ML20212N056
| ML20212N056 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/18/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20209G133 | List: |
| References | |
| TAC-57619, NUDOCS 8608280069 | |
| Download: ML20212N056 (14) | |
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UNITED STATES
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WASHINGTON, D. C. 20555 5"
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s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.17 TO FACILITY OPERATING LICENSE NO. NPF-29 MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
By letter dated May 6, 1985, the' Mississippi Power & Light Company, Middle South Energy, Inc., and South Mississinpi Electric Power Association (the licensees) submitted an application for a license amendment to increase the starage capa-city of the spent fuel pool and the upper containment pool for Grand Gulf Nuclear Station (GGNS) Unit I by replacing the originally installed fuel racks with new high density racks. By letters dated July 29, August 15, August 30, September 11, September 12, November 1, and December 18, 1985, and March 14, March 15, June 5. June 9, and July 25, 1986, the licensees revised and supple-3 mented their application. The notice of consideration of issuance of this license amendment was published in the Federal Register before the licensees' July 25, 1986, submittal. The July 25,'1986, suDmittal contained supplemental information provided in response to the staff's questions regarding plant pro-cedures that would be used during refueling in the event of a loss of offsite power and a subsequent single failure and a commitment to provide increased cooling capacity for the spent fuel pool. This supplemental information did not change the proposed Technical Specifications described in the notice, and the potential need for increased cooling capacity was described in the notice.
The notice of consideration accurately describes the license amendment request, and the supplemental information does not affect the substance of the requested amendment.
The GGNS Unit 1 is a boiling water reactor with a Mark III containment. The spent fuel pool is located in the' auxiliary building, which is similar to spent fuel pool arrangements for pressurized water reactors. Above the GGNS reactor, and within the containment, there is an upper containment pool with racks for holding new fuel to be placed in the reactor and spent fuel removed from the reactor during refueling; however, before reactor startup after refueling, all spent fuel is transferred to the spent fuel pool for storege.
The amendment would revise Section 5.6, " Fuel Storage," of the Technical Speci-fications to allcw increased upper containment pool capacity and increased spent fuel storage capacity. This increased capacity would be obtained by replacing the fuel racks in the upper containment pool and in the spent fuel storage pool with high density fuel racks. The center-to-center distance between fuel assem blies would be changad from 12 inches to 6.26 inches. This reracking would increase the upper containment pool capacity from 170,to 800 fuel assemblies in order to hold a complete core unloading, if necessary, and increase the spent fuel pool storage capacity from 1270 to 4348 fuel assemblies. This would provide spent fuel storage capability until the ' year 2003, assuming reloads of a third 8605280069 860818 PDR ADOCK 05000416 P
r of a core. The capability to off-load the entire core would be available until the year 2000. However, the number of fuel assemblies to be stored in the spent fuel pool would be limited by Technical Specifications to 2324 until spent fuel pool cooling capabiljty is increased.
The Technical Specifications would be changed to limit the spent fuel pool water temperature to 140 F rather than 150 F and require plant shutdown rather than a special report if the limiting temperature were exceeded.
The licensees have removed the originally installed spent fue1 racks, which were not used to store spent fuel assemblies, and have installed the new high density spent fuel racks during a planned outage in the fall of 1985.
The licensees determined, pursuant to 10'CFR 50.59, that removal of-th old racks and installation of the new racks would not involve an unreviewed safety ques-tion. A condition in the operating license prohibits the storage of spent fuel in the spent fuei pool until the standby service water system is modified.
Modification of the standby service water system will be completed during the i
first refueling outage.
2.0 EVALUATION 2.1 Criticality Considerations The high density spent fuel storage racks consist of double-walled stainless steel boxes with Boraflex neutron absorber sheets in the space between the walls. The inner dimension of the square boxes is 6.0 inches, and the boxes are arranged in an array having a 6.26-inch center-to-center spacing. The Boraflex sheets are 144 inches in length. As a result, 3 inches of the active fuel length of the assemblies extend beyond the Boraflex on each end. This fact was accounted for in the analysis. The Boraflex absorber contains 0.0204 gram of 8-10 per square centimeter of surface area.
2.1.1 Calculation Methods The nuclear criticality analysis of the sp nt fuel racks was performed with the AMPX-KEN 0 computer package using the 123-group GAM-THERMOS cross-section set with the NITAWL treatment of U-238 resonance effects. This calculation proce-dure is widely used for fuel cack criticality analyses and is acceptable.
It 1
has been verified by Southern Science, who did the analysis, by comparison with
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critical experiments. A calculational bias has been determined from the cum-parisons. The nominal design case assumes an 8x8R assembly having fuel rods with a uniform enrichment of 3.5 weight percent (w%) U-233. The following con-servative assumptions are made:
(1) The moderator is pure wa'ter at the temperature yielding the maximum reactivity.
(2) The racks are assumed to be infinite in extent in the lateral and vertical directions.
(3) The fuel is assumed to be fresh and no credit is taken for burnable poison.
Such fuel is more reactive than that which has burned out to its highest reactivity point.
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'(4) No credit is taken for absorption in minor structural members (spacers, etc.), and the cladding is assumed to be pure zirconium.
Uncertainties treated in the analysis include those due to Boraflex thickness, width, and B-10 concentration; fuel enrichment, density, and diameter; lattice pitch; stainless steel thickness; and flow channel distortion. The reactivity of the racks is maximum at the lowest temperature (39*F). The maximum reactiv-ity occurs with the assembly centered in the storage box. 1hese uncertainties 4
are the ones usually considered in spent fuel reactivity analyses and are acceptable.
2.1.2 Results and Conclusions Abnormal and accident situations analyzed included heating the pool water to boiling and reducing the pool water density to 0.1 gram per cubic centimeter, closing of the water gap between r.acks, dropping of a fuel assembly onto the racks, and positioning of a fuel assembly outside the racks. The analyses showed that in no case was the b 'ffective of the racks greater than that.of the design case. The staff conc.udes that the full range of ecciden,and ab-normal situations has been considered.
The results of the analyses show that, 'or 8x8R assemblies with uniform fuel enrichment of 3.5 w% U-235, the k-effective of the racks is 0.937 including all uncertainties (taken at a 93% probability with 95% confidence).
Since a uniform enrichment distribution bounds the results for anticipated enrichment distributions, the staff concludes that storage of fuel assemblies having aver-age planar enrienments of less than or equal to 3.5 w% U-235 enrichment may be safely stored in the GGNS Unit I high density storage racks. This conclusion is based on the following:
(1) Acceptablestate-of-the-artmethodsverifiedbycomparisonwit) experiment were used in the analysis.
(2) Acceptable, conservative assumptions were used in the analysis of the de-sign case.
(3) An acceptable set of uncertainties was considered.
(4) Acceptable abnormal and accident situations were analyzed.
(5) The results meet the staff's acceptance crit'erion of 0.95 for the k-effective value of the racks, including uncertainties.
2.2 Materials i
The safety function of the spent fuel pool and storage rack system is to main-tain the spent fuel assemblies in a suberitical array under all credible storage conditions. The staff ha: reviewed the compatibility and chemical stability of the materials, except the fuel assemblies, wetted by the pool water.
The high density fuel racks are constructed of Type 304 stainless steel, except for the neutron absorber material. The existing spent fuel pool liner is stain-less steel. The high density spent fuel storage racks utilize Borafler sheets as a neutron absorber. Bora11ex consists of boron carbide powder in a rubber-like silicone polymeric matrix. The spent fuel storage rack configuration 3
consists of individual storage cells interconnected to form an integral structure.
The space that contains the Boraflex is vented to the pool. Venting will allow gas generated by the" chemical degradation of the silicone polymer binder during heating and irradiation to escape, and will prevent bulging or swelling of the stainless steel tube.
The pool contains oxygen-saturated demineralized water. The water chemistry control of the spent fuel pool was previously reviewed and found to meet NRC recommendations.
2.2.1 Cor.rosion and Materials Compatibility The pool liner, rack lattice structure, and fuel storage racks are stainless steel, which is compatible with th.e storage pool environment.
In this environ-ment of oxygen-saturated high purity water, the corrosive deterioration of the Type 304 stainless steel should not exceed a depth of 6.00 x 10-5 inch in 100 years, which is negligible relative to the initial thickness.
Dissimilar metal contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, fuel storage tubes, and the Zircaloy in the spent fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore at similar potentials. The Boraflex is composed of noninetsllic materials and therefore will not develop a galvanic potential in contact with the metal components.
Boraflex has undergone extensive testing to determine the effects of gamma irradiation in various environments and to verify its structural integrity and suitability as a neutron-absorbing material. The tests have shown that the Boraflex is unaf'fected by tha pool water environment and will not be degraded by corrosion (Anderson, 1979). During tests performed at the University of 11 Michigan (Anderson, 1981), Boraflex was exposed to 1 x 10 rads of gama radiation with substantial concurrent neutron flux in deionized water.
These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of gamma irradiation.
Irradiation will cause some loss of flexibility, but will not lead to breakup of the Boraflex.
The annulus space that contains the Boraflex is vented to the pool at each storage tube assembly. Venting'of the annulus will allow gas generated by the chemical degradation of the silicone polymer binder during heating and irradia-tion to escape and will prevent bulging or swelling of the inner stainless steel wrapper.
The tests (Anderson, 1979) have shown that neither irradiation, environment, nor Boraflex composition has a discernible effect on the neutron transmission of the Boraflex material. The tests also have shown that Boraflex does not possess leachable halogens that might be released into the pool environment in the presence of radiation. Similar conclusions were reached regarding the leaching of elemental baron from the Boraflex.
Boron carbide of the grade normally in the Boraflex will typically contain 0.1 w% of soluble boron. The test results have confirmed the encapsulation function of the silicone polymer matrix in preventing the leaching of soluble species from the boron carbide.
To provide added assurance that no unexpected corrosion or degr'adation of the materials will compromise the integrity of the racks, the licensees have 4
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committed to conduct a long-term fuel storage cell inservice surveillance pro-gram. Surveillance samples are removable stainless steel clad Boraflex sheets, which are prototypical of the fuel storage cell walls. These specimens will be removed and examined periodically over the expected service life.
2.2.2 Conclusion From the evaluation above, the staff concludes that corrosion of the high den-sity spent fuel racks in the spent fuel storage pool environment will be of little significance during the life of the plant.
Compor.ents in the spent fuel storage pool are constracted of alloys that have a low differential galvanic potential between them and have a high resistance to general corrosion, local-ized corrosion, and galvanic corrosion.
Tests under irradiation and at elevated temperatures in deionized water indicate that the Boraflex material will not 4
undergo significant degradation during the expected service life.
u The staff further concludes that the environmental compatibility and stability of the materials used in the exphnded spent fuel storage pool are adequate on the besis of the test data cited above ar.J actual service experience in operat-ing reactors.
The staff has reviewed the surveillance program n d concludes that the monitor-ing of the materials in the spent fuel storage pool, as proposed by the li-censees, will provide reasonable asrurance that the Boraflex material will con-tinue to parf.orm its function for the design life of the pool. The materials surveillance program delineated by the licensees will reveal any deterioration of the Boraflex that might lead to the loss of neutron-absorbing capability during the life of the spent fuel racks. The staff expects that significant deterioration will not occur.
However, should deterioration occur, this moni-toring program will ensure that the licensees will be aware of it in sufficient time to take corrective action.
The staff, therefore, finds that the implementation of an inservice surveillance program and the selection of appropriate materials of construction by the li-censees meet the requirements of 10 CFR 50, Appendix A, Genir-' Cesign Criterion (GDC) 61, regarding a capability to perniit appropriate periodic inspection and testing of comoonents, and GDC 62, retarding the prevention of criticality by maintaining the structural integrity of components and of the boron neutron absorber and are, therefore, accer table.
2.2.3 References Anderson, J. S., "Boraflex N:. tron Shielding Material--Product Performance Data,"
Brand Industries, Inc., Report 748-30-1, August 1979.
Anderson, J. S., " Irradiation Study of Boraflex Neutron Shielding Materials,"
Brand Indust:-ies, Inc., Report 748-10-1 August 1981.
2.3 Structural Design The staff's evaluation of the high density racks is based on a review performed by NRC's consultant, Franklin Research Center (FRC). The FRC Technical Evalua-tion Report. TER-C5506-579, is appended to this Safety Evaluation as an appendix.
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2.3.1 Description of the High Density Racks and Spent Fuel Pool The new high density racks are stainless steel " egg-crate" structures.
Each 1
cell contains a spent fuel assembly, and a typical rack consists of approximately 300 cells. Weight of'the rack and fuel is transmitted to the floor of the pool through supporting legs.
Each rack is free standing on the pool floor, and a gap is provided between the racks a'nd between the racks and the pool wall so as to preclude impact during earthquake.
GGNS Unit I has an upper containment pool containing fuel storage racks addi-tional to those in the spent fuel storage pool. The upper containment pool is adjacent to the reactor cavity inside the containment. This upper containment pool was designed for temporary storage of spent or new fuel during refueling activities until the fuel could be moved to the spent fuel storage pool or replaced in the reactor vessel during core reload. The spent fuel pool is in the auxiliary buil5f ng and is desi.gned for long-term storage of spent fuel during reactor operation. Both the upper containment pool and the spent fuel pool are reinforcei concrete structures.
2.3.2 Applicable Codes, Standards, and Specifications The staff found that the licensees' load combinations and acceptance criteria i
were consistent with those in the " Staff Position for Review and Acceptance of i
Spent Fuel Storage and Handling Applications" dated April 14, 1978, and amended January 18, 1979. The staff evaluated the existing concrete pool structure for the new loads in accordance with the requirements of Grand Gulf Nuclear Station Final Safety Analysis Report (FSAR) Section 3.8.4, which was approved by the staff during the plant operating license review.
2.3.3 Seismic and Impact Loads Seismic loads for the rack design are based on the original design floor acceleration response spectra calculated for the plant at the licensing stage.
The plant design basis is a 0.075g operating basis earthquake (OBE) and a 0.159 safe shutdown earthquake (SSE).
The seismic loads were applied to the model in three orthogonal directions. Loads resulting from a fuel bundle drop accident were considered in a separate analysis.
The postulated loads from these events were found acceptable.
Further details are provided in the appendix.
2.3.4 Analyses of the Racks and Pool Structures The dynamic response and internal stresses and loads for the racks are obtained from a seismic analysis that is performed in two phases. The first phase is a time history analysis on a simplified nonlinear lumped mass model. The second phase is a stress analysis of a detailed linear three-dimensional finite ele-ment model. The trathodology is discussed further in the appendix. Calculated stresses for the rack components were found to be within allowable limits. The racks were found to have adequate margins against sliding and tipping.
An analysis was conducted to assess the potential effects of a dropped fuel assembly on the racks, and the,results were considered to be satisfactory. An analysis was conducted to assess the potential effects of a stuck fuel assembly causing an uplift load on the racks and a corresponding downward load on the lifting device as well as a tension load in the fuel assembly.
Resulting stresses were found to be within acceptance limits.
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The existing pool structures were analyzed for the modified fuel rack loads using a finite element computer program.
Original plant response spectra and damping values were used in consideration of the seismic loadings.
Design criteria, includirg loading combinations and allowable stresses, are in compli-ance with the Grand, Gulf FSAR, and it has been determined that the existing spent fuel pools can safely support the loads generated by the new fuel racks.
2.3.5 Conclusion On the basis of its review of the structural aspects of the information sub-mitted by the licensees in support of their request to allow installation of high density spent fuel racks in the existing spent fuel pool, the staff con-cludes that the high density spent fuel racks are structurally acceptable.
2.4 Installation of Racks and Load Handling Since GGNS Unit 1 is in its first fuel cycle, the originally installed spent fuel racks were not used for storing spent fuel. The licensees removed the old racks and installed the new high density spent fuel racks in a planned outage in the fall of 1985 in which other work was also accomplished.
The licensees performed a safety analysis pursuant to 10 CFR 50.59 to determine whether the removal of the old racks and installation of the new racks (without placing spent fuel in them) would involve an unreviewed safety question. The licensees concluded that it would not. A license condition pro 11 bits.the stor-age of spent fuel in the spent fuel pool until the standby service water system is modified.
Tnis work on the standby service water system is scheduled to be completed during the first refueling outage.
The rerack was completed using plant procedures for handling heavy loads that were developed from the guidelines in NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants," July 1980.
The licensees' compliance with the criteria of Phase I of NUREG-0612 was found acceptable by the staff in Supplement No. 5 to the Safety Evaluation Report for the Grand' Gulf Nuclear Station (NUREG-0831), dated August 1984, and compliance with Phase II of NUREG-0612 was found acceptable by the staff in its letter to the licensees dated April 4, 1985. The licensees stated that no heavy loads were dropped during the removal of the old racks and instal-lation of the new racks.
For carrying heavy loads over spent fuel in the high density racks, the proce-dures developed from the guidelines of NUREG-0612 are applicable.
In addition, because a postulated drop of the pent fuel pool gate onto the racks containing fuel could damage the fuel, administrative procedures will be used to prevent moving the gate over racks that contain spent fuel.
The staff concludes that procedures for handling heavy loads over the spent fuel stored in the high density racks are acceptable.
2.5 Radiological Consequences of Accidents The review of postulated accidents was conducted according to the guidance of Standard Review Plan Section 15.7.4 (NUREG-0800), NUREG-0554 (" Single-Failure-Proof Cranes for Nuclear Power Plants," May 1979), and NUREG-0612 with respect to accident assumptions.
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2.5.1 Cask Drop Accidents
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The licensees' submittal has indicated that no change would occur to the equip-i ment used in cask hydling or transport operations as a result of the proposed i
spent fuel pool and upper containment pool modifications.
Specifically, the potential cask drop distance is lim.ited to 30 feet and the cask handling crane j
is single failure proof. These cask handling conditions are acceptable without f
calculation of radiological consequences. The staff concludes, therefore, that.
l with respect to a cask drop accident, the assumptions and conclusions reported 1
in Section 15.3.'3 of NUREG-0831, the Grand Gulf Nuclear Station Safety Evaluation Report (SER) dated September 1981, remain valid and no additional analyses are l
necessary for the proposed modification.
f 2.5.2 Construction Accidents 1
Because the old racks were moved and the new racks were installed before any spent. fuel was stored in the originally installed racks, there was no potential for a construction accident involving stored spent fuel.
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2.5;3 Fuel Handling Accidents The licensees have' proposed to expand the storage capacity of the spent fuel pool from 1270 spent fuel assemblies to 4348 spent fuel assemblies and the stor-age capacity of the upper containment pool from 170 spent fuel assemblies to no more than 800 spent fuel assemblies. The maximum weight of the loads.that may be transported over spent fuel in either of the pools is limited to less than i
1140 pounds by Technical Specification 3/4.9.7 and would not be changed by the proposed amendment.
The spent fuel cask handling crane rails do not extend over i
any portion of the spent fuel pool. The proposed license amendment does not, there-fore, increase the radiological consequences of a postulated fuel handling acci-i dent considered in the SER of September 1981 because this. accident would still result in, at most, release of the gap activity of one fuel assembly because of the limitations on available impact k~1netic energy.
2.5.5 Conclusion The staff concludes that the assumptions and conclusions fo' the fuel handling accidents and cask drop accidents presented in the SER dated September 1981 for the originally installed spent fuel racks remain valid for the high density j
spent fuel racks. Therefore, the staff concludes that the high density spent fuel racks are acceptable with respect to fuel handling accidents because the calculated doses for the cask drop and fuel handling accidents remain unchanged j
and are within the NRC dose criterion.-
2.6 Occupational Radiation Exposure 2.6.1 Evaluation The staff has reviewed the licensees' removal and disposal of the low density racks and the installation of the high density racks, with respect to occup,a-tional radiation exposures.
Because the spent fuel pool for GGNS Unit I has never had spent fuel stored in it and is currently clean and uncontaminated, the dose to workers resulting from the spent fuel pool modification itself is estimated to be less than 1 person-rem. Thus, the staff concludes that exposure to workers resulting from the spent fuel pool modification is as low as is reasonably achievable (ALARA) and is acceptable.
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The staff has estimated the increment in the onsite occupational doses result-ing from the proposed future increase in stored fuel assemblies at GGNS Unit 1.
The estimate is based on information supplied by the licensees and on assumed occupancy times and Astimated dose rates in the spent fuel pool area from radio-nuclide concentrations in the spent fuel pool water.
The licensees have de-veloped a loading pattern for the h,igh density spent fuel racks in the spent fuel pool that will maintain occupational dose rates from spent fuel assemblies at less than 2.5 mr per hour. On the basis of present and projected operations in the spent fuel pool area, the staff estimates that the proposed modification should add less than 1% to the total annual occupational radiation dose at the plant. This small increase in the radiation dose in the spent fuel pool area should not affect the licensees' ability to maintain individual occupational
-doses at ALARA levels and within the limits of 10 CFR Part 20.
2.6.2 Conclusion,
The staff finds that storing additional fuel in the Unit 1 spent fuel pool, in accordance with the proposed loading pattern, will not result in any significant increase in doses received by plant personnel and should not affect licensees' ability to maintain individual occupational doses at ALARA levels and within the limits of 10 CFR Part 20.
2.7 Spent Fuel Pool Cooling System This section of the Safety Evaluation deals with the acceptability of.the capa-bility to provide adequate cooling to the spent fuel in the spent fuel pool, the proposed Technical Specifications, and the predicted decay heat generation rates from the spent fuel.
2.7.1 Evaluation The licensees calculated the spent fuel pool decay heat generated in the spent fuel pool from normal refuelings and considering a full core offload in accord-ance with Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling," and Standard Review Plan (NUREG-0800)
Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System." The licensees' calculations showed that the heat generation rate for the spent fuel pool with normal refuelings (normal heat load) is 18.93 MBTU per hour, and the heat gen-eration rate for the full spent fuel pool with the last 800 fuel assemblies a core offload (abnormal heat load) is 4L.81 MBTU per hour.
The staff has per-formed an independent calculation of the two cases, and the results confirm that the licensees have used the appropriate methods for determining the heat generation rates.
The staff also performed an analysis of the spent fuel pool water temperature based on the normal and abnormal heat load conditions. The analysis indicated that the bulk pool water temperature for the normal heat load case would be 171*F, and for the abnormal heat load case the water temperature would be 212 F.
Since the pool water temperature for the normal heat load case is higher than the 140*F pool water temperature identified in Section 9.1.3 of the Standard Review Plan, the licensees have committed, in a June 5, 1986, submittal, to limit the amount of spent fuel stored in the spent fuel pool to a maximum of 2324 spent fuel assemblies. This represents 10 reloads. The spent fuel pool water temperature was calculated for storing 2324 fuel assemblies and using the residual heat removal (RHR) system to remove the decay heat from the pool for the first 35 days following power operation. This is anticipated to 9
represent approximately 30 days following the removal of the reactor head and the beginning of the transfer of the first fuel assembly to the spent fuel pool.' On the basis of limiting the maximum number of spent fuel assemblies to j
2324 and the conunitgent to use the RHR system for the first 35 days, the spent fuel pool water temperature will be less than the 140*F identified in the Standard Review Plan and is, therefore, acceptable.
In the submittal dated June 5,1986, the licensees comitted to propose an ac-ceptable engineering solution to the current inadequacy of the spent fuel pool cooling system to bring the plant into conformance with the Standard Review Plan for the physical limit of the high density spent fuel racks (4348 fuel assemblies) before startup following the third refueling. The licensees also comitted to i'pht the solution before startup following the fifth refuel-ing.
In a submittal dated July 25, 1986, the licensees briefly identified two potential engineering solutions that would provide adequate spent fuel pool coolingsystemcafacity. These solutions were (1) increase the spent fuel pool
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pumping capacity or (2) increase the heat excnanger capacity either by replacing an existing heat exchanger or by adding new heat exchangers.
Either of these solutions appears to be an acceptable approach to increasing the capability of the' spent fuel pool cooling system so that the full capacity of the high density spent fuel racks could be utilized. Such utilization would require a future Technical Specification change in addition to the currently proposed change.
1 The spent fuel pool water temperature, based on the abnormal heat load case, is estimated to be 212 F with the storage of 2324 fuel assemblies and only using the spent fuel pool cooling system.
In this case, one loop of the RHR system is available to provide adequate cooling of the spent fuel pool.
The licensees have proposed a Technical Specification that would limit the maximum spent fuel pool water temperature to 140 F in accordar.ce with Standard Review Plan Section 9.1.3. >In addition, if the water temperature exceeds the 140 F limit, the licensees have committed to restore the water temperature to less than 140 F in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or to begin shutdown of the plant and to achieve hot shutdown %ithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Furthermore, the licensees committed to detemine if the spent fuel pool water temperature could be expected to exceed 210 F within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of i
exceeding the Technical Specification temperature limit of 140*F.
If the water temperature is projected to exceed 210*F within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, the action specified in Technical Specification 3.0.3 would be applicable. This represents the most rapid and orderly shutdown with the minimum transient on the reactor and re-lated systems, in order to achiere safe shutdown before removing one loop of the RHR system from the reactor service mode and placing it into the spent fuel pool cooling mode of operation.
Because the proposed Technical Specifications confom to the Standard Review Plan and the evaluation in the GGNS Safety Evaluation Report, NUREG-0831, the staff concludes that the proposed Technical Specifications are acceptable, q
Proposed Technical Specification 5.6.3 reflects the new physical storage capac-ity of the spent fuel storage facility with'the limitation of a maximum usable storage of 2324 fuel assemblies.
In addition, it reflects the limit of 800 fuel assemblies in the upper containment pool.
In a submittal dated June 9, 1986, the licensees modified the proposed Technical Specifications to require removal of all spent fuel from the upper containment pool before returning the reactor to a critical condition following a refueling. This confoms to the 3
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standard practice of not storing spent fuel inside the containment during reac-tor operation and is, therefore, acceptable.
The spent fuel pool cooling system, with the exception of the cleanup portion, isdesignedtoQualftyGroupCandseismicCategoryIrequirements. The spent fuel pool cooling system can be powered from redundant divisions of the Class 1E power system.
In case of a seismic' event, a seismic Category I bypass line and redundant seismic Category I isolation valves have been'provided at the cleanup system connections to the fuel pool cooling lines to isolate the nonseismic Category I portion of the system to ensure that failure in that portion of the system has no adverse effect on safety-related equipment. This design satisfies the requirements of GDC 2, " Design Bases for Protection Against Natural Phenom-ena." and the guidelines of Regulatory Guides 1.13. " Spent Fuel Storage Facility Design Basis," 1.26, " Quality Group Classification and Standards for Water,
Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants,"
and 1.29, " Seismic' Design Classification."
The nonsafety-related component cooling water system provides cooling water to the fuel pool heat exchangers under normal operating conditions.
Backup cool-ing is provided by the seismic Category I standby service water system (SSWS'),
which transfers the spent fuel pool heat loads to the ultimate heat sink.
Dur-ing testing of the system before the plant was licensed, the licensees deter-mined that the SSWS pumps were undersized and unable to provide design flows for safety-related components and the full spent fuel storage facility. Accord-ingly, License Condition 2.C.(20) was imposed, which prohibits the storage of spent fuel in the spent fuel pool until the SSWS is modified to provide design cooling water flows to all safety-related components, including the spent fuel pool cooling system heat exchangers. This modification of the SSWS will be completed during the first refueling outage. On the basis of its independent analysis, the staff concludes that once the SSWS has been properly modified, 2here should be adequate cooling water flow to the spent fuel pool cooling system heat exchangers to remove the decay heat generated by 2324 spent fuel assemblies with the reloading pattern specified in Table 1.1 of the licensees' May 6, 1985, submittal. Thus, the requirements of GDC 44, " Cooling Water,"
have been satisfied for the storage of 2324 spent fuel assemblies, subject to l
satisfactory modification of the SSWS.
The staff requested that the licensees discuss the redundancy of components 1
so that the spent fuel can be adequately cooled assuming a single active fail-ure concurrent with the loss of offsite power, as specified in the Standard Review Plan.
In a response dated July 25, 1986, the licensees described their 4
procedure to provide alternative cooling for spent fuel in the spent fuel pool and the reactor in the event of a loss of offsite power (LOOP) concurrent with j
a single failure for normal plant operating conditions.
In particular, the licensees addressed the operational conditions of cold shutdown and refueling with only the equipment and systems available that are required by the plant Technical Specifications.
For these two operational conditions, the worst single failure is the failure of the one required diesel generator. The li-j censees committed to revise emergency procedures to include the operator ac-tions necessary in the event of a LOOP with the failure of a diesel generator under cold shutdown or refueling conditions. The operator actions would re-quire the use of the station fire truck to pump water from the fire water stor-l age tanks via fire !oses through stairwells and into secondary containment and i
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I primary containment to maintain the water level in the spent fuel pool and in the reactor. Manual operation of the valves in the spent fuel pool cooling system, the low pressure core spray systems, and the RHR system will allow a controlled flow of yater from the reactor to the suppression pool to prevent bulk boiling in the pools.
The licensees have evaluated the water flow require-ments and have determined that the required flow rate of 720 gpm is well within the capability of the. fire truck's :1000-gpm capacity, and the two fire water 1
I storage tanks, which have a total capacity of 600,000 gallons, will provide more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of cooling for the spent fuel. The licensees committed to j
provide these procedures for~the staff's review and approval before entering i
the first refueling outage. Because the licensees have'a defined method of pro-viding adequate cooling for the spent fuel in the event of a LOOP and a single i
failure and have committed to provide adequate procedures, the staff concludes 3
that the design features of the plant in conjunction with the procedures are acceptable to provide alternative cooling for spent fuel in the spent fuel racks and in the reactor'during refueling operations in the event of a loss of offsite power and the worst single active failure.
2.7.2 Conclusion I
The staff concludes that the spent fuel pool storage capacity modifications are acceptable for the storage of 2324 fuel assemblies (out of a total capacity of 4348 fuel storage locations) with respect to the rack storage capacity, the developed heat loads, the pool water temperatures, and the capability of the spent fuel pool-cooling and support systems.
The staff further concludes that the approach described by the licensees for cooling spent fuel in the spent fuel pool, upper containment pool, and the reactor in the event of a loss of offsite power and a single failure during refueling is acceptable.
2.8 Radioactive Waste Treatment j
GGNS Unit 1 contains radioactive waste treatment systems designed to collect i
and process the gaseous, liquid, and solid waste that might contain radioactive material. The staff evaluated and found acceptable the radioactive waste treat-i i
ment systems in its Safety Evaluation Report (NUREG-0831) dated September 1981, in support of the issuance of the operating license. There is no change in the conclusions regarding the evaluation of these systems because of the use of the l
i high density spent fuel racks. Therefore, the staff concludes that the radio-l active waste treatment systems are acceptable for use with the high density spent fuel racks, i
I 3.0 ENVIRONMENTAL ' CONSIDERATION j
A separate Environmental Assessment has been prepared pursuant to 10 CFR 51.
1 The Notice of Availability of Environmental Assessment and Finding of No Sig-nificant Impact was published in the Federal Register on August 18, 1986
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4.0 CONCLUSION
i The Commission made a proposed determination that the amendment involves no j
significant hazards consideration, which was published in the Federal Register 4
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(51 FR 26078) on July 18, 1986, and consulted with the State of Mississippi.
No public comments were received, and the State of Mississippi did not have any comments.
3, The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in complia.nce with the Commission's regulations and the issu-ance of this amendment will not be inimical to the common defense and security nor to the health and safety of the public.
_P_rincipal Contributors:
M. Lamastra, Plant, Systems Branch, DBL L. W. Bell, Reactor Systems Branch, DPLA F. Witt, Plant Systems Branch, DBL W. Brooks, Reactor Systems Branch, DPLA S. B. Kim, Engineering Branch, DBL J. Ridgely, Plant Systems Branch, DBL L. Xintner, BWR Project Directorate No. 4, DBL Dated: August 18, 1986 13
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APPENDIX
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TECHNICAL EVALUATION REPORT b
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