adjudged to be adequately protected from heavy winds.
i A missile shield is located on the exterior face of the Reactor Building over the piping penetrations from the secondary side into the Reactor Building. The shield is not required by plant design and licensing criteria but is provided as an additional conservatism in the design.
These penetrations are therefore adequately protected from a tomado missile. Should a tomado strike the plant vicinity, a tomado missile could, however, conceivably impact a component of the secondary side which is exposed on the roof. Any single failure of a secondary side mechanical component is mitigated by the provision of its redundant component when the DHR system is operating in its nominal configuration. A failure of the offsite power supply is mitigated by the provision of backup power from a truck-mounted diesel during refueling outages. ' Therefore, adequate protection from natural phenomena, specifically a tomado missile, is provided for the DHR system.
The probability that a tomado would occur, times the probability of its strike during the relatively short time period that the DHR system is in use, times the probability that this event would occur during the even shorter time when the RHR system might be out of service, are considered sufficiently remote such that no specific design protective features are required for this postulated event. Moreover, in the unlikely event that the secondary side is affected such that the decay heat removal function is lost entirely, those consequences are no different than the total loss of decay heat removal regardless of the heat removal system. Although not j
previously analyzed, these consequences have been analyzed in support of the installation and use of the DHR system, and are demonstrated to be of no safety concem to public health j
and safety. The results of those analyses, presented in sections 9.2 and 9.3 of reference 2, j
show that a sustained total loss of decay heat removal capability would eventually result in j
l i
Page 9 of 17 j
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION JAF-SE-96-042, REV. O boiling in the combined SFP/ reactor cavity / RPV water volumes in approximately eight hours i
(worst case). Conservative estimations of the dose consequences of sustained boiling indicate the worst case control room dose would be only 15% of the applicable limit. All of the other calculated doses are much lower percentages of their associated regulatory limits.
4.6.4 Seismic Event : Probability and Conseauences The SFP is designed as a QA Category I structure and can withstand an earthquake while E
maintaining the minimum required water level above the spent fuel. Neither the FPCC nor the DHR system are designed as QA Category I systems. Therefore, a postulated seismic event could conceivably lead to a loss of SFP cooling since credit for the operation of these non-seismic heat removal systems cannot be assumed. The consequences of a seismically-induced loss of SFP cooling would be identical to consequences discussed in section 4.6.3 above, and detailed in reference 2.
5.
Action items to be Tracked:
1.
Plant Operating Procedure (s) shall be revised, or new procedures generated, to include use of the DHR system 2.
AP-10.02, " Outage Risk Assessment", and ODSO-32, " Shutdown Procedure" shall be revised to reflect use of DHR sveem and shall explicitely require SDC remain operable until the total decay heat load is within the capacity of the DHR operating in the nominal configuration.
3.
A UFSAR change package is included in Modification package F1-95-121 4.
Plant Operating Procedure (s) and the UFSAR shall be revised to specifically state: (1) the current restrictions on fuel movement (96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> delay in fuel movement and 4 assembly per hour maximum fuel transfer rate) are not applicable when the DHR is available, and (2), the current restrictions on fuel movement remain applicable when the DHR is not available.
6.
10CFR50.59(bM2) Summary of Ac:!vity and Nuclear SE:
A Decay Heat Removal (DHR) system is provided which is completely independent of the FPC and RHR systems and is powered independently of the in-plant electric distribution system.
The DHR is a non-safety related system with a design maximum heat removal capacity of 45 X 10' BTU /HR at a wet bulb temperature of 73 F. The DHR system has a normal heat remova!
configuration which providas two loops, either of which can remove 30 X 10' BTU /HR at a wet bulb temperature of 73 F. The DHR system is designed to supplement or to substitute for either the RHR system and/or the FPC system to provide additional operating flexibility. The DHR system can remove decay heat from fuel elements located in the SFP with the fuel pool gates in place. The DHR system can also remove decay heat from fuel elements in the reactor pressure vessel (RPV) when the refueling canal is flooded and the SFP gates are removed.
The discussions and evaluations provided above,in conjunction with the specific responses 6
below, provide the bases for the conclusion that the implementation and operation of the DHR system does not constitute an unreviewed safety question.
Page 10 of 17
I NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION JAF-SE-96-042, REV. 0 -
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' 6.A.1 The proposed activity does not increase the probability of an accident evaluated previously in the Safety Analysis Report.
The DHR system is independent of any accident initiators and/or accident mitigation systems, and can not increase the probability of an accident previously evaluated in the SAR.
Installation and operation of the DHR under the conditions and configurations detailed above provide an additionel and independent means of removing decay heat from the SFP or th>
combined RPV and SFP. The addition of the DHR reduces both the likelihood of elevated pool i
or cavity water temperature in the event of a single active failure and the peak resultant water 1
temperature.
The physical design of the DHR and the high degree of its independence from existing plant systems preclude the operation of the system from either increasing the probability of, or increasing the consequences of, any accident or malfunction discussed in the UFSAR.
6.A.2 The proposed activity does not increaae the consequences of an accident evaluated previously in the Safety Analysis Report.
The proposed mooift.pn does not increase the consequences of an accident previously analyzed in the SAR, since the DHR system is provided for an attemate decay heat removal system and is not used for any accident analysis decay heat removal (nor does it affect any existing system used for decay heat removal after an accident). Although not required by the plant licensing basis, an analysis of the dose consequences of protracted SFP boiling was performed.
That analysis is not directly specific to this modification as it conservatively considers a loss of all SFP decay heat removal regardless of the system (s) used. The results of the radiological assessment, presented above, indicate that the loss of all decay heat removal (no matter what source of decay heat removal is used), when the refuel cavity is flooded up and the fuel pool gates are removed, presents no threat to the public health and safety.
6.A.3 The proposed activity does not increase the probability of occurrence of a malfunction of equipment to safety evaluated previously in the Safety Analysis Report. BASIS:
The DHR s/Mam is mechanically and electrically independent of plant safety systems. Hence, postulated failures or malfunctions previously evaluated with respect to decay heat removal are not affected by the installation or use of the DHR system. The DHR system has no adverse effect on installed systems important to safety and cannot increase the probability of their malfunction.
6.A.4 The proposed activity does not increast me consequence of a malfunction of equipment importam to safetv *,aluated previously in the Safety Analysis Report.
The DHR system is mechanically and electrically independent of plant safety systems and equipment important to safety. Hence,the consequences from malfunctions previously evaluated are not adversely affected by the installation or use of the DHR system.
6.A.5 This change does not create the possileility of an accident of a different type from any j
previously evaluated in the SAR.
Operation of the DHR does not create the possibility of an accident not previously analyzed.
Page 11 of 17
NEW YORK POWER AUTHORITY l'
JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION JAF-SE-96-042, REV. O
!I While operation of the DHR in lieu of RHR SDC is permitted by the SE, under specified heat load conditions, the accident of interest, namely loss of decay heat removal, remains unchanged.
6.A.6 This change does not create the possibility of a malfunction of a different type from any j
previously evaluated in the SAR. BASIS:
The DHR system does not create the possibility of a malfunction of a different type, but rather presents an extra layer to the " defense in depth" concept of plant safety, and is less vulnerable to malfunctions than the existing FPC and RHR systems. The DHR system provides an improved, safer, more reliable means of achieving decay heat removal in the early stages of a refueling outage, when decay heat removal is most crucial.
6.A.7 This change does not reduce the margin of safety as defined in the basis for any Technical Specification. BASIS:
The plant Technical Specification related to the SFP is Technical Specifiestion 3/4.10.C, which states:
"Whenever irradiated fuel is stored in the spent fuel storage pool the pool water level shall be maintained at a minimum level of 33 feet."
Technical Specification 3/4.10.C !s r,ot changed or altered by this modification.
The provision of anti-siphori protection in the DHR design ensure the bases of Technical Specification 3.10.C is not adversely affected by DHR operation.
There are no other Technical Specifications or Bases that are affected by this modification.
Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specification.
Provision of the DHR system arfds another layer of defense in depth to SFP/RPV cooling capability which was previously not present. Since the Technical Specifications are written without consideration of an extra means of SFP/RPV cooling, now available with additional safety benefits from the DHR system, it is obvious that the margin of safety has been increased, rather than reduced.
6.A.8 This modification does not involve an unreviewed safety question based on statements and explanations 6.A.1 through 6.A.7.
6.A.9 This modification does not degrade the Security Plan, the Quality Assurance Program, or the Fire Protection system. BASIS:
The DHR system is entirely within the protected area and therefore no adverse impacts to the existing Security Plan are created by this modification. The installation, testing, and utilization of the system are performed under procedures and processes which conform to the QA l
l Program. Existing administrative controls and procedures ensure Modification package F1-l 95-121 will be reviewed by Fire Protection personnel.
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Page 12 of 17 l
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR SAFETY EVALUATION l
JAF-SE-96-042, REV. O H-7.
References:
1.
NYPA Nuclear Safety Evaluation JAF-SE-96-039, " Installation and Modification Pre-l Operational Testing of the Decay Heat Removal System" 2.'
NYPA Report JAF-RPT-DHR-02413, " Evaluation of the Decay Heat Removal System",
Revision 1, July 1996 3.
NUREG 1433, " Standard Technical Specification 'or General Electric Plants, BWR/4",
l USNRC, April 1995 l
4.
IE Bulletin 80-10," Contamination of Non-radioactive System and Resulting Potential for Unmonitored and Uncontrolled Release to the Environment", USNRC 5.
Significant Operating Experience Report 85-1," Reactor Cavity Seal Failure", January l.
1985 (and NYPA's evaluation thereof) 6.
General Electric Nuclear Energy letter W. H. Brown to D. Lindsey (NYPA), "ADHR Questions", dated November 15,1995 7.
Administrative Procedure AP-10.09," Outage Risk Assessment", Revision 3 l
8.
Report 91-06," Guidelines for Industry Actions to Assess Shutdown Management",
Nuclear Management and Reso fee Council, Inc., December 1991 9.
NYPA Design Basis Document for the Residual Heat Removal System I
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