DCL-86-019, Applicant Exhibit A-3,consisting of Responding to NRC 860108 & 15 Ltrs Requesting Addl Info Re Plant Spent Fuel Pool Reracking Rept Submitted to NRC on 850919

Applicant Exhibit A-3,consisting of Responding to NRC 860108 & 15 Ltrs Requesting Addl Info Re Plant Spent Fuel Pool Reracking Rept Submitted to NRC on 850919

ML20237J094
Person / Time
Site:	Diablo Canyon
Issue date:	06/17/1987
From:	Shiffer J PACIFIC GAS & ELECTRIC CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
DCL-86-019, DCL-86-19, OLA-A-003, OLA-A-3, NUDOCS 8709030492
Download: ML20237J094 (59)

Text

h -j 75 32 3 ~0Y l 6[/7[f7' PGandE Ex No. 3 h PACIFIC OAS AND E LE C T RI C ' .C

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• 26 P 3 :27 x; O M PANY 500 9 084/3 l 77 BEALE STREET . $AN FRANCISCO. CALIFORNIA 94106 . (415) 78142111..' .TWX 910 372 6587 awee o pesa

"'""""****^'"'" January 28, 1986

'PGandE Letter No.: DCL-86-019  ;

Mr. Steven A. Varga, Director PHR Project Directorate No. 3 Office of Nuclear Reactor Regulation i

U. S. Nuclear Regulatory Commission

Hashington, D.C. 20555 Re

Docket No. 50-275, OL-DPR-80 q Docket No. 50-323, OL-DPR-82 Diablo. Canyon Units 1 and 2 Spent Fuel Pool Reracking - Additional Information i

Dear Mr. Varga:

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In letters dated January 8 and January 15, 1986, the NRC Staff requested additional information regarding the spent fuel pool reracking report for Diablo Canyon Units 1 and 2 which was sent to the Staff.on-September 19, 1985 (DCL-85-305). The specific items and PGandE responses are included in l Enclosure 1 to this letter. Items were also identified in a meeting on i December 5,1985 between PGandE and the NRC Staff and summa.-ized in the NRC letter report dated January 6, 1986. These items and PGandE responses are included in Enclosure 2. -

Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.

Sincerely, iffer Enclosures (2) cc: L. J. Chandler R. T. Dodds R. C. Herrick (FRC)

J. 8. Martin {

B. Norton i N H. E. Schierling l s CPUC l

\ Diablo Distribution i l

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i PGandE Letter No.: DCL-86-019 ENCLOSURE 1

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PCandE RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORWJ1QM SPENT FUEL POOL RERACKING I

i Item 1:

The information in the Reracking Report regarding the radiological consequences for routine operations and certain accident scenarios is  !

qualitative (e.g., page 7-3 regarding doses; page 7-4 regarding filter and.

resin replacement; page 7-9 regarding plant man-rem, page 7-15 regarding off-site doses). Provide a quantitative comparison of the radiological consequences for the existing spent fuel storage and the proposed reracking of the spent fuel pool.

Resnonse 1:

The radiological 1' consequences of the proposed reracking of the spent fuel pool O are summarized below under the headings of occupational exposures, radwaste generation, environmental exposures resulting from normal operation, and envi vnmental exposures resulting from postulated accidents.

Occupational Ernosures There are four sources contributing to occupational exposures resulting from storage of spent fuel. These sources are: (1) exposures outside the fuel pool shield walls; (2) exposures above the fuel pool;.(3) exposures due to handling radwaste generated as a result of spent fuel storage and; (4) exposures due to shipment of spent fuel. The projected increases in occupational exposures are conservatively estimated not to exceed 2.4 man-rem per year and 96 man-rem over the life of the plant. This annual increase is less than 1% of PGandE's 1986 dose goal of 250 man-rem. The exposures noted  :

are the combined doses for both units and assume dry installation of the new )

spent fuel racks. l The projected increases are due entirely to exposures outside the fuel pool shield walls. These exposures are based on the dose rates tabulated in Table 7.3 and discussed in Section 7.6.2 of the Reracking Report. The following conservative assumptions were made in calculating the increased exposures:

the dose rates are based on storing freshly discharged assemblies along the shield walls and no credit for radioactive decay of the spent fuel has been taken. The 2.4 man-rem is the total calculated exposure outside the shield walls.

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i O The occupational exposures. received by personnel above'the pool are not 1 expected to change because the dose rates and occupancy times are not expected

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to change from present fuel handling practices. Pool surface dose rates are a function of the radiation from the stored assem'ulies, the activity.in the pool water, and the radiation from the assembly being handled during fuel transfer.

The maximum dose rate at the surface of the pool from the stored assemblies for the proposed reracking is 4.34 x 10-8 arem/hr. This dose rate constitutes a negligible contribution to exposure at the pool surface.

The radioactivity in the pool water results from three sources: (1) the introduction of reactor coolant during refueling operations; (2) the dislodging of crud from the fuel assemblius; and (3) the release of fission l products from the stored fuel. The introduction of reactor coc>1 ant is a function of refueling operations, and it will not change because of the .

reracking. The dislodging of crud from the fuel assemblies occurs during fuel handling. During the life of the plant, fuel handling operations will not increase because of the raracking, so no additional crud is introduced to the ,

1 pool. This conclusion is reached in NUREG-0575, Final Generic Environmental Impact Statement on Handling and Storage of Spent Light _ Water Power Reactor Fuel. NUREG-0575 also documents experiencs at GE's Morris Plant and at the NFS New York plant that shows little leakage of radioactivity from spent fuel which has been cooled several months. These plants store spent (power l reactor) fuel in water, and the experience at these plants is applicable to l Diablo Canyon. Because spent fuel that has been cooled for several months has l q- little leakage of radioactivity, the raracking is not expected to increase the amount of fission products released to the fuel pool. In conclusion, the amount of radioactivity in the fuel pool water is not expected to change over the life of the plant because of the reracking. The spent fuel pool cleanup system is designed to maintain the dose rate at the surface of the pool below 2.5 ares /hr.

During fuel transfer operations, the dose rate at the surface of the pool is increased by the radiation from the assembly being transferred. The maxima dose rate at the pool surface from this source has been calculated to be 6.9 arem/ hour. This dose rate is not a function of storage and is, therefore, not affected by the raracking.

As discussed in our response to Item 24 of this enclosure. PGandE does not project an increase in radwaste generation over the life of the plant as a result of the reracking. Also, because no increase in the pool water radioactivity is expected, there will be no increase in the dose rates around the filters and domineralizers. Therefore, there will be no increase in the exposures associated with radwaste handling because of_the reracking.

Exposures due to spent fuel shipment will be reduced because aged spent fuel rather than fuel recently discharged from the reactor will be shipped. This exposure savings has not been analyzed, and no credit has been taken for it.

Radwaste Generation No increase in radwaste generation over the life of the plant is projected.

This is fully discussed in our response to Item 24 of tnis enclosure.

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.O Environmental Ernosures from Normal Onoration Krypton 85 is the principal radioactive gas which, because of its long half-life, might be released as a result of additional spent fuel assemblies being stored in the spent fuel pool over extended periods.of time.

Pertinent experience at operating facilities has been documented, for. example, <

by the NRC in the Environmental Impact Appraisal for the reracking of Oconee Nuclear Station, Units 1, 2, and 3 (Amendments 123, 123, and 120;.Section 3.2 of the EIA, September 29, 1983); Rancho Seco Nuclear Generating Station (Amendment 52, Section 4.0 of the EIA, January 20, 1984); and Tro3an Nuclear Plant (Amendment 88. Section 3.0 of the EIA, June 8, 1984). This experience has shown that after spent fuel has decayed several months, there is no longer a significant release of fission products, including Kr-85, from stored fuel with cladding defects. This means that all, or most, of the Kr-85 i released from the spent fuel is released before the next load of spent fuel enters the pool. Thus, the release of Kr-85 from the spent fuel pool is not significantly increased by continuing to store spent fuel assemblies in the pool beyond the next refueling period. The proposed raracking of the spent fuel pool will not, therefore, significantly increase Kr-85 releases from the spent fuel pool over the life of the plant. i In addition, the amount of Kr-85 released from the spent fuel pool is very small compared with the total amount of Kr-85 released from the operating reactor, as shown in Table 11.3-3 of the FSAR Update. This is due to the much lower migration rate of the Kr-85 out of the fuel pellet matrix and through

/ fuel cladding defects at the lower temperature of the spent fuel pool compared with the temperature in the operating reactor. Increasing the number of storage locations in the spent fuel pool from 270 to 1,324 locations represents an increase of 4.90 in the amount of spent fuel to be stored in the pool. However, the amount of Kr-85 estimated to be released from the spent fuel pool, as shown in the FSAR Update, is orders of magnitude less than the total. amount of Kr-85 released from the operating reactor. .Even considering a factor of 4.90 increase in the quantity of Kr-85 released from stored fuel, the increase is only a minor fraction of the total Kr-85 released from the plant. This increase does not add significantly to the annual environmental dose, which is still less than the guideline values for such releases.

Environmental Ernosures from Accidents The design basis fuel handling accident is presented in the response to Item 9.

The radiological dose consequences of a postulated spent fuel pool boiling l event are presented in Section 7.7 of the Retacking Report on pages 7-11 through 7-15. The results, presented in Table 7.7, are a small fraction of l 10 CFR 100 guidelines and are considered acceptable. l O

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Item 2:

Provide the changes in projected annual doses and plant life doses (increases ,

or decreases) resulting from the proposed reracking.

Resnonse 2:

The' projected dose increases for both Units are conservatively estimated not to exceed 2.4 man-rem per year and 96 man-rem for the plant life. Increases in environmental doses from normal operations are discussed in the response to Item 1.

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O Item 3:

Verify that no changes in the plaat radiation zoning as identified in the FSAR are necessary as a result of the reracking. -

4 Resoonse 3:

Section 7.6 of the Reracking Report discusses the radiation shielding l evaluation that was performed for the spent fuel pool. The increased storage capacity with the new storage location geometry along with the effects of  !

increased enrichment and longer burnup of the fuel have been taken into  :

account. Results are provided in Section 7.6.4. The stated doses are the highest dose rates achieved outside the pool wall across from the midpoint of the active fuel region just after the assemblies have been placed along that  ;

particular wall. These are transient dose rates since they decay with time I and occur locally adjacent to where spent fuel has recently been placed. The dose rates from the recently stored fuel will reduce by approximately a factor of 3 after one month's storage and by a factor of 10 after one year. In  :

addit'. 9, the areas affected have a low expected personnel occupancy time, so l

the s' ects of the higher dose rates are minimal, as discussed in Response 1. l 4

The areas below the operating deck around the spent fuel pool are zoned at '

less than or equal to 1 arem/ hour, except for the fuel pool pump and heat exchanger rooms, which are zoned at less than or equal to 15 arem/ hour. The  ;

higher dose rates discussed in the reracking report are the maximum expected i in these areas immediately after placement of fresh spent fuel in the storage O locations closest to the pool walls. Revising the FSAR zone maps to reflect

.these temporary worst case conditions would result in the maps incorrectly identifying a higher than anticipated radiation level in these locations over i the vast majority of the plant's lifetime. The current radiation zoning  !

better reflects the normally anticipated dose rates in these areas.

Additionally, the transient elevated d se rates given in the reracking report  ;

are not expected to adversely affect plant operation, due to the low occupancy i times of personnel in these areas. If an elevated dose rate did become a i problem, operational fixes are available (e.g., shuffling of aged spent fuel to perimeter locations, installation of temporary shielding, etc.).

Table 7.4 of the Reracking Report provides results of the analysis for a '

single peak power assembly which is being transferred in the pool. It assumes the minimum allowed water shielding of 10 ft of water to the top of the active fuel. The maximum dose at the pool surface is 6.9 mrem / hour, but this reduces to 1.8 mrem / hour 10 ft above the surface of the pool. Typically, there will be more than 10 ft of water over the active fuel, and additional decay time will occur as fuel is being transferred. This local transient dose rate is acceptable and does not warrant a change to the radiation zone of less than or equal to 2.5 mrem / hour.

06705/0040K -___ __ _ ___ -

Item 4:

Verify that, as a result of the raracking, no changes to the ventilation 1 I

-system and fuel pool water cleanup system, such as shielding, are necessary for radiological reasons. Verify that the systems as described in the FSAR remain unchanged.

ResnonseJ:

A brief description of the existing ventilation and pool cleanup system is )

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' provided below.

A. Fuel Handlina Buildino Ventilation System The fuel handling area heating and ventilating system has the capability to provide ventilation air for the feel handling area separately'from the rest of the auxiliary building. The fuel handling area for each unit is physically

. isolated from the rest of the auxiliary building. The system consists of.

redundant supply and exhaust fans, and redundant HEPA and charcoal filter banks. A third set of full-capacity exhaust fans and HEPA filter bank trains is provided for normal operation. Each HEPA filter bank is preceded by a roughing filter bank. . The supply airflow, 29,850 cfs, .was selected on the-basis of the heat dissipated by the equipment. The exhaust air flow (35,750 cfe) consists of 28,700 cfm exhausted from the spent fuel pool area by drawing the air flow over the pool, and the remaining 7,050 cfm is ducted and exhausted by the exhaust fan from other areas in the fuel handling building.

O The prime function of the fuel handling area heating and ventilation system is to remove radiolytic gases from the surface of the spent fuel pool and to' treat the exhaust air in order to remove most of these gases. The purpose of the treatment of the exhaust air is to reduce the offsite dose to acceptable levels in the event of a fuel handling accident.

8. Snent Fuel Pool Cleanun System A portion of the spent fuel pool water may be diverted away from the heat exchanger through the spent fuel pit domineralizer, the spent fuel pit resin filter, and the spent fuel pit filter to maintain water clarity and purity.. A check valve in the piping to the demineralized prevents backflushing domineralizer resins to the spent fuel pool. Transfer canal water may also be circulated through the same demineralized and filter by opening the gate between the canal and the spent fuel pool. This purification loop is sufficient for removing all anticipated fission products and other contaminants which could be introduced if a fuel assembly with defective cladding is transferred to the spent fuel pool.

The spent fuel pool demineralized will limit the exposure rate at the surface of the pool due to radioactivity in the water to less than 2.5 mrem / hour. The i

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O exposure rate at the pool surface is routinely monitored with radiation surveys and monitoring equipment.

As discussed in the response to Item 1 above, the reracking will not result in increased radiological effects that would necessitate a modificati6n to the ventilation system nor the fuel pool water cleanup system. These systems will remain as described in the Diablo Canyon FSAR Update.

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O Item 5:

Regarding ALARA requirements and practices discuss the following:

~a. How the ALARA design review was conducted consistent with Regulatory Guide 8.8. Discuss the documentation of the review and provide examples.

b. How the experience of other individuals (non-PG&E) and at other facilities was utilized regarding post-modifications operations with respect to ALARA.

c. The Diablo Canyon ALARA program that will be applied during the spent fuel pool operations.

Resnonse Sa:

The design review was conducted consistent with the guidelines of Regulatory Guide 8.8. During the design review of the new racks and rack layout, ALARA considerations were an integral part of the review. Prist to finalizing the layout and design, a design review meeting was held at the Diablo Canyon Plant, with participation from project engineering, plant engineering, operations, construction, and regulatory compliance groups. This meeting and others caused ALARA-related design changes such as aligning of all storage cells and adding peripheral lead-in edges to the racks to help reduce handling time when transferring fuel. A shielding evaluation was performed which j confirmed that dose rates in the vicinity of the pool resulting from the increased storage capacity were acceptably low. During this evaluation, relatively high radiation streaming was identified through the fuel pool transfer gate to the operating deck when the fuel transfer canal is drained and spent fuel is stored near the gate. The design of the rack adjacent to this gate was modified to provide for miscellaneous equipment storage rather than spent fuel storage near the gate. This design change eliminated the potential for high doses from stored spent fuel to the operators while the transfer canal is dry.

Coordination of design and drawings with operations and plant staff personnel occurred throughout the rack design and pool layout process. This effort is documented on transmittal forms and memos.

In addition, project procedures call for an ALARA evaluation as part of the review and signoff of design change packages (DCP), and the reracking will be implemented through a DCP.

Resnonse 5b:

Contact with other plants was made during the entire reracking engineering process, starting with development of a design and fabrication specification for the racks. PGandE's consultants included several individuals with direct experience related to fuel racks and fuel handling operations. Many design 0

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b requirements incorporated into the specifications to facilitate rack l installation and fuel storage and handling are also in61rectly related to

- ALARA considerations, such as provisions for remote 1eteling and placing of fuel racks into a wet pool which contains spent fuel 1.1 the event that the reracking must be done after the first refueling outage.

'l Power plants that have completed spent fuel pool reracking projects (Nine Mile l

Point, Zion, Farley) were contacted to obtain information on operational experience. Additionally, published information (such as from NUREG-0575) provided insight into the magnitude of radiological changes expected from the reracking. Available information points toward minimal operational impact on -

radiation exposure to plant personnel. I Resconse Sc:

The ALARA program that is currently in place at Diablo Canyon will also be applied to spent fuel pool operations. The program includes Nuclear Plant Administrative Procedures for rad ution protection and for training, Administrative Procedures for operational ALARA implementation, temporary shielding, and Radiation Control Procedures for work permits. The program also includes regular meetings of the Joint ALARA Review Committee and the Plant Staff ALARA Committee. The p'iant's ALARA program was inspected by the NRC's Region V Office on August 26 to 30, 1985, and found acceptable. (Ref.

IE Inspection Report Nos. 50-275/85-30 and 50-323/85-28).

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Item 6:

Verify that spent fuel from other facilities will not be received, shipped or stored at the Diablo Canyon Plant.

Resnonse 6:

No spent fuel from other facilities, including those owned and operated by PGandE, will be received or stored at the Diablo Canyon Power Plant.

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. Discuss in more detail the features and operation of the spent fuel ' shipping ~

cask washdown area. ,

Resnonse 7:

The cask decontamination area is located on elevation 115 ft-adjacent to the spent fuel pool. -This area is bounded on three sides by 25 ft high concrete walls an'd a curb. The concrete walls have been sealed to minimize contamination. The decontamination area base and portions of the walls are lined with stainless steel, with a curb' provided to prevent the water and/or solvent used during decontamination from spreading over the building floor.

Drains in the floor of the area remove the contaminants to the radwaste' disposal system for processing.

During shipment of spent fuel offsite, the spent fuel shipping cask it first unloaded from its transporter inside the fuel' handling building.using the fuel handling area overhead bridge crane. If required, the cask may be washed in i the decontamination area prior to placing in the pool. The cask is lifted over the corner of the pool and placed into the fuel cask pit recessed area of the fuel pool floor. After the cover of the shipping cask is removed, spent fuel assemblies are removed from the storage racks and loaded into.the cask using the fuel handling bridge crane hoist. .The cask is then covered and

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lifted out of the pool and placed in the decontamination area. The outside surfaces of the cask are decontaminated prior to shipment by using O, high-pressure water and/or solvent. After appropriate radiation surveys, the cask is then returned to the transporter ready-for shipment.

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.i Item 8:

Discuss the materials compatibility of the'retack structures with the spent' fuel pool water environment. .

Resnonse 8:

The new spent fuel racks will' b'e constructed of stainless steel materials and neutron absorbing material of proven service in PHR spent fuel pool environments. The material list is given below:

- Materi al .ASME Designation h

Rack body Low carbon ASTM 240-304L stainless. steel Adjustable rack ' Stainless steel ASTM 564-630 support feet Stainless steel ASTM 276-521800 Fixed rack support feet- Stainless steel ASTM 240-304 Girdle bars Stainless steel ASTM 240-XM-29 Neutron poison Boraflex These material combinations effectively prevent the possibility of material

• wastage due to general corrosion, localized corrosion, and galvanic corrosion

as discussed below.

The pool-liner, rack honeycomb structure and support components, and fuel storage tubes are stainless steel, which is compatible with the storage pool environment . In this environment of oxygen-saturated borated water, the .

corrosive of 6.00x10 d fterioration of Type 304 stainless steel should not exceed a de inches in 100 years, which is negligible relative to the I

initial thickness. Dissimilar metal contact corrosion (galvanic attack)

I between the stainless' steel of the pool liner, rack components, fuel storage tubes, and the Inconel and the Zircaloy in the spent fuel assemblies will not be significant because all of these material are protected by highly' passivating oxide films and are therefore at similar potentials.

Boraflex (neutron poison-material) is used in Region I racks (290 out of 1,324 storage locations). It is composed of nonmetallic materials and, therefore, will not develop a galvanic potential in contact with the metal components.

Boraflex has undergone extensive testing to study the effects of gamma-irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material. The evaluation tests have shown that the Botaflex is unaffected by the pool water environment and will not be degraded by corrosion. . 'These tests indicate that Boraflex maintains' its neutron attenuation capabilities after being subjected to an environment of borated water and gamma irradiation. . Irradiation will cause some loss' of .

flexibility, but will not lead to breakup of the Boraflex. .

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O Long-term borated water soak tests at high temperatures have also been conducted. The tests show that Boraflex withstands a borated water immersion

. of 2400F for 260 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 space which contains the Botafiex is vented to the pool to prevent bulging or swelling of the stainless steel cover sheet during possible off-gassing.

The tests also show that Botaflex does not possess leachable halogens that might be released into the pool environment. Boron carbide of the grade normaily in the Boraflex will typically contain 0.1 wt% of soluble boron. The l

test results have confirmed the encapsulation function of the silicone polymer I matrix in preventing the leaching of soluble species from the boron carbide.

To provide added assurance that no unexpected corrosion or degradation of the poison material will compromise the integrity of the racks, PGandE has committed to conduct a long-term poison coupon surveillance program; the poison coupons will be represeritative of the material used in Region 1 of the pools.

PGandE concludes that the compatibility of the materials and coolant used in spent fuel pools is adequate based on tests, data, and actual service experience in operating reactors, and the selection of approprate materials and adoption of a surveillance program.

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Item 9:

Provide the results of the fuel handling accident analysis. If the radiological consequences do not change, this should be clearly stated and the calculated offsite doses be reported. The assumptions for the analysis should also be presented.

Resnonse 9:

The design basis. fuel handling accident (dropped assembly) in the fuel handling building discussed in Section 15.5.22 of the FSAR Update was reviewed to evaluate potential differences in the radiological dose consequences associated with the storage of fuel having design parameters different from the present fuel.

NUREG-0575 (Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel, U.S. NRC, NUREG-0575. August 1979, Appendix D, page D-9) states this accident involves the drop of a single-fuel assembly and the offsite dose calculation does not change as a result.of storing additional fuel assemblies in the spent fuel pool.

As discussed in the Reracking Report, the racks were designed for fuel of higher enrichment and longer burnup times than the current licensing basis for the plant. This requires consideration of possible changes in radionuclides source terms for this accident analysis.. Therefore, calculations were made of the radiological dose consequences from this accident for fuel of the maximum initial enrichment of 3.5% by weight of U-235 presently permitted by Technical

/ Specification 5.3.1, and fuel of maximum initial enrichment of 4.5% by weight of U-235 which is the design basis for the proposed high density spent fuel racks. Calculations were done for burnups of 33,000 20/MTU and 50,000 WD/MTU for each enrichment. These calculations indicated that the largest thyroid doses were 24.4 rem (exclusion area boundary) and 1.1 rem (low population zone). The largest whole body gamma-'deses were 0.582 rem

-(exclusion area boundary) and 0.0264 rem (Iow population zons). All results were well within 10 CFR 100 guidelines, and are considered acceptable.

The assumptions used for these analyses were consistent with the guidelines of Regulatory Guide 1.25. The radionuclides inventory source terms for these analyses were calculated by the ORIGEN code at a power level of 3.411 MHt and d delay of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown. A radial peaking factor of-1.65 was used as required by Regulatory Guide 1.25. All of the fuel. rods in the damaged assembly were assumed to be ruptured. The radionuclides gap' activity fractions, release fractions, and spent fuel pool decontamination factors were the standard values given in Regulatory Guide 1.25. The fuel handling building filter efficiencies and atmospheric dispersion factors were the values specified in Table 15.5-45 of the FSAR Update.

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. As discussed at the meeting on December 5, 1985 the backup source cooling water for the spent-fuel pool is the condensate storage tank and.the fire water tank. Evaluate the amount of water required in these tanks to meet both the ESF needs during a plant emergency and at the same time provide.the makeup for evaporation losses from the spent fuel pool in the event of the loss of normal cooling. Evaluate the plant Technical Specifications to assure that the necessary water level is maintained to accommodate both uses.

Resnonse 10:

Discussion of this item is provided in a supplement to the Reracking Report-and is contained in PGandE letter DCL-86-020. dated January 28, 1986.

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Item 11:

The rack module design criteria are based upon Section 3.8.4, Appendix D, of the NRC's Standard Review Plan. The staff requires that load combinations and allowable stresses for spent fuel rack modules be in accordat.ce with the allowable values of the NRC's Position Paper, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," U. S. Nuclear Regulatory Commission, January 18, 1979. Demonstrate that the modules meet the OT Position criteria.

Resoonse 11: 1 1

The analysig presented in the Reracking Report meets OT Position i guidelines. As stated in the Reracking Report, the stress levels in the rack feet greatly exceed those in the rack modules proper. Therefore, in essence, support feet stresses govern the rack design. Since there is no j interaction between To, Ta loads, and inertia loads (D, L E, E') in the '

support feet, the Standard Review Plan 3.8.4 (NUREG-0800) and OT Position criteria both yield the following effective limits:

Leadina D+L+E Normal limits as stated in Appendix XVII-2000 0 + L + E' Normal condition stress limits increased by Appendix F-1370.

@ Thus, although there are differences in the stress criteria between the OT position paper and Standard Review Plan 3.8.4, Appendix D, they amount to l

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identical limits for Diablo Canyon.

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• l "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications " April 14, 1978 and January 18, 1979 revision, USNRC, j Hashington, D.C. j l

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0670S/0040K

( Item 12:

Because of the added nonlinear complexity associated with the increased number of gap elements in the three-dimensional displacement analysis to account for rack-to-rack impact, the rack displacement analysis is based upon an 8 degree-of-freedom (DOF) model. Provide documentation showing the following:

a. That the 8 DOF model adequately portrays the mass-elastic dynamic motion of the fuel rack module (such as comparison of maximum displacement and time of recurrence with 32 DOF models).

b. How the single lumped mass represents the movement of the fuel assemblies and how it is referred to the top of the model.

c. How convergence and stability of the numerical integration on the nonlinear displacement solution are assured.

Resnonse 12a:

The 8 degree-of-freedom model was benchmarked against the older 32 00F model using Diablo Canyon rack modules.

Selected output from a typical convergence run is shown below (g = 0.2, no structurh1 damping):

• 32 DOF 8 00F i Time Step 1x10-5 sec 5x10-6 sec 5x10-6 sec Max. x-displacement 2.595 in. 2.710 in. 2.793 in.

Max. y-displacement 2.147 in. 2.213 in. 2.278 in.

Max. support load 1.464x105 lb 1.484x105 lb 1.22x105 lb on the foot Max. R6 in support 0.272 0.279 0.280 leg Max. R6 in the 0.115 0.106 0.123 rack body It should be noted that there is overall agreement between the two models.

O 0670S/0040K Resoonse 12bt The 8 degree-of-freedom model actually has two lumped nsses: one at the rack base plate location, another at the elevation of the top of the rack. The two masses are assumed to be connected by a rigid link. The ratio of the mass assigned at the top to that at the bottom is established by energy considerations, and further fine-tuned by making benchmark runs against the 32 DOF model.

Resoonse 12c:

Convergency and stability of the runs are assured oy running the same problem at different time steps and observing the close agreement between the results thereof. The output data of an 8 DOF run for a fully loaded 10x11 rack with coefficient of friction (COF) - 0.8 for two time steps are shown below.

Maximum Maximum Max.R6 Maximum Time Step x-displace- y-displace- in Floor ment ment Rack Load 5x10-6 sec 0.3402 in. 0.4780 in. 1.49 6.735x105 lb 2.5x10-6 sec 0.3401 in. 0.4783 in. 1.52 6.727x105 lb The above table confirms the excellent convergence characteristics of the 8 00F model.

O

.l 06705/0040K f),

k/ Item 13:

Describe the analysis methodology used to determine stresses in the rack module base metal and welds due to impact between the support pads and the

' floor following liftoff and due to rack-to-rack lateral impact. -

Resnonse 13:

The-locations in the dynamic model where impact may possibly occur are )

equipped with stop elements containing compression springs. The dynamic j simulation tracks the load in these springs, and stores it for further '

processing. The maximum value of the impact loads thus computed is applied as a static load on the actual structure. For example, the girdle bars at the top of the module are a designated impact location. The maximum load registered in the spring simulating the stiffness of the girdle bar is applied on the bar as a line load. The line load a.ssumption is verified by examining the girdle bar impact sequence which typically shows the impact commencing at one end and propagating to the other in less than .001 second. The bar itself can be modelled as a beam on an elastic foundation with simply supported ends where the elastic foundation modulus depends on the compression stiffness of j the cell walls transverse to the girdle bars.

Similarly, the support foot -Impact loads to the floor are obtained by filtering out the maximum load in the support spring; in the dynamic model.

The moments and shears produced by these loads in the support foot and rack ptoper are computed using standard " linear elasticity" principles.

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l 06705/0040K 1

i i

Item 14:

Justify the selection of the 6 X 11 cell rack module, stated to have the largest-aspect ratio, for the presentation of the analysis results. Include a discussion of why the natural frequencies of other racks would not'make them more susceptible to large displacements under seismic excitation. j i

1 Resoonse 14:

The natural frequency of the modules is eliminated from consideration'in selecting the bounding rack module due to the fact that the rack fundamental frequency in any mode is significantly higher than 30 Hz. Therefore, the principal contributors to dynamic amplification are rack inertia and the rack -

cross-section aspect ratio. The 10x11 rack module maximizes rack inertia; the <

6x11 rack maximizes the aspect ratio of the rack cross-section.

The above reasoning led to the selection of a 6x11 module as a limiting geometry.

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J 0670$/0040K i

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Item 15:

With respect to the spent fuel structure only limited information is provided in the Reracking Report regarding the adequacy of the analytical procedures, the load combination criteria, or the selection of allow:ble loads and stresses, other than to reference the original spent fuel pool analyses l

included in the FSAR. Accordingly, the staff requests that you provide the following:

a. Sketches and/or drawings of any changes to the spent fuel pool structure not considered in the FSAR analysis.

l l- b. A description of the mathematical model of the pool structure, including the finite-element model, if used, and the method of I analysis. Describe the assumptions employed and the limitations of I the model,

c. A detailed description of the loadings used, and justification for the load combination.

d. Identification of the source of the acceptance criteria and method of determining the allowable loads and stresses in various parts of the '

structure.

e. Description of the dynamic interaction between the pool structure and the rack modules, including the value of any dynamic amplification i factors. Include all assumptions made regarding the sumation and b phase of all rack module dynamic loads.

f. An analysis of the adequacy of the pool floor and liner under rack sliding and impact loads.

i

g. Identification of the critical regions of the pool structure. List the loads or stresses as appropriate. Compare the loads and/or  ;

stresses to allowable values, indicating the source of the allowable  ;

in accordance with Item 15.d above. .

l l Backaround l The spent fuel pools are located on each side of the east end of the auxiliary j l building. Figure 1 shows the plint layout and foundation elevations.

Generally, one half of the auxiliary building is symmetrical to the other, l with each half of the structure containing equipment for one unit. The Unit 1 '

spent fuel pool is shown in Figure 2. The walls of the spent fuel pool are 6  ;

feet thick- except for local areas around the fuel transfer tube, as shown on  !

Figures 3 and 4. The foundation slab under the spent fuel pool has a minimum thickness of 5 ft and is founded on approximately 5 ft of lean concrete placed  ;

on rock. The spent fuel pool sides and bottom are lined with stainless steel, i 1/4 in, thick on bottom and 1/8 in. nominal thickness on the sides. Figures 5 l and 6 show the typical layout and details of the liner.  ;

0670S/0040K l

p)

( The spent fuel pool structural analysis, addressed in the FSAR, is affected by the high density fuel racks due to the following:

o Weight of the proposed high-density racks o Thermal gradient across the walls o Interaction between the racks and the pool structure The effect of these items is addressed in the response to Question 15 below.

Resoonse 15a:

There are no physical changes to the spent fuel pool concrete structure addressed in the FSAR analysis. However, minor modifications to the liner plate will be made to accommodate the high-density racks. The modifications include relocation of spent fuel handling tool brackets to the east end of the north and south walls of Units 1 and 2, respectively. Figure 7 shows typical details of the modified brackets.

Resoonse 15b As described in the FSAR, Section 3.7.2.1.7.1, the seismic inertia loads were obtained using a time-history analysis of a spring and lumped mass model of the auxiliary building. Two horizontal models and a vertical model, shown in Figure 8, were used in the analysis.

V A detailed analytical static model of the auxiliary building (Figure 9) was used to distribute seismic inertia forces and moments to various walls, diaphragms, and columns, as described in the FSAR, Section 3.8.2.4.

The effect of the change in weight on the seismic models due to the high-density racks was insignificant, since increase in global mass was determined to be less than 17.. Therefore, there is no change in the seismic responses and forces reported in the FSAR.

The pool walls were analyzed for local out-of-plane effects due to hydrostatic, hydrodynamic, and seismic loads using U.S. Bureau of Reclamation  !

tables for plates (Ref. 1). Hydrodynamic effects were determined based on the i methods in TID 7024, SNuclear Reactors and Earthquakes" (Ref. 2), and

" Dynamics of Fixed-Base Liquid-Storagt Tanks" (Ref. 5). Thermal effects in the pool elements due to the increased thermal gradient generated from the use of the high-density racks were analyzed considering cracked concrete section.

The liner has been evaluated for loads and load combinations described in Response 15c. The mathematical model user' in the thermal evaluation consists of a set of springs representing the axial stiffness of the liner and flexural stiffness of the embedded anchors, as shown in Figure 10.

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i 0670S/0040K . - - _ _ _ _ - _ .

i l

Resoonse 15c:

I. Concrete Structure The loads and load combinations used for the evaluation of the. spent fuel.

pool structure are in accordance with the FSAR Update, Section 3.8.2.3, and are listed below. ,

A. Normal Conditions Dead load, . live load, loads from the DE, thermal loads, and pipe reactions are considered in all possible combinations. Inasmuch as working stress design is used for normal operating loads, the factored load approach is not used. For each structural member, the combination of these loads that produces the maximum stress is used for design. Stated in equation form: I C = D + L + DE + To + Ro -(1) where:

C = Required capacity of member based on working stress of i ACI 318-63 as supplemented by Section 3.8.2.5.3 of the FSAR Update, which is described in response to Item 15d.

D = Dead load of structure and equipment loads including hydrostatic loads L = Live load i

\

DE = Loads resulting from the design earthquake To - Thermal loads during normal operating conditions (inside face of wall 1760F, exposed face 910F, and bottom face of lean concrete 1010F)  :

Ro = Pipe reactions during normal operating conditions.

i B.' Abnormal Conditions l Dead load, live load, earthquake loads, and loads associated with .

i accidental pipe rupture are considered in the following l l combinations; for each structural member, the combination that 1 I

i l

1 I

06705/0040K l

i produces the maximum stress is used for design:

U=0+L+TA+RA + 1.5 PA (2)

U=D+L+T +R + 1.25 PA+ l 1.0(Yj + m+ r) + 1.25 DE (3T

, U=D+L+TA+R 1.0 PA+

(4) l 1.0(Yj + Ym +Trg +) + DDE U=D+L+T + Rg + 1.0 PA+

1.0(Yj + m + Tr) + HE (5) where:

TA = Thermal loads on structure generated by a postulated pipe break, including To (for T A inside face of walls 2140F, exposed face 930F, and bottom face of lean J concrete 1130F)

RA = Pipe reactions on structure from unbroken pipe generated by postulated pipri break conditions, including Ro PA = Pressure load within or across a compartment and/or building generated by a postulated pipe break, and including an appropriate dynamic factor (DLF) to account for the dynamic nature of the load Y3= Jet load on structare generated by a postulated pipe break, including an appropriate DLF Ya - Missile impact load on a structure generated by, or during, a postulated pipe break, such as a whipping pipe, including an appropriate DLF l Yr = Reaction on structure from broken pipe generated by a postulated pipe break, including an appropriate DLF U = Ultimate strength required to resist design loads based on the methods described in ACI 318-63. For load combinations involving Y3 , Ym, and Yr, local stresses due to those concentrated loads may exceed the l allowable provided there is no loss of function. See ,

Reference 3 Enclosure 3, Document (B), for more detailed information concerning the acceptance criteria for these load combinations.

{ DDE = Loads resulting from the double design earthquake HE = Loads resulting from a Hosgri earthquake i For the spent fuel pool structural evaluation, PA, Yj, Ym, and Yr are (O

x) not applicable. l 06705/0040K _ _ _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ .

O 1I. m

. To assure leaktightness of the liner, the following loads and load combinations are used in the evaluation of the liner:

D+L+To (65

.D + L + Ta + HE. (7)

Resnonse 15d:

I. Concrete Structure The acceptance criteria as described in the FSAR Update, Section 3.8.2.5, -

are summarized below.

For DE and DDE load combin' ations, the nominal design strength for concrete and specified yield strength for reinforcing and structural

-steel are considered. For load combinations including HE, however, the actual material properties are used.

1. Harmal Leads For normal loads, the auxiliary building is designed for the allowable working stresses of ACI 318-63, as supplemented by. Item O 2.

3 below.

Abnormal Loads For abnormal loads, the auxiliary building is designed for overall elastic behavior. For concrete elements, the strength design of ACI 318-63 applies, as supplemented by Item 3 below.

For load combinations involving.Y4, Ya, and Yr as applicable, local stren es due to'those concentrated loads may exceed the allowables provided there is.no loss of function. See Reference 3, Enclosure 3, Document (B), for more detailed information concerning the acceptance criteria for these load combinations.

3. In-Plane Loads on Concrete Elements The design of slab diaphrages and shear walls for in-plane forces is not explicitly t. overed by ACI 318-63.. Section 104 of ACI 318-63 allows criteria based on test data to be used for the -

I design of elements not covered.by its provisions. Consequently,

~

O 06705/0040K - _ - _ _ ._ _ _ ___ _ - -_ - _ _ _ - _ _ _ _ -

I the document titled " Recommended Evaluation Criteria for Diablo Canyon Nuclear Power Plant Auxiliary Building Walls and Diaphragms" (Ref. 4) is developed to provide criteria for

~

evaluation of auxiliary building shear walls and floor diaphragms

.for in-plane seismic forces, including the simultaneo.us effects of out-of-plane forces.

Accordingly, the structural elements are evaluated as follows: ,

(1) The columns are evaluated by the provisions of ACI 318-63 for all loading conditions.

(2) The slabs and walls are evaluated for out-of-plane loads l according.to ACI 318-63, and for in-plane loads according to Reference 4.

II. Ling The allowable concrete bearing pressure under rack dead loads is in accordance with ACI 318-63. The design for local impact loads is not covered by ACI 318-63. ACI 349-80 is used to determine the allowable bearing pressure on concrete due to rack impact.

The allowable foundation pressure is in accordance with FSAR Update l Section 3.8.4.1.4.

b The allowable liner strains and anchor displacements for normal operating and accident thermal conditions are in accordance with ASME Section'III, Division II,1983.

Resoonse 15e:

The interaction between the pool structure and the rack modules is primarily between the support feet and the pool slab. The table below shows the maximum cumulative support reaction (sum of 4 feet) for the two representative modules studied (Hosgri earthquake):

Module Maximal value of Impact Tyne Dead Load (b) cumulative reaction (a) Factor a/b 10 x 11 158,676 lb 677,000 4.27 6 x 11 97,652 lb 405,500 4.153 l

O 0670S/0040K ..

I An impact factor f of 4.5 can be safely taken to bound the data for all modules. Since there are 13 types of modules in a pool containing 16 modules,

~

these maximal reactions will occur at different instants of time, and' statistical summation of these. reactions is appropriate. Dynamic amplification of loads due to soil-structure interaction is determined to be insignificant because the structure is supported on bedrock.;

Those empty racks which are located on the outer periphery of the pool to within 4 in. of the pool wall may impact the wall under the Hosgri i

. earthquake. The associated impact loads are smail: the 10x11 module (empty)

' develops a maximum load of 38,550 lb with a total duration of .001 sec. The wall impact loads for all proximate modules may be conservatively estimated to be 80,000_ lb under HE conditions, applied as a line load over the length of the girdle bar' facing the wall. This effect is included in Response 15(g).

Resoonse 15f:

The maximum vertical impact load under the support leg of rack modules is 189 kips for the Hosgri Ioad combination. This load is transmitted to fcundation slab and bedrock via bearing. Two cases are considered:

(1) 6-1/2 in, diameter adjustable rack foot over leak chase (2) 6-1/2 in. diameter adjustable rack foot away from leak chase The bearing pressure is computed by discounting the area of leak chase drains below the rack leg. The maximum bearing pressure computed is 7.5 ksi, which is below the concrete (8.4 ksi) and steel (24 ksi) allowables. However, for ease of installation, provisions are being make to provide bearing plates with minimum dimensions of 12.5 in. x 12.5 in. x 1 in, under each foot of the racks. This would further reduce the bearing pressure to 4.5 ksi and increase the factor of safety against the allowable value.

The effect of the impact load on the foundation bearing pressure is determined to be insignificant because the maximum computed contact pressure'underneath

, the fuel pool is 7.2 ksf, which is significantly lower than the allowable l foundation. bearing pressure of 80 ksf used in design of adjacent structures I

(FSAR Update Section 3.8.4.1.4).

The maximum horizontal sliding force on a support leg of rack modules is estimated to be 151 kips. This sliding force is resisted by frictional resistance between the steel liner and concrete floor slab, and by bearing of the stiffener angles embedded in concrete. Two loading cases are considered:

(1) Sliding force resulting from four legs of interior rack modules clustered together (2) Sliding force resulting from one leg of a module The liner plate weld seams are checked for the maximum axial force developed as a result of difference between the rack sliding forces and the resisting A forces. The maximum force computed in the liner weld seam is 1.5 k/in.

D compared with allowable force (,f 6 k/in.

06705/0040K ,

___________.._m. _ _ _ _ _ _

Resnonse 15a:

. I. Concrete Two effects are considered in the analysis of the pool structure: global effects and local out-of-plane effects.

(1) fa]obal loads - Controlling regions in the walls of the auxiliary' building due to in-plane effects are summarized in FSAR Update Tables 3.8-14,15, and 16 for the DE, DDE, and Hosgri events .

respectively. The pool structure wall stress ratios from these tables are tabulated in Table 1.

(2) Local out-of-olane loads - Tables 2, 3, and 4 summarize the demand and capacity of_ the controlling regions of the pool structure.

These values take into account the increase in thermal gradients due to high-density racks, structure / rack interaction, and the other loads as applicable to the specified load combination.

In all cases, the stress ratios are greater than the minimum allowable value of 1. Therefore, all structural elements satisfy the criteria.

II. L1DLC The liner plate strain and anchor displacements are summarized in Table 5

! .A for the governing load combination. In all cases the strains and displacements are within allowable values.

Q/

l O

06705/0040K  !

1 References for Ouestion 15

1. " Moments and Reactions for Rectangular Plates " Engineering Monograph No.

27, U.S. Department of Interior, Bureau of Reclamation.

2. " Nuclear Reactors and Earthquakes," TID 7024, NTIS, U.S. Department of Commerce.

3. Letter (Docket Nos. 50-275 and 50-323) from A. Giambusso of the U.S.

Atomic Energy Commission to F. T. Searls of the Pacific Gas and Electric Company, dated August 23, 1973, including enclosures.

4. " Recommended Evaluation Criteria for Diablo Canyon Nuclear Power Plant Auxiliary Building Halls and Diaphragms," Jack R. Benjamin & Associates.

Inc., February 11, 1983.

5. A. S. Velets::s and J. Y. Yang, " Dynamics of Fixed-Base Liquid-Storage l

Tanks," Proceedings of U.S.-Japan Seminar on Earthquake Engineering l Research with Emphasis on Lifeline Systems, November 1976.

6. " Containment Building Liner Plate Design Report, BC-TOP-1," Bechtel Power Corporation, San Francisco, California.

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MOTATION S-AMcHOR DISPLACEMENT I

Kc-AncaoA sTzrrNEss Kgp*ST3FFMEs$ OF LINER PANEL A l K ap=5TIFFNESS OF LINER PANEL S i SPENT FUEL POOL

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5 FIGURE 10 5

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(J Table 1 AUXILIARY BUILDING SPENT FUEL POOL CONCRETE HALLS STRESS RATIOS Moment x 10 3 . k-ft Shear. esi Stress Hall Location (a) El. ft Demand Canacitv(b) Demand Canacitv(b) Ridiq(c)

A. DESIGN EARTHQUAKE HALL AD JDQ 7]i 100 141 143 14 85 85 220 164 664 2.6 HALL BC 100 45 155 20- 113 3.4 HALL CD 100 80 250 54 146 2.7 HALL AB 100 80 240 54 145 2.7 B. DOUBLE DESIGN EARTHQUAKE

/ HALL AD JfD 4 J_4.0 34Q 223 ._.J182 2d 85 170 370 316 1,086 2.2 l HALL BC 100 85 260 40 244 3.0 NALL CD 100 160 420 105 280 2.6 l HALL AB 100 160 410 106 268 2.5 l

C. HOSGRI EARTHQUAKE HALL AD 10Q Jjl0 31Q 118 B.fiQ l.Ji 85 230 320 493 979 1.4 HALL BC 100 130 190 60 197 1.5 HALL CD 100 330 480 220 339 1.4 HALL AB 100 270 460 199 324 1.6 (a) For wall identification, see Figure 3. Counterparts in Unit 2 are similar.

(b) Axial demand effect is included in the capacities.

(C) Stress ratio - capacity / demand for shear or moment, whichever is smaller.

06705/0040K Table 2 AUXILIARY BUILDING SPENT STRESS RATIOS FUEL POOL (CQNCRETE STRUCT (DE) a> ,

Demand Values Canacity Values Moment Shear Moment Shear Stress Ball (b) k-ft/ft kl.f1_ k-ft/ft k/ft Ratio AB 195 31 220 56 1.1 BC 220 35 245 56 1.1 AD 225 42 245 56 1.1 CE 145 20 220 56 1.5 ED 180 27 250 65 1.4 GF 215 27 260 46 1.2 l Basemat 210 small 275 large 1.3 (a) Stress ratio - capacity / demand for shear or moment, whichever is sma)1er.

(b) For wall identification, see Figure 3.

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" Table 3 i

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. AUXILIARYBUILDINGSPENTFUELPOOLgCRETESTRUCTURE l STRESS RATIOS (DDE)  ;

Demand Values Canacity Values Homent Shear Moment Shear Stress Egil(b) k-ft/ft KLf1 k-ft/ft g/_fj; Ratio AB 265 38 420 85 1.6 BC 295 42 460 85 1. 6 '

AD 305 52 460 85 1.5 CE 195 25 420 85 2.2 ED 235 30 470 99 2.0 GF 285 32 490 71 1.7 Basemat 265 small 520 large 2.0 (a) Stress ratio - capacity / demand for shear or moment, whichever is smaller.

(b) For wall identification, see Figure 3.

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I O l 0670S/0040K O. Table 4

. AUXILIARYBUILDINGSPENiFUELPOOL(CgNCRETESTRUCTURE STRESS RATIOS (HE) A .

Demand Values Canacity Values I

g(b) !3E$7ht b 3I$7ht b h!b 300 45 500 92 1.7 AB BC 355 55 550 92 1.5 AD 365 68 550 92 1.4 CE 200 31 500 92 2.5 ED 245 34 570 107 2.3 GF 295 37 590 76 2.0 j Basemat 345 small 630 large 1.8 (a) Stress ratio - capacity / demand for shear or moment, whichever is smaller.

(b) For wall identification, see Figure 3.

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Table 5 AUXILIARY BUILDING SPENT FUEL POOL LINER

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STRAINS AND DISPLACEMENTS LINER PLATE STRAINS i Liner Governing Location Lead Comb Demand Allowable (1) Safety Factor (1)

Wall D+Ta+HE 0.00112 0.005 4.5 Floor D+Ta+HE 0.00156 0.005 3.2 LINER ANCHOR DISPLACEMENTS {

Liner Governing Location Load Comb Demand A110wable(3) Safety Factor (2) 1 Hall D+Ta+HE 0.01220 0.06200 5.2 ,

Floor D+Ta+HE 0.01375 0.06200 4.5 (1) From ASME B&PV Code Section III, Division 2,1983 Edition, Table CC-3720-1.

(2) Safety factor is computed relative to the code allowable.  ;

(3) From ASME B&PV Code Section III, Division 2,1983 Edition, Table C-3730-1, using Su - 0.125 in. based on test results given in BC-TOP-1 Appendix B (Ref. 6).

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l O Item 16:

. Discuss the spent fuel pool cooling system with respect to the requirements of General Design Criterion 44 of 10 CFR 50. Appendix A. .

Resnonse 16:

The spent fuel pool cooling system is discussed in a supplement to the );

Reracking Report and is contained in PGandE letter DCL-86-020 dated January 28, 1986, O

<l 0670S/0040K-Item 17:

The.Reracking Report does not include the' calculated decay heat loads for t'he l proposed pool modifications or sufficient information for the staff to

. calculate the loads. independently. . Provide the following information (a) all f anticipated future discharges as a function of decay time and (b) the decay l heat load for each discharge for the maximum normal and_ the maximum abnormal j conditions (the maximum normal. heat load is the heat load assuming' the pool is j filled with successive normal refueling discharges; the maximum abnormal heat load is the heat load assuming one_ full core. discharge and successive normal refueling discharges).

1 Resnonse 17:-

Section 5 of the Reracking Report discusses the thermal hydraulic-analysis performed in support of the increased pool storage capacity. Results for three cases were presented which constitute the worst case design conditions.

-Cases 1 and 2 are for a partial core offload (refueling), while Case 3 is for a full core offload. Table 5.1' lists these cases, and Table 5.2 provides the results of the calculations. The decay heat load (Q x10-0 8tu/ hour) in Table 5.2 represents the total decay heat in the poo' for the respective cases at the time of maximum pool temperature. For example, for Case 1, the pool decay heat load is 21.89x106 Btu / hour at 136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br /> after reactor shutdown.

It is based on 76 fuel assemblies from the present refueling plus 15 previous ~

l refuelings of 76 assemblies each dischat'g ed at 18 month intervals in the pool. The decay heat present in the pool as a function of decay time is O provided in Figures 5.1c and d through 5.3c and d for the three base cases.

These cases represent and of pool life conditions where all storage locations are filled with spent fuel from previous refuelings and, therefore. envelope the results for any individual previous refu' ling.

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Item 18:

Provide the time interval between reactor shutdown and the commencement of the discharge of assemblies. Provide the time.to complete a normal discharge and a' full core discharge.

Resnonse 18:

Placement of fuel assemblies into the pool is assumed to commence 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown, which is the minimum time required by plant Technical Specifications prior to moving irradiated fuel from the reactor vessel. The rate of discharge is assumed to be four assemblies per hour. . Therefore, it takes 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> to discharge 76 assemblies (normal discharge) and 48.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to discharge a full core after the start of discharge.

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Item 19:

. Provide a P&I diagram of the spent fuel pool cooling system, and provide the i assumptions made in establishing its rated heat removal capability. l 1

Resoonse 19:

The spent fuel pool cooling system is discussed (including the P&I diagram) in a supplementfto the Reracking Report and is contained in PGandE letter DCL-86-020 dated January 28, 1986.

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0670S/0040K-Item 20:

For the maximum normal and maximum abnormal heat load conditions provide the pool water temperature as a function of time and all assumptions on which the calculations are based.

Resoonse 20:

Figures 5.la and b through 5.3a and b of the Reracking Report provide the pool bulk temperature as a function of time for the three base cases. Case 1 represents a refueling case of 76 assemblies for the maximum normal heat load. Case 2 is a bounding case for a projected maximum refueling of one half of the core. Case 3 represents the maximum full core discharge heat load.

All'the assumptions on which the calculations are based are contained in Section 5 of the Reracking Report.

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O Item 21:

For the maximum normal and maximum abnormal heat loads, assuming the pool

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cooling.is lost, provide the time before boiling occurs, the boiloff rate, and the time before the boiloff causes the top of the storage racks to become uncovered.

Resnonse 21:

Table'5.3 of the Reracking Report provides the time before boiling occurs and' the boiloff rate after loss of cooling for the three base cases previously described. These are based on the fuel pool being filled with previously discharged fuel. The time to boil is taken from the instant all cooling is

lost. Condition'1 assumes loss of cooling occurs at the time the maximum pool l

bulk temperature is reached, while Condition 2 assumes loss of cooling at the time of maximum decay heat in the pool (i.e., at the time that the last fuel assembly is placed in the pool). The results show that for ' loss of cooling after discharging 76 assemblies into a pool containing 15 previous discharges.

of 76 assemblies each at 18 month intervals (Case 1. Condition 1).-boiling would occur 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after cooling was lost, and the vaporization rate at the time boiling starts would be 22,145 lbs/ hour. If cooling is lost after a full core discharge (Case 3, Condition 1), boiling would occur in 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the vaporization rate would be 43,676 lbs/ hour. These values are conservative since no heat loss is assumed through the walls or floor of the pool, and heat loss due to evaporation prior to boiling'is not included.

The minimum time it would take to boil off all of the water in the. pool down to the top of the racks for Case 1 is 85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> after boiling starts. For Case 3 a full core discharge, the racks would not begin to be uncovered until a-minimum of 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> after boiling begins. This is based on the water level in the pool starting at its lowest normal level, elevation 137 ft 4 in., with the top of the racks at elevation 113 ft 9 in. The decay heat rate and hence the vaporization rate are conservatively assumed to remain constant, and the only heat loss from the pool is assumed to be through evaporation.

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^ Item 22: l Describe the pool water level monitoring rystem and indicate the location of the alarm. .

Resoonse 22:

The level monitoring system for the spent fuel pool consists of a level i

transmitter LC-650 and temperature indicator TI-651. This level indicating channel has alarms with a high setpoint of 16 in, above normal water level and a low setpoint of 4 in. below normal level. The tolerance on setpoints is plus or minus 1 in. The temperature indicating channel presently has a high alarm se %oint of 125'F plus or minus 2*F. Both of these transmitters are powered from instrument AC panel 16 (breaker 1620), which is normally powered from Class 1E vital bus G with a backup power supply from bus F.

The high/ low' level and temperature alarms are located on main annunciator panel 11, window 04, on vertical board 3 in the main control room.

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Item 23:

. Describe the available makeup water systems, the quantity available from each source, and their seismic classification. Indicate their respective makeup rates and the time interval between their activation and when the makeup flow rate is achieved.

i Resnonse 23:

A discussion of the spent fuel pool makeup sources is contained in PGandE letter DCL-86-020 dated January 28, 1986.

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0 06705/0040K Item 24:

Provide an estimate of the expected increase in radwaste volume teased on past plant experience or experience at similar sites.

Reseense 24:

PGtndE does not expect an increase in radwaste. volume due to the spent fuel pool reracking. PGandE plans to accomplish the reracking modification prior to any spent fuel being placed into the pools. Materials utilized in the teracking should not be contaminated and, therefore, not treated as radwaste.

The increased storage capacity is not expected to result in an increase in radwaste from the spent-fuel pool cleanup system over the life of the plant.

As described in the response to Item 2 of this enclosure, there is no projected increase in the radioactivity in the spent fuel pool, so no additional load is expected to be added to the cleanup system. PGandE has contacted personnel at Nine Mile Point, Zion, and Farley. Those plants have not increased the volume of radwaste generated as a result of spent fuel storage after reracking.

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O k/ Item 25:

Indicate the depth of spent fuel pool water which will normally lie over stored-in-place fuel elements, and provide the resulting pool surface dose rates for this condition. .

Resoonse 25:

The normal water level in the spent fuel pool ranges from elevation 137 ft 4 in. to elevation 139 ft 0 in. The top of the spent fuel racks will be approximately at elevation 113 ft 9 in. Therefore, there will typically be from 23 ft 7 in. to 25 ft 3 in. of water over the top of the new spent fuel <

racks. In addition, the top of a spent fuel assembly stored in the rack is nominally 1.5 in, below the top of the rack, and the top of the active fuel is approximately 1 ft 6 in. below the top of the fuel assembly. The minimum water depth above stored spent fuel required by Plant Technical Specification is 23 ft.

The maximum dose rate at the surface of the pool, except during fuel handling operations, is 2.5 mrem / hour. This dose rate is principally due to the radioactivity in the water. The spent fuel pool cleanup system is used to maintain radienuclide concentrations in the spent fuel pool as low as is reasonably achievable. The dose rate at the surfact of the pool due to the stored assemblies has been calculated to be 4.3x10-o mrem / hour.

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PGandE Letter No.: DCL-86-019 O ENCLOSURE 2 PGandE RESPONSE TO .

ITEMS RAISED AT DECEMBER 5. 1985 MEETING SPENT FUEL POOL RERACKING Item 1:

Identify proposed changes to mechanical, cleanup, HVAC and radiation monitoring systems as resulting from the reracking.

Resoonse 1:

This item is addressed in Item 4 of Enclosure 1.

l Item 2:

Identify the sources of makeup water and flow capacity. Consider the need for changes to the Technical Specifications if the condensate storage tank is relied on as a source for makeup water.

Q Resnonse 2:

This item is addressed in Items 10 and 23 of Enclosure 1. ]

Ltem_3:

Discuss the spent fuel pool cooling system with respect to the requirements of General Design Criterion 44 of 10 CFR 50, Appendix A.

Eminonse 3:

This item is addressed in Item 16 of Enclosure 1.

Item 4:

Clarify if the allowable stress limits for Level A and Level B service are identical (see page 6-16 of the Retacking Report).

Resoonse 4: ,

The allowable stress limits for Level A and Level 3 service are not identical; however, the allowable stress limits for Level A service were used for the tO. first five loading combinations on page 6-16 of the Reracking Report.

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() Item 5:

Specify the water level above the top of the new fuel racks for various conditions.

Resoonse 5:

This item is addressed in Item 25 of Enclosure 1.

Item 6:

Discuss the compatibility of the new fuel racks with the fuel pool material and environment.

Resoonse 6:

This item is addressed in Item 8 of Enclosure 1.

l Item 7:

Describe in detail how ALARA commitments were considered in the spent fuel pool layout and design.

,q Resoonse 7:

! This item is addressed in Item 5 of Enclosure 1.

Item 8:

The change in offsite doses for postulated accidents should be addressed.  !

Resoonse 8-This item is addressed in Item 1 of Enclosure 1.

Item 9: )

What is the incremental radiological dose due to the increased spent fuel l storage capacity to plant operations personnel during the life of the plant from routine fuel handling operation?

Resoonse 9:

This item is addressed in Item 2 of Enclosure 1.

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1 Item 10:

Clearly indicate that fuel from other facilities will not be stored in the Diablo Canyon fuel pools and that PGandE does not intend to ship fuel offsite in the near future. -

Resnonse 10:

This item is addressed in Item 6 of Enclosure 1.

Item 11:

Address the consequences of thermal expansion of the fuel racks under maximum fuel temperature, e.g., with respect to the gap between modules.

Resoonse 11:

The thermal expansion of a typical module under the conditions of the maximum pool bulk temperature is 1sss than 0.1 in. (assuming 700F initial installation temperature). This length change is approximately equal to manufacturing tolerances in the rack, and is inconsequential as to the development of any stresses in the freestanding racks either individually or due to closure of the gaps between the rack modules.

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