ML20211B679
| ML20211B679 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/30/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211B676 | List: |
| References | |
| NUDOCS 8606110851 | |
| Download: ML20211B679 (39) | |
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Safety Evaluation By The Office of Nuclear Reactor Regulation Relating to the Reracking of the Spent Fuel Pools At the Diablo Canyon Nuclear Power Plant, Units 1 and 2 As Related to Amendment No. 8 to Unit 1 Facility Operating License No. DPR-80 And Amendment No. 6 to Unit 2 Facility Operating License No. DPR-82, Pacific Gas and Electric Company Docket Nos. 50-275 and 50-323 8606110851 860530 PDR ADOCK 05000275 P
TABLE OF CONTENTS PAGE 1
1.
INTRODUCTION 1.1 Licensee Submittal and Staff Review 1
1.2 License Amendment Request 2
1.3 Summary Description of Reracking 2
2.
CRITICALITY CONSIDERATIONS 3
3.
MATERIAL COMPATIBILITY AND CHEMICAL STABILITY 6
4.
STRUCTURAL DESIGN 9
4.1 Description of Speat Fuel Pool and Racks 9
4.2 Applicable Codes, Standards and Specifications 10 4.3 Loads and Load Combinations 10 4.4 Design and Analysis of Racks 11 4.5 Design and Analysis of Pool Structure 13 4.6 Fuel Handling Accident Analysis 14 4.7 Conclusions 14 5.
INSTALLATION OF RACKS AND LOAD HANDLING 14 6.
SPENT FUEL POOL COOLING SYSTEM 15
.16 7.
SPENT FUEL POOL CL,EANUP SYSTEM 8.
RADIATIONPROTECTiONANDALARACONSIDERATIONS 18 9.
FUEL HANDLING ACCIDENT AND CASK DROP ACCIDENT 20
- 10. RADIOACTIVE WASTE TREATMENT 22
- 11. SIGNIFICANT HAZARDS CONSIDERATION COMMENTS 22
- 12. FINAL NO SIGNIFICANT HAZARDS CONSIDERATION 25
- 13. ENVIRONMENTAL CONSIDERATIONS 31
- 14. CONCLUSIONS 31
- 15. REFERENCES 32 APPENDIX A: Technical Evaluation Report by Franklin Research Center
1.
INTRODUCTION 1.1 Licensee Submittal and Staff Review This report presents the NRC staff safety evaluation for the reracking of the spent fuel pools of the Diablo Canyon Nuclear Power Plant, Units 1 and 2.
By letter dated October 30, 1985 (DCL-85-333, Ref. 1) the Pacific Gas and Electric Company (PG&E or the licensee) submitted, as License Amendment Request LAR-85-13, an application to increase the storage capacity of the spent fuel pools (SFP) for Unit I and Unit 2 of the Diablo Canyon Nuclear Power Plant (Diablo Canyon or the Plant), including the appropriate and necessary changes to the combined Technical Specificttions for Units 1 and 2.
The licensee requested the increase in storage capacity in order to be able to store onsite the spent fuel resulting from approximately 20 years of normal operation of both units.
The application is based on the licensee's " Report on Reracking of Spent Fuel Pools For Diablo Canyon U6its 1 and 2" (the Reracking Report) which was submitted to the staff by letter dated September 19, 1985 (DCL-85-306, Ref. 2).
During its review of the Reracking Report, including discussions at meetings with the licensee and audits of records, the staff requested additional information by letters dated January 8 (Ref. 3), January 15 (Ref. 4),
February 18 (Ref. 5) and February 28, 1985 (Ref. 6). The licensee provided additional infomation supplementing its application by letters of December 20, 1985 (DCL-85-369, Ref. 7), December 24, 1985 (DCL-85-371, Ref. 8),
1 January 28, 1986 (DCL-86-019, Ref. 9), January 28, 1986 (DCL-86-020, Ref. 10),
March 11,1986 (DCL-86-067, Ref.11), April 24,1986 (DCL-86-108, Ref. 37) and April 24,1986 (DCL-86109, Ref. 38).
During its review of the application the staff met with the licensee and representatives of the Joseph Oat Corporation, Camden, New Jersey, the designer and manufacturer of the fuel racks, to discuss in further detail the fuel racks and the fuel pool structure. Meetings were held by the staff with PG&E on December 5, 1985 (Ref. 12), January 8, 1986 (Ref. 13) and February 20, 1986 (Re f. 14 ).- In addition, the staff visited the Joseph Oat Corporation on January 30, 1986 to tour the facilities and observe the manufacturing process l
of the fuel racks (note: as a result of this visit a request for additional infomation was sent to the licensee as Reference 13), and on March 24-25, 1986 (Ref.15) to audit the structural analysis procedurcs and calculation packages l
for the racks and the pool structure. On April 14, 1986 the staff toured the l
Unit 1 fuel handling building at the Diablo Canyon Plant to observe the l
structural layout and features of the fuel pool (Ref. 16).
l This report was prepared by the staff of the Office of Nuclear Reactor Regulation (NRR).
Technical assistance for the structural evaluation of the spent fuel racks and spent fuel pools was provided by the Franklin Research 1
Center, Philadelphia, Pennsylvania. The principal contributors to this report are:
W. Brooks Reactor Systems Branch, PWR-A R. Fell Plant Systems Branch, PWR-A H. Gilpin Plant Systems Branch, PWR-A C. Herrick Franklin Research Center (Consultant)
D. Jeng Engineering Branch, PWR-A F. Rinaldi Engineering Branch, PWR-A H. Schierling Project Directorate #3, PWR-A R. Serbu Plant Systems Branch, PWR-A A. Singh Plant Systems Branch, PWR-A J. Wing Plant Systems Branch, PWR-A The NRC Project Manager for the Diablo Canyon Nuclear Power Plant is Mr. H.
Schierling (Address:
U.S. Nuclear Regulatory Comission, Office of Nuclear Reactor Regulation, Washington, D.C.
20555 - Telephone: 301-492-8856).
i 1.2 License Amendment Request l
The license amendment request (Ref. 1) is for changes to the spent fuel pool of each unit and for the accompanying changes to the combined Technical Specifications for both units. Accordingly, this safety evaluation applies to two amendments, one for Unit 1 Facility Operating License DPR-80, the other for Unit 2 Facility Operating License DPR-82. The request for the amendments, including the staff's proposed "No Significant Hazards Consideration", was noticed in the Federal Register on January 13, 1986 (Ref. 17).
Further details are addressed in Section 11 of this report.
1.3 Sumary Description of Reracking The current spent fuel system is described and analyzed in the licensee's Final Safety Analysis Report (FSAR) Update (Ref. 48), Section 9 and was evaluated by the staff in its Safety Evaluation Report (SER) and supplements (Ref. 30).
The Diablo Canyon Units 1 and 2 have separate, identical spent fuel handling and storage facilities. Both are located in a common fuel handling building, I
separated by the hot shop and parts of the auxiliary building.
In the current design the only component shared by both units is a portable backup pump in the l
2 l
spent fuel pool cooling system as discussed in Section 6.
The evaluation in this report applies equally to both units.
The proposed amendments will allow the licensee to expand the storage capacity of each spent fuel pool (SFP) from the current capacity for 270 fuel assemblies to 1324 assemblies. The expansion will be accomplished by removing the current racks in the SFPs and replacing them with high density racks in which the individual stainless steel cells are more closely spaced. Each cell provides storage for one fuel assembly. The spent fuel storage racks will be arranged in two discrete regions within each pool.
In Region 1 storage racks with poisoned cells will be located, that is, cells surrounded with "Boraflex", a neutron absorbing material.
Region 1 provides for storage for 290 fuel assemblies, which is adequate for approximately one and one-half core (a full core is 193 assemblies).
It will normally be used for storage of fuel assemblies with an enrichment of equal to or less than 4.5 weight percent of Uranium-235 (U-235) at their 'nost reactive point in life, i.e., as new fuel.
I In Region 2 storage racks wif.h unpoisoned cells will be located, that is cells without Boraflex, providing 1034 storage locations for spent fuel assemblies meeting specified burnup conditions.
The existing fuel storage racks have a nominal center-to-center cell spacing of 21 inches. The new storage racks for Region 1 and Region 2 will have a nominal center-to-center cell spacing of 11 inches. The major components of the fuel I
racks are the cells, 31ther poisoned or unpoisoned, gap channels welded between l
the cells, a rack base plate assembly, including support legs, and a girdle bar around the upper part of the rack assembly.
In the case of the poisoned racks, each cell is surrounded by the Boraflex neutron absorber, held in place by a stainless steel cover sheet which also pennits venting of the Boraflex to the pool environment. The fuel racks are assembled as an essentially fully welded construction. There will be a total of 16 fuel racks of various sizes in each pool. The racks are free standing, not connected to the pool floor, walls or to each other. The proposed reracking will not involve major changes to the spent fuel pool structures.
I 2.
CRITICALITY CONSIDERATIONS The current capacity of each spent fuel storage pool is 270 fuel assemblies; the reracking will increase the capacity to 1324 assemblies in each pool. This is accomplished by replacing the present racks with high density racks, i.e.,
reracking. A two-region design has been applied (Technical Specifications, Figure 3.9-1, Ref. 1). Region 1 is designed to accept fresh fuel and will store the fresh fuel assemblies on a 10.93 inch center-to-center cell spacing.
Each storage location will be surrounded on all four sides by a neutron i
absorber material, Boraflex, with sufficient boron to result in an effective multiplication factor, k-effective, of a value of 0.95 or less. This region will have the capacity to store 290 fuel assemblies, approximately one and one-half core. Region 2 has racks of the same design except that no neutron absorber material is used. Storage in Region 2 is restricted to spent fuel 3
assemblies which have achieved a minimum burnup, dependent on the initial enrichment, as discussed below. Space for 1034 spent fuel assemblies is available in Region 2.
The criticality analysis was perfonned by Southern Sciences, a company which has perviously performed such a analyses for other fuel pool expansion applications, including Rancho Seco (Docket No.80-312), Quad Cities Units 1 and 2 (Docket Nos. 50-254 and 50-265), and Virgil Sumer (Docket No. 50-395).
Three different calculational methods were used in the criticality analyses of the spent fuel racks as follows:
(1) The AMPX-NITAWL-KEN 0 code package, which is a multigroup Monte-Carlo calculation using both the 123 and 27 group cross section sets. This is the primary analysis method.
(2) The CASM0-2E code, which is a two dimensional, multigroup transport theory method.
It was used both as a primary calculational tool and as a means of evaluat %g small reactivity effects associated with The statistical nature of the KEN 0 manufacturing (tolerances. item 1 above) makes their use for such effects calculations impractical. The CASMO-2E code was also used for performing the burnup calculations for the fuel.
(3) Two multi-group diffusion codes, PDQ07 and SNEID, were used for the calculation of small reactivity effects that could not be obtained by CASM0-2E because of geometry limitations in that code.
Each code listed above h,as been qualified for use by comparison of its calculated k-effective value to that of critical experiments. On the basis of such comparisons calculational bias and uncertainty values have been obtained for application to the calculated results for the spent fuel racks.
The KEN 0 code was developed by Oak Ridge National Laboratory (Ref. 45).
The code has been widely used by Southern Sciences for the calculation of the reactivity of spent fuel pool racks and has been found acceptable by the staff.
Its use is applicable and acceptable for the Diablo Canyon fuel racks. Use of non-statistical codes such as CASMO-2E, PDQ07 and SNEID for calculation of small reactivity effects is standard industry practice and is acceptable. The CASMO-2E and similar codes are routinely used for obtaining lattice physics parameters as a function of burnup. The licensee's Reracking Report (Ref. 2) gives a comparison of CASMO-2E with several similar codes for the reactivity of fuel assemblies as a function of burnup. The codes agree to within 0.8% for k-effective, with the CASMO code usually being conservative, that is, predicting a higher reactivity. The staff finds that the CASMO-2E code is acceptable for calculating burnup effects.
The following assumptions were made in the calculations to assure conservatism:
(1) The moderator is water at a temperature which yields the highest reactivity.
4
o 0
(2) The storage racks are assumed to b< infinite in extent in all directions.
(3) No credit is taken for absorption in minor structural members, for example in spacer grids.
The staff concludes that the methods used for the analyses are acceptable.
The effect of uncertainties in the Boraflex width, thickness and Boron-10 concentration in conjunction with fuel assembly dimensions, fuel enrichment and density, and off-center position of the fuel in the storage cell were determined. These represent the usual uncertainties considered and are appropriate and acceptable.
For the Region 2 racks an additional allowance for uncertainty in burnup was included.
The analyses result in effective multiplication factors of 0.912 0.008 and 0.932 1 0.006 for the r5cks in Region 1 and Region 2, respectively. These values meet the staff criterion of 0.95 or less for k-effective and, therefore, are acceptable.
The k-effective value for Region 2 was obtained with the CASM0 2E code for fuel having an initial enrichment of 4.5 weight percent U-235 and a burnup of 34,500 MWD /MTU.
In order to determine the required burnup as a function of initial enrichment a series of iterative calculations were performed with CASMO. An initial enrichment was chosen and the burnup required to obtain the target k-effective value was obtained. The resulting curve was extrapolated to obtain the enrichment for which no burnup was required. The k-effective for this enrichment was then calculated by KENO. Comparison of the two calculations showed that the CASM0 calculation is conservative. This procedure is coninonly used for such an analysis and is acceptable.
The effect of abnormal and accident conditions on k-effective included considerations of heatup of the pool water, dropping of an assembly onto the racks, misplacement of an assembly, and lateral motion of the racks.
For Region 1 the effect of a moderator temperature increase is to decrease the reactivity.
For Region 2 an increase in reactivity results and k-effective reaches a value of 0.95 at the boiling temperature after which the reactivity decreases.
In the case of postulated boiling credit may be taken for the presence of soluble boron in the pool which would reduce the reactivity by as much as 0.2.
The potential for pool boiling is discussed in Section 6.
Dropping a fuel assembly onto the top of the racks has a negligible effect on reactivity since the dropped assembly would be separated from the fuel racks by approximately 12 inches of water. Lateral movement of the racks is prevented by design features from reducing the spacing by an amount sufficient to increase the k-effective value above 0.95.
Credit for the presence of boron in the pool water was taken in the analysis of the abnormal location of a fuel assembly in order to ensure the k-effective value does not exceed 0.95. Both the placement of an assembly outside the racks 5
4 and the misloading of a fresh fuel assembly into the Region 2 racks were analyzed.
In both cases the k-effective value was below the staff acceptance criterion of 0.95.
The staff has determined that the analysis of abnormal events is acceptable.
Conclusions The staff has reviewed the licensee's proposed spent fuel pool reracking with respect to criticality and finds that:
(1) Acceptable analysis methods were used which have been qualified for use by comparison with experiment.
(2) Conservative input assumptions were made.
(3) Mechanical and calculational uncertainties were included.
(4) The results meet the staff acceptance criterion of 0.95 for k-effective of the racks.
In sumary, the staff concludes that the criticality analysis for the proposed reracking of the fuel pools of the Diablo Canyon Nuclear Power Plant, Units 1 and 2 is acceptable. The staff further concludes that up to and including 290 Westinghouse design fuel assemblies of enrichment 4.5 weight percent U-235 can be stored in Region 1 of the pool and that up to and including 1034 of exposed assemblies meeting the burnup criteria in the proposed Technical Specifications (Figure 3.9-2) can be safely stored in Region 2.
3.
MATERIAL COMPATIBILITY AND CHEMICAL STABILITY The Diablo Canyon Nuclear Power Plant, Units 1 and 2 each have a facility for the wet storage of spent fuel assemblies. The staff has reviewed the compatibility and chemical stability of the materials wetted by the pool water.
l During its review the staff requested additional information regarding the material compatibility of the rack structure with the spent fuel pool water
,env ronment (Ref. 3). The licensee provided the information in the submittal of i
January 28, 1986 (Ref. 9).
The proposed spent fuel racks are constructed of Type 304 stainless steel, except for the nuclear poison material Boraflex. The spent fuel pool liner also is constructed of Type 304 stainless steel.
Some of the spent fuel storage racks utilize Boraflex sheets as a neutron absorber (Ref. 18) as discussed in Section 2.
Boraflex consists of 42 weight percent of boron carbide powder in a rubber-like silicone polymeric matrix. The spent fuel storage racks are constructed of individual fuel storage cells interconnected by channels to form an integral welded structure. The major components of the fuel racks are listed in Section 1.3.
The spent fuel pools are filled with oxygen-saturated demineralized water containing boric acid.
The fuel pool liner and the fuel rack assemblies, except the Boraflex used in Region 1, are stainless steel which is compatible with the pool water and 6
radiation environment.
In this environment of oxygen-saturated borated water, exceed a depth of 6 x 10 9n of Type 304 stainless steel is not expected to the corrosive deteriorat1 inch per year (Ref. 35), which is negligible relative to the initial minimum thickness of 0.08 inch of the walls of the poisoned cells. Dissimilar metal contact corrosion (i.e., galvanic attack) between the stainless steel of the pool liner or rack assemblies, and the Inconel and the Zircaloy in the fuel assemblies stored in the racks, will not be significant because all these materials are protected by highly passivating i
oxide films which are at similar galvanic potentials.
Boraflex has been used previously as a neutron poison material in a number of spent fuel pool expansions, including Prairie Island Units 1 and 2 (Docket Nos.
50-282 and 50-306) and Oconee Units 1 and 2 (Docket Nos. 50-269 and 50-270).
It has been found acceptable by the staff as a neutron poison material.
For each proposed storage cell assembly, the space between the outside of the fuel cell wall and the cover sheet which contains the Boraflex is vented to the pool. This will allow gas generated by chemical degradation of the silicone polymer binder of the Boraflex during heating and irradiation to escape to the pool and will prevent bulging or swelling of the cover sheet.
Boraflex is composed of non-metallic materials and therefore will not develop a galvanic potential in contact with the metal components. Boraflex has under-gone 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 (Ref. 19). The evaluation tests have shown that Boraflex is unaffected by the pool water environment and will not be degraded by corrosion. Tests g re performed at the University of Michigan, exposing Boraflex to 1.03 x 10. rads of gamma radiation with substantial concurrent neutron flux in borated ~ water. These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gama irradiation.
Irradiation will cause some loss of flexibility, but will not lead to breakup of the Boraflex.
Long-tem borated water soak tests at high temperatures were also conducted l
(Ref. 20). The tests show that Boraflex withstands a borated water immersion I
at 240*F for 251 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide. The spent fuel pool water temperature under normal operating conditions will be approximately 105*F which is well below the 240'F test temperature.
In general, the rate of a chemical reaction decrease expontially with decreasing temperature. Therefore, the staff does not I
anticipate any significant deterioration of the Boraflex at the pool normal l
operating conditions over the design life of the spent fuel racks.
The tests have shown that neither irradiation, environment, nor Boraflex composition have 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 are reached regarding the leaching of elemental boron from +he Boraflex. Boron carbide of the grade normally 7
l present in the Boraflex will typically contain 0.1 weight percent of soluble boron. The test results have confirmed the encapsulation capability of the silicone polymer matrix to prevent the leaching of soluble species from the boron carbide.
To provide added assurance that no unexpected corrosion or degradation of the neutron absorber material will compromise the long-term integrity of the Boraflex, the licensee will conduct a long-term poison coupon surveillance program as described in Section 8 of the Reracking Report. Surveillance samples in the form of removable Boraflex sheets will be exposed to a realistic environment in the pool. They will be periodically removed, and examined for deterioration.
The staff recently identified concerns related to the terminology used for various welding processes, the applicability of the ASME Code Section IX to the welding, and the long-term compatibility of certain structural welds with a spent fuel pool environment (Ref. 39). The staff had discussions with PG&E on May 7, 1986, on these concerns regarding the proposed reracking for the Diablo Canyon spent fuel racks and received additional infonnation frem PG&E (Ref. 40).
Based on the information provided by the licensee these concerns have been acceptably resolved for the proposed reracking.
Conclusions Based on the above discussion, the staff concludes that the corrosion of the spent fuel pool components due to the spent fuel storage pool environment should be of little significance during the life of the facility. Components in the spent fuel storage pool are constructed of alloys which have a low differential galvanic p.otential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion. Tests under irradiation and at elevated temperatures in borated water indicate that the Boraflex material will not undergo significant degradation during the projected service life of approximately 40 years for the racks.
The staff further concludes that the environmental compatibility and stability of the materials used in the spent fuel storage pools is adequate based on the A
test data cited above and actual service experience at operating reactor facilities.
l The staff has reviewed the description of proposed surveillance program for monitoring the Boraflex in the spent fuel storage pools and concludes that the program can reveal deterioration that might lead to loss of neutron absorbing capability during the life of the spent fuel racks. The staff does not anticipate that such deterioration will occur, which would be gradual.
In the unlikely event of Boraflex deterioration in the pool environment, the monitoring program will detect such deterioration and the licensee will have sufficient time to take corrective action, for example, replacement of the Boraflex sheets in Region 1.
The staff, therefore, finds that implementation of the proposed monitoring program and the selection of appropriate materials of construction by the licensee meet the requirements of 10 CFR 50, Appendix A, General Design 8
Criterion 61 regarding the capability to permit appropriate periodic inspection and testing of components and General Design Criterion 62 regarding preventing criticality by maintaining structural integrity of components and of boron poison, and are, therefore, acceptable.
4.
STRUCTURAL DESIGN The staff's evaluation of the structural design aspects of the proposed spent fuel storage expansion is based on a review performed by the NRC staff and the Franklin Research Center (FRC), as NRC consultant. The FRC Technical Evaluation Report TER-C5506-625 is appended to this Safety Evaluation as Appendix A.
The staff accepts the findings and conclusions of the FRC Technical Evaluation Report and incorporates this report as part of its evaluation.
During the course of the review by the staff and its consultant the staff requested additional information (Ref. 3, 5, and 6) which was provided by the licensee (Refs. 7, 9, 11, 37 and 38). Structural design aspects were the subject of two meetings between the NRC and the licensee (Ref.12 and 14).
In addition, as discussed in Section 1.1, the staff and its consultant visited the rack manufacturing facilities (see at Reference 13) and audited structural analysis procedures and calculation packages (Ref. 15) at Joseph Oat Corporation, and toured the Unit I fuel handling building at the Diablo Canyon Plant to observe the structural layout and features of the fuel pool (Ref.16).
4.1 Description of the Spent Fuel Pool and Racks There are two spent fuel pools at the Diablo Canyon Plant, one for each unit.
They are constructed of,'y 58 feet long, and 46 feet deep reinforced concrete. The overall dimensions of each pool are 48 feet wide, b The walls of the pools are 6 feet thick, except near the fuel transfer tube. The foundation slabs have a minimum thickness of 5 feet and are founded on approximately 5 feet of lean concrete that rests on rock strata. The walls and floor of the spent fuel pool are lined with a stainless steel liner 1/8 inch and 1/4 inch thick, respectively. This liner serves only as a water tight boundary, not as a structural member.
The proposed reracking for Diablo Canyon Units 1 and 2 will utilize 16 high-density fuel racks comprising 13 storage rack configuration designs as defined in Table 1 for each spent fuel pool.* The 16 racks are to be arranged
- Tables and Figures mentioned in this Section 4 refer to the FRC Report.
Appendix A.
9
in the pool of each unit as shown in Figures 1 and 2.
An elevation view of a typical rack module is shown in Figure 3.
Fixed and adjustable mounting support legs are provided for each rack module to achieve leveled, freestanding positions on the pool floor (Figures 4-a and 4-b).
l Horizontal and vertical cross sections showing typical construction of a rack module are provided in Figures 5 and 6, which also show the neutron absorbing material Boraflex required for the three (3) rack modules in Region 1 (Table 1). The construction details of the 13 rack modules in Region 2 (Table 1) are the same except for the deletion of the neutron absorbing material.
Clearance between the rack modules and the pool walls is a minimum of 3.0 inches and generally greater for most modules. Clearance between adjacent modules is 2 1/4 inches, which includes 7/8 inch thick girdle bars on each rack module for protection in the event of postulated rack-to-rack or rack-to-wall impacts.
The materials for both Region 1 and Region 2 rack modules are:
stainless steel sheet and plate ASTM A-240-304L forging material ASTM A-182 weld filler material ASME SFA-5.9, Type 308L and 308 LSI Typical module storage cells of both Region 1 and Region 2 racks have an 8.85 inch square cross section to ensure that fuel assemblies with the maximum expected axial bow can be inserted and removed from the storage cells without damage to the fuel assemblies or the rack module.
4.2 Applicable Codes, S'tandards and Specifications The staff has evaluated the acceptance criteria for reracking used by the licensee with respect to those in " Staff Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (Ref. 34). These criteria were previously reviewed and accepted by the staff during other spent fuel pool expansion applications and, therefore, are also acceptable for this reracking design. The design adequacy of the existing Diablo Canyon Units 1 and 2 concrete pool structures was evaluated for the new loads, which result from the proposed fuel racks, in accordance with the design criteria for Category I concrete structures provided in the Diablo Canyon Final Safety Analysis Report (FSAR), Section 3.8.4.
These criteria were previously reviewed during the licensing review stage and were found acceptable.
4.3 Loads and Load Combinations Loads and load combinations considered for the analysis and design of the racks and the pool structure were reviewed and found to be in agreement with the applicable portions of the staff position and the Diablo Canyon FSAR Section 4.2, which was previously reviewed and approved by the staff. Additional details are provided in the FRC Technical Report (Appendix A).
10 l
l
4.4 Design and Analysis of Racks a.
Nonlinear Dynamic Analysis Model The licensee's nonlinear model of a spent fuel rack module for dynamic displacement analysis considers the rack module as a single cell, or stick model on a rigid base with four supports. Additional impact springs that simulate impact with adjacent rack modules and/or the pool walls and the fuel assembly in a rack storage cell are used in the modelling of the racks.
Springs acting through gap elements were used to represent movement of the fuel mass in clearance space before impacting the cell walls.
The licensee's model includes two mass points comprising 8 degrees of freedom.
Mass 1 is located in the rack module and has 6 degrees of freedom, (i.e.,
3-dimensional space with three linear translations and three rotations). Mass 2 is located at the top of the fuel assembly and moves with 2 translational degrees of freedom. The lower mass point of the fuel assembly is lumped with the rack module mass.
The racks' storage cells provide clearance that accommodates possible bowed fuel assemblies. This situation is considered in the licensee's analysis for the impact of fuel assemblies on the storage cells. The analysis accounts for the hydrodynamic coupling effects of the fluid, ignores fluid and structural damping, but considers the effect of the Hosgri earthquake on all of the rack modules. These assumptions provide conservatism in the analyses and design of the racks. To validate the dynamic results from the simpler model, the licensee compared the computed dynamic response of the 8 degree of freedom model to that for a 32 degree of freedom model under the same acceleration time l
history. Both the 8 and 32 degree of freedom modules were based on a Diablo Canyon rack module with a coefficient of friction of 0.2 between the rack module supports and the pool floor liner using zero structural damping. The close agreement between x and y displacement, support loads, and stress ratios (see Section 3.1.4 of Appendix A) indicates that the 8 degree of freedom model based upon the assumption of a rigid body rack module provides satisfactory analysis of the dynamic displacements under seismic excitation, b.
Frictional Forces Between Rack Support and Pool Liner The licensee used a minimum value of 0.2 and a maximum value of 0.8 for the range of static friction coefficients between the rack support pads and the pool liner. This range of friction coefficients is determined to be adequate to bound a realistic friction coefficient based on comparisons with available data (see Reference 5 in Appendix A).
c.
Hydrodynamic Coupling The licensee has used a state-of-the-art analysis to account for fluid coupling. The assumptions consider a rigid boundary between two adjacent racks. This approach does not make a distinction for interaction of non-identical racks. This distinction is judged by the staff to be unnecessary in 11
4 light of the minimal difference in calculated results for identical and nonidentical racks. Furthermore, this effect would be amply accounted for by the fact that conservative assumptions are adopted in other aspects of modelling, analysis and design, d.
Rack loads Seismic loads for the rack design are based on the design acceleration response spectra for the floor calculated for the plant at the licensing stage. This was based on a design earthquake, a double design earthquake (DDE) and the postulated Hosgri earthquake. The Hosgri earthquake (0.75 g) controls the design of the racks and the evaluation of the spent fuel pool. The seismic loads were applied to the model in three orthogonal directions, e.
Analysis Method The n. n-linear dynamic displacement analysis is accomplished with the use of the DYNAHIS computer code as discussed in Appendix A.
This code has been used in the evaluation of previous rack designs.
It utilizes a time-step integration method based upon the central-difference method. The solutions by the use of DYNAHIS have been found acceptable, as stated in the Appendix A.
f.
Results of Rack Analysis The licensee has considered the two limiting racks (10x11 and 6x11 cell configurations) that address the largest size rack and the one with the largest aspect ratio. The results indicate that the largest tipping or rocking response results for the rack with the largest aspect ratio.
The results are present'6d in terms of stress ratios (R) of actual-over-allowable stresses. There are six stress ratios categories, as follows:
Rg = ratio of direct tensile or compressive stress on a net section to its allowable value R2 = ratio of gross shear on a net section to its allowable value R3 = ratio of maximum bending stress due to bending about the x-axis to its allowable value for the section R4 = ratio of maximum bending stress due to bending about the y-axis to its allowable value RS = combined flexure and compressive factor 6 = combined flexure and tension (or compression) factor R
The results were reviewed to determine that they meet the acceptance criteria of less than one for the nonnal condition and less than two for severe accident conditions (see Appendix A Tables 3 and 4).
12
g.
Audit of Rack Stress Analysis The staff and its consultant conducted an audit of calculations and results at Joseph Oat Corporation, the rack manufacturer. The results were found to meet the acceptance criteria. However, some of the results were found slightly lower than reported earlier in the Reracking Report (Ref. 2) submitted to the staff. This is considered acceptable, because it indicates additional margin of safety incorporated in the design. This finding resulted from the combined change in the area of the four support legs for the rack assemblies and from the reduction of the material allowable stress. The fuel rack support legs were enlarged to provide a diameter of 9 inches. The material stress allowables were reduced to account for the reduction factor related to postulated temperature effects.
The results were validated by checking the change between the final and original results. The change in size of support legs would reduce the applicable stress level ratios, while the reduction in the value of allowable stress would increase the applicable stress level ratios. These conditions were verified by checking the results at specific locations in the fuel racks, and were found acceptable.
4.5 Design and Analysis of Pool Structure a.
Pool Structural Analysis The existing pool structure was evaluated for postulated interactions of the rack modules with the structure as a result of a seismic event. This interaction results from the vertical dynamic response between the rack legs and the pool slab, the postulated lateral impacts on the pool walls, other localized bearing loads'on the liner, leak-chase drains and pool slab, and postulated fuel accidental impacts.
In addition, the spent fuel pool structure with the increased loading of the additional spent fuel was analyzed for the previously approved seismic loadings as stated in the FSAR which showed that the increased fuel density did not make a significant change in the response of the pool structure (see Appendix A, Tables 5 through 11).
The vertical dynamic response evaluation considered the two representative modules with regard to weight and aspect ratio (i.e., 10x11 and 6x11 rack modules). The large 10x11 module was evaluated for the potential impact on the pool walls. The impact force would be applied to the wall through the girdle bar as a line load. The other localized loads inc' de bearing and sliding forces affected by the controlling Hosgri earthquake. The resulting stresses are found acceptable because they meet the requirements identifiet in respective design codes:
concrete components ACI 318-63 liner ASME Section III Division 2 1983 For details on the spent fuel pool liner analysis results, refer to Table 12 of Appendix A.
Significant safety margins are incorporated in the design of the spent fuel pool liners by the licensee.
l 13
b.
Cask Pit and Cask Restraint The licensee has comitted to install a removable cask lateral support prior to filling the pools with water and storing the spent fuel (Ref. 37 and 38). The lateral support was designed to prevent a shipping cask from tipping into the spent fuel racks during a postulated seismic event. Also, this restraint will remain in place when spent fuel is stored in the rack modules adjacent to the cask pit. This support will provide protection from a cask impact on the rack modules and at the same time will provide a physical barrier to prevent the spent fuel racks adjacent to the cask pit from falling into the cask pit. The staff finds this acceptable.
4.6 Fuel Handling Accident Analysis Loads due to a fuel assembly drop accident were considered in separate analyses. The postulated loads from these events and the assumptions used in the analyses were found to be acceptable. Additional description and details are provided in the FRC Technical Report (Appendix A). The radiological aspect are addressed in Section 9.
4.7 Conclusions The staff concludes, based on its evaluation of information provided by the licensee, discussions with the licensee at meetings, and information audited by the staff and its consultant, that the licensee's structural analyses of the spent fuel rack modules and the spent fuel pool are in compliance with the acceptance criteria set forth in the FSAR and are acceptable. The analyses indicate the rack modules and pool structure are satisfactory for high density l
fuel storage.
l Therefore, it is concluded that the proposed spent fuel rack installation will satisfy the requirements of General Design Criteria 2, 4, 61 and 62 of 10 CFR 50, Appendix A, as applicable to the structural components.
5.
INSTALLATION OF RACKS AND LOAD HANDLING As of this time no spent fuel has been stored in either of the two fuel pools.
Unit I new fuel assemblies for one core load were stored in the Unit I spent fuel pool prior to the initial loading of the reactor, under water from March 1976 to May 1981 and dry from May 1981 to November 1983; Unit 2 new fuel assemblies for one core load were stored dry in the Unit 2 pool prior to loading the reactor, from May 1977 to May 1985. At the current time both pools are dry and no fuel is stored in either of the pools.
The proposed reracking is to be performed in a dry pool condition, prior to the first refueling of each unit, to preclude any potential contamination from l
spent fuel storage as discussed in Section 8 of this report. The final I
disposal of the existing racks has not been decided by the licensee.
It will be performed in accordance with applicable state and federal requirements.
14
Y The handling of both the old and the new racks fall inta the category of heavy loads as defined in NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants" (Ref. 21).
Installation will be performed consistent with the licensee's previous responses to the NUREG-0612 guidelines.
In SSER-27 of July 1984 and SSER-31 of April 1985 (Ref. 30) the staff concluded that the Diablo Canyon program for the control of heavy loads is in compliance with the guidelines of NUREG-0612. The staff concludes that an installation in a dry and uncontaminated condition will not result in an accident with the potential release of radioactivity.
6.
SPENT FUEL P0OL COOLING SYSTEM The spent fuel pool cooling and cleanup system for each of the two uhits fs designed to re:nove decay, heat generated in the stored spent fuel assemblies and maintain the cleanliness of the spent fuel pool water. The cleanup function of the system is addressed in Section 7.
Ddring its review of the Reracking Report (Ref. 2) the staff discussed with the licensee the sgent fuel pool cooling system, in particular the system operation, redundaricy of components, water makeup sources and heat loads to the po61 for various spent fuel loading conditions in the pool (Ref. 12 and Ref. 13). The staff requested additional infomation in its letter of January 8,1986 (Ref. 3) which was provided by the licensee in two letters of January 28, 1986 (Ref. 9 and Ref. 10).
The current spent fuel pool cooling portion. of the system is a s_ ingle loop with one pump and one heat exchanger. Provisions exist for connecting a portable pump, shared by both units, as a backup to the permanently instalied' pump. The staff discussed this design feature'in detail with the licensee regarding the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 44.
In its letter of January 28,1986 (Ref.10) the licensee committed to permanently install in parallel a s'econd, redundant, 100% capactty pump, in the spent fuel pool cooling system for each of the two units. The arrangement for the portable pump was found acceptable for the current capacity'of 270 fuel assemblies (Ref. 30). The installation of the: additional pump will be completed prior to exceeding the current spent fuel capacity for 270 fuel assemblies. The staff finds this acceptable.,
The spent fuel pool cooling system heat exchanger is cooled by the component l
cooling water system which in turn is cooled by the auxiliary saltwater system which rejects the waste heat to the Pacific Ocean. The fuel pool water is pumped from the pool through the tube side cf the heat exchanger and returned to the pool. The pump suction line is protected by a strainer and is located 4 feet below the normal spent fuel pool water level. The connections'to the spent fuel pool are provided with antisiphon devices to preclude possible draining of the pool water. The piping of the spen.t fuel pool cooling system is arranged so that failure of any pipe will not drain the spent fuel pool below the level required for acceptable radiation shielding. The spent fuel pool cooling system, including the additional pump to be installed, is designed to Seismic Category I criteria and is powered by,a Class IE electrical power system.
15
5 The licensee has calculated the heat load to the pool resulting from a spent fuel discharge in accordance with Standard Review Plan Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System "(Ref. 22) and Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling" (Ref. 23)7 The maximum normal heat load was calculated to be 2.28 x 10 Btu / hour. The maximum normal heat load is for the last normal off-load of 76 fuel assemblies (a normal off-load of 76 assemblies is larger than a one-third core discharge of 64 assemblies). This heat load will result in a maximum bulk pool temperature of 140*F at 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> after the transfer of the last assembly. This meets the the Standard Review Plan acceptance criterionof140*Fforthebulkpoo} temperature. The maximum abnormal heat load was calculated to be 4.38 x 10 Btu / hour. The maximum abnormal heat load is for the offloading of 193 fuel assemblies (i.e., one full-core) into the pool after the last normal offload. This abnormal heat load results in a maximum bulk pool temperature of 174 F which meets the Standard Review Plan acceptance criterion of no bulk pool boiling for this condition.
The licensee has also considered the complete loss of the SFP cooling system for an extended period of time. For this condition natural surface cooling will maintain the water temperature at or below the boiling point.
Boiling would comence after 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> for the maximum normal heat load condition and after 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the maximum abnonnal heat load condition. The corresponding boil-off rates are 22.145 lb per hour and 43,676 lb per hour, respectively.
Four sources of make-up water are available in the event of such complete loss of cooling, two of which meet the criteria for Seismic Class I.
These sources provide adequate make-up for the boil-off rates.
The staff has evaluated the spent fuel pool cooling sy' tem design with respect s
to the criteria of the Standard Review Plan, including the performance of independent calculations, and has found they comply with these guidelines.
Conclusions Based on the above the staff has concluded that the proposed spent fuel pool storage reracking to accomodate the storage of 1324 fuel assemblies in each pool is acceptable with respect to the expected maximum heat loads, pool water temperatures, and the spent fuel pool cooling and support system capabilities.
The staff finds the licensee's commitment to install a redundant pump in the cooling system for each unit acceptable.
- 7. SPENT FUEL POOL CLEANUP SYSTEM The function of the spent fuel pool cleanup system is to maintain water clarity and purity. A portion of the spent fuel pool water can be diverted through the demineralizer, the resin filter, and the mechanical filter (strainer) in the clean-up system. A check valve in the piping to the demineralizer prevents backflushing denineralizer resins to the spent fuel pool. Transfer canal water may also be circulated through the same demineralizer and filter by opening the gate between the canal and the spent fuel pool. This purification loop will remove fission products and other contaminants which could be introduced into 16
s the pool water if a fuel assembly with defective cladding is transferred to the spent fuel pool.
Radioactivity and impurity levels in the water of a spent fuel pool increase primarily during refueling operations as a result of fission product leakage from defective fuel elements being discharged into the pool and to a lesser degree during other spent fuel handling operations. The reracking of the spent fuel pools at the Diablo Canyon Plant will not increase the refueling frequency and fraction of the core replaced after each fuel cycle. Therefore, the frequency of operating the spent fuel pool cleanup system likewise is not expected to increase. Similarly, the chemical and radionuclide composition of the spent fuel pool water will not change as a result of the proposed reracking. Following the discharge of spent fuel from the reactor into the pool, the fission product inventory in the spent fuel and in the pool water One year after the refueling (i.e., after will decrease by) radioactive decay. reactor shutdown, which is less thin the 18 months cycle, the radioactive inventory has decreased by 99% (Ref 41). Furthermore, experience also shows that there is no significant leakage of fission products from spent fuel stored in poois after the fuel has cooled for several months.
Thus, the increased quantity of spent fuel to be stored in the Diablo Canyon fuel pools will not increase significantly the total fission product activity in the spent fuel pool water during the operation of the pools.
The staff has evaluated the information provided by the licensee. Based on this evaluation and its experience with other high density spent fuel storage facilities, including evaluation of operating data, the staff has determined that the proposed reracking of the spent fuel pools at Diablo Canyon will not adversely affect the performance capability or capacity of the spent fuel pool cleanup system. The ra,dioactivity and impurities in the pool water are not expected to increase as a result of the reracking.
Replacement of filters or demineralizers would offset any unanticipated increase of the radioactivity and impurity level of the water in the event of a reduction of the decontamination effectiveness.
The staff has determined that for the proposed fuel storage expansion the existing spent fuel pool clean-up system (1) provides the capability and capacity to remove radioactive materials, corrosion products, and impurities from the pool water, and thus meets the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 61 as it relates to appropriate filtering systems for fuel storage; (2) is capable of reducing occupational exposures to radiation by removing radioactive products from the pool water, and thus meets the requirements of 10 CFR Part 20, Section 20.1(c) as it relates to maintaining radiation exposures as low as is reasonably achievable; (3) confines radioactive materials in the pool water to the demineralizer and filters, and thus meets Regulatory Position C.2.f(2) of Regulatory Guide 8.8 (Ref. 24), as it relates to reducing the spread of contaminants from the source; and (4) removes suspended impurities from the pool water by filters, and thus meets Regulatory Position C.2.f(3) of Regulatory Guide 8.8 (Ref 24),
as it relates to removing crud from fluids through physical action.
17
9 Conclusions On the basis of the above evaluation, the staff concludes that the spent fuel pool cleanup system meets the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 61,10 CFR Part 20. Section 20.1(c) and the appropriate sections of Regulatory Guide 8.8 and, therefore, is acceptable for the proposed reracking of the spent fuel pool.
8.
RADIATION PROTECTION AND ALARA CONSIDERATIONS The staff has evaluated the radiological aspects of the proposed reracking of the spent fuel pool using(the criteria of the Standard Review Plan, Section 12 and Regulatory Guide 8.8 Ref. 24). During its review of the proposed reracking the staff requested additional information regarding radiation protection and ALARA considerations (Ref. 3). The licensee responded in a letter of January 28, 1986 (Ref. 9).
The licensee has proposed to perform the reracking to the Diablo Canyon spent fuel pools with the pools in a dry and radiologically clean condition, that is prior to the first refueling. This will preclude the need for contamination and airborne radioactivity controls, and will result in essentially no occupational dose incurred from the reracking process or from handling radioactive materials. This aspect is also addressed in the Environmental Assessment on the proposed reracking (Ref. 25).
During the design review process for the spent fuel pool reracking effort, the licensee utilized ALARA design considerations consistent with Regulatory Guide 8.8.
These included the following:
(1) ALARA design' review meetings during the design and layout phase.
(2) Identification and incorporation of ALARA design changes for post-reracking operations, such as improvement of cell alignment capability to reduce spent fuel handling time.
(3) Evaluation of concr9te and water shielding to determine dose rate impacts on plant areas and spent fuel pool operations from increased spent fuel storage.
(4) Utilization of utility and consultant personnel experienced in spent fuel pool reracking modifications.
(5)
Incorporation of experience and lessons learned from other utilities in the reracking process and post-reracking operations.
(6) Overall evaluation of the radiological impact on occupational exposure and spent fuel pool and other plant operations.
(7) Evaluation of changes in potential radiological releases to the environment.
18
(8) Evaluation of potential changes in radioactive waste generation and handling at the facility.
(9) Evaluation of the radiological impact of increased fuel storage on the ventila ion system and spent fuel pool water cleanup system capacitv :cd function.
The impact on the or cupational dose from spent fuel pool operations following reracking is expected to be minimal. The licensee has conservatively calculated increases in doses and dose rates which could result from the increased spent fuel storage. The increase in annual occupational dose from spent fuel pool operations for both units is estimated not to exceed 2.4 person-rem per year, and 96 man-rem over the plant life. This is less than 1% of the anticipated occupational dose for the Diablo Canyon facility, and much less than 1% of the overall average dose for PWR's in the United States.
The water level above the spent fuel will remain unchanged at 10 feet during refueling operations and at 23 feet during storage. This continues to meet the staff criteria and requirements for water shielding for spent fuel pools. As discussed in Section 7 the proposed spent fuel pool expansion will not increase the frequency of refueling operations and, therefore, the dose rates associated with this activity will not change as a result of the reracking. Also, as discussed in Section 7, the fission product leakage from defective spent fuel into the pool water is highest at the time of the refueling operations and decreases significantly after the fuel has cooled for several months in the pool. The activity in the spent fuel and in the pool water significantly decreases by radioactive decay during the first year after refueling. There-fore, the dose rates from stored spent fuel and the pool water activity will not change significantly as a result of the proposed reracking. Finally, since there will be no significant increase in spent fuel pool water activity as a result of the reracking there will be no significant increase in the dose rates and doses associated with spent fuel pool water cleanup system filter change outs. Radioactive waste volumes are not expected to change.
The licensee has considered that, as a result of the reracking, some radiation zones next to the spent fuel pool walls could experience temporary dose rates
~ bove the FSAR designated zone dose rates; however, this could occur only if a
freshly discharged fuel elements were placed in certain storage locations close i
to the pool walls and decay is not considered. These affected zones are already designated and controlled as radiation areas and are typically low occupancy areas. The applicable radiological control requirements would remain the same within these zoned areas.
Rezoning these areas would not accurately reflect the nonnal zone radiation levels that will exist after only a short decay period and for most of the plant life.
Since the airborne radioactivity and water activity will not significantly increase, no design or capacity changes in the spent fuel pool ventilation system or spent fuel pool water cleanup systems are needed for radiological reasons. Similarly, the radiation monitoring system for the spent fuel pools as described in the FSAR are adequate and remain unchanged.
19
8 The licensee has comitted that only fuel from the Diablo Canyon Plant will be stored in the spent fuel pools.
Fuel from other facilities will not be received or stored at the facility.
The radiation protection program at the Diablo Canyon Plant incorporates ALARA measures which include procedures and training for radiation protection, application of ALARA engineering controls such as temporary shielding, and ALARA work control procedures. These will be applied to spent fuel pool operations at the Diablo Canyon Plant.
i The licensee also provided a description of contained and airborne radioactivity sources related to the spent fuel pool water which may become airborne as a result of failed fuel and evaporation. The staff has reviewed these source terms and finds them acceptable.
Conclusions Based on its review of the licensee's submittals, the staff concludes that the projected activities and exposure (person-rem) estimates resulting from the fuel pool expansion are reasonable. The licensee has factored ALARA considerations into the design of the reracking and into post-reracking spent fuel pool operations. Previous staff evaluations of the licensee's radiation protection programs and performance as documented in the Safety Evaluation Report and in regional inspection reports indicated that the licensee has the capability to implement adequate radiation protection measures and ALARA dose-reducing activities.
The staff concludes that the licensee will be able to maintain individual occupational exposures within the applicable limits of 10 CFR Part 20, and maintain doses ALARA, consistent with the guidelines of Regulatory Guide 8.8.
Therefore, the radiation protection aspects of the proposed spent fuel pool reracking are acceptable.
9.
FUEL HANDLING ACCIDENT AND CASK DROP ACCIDENT The staff has reviewed and evaluated the proposed spent fuel pool reracking with regard to a postulated fuel handling accident and cask drop accident. The review was performed to identify any change in radiological consequences of these accidents as a result of the reracking.
Calculations of offsite radiological co1 sequences due to a fuel handling accident (i.e. the dropping of a spent fuel pool assembly onto stored spent fuel assemblies) were performed using the standard assumptions given in Regulatory Guide 1.25 (Ref. 26) and Standard Review Plan Section 15.7.4 (Ref. 27). Regardless of how many spent fuel assemblies are stored in the pool, all fuel rods in one entire, freshly discharged fuel assembly were assumed to be breached, releasing their entire gap activity in accordance with conservative assumptions ir. Regulatory Guide 1.25.
The assumption that all fuel rods in one assembly rupture is conservative because the kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop 20
height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly. The accident evaluation conservatively assumes all rods in one entire assembly rupture and release their gap activity.
Because the energy that causes the rupture of the fuel rods is fixed, impacting other fuel assemblies will not increase the total number of fuel rods ruptured.
In addition, the spent fuel racks protect the stored fuel assemblies from impact and absorb some of the kinetic energy from the dropped assembly. Therefore, the assumption that one full assembly releases its entire gap contents during a fuel handling accident is not affected by the proposed spent fuel pool expansion.
In the analysis of a fuel handling accident the staff postulates the rupture of all fuel rods in one full, freshly discharged assembly (i.e., no decrease in radioactivity due to decay after removal from the reactor) and the subsequent release of the entire fuel clad fission product gap activity from the assembly.
This assumption remains unchanged for the proposed spent fuel pool expansion.
As stated by(the licensee in a response to a staff request for additionalRef. 9) the ne information initial enrichment (up to 4.5% U-235) and longer burnup times than the current licensing evaluation basis of 25,000 MWD /MTU for the average burnup, as stated in the FSAR, Section 15.5.22. The licensee, accordingly, performed the fuel handling accident analysis for a U-235 enrichment of 3.5% and 4.5% and for a burnup of 30,000 MWD /MTU and 50,000 MWD /MTU.
The licensee has provided further information (Ref. 49) regarding the application of Regulatory Guide 1.25 (Ref. 26) assumptions in calculating the consequences of a postulated fuel handling accident involving spent fuel w':h an extended burnup of up to 50,000 MWD /MTU for Diablo Canyon Units 1 and 2.
The additional information is largely based on the Westinghouse Topical Report
~
WCAP-10125 (Ref. 46) which supports operation of Westinghouse fuel beyond 50,000 MWD /MTV with accumulated gap activities of less than the values in Regulatory Guide 1.25. The staff has reviewed the Westinghouse Topical Report (Ref. 47) and concludes that it is acceptable for the Diablo Canyon analysis.
The staff has reviewed the additional information and concludes, based on the maximum rod average linear heat generation rate of 8.43 kw/ft, that the licensee's basis for using the gap activity consistent with Regulatory Guide 1.25 is acceptable for the analysis of the fuel handling accident involving l
extended burnup fuel. The offsite doses remain well within the guideline l
values of 10 CFR Part 100 and meet the acceptance criteria of Standard Review Plan Section 15.7.4 (Ref. 27). The offsite doses calculated in accordance with l
staff methodology will not significantly change as a result of increasing the i
number or burnup of assemblies stored in both pools.
The cask drop accident was also reviewed. The licensee has proposed administrative controls by Technical Specifications which would preclude the movement of a spent fuel shipping cask in an exclusion zone over and in the vicinity of stored spent fuel that could result in a cask drop or tipping accident damaging stored spent fuel. The change to Technical Specification 3/4.9.13 prohibit cask handling operations near the spent fuel pool while fuel is stored in the spent fuel cask exclusion zone.
21
h O
Conclusions The proposed reracking will not change the consequences of a fuel handling accident from those previously reported in the Diablo Canyon Safety Evaluation Report. The administrative controls proposed to be placed on cask movement will prevent a cask drop or tipping accident that could impact spent fuel assemblies.
For these reasons, the staff finds the proposed reracking and associated changes to the Technical Specifications acceptable.
- 10. RADI0 ACTIVE WASTE TREATMENT The Diablo Canyon Plant includes a separate radioactive waste treatment system for each of the two units, designed to collect and process gaseous, liquid and solid waste that might contain radioactive material. The radioactive waste treatment system was evaluated by the staff in the Final Environmental 1973 (Ref. 29) and in the Safety Evaluation Report Statement (FES), dated May(Ref. 30), in support of the issuance of operating (SER), dated October 1974 licenses for the two units. There will be no change in the waste treatment systems as a result of the proposed reracking. Further details of the waste treatment are addressed in Section 2 of the Environmental Assessment (Ref. 25) on the proposed reracking.
- 11. SIGNIFICANT HAZARDS CONSIDERATION COMMENTS The licensee's request for these amendments was individually noticed on January 13, 1986 (Ref. 17) followed by a bi-weekly notice on May 21, 1986 (Ref. 44). Separate comments, request for a hearing, and a petition for leave to intervene were filed by (1) the Mothers for Peace (Ref. 31), (2) Consumers Organized for Defense of Environmental Safety (C0 DES), and (3) the Santa Lucia Chapter of the Sierra Club (Sierra Club) (Ref. 33). The relevant comments of these groups are identified as M L and S respectively and addressed below.
M-1: Mothers for Peace state "By creating more storage capacity, the radioactive inventory will be greatly increased and would therefore, present a significant hazard to our comunity in the event of an accident caused by an earthquake."
The highest levels of radioactivity in a spent fuel pool occur imediately after the offloading of spent fuel elements. As discussed in Section 7, after one year of storage the initial radioactivity inventory has decreased by 99%
due to radioactive decay (Ref. 41). There would be a gradual buildup over the potential storage period of long lived radioisotopes in the inventory stored in the pool.
However, this is principally activity contained in the fuel matrix within the cladding.
Potential accidents have been evaluated (see Section 12) and there are no sources of failure of the pool or its cooling system that could provide a niechanism for the dispersal of the fuel pool inventory. The slight increase in inventory of long lived fission products has no effect on the potential for any such accident, and in the absence of such potential has no effect on consequences to the public. The only radioactive gas of significance due to the storage of additional spent fuel would be the noble gas Krypton-85 (Kr-85), which has a half life of 10.8 years (Ref. 25). However, experience 22 I
./
has demonstrated that after spent fuel has been moved from the reactor to the spent fuel pool and has cooled for 4 to 6 months, the gaseous fission products, including Kr-85, have been released from the stored spent fuel with cladding detects. The release of gaseous fission products, including Kr-85, from non-defective fuel is insignificant in comparison with the overall releases from routine plant operations.
The staff and its consultant have evaluated the seismic design of the proposed j
changes to the spent fuel pool and storage of additional fuel assemblies as discussed in Section 4 of this report and summarized in Section 12 for the seismic event. The racks and the pools were designed to seismic Category I requirements and meet the structural acceptance criteria and seismic design bases for Diablo Canyon. The staff concludes that the spent fuel pool and spent fuel racks have been designed to withstand the postulated seismic loadings and, therefore, there will be no significant hazard resulting from a postulated seismic event.
M-2: Mothers for Peace refer to the past seismic design evaluation for Diablo Canyon and to the currently ongoing Long Term Seismic Program (LTSP) for the plant.
In summary, it is stated that, "Without a complete seismic record, the margin of safety could be seriously jeopardized, and thus reracking would again pose a significant hazard."
In April 1984 the Comission included in the Diablo Canyon Unit I low-power license a condition requiring the licensee to develop ar implement a program d
to reevaluate the seismic design bases for Diablo Canyon (Ref. 36). The background for this condition was addressed in SSER-27 of July 1984. The staff concluded that there is r.o reason to modify previous conclusions on the seismic design bases while the. program was being carried out. Additional details were provided in the license condition when restated in the Unit 1 full-power license in November 1984 by including four specific elements. The applicability of the long term seismic program (LTSP) to Diablo Canyon Unit 2 wr.s addressed in SSER-31 of April 1985.
The LTSP was established to reevaluate the seismic design bases for Diablo Canyon, taken into consideration all relevant infomation that has become available since the review by the ACRS in 1978 (Ref. 42). The program was not required because the staff questioned the adequacy and applicability of the current seismic design bases. This position was a basis in the staff's determination that low and full power licenses be issued for both units, a view which was confimed by the Commission itself (Ref. 43). The current seismic design basis for the Diablo Canyon Plant is therefore appropriate for the proposed reracking. The staff concludes that it is not necessary to complete the LTSP prior to the reracking.
S-1: Sierra Club does not agree with the licensev s evaluation that the proposed reracking does not involve a significant increase in the consequences of a seismic event.
23
D This concern is the same as raised by Mothers for Peace, in part, and is addressed at H-1, second paragraph. The staff concludes there will be no significant change in the consequences resulting from the postulated seismic event.
S-2: Sierra Club does not agree with the licensee's evaluation that the proposed reracking does not involve a significant increase "in the consequences of a loss of spent fuel cooling."
The staff has evaluated the loss of spent fuel pool cooling as discussed in Sections 6 and 12 of this report. The staff has concluded that the consequences will not be significantly increased form those previously evaluated.
5-3: Sierra Club disagrees with the licensee's finding the "the proposed reracking does not involve a significant reduction in a margin of safety."
The staff has evaluated the change in margin of safety as a result of the proposed reracking as discussed in detail in Section 12 of this report. The staff has concluded that the reracking does not result in a significant reduction in a margin of safety with respect to criticality, cooling or structural considerations.
C-1:
C.0.D.E.S. believes that the License Amendment Request "does involve a significant increase in the consequences previously evaluated."
The licensee has evaluated certain accidents and events (Ref. 1). TM staff has reviewed and evaluated the information provided by the licensee and has concluded, as discussed,in Section 12 of this report that the selection of accidents and events was appropriate and that the consequences of previously analyzed accidents and events will not be significantly increased as the result of the proposed expansion of the spent fuel pool capacity and storage of spent fuel assemblies.
C-2:
C.0.D.E.S. believes that the License Amendment Request "does create the
. possibility of a new or different kind of accident from any accident previously evaluated."
The licensee has evaluated the possibility of a new or different kind of any accident previously evaluated (Ref. 1). The staff has reviewed and evaluated the information provided by the licensee and has concluded, as discussed in Section 12 of this report, that the proposed reracking does not create the possibility of a new or different kind of accident from any previously evaluated for the Diablo Canyon spent fuel storage facility.
C-3:
C.O.D.E.S. believes that the License Amendment Request "does involve a significant reduction in the margin of safety."
This concern is the same as raised by Sierra Club and is addressed at S-3 above.
24
l 7
C-4:
C.0.D.E.S. states that "the licensee has neglected entirely the part of human error or foibles may play in the probability or consequences of an accident involving the re-racked spent fuel ponds."
The high density spent fuel racks are being manufactured by the Joseph Oat Corporation in accordance with the company's quality assurance program and control procedures. This process was audited by the licensee. The spent fuel pool modification and installation of the new racks will be perfomed in accordance with the licensee's quality assurance program. Operation of the spent fuel pool including loading and unloading of fuel elements will be performed in accordance with the Technical Specifications and plant operating procedures. The licensee has evaluated a spent fuel element drop accident and spent fuel cast tipping and drop accident, which are not expected to occur, but is postulated as a result of equipment design, mechanical or human error. As discussed in Section 9 the consequences will not be changed as a result of increasing the number of assemblies stored in the pool. The licensee also evaluated the misplacement of a fresh fuel assembly into the unpoisoned Region 2 and the placement of an assembly outside the racks as discussed in l
Section 2.
In both cases the k-effective value was below 0.95 and the staff determined the analysis to be acceptable.
4 12.
FINAL NO SIGNIFICANT HAZARDS CONSIDERATION The licensee's request for amendments to the operating licenses for Diablo Canyon Units 1 and 2, including a proposed determination by the staff of no significant hazards consideration was individually noticed in the Federal Register on January 13, 1986 (Ref. 17) followed by a bi-weekly notice on May 21, 1986 (Ref. 44). This is the staff's final detemination of no significant hazards consideration.
I The Commission's regulations in 10 CFR 50.92 include three standards used by the NRC staff to arrive at a detemination that a request for amendment (s) involves a no significant harards consideration. These regulations state that the Comission may make such a final detemination if op(eration of a facility inaccordancewiththeproposedamendment(s)wouldnot 1) involve a i
significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
j The proposed spent fuel pool expansion amendments are similar to more than 100 earlier requests from other utilities for spent fuel pool expansions. The majority of these requests have already been granted by the NRC, others are under staff review. The knowledge and experience gained by the NRC staff in reviewing and evaluating these similar requests were utilized in this evaluation. The licensee's request does not use any new or unproven technology in either the analytical techniques necessary to support the expansion or in the construction process.
The staff has determined that the licensee's request for amendments to expand 4
I the spent fuel pool storage capacity for Diablo Canyon Units 1 and 2 by 25 i
~
reracking to allow closer spacing of spent fuel assemblies does not significantly increase the probability or consequences of accidents previously evaluated; does not create new accidents not previously evaluated; and does not result in any significant reduction in the margins of safety with respect to criticality, cooling or structural considerations.
The following staff evaluation in relation to the three standards demonstrates that the proposed amendments for the SFP expansions do not involve a significant hazards consideration.
First Standard
" Involve a significant increase in the probability or consequences of an accident previously evaluated."
The following postulated accidents and events involving (spent fuel storage had been identified previously by the licensee in the FSAR Ref. 48) and were evaluated by the staff (Ref. 30) during the operating license review.
a (1) Spent Fuel Assembly Drop Accident (2) Spent Fuel Shipping Cask Drop Accident (3) Loss of Spent Fuel Pool Cooling (4) Seismic Event (5) Tornado Generated Missile
~
(6) Criticality Accident (7) Reracking Installation The licensee has considered these accidents and events as part of the reracking amendment request (Ref. 1) and the Reracking Report (Ref. 2). The staff has evaluated the accidents and events as discussed below.
In general, as discussed in Section 7 and elsewhere in this report, the increase in the number of spent fuel assemblies to be stored in the spent fuel pool (i.e., from the current 270 assemblies to the proposed 1324 assemblies) will not result in a corresponding increase in radioactive fission product inventory in the stored spent fuel assemblies or in the pool water throughout the life of the plant.
The total increase in the number of stored spent fuel elements by approximately a factor of five will not be reached until about the year 2007 through successive normal offloads of 76 assemblies at each discharge and one full-core discharge. The release of fission product gap activity from defective fuel elements into the fuel pool decreases significantly after the elements cool down for several months.
Furthermore, the radioactivity decreases by about 99%
due to decay within, bout one year after reactor shutdown. Therefore, most of l
the spent fuel storeo in the pool will be aged and will not contribute signifi-cantly to the total radioactive inventory of the spent fuel or the pool water.
26
The major contribution to the total inventory of radioactivity at any time is the most recently discharged spent fuel.
1.
Spent Fuel Assembly Drop Accident The spent fuel assembly drop in a spent fuel pool is evaluated in Section 9 of j
this report. The frequency of refueling operations will not change as a result of the proposed reracking. Therefore, the probability of occurrence for this accident will not increase. As with the existing rack configuration, it was conservatively assumed that the equivalent of all rods in one spent fuel assembly would be breached releasing the entire gap activity. There exists the potential for an increase in offsite doses as a result of an increase in fuel burnup. However, as discussed in Section 9, such increase would be insignifi-cant. The staff concludes that there is no significant increase in the con-sequences of this accident.
2.
Spent Fuel Shipping Cask Drop Accident l
The spent fuel shipping cask drop is addressed in Section 9 of this report.
The implementation of the proposed Technical Specifications will preclude the movement of the spent fuel cask over any spent fuel and would prevent cask handling operations near the spent fuel pool while fuel is stored in the spent fuel cask exclusion zone. Therefore, the probability of occurrence for this accident will not increase significantly as a result of the spent fuel pool expansion. The postulated dropping or tipping of the cask outside the exclusion zone will not impact any spent fuel assembly. Therefore, the consequences of this accident will not increase from those previously evaluated and reported in the FSAR.
3.
Loss of Spent Fuel
- Pool Cooling The spent fuel pool cooling system is addressed in Section 6 of this report with respect to its redundancy, water makeup sources, postulated maximum heat loads, and the potential for boiling. The licensee has comitted to replace the portable second pump, currently shared by the spent fuel pool cooling systems for Units I and 2, by installing a permanent second pump in each system. This will increase the reliability of the system. The expansion of the spent fuel pool capacity will not increase the probability of a loss of 1
spent fuel pool cooling previously evaluated by the staff in Section 9.3.2 of the SER (Ref. 30). The most severe consequences of a loss of spent fuel pool cooling would occur with the full-core offload when the pool is nearly filled as discussed in Section 6.
The increase in consequences will be an increase in the pool bulk water temperature which will remain below or at the boiling point. Adequate source of water are available to make-up for evaporation losses. The staff concludes that the increase in spent fuel stored in the pools will not significantly increase the consequences of loss of the spent fuel pool cooling. The licensee also considered the offsite radiological consequences in the event that, in addition to the postulated loss of spent fuel cooling system, all water make-up sources are postulated to be lost.
In this case pool boiling and evaporation would occur, however, the radiological consequences are insignificant.
l 27
. - -- = -__. _ _. _ _
1 4.
Seismic Event The structural design of the spent fuel rack assemblies and the spent fuel pool structures with respect to a postulated seismic event have been evaluated as discussed in Section 4 and Appendix A of this report. The spent fuel pool expansion does not increase the prob 6bility of such an event. The racks and the pools were designed to seismic Category I requirements. The results of the structural analyses show that the racks and the fuel pool structure meet the structural acceptance criteria for the Diablo Canyon Plant. The sr.ismic loads resulting from a seismic event will not result in a failure of the racks or pool structure, thus their integrity will be maintained. The staff concludes that there will be no significant change in the consequences resulting from a i
postulated seismic event from those previously determined.
5.
The spent fuel pool expansion does not increase the probability of occurrence i
i of a tornado, the probability of generating a missile or the probability of such missile striking the fuel pool as previously evaluated by the staff in Section 3.3 of its SER (Ref. 30). Similarly, the consequences of a postulated tornado generated missile impacting the fuel pool structures have been previously analyzed by the licensee and evaluated by the staff. The spent fuel storage pools have adequate protection against tornado forces and tornado generated missiles. The proposed reracking does not affect the evaluation.
The consequences of tornado generated missiles will not be significantly
{
increased from those previously determined.
i 6.
Criticality Accident l
Considerations regardin'g a criticality accident have been evaluated in Section 2.
The effect of various abnormal and accident conditions on criticality were considered. With the inclusion of administrative controls in accordance with the proposed amended Technical Specifications to maintain the boron concentra-i tion in the spent fuel pools at a minimum of 2000 ppm, and to limit the storage in Region 2 spent fuel racks to spent fuel assemblies based on initial enrich-ment and cumulative exposure, none of the postulated conditions will result in a criticality accident. There is no significant increase in the probability of l
a criticality accident due to the proposed reracking.
The change to the two-region spent fuel pool required the performance of an additional evaluation to ensure that the criticality criterion, k-effective less or equal to 0.95, is maintained. This included the evaluation of the limiting criticality condition, caused by dropping or misplacing of an unirradiated fuel assembly of 4.5 weight percent U-235 enrichment into a Region 2 storage cell (i.e., unpoisoned) or outside and adjacent to a Region 2 rack j
module. The evaluation for this case showed that, for the required boron concentration of 2000 ppm in the fuel pool water in accordance with the
{
proposed Technical Specifications, the criticality criterion is maintained and there is no significant increase in the consequences of this accident.
i 28 1
e
_--n_-._
---~.__,__.,.m-_,
e..
7.
Reracking Installation The staff evaluation of the proposed spent fuel pool expansion by removing the existing racks and installing the new racks is presented in Section 5 of this report. The reracking is performed prior to the first refueling in a dry and uncontaminated pool. Therefore, the staff concludes that the probability of an accident involving the relaasa -# radioactivity and its consequences are insignificant.
In sumary, therefore, baseo on the above discussion, the probability or consequences of previously analyzed accidents and events for Diablo Canyon Units 1 and 2 will not significantly increase as the result of the proposed expansion of the spent fuel pool capacity and storage of spent fuel assemblies.
Second Standard
" Created the possibility of a new or different kind of any accident previously evaluated" The proposed reracking of t b spent fuel pools has been evaluated in accordance with the design bases specified in the Diablo Canyon FSAR, the guidance contained in NRC position paper "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (Ref. 34), appropriate NRC Regulatory Guides and Standard Review Plans,(and appropriate industry Codes and Standards as listed in the Reracking Report Ref. 2) and listed in this report.
In addition, several previous NRC safety evaluations for similar spent fuel pools expansions at other nuclear facilities have been considered. No unproven techniques and methodologies were utilized in the analysis and design of the proposed high density facks and the reevaluation of the spent fuel pool structure. No unproven technology will be utilized in the construction and installation process of the new racks. The basic reracking technology in this instance has been developed and demonstrated in numerous applications for a fuel pool capacity increase which have already received NRC staff approval.
As a result of the safety evaluation in this report and based on its evaluations of similar spent fuel pool expansions, the staff concludes that the e
proposed reracking does not create the possibility of a new or different kind of accident from any accident previously evaluated for the Diablo Canyon spent fuel storage facilities.
Third Standard
" Involve a significant reduction in margin of safety".
The NRC staff safety evaluation review process has established that the issue of " margin of safety " when applied to a spent fuel pool modifications will need to address the following areas:
(1) nuclear criticality considerations, 4
29
(2) thermal-hydraulic considerations, (3) material, structural and mechanical considerations.
The established acceptance criteria used to assess the adequacy of facilities assure maintenance of the necessary margins of safety. This safety evaluation by the staff addresses the three areas identified above.
The margin of safety that has been established for nuclear criticality considerations is that the effective neutron multiplication factor (k-effective) in the spent fuel pool is to be less than or equal to 0.95, including all reasonable uncertainties and under all postulated conditions. As noted in Section 2 of this report, the criterion is met for all normal and abnormal conditions for the storage of spent fuel in the proposed configuration. The proposed amendments, therefore, do not significantly reduce the margin of safety for criticality.
The criteria used to evaluate the margin of safety with respect to thermal-hydraulic considerations of the storage of spent fuel are the methodologies.
assumptions, and requirements identified in the staff's Standard Review Plan (SRP) Section 9.1.3 (Ref. 22), including Branch Technical Position ASB 9.2 (Ref. 23), to assure that the fuel pool temperature does not exceed the criteria of 140'F for nonnal reload conditions (i.e., unloading of 76 assemblies) and 212*F for abnormal reload conditions (i.e., full-core discharge). As noted in Section 6 of this report these criteria are met for the normal reload conditions and the abnormal full-core discharge conditions for the bulk pool temperatures. The proposed reracking, therefore, does not significantly reduce the margin of safety for spent fuel cooling with respect to thermal-hydraulic aspects.
The criteria used to evaluate the margin of safety (with respect to material, l
structural and mechanical considerations are that 1) the compatibility and chemical stability of the materials wetted by the pool water be demonstrated I
and no significant corrosion occur, and (2) the structural and mechanical design of the spent fuel pool and the storage racks maintain the fuel assemblies in a safe configuration for all environmental and abnonnal loadings using the codes, standards and specifications identified in Section 4 of the report.
As noted in Section 3 of this report, the corrosion that will occur in the spent fuel pool environment will be of little significance for the life of the plant and the environmental compatibility and stability of the materials used is adequate based on test data and actual service experience in operating reactors. As noted in Section 4 and Appendix A of this report, the structural and mechanical design of the spent fuel pools and storage racks can withstand the environmental and abnormal loadings and the pool structure can sustain the increased floor loadings with adequate margin. The proposed reracking of both pools, therefore, does not significantly reduce the margin of safety with regard to materials, structural, and mechanical integrity.
30
As a result o'f the review and evaluation by the staff as reported in this document, the staff concludes that the proposed reracking of the Diablo Canyon Units 1 and 2 spent fuel pools to expand the capacity of the pools does not result in a significant reduction in a margin of safety with respect to criticality, cooling or structural considerations.
Sumary In sumary, based on the foregoing and the fact that the reracking technology in this instance has been well developed and demonstrated, the Comission has concluded that the standards of 10 CFR 50.92 are satisfied. Therefore the Comission has made a final determination that the proposed amendment for spent fuel pool expansion does not involve a significant hazards consideration.
- 13. ENVIRONMENTAL CONSIDERATIONS A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51. The Notice of Issuance of Environmental Assessment and Finding of No Si nificant Impact was published in the FEDERAL REGISTER on May 29,1986(Ref.
25.
- 14. CONCLUSIONS The staff has reviewed and evaluated the licensee's request for amendments for the Diablo Canyon Nuclear Power Plant Unit I and Unit 2, operating licenses regarding the expansion of the spent fuel pools. Based on the considerations discussed in this safety evaluation the staff concludes that:
(1) these amendments will not (a) significantly increase the probability or consequences of actidents previously evaluated, (b) create the possibility of a new or different accident from any previously evaluated of (c) significantly reduce a margin of safety and, therefore, the amendments do not involve significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endagered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public.
Dated: May 30, 1986 i
31
f
- 15. REFERENCES 1.
PG8E Letter No. DCL-85-333, October 30, 1985 from D. A. Brand (PG&E) to H.
R. Denton (NRC),
Subject:
License Amendment Request 85-13. Reracking of Spent Fuel Pools.
2.
PG&E Letter No. DCL-85-306, September 19, 1985 from J. D. Shiffer (PG&E) to G. W. Knighton (NRC),
Subject:
Report on "Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2."
3.
U.S. Nuclear Regulatory Comission, letter dated January 8,1986, from H.
E. Schierling (NRC) to J. D. Shiffer (PG8E),
Subject:
Spent Fuel Pool Reracking, Request for Additional Information.
4.
U.S. Nuclear Regulatory Comission, letter dated January 15, 1986 from H. E. Schierling (NRC) to J. D. Shiffer (PG8E),
Subject:
Spent Fuel Pool Reracking, Request for Additional Information.
5.
U.S. Nuclear Regulatory Comission, letter dated February 18, 1986 from H. E. Schierling (NRC) to J. D. Shiffer (PG8E),
Subject:
Spent Fuel Pool Reracking, Request for Additional Information.
6.
U.S. Nuclear Regulatory Comission, letter dated February 28, 1986 from H. E. Schierling (NRC) to J. D. Shiffer (PG&E),
Subject:
Sumary of Meeting with PG8E held on February 20, 1986 on Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2, including a Request for Additional Information.
i 7.
PGE8E Letter No. D.CL-85-369. December 20, 1985 fromJ.D.Shiffer(PG&E) to S. A. Varga (NRC),
Subject:
Reracking of Spent Fuel Pools - Peak Decay Heat Loads and Water Temperatures.
8.
PG&E Letter No. DCL-85-371, December 24, 1985 from J. D. Shiffer (PG&E) to S.A.Varga(NRC),
Subject:
Reracking Reference Material.
9.
PG&E Letter No. DCL-86-019. January 28, 1986 from J. D. Shiffer (PG&E) to S. A. Varga (NRC)
Subject:
Spent Fuel Pool Reracking - Additional Information.
10.
PGAE Letter No. DCL-86-020, January 28, 1986 from J. D. Shiffer (PG&E) to S. A. Varga (NRC),
Subject:
Supplement to the Spent Fuel Pool Reracking l
Report - Spent Fuel Pool Cooling System.
11.
PG&E Letter No. DCL-86-067, March 11, 1986 from J. D. Shiffer (PG8E) to S.
A. Varga (NRC),
Subject:
Response to Questions on Spent Fuel Racks.
U.S. Nuclear Regulatory (Comission January 6,1986, from H. E. SchierlingSu 12.
(NRC)toJ.D.Shiffer PG8E),
Subject:
I on December 5, 1985 on Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2.
l l
32 l
- 13. U.S Nuclear Regulatory Comission, January 29, 1986 from H. E. Schierling (NRC) to J. D. Shiffer (PG&E),
Subject:
Sumary of Meeting with PG8E held 8, 1986 on Detailed Control Room Design Review and Three Other on January (Note: Spent Fuel Pool Reracking was one of the other three Subjects subjects).
14.
U.S. Nuclear Regulatory Comission, February 28, 1986, from H. E. Schierling (NRC) to J. D. Shiffer (PG8E),
Subject:
Sumary of Meeting with PG&E held on February 20, 1986 on Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2.
U.S. Nuclear Regulatory (Comission, May 9,1986, from H. E. SchierlingSumar 15.
(NRC) to J. D. Shiffer FG&E),
Subject:
and 25, 1986 at Joseph Oat Corporation, Camden, New Jersey.
23, 1986, from H. E. Schierling U.S. Nuclear Regulatory (Comn.ission, May Sumary of Site Visit on Spent 16.
(NRC) to J. D. Shiffer PG&E),
Subject:
Fuel Pools on April 14, 1986.
- 17. Federal Register Notice, Vol. 51, No. 8,1451, January 13, 1986, Pacific Gas and Electric Corrpany; Consideration of Issuance of Amendments to Facility Operating Licenses DPR-80 and DPR-82 for Diablo Canyon Nuclear Power Plant, Units 1 and 2, Respectively, and Proposed No Significant Hazards Detennination and Opportunity for Hearing.
18.
J. S. Anderson, "Boraflex Neutron Shielding Material Product Perfomance Data." Brand Industries. Inc., Report 748-30-1, August 1979.
19.
J. S. Anderson, " Irradiation Study of Boraflex Neutron Shielding Materials," Brand'1ndustries, Inc., Report 748-10-1, July 1979.
20.
J. S. Anderson, "A Final Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials," Brand Industries, Inc., Report 748-21-1 August 1978.
21.
U.S. Nuclear Regulatory Comission, Report NUREG-0612. " Control of Heavy i
Loads at Nuclear Power Plants, Resolution of Generic Technical Activity A-6," July 1980.
22.
U.S. Nuclear Regulatory Comission, Report NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"
Rev. 1. July 1981 Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup Systems."
23.
U.S. Nuclear Regulatory Comission, Branch Technical Position ASB 9-2,
" Residual Decay Energy for Light-Water Reactors for Long-Term Cooling",
included in Reference 22, Section 9.2.5, " Ultimate Heat Sink".
24.
U.S. Nuclear Regulatory Comission, Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear i
33 O
a r - - - -. _ - -
- -. ~,,, -, -. -. _ -,, - - - -,. - - -. - -
i Power Stations Will Be As Low As Is Reasonably Achievable," Rev. 3 June 1978.
21, 1986 from H. E. Schierling U.S. Nuclear Regulatory (Comission, May 25.
(NRC) to J. D. Shiffer FG&E)
Subject:
Environmental Assessment and Finding of No Significant Impact - Spent Fuel Pool Expansion, Diablo Canyon Nuclear Power Plant, Units 1 and 2.
Also:
Federal Register Notice, 51 FR 19430, May 29, 1986.
26.
U.S. Atomic Energy Comission, Regulatory Guide (Safety Guide) 1.25
" Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 1972.
27.
U.S. Nuclear Regulatory Comission, Report NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"
July 1981, Section 15.7.4, " Radiological Consequences of Fuel Handling Accidents."
28.
U.S. Nuclear Regulatory Comission, Report NUREG-0575, " Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel," Volumes 1, 2 and 3, August 1979.
29.
U.S. Atomic Energy Comission, " Final Environmental Statement Related to the Nuclear Generating Station Diablo Canyon, Units 1 and 2 " May 1973.
30.
U.S. Atomic Energy Comission, " Safety Evaluation of the Diablo Canyon Nuclear Power Station, Units 1 and 2" NUREG-0675, October 1974 and Sepplements No. I through No. 33, as applicable.
- 31. Mothers for Peace to U.S. Nuclear Regulatory Comission, Secretary of the Comission, letter dated February 7,1986.
- 32. Consumers Organized for Defense of Environmental Safety (C.O.D.E.S.) to U.S. Nuclear Regulatory Comission, Secretary of the Comission, letter dated February 12, 1986.
- 33. Sierra Club - Santa Lucia Chapter to U.S. Nuclear Regulatory Comission, Secretary of the Comission, letter dated February 10, 1986, 34.
U.S. Nuclear Regulatory Comission, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978, and
" Modifications to the OT Position," January 18, 1979 letters from B. K.
j Grimes (NRC) to All Power Reactor Licensees.
l 35.
E. G. Brush and W. L. Pearl, " Corrosion and Corrosion Product Release in 1
Neutral Feedwater, " Corrosion, Vol. 28, p. 129 (April 1972).
I 34 l
36.
U.S. Nuclear Regulatory Comission, " Amendment No. 9 to Facility Operating License DPR-76", Letter from D. G. Eisenhut (NRC) to J. O Schuyler (PG&E),
dated April 18, 1984.
- 37. PG&E Letter No. DCL-86-108, April 24, 1986 from J. D. Shiffer (PG&E) to S. A. Varga (NRC),
Subject:
Additional Infonnation on Spent Fuel Racks.
- 38. PG&E Letter No. DCL-86-109, April 24, 1986 from J. D. Shiffer (PG&E) to S. A. Varga (NRC),
Subject:
Supplemental Information Regarding Spent Fuel Racks.
39.
U.S. Nuclear Regulatory Comission, April 25, 1986, from T. V. Wambach (NRC) to K. W. Berry (Consumers Power Company),
Subject:
Expansion of Spent Fuel Storage Capacity at Pallisades Plant.
- 40. PG8E Letter No. DCL-86-126, May 9, 1986 from J. D. Shiffer to S. A. Varga (NRC),
Subject:
Supplement Information on Spent Fuel Racks.
41.
U.S. Nuclear Regulatory Comission Information Report SECY-83-337, dated August 15, 1983, from Executive Director of Operations to Commissioners,
Subject:
Study of Significant Hazards.
42.
U.S. Nuclear Regulatory Comission, July 14, 1978, from S. Lawroski, Chairman - ACRS, to J. Hendrie, Chairman - NRC,
Subject:
Report on Diablo Canyon Nuclear Power Station Units 1 and 2.
43.
U.S. Nuclear Regulatory Comission (Diablo Canyon Nuclear Power Plant, Units 1 and 2):
a.) Memorandum an'd Order CLI-84-5, dated April 13, 1984, 19 NRC 953, 961-2(1984).
b.) Decision CLI-84-12, dated August 10, 1984, 20 NRC 249, 251 n.3 (1984),
c.) Memorandum and Order CLI-84-13, dated August 10, 1984, 20 NRC 267, 276(1984).
- 44. Federal Register Notice, Vol. 51, No. 98, 18699, May 21, 1986, Pacific Gas and Electric Company; Consideration of Issuance of Amendments to Facility Operating Licenses DPR-80 and DPR-82 for Diablo Canyon Nuclear Power Plant, Units 1 and 2, respectively, and Proposed No Significant Hazards l
Detennination and Opportunity for Hearing.
45.
L. M. Petrie and N. F. Cross, " KENO-IV, An Improved Monte Carlo Criticality Program." ORNL-4938, Oak Ridge National Laboratory, November 1985.
- 46. Westinghouse Electric Corporation, Report WCAP-10125 (proprietary and non-proprietary version), " Extended Burnup Evaluation of Westinghouse Fuel," July 1982, i
35
47.
U.S. Nuclear Regulatory Commission, Letter from C. O. Thomas (NRC) to E. P. Rahe (Westinghouse), dated October 11, 1985,
Subject:
Acceptance for Referencing of Licensing Topical Report WCAP-10125 (P), " Extended Burnup Evaluation of Westinghouse Fuel."
- 48. PG8E Lette, No. DCL-85-308, September 20, 1985 from J. D. Shiffer (PG&E) to H. R. Denton (NRC),
Subject:
FSAR Update Rev. 1.
- 49. PG8E Letter No. DCL-86-149, May 28, 1986 from J. D. Shiffer (PG&E) to S. A. Varga (NRC),
Subject:
Spent Fuel Pool Reracking Additional Information.
1 l
l l
1 36 a
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APPENDIX A TECHNICAL EVALUATION REPORT FRANKLIN RESEARCH CENTER DATED APRIL 30, 1986 6