ML20199K346
| ML20199K346 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 01/22/1999 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F1-99-0010, W3F1-99-10, NUDOCS 9901260399 | |
| Download: ML20199K346 (25) | |
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Enti y Operzthns. Inc.
Killona LA 70066-0751 Tel 504 739 6660 v ce esiden x a ons Waterfcxd 3 W3F1-99-0010 A4.05 PR January 22,1999 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555
Subject:
Waterford 3 Steam Electric Station Docket No. 50-382 License No. NPF-38 Additional Information Regarding Operating License Amendment 144 Re:
1.
Letter WSF1-97-0061, dated March 27,1997 2.
Letter from Mr. Chandu P. Patel, NRR Project Manager, to Mr. Charles M. Dugger, Vice President, Operations, dated November 19,1997 3.
Letter W3F1-97-0270, dated December 12,1997 4.
Letter from Mr. Chandu P. Patel, NRR Project Manager, to Mr. Charles M. Dugger, Vice President, Operations, dated July 10,1998 Gentlemen:
f In order to increase the number of fuel assemblies that can be stored at the Waterford 3 Steam Electric Station, Entergy Operations, Inc. (Entergy) decided to jj replace the current spent fuel storage racks with high-density storage racks. The l
new racks increase spent fuel storage capacity and allow an increase in the maximum initial nominal enrichment of stored fuel from 4.9 nominal weight percent (w%) U-235 to 5.0 w% U-235. To support this modification, Entergy submitted Technical Specification Change Request (TSCR) NPF-38-193 for Waterford 3 (Reference 1). The NRC approved the request and issued Amendment 144 to the Waterford Technical Specifications (Reference 4).
While reviewing TSCR NPF-38-193, the NRC requested a detailed description of how
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the decay heat generated from the spent fuel assemblies stored in the cask storage pit and the refueling canal will be removed (Reference 2). In our response (Reference 3), Entergy provided the requested information, which contained the 99012603N 990122
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PDR ADOCK 05000382~
P PDR j
Additiorial Information Regarding Operating License Amendment 144 W3F1-99-0010 i
Page 2 January 22,1999 statement, "The component cooling water total maximum flow, for fuel pool cooling, is 5000 gpm." The 5000-gpm flow would be needed to remove the spent fuel pool heat load and maintain pool temperature at 140 F under analysis assumptions with only one fuel pool pump operating. Entergy did not discuss in detail how the 5000 gpm flow would be achieved.
While conducting a review of the high-density storage racks design change package, Waterford 3 personnel discovered the design did not adequately address the worst case single active failure for a partial core offload condition as required by NUREG-0800, " Standard Review Plan," Section 9.1.3.
Specifically, the design change package did act adequately address the ability of the Fuel Pool System to maintain spent fuel poo! temperature 5140 F under the assumptions of the high-density rack analysis (e.g., power uprate and lengthened fuel cycle, etc.) while simultaneously considering the loss of a divisional electrical
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bus. At issue was the ability of the Component Cooling Water System to provide cooling water to the fuel pool heat exchanger in order to maintain spent fuel pool temperature 5140 F with a partial core heat load while supplying adequate cooling j
water to essentialloads.
Current fuel design and cycle length ensure the analyzed elevated spent fuel pool heat loads cannot be realized. Therefore, this issue does not impact current refueling operations.
A description of system design and operation is provided in Attachment 1. A revised No Significant Hazards Consideration determination is provided as Attachment 2.
The Technical Specification pages issued as Amendment 144 are not affected by this issue. As discussed in the attachments to this letter, procedural limitations and operator actions are required to maintain the validity of the high4ensity rack analysis.
Therefore, Entergy requests NRC review and approval of these proposed actions.
Per 10CFR50.9(a), information provided to the NRC is to be complete and accurate.
Entergy believes we failed to meet the requirement of 10CFR50.9(a) in our response to the NRC's Request for Additional Information (Reference 3) in that the information provided was not accurate. On July 15,1997, Entergy informed the NRR Project Manager of this situation via telephone.
This letter corrects the inaccuracies and provides complete information. Entergy does not believe the inaccuracies meet the reporting requirement of 10CFR50.9(b),
in that the information in question does not represent a significant implication for
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j Addition'al Information Regarding Operating
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License Amendment 144 W3F1-99-0010 Page 3 January 22,1999 i
public health and safety or common defense and security. The basis for this determination is provided in Attachment 3.
Should you have any questions, please contact Early Ewing at (504) 739-6242.
Very truly yours,
/
W C.M. Dugger Vice President, Operations
. Waterford 3 CMD/RWP/rtk
- Attachments:
1.
Affidavit 2.
Technical Specification Change Request NPF-38-193 i
Revised Information J
3.
Revised No Significant Hazards Consideration Determination 4.
10CR50.9(b) Reportability Determination cc:
E. W. Merschoff, NRC Region IV C. P. Patel, NRC-NRR J. Smith N. S. Reynolds NRC Resident inspectors Office Administrator, Radiation Protection Division (State of Louisiana)
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I
In the matter of
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Entergy Operations, Incorporated
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Docket No. 50-382 Waterford 3 Steam Electric Station
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Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is Vice President, Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Additional Information Regarding Operating License Amendment 144; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
w Charles Marshall Dugger Vice President, Operations i
Waterford 3 1
STATE OF LOUISIANA
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) ss PARISH OF ST, CHARLES
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l Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this 2 e-I'_ day of (1 m
.1999.
,m (J
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't m
l Notary Public i
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1 ATTACHMENT 1 to W3F1-99-0010
Att: chm:nt 1 to W3F1-99-0010 Page 1 of 7 TECHNICAL SPECIFICATION CHANGE REQUEST NPF-38-193 REVISED INFORMATION BACKGROUND l
In order to increase the number of fuel assemblies that can be stored at the Waterford 3 Steam Electric Station, Entergy Operations, Inc. (Entergy) decided to replace the current spent fuel storage racks with high-density storage racks. The new racks increase spent fuel storaga capacity and allow an increase in the maximum initial nominal enrichment of stored fuel from 4.9 nominal weight percent (w%) U-235 to 5.0 w% U-235. To support this modification, Entergy submitted Technical Specification Change Request (TSCR) NPF-38-193 for Waterford 3.' The NRC approved the request and issued Amendment 144 to the Waterford Technical Specifications.2 While reviewing TSCR NPF-38-193, the NRC requested a detailed description of how the decay heat generated from the spent fuel assemblies stored in the cask storage pit and the refueling canal will be removed. In our response, Entergy provided the requested information, which contained the statement, "The component cooling water total maximum flow, for fuel pool cooling, is 5000 gpm." The 5000-gpm flow would be needed to remove the spent fuel pool heat load and maintain pool temperature at 140 F under analysis assumptions with only one fuel pool pump operating. Entergy did not discuss in detail how the 5000 gpm flow would be achieved.
While conducting a design review of the design change package for installing the high-density storage racks, Waterford 3 r,ersonnel discovered the design did not adequately address the worst case single active failure (loss of an electrical bus) for a partial core offload condition. At issue was the ability of the Component Cooling Water System (CCW) to provide cooling water to the fuel pool heat exchanger in order to maintain the spent fuel pool temperature < 140 F with a partial core heat load while supplying adequate cooling water to essentialloads. With this issue raised, a condition report was written to initiate corrective actions.
' Letter #W3F1-97-0061, dated March 27,1997 2 Letter from Mr. Chandu P. Patel, NRR Project Manager to Mr. Charles M. Dugger, Vice President, Operations. Entergy Operations, Inc.," Issuance of Amendment 144 to Facility Operating License NPF-38 - Waterford Steam Electric Station, Unit 3 (TAC No. M98325)," dated July 10,1998
- Letter from Mr. Chandu P. Patel, NRR Project Manager to Mr. Charles M. Dugger, Vice President, Operations, Entergy Operations, Inc.,
- Request for Additional Information (RAI) Regarding Technical Specification Change Request NPF-38-193," dated November 19,1997, item #1, third asterisk
' Letter #W3F1-97-0270, dated December 12,1997, Enclosure, Response to item #1 - third asterisk, page 4 i
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Att: chm:nt 1 to W3F1-99-0010 Page 2 of 7 1
DISCUSSION The cooling portion of the Fuel Pool System is a closed loop system consisting of two 50% capacity pumps, one 100% capacity heat exchanger, and a backup fuel pool heat exchanger. The fuel pool pumps circulate spent fuel pool water through the fuel pool heat exchanger, where heat from the water is rejected to CCW. CCW utilizes two independent trains with one 100% capacity pump per train to supply cooling water to its respective heat loads. Also, CCW has a 100% capacity " swing" pump that can be used in either train.
The original FSAR Section 9.1.3.1 b) stated:
"The Fuel Pool System is designed to remove the decay heat produced in the spent fuel from approximately 45% of a core placed in the spent fuel pool after reactor shutdown in addition to the decay heat from 11 previous refueling batches.
With one fuel pump operating, the maximum spent fuel pool water temperature will not exceed 140 F [ emphasis added)."
NUREG-0800, " Standard Review Plan," (SRP) requires the spent fuel pool cooling system design to consider a single active failure for partial core offload conditions.
SRP Section 9.1.3, Subsection Ill.1.d states, in part:
"For the maximum normal heat load with normal cooling systems in operation, and assuming a single active failure (emphasis added), the temperature of the pool should be kept at or below 140 F and the liquid level in the pool should be maintained."
The spent fuel pool heat load resulting from a partial core offload plus 11 previous 8
batches is 22.2 x 10 BTU /hr five days after shutdown.5 With this heat Icad, the fuel pool heat exchanger can maintain spent fuel pool temperature at 140 F with one fuel pool pump supplying 2000 gpm @ 140 F on the tube side and one CCW pump supplying 1600 gpm @ 90 F on the shell side.8 Therefore, the original design met the SRP requirements, as documented in NUREG-0787, " SAFETY EVALUATION REPORT related to the operation Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382," Section 9.1.3.
Under the assumption of the high-density storage rack analysis, the spent fuel pool decay heat load will increase. As documented in TSCR NPF-38-193, the spent fuel pool heat load resulting from a partial core offload plus previous refueling batches that 5 Letter W3F1-98-0049, dated March 28,1998; Enclosure 1, Response 8, page 9 l
- FSAR Table 9.1-3, Sheet 7 l
. to W3F1-99-0010 Page 3 of 7 must be removed by the Fuel Pool System will increase from 22.2 x 10 BTU /hr to 8
6 33.7 x 10 BTU /hr.7 The table below identifies the major assumptions used to calculate the original and new heat loads. As seen in the table, most of the new heat load assumptions are more conservative than the original assumptions.
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PARAMETERS ORIGINAL NEW ASSUMPTIONS ASSUMPTIONS' Length of Fuel Cycle 18 months 24 months Fuel Enrichment of Transferred Assemblies 4.9 w% U-235 5.0 w% U-235 Reactor Power 3390 MWt 3390 MWt (Cycles 1 - 9) 3661.2 MWt (Cycles 10 - 22)
(8% power uprate)
Elapsed Time After Shutdown Prior to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 72 hours Beginning Partial Core Transfer Transfer Rate from Reactor to Spent Fuel 96 assemblies in the 4 assemblies / hour Pool pool 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown Spent Fuel Assemblies Exposure 4.5 EFPY 4.5 EFPY
- of Fuel Assemblies Discharged from 96 116 Reactor to Spent Fuel Pool for Partial Core Transfer
- of Fuel Assemblies in the Spent F;el 992 2369 Pool Prior to Partial Core Transfer (end of plant life)
Resulting Spent Fuel Pool Heat Load 22.2 x 10" BTU /hr 33.7 x 10" BTU /hr
- Effective Full Power Years In support of TSCR NPF-98-193, Entergy assumed a failure of the more efficient fuel pool pump leaving the remaining pump and two CCW pumps to provide cooling water to the heat exchanger. Entergy relied upon the two CCW pumps to supply 5000 gpm.
Upon further investigation, Entergy discovered the design change package did not adequately identify the results of the worst case single active failure (i.e., a loss of divisional electrical power) impacting the ability of the Fuel Pool System to maintain spent fuel pool temperature 5140 F with high-density storage racks. The loss of one divisional electrical bus would not only disable one fuel pool pump (as assumed in the design change), but also one CCW pump as well. Assuming this single active failure, CCW would not be able to supply the 5000 gpm cooling water flow to the fuel pool heat exchanger.
To determine the impact of this single active failure on TSCR NPF-38-193 and its supporting information, Entergy performed a new engineering calculation. The purpose of the calculation was to determine the most limiting condition, assuming a single active failure, during a normal refueling (partial core discharge) for the fuel pool heat exchanger to maintain the spent fuel pool 5140 F. The results of this calculation 7 Letter #W3F1-97-0061, dated March 27,1997, Enclosure, page 9 8 Letter #W3F1-97-0288, dated January 21,1998
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' to W3F1-99-0010 Page 4 of 7 indicate that with a divisional bus failure, the remaining fuel pool pump and CCW pump (supplying 2440 gpm @ 140 F and 2768 gpm @ 90*F, respectively) would 8
remove 29.0 x 10 BTU /hr and maintain the spent fuel pool at 140 F. (The calculation assumed design basis fouling and 5% tubes plugged.) This value is less than the newly calculated partial core heat load of 33.7 x 10 BTU /hr (assuming 24 month fuel cycle, power uprate,116 assemblies discharged and 5.0 w% U-235).
Entergy then evaluated a configuration in which electrical power is restored to the disabled fuel pool pump. This evaluation indicated that with two fuel pool pumps and one CCW pump operating (supplying 3650 gpm @ 140 F and 2768 gpm @ 90 F, respectively), the fuel pool heat exchanger can remove 33.79 x 10 BTU /hr, thereby 6
providing adequate heat removal capacity for the assumed partial core offload configuration. (As with the calculation discussed above, this evaluation assumed design basis fouling and 5% tubes plugged.)
Furthermore, Entergy determined the disabled fuel pool pump could be powered from the available divisional bus via a temporary electrical connection. Based on the following points, Entergy concluded a temporary electrical connection could be used to satisfy spent fuel pool cooling requirements:
- 1. The spent fuel pool temperature is procedurally controlled at 130 F.8 With the 6
most limiting configuration (33.7 x 10 BTU /hr heat load, one fuel pool pump and one CCW pump supplying cooling water to the fuel pool heat exchanger), spent fuel pool temperature will increase from 130 F to 140 F in approximately 6 %
hours. Previous experience indicates the temporary electrical connection can be installed within the 6 %-hour time period.
- 2. The additional electrical load of the second fuel pool pump presents no adverse impact on the available divisional bus.
- 3. Electrical separation is provided between the safety-related portion of the electrical motor control center and the pump motor. The separation requirements of Regulatory Guide 1.75, " Physical independence of Electric Systems," are maintained at safety-related/nonsafety-related interfaces.
- 4. In the event of a Loss of Offsite Power event while the temporary electrical connection is installed, the Load Shedding and Sequencing System automatically sheds all non-essential loads (which includes both fuel pool pumps).' The fuel pool pumps can be manually realigned, if needed. Therefore, the installed temporary connection would not affect the plant's ability to respond to such a condition.
- Letter W3F1-97-0270, dated December 12,1997
- FSAR Table 8.3-1, Sheets 7 and 14
Atttchm::nt 1 to W3F1-99-0010 Page 5 of 7 Root Cause Determination Results A Root Cause Determination was performed as part of the corrective action process.
The results identified that Entergy personnel did not adequately confirm or verify the assumptions used in the calculations supporting TSCR NPF-38-193. Specifically, Entergy personnel did not recognize the worst case single active failure as a divisional bus failure resulting in a loss of one CCW pump as well as the loss of one fuel pool pump.
Corrective Actions As discussed above, Entergy determined a temporary electrical connection can be locally stationed to allow powering either fuel pool pump from either divisional bus (i.e., Pump A can be powered from Divisional Bus B; Pump B can be powered from Divisional Bus A). Current fuel design and cycle length ensure the analyzed elevated spent fuel pool heat loads associated with the high-density storage rack modification cannot be realized. However, as a conservative measure, the following administrative controls will be established to ensure spent fuel pool temperature is maintained
< 140 F following a failure of a divisional electrical bus. These actions serve as the basis for the acceptability of TS issued under Amendment 144; therefore, Entergy requests NRC approval of these actions.
- 1. Prior to beginning a partial core offload during refueling operations, Entergy will determine if the maximum anticipated spent fuel pool heat load is expected to 8
exceed 29.0 x 10 BTU /hr.
- 2. If the spent fuel pool heat load is expected to exceed 29.0 x 10e BTU /hr, Entergy will implement steps necessary to pre-stage the temporary electrical connection.
- 3. If the plant experiences a loss of a divisional bus powering a fuel pool pump, plant personnel will install the temporary electrical connection to restore power to the disabled fuel pool pump.
Entergy will formalize these actions by incorporating them into the Reload Safety Analysis Ground Rules Document and appropriate plant procedures by February 19, 1999.
In addition to the above operator actions, Entergy will revise the FSAR, as l
appropriate, to discuss the actions as well as reflecting changes resulting from installing the high-density storage racks.
i Waterford 3 management has met with supervisory personnel to reinforce expectations regarding appropriate technical reviews of design documents.
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j Attachmtnt 1 to W3F1-99-0010 Page 6 of 7 Review of TSCR NPF-38-193 No Significant Hazards Consideration Determination Entergy has determined the No Significant Hazards Consideration (NSHC) determination" is impacted by the information contained in this letter. A revised NSHC determination is provided in Attachment 2.
l Review of the NRC's Safety Evaluation for Amendment 144 Entergy has reviewed the NRC's Safety Evaluation for Amendment 144.12 The portion of the Safety Evaluation impacted by this identified condition is Section 2.2, i
. Spent Fuel Pool Cooling and Cleanup System. Entergy has identified the following i
. differences between the Safety Evaluation and the corrected information presented in l
this letter.
l 1.
Page 6, Table providing information on Cases 1 and 2:
a.
The maximum spent fuel pool temperature value for Case 1 is 140'F, rather than the stated 139.41*F.
1 b.
Case 1 is based on the spent fuel pool being cooled via two fuel pool pumps, one CCW pump, and the fuel pool (primary) heat exchanger.
2.
Page 6, second paragraph, last sentence:
I The last sentence should read, "The peak SFP water temperature resulting from this heat load and with a single active failure is 140*F, which is equal to the i
design basis and the guidance of Standard Review Plan (SRP) Section 9.1.3 for i
SFP water temperature during planned refueling."
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3.
The Safety Evaluation does not reflect the actions pertaining to installing a temporary electrical connection as identified in the Corrective Actions section
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l above.
In addition to the differences presented above, Entergy would like to comment on the third paragraph of Page 7. The third paragraph states:
" Letter W3F1-97-0258, dated November 13,1997
" Letter from Mr. Chandu P. Patel, NRR Project Manager to Mr. Charles M. Dugger, Vice President, Operatio".3, Entsrgy Operations, Inc., *lssuance of Amendment 144 to Facility Operating License NPF Waterford Steam Electric Station, Unit 3 (TAC No. M98325)," dated July 10,1998
Attachm:nt 1 to W3F1-99-0010 Page 7 of 7 "The calculated peak bulk temperature of the proposed SFP storage capacity resulting from a full core off-load with two SFP pumps operating are 151.61 F, 155.50 F, and 158.50 F in the SFP, cask storage pit, and refueling canal, respectively. The licensee also performed an analysis to determine the bulk temperature of the combined three regions for a full core off-load assuming a single active failure of one of the two SFP cooling pumps. The calculated peak bulk temperature is 163.490 F."
The intent of the analysis mentioned in this paragraph was to show that, even considering a single active failure, the Fuel Pool System could maintain the pool temperature below SRP temperature limits with a full core off-load. The information contained in the paragraph is accurate; however, the single active failure identified in the second sentence of the paragraph is not the worst case single active failure identified for the partial core off-load. SRP Section 9.1.3, Subsection Ill.d states a single active failure need not be considered for a full core off-load. Therefore, the Fuel Pool System complies with the SRP requirements as justified in the first sentence of the paragraph.
J Assuming credit for the compensatory operator actions discussed above, it is Entergy's position the Fuel Pool System continues to meet the design requirements of SRP, Section 9.1.3. The basic conclusions of the Safety Evaluation remain valid.
CONCLUSIONS
- 1. The worst case single active failure impacting the ability of the Fuel Pool System to maintain spent fuel pool temperature 5140 F with high-density storage racks is a loss of divisional electrical power. This single active failure results in the loss of one fuel pool pump and one CCW pump. Two fuel pool pumps and one CCW pump are required to remove the new calculated spent fuel pool heat load of 6
33.7 x 10 BTU /hr and maintain the spent fuel pool 5140 F for a partial core offload. Waterford 3 has established administrative controls via procedure that, upon loss of divisional electrical power with elevated heat loads in the spent fuel pool, install a temporary electrical connection to restore power to the disabled fuel pool pump. This temporary alteration can be insta!!ed before the spent fuel pool temperature reaches 140 F. Therefore, Entergy believes the established administrative controls in concert with Amendment 144 provide adequate protection for maintaining spent fuel pool temperature 5140 F.
- 2. The NRC's Safety Evaluation for Amendment 144 is based upon previously submitted information. The corrected information presented in this letter impacts information contained in the Safety Evaluation as identified above. Entergy believes the basic conclusions of the Safety Evaluation are not impacted.
However Entergy requests the NRC to review and approve the compensatory operator actions discussed in this letter.
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ATTACHMENT 2 to W3F1-99-0010 1
Att: chm:nt 2 to l
1' W3F1-99-0010 Page 1 of 9 I
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NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION (Note: Revisions are indicated in bold type and deletions are lined-out.)
I In accordance with 10CFR50.92, Entergy has reviewed the proposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC).
The basis for this conclusion is that the three criteiia of 10CFR50.92(c) are not l
compromised. The proposed changes do not involve a SHC because they would not:
l 1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
l In the analysis of the safety issues concerning the expanded pool storage capacity, the following previously postulated accident sconarios have been considered:
l a.
A spent fuel assembly drop in the spent fuel pool l
b.
Loss of spent fuel pool cooling flow I
c.
A seismic event d.
An accidental drop of a fully loaded fuel shipping cask The probability that any of the accidents in the above list can occur is not significantly increased by the modification itself. The probabilities of a seismic event or loss of spent fuel pool cooling flow are not influenced by the proposed changes. The probabilities of accidental fuel assembly or shipping cask drops are primarily influenced by the methods used to lift and move these loads. The I
method of handling loads during normal plant operations remains unchanged, I
since the same equipment (i.e., spent fuel handling machine and cask handling crane) and procedures will be used. A new offset handling tool will be required to access some storage rack cells located adjacent to the pool walls. The grapple mechanism, procedures, and fuel manipulation methods will be very similar to those used by the standard fuel handling tool on the spent fuel handling machine.
l Therefore, this tool does not represent a significant change in the methods used to lift or move fuel in the Fuel Handling Building. Since the methods used to move loads during normal operations remain nearly the same as those used previously, there is no significant increase in the probability of an accident.
During rack removal and installation, all work in the pool area will be controlled i
and performed in strict accordance with specific written procedures. Any movement of fuel assemblies required to be performed to support the
Att: chm:nt 2 to W3F1-99-0010 Page 2 of 9 modification (e.g., removal and installation of racks) will be performed in the same manner as during normal refueling operations. Shipping cask movements will not i
be performed during the modification period.
Accordingly, the proposed modification does not involve a significant increase in the probability of an accident previously evaluated.
The consequences of the previously postulated scenarios for an accidental drop of a fuel assembly in the spent fuel pool have been re-evaluated for the proposed change. The results show that the postulated accident of a fuel assembly striking the top of the storage racks will not distort the racks sufficiently to impair their functionality. The resulting structural damage to a falling assembly and/or a stored assembly has been determined to remain unchanged. The minimum subcriticality margin, K l
eff ess than or equal to 0.95, will be maintained. The structural damage to the Fuel Handling Building, pool liner, and fuel assembly resulting from a fuel assembly drop striking the pool floor or another assembly located within the racks remains unchanged. The resulting structural damage to these items subsequent to this event is not influenced by the proposed changes.
The radiological dose at the exclusion area boundary will increase due to the changes in fuel enrichment and burnup. The previously calculated doses to the thyroid and whole body were 0.47 and 0.11 rem, respectively. The new thyroid and whole body doses based on the proposed change will be 0.553 and 0.304, respectively. These dose levels are extremely small when compared to the levels required by 10CFR100. Therefore, the increase in dose is not considered a significant increase in consequence. Thus, the results of the postulated fuel drop accidents remain acceptable and do not represent a significant increase in consequences from any of the same previously evaluated accidents.
The consequences of a loss of spent fuel pool cooling have been evaluated and found to have no increase. The concem with this accident is a reduction cf spent fuel pool water inventory from bulk pool boiling resulting in uncovering fuel assemblies. This situation 'would lead to fuel failure and subsequent significant increase in offsite dose. Loss of spent fuel pool cooling at Waterford 3 is mitigated by ensuring that a sufficient time lapse exists between the loss of forced cooling and uncovering fuel. This period of time is compared against a reasonable period to re-establish cooling or supply an alternative water source (such as firewater). Evaluation of this accident usually includes determination of the time to boil. This time period is much less than the onset of any significant increase in offsite dose, since once boiling begins it would have to continue unchecked until the pool surface was lowered to the point of exposing active fuel.
The time to boil represents the onset of loss of pool water inventory and is commonly used as a gage for establishing the comparison of consequences before and after a refueling project. The heat up rate in the spent fuel pool is a nearly linear function of the fuel decay heat load. The fuel decay heat load will
Attrchm:nt 2 to W3F1-99-0010 Page 3 of 9 increase subsequent to the proposed changes because of the increase in the number of assemblies and higher fuel burnups. The heat up rate established for the limiting normal heat load conditions prior to reracking was 5.41 F per hour.
This would result in the pool temperature increasing from the maximum normal temperature of 140 F to boiling in a period of 13.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The heat up rate established for the limiting normal heat load conditions subsequent to the proposed changes has been determined as 13.6 F per hour. This would result in the pool temperature increasing from the maximum normal temperature of 140 F to boiling in a period of 5.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
This time to boil comparison was made for limiting normal heat load conditions.
However, the end of this period of time does not represent the onset of any significant increase in offsite doses. As stated above, this consequence would l
result subsequent to fuel being uncovered through unchecked boiling and resulting water level drop of approximately 24.5 feet from normal surface to the top of the fuel storage racks. This depth is conservative, since the top of active fuel is below this level. Subsequent to the proposed changes under limiting normal heat loads the time lapse between the onset of unchecked boiling and uncovering of the racks has been determined to exceed 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.
i As stated above in the safety assessment, subsequent to reracking, the time to boil after loss of forced cooling in the most severe scenario is 2.89 hours0.00103 days <br />0.0247 hours <br />1.471561e-4 weeks <br />3.38645e-5 months <br /> (the ensuing rate of evaporative loss would not result in the fuel being uncovered until after an additional 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />, which is 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after reactor shutdown).
However, the design basis limiting pool heat load under these conditions actually decreases after the proposed modification, because of conservatisms previously used to determine the heat load for this condition. Therefore, the calculated time to boil in this most severe scenario will increase subsequent to the proposed modification. In the unlikely event that all pool cooling is lost, sufficient time will be available subsequent to the proposed changes for the operators to provide alternate means of cooling (i.e., fire water) before fuelis uncovered. Therefore, the proposed changes represent no increase in the consequences of loss of pool cooling.
The consequences of a design basis seismic event are not increased. The consequences of this accident are evaluated on the basis of subsequent fuel damage or compm;nise of the fuel storage or building configurations leading to radiological or cr;ticality concerns. The new racks have been analyzed in their new configuration and found safe during seismic motion. Fuel has been determined to remain intact and the storage racks maintain the fuel and fixed poison configurations subsequent to a seismic event. The structural capability of the pool and liner will not be exceeded under the appropriate combinations of dead weight, thermal, and seismic loads. The Fuel Handling Building structure
i Attrchrn:nt 2 to W3F1-99-0010 Page 4 of 9 l
will remain intact during a seismic event and will continue to adequately support and protect the fuel racks, storage array, and pool moderator / coolant. Thus, the consequences of a seismic event are not increased.
The consequences of a spent fuel cask drop into the cask storage pit have been analyzed along with the new rack storage configuration. This evaluation concluded that there is no increase in consequences. Administrative controls, appropriate changes in load paths, and crane travel limits will continue to preclude handling heavy loads above stored fuel. Therefore, a cask impacting stored fuel is not a postulated event. Potential damage to the cask and contained fuel remain unchanged, since the pertinent parameters for this analysis (i.e., lift height, weight, impact zone configurations, etc.) are not affected by the new rack configurations. The floor was reanalyzed to assess the effect of the additional loading from higher density fuel storage. It was determined that the floor remains intact with minor local crushing of concrete. The liner plate would sustain limited damage, which is repairable. Leakage would be limited to flow through the leak chase system and would be collected at the sump. The Fuel Handling Building 1
integrity would not be compromised; therefore, there would be no release of contaminated pool water outside of the building. Makeup water from the condensate storage pool and/or the refueling water storage tank would be adequate to offset loss of water inventory due to any leakage. This accident does not result in any increase in offsite or Fuel Handling Building doses. Thus, the proposed changes do not represent any increase in the consequences of a postulated spent fuel cask drop.
Therefore, it is concluded that the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated.
2.
Create the possibility of a new or different kind of accident from any previously analyzed.
To assess the possibility of new or different kind of accidents, a list of the critical parameters required to ensure safe fuel storage was established. Safe fuel storage is defined here as providing an environment that would not present any significant threats to workers or the general public. In other words, meeting the requirements of 10CFR100 and 10CFR20. Any new events that would modify these parameters sufficiently to place them outside of the boundaries analyzed for normal conditions and/or outside of the boundaries previously considered for accidents would be considered a new or different accident. The criticality and j
radiological safety evaluations were reviewed to establish the list of critical parameters. The fuel configuration and the existence of the moderator / coolant were identified as the only two parameters that were critical to safe fuel storage.
Significant modification of these two parameters represents the only possibility of an unsafe storage condition. Once the two critical parameters were established, 1
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W3F1-99-0010 Page 5 of 9 an additional step was taken to determine what events (which were not previously considered) could result in changes to the storage configuration or moderator / coolant presence during or subsequent to the proposed changes. This process was adopted to ensure that the possibility of any new or different accident scenario or event would be identified.
Due to the proposed changes, the following events were considered as the only events that might represent a new or different kind of accident:
a.
An accidental drop of a rack module during construction activity in the pool b.
Draining the cask storage pit and refueling canal through the floor drains c.
Fuel assembly mispositioning accident in Region 2.
A construction accident resulting in a rack drop is an unlikely event. A new rack lifting rig will be introduced to lift and suspend all but one of the racks using the existing Fuel Handling Building cranes. Either a new temporary hoist or a combination of one of the existing 15 ton cranes and a lifting bag will be used to lift one of the existing eighty cell racks that is adjacent to the east wall of the spent fuel pool. The cranes, hoists and lifting rig have been or will be designed using the guidance of NUREG-0612 and ANSI N14.6. The postulated rack drop event is commonly referred to as a " heavy load drop" over the pools. Heavy loads will not be allowed to travel over any racks containing fuel assemblies. The danger represented by this event is that the pool structure will be compromised leading to loss of moderator / coolant, which is one of the two critical parameters identified above. However, although the analysis of this event has been performed and shown to be acceptable, the question of a new or different type of event is answered by determining whether heavy load drops over the pool have been considered previously. The postulated drcp of a pool gate was previously evaluated and represents a heavy load drop similar to a rack drop. All movements of heavy loads over the pool will comply with the applicable administrative controls and guidelines (i.e. plant procedures, NUREG-0612, etc.).
Therefore, the rack drop does not represent a new or different kind of accident.
The cask storage pit and refueling canal both have floor drains that will be plugged (a welded closed cover plate) prior to installation of the new storage racks in each of the respective areas. The plugs will preclude any water loss through the drain system. Therefore, draining the cask storage pit and refueling canal through the floor drains is not a postulated event.
Fuel assembly mispositioning in Region 2 is an unlikely event, since locating assemblies that do not meet the burnup criteria will be administratively controlled.
Administrative controls will consist of developing a checker-boarding storage
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Attachm:nt 2 to W3F1-99-0010 Page 6 of 9
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pattern in the Region 2 racks prior to storage or placement of the non-compliant l
fuelin Region 1 racks. The Region 2 mispositioning event represents a change from the previously analyzed condition, since Waterford 3 currently has only l
Region 1 style storage. Therefore, a new fuel storage configuration is possible.
However, the event does not represent a new or different kind of accident, since l
fuel assembly mispositioning is possible with the existing racks through controlled or uncontrolled (assembly drop) lowering of an assembly adjacent to the outside of the storage racks. This condition was previously evaluated and found to be acceptable. The new event was evaluated using similar techniques with similar acceptance criteria and was shown to remain acceptable. Therefore, due to the l
similarity of this new event with that which was previously analyzed it is not l
considered to represent a new or different kind of accident.
The proposed change does not alter the operating requirements of the plant or of l
the equipment credited in the mitigation of the design basis accidents. The l
proposed change does not affect any of the important parameters required to ensure safe fuel storage. Therefore, the potential for a new or previously unanalyzed accident is not created.
l 3.
Involve a significant reduction in the margin of safety.
I The function of the spent fuel pool is to store the fuel assemblies in a subcritical and coolable configuration through all environmental and abnormal loading, such as an earthquake or fuel assembly drop. The new rack design must meet all i
applicable requirements for safe storage and be functionally compatible with the spent fuel pool.
Entergy has addressed the safety issues related to the expanded pool storage capacity in the following areas:
a.
Material, mechanical and structural considerations b.
Nuclear criticality c.
Thermal-hydraulic and pool cooling The mechanical, material, and structural designs of the new racks have been reviewed in accordance with the applicable provisions of the NRC Guidance entitled," Review and Acceptance of Spent Fuel Storage and Handling J
Applications". The rack materials used are compatible with the spent fuel i
assemblies and the spent fuel pool environment. The design of the new racks 2
preserves the proper margin of safety during abnormalloads such as a dropped l
assembly and tensile loads from a stuck assembly. It has been shown that such I
e
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Attachm:nt 2 to W3F1-99-0010 Page 7 of 9 loads will not invalidate the mechanical design and material selection to safely store fuel in a coolable and subcritical configuration.
The methodology used in the criticality analysis of the expanded spent fuel pool meets the appropriate NRC guidelines and the ANSI standards (GDC 62, NUREG 0800, Section 9.1.2, NRC Guidance entitled, " Review and Acceptance of Spent Fuel Storage and Handling Applications", Reg. Guide 1.13, and ANSI ANS 8.17). The margin of safety for subcriticality is maintained by having the neutron l
multiplication factor equal to, or less than,0.95 under all accident conditions, including uncertainties. This criterion is the same as that used previously to establish criticality safety evaluation acceptance and remains satisfied for all analyzed accidents. Therefore, the accepted margin of safety remains the same.
l Normal operating temperature for the spent fuel poolis procedurally
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controlled at 130'F. During the worst case single active failure of a divisional electrical bus, power can be restored to the disabled fuel pool pump in accordance with procedural requirements well before pool l
temperature reaches 140'F. Therefore, the procedural requirements ensure j
the spent fuelpool temperature can be maintained < 140'F.
The ther 2! Sydrre'!r 2nd r^e!' g eur'ert!c ef the ^^e! demenetreted that the p^^! crn be 2!nte! ed ' 'er' the rper! fed t"er
!"-!!r unde-the r^ndit!cne cf the r" u 'ert !^rd 2nd du-5; :!! rred5!c cre!de-t reque cer end re!r-!r m._.
vum _ i.m _ _ m...
m.m _. _ -a u n c a...im..w m.m
.. ~.i.
I i 5rb hf r$E5; hbkThe maxirnum local water temhrature in the IEot r
channel will remain below the boiling point. The fuel will not undergo any significant heatup after an accidental drop of a fuel assembly on top of the rack blocking the flow path. ~ A loss of cooling to the poci will allow sufficient time (5.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the limiting normal heat load) for the operators to intervene and line up l
alternate cooling paths and the means of inventory make-up before the onset of l
pool boiling. The thermal limits specified for the evaluations performed to support the proposed change are the same as those that were used in the previous evaluations. Therefore, the accepted margin of safety remains the same.
i Entergy has also evaluated the impact of the installed temporary electrical connection on the ability of the plant to respond to accident and transient conditions. The installation of the temporary connection meets the electrical separation requirements of Regulatory Guide 1.75, " Physical Independence of Electric Systems." Also, the additional electricalload placed on the available divisional bus by the previously disabled fuel pool pump was determined to have no adverse impact on the available bus.
Therefore, neither the additionalload of the disabled pump nor a fault condition in the disabledpump motor would impact the plant's ability to l
respond to accident or transient conditions.
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Page 8 of 9 i
Thus, it is concluded that the changes do not involve a significant reduction in the margin of safety.
The NRC has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (51FR7751, March 6,1986) of amendments that are considered not likely to involve a SHC. The proposed changes for Waterford 3 are similar to Example (x): an expansion of the storage capacity of j
spent fuel pool when all of the following are satisfied:
i (1) The storage expansion method consists of either replacing existing racks with a
'i design that allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits.
The Waterford 3 spent fuel pool rerack modification involves replacement of the I
existing racks with a design that will allow closer spacing of the stored fuel i
assemblies.
(2) The storage expansion method does not involve rod consolidation or double tiers.
The Waterford 3 reracking does not involve fuel consolidation. The racks will not be double tiered; no fuel assemblies will be stored above other assemblies.
)
(3) The Keg of the pool is maintained less than, or equal to,0.05.
The design of the new racks integrates a neutron absorber, Boral, within the racks to allow cicser storage of spent fuel assemblies while ensuring that K g e
remains less than 0.95 under all conditions. Additionally, the water in the spent fuel pool does contain boron as further assurance that K g remains less than e
0.95. The boron that is contained in the pool is not credited, except in the accident condition.
(4) No new technology or unproven technology is utilized in either the construction j
process or the analytical techniques necessary to justify the expansion.
The rack vendor has successfully participated in the licensing of numerous other racks of a similar design. The construction process and the analytical techniques of the Waterford 3 pool expansion are substantially the same as in the other
. completed rerack projects in the industry. Thus, no new or unproven technology is used in the Waterford 3 reracking.
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Attachm:nt 2 to W3F1-99-0010 Page 9 of 9 1
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l Environmental Considerations l
Entergy has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed changes neither significantly increase the types and amounts of effluents that may be released offsite nor significantly increase individual or cumulative occupational radiation exposures.
Based on the foregoing, Entergy concludes that the proposed changes meet the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the l
requirements for an environmental impact statement.
S90
'S o.egerd!ng the prepered schedu!e fer +h!r 2mendme"+, "!r re';uerted +52t !rruence be ne !:ter ther J2ausa; 9,1999. Th!r 2ppre"2! d2tc !r ",+ecerter; for the exped!!! cut remov2! cf the er! sting oer2ngy 73cgg ggg 3;gg te ggppe ggg gengggggy 7g73 9 j
remp!et!cr d2te of Ju!y 1999,">hich !r v'her Cyc!e 10 fue!"'"! beg!n arri"Mg er r!+e.
Conclusion As discussed herein, the proposed changes to the Technical Specifications In conjunction with the procedural requirements do not involve a SHC pursuant to 10CFR50.92. Reracking the Waterford 3 spent fuel pool has been determined to be safe. Additionally, Entergy has determined that this license amendment meets the criteria delineated in 10CFR51.22 (c) (9) for a categorical exclusion from the requirements for an environmental impact statement.
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ATTACHMENT 3 to W3F1-99-0010 l
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Attachm:nt 3 to W3F1-99-0010 Page 1 of 2 10CFR50.9(b)
REPORTABILITY DETERMINATION in a Request for Additional Information letter, the NRC requested a detailed description of how the decay heat generated from the spent fuel assemblies stored in the cask storage pit and the refueling canal will be removed. In our response, Entergy provided the requested information, which contained the statemnt, "The component cooling water total maximum flow, for fuel pool cooling, is 5000 gpm. 2 Entergy did not discuss in detail how the 5000 gpm flow would be achieved and has since discovered the statement is inaccurate.
10CFR50.9(b) states in part:
"Each applicant or licensee shall notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security."
Entergy does not believe the inaccuracies presented in our response to the NRC's Request for Additional Information meet the reporting criteria of 10CFR50.9(b), as discussed below.
- 1. The Fuel Pool System is not required for safe plant shutdown.
6
- 2. The new calculated partial core heat load (33.7 x 10 DTU/hr) is based on conservative assumptions reflecting future core management plans (i.e., core thermal power uprate, extended fuel cycle, increased number of transferred fuel assemblies, etc.). Such plans are not scheduled for several fuel cycles.
- 3. The calculated spent fuel pool heat load is based on end-of-plant life conditions.
Calculations show that with partial core offloads, the heat load is not expected to 6
exceed 29.0 x 10 BTU /hr (the heat removal capability with one Fuel Pool pump and one CCW pump) for current licensed core power level, fuel designs and cycle lengths.
' Letter from Mr. Chandu P. Patel, NRR Project Manager, to Mr. Charles M. Dugger, Vice President, Operations, dated November 19,1997 2 Letter #W3F1-97-0270, dated December 12,1997
Attachm:nt 3 to W3F1-99-0010 Page 2 of 2
- 4. A complete loss of spent fuel pool cooling was addressed in the NRC's Safety Evaluation for Amendment 144, Section 2.2.1, " Effects of SFP Boiling." In this situation, makeup water for the spent fuel pool can be provided from either the refueling water storage pool (via the refueling water storage pool purification pump
@ 150 gpm), the condensate storage pool (via the CCW makeup pumps @ 600 gpm), or the Fire Water System (@ 100 gpm). The NRC staff found that cooling l
the spent fuel pool by allowing the pool to boil and by adding makeup water in the i
event of a complete loss of cooling capability acceptable since such actions conform with the guidance provided in the SRP. Therefore, there is no threat to public health and safety.
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