ML19225C835
| ML19225C835 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/20/1979 |
| From: | East J, Mckay H, Neil R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML19225C832 | List: |
| References | |
| NUDOCS 7908020567 | |
| Download: ML19225C835 (38) | |
Text
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O g umic UNITED STATES CF AMERICA 22 -
MUCLEAR REGULATORY COMMISSION Qs JUN22 em .e s. r-w79I4 "l Gottas** % $www BEFORE THE ATOMIC SAFETY AND LICENSING BOAa N c-In the Matter of )
)
VIRGINIA ELECTRIC AND POWER COMPANY ) Doc. Nos. 50-338 SP
) 50-339 SP
)
)
(North Anna Power Station, ) (Proposed Amendment Units 1 and 2) ) Operating License NPF-4)
VEPCOS ANSWERS TO CEF INTERROGATORIES Tnese are the answers of Virginia Electric and Power Company (Vepco) to the interrogatories served on the company June 1, 1979, by in tervenor Citizens ' Energy Forum, Inc. (CEF).
Before responding to the interrogatories, Vepco will address several T.atters raised in the introductory section of CEF 's request.
Vepco proposes to have a panel of witnesses present its testimony in this pr oceeding . At this time Vepco plans to have the following witnesses participate on that panel: (1) Dr.
Morris L. 3rehmer, (2) Mr. Henry H. Barbour, (3) Mr. c Rc'ert W.
Calder and (4) Mr- H. Stephen McKay. The professional qu,lifications of Mr. Barbour are attached; qualifications of the other gentler.en have already been supplied to tne intervenors. If for some reason one or more of the individuals 431 OGr we3 79 o 3Fff7
.t identified above will not be able to particinate, or if Vepco finds it necessary to add an additional witness to the panel, Vepco will notify CEF as soon as possible.
Because many of the issues overlap, the participants on Vepco 's panel will be generally f amiliar with all the intervenors' contentions. When a question calls for expertise in a particular area, the T. embers nf the panel will decide who is best able to answer tne question. Mr. McKay has general knowledge of all the issues involved in this proceeding. Mr.
Calder has expertise in the area of materials integrity and corrosion. Dr. Brehmer will be prepared to discuss the no n-r ad iolog ical environmental impacts of the proposed action.
Finally, Mr. Barbour will be prepared to answer questions about alternatives to the proposed c.tlon, particularly those 'aving to do with the storage and disposition of spent nuclear fuel.
At this timt Vepco has not completed its written testimony and is the efore unable to provide a summary of that testimony. Any JocuL nts Vepco relies upon to support its presentation wi:1 be identified in the written testimony. This testimony should be filed within the next few days.
Vepco objects to CEF's request for production of detailed drawings of the new spent fuel pool racks (see Cuestion 5-1). As explained in Vepco's statement of objections, these drawings are considered proprietary to the
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- t VEPCO'S ANSWERS Contention 1: Thermal Effects 1-1 Describe in detail, and provide sketches to illustrate, the spent fuel pool cooling system; specifically,
- a. its relation to the component cooling system,
- b. its relation to the service water cooling system, and
- c. its relation to the pumphouse.
Vepco answer 1-1 The spent fuel pool cooling system takes water from the pool, pumps it through the tubes of the shell and tube heat exchangers (or coolers, and returns it to the pool. Heat is transferred through the heat exchanger tube walls to component cooling water on the shell side of the heat exchanger.
S im ila rly, the component cooling system has pumps and heat exchangers chat transfer heat to the service water. Service water is pumped from the service water reservoir by pumps in the service water pump house to various heat exchangers in the Station, including those in the component cooling system, and then back to the service water sprays in the service water reservoir. The sprays transfer heat to the air as the water droplets fall back into the reservoir.
A more detailed description, including a diagram of the spent fuel pool cooling system, is in section 9.1.3 of the 4 3 _4 -
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- t Nor th Anna Ur.its 1 snd 2 FSAR, which is in the NRC Public Document Room. The component cooling and service water systems are described in sections 9.2. 2 and 9.2.1, respectively. The general layout can be seen in FSAR figures 1.2-1 and 1.2-10 through 1.2-13. There is also a diagram ( figure 5-2) in vepco's Application.
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t 1-2 On page 54 of the Summary of Proposed Mod if ic a t io ns to the Spent Fuel Storage Pool Associated with Increasing Storage Capacity for North Anna Power Station Unit Nos. 1 & 2 (hereafter referred to as Application) is stated "the existing cooling system has suf ficient design margin to remove the additional heat load when uranium fuel is stored in the pool." What is the basis for this statement? What tests have been conducted (e.g. in other operating plants) to assure that this is the case?
Vepco answer 1-2 ,
The basis of this statement is calculations performed using the assumptions in section 7.2 (pages A7 and 43) and Table 7-1 of the Application.
No specific tests have been conducted to assure that this is the case, but the operating experience of the spent fuel pool at Surry Power Station, presented in section 5.5.1, page 21, of the Application is an indication of the conservatism that is included in both the amount of heat assumed to be generated by the spent fuel and in the design of the spent fuel pool cooling system. The conservatism inherent in these calculations is also discussed in NUREG-0404 on pages D-6 throtgh C-3.
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1-3 What is the volume of water that will be kept in the spent fuel pool at all times, both in level ano in volume?
At what rate and force (psi) does the cooling water flow into the spent fuel pool? At what temperature is the cooling water upon entrance to the pool, in the pool, and at exit from the pool?
vepco Answer 1-3 The water level that will be maintained in the spent fuel pool is shown in Figure 5-13 of the Application as elevation 289'10". The botton of the pool is at elevation 249'4", which results in a water depth in the pool of 40'6".
As described on page 20 of the Application, if tho water level drops six inches below the normal level, the high/ low water level annunciator will alarm in the control room. The volume of water in the pool at its normal operating level is 4.932 x 5
10 gallons. The operating characteristics of the fuel pool cooling pumps are shown on page 16 of the Application.
The temperature of the cooling water entering and leaving the pool depends upon (1) the temperature of the component cooling water, (2) the amount of heat transferred to the water f r om spent fuel, and (3) the number of components of the spent f zel pool cooling system that cre in operation.
Without making assumptions about these variables, it is impossible to calculate the entrance and exit temperaturer of f
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the water. The bulk temperature of the water in the pool is the seme as the temperature of the water as it ex its the pool .
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1-4 What heat is generated per fuel assembly at the time of initial storage in the pocl? What total heat is generated at the time of refueling (150 spent fuel assemblies)?
Define stabls temperature conditions as they apply to spent fuel assemblies. How long does it take spent fuel assemb.'ies to reach stable temperature after reactor shutdown?
Veoco answer 1-4 The heat generated by a spent fuel assembly is a function of two factors: the amount of exposure the fuel has received while in the reactor core and the amount of time tnat has passed since the reactor was shut down and the fuel was removed from the core. A rough estimate of the heat generated by one assembly can be calculated by dividing the heat generation rates in Table 7 ' of Vepco's Application by the number of assemblies involved (i.e., 53 for the ncrmal case and 157 for the abnormal case). Note, however, that the two answers will not be the same, because the bases for the numbers are different. For the normal case it was assumed tnat the fuel had seen 816 effectiva full-pcwer days (EFPD) ar . load factor of 35% and tha t these assemblies were all discharged to the pool 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after unit shutdown. The abnormal case assumes a full core discharge with each one-third of the core having 272, 54 1, and 316 EFPD, respectively, a load factor of 85s and a 1 the fuel assemblies in the pool 252 hours0.00292 days <br />0.07 hours <br />4.166667e-4 weeks <br />9.5886e-5 months <br /> after Jf g.
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unit shutdown, as described in section 7.2 of vepco's Application. To calculate the exact estimated heat generation rate for a given fuel assembly, consult the NRC's Branch Technical Position 9 4 0g *he
- Standard Review Plan.
The temperature of a spent #uel
& assembiv
. actually never stops decreasing. However, the rate of temperature change decreases with time (see Branch Technical Position 3-2).
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- 31) 1-5 Describe the effect, if any, of the spent fuel pool building ventilation system on maintaining an acceptable temperature in the spent fuel pool. In che event of an accident, would the ventilation system be relied upon to maintain acceptable spent fuel pool temperatures? If so, how?
Vepco answer 1-5 The fuel pool building ventilation system is not designed to control the temperature of the spent fuel pool.
However, as it does remove air (and thereby moisture in the air) from the fuel building ataosphere, it removes heat that is removed from the pool through evaporation. This heat is not considered in the SFP cooling calculation, and it would not be relied upon for cooling the pool in the event of an accident.
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1-6 Provide an analysis of the flow c:7 water .n the pool with filled and partially-filled high-density racks. Show anticipated water temperatures throughout the pool, identifying areas of highest and lowest temperatures under normal and emergency conditions, assuming 1/3 core in the pool 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after reactor shutdown.
Veoco answer 1-6 A thermal-hydraulic analysis of the fL.ow of water in and around the new spent fuel rack structure has been performed by NUS Corporation, the rack vendor, and is summarized in section 6.6 of Vepct s Apolication. This analysis assumed that the racks were fully loaded with spent fuel and demonstrated that sufficient flow would be induced by natcral convection to preclude local boiling in the hottest atorage location. This analysis is considered proprietary by NUS Corporation and is not being produced (see Vepco's statement of objections). A thermal-hydraulic analysis of partially filled racks has not been performed.
To acdress the question of anticipated water temper-atures throughout the fuel pool, a fuel building temperature study was performed in November 1976 during the refueling at the Surry Power Station. Temperatures at different locations and depths in the fuel pcol were measured periodically during the fuel transfer operation. Throughout this test the
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temperat' ire of the water in the fuel pool wat relatively uniform (within 1 or 2*F). A copy of the special test procedure and the data sheets are attached.
The area of highest spent fuel pool temperature would be in the storage cell containing the fuel assembly with the highest heat generation rate. The lowest temperature in the fuel pool would be where the fuel pool cooling water enters the fuel pool from the fuel pool coolers.
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2-7 (a) erovide detailed information or. the makeup and coolinc water systems specifically mentioned on page 54 of the Application, including a Jelcription of the relationship to one another of the four systems mentioned. (b) How wou'.d diversion of water trom these systems affect other components of the plant? (c) Are any of these systems also relied upon as back-up systems to other plant components? (d) What type of changes in " valve lineup" c -hor " temporary measures, such as the use of temporary pumps or noses" would be required in order to make use of these makeup and cooling water sources?
How would such changes be accortplished? Would these changes be manual, or automatic? In the event of a release of excessive radioactivity in the spent fuel pool area, would it be possible to make these changes? (e) What are the "n 2=ber of installed station systems" cited on page 55 of the App'ication that could provide makeup and cooling water if needed, and how would the ir diversion to the spent fuel pool affect other plant components?
Vepco answer 1-7 (a) Primary grade water is stored in the two primary grade water tanks. Interco nections with these tanks are shown on FSAR figure 9.3.5-3.
The fire protection system eces ater from either rne lake or the service water reservoir. It is described in detail in FdAR sectior 9.5.1.
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The boron recovery system is used to remove boric acid from reactor coolant and purify the water. This purified water can be uscd as makeup. A detailed description is in FSAR section 9.3.5.
The refueling water storage tanks are described in FSAR section 6.2.2.2. They receive makeup water from the primary grade water system through the bor ic acid blender .
(b) The effect that diversion of water from one of these systems would have on othe; comporents of the plant would depend on the particular demands on that system at the time.
If diverting water from one system vould have an adverse effect, Vepco would simply obtata makeup from another system.
'c) Yes.
(d) Each of the four water sources can be intreduced by manual changes in t ' valve lineup. The valve lineup for use of water from the primary grade water sysuem, baron recovery system, or refueling water storage tank can be seen in FSAR figures 9.1.3-1 and 9.3.5-3. The valve lineup for use of the fire protection system can be seen in JSAR figure 9.5.1-1.
Normal makeup is done in accordance with procedure 1-OF-16.5.
(attached). Under abnormal conditions procedure 1-AP-27 is used (attached). Makeup water can be introduced even in the event of an accident in the fuel pool, because the valves involved are outside the fuel building.
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(e) The installed syste:ns mentioned on page 55 of the Application are those listed on page 54.
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Contention 2: Rad io ac t ive Emissions 2-1 What systems currently exist to notify Louisa County officials and residents, as well as persons in surrounding counties, in the event of an unusual release of radioactivity f rem the spent fuel pool? If no such systems exist, are plans underway for their development?
Vepco answer 2-1 The procedures for notification of the Louisa County officials in the event of unusual release of radioactivity are set forth in the North Anna Power Station Emergency Plan.
Notification of residents is provided in che Commonwealth of Vir g in ia , and the County of Louisa Radiological Emergency Response Plans. These documents may be found in the NRC Public Document Room. Notification procedures are also discussed in considerable detail in the record of the .; orth Anna 1 and 2 aperating license hearing held November-December 1976; the transcripts and exhibits should be in the NRC Public Document Room.
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2-2 Answer for each of the following: Lake Anna, groundwater, air and land surrounding the North Anna Power Station. (a) How often are these monitored for radiation levels, to what distance from the plant are they monitored, and by whom are they monitored? (b) Are the results of such monitoring reported to the public? If so, how? If not, why not? (c) What is the normal background radiation level for each? What is the current average desage above background for each?
Vepco answer 2-2 (a) Monitoring frequencies and locations are listed in the Environmental Technical Specifications, which are Appendix B to the North Anna Unit 1 operating license and may be found in the NRC Public Document Room. Vepco does the monitoring.
(b) Yes. The results are reported to the NRC in the Environmental Radiological Monitoring Report, which may be found in the NRC Public Documer.t Room.
(c) Background levels and doses can tu n etermined from the Environmental Radiological Moniroring Reports.
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2-3 What would the effect be on the pool and racks, and on the keff in the pool, if seismic conditions were to cause two or more racks to slide closer together than the planned 14" center-to-center spacing? Paw far, and in what directions, are the racks designed to slide unjer seismic cond it io n s ?
Veoco answer 2-3 The spent fuel racks are restrained by embedment clips on the fuel pool embedment plates and are not free to slide together such that two storage cells in adjacent racks could be closer than 14". Vepco has therefore not analyzed these effects.
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2-4 In the sentence "Mec..anical restriction will be provided to prevent an unptotected fuel assembly from being brought closer than 5" to the side of any rack assembly in the side water ch an n e.l . " , found on page 45 of the Applico; ion, wnat is meant by the terms " mechanical restriction" and " s id 's water channel"?
Vepco answer 2-4 The term " mechanical restriction" refers to a metal plate that pretrudes from the top of certain spent fuel racks that are adjacent to the pool walls. It physically prevents b r ing ing the fuel assembly closer than 5 inches from a side of the spent fuel rack.
The term ' water channel" refers to the region between the spent fuel rac and the fuel pool wall.
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2-5 In the Application, an accident is assumed in which a fuel assembly is dropped so it is " parallel to and at the same level as the stored fuel in rack assemblies." Has an accident been postulated to involve the dropping of a spent fuel assembly so that it is not parallel, but rather perpendicular to, the spent fuel assemblies in the re:ks? If so, provide details and results. If not, explain why not.
Vepco .1nswer to 2-5 Yes. An accident involving a fuel assembly dropped on top of the racks, perpendicular to the stored fuel assemblies, is discussed in section 6.7.1 of the Application.
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2-6 W'/.. n we r e the two new embedments added to the spen, c- el pcol? Provide copias of documents
- o and from the Nuclear Regulatory Commission mequesting and approving, respective 17, the addition of t n e new embedments.
vepco answer 2-6 The two new embedmento- in the seen.. fue1 pco1 werg added in the spring of 1977*
There was no such cor-aspondence.
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2-7 Has the potential exposure of the populace within a 10-mile radius (or other such radius) of the North Anna Power Station in the event of an accident in the spent fuel pool that releases radiation, been analyzed? If so, provide the results of any such analycis. If not, explain why not.
vepco answer 2-7 Yes. The results of an analysis of potential exposure due to an accident in the spent fuel pool are discussed in FSAR section 15.4.5.
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2-8 In a letter from Sam C. Brown, Jr. to Harold Denton, dated September 7, 1978, and responding to requests for additional information regarding expansion of spent fuel capacity for North Anna Units 1 and 2, is the statement:
If the spent fuel racks become contaminated, they will be removed from the spent fuel pool by the overhead crane and taken to the decontamination building where they are decentaminated to the lowest possible level.
The racks are then moved to a tent in a suitable location, cut up, packaged in wcoden boxes, and shipped off-site for burial.
(a) What is the " lowest possible level" of contamination, in terms of measured radiation? (b) Where is a " suitable location"? (c) What special provisions are taken in constructing the tent and wooden boxes mentioned, to insure that they will contain any residual centamination? (d) What exposure to the public is anticipated from such activity? How is this expected exposure arrived at? (e) If such activity is c.ecessary, where will the cut-up racks be shipped for burial?
Vepco answer 2-8 (a) The term " lowest possible level" means lowest possible level given existing decontamination techniques. At this time we cannot quantify what this level will be.
(b) A " suitable 1ccation" would be an area within the Station's restricted area that is accessible by the fuel building overhead crane.
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(c) The tent would be constructed of wood and herculite with a filtered ventilation system that would draw a slight negative pressure on the tent so that any leakage would be inleakage, thereby preventing a release of contamination.
(d) The exposure to the public from this activity should be negligible. As discussed in the previous answers, the decontamination procedures and the use of a tent should prevent any offsite exposure to radiation.
(e) A licensed burial site, sach as the facility in Barnwell, South Carolina.
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2-9 Section 7-4 of the Application states "the escape of gaseous or volatile fission products from even vefective fuel is expected to be negligible". Define the word negligible in this case. What i; the procedure by which negligible amounts of radioactivity are quantified.
Vepco answer 2-9 Webster 's New Collegiate Dictionarv, defines n gligible as "so small or uni.mportant or of so little consequence as to warrant little or no attention." In this application these releases are negligible in comparison to releases resulting from normal station creration, which in turn are well within NRC regulatory limits.
$?y Contention 5: Corrosion 5-1 Provide detailed drawings of the proposed new spent fuel pool racks, clearly noting all dimensions (including w.- ght). Include all partitions from side to side, support structures, projections, and points of attachment to the pool or liner.
Vepco answer 5-1 See Vepco's statement of objections. The detailed engineering drawings referred tc in this interrogatory are considered proprietary to NUS Corporation, the rack supplier.
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5-2 Given the additional amount of corrosion and fission products to be found in the spent fuel pool in light of the proposed modification, what will the effect be on worker exposures incurred in maintaining the fuel pool purification system (e.g., changing of filters) over the lifetime of the pool?
Vepco answer 5-2 A discussion of the personnel exposures experienced at Surry Power Station for the operation and maintenance of a spent fuel pool is contained in Vepco's Application in section 5.5 and in our letter of September 7, 1978 to the NRC in response to question number 5 in its letter of August 17, 1978.
Vepco expects similar exposures at North Anna. A copy of this letter is attached.
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3-3 What will be the effc:ts of the heat load increase in the spent fuel pool on the rate of corrosion of the circonium allc / cladding of the spent fuel assemblies? What will be the effect on the corrosion rate of the stainless steel racks? Provide references to studies which s,Oport your answers.
Veoco answer 5-3 The corrosion rate of materials is te.rperature dependent. Even with the sligbcly increased heat loads, however, th e temperature of the spent fuel pool will be roughly the same as it would have been without the cdditional storage capacity. Thus, the corrosion rate of the circoriium cladding and the stainless steel racks will be unaffected. Please refer to NUREG-0404, Append ix H.
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5-4 (a) What possible contaminants may be released in the spent fuel pool by defective spent fuel assemblies? (b)
What number of defective rods may be stored in the pool without overloading the capacity of the pool filtration system to maintain water purity? (c) What will be the effect of defective rods stored in the pool on the rate of corrosion of the zircalloy (sic) cladding of the spent fuel assemblies to be stored in the pool?
Vepco answer 5-4 (a) The " contaminants" that would be released to the spent fuel pool would be the fission product inventory that is in the " gap" between the fuel pellets and the cladding. A list of these isotopes is contained in section 15.4.5 of the FSAR.
(b) Vepco does not know. A large number of defective fuel assemblies should not overload the spent fuel pool purification system but might require more frequent filter and demineralizer resin changes.
(c) Defective fuel rods will have no effect on the rate of corrosion of zircaloy cladding. Circaloy was chosen because it is compatible witn reactor fuel materials and fission products.
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5-5 Much of the information in VEPCO's Application draws upon operating experience at the Surry Power Station.
Provide the answers to the Nuclear Spent Fuel Questionnaire, dated November, 1977, to Congressman John Moss' Subcommittee on Oversight investigation concern.ing the Surry spent fuel pool.
Vepco answer 5-5 These answers are attached.
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5-6 What is VEPCO's defective rack detection mechanism?
Vepco answer 5-6 After the new racks were manufactured but before they left the factory, a " dummy" fuel assembly was inserted in each storage cell to ensure the proper fit. Once the new racks are installed, they will be tested again in the same manner. If a rack suffers some damage after it is installed, it will be rechecked in the same man.ter. Other than these checks, Vepco has no program for in-service inspection of sper.t fuel racks.
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5-7 If installed, how will any defective racks be corrected and/or remo. -Om the spent fuel pool?
Vepco answer 5-7 If defective storage locations are identified, the affected rack will not be repaired or removed. Any defective storage location will not be used.
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5-8 Were the propos2d high-density type 304 austenitic stainless steel racks fabricated by Brooks ar.d Perkins?
Vepco answer 5-9 No.
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.: .g-5-9 What was the basis for VEPCO's determination not to use boren panels or borated water in the spent fuel pool.
What are the advantages, both economically and otherwise, for not using either or both? What are the disadvantages?
Vepco answer 5-9 No determination was made not to use boron panels.
Boron panels were not necessary to meet the NRC's criteria for maintaining subcriticality. The advantage of these panels is that they permit closer spacing of fuel assemblies while maintaining subcriticality, which in turn permits storage of more spent fuel i.n a given space. The principal disadvantage of such panels is that they add additional cost to the expense of purchasing the racks, and in Vepco's case, withour any benefit to the proposed modification.
We do intend to use borated water in the spent fuel pool. Borated water is required in the refueling canal, which is connected to the spent fuel pool during refueling (see Technical Specification 3.9.1). A boron concentration this high makes criticality impossible even with the fuel in ics most reactive configuration. The disadvantage is the potential for dilution or inadvertent filling of the pool with unborated water. This is why all our criticality analyses assume unborated water to calculate the maximum possible effective multiplication factor (keff).
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I prepared Vepco answers 1-2, 1-3, 1-4, 1-5, 1-6, 2-3, 2-4, 2-6, 2 3. , 2-9, 5-2, 5-3, 5-4, 5-5, 5-7, and 5-3 abcve.
They are true and correct to the best of my knowledge an?
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- d. Stgonen 4c Ka y ,f/As soc ia ta Eng ineer , Vepcff COMMONNEALTH OF VIRGINIA.
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The foregoing instrument was acknowledged before me this [c'/ O da y o f p;n , 1979, by H. Stephen >!cKay.
My Co7miss on expires 7- /k /[bA
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s I prepared Vepco anssers 1-1, 1-7, 2-5, 2-7, and 5-9.
They are true and correct to the best of my knosledge and belief.
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> Robert v. 'Je i r, As so c ia "e' Engineer, Vepco
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f The foregoing instrument was acknowledged before me thie J o O day of '7~o ,v c , 1979, by ?obert M. Neil.
My commission expires de; /d, /78A
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Nota blic I prepared Vepco answers 2-1 and 2-2. They are crue and correct to the best of my knowledge and ho~ te f .
JYld W James Sacc - Associate p' Eng ineer . Vepco COMMONWEALTH OF VIRGINIA.
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/ The fo r ego ing instrument was acknowledged before me this .,74 5 iay of g/ .c. , 1979, by hmes East.
My commiss on expires ,Z /6 AIA /
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! located the documents for Vepco answer 5-5.
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Dennis P. F i s n c a c .< , Eng inee r ,
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The foregoing in s t r umen t was acknowledged before me this .h '- d a y o f fpc , 1979, by Dennis Fishback.
v,y commission expires CT_/O /NIS 1
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