ML20045B407
| ML20045B407 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/19/1985 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20045B367 | List: |
| References | |
| NUDOCS 9306170314 | |
| Download: ML20045B407 (25) | |
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V . SUBTASK REPORT l t FOR POOL COOLING AND PURIFICATION REQUIREMENTS- i r DURING FUEL CONSOLIDATION Revision 02 February 19, 1985 [ t i
" DISCLOSURE" 1 i
This report has not been reviewed to determine whether it contains patentable subject matter, nor has the accuracy of its information or f conclusions been evaluated. Accordingly, the report is not to be i considered a published report and is not available for general distribution, and its distribution is limited to employees and l advisors of Northeast Utilities Service Company and EPRI for the sole purpose of evaluating the progress and future course of the project described in the repo'rt. Until the report. has been reviewed and evaluated by Northeast Utilities Service Company and EPRI, it.should be neitner disclosed to others nor reproduced, wholly or partially, without written consent of Northeast Utilities Service Company and ;
\ EPRI.
' i 9306170314 930610 j PDR ADOCK 05000336 P PDR f{ j
vt { -- TABLE OF CONTENTS l t i
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Page
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1.0 Objective 3 ,-
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'I 4 e 2.0 Parameters / Considerations i
Analyses and Evaluations 6 3.0 i 9 i 4.0 Discussion of Results e 12 5.0 Conclusions and Recommendations l 1 I t i Appendix A Spent Fuel Decay Energy Calculations ? i Appendix B Spent Fuel Total Decay Energy Curves : Cask Laydown Area Heat Input ; Appendix C 3 e ! i t , e
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Section 1
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{ OBJECTIVE ,, f! (I The purpose of this study was to review incremental cooling requirements in the ;j i cask laydown pool during the fuel consolidation operation. Also, the existing spent fuel pool cooling system was to be evaluated to determine its capability to provide sufficient cooling in relation to the approximate quadrupling of the ! spent fuel rod inventory toward the end of plant life. Additionally, the } l possibility of crud escaping the fuel consolidation area and entering the spent fuel pool was to be examined. If necessary, recommendations were to be made for equipment and operational modifications. Radiological considerations for the case of fuel bundle crud that may be freed in the pool by handling and packaging of the waste hardware were evaluated separately as part of the ALARA report. This report was forwarded to NUSCO via This study provided information on the crud l MP-2-84-122, dated May 22, 1984 filtration system and amount of crud generated and captured. . i 9 h I l t 4 1 I
Section 2 . I PARAMETERS / CONSIDERATIONS
' il 11 a
r In this study there were several parameters and considerations that impacted 4J the pool cooling and purification system during the fuel consolidation ef fort. ! These included the decay energy of spent fuel as a function of time, the spent fuel pool fuel storage capacity, the cooling capability of the spent fuel t system, the size of the fuel consolidation work area (cask laydown ar'ea), the maximum number of fuel rods held in the consolidation work platform, and the excess heat generated by the fuel consolidation machinery and anci.llary equi pment. Due to site specific differences in design and equipment, none of ; tne above parameters can be considered generic to other plants. - , For the Millstone Nuclear Power Station Unit 2, fuel bundles will not be consolidated until they have decayed for a minimum of five years. The spent ij fuel pool is composed of two regions: region one has 384 storage slots, and - region two has 962 storage slots. For this analysis, it will be assumed that all of region one can be utilized with fuel bundles af any age and decay . status. Region two, however, can be 100% utilized only if consolidated fuel bundles are stored there. Otherwise, region two has a 75% utilization factor. i Based on these limitations, the following scenario for utilization of storage i slots is assumed: , h l I i l l , (- , t
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Type Fuel No. Fuel Region Storage Slots Remarks Bundles Required , Full Core 217 1 217 167 Storage Slots , l Offload remain in Region 1 I Unconsolidated 362 1 167 195 Unconsolidated i Fuel (5 yr.) bundles must go into Region 2 (195/0.75 = > l 260 slots required) l, 2 260 702 storage slots remain _ available in Region 2 , Consolidated 1375 2 688 14 storage slots remain ; Fuel (19 yr.) , For conservatism, the 14 remaining storage slots will be assumed to have 14 bad fuel assemblies that had been in the core for two years and have decayed for ; I one year.- The reason for choosing two years in the core is that if three years had been assumed, the assemblies would have been among the one-third core replaced during refueling. i n
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Section 3 i i - ANALYSES AND EVALUATIONS 3.1 Spent Fuel Pool Cooling System (SFPCS) l Since the new racks and fuel consolidation process essentially quadruples the capacity of the spent fuel pool, an evaluation was performed to determine if the SFPCS was adequate for its design bases. From the Millstone Nuclear Power Station Unit 2 FSAR, the first design basis is for the normal operating condition, and it predicates an operating schedule in , which the unit is on an equilibrium refueling period with approximately one-third of a core being removed from the unit at the end of each period..'The removed assemblies have been in the reactor for three cycles at the time of , 1 discharge, and the heat load is based on removing decay heat from 1064 occupied storage slots (1281 useable storage slots less 217 slots for potential core off load). r i t For normal operation, the NRC Regulatory Guide 1.13, proposed Revision 2, requires the cooling system to be capable of maintaining a bulk pool water temperature of 140 F or less with one cooling pump out of commision (single ; dCtiVe f ailure). I The second design basis for tne system (abnormal operation) considers the ; possible necessity of a total core off lodd when spent fuel is already , residing in the spent fuel pool. Worst-case analysis assumes all 1281 l f useable potential storage slots are utilized - 217 slots for core off load, ' 376 storage slots for unconsolidated fuel-that has decayed less than five years (362 assemblies from normal refueling plus 14 damaged fuel . assemblies), and the remaining 688 slots filled with consolidated fuel ! assemblies. l j For abnormal operation, one shutdown. cooling system heat exchanger and one ~ low pressure safety injection pump are utilized to supplement the SFPCS. L
i
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NRC Regulatory Guide 1.13 proposed Revision 2, requires the bulk pool water : temperature to be maintained below boiling. No allowance need be made for .: t l a single active f ailure. 3.2 Incremental Cooling Requirements in Consolidation Work Area { The consolidation work is to be done in the cask laydown pit, which is 9' x , 9' x 40'. The consolidation work area under consideration can accommodate up to ten fuel bundles. Assuming that the fuel bundles to be consolidated have decayed for five years, the decay heat generated by ten fuel bundles , would be approximately 30,000 BTU /hr. Also, taking into account the heat generated by the filter system, pool lighting and other miscellaneous equipment, the total heat being generated in the consolidation work area will be approximately 145,000 BTU /nr (See Appendix C). Tnis amount of heat would cause the temperature 'in the consolidation work pool to rise - ; approximately 18"F per 24 hours. t 3.3 Spent Fuel Consolidation Filtration The cask laydown drain pit pump (P-151) should not be used during fuel consolidation operations. Tnis will preclude the introduction of fuel bundle crud into the spent fuel pool. By isolating the cask laydown area, any potential crud release will be confined to a small area. C-E's Engineering Development and Services Department has determined that the expected crud release from each bundle during consolidation could be as hign as 1.6 pounds (730 gr) or 26.5 cubic inches. The major constituents of the crud are expected to be oxides of nickel and iron. l C-E's Design Specification for Spent Fuel Consolidation Vacuum / Filtration } System for Millstone Point Unit 2 (00000-ESE-538) requires two filtration f t skids designed to operate at the bottom of the cask laydown area. Each skid shall consist of one pump with a strainer upstream and two filters in series downstream. The strainer will take out particles as small as ten microns. The filters will be placed in series, providing a two stage j i filtration system. The upstream filter rating shall be five microns ,
I absolute, and the second filter shall_be two microns absolute. The pump selection shall be made to provide a minimum flow of 250 gpm through the + ( filters. 6 f 1; f s 1 8 i 1
1 Section 4 i DISCUSSION OF RESULTS 4.1 Spent Fuel Pool Cooling System - l Since only spent fuel assemblies thet have decayed for at least five years e wil be consolidated, 362 storage locations will be occupied by unconsolidated assemblies. The worst case analysis for the normal operation design basis considered the decay heat generated by 688 consolidated bundles of spent fuel and 376 unconsolidated bundles (including 14 damaged fuel bundles and a recent one-third core due to refueling). Data were obtained from decay energy curves supplied by I C-E's Physics and Computer Analysis Department (See Appendices A.and B). , The analysis showed that the spent fuel heat load would not be less than ; the SFPCS cooling capacity of II.S x 106 BTV/hr until the most recent one-
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third core discharge has decayed for at least nineteen days. Upon initial loading of the one-third core into the spent fuel pool, the maximum heat , load would be 15.2 x 10 6BTU /hr (assuming fuel was not moved for 150 , hours per NRC Standard Review Plan for FSAR Section 9.1.3). Thus, for dbout fourteen days after unloading fuel from the reactor, the Capacity of ; the SFPCS would be insufficient to meet the requirements of NRC Regulatory ; Guide 1.13, proposed Revision 2, for normal operation. j i The worst case analysis for the abnormal operation Dasis added the decay heat from a full core off-load to that from the maximum inventory of spent fuel stored in the pool. In accordance witn NRC SRP 9.1.3, tne full core I was assumed to have decayed for 150 hours before removal from the reactor, l and the most recently removed one-third core.of spent fuel was assumed to have decayed for 36 days. The resulting maximum decay heat load was 37.8 x 106 BTU /hr. The combined cooling capacity of the SFPCS and one shutdown ! cooling heat exchanger supplied by a low pressure safety injection pump is 38.5 x 106BTU /hr, which is more than adequate to meet the requirements of Regulatory Guide 1.13, proposed Revision 2, for abnormal operation.
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l 1 Since only fuel assemblies that have decayed for at least five years will be consolidated, the decay heat loads from consolidated bundles will be 4 relatively low. For example, the decay heat of a consolidated bundle of' five years decayed fuel will be approximately equal to the decay heat of an unconsolidated bundle of three years decayed fuel, approximately 950 ? watts. By contrast, an unconsolidated fuel bundle that has decayed for only ten days nas a decay heat of approximately 32,000 watts, for one month j
- 21,000 watts, and for one year -_4,700 watts. lh 4
j 4.2_ Cask Laydown Pit Cooling Requirements b li As noted in Section 3.2, the heat generation rate in the cask laydown pit ] during consolidation will be approximately 145,000 BTU /hr (42.5 KW). The q easiest way to remove this heat would be to use the cask laydown pit drain pump (P-151) to pump a portion of the consolidation area water inventory t into the spent fuel pool. Makeup into the cask laydown area would be from a spent fuel pool heat exchanger return line tnrough valve RW-350 as shown l on Bechtel drawing number 25203-26023, Piping and Instrument Diaram, Spent 1 l Fuel Pool Cooling and Cleanup System for Millstone Nuclear Power Station Unit 2. This method would, however, allow any contamination that escapes the consolidation filtration system to be introdJCed into the spent fuel pool. Another metnod of heat removal would be by using a heat exchanger. However, due to the small heat generation rate, a length of pipe with cooling water circulating through it would perform adequately as a heat l exchanger. A cooling loop is more advantageous tnan a heat exchanger since it is mechanically simpler and less costly. f i f 4.3 Spent Fuel Consolidation Filtration .; r I As noted in Section 3.3, approximately 26.5 cubic inches of crud is j expected to be released per fuel bundle.. It is targeted to use one filter f for each two to four fuel bundles processed. ( l l f
m n The filtration industry at large markets a nuclear compatible pleated t cartridge filter made of microfiberglass. The filter media is constructed , g
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so that a graduated porosity exists. For example, the graduation could be l i from 20 microns upstream to two microns downstream. lj 4: II Therefore, based on a cylinder six inches in diameter and 30 inches high, a l
-i cartridge volume of 850 cubic inches is obtained. Assuming that the ;
maximum solid volume for a spent filter is 25% of tnis volume, the filter l usage f actor for this case would be approxinately eight fuel bundles cer l filter. However, tnis filter usage factor is only an approximation of what ; can be expected and will probably be less. i e i
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Section 5 q CONCLUSIONS / RECOMMENDATIONS 5.1 Spent Fuel Pool Cooling System f;
;i The cooling capacity of the existing SFPCS, supolemented by one shutdown .
cooling heat exchanger and one low pressure safety injection punp is I adequate for the abnormal operating condition, in which the full core has to be off-loaded into the spent fuel storage pool containing the maximum inventory of spent fuel. No changes are required for this design basis. : I,i I' As indicated in Section 4.1, the cooling capacity of the existing SFPCS would not meet proposed regulatory requirements for the normal operating condition design basis for about fourteen days following the transfer of the one-third replacement core into the spent fuel storage pool. The ' addition of one heat exchanger and two cooling pumps to the SFPCS would rectify this deficiency. i - However, before committing to this change, wnich may be expensive, difficult to accommodate, and would have to be accomplished without interrupting fuel pool cooling, it is recorrnended thet alternative courses , of action be investigated. The following examples are suggested, but , except as noted, it has been beyond the scope of this task to evaluate their impact or probability of success.
- 1. Preliminary scoping calculations indicate that the two existing heat {
- exchangers operating in parallel and supplied by one existing cooling l pump (i.e., one active f ailure) may be able to maintain the bulk pool water temperature at 140*F during periods when the reactor building closed cooling water system inlet temperature to the spent fuel pool f cooling system heat exchangers can be reduced to 75 F instead.of tne 85 F temperature assumed for design. If both cooling pumps are operating during these periods, the bulk pool cooling water may be able to be maintained at about 125 F. More detailed-analysis of.tne j i
1 f
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-~c , - . _ - , . off-design performance of the heat exchangers and cooling pumps is necessary to verify this.
- 2. Monitoring of the bulk pool water temperature combined witn pacing the ;
rate of removal from the reactor of the spent fuel assemblies being replaced, would take advantage of the fairly rapid decline in decay i heat production during the first few days after reactor shutdown. It f might be feasible to rearrange the sequencing of work to be performed in the refueling shutdown so as to avoid lengthening the critical #h schedule. This approach might also include a Technical Specification modification to require that, in the event of a spent fuel pool j, cooling system pump f ailure, fuel transf er from the reactor to the : storage pool be stopped until the bulk pool water temperature stabilizes at or below 140 F.
- 3. The use of supplementary cooling water from the shutdown cooling system during normal operation in a refueling shutdown would reduce ,
the heat load. The impact of this on the present design bases for the snutdown cooling system would have to be evaluated since there would be at least two-thirds of a core of irradiated fuel in the-reactor vessel. Use of both cooling systems probaoly complicates the determination of the most significant single active failure in normal ; operation. A possible adverse effect would be the spread of contamination from the reactor vessel to the spent fuel storage pool. Some equipment modification would probably be required since Ine two systems are presntly isolated from each other by blind flanges in the interconnecting piping. 5.2 Consolidation Work Area ; The actual consolidation of fuel bundles will take place in the cask laydown pit. Heat generation in this area will be a maximum of 145,000 BTU /hr. As noted in section 4.2, this heat could be removed utilizing i existing piping. However, this is not advisable since any crud escaping the consolidation area filtration system could be introduced into the spent ' fuel pool. i Obeiously it is desireable to limit the potential for crud escape to the smallest possible area. This can be best accomplished by isolating the ! ( cask laydown pit from the spent fuel pool except during fuel transfer. To j do this, an auxiliary cooling system would be required in the consolidation area. g, I' 1 , Since the heat load is small, a very simple heat exchanger will be i sufficient. It is recommended that a stainless steel pipe with cooling water circulating through it be used. For example, a 75 foot length of two j and one-half inch schedule 40 pipe would provide the necessary surface area f i f or cooling the consolidation area. The circulating cooling fluid can be j supplied from either the Reactor Building Closed Cooling Water System j (RBCCW) or from the Spent Fuel Pool Heat Exchanger (SFPHX) process outlet. ; Since either source is capable of supplying the necessary cooling, the f selection depends on the piping arrangement that-is more readily f accessible. A flow rate of 25-50 gpm provided by a small pump would be sufficient. Effluent cooling water shall be returned to the RBCCW or SFPHX ! i - outlet downstream of the point from which the influent is drawn. l' 5.3 Spet Fuel Consolidation Filtration { l? A vouum/ filtration system shall be provided in dccordance with Combustion [ Engineering Development Department Specification Number 00000-ESE-538. , Tnis specification provides for two independent filter trains with remote j4 filter change out capability. Selection of filter size and style shall be ' made on the basis of minimum filter usage. If the filtration system should f ail end large amounts of crud are released ( into the cask laydown area, the transfer gate will remain closed. In this t event, a submersible filter system must be utilized to remove the contamination. Keeping the transfer gate closed while the crud is filtered j will preclude contamination from entering the spent fuel pool. : ( i k , i l i
Appendix A ( SPENT FUEL DECAY ENERGY CALCULATIONS l I Full Core Off LoadII) { t In accordance with refueling procedures, one third of the core has been in -j operation for up to three years, one tnird for up to two years, and one third for up to one year. Using data obtained f rom Appendix B, the maximum heat output for a 2700 MWI2) full core off load is determined as follows: ; t Quantity Operating Time P/P g P (MW) . 1/3 Core 3 years 3.3 x 10-3 2.97 I 1/3 Core 2 years 3.0 x 10-3 2.70 1 year 2.8 x 10-3 2.52 ! 1/3 Core 6 217 Assemblies 8.19 MW (27.94 x 10 BTU /hr) t 11 Unconsolidated Spent Fuel-Normal Operation (less than five years decay time) All spent-fuel bundles in the category have had e full three year operating life. ' Quantity Decay Time P/P g P (MW) ' 1/3 Core 5 a) 150 hours (5.4 x 10 sec) 3.3 x 10-3 2.97 b) 19 days (1.6 x 106 sec) 2.0 x 10-3 1.80 - 1/3 Core 1 year (3.2 x 107 sec) 3.7 x 10-4 .333 2.1 x 10-4 .189 l 1/3 Core 2 years (6.3 x 107 sec) 1/3 Core 3 years (9.5 x 107 sec) 1.3 x 10-4 .117 1/3 Core 4 years (1.3 x 108 sec) 9.0 x 10-5 .081 362 Assemblies (1) Full core off load is assumed to take 150 hours (2) Each one-third core is assumed to be 900 MW
' A-1
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'I a) The maximum heat load from unconsolidated spent fuel (most recent one-third 6 ' i core decayed for 150 hours): 3.69 MW (12.59 x 10 BTU /hr.).
b) The heat load from unconsolidated spent fuel with the most recent one-third 6 core having decayed for nineteen days: 2.52 MW (8.60 x-10 BTU /hr). III Unconsolidated Spent Fuel-Abnormal Operation (less than five years decay time) Quantity Decay Time P/P o P (MW) 1/3 Core 36 days (3.1 x 106 sec) 1.6 x 10-3 1.44 1/3 Core 1 year +36 days (3.5 x 107 sec) 3.5 x 10-4 .315 7 2 years +36 days (6.6 x 10 sec) 2.0 x 10-4 .180 1/3 Core 1/3 Core 3 years +36 days (9.8 x 107 sec) 1.3 x 10-4 .117 1/3 Core 4 years +36 days (1.3 x 108 sec) 9.0 x 10 " .081 362 Assemblies 2.13 (7.27 x106 BTV/hr) - IV Unconsolidated Damaged Fuel Bundles , 3 Fuel bundles in this category are assumed to have had a two year operating life and been removed from the core because of damage.. For conservatism, it will be assumed they have decayed for one year. , Quantity Decay Time P/P g P (MW) P(BTV/ne) 14 bundles 1 year (3.2 x 107 sec) 2.8 x 10 # .049 .17 x 10 6 ; 14 bundles 1 year +36 days (3.5 x 2.5 x 10-4 .044 .15 x 10 6 7 l 10 ) t A-2 l 4
V Consolidated Fuel Bundles ( ' All consolidated fuel bundles have decayed at least five years. A full ~ core of consolidated bundles will be stored in 109 storage slots as opposed , I1 to 217 slots for an entire core load of unconsolidated fuel bundles. i These decay energy values are applicable to both Normal and Abnormal Operation since an additional 36 days decay time has a negligible effect on the heat load. ji Quantity Decay Time P/P o P(MW) 1/3 Core 5 years (1.6 x 108 sec) 7.2 x 10-5 .065 '! 1/3 Core 6 years (1.9 x 108 sec) 6.1 x 10-5 .055 lf' 1/3 Core 7 years (2.2 x 108 sec) 5.6 x 10-5 .050 1/3 Core 8 years (2.5 x 108 sec) 5.0 x 10-5 .045 , 1/3 Core 9 years (2.8 x 108 sec) 4.7 x 10-5 .042 1/3 Core 10 years (3.2 x 108 sec) 4.5 x 10-5 .041 > 1/3 Core 11 years (3.5 x 108 sec) 4.2 x 10-5 .038 1/3 Core 12 years (3.8 x 108 sec) 4.0 x 10-5 .036 1/3 Core 13 years (4.1 x 108 sec) 3.9 x 10-5 .035 1/3 Core 14 years (4.4 x 108 sec) 3.8 x 10-5 .034 . 1/3 Core 15 years (4.7 x 108 sec) 3.7 x 10-5 .033 1/3 Core 16 years (5.0 x 108 sec) 3.5 x 10-5 .032 1/3 Core 17 years (5.4 x 108 sec) 3.4 x 10-5 .031 1/3 Core 18 years (5.7 x 108 sec) 3.3 x 10-5 .030 { I/3 Core 19 years (6.0 x 108 sec) 3.2 x 10-5 .029 1/3 Core 20 years (6.3 x 108 sec) 3.1 x 10-5 .028
. 1/3 Core 21 years (6.6 x 108 sec) 3.1 x 10-5 .028 1/3 Core 22 years (6.9 x 108 sec) 3.0 x 10-5 .027 1/3 Core 23 years (7.3 x 108 sec) 2.9 x 10-5 .026 1375 Assemblies .705 MW (688 Storage Slots) (2.41 x 10 6BTU /hr; A-3 l
lr
- I Appendix B
{ SPENT FUEL TOTAL DECAY ENERGY CURVES ; l, This appendix contains the following decay energy curves applicable to the calculations contained in Appendix A. Total Decay Energy, One Year Operation B-2, B-3
- Total Decay Energy, Two Years Operation B-4, B-5 Total Decay Energy, Tnree Years Operation B-6, B-7 n
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~ . . . . . . . Appendix C' i
'('
CASK LAYDOWN AREA HEAT INPUT The following heat inputs from miscellaneous equipment operating in the f cask laydown area were provided by C-E's Engineering Services and , Development Department. I
}
Component Heat Input (BTU /hr) 5 Filter System (Two 15 HP Pumps) 96,000 Compactor System 120 Transfer Piston 5 Pool Lighting (Four 500 Watt Lights) 7,000 Camera Lights (Three 1000 Watt Lights) 10,500
- Ten Fuel Bundles 30,000 143,625 .
l The total heat input is approximately 145,000 BTU /hr. Neglecting heat losses through the walls, evaporation, radiation, etc., the cask laydown water temperature will rise approximately 6.0*F per eight hours of fuel -l consolidation activity. ;
. l 4
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