Text

Supplemental Blasts & Explosions Responses to NRC Additional Questions for Diablo Canyon Independent Spent Fuel Storage Installation Application, Attachments 1 and 2 - Appendix D
	ML030940368
Person / Time
Site:	Diablo Canyon
Issue date:	03/27/2003
From:	Womack L; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	+sispmjr200505, -RFPFR, DIL-03-005, TAC L23399
	Download: ML030940368 (110)
	v • d • e

Enclosure 1 PG&E Letter DIL-03-005 Sheet 1 of 5 PG&E Supplemental Response to NRC Additional Blasts and Explosions Questions For Diablo Canyon Independent Spent Fuel Storage Installation (ISFSI)

License Application NRC Comment No. I PG&E should provide an analysis estimating the potential effects of heat flux from a postulated fire of 7,571 1[2,000 gal fuel tankernearthe storage pads. This analysis should demonstrate all safety-related structures, systems, and components will be able to continue performing their intended safety functions in the event of a tanker fire.

PG&E Response Calculation M-1052 (Attachment 1) analyzed the heat gain of a transfer cask due to a 2,000-gallon gasoline fire. The results show that the heat gain on the transfer cask does not exceed the design-basis heat load determined to be acceptable by the cask manufacturer and is acceptable. This calculation was performed with a 4-meter setback distance and is very conservative. For the other structures, systems, and components (SSCs) potentially involved with the storage cask, such as the seismiclbolting, the actual setback distance being maintained by the truck route will place this hazard at a distance where there will be no significant effect on these components. In addition, since this 2,000-gallon tanker truck will be administratively controlled with no area access when a cask is in transport or in the cask transfer facility (CTF), none of the transport or CTF SSCs will be potentially exposed to this hazard. When the truck is moving on its route, the roadbed in all cases is below the level of the ISFSI pad, which ensures that even if there was a tank rupture the fuel would not run toward the ISFSI.

In conclusion, all SSCs will be able to continue performing their intended safety functions in the event of a tanker fire.

NRC Comment No. 2 PG&E should provide information to clarify whether the area surrounding the storage pads would be covered with non-combustible materials.

PG&E Response The final ISFSI site development design has not yet been completed. However, the restricted area not covered by the storage pads will be covered with crushed rock approximately 12 inches deep. The outer fence will be separated from the inner fence by a distance of approximately 20 ft. The isolation zone (i.e., the region between the fences) will also be covered with crushed rock approximately 12 inches deep. A maintenance program will control any significant growth of vegetation through the crushed rock. Therefore, the surface of the restricted area will be noncombustible.

PG&E Letter DIL-03-005 Sheet 2 of 5 NRC Comment No. 3 PG&E should clearly specify the design-basis air overpressure criterion for aml structures, systems, and components important to safety.

PG&E Response See response to NRC Comment No. 4.

NRC Comment No. 4 PG&E should provide additional analysis to demonstrate that all safety-related structures, systems, and components would be able to continue their functions after being subjected to explosion-generated missiles, if an air overpressure criterion higher than 6.9 kPa [I psi] is used. This comment is same as RAI 15-8.

PG&E Response PG&E's response to RAI 15-8, submitted in PG&E Letter DIL-02-009, dated October 15, 2002, indicated that for all important-to-safety SSCs,.administrative controls would be relied upon to ensure that: (a) onsite gasoline-powered vehicles would be kept a, sufficient distance from important-to-safety SSCs such that any explosion overpressure is less than 1 psi: or (b) a probabilistic risk assessment (PRA) is performed to demonstrate that the risk meets Regulatory guide (RG) 1.91 acceptance criteria; or (c) diesel-powered vehicles are required. In addition, administrative controls will be relied upon to ensure all other potential transport hazards are either: (a) at a sufficient distance from the storage pad (at all times) and the CTF (while transferring a loaded MPC) to ensure the explosion overpressure is less than the design-basis overpressure of 1 psi, or (b) provide determination of acceptable risk based on RG 1.91.

The Diablo Canyon ISFSI SAR is revised to specify a design basis for shipment of multiple compressed gas bottles. Refer to draft SAR Table 3.4-1 in Enclosure 3 to PG&E Letter DIL-03-005, dated March 31, 2003. Per RG 1.91, any hazards that do not meet the 1-psi or the acceptable risk criteria would require explosion-generated missiles to be evaluated. However, based on PRA Calculation No. PRA01-01, Revision 3 (Attachment 2), and administrative controls being provided as discussed above, all hazards for the ISFSI pad or transporter will meet the RG 1.91 criteria or methodologies. As a result, no further missile analysis is necessary.

NRC Comment No. 5 PG&E did not sum the annual frequency of explosion hazard from each individual source to estimate the cumulative hazard to the proposed facility, as recommended in Regulatory Guide 1.91 (U.S. Nuclear Regulatory Commission, 1978).

PG&E Letter DIL-03-005 Sheet 3 of 5 PG&E Response Per RG 1.91, "If the substance in question is shipped on more than one transportation mode near the plant, exposure rates calculated for the modes should be summed." As a result, PG&E has revised the PRA Calculation No. PRA01-01 (Attachment 2) to sum the risk from the same substance transported by different modes per the RG. The results show that the risk remains insignificant from all hazards.

The NRC comments implied that RG 1.91 also requires the exposure potential for all hazards along the transport route path to be summed together to get a total potential exposure for the transport route. Although PG&E has reviewed RG 1.91 and has not identified this requirement, PG&E has evaluated the total risk along the transport route.

From PRA Calculation No. PRA01-01, Revision 3, the total risk for all hazards along the transport route is equal to: 6.1e-08 (parking lot cars)+ 3.7E-07 (hydrogen facility) +

3.7E-7 (acetylene facility) + 5.9E-08 (transformer) = 8.6E-7lyear.

NRC Comment No. 6 PG&E (Afzali, 2002) used loaded transporter speeds that are three times higher than that given in Table 3.4-4 of the SAR to perform the probabilistic explosion hazard assessment and, as a result, the estimated annual frequencies of different events in Afzali (2002) are not conservative.

PG&E Response The 0.4 mph transporter maximum speed shown in SAR Table 3.4-4 is correct. SAR Sections 2.2.2 and 8.2, SAR Tables 3.4-1, 7.4-1, 7.4-2, 7.5-3, 7.5-4, 8.1-1, and 8.2-11, and Environmental Report Table 5.1-1 have been revised based on the speed. Draft sections and tables are being submitted with this response as Enclosure 3. In addition, the nonproprietary and proprietary versions of the Holtec dose analysis Hl-2002563 have been revised to reflect this speed and a copies of these calculations are being provided in Enclosures 2 and 1 of PG&E Letter DIL-03-005, respectively.

The NRC has indicated that they believe that the PRA calculation provided previously did not use the 0.4 mph speed, but used a speed that was three times faster than the 0.4 mph value. PG&E believes that this is a misunderstanding because PRA Calculation No. PRA01-01, Revision 2 states that the parking lot is approximately 1,000 ft long and that the transporter would conservatively be in that area less than one hour. Using that data it equates to a speed of about 0.19 mph. Based on this, PRA Calculation No. PRA01-01, Revision 2 was very conservative. However, PRA Calculation No. PRA01-01 has been revised to consider more current data and to clarify the results. Refer to Attachment 2.

PG&E Letter DIL-03-005 Sheet 4 of 5 NRC Comment No. 7 PG&E should clarify the following:

(1) Maximum fuel capacity of all gasoline powered onsite vehicles.

(2) Technical bases for two different TNT equivalent calculations for the same amount of gasoline.

PG&E Response The PRA Calculation No. PRA01-01 has been revised to conservatively use a 50-gallon fuel capacity for all onsite gasoline powered vehicles. SAR Sections 2.2.2.2, 8.2.5 and 8.2.6 have been revised to reflect this value.

The PRA Calculation No. PRA01-01, Revision 2 was based on RG 1.91 methodologies and used the equivalent energy number from the Army Manual TM 5-1300 (referenced in RG 1.91), which is 2531.5 btu/lb. Holtec used an equivalent energy number from the Perry's Chemical Engineers Handbook, which is 4.5E+6 joules/kg = 1933.39 btu/lb. As

-a result, the Holtec number is approximately 76 percent of the PRA value. This created the inconsistency, however, use of a smaller number for TNT detonation energy in the Holtec calculation is at least as conservative as the PRA since the equivalent weight used is higher and the separation distance is less in the Holtec calculation.

As a result, both calculations are considered conservative and meet RG 1.91 guidance.

Revision 3 of PRA Calculation No. PRAOI-01,(Attachment 2) also uses the TNT equivalency for 50 gallons of 66.7 kg [147.13 lb] TNT, which is based on RG 1.91 and the Army Manual TM 5-1300.

NRC Comment No. 8 HI-STORM 100storage cask do not have sufficient safety margin from potential detonation of the fuel tank of an onsite gasoline-powered vehicle with a capacity of 189 1 f50 gal], as assumed by Afzali (2002), at the controlled area boundary, assumed to be 15 m [50 ft] from the storage pads.

PG&E Response Per PG&E's response to NRC Comment No.4 above, no significant hazard will exist for the ISFSI that would threaten the margin of safety of the ISFSI or its SSCs. This is based on the administrative controls that will be in place to control all hazards and to maintain the risk to an acceptable level per RG 1.91 requirements and methodologies.

PG&E Letter DIL-03-005 Sheet 5 of 5 NRC Comment No. 9 PG&E has not provided an analysis demonstrating that the transporter will be able to withstand and carry out all of its safety functions when subjected to an explosion with an air overpressure of 2.65 MPa (384 psi].

PG&E Response Per PRA Calculation No. PRA01-01, Revision 3, the loaded transporter and its SSC's will not be exposed to credible hazards that exceed RG 1.91 acceptance criteria or methodologies. All potential hazards will either be eliminated; meet the setback requirements of a 1-psi overpressure exposure; or the potential exposure will be determined to be acceptable in accordance with the methodologies in RG 1.91.

NRC Comment No. 10 PG&E analysis does not consider more than one 26.51[7 gal] propane tank or one acetylene bottle being transported near the proposed storage facility.

PG&E Response There are no storage facilities for-explosive gases within 1,000 ft of the ISFSI pad and there will be no use of any equipment powered by propane within 1,000 ft of the ISFSI pad. However, some amount of explosive gas bottles may be transported past the ISFSI facility.

In PRA Calculation No.PRAOI-01, Revision 3, a truck with multiple gas bottles was evaluated based on the hazard being bounded by the risk of the 2,000-gallon tanker truck explosion. The basis for this reasoning is that there are no bottle storage facilities close to the ISFSI pad, the actual number of bottles that would potentially be transported past the ISFSI area is very limited, and the potential for an explosion is considered limited to a transportation accident similar to a large truck accident. In these accidents, the bottles usually fail from valves being broken during a rollover or the bottles being pierced by some external object. PRA Calculation No. PRAO1-01, Revision 3, assumes that the transportation of the bottles would be under similar traffic controls as for the 2,000-gallon tanker truck and that the gas bottles would be properly restrained.

As shown in the PRA calculation, the risk of a truck accident explosion is 1.57E-7, which is not risk significant. If every truck crash is conservatively considered as causing a bottle failure that will potentially affect the ISFSI, the risk is not significant per RG 1.91.

PG&E Letter DIL-03-005 Sheet 1 of 1 List of Attachments Attachment No.

TITLE 1

Determine the Heat Gain of the HI-TRAC Casks Due to a 2000 Gallon Gasoline Fire, Calculation M-1052 2

Risk Assessment of Dry Cask/Spent Fuel Transportation within the DCPP Owner Controlled Area, Probabilistic Risk Assessment, Calculation File No. PRA01-01, Revision 3

ATTACHMENT I

69-20132 04110100 NUCLEAR POWER GENERATION CF3.1D4 AlTACHMENT 7.2 TITLE:

DESIGN CALCULATION COVER SHEET Page 1 of 2 Unit(s):

UO Responsible Group:

PTEB No. of Pages 6

File No.:

131.9 Calculation No.: M-1052 Design Calculation YES 0 Quality Classification G

5 NO 3 Structure, System or Component: IFISI

Subject:

Determine heat load on Hi-Trac casks from a 2000 gallon gasoline fire.

d*

Electronic calculation YES a NO 0 Comnputer Model Computer ID l

Program Location Date of Last I __Change Registered Engineer Stamp: Complete A or B NTI l: Update DCI promptly after approval.

wnTD v: Forward electronic calculation file to CCTG for uploading to EDMS.

calccover.doc

69-20132 04/10/00

('2 Page 2 of 2 CF3.D4 ATTACHMENT 7.2 TITLE: DESIGN CALCULATION COVER SHEET RECORD OF REVISIONS CALC No. M-1052 Rev Status Reason for Revision Prepared LBIE LBIE Check LBIE Checked Supervisor Registered No.

By:

Screen Method*

Approval Engineer Remarks Initials/

Yes/

PSRC PSRC Initials/

Initials/

Signature/

LAN ID/

No/

Mtg.

LAN ID/

Date NA NA No.

Date Date Date Date 0

F new calc 0 Yes 0 Yes 0 A 60-" A SPB8 0 No 0 No 0 B FCL2 LLF3 SPB8 O NA O NA aC (4C/o 7(a/ov

//i D Yes D Yes Q A O No D No Q B O NA O NA C

C Q Yes D Yes O A O No D No 0 B NA O NA C

C Yes D Yes Q A O No O No 0 B O NA O NA C

C D Yes D Yes D A D No O No a B NA O NA OC

Check Method: A: Detailed Check, B: Alternate Method (note added pages), C: Critical Point Check caker.d=c

Calculation Sheet CALC.NO.

M-1052 REV.NO.

0 Unit( )1

)2(x)1&2 SHEET NO.

I OF 6

SUBJECT Determine the heat gain of the HI-TRAC casks due to a 2000 gallon gasoline fire 1.

Purpose:

This calculation will determine the amount of heat transferred to the storage or transfer casks due to a gasoline fire of 2000 gallons. The heat gain as a result of this fire will be compared to the design basis fire calculated by the HI-TRAC manufacturer (HOLTEC).

2.

Background:

The licensing submittals for the Dry Cask Storage project (ref 11.2.a) included the commitment to determine the fire loading impact on the storage or transfer casks as a result of a 2000 gallon gasoline fire. This 2000 gallon fire would be the result of a hypothetical tanker truck accident where the entire contents of the tank spill onto the road and ignites.

3.

Assumptions:

a. The entire 2000 gallon load of gasoline instantaneously leaks from the truck and forms a pool that is a minimum of 8 meters in diameter, and

.15 meters deep (6 inches).

b. The truck is parked on a smooth surface that will not contain the fuel in a pool smaller than 8 meters in diameter.

c. The gasoline in the pool is ignited as soon as the pool forms.

d. There are no wind effects considered in the calculation.

e. Conduction is not a heat transfer mechanism.

f. The only heat source is the burning 2000 gallons of gasoline.

g. The ambient temperature of the storage or transfer casks is 70 deg F.

h. The transfer cask is located right next to the pool fire.

Calculation Sheet CALC. NO.

M-1052 REV. NO.

0 Unit ) 1 2 (x) 1 & 2 SHEET NO.

2 OF 6

SUBJECT Detenrine the heat gain of the HI-TRAC casks due to a 2000 gallon gasoline fire

4.

Inputs:

a. The maximum allowed heat input is 906,213 BTUs (ref 11.1.a)

b. The 2000 gallon truck size is based on largest tanker truck used a DCPP.

c. The density of gasoline was obtained from Marks' Standard Handbook

5.

Methodology:

The method used in this calculation is derived from the discussion on pool fires in the Fire Protection Handbook, Sixteenth Edition, Section 21, Chapter 6. The approach is to first determine the duration of the fire based on the fuel supply, and burning rate. The total heat gain of the transfer cask due to convection, and radiation is then calculated using heat transfer methodology. The Hi-TRAC transfer cask is bounding due to lower amount of heat input allowed.

6.

Acceptance Criteria:

The postulated fire will not pose a heat load threat to the transfer cask if the total amount of heat transferred does not exceed 906,213 BTUs.

7.

Body of the Calculation:

Fire Duration; M = burning rate for gasoline pool fires = 0.055 kg/m2-s (ref 11.1.b pg 21-

37) 2 Where m = area of pool 0.055 kg/M2-s x 60 sec/min = 3.3 kg/m2-min area of pool = 7cr2 r = d/2 = 812 = 4 meters area = i x 42 = 50.2 m2 3.3 kg/M2-min x 50.2 m2 = 165.9 kg/min bum rate

Calculation Sheet CALC. NO.

M-1052 REV. NO.

0 Unit 11( )2(x)1&2 SHEET NO.

3 OF 6

SUBJECT Determine the heat gain of the HI-TRAC casks due to a 2000 gallon gasoline fire Equating the burn rate to the fuel load will determine the fire duration fuel load = 2,000 gallons 2,000 gals x 6.2 lbs/gal = 12,400 Ibs 12,400 Ibs *. 2.2 lb/kg = 5,660.3 kg 5,660 kg + 165.9 kg/min = 34.1 minutes fire duration Heat Transfer; Radiation The radiant heat impinging on a surface due to a pool fire can be calculated by the following; Q XrxbhxMxA.16L2 (ref 11.1.b, pg2l-39)

Where; Q = heat transfer in kw X, =.10 (ref 11.1.b, pg 21-41) ah = heat release = 43.7 mj/kg (ref 11.1.b, pg 21-37)

A = surface area of transfer cask (height -196 in, diameter - 94.6 in)

A = 196 x or x 94.6 = 58250 in2 x 6.45613 x 0l m2nin2 = 37.6 M2 (ref 11.1.d)

M = burning rate for gasoline pool fires L = I/D where I is the distance between the fire and the cask, and D is the diameter of the pool fire L=4./8=.5 Q =.10 x 43.7 mj/kg x.055 kg/m2-s x 37.6 m2. 16 x.52 = 2.26 kw 2.26 kw x 3412.9 btu/hr-kw = 7710 btu/hr 7710 btu/hr x 34.1 min -. 60 min/hr = 4,382.3 btus Therefore, the heat input from radiation is 4,382.3 btus

Calculation Sheet EMAILS CANYON 0

CALC. NO.

M-1052 REV. NO.

0 Unit(

1(

2(x)1&2 SHEET NO.

4 OF 6

SUBJECT Deermine the heat gain of the HI-TRAC casks due to a 2000 gallon gasoline fire Convection The heat gain due to convection can be calculated by the following; Q=AxfxAtbtulhr(ref11.1.c, pg 10-12)

Where; A = surface area of transfer cask (height -196 in, diameter - 94.6 in)

A = 196 x 7t x 94.6 = 58250 in2. 144 in2lft2 = 404.5 ft2 (ref 11.1.d) f = film coefficient for air At = temperature differential between the fire and the cask the flame temperature for gasoline is 2150 deg F (ref 11.1.b, pg 21-37)

At = 2150 -70 = 2080 deg F f must be calculated from the following relationship:

f x L/k = C(Ng x N,)33 (ref 11.1.c, pg 10-12) f = C(N. x Nr) 33 x k/L where; L = height of the transfer cask = 16.3 feet K =.04 btulhr-ft-deg F (ref 1.1.c, pg 10-24)

C =.13 (ref 11.1.c, pg 10-12)

N, =.74 (ref 11.1.c, pg 10-24)

Ng=L3 Xp2Xp XAtx pF At=t

,=2150-0=2150degF,degR p=.0202 1bmlft3 (ref 11.1.c, pg 10-24)

,B =.510 EE-3 (ref 11.1.c, pg 10-24) g=32.2

= 3.00 EE-5 (ref 11.1.c, pg 10-24)

Ng = 16.33 X.02022 x.510 EE-3 x 2150 x 32.2

3.00 EE-52 = 6.98 EE10 f =.13 x (6.98 EE10 x.74)33 x.04/12 = 1.49 btu/hr-ft2 -deg R from above, Q = A x f xAt Q = 404.5 ft2 x 1.49 x 2080 deg F = 1,253,626.4 btulhr The fire duration was previously determined to be 34.1 minutes Therefore the heat transfer is 1,253,626.4 btu/hr x 34.1 min + 60 min/hr =

712,477.7 btus.

Calculation Sheet 96aNIo eraseN CALC. NO.

M-1052 REV. NO.

0 Unit (

1 2(x 1 &2 SHEET NO.

5 OF 6

SUBJECT Determine the heat gain of the HI-TRAC casks due to a 2000 gallon gasoline fire Adding the radiant and convective loads yields, Total = 712,477 + 4382.3 = 716,860 btus.

8.

Results:

This calculation determined the total heat gain experienced by the Hl-TRAC due to a 2000 gallon gasoline fire is 716,825 btus.

9.

==

Conclusion:==

The results of this calculation show that a 2000 gallon gasoline fire will not generate enough heat to exceed the 906, 213 btu design basis heat load determined acceptable by the manufacturer (HOLTEC). This calculation is very conservative and does not consider any separation distance between the fire and the transfer/storage cask, even though administrative controls will maintain a minimum of 50 feet separation.

10.

Impact Evaluation:

The licensing basis impact is contained in the Dry Cask Project submittals.

11.

References:

11.1. Input

References:

a. Holtec letter from Evan Rosenbaum to Eric Lewis dated 02/15/2001.

b. NFPA Fire Protection Handbook, sixteenth edition.

c. Mechanical Engineering Review Manual, seventh edition

d. Table 4.2-3, pg 72 of SAR 4.2

Calculation Sheet CLC. NO.

M-1052 REV. NO.

0 Unit( )1( )2(x)1&2 SHEET NO.

6 OF 6

SUBJECT Determine the heat gain of the HI-TRAC casks due to a 2000 gallon gasoline fire Output

References:

a. PG&E 10CFR72 License Application SAR Section 8.2.5 Other:

None.

12.

Enclosures and Attachments:

None.

ATTACHMENT 2

PACIFIC GAS & ELECTRIC COMPANY PROBABILISTIC RISK ASSESSMENT CALCULATION FILE NO. PRAOI-01 Revision 3

SUBJECT:

Risk Assessment of Dry Cask/Spent Fuel Transportation within the DCPP Owner Controlled Area PREPARED BY:

VERIFIED BY:

E. G. Davis DATE:

DATE: ?,i-6/03 VERIFIED IN ACCORDANCE VITH: CF3.1D15

/i Af.fA DATE:

3_/_2_t__ 3 APPROVED BY:

NOTE: This Document contains assumptions and results that are the basis for parts of SAR Sections 2.2.2.3 and 8.2.6, Modification of this document will require evaluation under 10CFR72.48.

This file contains:

15 pages

CALCULATION FILE PRAOI01 Rev.3 Sheet 2 RECORD OF REVISIONS REV. 0 Original Calculation.

REV. I In this revision, the risk of damage to dry casks due to explosion of an acetylene carrying truck is added to the analysis of risk of damage due to other hazards. This is done to support response to RAI 15-12.

REV. 2 In this revision, the date of the final HI-TRAC evaluation was changed to reflect the actual issuance date (March 6, 2001). The February 2001 date was a preliminary draft, which was the version available at the time this PRA calculation was initially issued (Revision 0 dated February 28, 2001).

REV. 3 In this revision, the evaluations have been revised to address the RG 1.91 1 PSI criteria where possible, to use more up to date vehicle crash data, and to clarify the previous evaluations.

INTRODUCTION Per the requirements of RG 1.91, Revision 1, explosive hazards in transit are to be evaluated to ensure that there is no significant potential for damage to any safety related components. This RG also provides guidance that allows determination of risk based on location and amount of the hazard. When using a potential design limit of 1 psi for a component, if the risk is found to be less than 1 06 using conservative data or less than 10- using more site-specific data, then the risk is considered non-significant and acceptable. The methodology in the RG allows a hazard to be dismissed as non-significant if the 1 psi criteria is met based on a setback distance and volume of hazard.

However, if the setback distances or volume of a hazard will potential exceed the I psi criteria the RG requires further evaluation based on probability and risk. Based on this the RG guidance risk evaluations were performed to assess the risks of an explosion causing damage to the dry casks while being transported, or while the casks are stored at the Independent Spent Fuel Storage Installation (ISFSI). As a result, several explosion sources were identified that potentially don't meet the setback and volume limitations and required further evaluation. Specifically, six bounding explosion sources were identified requiring this risk evaluation:

1. An explosion of a parked vehicle's fuel tank while the HI-TRAC transporter is passing on the road near or through the parking lots on its route to the ISFSI. (Lot 8 only of concern if the lot is occupied with vehicles)

2. A hydrogen explosion in the protected area from the hydrogen tanks while the HI-TRAC transporter is in the vicinity.

3. An explosion of a 2000-gallon fuel truck while i passes within 812 feet of the ISFSI facility.

CALCULATION FILE PRA01D11 Rev.3 Sheet 3

4. An explosion of a Unit 2 transformer while the HI-TRAC transporter is passing on K-the elevated road inside the Radiologically Controlled Area (RCA).

5. Explosion of one of the 140 vehicles per day (with a maximum of 50 gallons of gas on board) that pass within 238 feet of the ISFSI facility.

6. An explosion at various compressed gas cylinder storage facilities and the movement of multiple propane or acetylene bottles past the ISFSI area.

In addition, similar hazards that are potentially transported by different means are summed in this calculation to show the total exposure potential for that hazard per RG 1.91.

DISCUSSION One of the requirements in the Diablo Canyon Spent Fuel Storage Installation (ISFSI)

Safety Analysis Report (SAR) is the evaluation of explosions. During the evaluation process, six bounding potential explosion hazards were identified as needing a risk evaluation to ensure that they meet the RG 1.91 1 psi limit, or that they are acceptable risks per the methodologies provided in RG 1.91. These six explosion hazards are:

1. Explosion of a parked vehicle, while the HI-TRAC transfer cask transporter is passing in the vicinity of the parking lots.

2. Explosion of the bulk hydrogen storage facility while the cask transporter is in the area.

3. Explosion of one of the 2000-gallon fuel trucks while the truck is passing within 812 feet of the ISFSI facility.

4. Explosion of a Unit 2 transformer while the cask transporter is moving through the RCA.

5. Explosion of one of the 140 vehicles per day (with a maximum of 50 gallons of gas on board) that pass within 238 feet of the ISFSI facility.

6. Explosion of various compressed gas cylinder storage facilities and the explosion of multiple propane or acetylene bottles moving past the ISFSI facility.

ACCEPTANCE CRITERIA Regulatory Guide 1.91 (Reference 1) contains guidance on acceptable risk from explosions for nuclear plants. Regulatory Guide 1.91 states, "if estimates of explosion rate, frequency of shipment, and exposure distance are made on a realistic or best estimate basis, an exposure rate less than 1017 per year is sufficiently low. If conservative estimates are used, an exposure rate less than 104 is sufficiently low.'

CALCULATION FILE PRADI-01 Rev.3 Sheet 4 ASSUMPTIONS

1. It is estimated that there will be eight shipments per year of the HI-TRAC transfer casks from the spent fuel building to the ISFSI.

2. The hydrogen tanks will not be filled, tested or vented while the HI-TRAC transfer cask transporter is in the vicinity of the hydrogen tanks. (Ref.: AR A0524878)

3. Additional physical barriers will be erected to prevent the transporter from getting too close to the hydrogen tanks. (Ref.: AR A0524878)

4. The 2000-gallon trucks will be maintained at least 812 feet from the transporter path during spent fuel transport. See Assumption 6 below for specific setback required to meet RG 1.91 criteria of 1 psi. (Ref: AR A0524878)

5. A 2000 gallon truck will pass by the ISFSI facility once per day. (Reference 9)

6. The 2000-gallon truck does not stop within 812 feet of the ISFSI. Administrative controls will designate parking areas for this vehicle. However, if there is a mechanical breakdown within the 812 feet of the ISFSI, administrative controls will require immediate removal to its designated parking area or to a distance of more than 812 feet from the ISFSI. (Ref: AR A0524878).

The setback distance is based on the maximum amount of gasoline in the truck being 2000 gallons. Using the RG 1.91 methodology, reference (1) the required setback or separation distance (at which the pressure wave is equal to or less than 1 psi) can be calculated.

R ->-kW"/3 Where R is the setback in feet WtKt is the explosion hazards in equivalent pounds on TNT k is 45 when R is in feet and W is in pounds Using the formula in reference (2) section 1) for 4000 gallons to determine the equivalent lb-TNT and using a 2000 gallons capacity, Wnt =11770.6*2000/4000 = 5885.3 lb-TNT And R 2 (45)(5885.3)113 R 2 812 feet in setback or separation distance

CALCULATION FILE PRAOI-01 Rev.3 Sheet 6 Note: the Wtm calculated in reference (2) is conservative as it assumes "that b

100% of the liquid has been vaporized and mixed with air between the upper and lower flammability limits" and that "No credit for partial shielding between the casks and the location of the explosion is considered."

7. The speed of all vehicles in the area of the ISFSI will be administratively controlled to less than 25 miles per hour. (Ref: AR A A0524878)

8. No vehicles will be allowed to pass the 2000-gallon truck in either direction while it is in the 812 foot exclusion area around the ISFSI facility. (Ref. AR A0524878)

9. All gasoline powered vehicles, which must pass within 238 feet from the closest point of an ISFSI pad will be under administrative controls. The administrative controls will control speed, movement and provide designated parking areas outside the 238 feet for these vehicles. If there is a mechanical breakdown within the 238 feet, administrative controls will also require immediate removal to its designated parking area or to a distance of more than 238 feet from the ISFSI. (Ref: AR A0524878)

The setback distance is based on the maximum amount of gasoline in any vehicle fuel tank is 50 gallons. Using the RG 1.91 methodology the required setback or separation distance (at which the pressure wave is equal to or less than I psi) can be calculated.

R > kW1"3 Where R is the setback in feet Wun is the explosion hazards in equivalent pounds on TNT k is 45 when R is in feet and W is in pounds Using the formula in reference (2) section 1) for 4000 gallons to determine the equivalent lb-TNT and using 50 gallons as the tank maximum capacity, Wmt = 11770.6

50/4000 = 147.13 lb-TNT And R > (45)(147.13)Ž3 R > 238 feet in setback or separation distance Note: the Wtnt calculated in reference (2) is conservative as it assumes "that 100% of the liquid has been vaporized and mixed with air between the upper and lower flammability limits" and that 'No credit for partial shielding between the casks and the location of the explosion is considered."

CALCULATION FILE PRAOI01 Rev.3 Sheet 6

10. It is assumed that no more than 140 gasoline-powered vehicles per day will pass by the ISFSI facility. The actual number of vehicles that pass the ISFSI facility per day is approximately 50 vehicles and they pass it in two directions. However, 140 vehicles was used to be conservative.

11.The 4000-gallon fuel truck will be not allowed in the owner-controlled area during spent fuel transport. (Reference 2) (Ref: AR A0524878) 12.Administrative controls will prevent the 4000-gallon truck from moving up the hill and passing by the ISFSI facility. (Ref: AR A0524878)

13. Physical and administrative controls will be in place to prevent vehicle movement with in 238 feet of the moving cask transporter. (Ref: AR A0524878) 14.The cask transporter, while loaded with a HI-TRAC transfer cask, will be in the vicinity of the bulk hydrogen storage facility less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during each shipment from the spent fuel building to the ISFSI facility.

15.The cask transporter, while loaded with a HI-TRAC transfer cask, will be in the vicinity of the parking lots less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during each shipment from the spent fuel building to the ISFSI facility. This estimated time is considered very conservative considering that the parking lots are approximately 1000 feet in length, the transporter moves at about 0.4 mph, and that the transporter will pass straight through these areas with minimal maneuvering required. At the 0.4 mph speed the transporter would be in the area less than 30 minutes. This time has been doubled to be conservative in this evaluation.

16.Administrative Procedure AD4.ID1 requires that all bottles w-ithin the RCA and outside the protected area will be appropriately secured and chained so they are not hazards. A walkdown will occur prior to the transporter beginning its trip, from the Spent Fuel Pool Building to the ISFSI, to confirm the bottles are appropriately chained. (Ref: AR A0524878) 17.Walkdowns will occur in the parking lots, which have the potential to affect the transportation route, to assure no potential explosive hazards (such as leaking gas tanks) exist prior to any movement of the loaded cask transporter in the vicinity of these parking lots. (Ref: AR A0524878) 18.Additionally, a walkdown will occur outside the protected area prior to movement of the cask transporter, while loaded with a HI-TRAC transfer cask, to evaluate any transient hazardous material located along the pathway. (Ref: AR A0524878)

19. It is assumed that all gas bottles transported past the ISFSI facility will be appropriately secured in the upright position within the transporting vehicle.

CALCULATION FILE PRA01D-1 Rev.3 Sheet 7 CALCULATIONS Hydroqen Tank Explosion The bulk hydrogen facility contains 6 hydrogen tanks. The hydrogen tanks and hydrogen piping contain relief valves, which are vented to atmosphere. Because of the design, a hydrogen explosion is considered almost incredible. However, hydrogen fires are credible, and appear in the EPRI Fire Events Database (Reference 5), including fires caused by a stuck open or leaking relief valve.

The EPRI Fire Events Database gives an annual frequency of Hydrogen fires of 3.2e-3/year (Reference 5). It is conservatively assumed the entire frequency of fires can be assigned to the bulk tank facility. Thus, the hourly frequency of hydrogen fires at the bulk tank facility is:

Hydrogen Fire Frequency = 3.2e-3/yr

yr/8760 hrs = 3.7e-7/hr Because of the design of the hydrogen system, which does not allow hydrogen to accumulate in confined spaces, there is an extremely low probability of a hydrogen explosion, even if a hydrogen fire occurs. If we conservatively assume a conditional probability of 0.1 that a hydrogen explosion occurs, given a hydrogen fire has occurred, then the Hydrogen Explosion Frequency is:

Hydrogen Explosion Frequency = 3.7e-7/hr

0.1 = 3.7e-8hr The hydrogen explosion is a concern when the HI-TRAC transfer cask transporter is in the vicinity of the hydrogen tanks. As noted in the assumptions section, the transporter should be in the vicinity of the hydrogen tanks less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for each shipment, with eight shipments per year. It will be assumed that the transporter will be in the vicinity of the hydrogen tanks a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per year.

On a yearly basis, the probability (exposure rate) of a hydrogen explosion potentially damaging the HI-TRAC cask is:

3.7e-8hr

10 hr/yr = 3.7e-7/yr According to Regulatory Guide 1.91, "if conservative estimates are used, an exposure rate less than 10-6 per year is sufficiently low."

2000 Gallon Truck Explosion There is data available from the Department of Transportation on truck crashes and truck fires. Table 38 of the 2001 NHTSA statistics (Reference 6) show that in 2001 large trucks were involved in 429,000 crashes. Table 1 of the 2001 FMCSA data (Reference 7) shows that in 2001 large trucks traveled approximately 207,686 million

CALCULATION FILE PRAOI-01 Rev.3 Sheet 8 miles. This results in a 'Large trucks involvement rate" of 207 per 100 million miles.

This data includes crashes from all hazards including weather and natural causes.

Per the assumption that no other vehicle will be traveling near the ISFSI when the 2000-gallon truck is in motion it is assumed that only a single vehicle accident can occur. Based on Table 46 of the 2001 NHTSA statistics (Reference 6), single vehicle crash data, shows a total of 96,000 crashes occurred in 2001, which is approximately 22 percent of all large truck crashes.

Per the listed assumptions the speeds in the area of the ISFSI are to be controlled below 25 miles per hour at all times. Based on this, Table 29 of 2001 NHTSA statistics (Reference 6) shows that for all single vehicle crashes approximately 31 percent are less than 30 miles per hour. Although there is no specific data provided for large trucks as compared with all vehicles, for our calculations we are conservatively assuming that for trucks it is 40 percent of all accidents are at less than 30 miles per

hour, Although the Table 29 data does not provide a direct correlation between the large truck data and the all vehicle data, the use of 40 percent is believe reasonable. In support of this assumption, Table 26 of the 2001 FMCSA data (Reference 7) provides data that shows that the percentage of fatal crashes involving a single large truck at 25 miles or less is about 6.8 percent of all fatal large truck crashes. Table 29 of 2001 NHTSA statistics (Reference 6) shows that for all fatal single vehicle crashes approximately 13 percent are at less than 30 miles per hour. For speeds above this level the percentage of fatal accidents involving large trucks continues to remain below the percentage for all vehicles. Based on this trend it is reasonable to consider that the percentage of single large truck accidents under 30 mph will remain below the percentage for all vehicles crashes. However, to ensure this is conservative this evaluation used a 40 percent figure as discussed above.

Table 38 of the 2001 NHTSA statistics (Reference 6) shows that 0.5 percent of all large truck crashes result in fires. Thus, the frequency of truck fires is:

207

0.22

0.40 / 100e6 *0.005 = 9.1 le-10mile For the purposes of this analysis, it is conservatively assumed that all truck fires result in an explosion. Thus, the explosion rate for truck fires is 9.1le-10/mile.

Regulatory Guide 1.91 provides the following equation for use in the evaluation of explosions:

r = nfs, where n= explosion rate for the substance and transportation mode in question in explosions per mile f = frequency of shipment for the substance in question in shipments per year, and s = exposure distance in miles

CALCULATION FILE PRADI-01 Rev.3 Sheet 9 tf' It is assumed that the ISFSI exposure distance, s is conservatively 2500 feet. As noted above in the assumptions section, the frequency of shipments is 1 per day, or 365 per year.

Thus, r = 9.11e-10mile

365

250015280 = 1.57e-7/year Note: This is conservative because we are assuming that all vehicle fires lead to an explosion 50 Gallon Gasoline Powered Vehicle Explosion near ISFSI There is data available from the Department of Transportation on all vehicle crashes and all vehicle fires. The 2001 National Statistics summary of the NHTSA statistics (Reference 6) show that in 2001 all motor vehicles were involved in 6,323,000 total crashes. It also shows that motor vehicles traveled approximately 2,781,462 million miles. This results in a "vehicle involvement rate" of 227 per 100 million miles." This data conservatively include all motor vehicles including large trucks and from all hazards including weather and natural causes.

Per the assumption that vehicle travel will be limited within the 238 feet setback distance from the closest part of the ISFSI facility and will result in no vehicle being allowed to pass each other within that setback distance As a result, only a single vehicle accident can occur. Based on this Table 29 of the 2001 NHTSA statistics (Reference 6) provides single vehicle crash data, which shows a total of 1,907,000 crashes in 2001, which is approximately 30 percent of all vehicle crashes.

Per the listed assumptions the speeds in the area of the ISFSI are to be controlled below 25 miles per hour at all times. Based on this, Table 29 of 2001 NHTSA statistics (Reference 6) shows that for all single vehicle crashes approximately 30 percent are less than 30 miles per hour.

Table 38 of the 2001 NHTSA statistics (Reference 6) shows that 0.1 percent of all vehicle crashes result in fires. Thus, the frequency of vehicle fires is:

227

0.30

0.30/ 100e6

0.001 = 2.04e-10/mile.

For the purposes of this analysis, it is conservatively assumed that all vehicle fires result in an explosion. Thus, the explosion rate for vehicle fires is 2.04e-1 0/mile.

Regulatory Guide 1.91 provides the following equation for use in the evaluation of explosions:

r = nfs, where n= explosion rate for the substance and transportation mode in question in explosions per mile f = frequency of shipment for the substance in question in shipments per year, and

CALCULATION FILE PRAD"I-Rev.3 Sheet 10 s = exposure distance in miles It is assumed that the ISFSI exposure distance, s is approximately 400 feet, which is approximately double the exposure length of the ISFSI facility within the setback of 238 feet from any point on the road.

As noted -above in the assumptions section, the frequency of 140 per day, or 51,100 per year.

Thus, r = 2.04e-10/mile

51,100

400/5280 = 7.89e-7/year Summing the potential exposure rates from a gasoline hazard for the ISFSI facility per RG 1.91:

Total potential exposure = 1.57e-7/year + 7.89e-7lyear = 9.46e-7/year According to Regulatory Guide 1.91 Revision 0, if conservative calculations are used, an exposure rate, of less than 1e-6/year is acceptable.

For explosions from gasoline hazards at the ISFSI, all of the above date and results are very conservative as they are based on national highway statistics and do not take into consideration the very controlled nature of the activities at the ISFSI, the limited maneuvers performed by any vehicle in the area of the ISFSI, the normal speeds which are below 25 miles per hour, the over stated potential for fires resulting in explosions, and the number of vehicles in the area.

Parked Vehicle Explosion Risk Vehicle (defined as non-large truck) explosions almost always are the result of a crash or collision, and rarely, if ever occur in parked vehicles. Based on the known history of DCPP there has been only one fire in a vehicle in the parking lots. The fire ignited at the time the car was being started and occurred in the ignition system. That fire would not be considered a credible scenario during fuel transport because of administrative controls. Prior to a loaded transporter being moved in the vicinity of a parking lot, administratively controlled walk downs of the parking lots will be performed looking for any possible fire or explosion hazards such as leaking gas. During the transportation of the HI-TRAC transfer cask, administrative and physical controls will be in place to prevent vehicles from moving within 238 feet of the transporter, while it is in the vicinity of the parked vehicles. Thus, it is concluded that it is extremely unlikely and not credible for a stationary vehicle to explode while the transporter is in transit by the parking lots.

In an effort to provide support for this conclusion, a search was conducted for industry data concerning the frequency of explosions of parked vehicles. However, no industry information was found. Per the previous analysis of a 50 gallon vehicle, it was found that the frequency of fires/explosions in a single moving vehicle crash is 2.04e-1 0/mile.

CALCULATION FILE PRAOI-01 Rev.3 SSheet II Since none of the parked cars are moving or allowed to move within 238 feet of the transporter, use of this frequency is very conservative.

Regulatory Guide 1.91 provides the following equation for use in the evaluation of explosions:

r = nfs, where n= explosion rate for the substance and transportation mode in question in explosions per mile f = frequency of shipment for the substance in question in shipments per year, and s = exposure distance in miles The frequency (s) of shipment in the assumption section is 8 trips will be made per year. It is assumed that the ISFSI exposure distance, s is conservatively 1000 feet. In addition, it is estimated that a maximum of 200 vehicles will be within that 238 feet setback distance at any moment while the transporter is moving through the exposure distance.

Thus, conservatively r = exposure rate

frequency

exposure distance *number of exposure vehicles = 2.04e-10

8

100015280

200 = 6.18e-8 Note: This is conservative because we are assuming that all vehicle fires lead to an explosion and we are using moving vehicle explosion rates for stationary vehicles.

According to Regulatory Guide 1.91 Revision 1, if conservative calculations are used, an exposure rate, of less than le-6lyear is acceptable.

Transformer Explosion There are 6 active, normally energized transformers located on the Unit 2 side (south side) of containment. Three are single-phase 500kV transformers, two are three phase 25kV transformers and the last is a three-phase 12kV transformer. There are also two spare transformers stored adjacent to the active transformers. The transformers are located at elevation 85'. The road for the transporter is at elevation 115' and runs perpendicular to the potential explosion zone for under 600 feet. There is a sloped, rock-covered embankment located approximately 120 feet from the transformers. That embankment is 30 feet tall and would take the majority of the explosions force heading that direction. On the top of the embankment at elevation 115' is a paved lot that is used as a storage area - this has already been evaluated in Reference 4.

Approximately 60 feet away from the ledge of the embankment is the road that the Hl-TRAC transporter will traverse.

The layout of the transformers is such that the three single-phase 500kV transformers are located closer to the pathway, while the two three phase 25kV transformers are

CALCULATION FILE PRAOI-01 Rev.3 Sheet 12 mostly shielded from the pathway by the 500kV transformers. The 12kV transformer is located further yet, and is also shielded by the other transformers. All of the active transformers have a fire suppression system surrounding them, which will activate in the event of a fire.

Failure rate for transformers was obtained from a standard nuclear industry source (Reference 10) using catastrophic failures, which are composed of open and short circuits, plus not crediting mitigation by protective features; i.e. potential explosive failures. The recommended values from the reference were utilized, which is conservative for the new 500kV transformers installed on Unit 2.

Transformer type Failure Rate Reference 500 kV Single Phase 2.6E-7 failures per hour Page 369, Section 6.3.1.6 Transmission 12kV - 25kV Three Phase 6.6E-7 failures per hour Page 371, Section 6.3.2.1 Transmission I

I Consistent with the assumptions throughout this calculation, the conservative time of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of transport will be used. There are three transformers of each type.

Additionally, there is a geometric factor to consider, for an explosion would have to impact the transporter traveling approximately 240 feet away at an elevated height of 30 feet above the transformer. The geometric factors utilized assumed that there would be no bounce, or reflection off structures or the rock embankment that could impact the transporter and that the blast was evenly distributed over the potential 180 degrees of blast direction. As noted earlier, the transporter will be approximately 240 feet away from the transformers and elevated approximately 30 feet. As a result, the ratio of the transformer target area to the total blast area is quite small, and is judged to result in geometry factors of 5e-3 for the 500 kV transformers and le-3 for the 12-25kV transformers, assuming a 20% throughput of energy past the 500kV transformer acting as a directional shield.

Thus: 500kV: (2.6e-7 failures per hour per transformer) x (3 transformers) x (10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per year) x ( 5e-3 geometric) = 3.9e-8 per year And 12-25kV: (6.6e-7 failures per hour per transformer) x (3 transformers) x (10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per year) x (1 e-3 geometric) = 1.98e-8 The total risk is the sum of the risks from the two calculations for the 6 transformers, which is 5.9e-8 per year.

The risk from a transformer explosion, based on exposure time and distance, is estimated to be less than the le -6 acceptance criteria stated by R G 1.91.

CALCULATION FILE PRAD"I-O Rev.3 Sheet 13 Gas Bottle Explosion There are no storage facilities for explosive gases within 1000 feet of the ISFSI facility.

It is intended that propane or any other explosive gas transportation past the ISFSI facility will be controlled through administrative procedures. These procedures will not only control the amount of gas being transported, but how it is physically restrained during transport to ensure limited potential for failure.

In this evaluation the risk of damage from this hazard is judged to be bounded by the risk of damage due to 2000-gallon truck explosion. The basis for this judgement is that there are no bottle storage facilities close to the ISFSI facility, the potential for an explosion is considered limited to a transportation accident, and the administrative controls on the transport through the ISFSI facility area limit speed, number of vehicles and setback distances.

Explosions or failures of any type involving compressed gas bottles usually are caused from valves being broken or the bottles being pierced by some external object. These types of failures require some motive force and are considered to be limited in the case of the area around the ISFSI to a vehicle crash. In addition, because of the limitations on vehicle motion the possibility of a crash is limited further to single vehicle crashes similar to the large truck evaluation, which was found to be 1.57e-7/year in this calculation. If we conservatively consider that every truck crash causing a bottle failure that will potentially affect the ISFSI, the risk is less than significant per RG 1.91.

When using RG 1.91 approach for evaluating the risk of damage due to explosion from gas bottles transported in a truck, the estimates of parameters used for calculating exposure rate (r) are bounded by the estimates used for the 2000-gallon truck explosion on the basis that:

n (explosion rate) is the same since the primary reason for a bottle explosion is judged to be due to large truck crashes.

f (frequency of gas bottle shipments) is judged to be less than the frequency of 2000-gallon truck shipment of one per day.

s (exposure distance in miles) is judged to be less than that of the 2000 gallon truck hazard due to lower potential energy release from explosion of a few gas bottles than the potential energy release from explosion of 2000 gallon of gasoline.

There is one other potential failure mode for these pressurized bottles and that is a welded seam failure. Aithough this is a possibility, the gas bottles provided on the site are required to meet current industry testing standards. These standards are provided to ensure that a weld failure is not a significant risk.

Therefore, it is judged that, based on RG 1.91 criteria, the risk of damage due to gas bottle explosion is less than 1.0E-6 and is therefore considered insignificant.

CALCULATION FILE PRAOI-01 Rev.3 Sheet 14 Note that the risk of damage is a conservative measure to use as a surrogate for the risk to the public on the basis that the damage to the casks does not constitute failure of the cask barrier integrity or the fuel cladding integrity.

In addition to the transportation of gas bottles past the ISFSI facility, there is one stationary gas bottle facility along the transporter route that could potentially affect the transporter. This facility is located on the east side of the cold machine shop and contains acetylene bottles. This facility is more than twenty five feet from the transporter route and is a few feet below the transport roadbed. The facility only holds a maximum of 10 bottles and is protected on two sides by concrete block walls and on the third side by a building. The Diablo Canyon procedural requirements for the storage of gas bottles ensure that these bottles are restrained in a vertical position ensuring that no potential missiles are aimed at the transporter route. These restraint requirements are provided for seismic considerations. In addition, the exposure time for the transporter being in the area is less than 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per year.

Explosions or failures of any type involving compressed gas bottles usually are caused from valves being broken or the bottles being pierced by some external object. These types of failures require some motive force. The location of this facility provides limited access by vehicles and administrative controls will not allow any vehicle motion within 238 feet of the transporter, therefore, there is no motive force and the possibility of an explosion is not considered credible. The risk provided by this facility is conservatively bounded by the hydrogen facility risk potential.

RESULTS The risks associated with explosions, which could potentially damage the HI-TRAC transfer cask or the HI-STORM storage cask and their associated SSC's, were evaluated. All the hazards evaluated resulted in conservative estimates for exposure rates of less than 106, which is risk insignificant. According to RG 1.91, this risk level is acceptable.

RECOMMENDATIONS Each of the assumptions listed in the assumptions section of this calculation should be implemented through administrative procedures. This is being tracked by AR A0524878.

REFERENCES

1. "Regulatory Guide 1.91, Revision 2, "Evaluation of Explosions Postulated to Occur On Transportation Routes Near Nuclear Power Plants," February 1978.

2. PG&E Calculation M-1046, Revision 0, "Minimum Separation Between Fuel Tanks and Storage Casks."

CALCULATION FILE PRADOl Rev.3 Sheet 16

3. PG&E White Paper, "Bulk Hydrogen Facility Risk Evaluation," Doug Spaulding, dated February 2001.

4. "Evaluation of the HI-TRAC Transportation Route", Dave Hampshire, dated March 6, 2001.

5. EPRI, 'Fire Events Database for U.S. Nuclear Power Plants."

6. US Department of Transportation, National Highway Traffic Safety Administration (NHTSA), "Traffic Safety Facts 2001," "2001 Motor Vehicle Crash Data from FARS and GES".

http:/twww-nrd.nhtsa.dot.gov/pdf/nrd-30/NCSA/TSFAnnITSF200I.pdf

7. US Department of Transportation, Federal Motor Carrier Safety Administration (FMCSA) 'Large Truck Crash Facts 2001', January 2003 http://ai.volpe.dot.gov/CarrierResearchResults/PDFs/LargeTruckCrashFacts200l.pd f

8. PG&E Calculation M-1047, Revision 0, "Minimum Separation Between Acetylene Tanks and Transfer Casks."

9. "Telecon with Lou Ricks on Fuel Truck Frequency," Dave Hampshire and Lou Ricks, 2113/01.

10. IEEE Std. 500-1984, "IEEE Guide to the Collection and Presentation of Electrical, Electronic, Sensing Component, and Mechanical Equipment Reliability Data for Nuclear-Power Generating Stations."

ii.Army TM 5-1300 - Structures to Resist the Effects of Accidental Explosions, November 1990, page 2-57.

PG&E Letter DIL-03-005 Sheet 1 of 1 LIST OF HOLTEC NONPROPRIETARY CALCULATION PACKAGES

1.

Dose Evaluation for the ISFSI at Diablo Canyon Power Station Holtec Report No. HI-2002563, Revision 5.

MEEM H O LT E C INTERNATIONAL Hoftec Center, 555 kidn Dive West Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 - 0909 I

DOSE EVALUA TION FOR THE ISFSI AT DIABLO CANYON POWER STATION FOR Pacffic Gas and Electdc Company Holtec Report No: HI-2002563 Holtec Project No: 1073 Report Class: SAFETY RELATED U

HOLTEC INTERNATIONAL DOCUMENT ISSUANCE AND REVISION STATUS, DOCUMENT NAME: Dose Evaluation for the ISFSI at Diablo Canyon Power Station DOCUMENT NO.:

HI-2002563 CATEGORY: O GENERIC PROJECT NO.:

1073 l

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Date Author's Rev.

Date Author's No.2 Approved Initials VIR #

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9/28/01 ERD 223887 3/22/01 ERD 1 93245 4

10/24/01 ERD 163444 2

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DOCUMENT FORMATTING The formatting of the contents of this document is in accordance with the instructions of HQP 3.2 or 3.4 except as noted below:

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-This dcumenthasbeensubjected to review, verification and appvalp e set forth i the Holec Quality Auane iProcedures7Mniiat Psword controlled signatures of Holtec persolwho participated infthe preparation, review, and A -validatio of this documient are save in -the n-drive of thecompanys networ. The Vidation identifieRct cord (VR) umiber is a randoi number that is generated Ibythe coiputerafter the specific revision ofthis document hasurgone he red r

ad approval process, and the te Holtec personnel have icorded?

their passw onolled electronic curn tohe document.

2.

A revision to this d will be ordered "byte Pject Manager and carried out if any of itontents is mat ly afected dung evolution of tis prect he determination as to te need tor revision will be made by the Project Manager with inputfrom others, as deemedl necessaiy y him.

-3. fI;vis3io to this d ocument may be made by addingsupplements to t ie document and replacing the "Table of Conent",e this pae and th;4Ue~

"RevsT:

ionLi tog.

U-5:: fi-Ei:x ::-

Summary of Revisions Revision 0 Original Issue Revision I

1. All changes noted with revision bars.

2. Added Section 1.2 in accordance with Holtec administrative memos. The introduction from Revision 0 has been moved into Section 1.1.

3. Made changes per PG&E comments.

4. A footnote was added to the reference section in accordance with Holtec administrative memos.

5. Four was changed to three in Figure 2 caption.

6. The table of contents was changed appropriately.

Revision 2

1. All changes noted with revision bars. Revision bars from rev I have been removed.

2. Fixed an incorrect reference to a section.

3. Removed a paragraph discussing analysis of the dose at the nuisance fence behind and on the sides of the array. This analysis is not contained in the appendices that were referenced.

Revision 3

1. The HI-STORM 100S drawings were updated to include design changes. The significant change from the perspective of this report was the removal of the inner shield shell and the increase in the concrete density of the body to compensate. The lid design was also changed to include the shear bar as part of the lid design. This affected the calculation of the dose from a HI-STORM without a lid. All analyses of the HI-STORM 100S in this report were updated accordingly. In some cases the previous models of the overpacks were used but justification is appropriately provided.

2. Revision bars mark all revisions to text. The following pages of results were completely replaced: A2-A9, D4-D7, E2-E14, F3-FI 1, G3-G14, H5-H14, I2, K2-K 1I, L5-L7, M19-M21.

Revision 4

1. Appendix N was added to discuss the determination of the allowable burnup and cooling times for the BPRAs and TPDs.

2. The approved computer code list in Attachment A was updated to show the report number in the footer and the version of the code used.

Revision 5

1. The duration for step 47 during loading operations in App. K and step 19 during unloading operations in App. K was increased from 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (90 minutes) to 3.0 hours0 days 0 hours 0 weeks 0 months (180 minutes). This resulted in a change to the total dose rates in App. K.

Report: HI-2002563 P a-ge: I

Table of Contents

1. Introduction......................

2 1.1 Statement of Purpose......................

2 1.2 About This Document....................

2

2.

General Methodology......................

4

3.

Acceptance Criteria.......................

4

4.

Assumptions.......................

5

5.

Input Data.......................

7 5.1 ISFSI Geometry......................

8

6.

Computer Codes.......................

8

7.

Analysis and Results......................

9 7.1 Source Terms...........................................

9 7.2 MCNP Modeling of the MPC, HI-STORM Overpack, and 125-ton HI-TRAC Transfer Cask 9 7.3 Method of Tallying 10 7.4 Choice of Design Basis Burnup and Cooling Time

.10 7.5 Dose Calculations for the HI-STORM 100S on the ISFSI I

7.5.1 MCNP Surface Source Calculations.................................................

I 1 7.5.2 Dose Rates Adjacent to the Overpack.................................................

12 7.5.3 Cask Configurations for Dose Versus Distance Calculations............................. 12 7.5.4 Final Dose Rate Calculations............................................

.... 14 7.6 Dose Rates from the HI-STORM 100S Without a Lid 14 7.7 Dose Rates from the 125-ton HI-TRAC Transfer Cask.

15 7.7.1 Normal Conditions............................................

15 7.7.2 Accident Conditions............................................

16 7.8 Doses During Loading and Unloading Operations.

........................................... 16

8.

Computer Files............................................

17

9.

Summary..............................................

23

10.

References 28 10.1 Drawings...........

29 Appendix A:

Bounding Burnup and Cooling Time for HI-STORM.................................. A-1 Appendix B:

Bounding Bumup and Cooling Time for the 125-ton HI-TRAC.................. B-l Appendix C:

Source Terms............................................

C-l Appendix D:

Near Dose Rates for HI-STORM Overpack............................................

D-I Appendix E:

Single Cask Dose Rates Versus Distance............................................

E-I Appendix F:

Results of Various Overpack Configuration Analyses.................................. F-l Appendix G:

Final ISFSI Dose Rate Calculations....................

........................ G-l Appendix H:

Annual Personnel Doses from ISFSI Operations......................................... H-l Appendix I:

Dose Rates from a HI-STORM 100S Without a Lid.1-1 Appendix J:

Dose Rates from the 125-ton HI-TRAC Transfer Cask

.J-1 Appendix K:

Occupational Exposures During Loading and Unloading Operations.

K-l Appendix L:

Effect of Air Composition on Dose Calculations.

L-I Appendix M:

Comparison of B&W 15x15 and W 17x17 Assemblies.

M-1 Appendix N:

Allowable Burnup and Cooling Times for BPRAs and TPDs.N-I Attachment A: List of Approved Computer Codes (4 pages)

Report: HI-2002563 Page: I

1. Introduction 1.1 Statement of Purpose This report documents the radiation shielding analysis that was performed for the Independent Spent Fuel Storage Installation (ISFSI) at Diablo Canyon Power Plant (DCPP). This shielding analysis includes calculation of the dose rates from the HI-STORM 100S overpack on the ISFSI and in the Cask Transfer Facility (CTF), the 125-ton HI-TRAC transfer cask, and the entire ISFSI filled with HI-STORM IOOS overpacks. Occupational exposures during loading and unloading operations of a HI-STORM IOOS overpack and maintenance and surveillance operations around the ISFSI are estimated in this report. This report also addresses the radiation consequences of a lead slump in the 125-ton HI-TRAC resulting from a drop accident. This report only considers the direct radiation source emanating off the sides and top of the overpack and the sides, top, and bottom of the transfer cask. Since the MPC is seal welded there is no effluent release of radiation. However, the current NRC regulations require the calculation of the off-site dose rate associated with normal, off-normal, and accident effluent release of radiation. These calculations are performed in reference [8] and are summarized in this report in Section 9.

In its fully implemented final configuration, this facility will consist of 140 HI-STORM 100S casks loaded with the MPC-24 or MPC-32. Up to seven ISFSI pads may be constructed with each pad able to store a 4x5 array of casks [12]. At the completion of the ISFSI, the casks would be in a 5x28 configuration. The center to center pitch between HI-STORM casks is 17 feet [12].

The pads will be constructed in such a manner as to maintain the 17 foot pitch between casks on adjacent pads. Dose rates from the cask array are calculated as a function of distance. Distances include relevant onsite and offsite dose locations. For offsite dose rates, the results presented in this report must be added to the dose rates from other Uranium Fuel Cycle operations to determine whether the regulatory requirements for normal (I OCFR72.104) conditions are met.

1.2 About This Document This work product has been labeled a safety-significant document in Holtec's QA System. In order to gain acceptance as a safety-significant document in the company's quality assurance system, this document is required to undergo a prescribed review and concurrence process that requires the preparer and reviewer(s) of the document to answer a long list of questions crafted to ensure that the document has been purged of all errors of any material significance. A record of the review and verification activities is maintained in electronic form within the company's network to enable future retrieval and recapitulation of the programmatic acceptance process leading to the acceptance and release of this document under the company's QA system. Among the numerous requirements that a document of this genre must fulfill to muster approval within the company's QA program are:

Report: 1I-2002563 Page: 2

The preparer(s) and reviewer(s) are technically qualified to perform their activities per the applicable Holtec Quality Procedure (HQP).

The input information utilized in the work effort must be drawn from referencable sources. Any assumed input data is so identified.

All significant assumptions, as applicable, are stated.

The analysis methodology, if utilized, is consistent with the physics of the problem.

Any computer code and its specific versions that may be used in this work has been formally admitted for use within the company's QA system.

The format and content of the document is in accordance with the applicable Holtec quality procedure.

The material content of this document is understandable to a reader with the requisite academic training and experience in the underlying technical disciplines.

Once a safety significant document produced under the company's QA System completes its review and certification cycle, it should be free of any materially significant error and should not require a revision unless its scope of treatment needs to be altered. Except for regulatory interface documents (i.e., those that are submitted to the NRC in support of a license amendment and request), revisions to Holtec safety-significant documents to amend grammar, to improve diction, or to add trivial calculations are made only if such editorial changes are warranted to prevent erroneous conclusions from being inferred by the reader. In other words, the focus in the preparation of this document is to ensure accuracy of the technical content rather than the cosmetics of presentation.

In accordance with the foregoing, this Calculation Package has been prepared pursuant to the provisions of Holtec Quality Procedures HQP 3.0 and 3.2, which require that all analyses utilized in support of the design of a safety-related or important-to-safety structure, component, or system be fully documented such that the analyses can be reproduced at any time in the future by a specialist trained in the discipline(s) involved. HQP 3.2 sets down a rigid format structure for the content and organization of Calculation Packages that are intended to create a document that is complete in terms of the exhaustiveness of content. The Calculation Packages, however, lack the narrational smoothness of a Technical Report, and are not intended to serve as a Technical Report.

Because of its function as a repository of all analyses performed on the subject of its scope, this document will require a revision only if an error is discovered in the computations or the equipment design is modified. Additional analyses in the future may be added as numbered supplements to this Package. Each time a supplement is added or the existing material is revised, the revision status of this Package is advanced to the next number and the Table of Contents is amended. Calculation Packages are Holtec proprietary documents. They are shared with a client only under strict controls on their use and dissemination.

This Calculation Package will be saved as a Permanent Record under the company's QA System.

Report: HI-2002563 Page: 3

2. General Methodology The analysis of the 140 cask ISFSI can be separated into two distinct parts. The first is the generation of the radiation source terms to represent the spent nuclear fuel at the appropriate burnup and cooling time. The second part is the radiation transport simulation to calculate the dose rates near and far from a cask and an array of casks.

The radiation source terms were calculated using the SAS2H and ORIGEN-S modules from the SCALE 4.3 [1,2] code system from Oak Ridge National Laboratory. This is a widely accepted means of generating radiation source terms from spent nuclear fuel.

The radiation transport simulation was performed with MCNP 4A [3] from Los Alamos National Laboratory. This is a state of the art Monte Carlo code that offers coupled neutron-gamma transport using continuous energy cross sections in a full three-dimensional geometry.

The specifics of the radiation source term calculations and radiation transport simulation are discussed below.

3. Acceptance Criteria The acceptance criteria for offsite dose rates are dictated by IOCFR72.104 and IOCFR72.106 and are summarized below.

Normal condition requirements from I OCFR72.104.

1. During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area, must not exceed 25 mrem to the whole body, 75 mrem to the thyroid and 25 mrem to any other critical organ.

2. Operational restrictions must be established to meet as low as reasonably achievable (ALARA) objectives for radioactive materials in effluents and direct radiation.

Accident condition requirements from IOCFR72.106 Any individual located on or beyond the nearest boundary of the controlled area may not receive from any design basis accident the more limiting of a total effective dose equivalent of 5 Rem, or the sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue (other than the lens of the eye) of 50 Rem. The lens dose equivalent shall not exceed 15 Rem and the shallow dose equivalent to skin or to any extremity shall not exceed 50 rem. The minimum distance from the spent fuel or high level radioactive waste handling and storage facilities to the nearest boundary of the controlled area shall be at least 100 meters.

References [4] and [10] and the Diablo Canyon SAR demonstrate that there are no accidents which would significantly affect shielding effectiveness of the HI-STORM. References [4] and

[10] further demonstrate that the requirements of 10CFR72.106 are easily met by the HI-STORM 100 system. Therefore, explicit analysis of an accident scenario and demonstration of Report: HI-2002563 Page: 4

compliance with IOCFR72.106 was not performed. References [4] and [10] offer further discussion on this topic.

This report demonstrates that the ISFSI meets the above stated acceptance criteria.

For onsite dose rates, the following dose rate limits are used which are consistent with the requirements specified in 10CFR20:

5 rem/year for personnel with dose rate monitors (IOCFR20.1201) 0.5 rem/year for personnel without dose rate monitors (IOCFR20.1201 and 1502) 2 mrem in any one hour and 100 mrem/year for individual members of the public (IOCFR20.1301)

This report demonstrates that the ISFSI is capable of meeting the above stated acceptance criteria. Compliance with 10CFR20 will be demonstrated by personnel dose monitoring in accordance with the DCPP Health Physics Program.

4. Assumptions The following assumptions are used in this analysis:

1. It is assumed that the occupancy factor for the closest resident beyond the site boundary is 8760 hr, which is full occupancy for the entire year.

2. In compliance with the applicable portions of [9], it is assumed that the occupancy factor for the nearest site boundary, exclusion area boundary and unrestricted area boundary is 2080 hr.

This assumption is based on the approach to identify individuals within the geographic location of the ISFSI, and estimate their maximum radiological exposure. The area directly outside the unrestricted area boundary is uninhabited. As a bounding approach, it is estimated that the individual with the maximum exposure would be an individual working outside the boundary for the entire year. The occupancy is then calculated based on a working week of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months and 52 weeks/year.

3. It is assumed that the occupancy factor for the occupational dose rate is 2080 hr, which is based on a working week of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months and 52 weeks/year.

4. It is assumed that the occupancy time for individual members of the public is administratively limited to ensure compliance with the dose rate limitation of 100 nirem/year specified by IOCFR20. Therefore, only compliance with the 2 mrem in any one hour limitation is evaluated for members of the public in this report. Compliance is demonstrated by showing that the dose rate is less than 2 mrem/hr.

5. All PWR fuel assemblies are assumed to be B&W 15x15. This is the design basis fuel assembly from reference [4]. This assembly type has a higher source term than all 17x1 7 assembly types used at DCPP, as demonstrated in [4]. Additional discussion on this assumption is provided in Appendix M.

Report: HI-2002563 Page: 5

6. The analyses account for the increase in cooling time of the fuel in each cask over the years of operation of the facility. Cooling times are calculated based on the assumption that 8 casks are placed into the ISFSI per year. It is assumed in the analysis that each pad is completely filled before the next pad is put into use. It is also assumed that the casks with the youngest fuel are placed closest to the front of the ISFSI and therefore the highest dose rate contribution is calculated. At a rate of eight casks per year, it will take 17.5 years to fill the ISFSI to capacity for a total minimum cooling time after core discharge of 22.5 years for the first casks deployed. However, the oldest fuel in the casks in the ISFSI was conservatively assumed to be 20 years old. No credit was taken for additional cooling from 20-22.5 years.

Note that this approach also conservatively assumes that all fuel is loaded in the HI-STORM 100 System casks at 5 years' cooling time - the shortest cooling time allowed by the Technical Specifications. The distribution of cooling times within the ISFSI is shown in Figure 1.

7. It is assumed that the facility is filled to its maximum capacity in each phase, and that all HI-STORM casks are loaded with the MPC-32 with fuel of 32,500 MWD/MTU burnup, 2.9 wt0/o initial enrichment, and 5 year cooling time at the time of loading. This burnup and cooling time was chosen based on a comparison of the dose rates for different burnup and cooling times. It is demonstrated in Appendix A that this burnup and cooling time in the MPC-32 for the HI-STORM IOOS bounds other burnup and cooling times for the MPC-32 and MPC-24 based on the allowable burnup and cooling times.

8. The burnup and cooling time assumed for the 125 ton HI-TRAC analysis is 55,000 MWD/MlrJ and 12-year cooling. This is based on a comparison of the dose rate for different burnup and cooling times on the side of the 125-ton HI-TRAC loaded with the MP("-24 end MPC-32. Appendix B presents the results of the comparisons and demonstrates that the burnup and cooling time chosen is bounding. The 55,000 MWD/MTU differs slightly from the value of 57,500 MWD/MTU used in reference [10]. The 55,000 MWD/MTU is basically the same value as the allowable for the MPC-24E at 12 year cooling for uniform loading.

9. The enrichment assumed for the 32,500 MWD/MTU was 2.9 wt.% 2 35U which is consistent with reference [4]. The enrichment used for 55,000 MWDIMTU was 4.0 wt.% 235U which is less than, and more conservative, than the value used in reference [4]. These values were chosen based on a review of the current DCPP fuel inventory provided in [5].

10. The cobalt-59 impurity level was assumed to be 1.0 gm/kg for the hardware above and below the active fuel region and for the grid spacers. This is a conservative value for the cobalt-59 impurity level as modem fuel is manufactured with cobalt impurity levels typically 0.3-0.5 gm/kg or less. Consistent with reference [10], the cobalt-59 impurity level in the steel in the BPRAs was assumed to be 0.8 gn/kg and in the inconel it was assumed to be 4.7 gm/kg.

11. It is conservatively assumed that all in-core grid spacers are non-zircaloy with a cobalt-59 impurity level of 1.0 gm/kg. Some of the fuel assemblies in the core periphery locations at Diablo Canyon use fuel straps which provide additional support for the fuel rods. These fuel straps are inconel and it is assumed that the mass of these straps is 30 gm/each and it is assumed that there are 10 straps per assembly. This results in an additional 300 gm of inconel in the active fuel region in addition to the 4.9 kg assumed for the grid spacers. This is a slightly more than 6% increase in the mass. Since the cobalt-59 impurity level being used for the grid spacers is 1.0 gm/kg, which is more than 20% conservative, the fuel straps are bounded by the grid spacers and therefore are not explicitly analyzed.

Report: HI-2002563 Page: 6

12. It is assumed that each fuel assembly in the ISFSI contains a BPRA with stainless steel clad and a cooling time of 13 years at the time of loading. The activation of the BPRA is based on an assembly burnup of 40,000 MWD/MTU and 24 rods per BPRA. These are conservative assumptions based on the current inventory. The current fuel inventory [11] lists only a total of 324 BPRAs, compared to a maximum of up to 4480 assemblies in the ISFSI. On average, the BPRAs have less than 14 rods each. Out of the 324 BPRAs, only 284 have stainless steel clad. The remaining 40 BPRAs have Zr-4 clad which results in lower dose rates. BPRAs have only been used in the first 3 operational cycles of the plants and it is anticipated that there is no further use of BPRAs in the future. At the time the first casks are loaded, the cooling time of the BPRAs will therefore exceed 13 years.

13. The current inventory shows 194 thimble plugs (TPs), and no future use of TPs is anticipated

[5]. Typically, TPs have a lower activation than BPRAs, as TPs do not protrude into the active region of the assemblies. Each assembly can only accommodate a BPRA or a TP, but not both. Therefore, the presence of TPs is bounded by the conservative assumption regarding BPRAs (see previous assumption) and no further evaluations are required for TPs.

14. The air density in the dose versus distance calculations was assumed to be 1.17E-03 gm/cc and the air was assumed to be composed of only Nitrogen and Oxygen. Appendix L provides additional discussion on this assumption.

15. Occupancy times and dose rates for surveillance, maintenance and repair of the ISFSI are based on the following assumptions

Daily walk-dowrn of the ISFSI to inspect all casks and vent openings. This requires a person to walk the full length of the ISFSI (approximately 493 R) outside the pads and between the lines of casks on each pad. The individual will walk a maximum of 6 ISFSI lengths. This walk-down would occur once a day every day. For I person, the time for this walk is estimated as less than 20 min. This is based on an assumed walking speed of 2 miles/hour, the length of the ISFSI and the number of lines of casks. This results in a total occupancy of 122 hour0.00141 days 0.0339 hours 2.017196e-4 weeks 4.6421e-5 months s/year.

Repair operations: I repair operation per month, I hour each repair, 2 persons. This results in a total occupancy of 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s/year/person.

These occupancy times are estimated to demonstrate that the ISFSI can be operated within the requirements specified by IOCFR20 (see Section 3). Compliance with IOCFR20 will be demonstrated by personnel dose monitoring in accordance with the DCPP Health Physics Program. Therefore, these occupancy times do not represent any limitations or requirements.

Other assumptions are stated in the text as necessary. Since the MCNP models used in this report are based on previous HI-STORM analyses, additional assumptions and discussion can be found in references [4], [6], and [10].

5. Input Data The input data for generating the radiation source terms is provided in references [4] and [10]

and can be found in Appendix C. The input data for the MCNP models of the overpack and the MPC, including the density and composition of all materials used in the models is also available Report: H1-2002563 Page: 7

in these references. In addition, input data for the HI-STORM 100S and a discussion of the modeling can be found in reference [15].

5.1 ISFSI Geometry The ISFSI configuration is specified in [12] and illustrated in Figure 1. The number associated with each cask location in Figure 1 is the minimum cooling time of the fuel in each cask (see Section 4, Assumption 6). There are a maximum of 7 pads being constructed, each of which will hold a 4x5 array of HI-STORM 100S casks. In the final configuration the ISFSI will be a 5x28 array of casks. The center to center pitch for the casks is 17 feet and the pads will be positioned such that the pitch between casks on adjacent pads is maintained at 17 feet. An overpack is 11 feet 1/2 inch in diameter. There will be a security fence and two nuisance fences placed around the ISFSI. The outer nuisance fence will be located to maintain the dose rate below 2 mrem/hr.

The ISFSI site is cut into a hill slightly. Only one of the long sides of the ISFSI is on the same level with the surrounding area. This is the front side facing the controlled area boundary. The sides and the back of the ISFSI face the excavated slopes of the hill.

Dose rates are calculated as a function of distance from the cask array. The distances include the following dose locations:

Offsite dose locations:

Nearest resident 1.5 miles / 24i4 m

[7]

Unrestricted Area Boundary 1400 ft / 426.7 m

[7]

Onsite dose locations:

Aux. Building Wall 798 ft 243.2 m

[7]

Make-up water Facility 223 ft / 68.0 m

[7]

Nuisance Fence, Front 100 ft /31.4 m assumed It is conservatively assumed that there are no obstructions (i.e. hills, buildings, earth berms) between the cask array and any of the dose locations.

6. Computer Codes The computer codes used for these calculations were the following.

1. SAS2H module from SCALE 4.3 - reference [I]

2. ORIGEN-S module from SCALE 4.3 - reference [2]

3. MCNP 4A - references [3] and [6]

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7. Analysis and Results This section of the report describes the calculations that were performed to determine the dose rates for various distances and locations. The basic development of the MCNP models, including source terms and tally normalization, had already been accomplished during the HI-STORM 100 project. This information is appropriately referenced as needed.

7.1 Source Terms There are three distinct primary radiation source terms that must be accounted for in the analysis of the HI-STORM 100 system. These are:

1. Neutron source from the decay of spent nuclear fuel.

2. Photon source from the decay of spent nuclear fuel.

3. Photons from the decay of Cobalt-60 in the end-fittings of the fuel assemblies. This source represents the activation of the steel components in the fuel assemblies.

These radiation source terms were calculated using the SAS2H and ORIGEN-S modules of SCALE 4.3. A full description of the methodology can be found in reference [4]. The source terms for the 32,500 and 55,000 MWD/MTU design basis fuel can be found in Appendix C.

A secondary source of radiation is from the following.

1. Secondary neutrons from fast fission in the fuel.

2. Secondary photons from prompt neutron interactions in the MPC and overpack.

These secondary sources are automatically accounted for during the MCNP calculation by running a coupled neutron-gamma calculation.

Curies of Cobalt-60 in the BPRAs are also listed in Appendix C. The allowable burnup and cooling times for BPRAs and TPDs are calculated in Appendix N.

7.2 MCNP Modeling of the MPC, HI-STORM Overpack, and 125-ton HI-TRAC Transfer Cask All MCNP calculations are performed for a HI-STORM IOOS loaded with the MPC-24. Results for the HI-STORM loaded with the MPC-32 are obtained from the MPC-24 results through multiplication with the ratio of assemblies in each MPC. This is slightly conservative, as it neglects the increased self shielding within the MPC-32.

The MPC and the overpack were modeled in full three-dimensional detail using MCNP. The description of the modeling process can be found in references [4], [6], and [10].

References [4], [6], and [10] identify a couple of modeling discrepancies between the MCNP model of the overpack and the design basis drawings. These discrepancies are:

Report: HI-2002563 Page: 9

1. The steel channels in the cavity between the MPC and overpack were not modeled. This is conservative since it removes steel that would provide a small amount of additional shielding.

2. The bolt anchor blocks were not explicitly modeled. Concrete was used instead. These are small localized items and will not impact dose rates.

3. The exit vents in the overpack were modeled as being inline with the inlet vents. In practice, they are rotated 45 degrees and positioned above the short radial plates. Therefore, this modeling change has the exit vents positioned above the full length radial plates. This modeling change has minimal impact on the dose rates at the exit vents.

4. The short radial plates in the overpack were modeled in MCNP event though they are optional on the drawings.

The MCNP models of the 125-ton HI-TRAC described in References [4], [6], and[10] have been updated to incorporate the latest Engineering Change Orders (ECOs). The only difference between the models used in this analysis and the design of the 125-ton HI-TRAC is the following.

The pocket trunnions on the 125-ton HI-TRAC transfer cask have been removed.

Section 10.1 provides a listing of the drawings that were used to generate the MCNP models used in this report. In certain calculations, an earlier version of the HI-STORM 100S design was used for some of the MCNP models. This is described in more detail and justified in Section 10.1.

7.3 Method of Tallying In MCNP, the calculation of a user requested quantity (e.g. dose rate) is referred to as tallying.

The tally results calculated in MCNP are normalized per starting particle. Therefore, the MCNP results must be normalized to the actual source strength for the system being analyzed. This normalization is done after the computer runs are completed and is done in EXCEL. The method of the tally normalization is described in reference [6].

7.4 Choice of Design Basis Burnup and Cooling Time The burnup and cooling times used in this analysis were chosen based on a review of the dose rates from the HI-STORM IOOS and the 125-ton HI-TRAC with different bumup and cooling times from the allowable bumup and cooling times. The results indicate that the 32,500 MWD/MTU and 5 year cooling for the MPC-32 bound the allowable burnup and cooling times for both the MPC-24 and MPC-32 for the HI-STORM lOOS. The results also indicate that 55,000 MWD/MTU and 12 year cooling for the MPC-24 bound the allowable bumup and cooling times for both the MPC-24 and MPC-32 for the 125-ton HI-TRAC analysis. The results of these comparisons are presented in Appendices A and B for the HI-STORM and HI-TRAC respectively.

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The burnup and cooling times analyzed were for uniform loading. Results presented in reference

[IO] indicate that the dose rates for the HI-TRAC transfer cask from burnup and cooling times taken from the allowable contents for uniform loading bound or are equivalent to the dose rates from burnup and cooling times for regionalized loading. The same conclusions would be true for the HI-STORM cask. Therefore, explicit analysis of the regionalized loading pattern was not performed.

7.5 Dose Calculations for the HI-STORM 100S on the ISFSI 7.5.1 MCNP Surface Source Calculations MCNP offers the capability to generate a surface source file in one calculation which can then be used in other calculations. The surface source file contains information for particle tracks that cross user designated surfaces. Particles are either neutrons or photons. This method of using a surface source file has a major benefit because the user can generate a surface source file for particles leaving the overpack and then use this file in different runs with different overpack arrangements. The advantage is that, for each overpack arrangement, the user does not have to waste valuable computer time tracking particles out of the overpack since this information is already contained on the surface source file.

Reference [6] describes in detail the use of the surface source file for analysis of the HI-STORM 100 system.

Since there are three separate radiation sources (neutron, gamma, and Cobalt-60) a minimum of three different MCNP calculations had to be performed. A brief description of these MCNP calculations is provided here. The computer input file names are listed in Section 8.

1. A coupled neutron-gamma calculation using the neutron source was performed to generate a surface source file for particles leaving the side and top of the overpack.

2. A gamma only calculation was performed using the decay gamma source. This run generated a surface source file for particles leaving the side and top of the overpack. The energy range of starting particles was 0.7 MeV to 1.5 MeV.

3. A gamma only calculation was performed using the decay gamma source. This run generated a surface source file for particles leaving the side and top of the overpack. The energy range of starting particles was 1.5 MeV to 3.0 MeV.

4. A gamma only calculation was performed using the decay gamma source. This run generated a surface source file for particles leaving the side and top of the overpack. The energy range of starting particles was 0.3 MeV to 0.7 MeV.

5. A gamma only calculation using the Cobalt-60 source was performed. This calculation generated a surface source file for particles leaving the side and top of the overpack.

In some cases, the MCNP runs that generated the surface source file for the side and top of the HI-STORM overpack were different. The specifics of each MCNP calculation are listed in Section 8. The surface source technique was only used for the HI-STORM analysis.

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7.5.2 Dose Rates Adjacent to the Overpack MCNP calculations were performed to determine the dose rate at the surface of the overpack and at a distance of 1 meter from the overpack. The computer input files used for the calculations are listed in Section 8.

Appendix D contains the EXCEL output showing the doses at the surface and 1 meter from the HI-STORM 100S overpack. Theses dose rates are expressed in mrem/hr. A summary of the important dose locations is provided in Section 9.

7.5.3 Cask Configurations for Dose Versus Distance Calculations Figure 1 shows a diagram of the ISFSI. The estimation of the dose from a facility of this size and orientation is a complicated calculation. It is almost impossible and certainly impractical to try to model the entire facility in MCNP or any other computer code. Therefore, numerous smaller calculations had to be performed. The results from these smaller calculations were combined in a conservative fashion to estimate the dose from the entire facility.

The radiation source from this facility can be separated into two components. The first will be referred to as the top-dose. This is the dose from radiation that leaves the tops of the overpacks.

The second component will be referred to as side-dose. This is the dose from radiation that leaves the sides of the overpacks.

The geometry of the facility will impact each of these dose components in a different fashion.

The total top-dose will be a summation of the top-doses from all 140 casks where the actual distance from the dose location to the individual cask is accounted for.

The total side-dose will be a summation of the side-doses from all 140 casks where the distances within the facility and the self-shielding of one row of casks to another row are accounted for.

Since the side-dose is from particles leaving the side of the overpack, this dose contribution will be greatly reduced if the cask is situated behind another cask. The front cask blocks radiation from reaching the site-boundary. However, it is incorrect to say that the front cask completely blocks all radiation from the back cask. The fraction of radiation blocked was therefore calculated with MCNP and used in the determination of the total side-dose.

Dose locations along the long side of the cask array are facing 28 casks directly, i.e. without being shielded by other casks. Dose locations along the short side of the array only face 5 casks directly. Dose rates at dose points along the long side of the array will therefore always be higher than dose rates at dose points along the short side of the array. As a bounding approach, dose rates are generally calculated for the long side of the array, regardless of the actual orientation of the dose location relative to the cask array. In the particular case of the ISFSI shown in Figure 1, the peak dose location is perpendicular to the long side of the array but is not in the center of the array. Rather the peak dose location is approximately in the center of Pad 6 in the North direction.

Figure 2 shows three different overpack configurations that were analyzed to determine the self-shielding effects. Since particles leaving the top of the overpack are not self-shielded, only particles leaving the side of the overpack were used in the sources for these configurations.

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Therefore, the results from these configurations will apply only to the side-dose. These configurations are described in detail below.

7.5.3.1 Configuration I Configuration I is a single cask surrounded by 1050 meters of air in the radial direction and 700 meters of air in altitude. The cask is sitting on an infinite slab of dirt. This configuration was used to calculate the average dose rate versus distance for a single cask. The dose rate from particles leaving the side of the cask was calculated separately from the dose rate from particles leaving the top of the cask. The total dose rate was also calculated.

These dose rates were combined in a manner, which is discussed later, to determine the dose rate from the entire facility at the site boundary and security fence.

The MCNP input files used to analyze this configuration are listed in Section 8. All three sources (neutron, photon, and cobalt) were analyzed in this configuration. The results for this configuration are presented in Appendix E.

7.5.3.2 Configuration 2

[

7.5.3.3

[

I Configuration 3 Report: HI-2002563 Page: 13

I 7.5.4 Final Dose Rate Calculations Appendix G presents the results of the calculations to determine the dose rate as a function of distance from the ISFSI. [

] The results are summarized in Section 9.

The dose rates at a distance of 1.5 miles were estimated by extrapolation using appropriate curves. The curves were calculated by fitting the data from the total dose rates versus distance for the entire cask array for distances between 300 and 600 m. The curves reproduce the calculated dose rates between 300 m and 600 m within less than 2.5 %.

Based on the assumed occupancy times and ISFSI dose rates, personnel dose values for the operation and maintenance of the ISFSI are calculated in Appendix H. The accumulated dose for the construction of the last pad is also calculated in Appendix H. Since these operations will occur inside the ISFSI array, a conservative calculation was performed to estimate the dose rate.

[

] The results are presented in Appendix H and summarized in Section 9. The MCNP calculations are listed in Section 8.

7.6 Dose Rates from the HI-STORM 100S Without a Lid The MPC transfer operations from the 125-ton HI-TRAC transfer cask to the HI-STORM IOOS overpack will occur outside the Part 50 structure at the Cask Transfer Facility (CTF). As a result of this, the off-site dose rate from the HI-STORM overpack in the CTF must be calculated. The overpack configuration in the CTF and on the ISFSI are essentially identical with the following exception. After the MPC has been lowered into the HI-STORM overpack and the HI-TRAC transfer cask has been removed, the HI-STORM overpack will be temporarily without a lid. This lidless configuration will permit a significant amount of radiation streaming out the top of the overpack because of the annular gap between the MPC and the overpack. Therefore, the dose rate as a function of distance from a lidless overpack in the CTF was calculated.

Report: HI-2002563 Page: 14

The dose rate was calculated only for radiation leaving the top surface of the lidless overpack.

The dose rate from radiation leaving the side of the overpack is already accounted for in the ISFSI dose rate calculations. [

] The surrounding air and ground region were as described in Section 7.5.3.1. The MCNP surface source technique, as described in Section 7.5.1, was used for these calculations.

The specific input files are listed in Section 8.

The results of these calculations are presented in Appendix I.

7.7 Dose Rates from the 125-ton HI-TRAC Transfer Cask The 125-ton HI-TRAC transfer cask will be used at DCPP for loading the MPC and transferring the MPC to the HI-STORM overpack. Therefore, the dose rates from the 125-ton HI-TRAC were calculated for use in estimating the occupational exposure during loading operations.

7.7.1 Normal Conditions The physical configuration of the HI-TRAC used for calculating the dose rate from radiation leaving the side of the overpack was as described in Section 7.2. [

I The normal condition for the 125-ton HI-TRAC may or may not have water in the water jacket and may or may not have water in the MPC. These conditions vary depending on the process being performed in the loading evolution. For example, water is not present in the water jacket when the HI-TRAC is removed from the spent fuel pool; however, water is in the MPC. For all models, the outer water jacket was assumed to be filled with water and the MPC was assumed to be dry. This is the configuration that exists after the MPC has been seal welded and undergone the moisture removal process.

The configuration with water in the water jacket and no water in the MPC was used for all personnel dose calculations. After the MPC is removed from the spent fuel pool there is water in the MPC and no water in the water jacket. The HI-TRAC is then placed in the cask washdown area where the water jacket is filled to provide additional shielding before significant work begins. This configuration of water in both the water jacket and the MPC exists through the MPC closure operations until the moisture removal phase begins. Therefore, the dose calculations with water in the water jacket but not in the MPC are conservative for the conditions existing prior to the moisture removal since the additional shielding provided by the water in the MPC is not accounted for. The temporary shielding which will be placed above the water jacket Report: HI-2002563 Page: 15

surrounding the HI-TRAC transfer cask consists primarily of water. Credit was taken for this shielding in one set of calculations.

The following three variations of the normal configuration were analyzed.

Normal conditions with water in the water jacket and no water in the MPC. HI-TRAC lid is installed.

Normal conditions with water in the water jacket and no water in the MPC. HI-TRAC lid is not installed.

Normal conditions with water in the water jacket and no water in the MPC. HI-TRAC lid is not installed but the temporary shielding is installed. The temporary shielding was modeled as water extending from the water jacket upward to the top of the upper forging and out radially to the edge of the water jacket.

The results of these calculations are presented in Appendix J and the input files used are listed in Section 8.

7.7.2 Accident Conditions In addition to the normal condition dose rates, the dose rate around the transfer cask was calculated Ior a postulated accident scenario in which the transfer cask undergoes a vertical drop.

The potential consequence of this vertical drop is a slump in the lead below the uipper forging and below the lifting trunnion. The magnitude of this lead slump was calculated to be less than 0.7 inches in reference [13]. Conservatively, the lead slump was modeled as 0.8 inches. The dose rate was calculated in a localized area around the lifting trunnion. This area was the width of the lifting trunnion and 4.125 inches in height and fully encompassed the lead slump below the lifting trunnion. The region of highest dose for the lead slump accident occurs below the lifting trunnion since this region is closest to the active fuel zone.

The dose rate was calculated over the same area at radial distances equal to the edge of the water jacket, and 6 inches, I foot, 2 foot, and I meter from the edge of the water jacket. For the lead slump calculations, the water in the water jacket and MPC were not modeled. The dose rate was calculated before and after the lead slump event.

The results of these calculations are presented in Appendix J and the input files used are listed in Section 8. The dose consequences from the accidental loss of the water jacket are reported in references [4] and [10].

7.8 Doses During Loading and Unloading Operations The results from the calculation of dose rates around the HI-STORM overpack and the HI-TRAC transfer cask with design basis fuel were used to estimate the occupational exposure Report: HI-2002563 Page: 16

during the loading and unloading operations. The results of this estimate are presented in Appendix K.

The sequence of steps listed for loading and unloading is meant to be representative of the loading and unloading process and not an exhaustive list of the processes. The estimated duration of the step is provided as well as the time in the dose field. The difference between the two is that the process may take X amount of time but the workers only have to be near the cask for Y amount of time. Consistent with ALARA principles, the workers will be in an area of lower dose when their presence is not required around the transfer or storage cask. Likewise, the number of workers around the transfer or storage cask will be kept to a minimum. The dose rates reported in Appendix K were calculated by taking the number of workers times the dose at the location times the time in the dose field. Appendix K provides additional discussion about the dose calculations for the occupational exposure estimates.

As mentioned in Section 7.7, the MPC will be filled with water during a portion of the loading and unloading operations. Conservatively, this additional shielding was not credited in the analysis.

8. Computer Files All computer runs listed here were made on PCs at Holtec's main office. All files are stored on the lioltec computer server in directory \\projects\\1073\\eredmond.

The following is a list of all MCNP runs that were used in this report. See Section 7 for details of the calculations.

Input File Description MCNP HI-STORM IOOS calculations for the ISFSI h4c5dl I h4c5dM2 h4c5eO7 h4n5dl I h4n5dl2 h4n5eO7 h4p5dl 1 h4p5dl2 Report: HI-2002563 Page: 17

Input Fe Description h4q5dl I h4q5dl2 h4r5dl 1 h4r5dl2 h4p5eO8 h4pSell h4q5eO7 h4q5elO h4r5eO7 h4r5el 0 h4c5d21 h4n5dl5 h4p5dl5 h4q5dl5 h4r5d15 h4c5d22 h4nMdl6 h4p5dl6 h4q5dl6 h4r5dl6 MCNP HI-STORM IOOS calculations inside the ISFSI h4c5d23 Report: HI-2002563 Page: 18

Input File Description h4n5dl 7 h4pSdl 7 h4q5dl7 h4rSdl7 h4p5el4 h4q5el3 h4r5el 3 MCNP HI-STORM IOOS without a lid calculations h4c5eO5 MPC-24 cobalt run to generate surface source out top of overpack without lid h4c5eO6 MPC-24 run using h4c5eO5 calculating dose versus distance h4n5eO5 MPC-24 neutron run to generate surface source out top of overpack without lid h4n5eO6 MPC-24 run using h4n5eO5 calculating dose versus distance h4p5eO9 MPC-24 photon run to generate surface source out top of overpack without lid

- 0.7-1.5 MeV J

h4p5elO MPC-24 run using h4p5eO9 calculating dose versus distance h4q5eO8 MPC-24 photon run to generate surface source out top of overpack without lid

- 1.5-3.0 MeV h4q5eO9 MPC-24 run using h4q5eO8 calculating dose versus distance h4r5eO8 MPC-24 photon run to generate surface source out top of overpack without lid

_ _ - 0.3-0.7 MeV h4r5eO9 MPC-24 run using h4r5eO8 calculating dose versus distance h4c3dO3 h4p3dO3 h4q3dO3 h4r3dO3 h4n3dO3 h4c5d24 h4c5d25 h4n5dI9 h4n5d21 Report: HI-2002563

-Pa-ge: 19

Input File Description h4p5dl9 h4p5d23 h4q5dl9 h4q5d23 h4r5dl 9 h4r5d23 MCNP 125-ton HI-TRAC calculations with tallies along the axial length at various radial distances - normal conditions with water in the water jacket but not in the MPC u4c5aO2 MPC-24 cobalt source - HI-TRAC lid installed u4p5aO2 MPC-24 photon source - HI-TRAC lid installed - 0.7-3.0 MeV u4q5aO2 MPC-24 photon source - HI-TRAC lid installed - 0.45-0.7 MeV u4n5aO2 MPC-24 neutron source - HI-TRAC lid installed u4c5a04 MPC-24 cobalt source - HI-TRAC lid not installed u4p5aG4 MPC-24 photon source - HI-TRAC lid not installed - 0.7-3.0 MeV u4q5aO4 MPC-24 photon source - HI-TRAC lid not installed - 0.45-0.7 MeV u4n5aO4 MPC-24 neutron source - HI-TR-AC lid not installed u4c5aO5 MPC-24 cobalt source - HI-TRAC lid not installed - temporary shielding installed u4p5aO5 MPC-24 photon source - HI-TRAC lid not installed - temporary shielding installed - 0.7-3.0 MeV u4q5aO5 MPC-24 photon source - HI-TRAC lid not installed - temporary shielding installed - 0.45-0.7 MeV u4n5aO5 MPC-24 neutron source - HI-TRAC lid not installed - temporary shielding installed MCNP 125-ton HI-TRAC calculations with tallies along the axial length at various radial distances - accident conditions with no water in the water jacket or the MPC u4c5c04 MPC-24 cobalt source - HI-TRAC lid installed u4c5c05 MPC-24 cobalt source - HI-TRAC lid installed - 0.8 inch lead slump u4p5c04 MPC-24 photon source - HI-TRAC lid installed - 0.7-3.0 MeV u4p5c05 MPC-24 photon source - HI-TRAC lid installed - 0.8 inch lead slump - 0.7-3.0 MeV u4q5cO4 MPC-24 photon source - HI-TRAC lid installed - 0.45-0.7 MeV u4q5cO5 MPC-24 photon source - HI-TRAC lid installed - 0.8 inch lead slump - 0.45-0.7 MeV u4n5cO4 MPC-24 neutron source - HI-TRAC lid installed u4n5cO5 MPC-24 neutron source - HI-TRAC lid installed - 0.8 inch lead slump MCNP 125-ton HI-TRAC calculations with radial tallies along the top and bottom surface -

normal conditions with water in the water jacket but not in the MPC Report: HI-2002563

-Page: 20

Input File Description ht24c03 MPC-24 cobalt source - HI-TRAC lid installed - pool lid installed ht24nO3 MPC-24 neutron source - HI-TRAC lid installed - pool lid installed ht24pO3 MPC-24 photon source - HI-TRAC lid installed - pool lid installed - 0.7-3.0 MeV ht24p23 MPC-24 photon source - HI-TRAC lid installed - pool lid installed - 0.45-0.7 MeV u4c5aO6 MPC-24 cobalt source - HI-TRAC lid installed - pool lid with 2.5 inches Holtite-A on bottom u4n5aO6 MPC-24 neutron source - HI-TRAC lid installed - pool lid with 2.5 inches Holtite-A on bottom u4p5aO6 MPC-24 photon source - HI-TRAC lid installed - pool lid with 2.5 inches Holtite-A on bottom - 0.7-3.0 MeV u4q5aO6 MPC-24 photon source - HI-TRAC lid installed - pool lid with 2.5 inches Holtite-A on bottom - 0.45-0.7 MeV u4c3aOl u4p3aO1 u4q3aOl u4n3aOI B&W 15x15 ORIGEN-S and SAS2H input files a7Oa4ala.inp SAS2H input file for 70,000 MWD/MTU 4.0 wt.% 235U B&W 15xl5 a45a4a3a.inp ORIGEN-S input file for 45,000 MWD/MTU 4.0 wt.% 235U B&W 15x15 a45a4a3c.inp ORIGEN-S input file for 45,000 MWD/MTU 4.0 wt.% 235U B&W 15xl5 with.

I gm Co-59 a5Oa4a3a.inp ORIGEN-S input file for 50,000 MWD/MTU 4.0 wt.% 235U B&W l5x15 aSOa4a3c.inp ORIGEN-S input file for 50,000 MWD/MTU 4.0 wt.% 235U B&W l5xl5 with I g Co-59 a52m4a3a.inp ORIGEN-S input file for 52,500 MWD/MTU 4.0 wt.% 235U B&W 15x15 a52m4a3c.inp ORIGEN-S input file for 52,500 MWD/MTU 4.0 wt.%... U B&W 15xl5 with 5I gmCo-S9 a55a4a3a.inp ORIGEN-S input file for 55,000 MWD/MTU 4.0 wt.% 235U B&W 15xl5 a55a4a3c.inp ORIGEN-S input file for 55,000 MWD/MTU 4.0 wt.%.. U B&W l5xl5 with 1 gin Co-59 a7Oa4kla.inp SAS2H input file for 70,000 MWD/MTU 4.5 wt.% 235U B&W 15x15 a6Oa4k3a.inp ORIGEN-S input file for 60,000 MWD/MTU 4.5 wt.% 235U B&W 15x15 a6Oa4k3c.inp ORIGEN-S input file for 60,000 MWD/MTU 4.5 wt.% 235U B&W 15x15 with 1 gin Co-59 15bwxsc.in SAS2H input file for 42,500 MWD/MTU 2.9 wt.% 235U B&W 15x15 15bw325.in ORIGEN-S input file for 32,500 MWD/MTU 2.9 wt.% 2-`U B&W 15xl5 15bw325c.in ORIGEN-S input file for 32,500 MWD/MTU 2.9 wt.%... U B&W l5x15 with I gm Co-59 l5bw xsd.in SAS2H input file for 42,500 MWD/MTU 3.2 wt.% 235U B&W 15xl5 15bw375.in ORIGEN-S input file for 37,500 MWD/MTU 3.2 Wt.% 235U B&W 15x15 Report: HI-2002563 Page: 21

Input File Description 15bw375c.in ORIGEN-S input file for 37,500 MWD/MTU 3.2 wt.% 235U B&W 15x15 with I gn Co-59 15bwxs.in SAS2H input file for 45,000 MWD/MTU 3.4 wt.% 235U B&W 15x15 15bw415.in ORIGEN-S input file for 41,500 MWD/MTU 3.4 wt.% 2"U B&W 15x15 15bw415c.in ORIGEN-S input file for 41,500 MWD/MTU 3.4 wt.% `-U B&W 15x15 with I gm Co-59 15bw_xse.in SAS2H input file for 50,000 MWD/MTU 3.6 wt.% 235U B&W 15x15 15bw475.in ORIGEN-S input file for 47,500 MWD/MTU 3.6 wt.% 235U B&W 15x15 15bw475c.in ORIGEN-S input file for 47,500 MWD/MTU 3.6 wt.% 235U B&W 15x15 with I gmn Co-59 Westinghouse 17x17 ORIGEN-S and SAS2H input files 17w29xs.in SAS2H input file for 42,500 MWD/MTU 2.9 wt.% 35U W17x17 17w4Oxs.in SAS2H input file for 60,000 MWD/MTU 4.0 wt.% 235U W17xI7 17w325.in ORIGEN-S input file for 32,500 MWD/MTU 2.9 wt.% 2--U W17x17 17w325c.in ORIGEN-S input file for 32,500 MWD/MTU 2.9 wt.%... U W17xl7 with I gm Co-59 17w550.in ORIGEN-S input file for 55,000 MWD/MTU 4.0 wt.% 235U W17x17 17w550c.in ORIGEN-S input file for 55,000 MWD/MTU 4.0 wt.% 235U W17x17 with 1 gm Co-59 Report: HI-2002563 Page: 22

9. Summary The shielding analysis of the Independent Spent Fuel Storage Installation (ISFSI) at Diablo Canyon Power Plant (DCPP) is presented in this report. The facility consists of up to 140 HI-STORM 100S casks and is assumed to be filled with fuel of 32,500 MWD/MTU and a minimum cooling time of 5 years. Figure 1 shows the ISFSI configuration with the age of the fuel in the cask shown inside the circle representing the cask. The dose rates were calculated for a single cask and the entire array. The results of the single cask analysis are presented below for the surface and I meter dose rates.

Dose Rates on the Surface of the HI-STORM 100S 32,500 MWD/MTU and 5 Year Cooling (mrem/hr)

Location Neutron Photont Cobalt Total Bottom duct 2.90 6.80 17.01 26.72 Fuel midplane 0.76 33.92 0.11 34.79 Top duct 2.55 9.90 18.32 30.77 Top of overpack, above shield block 0.91 1.56 1.47 3.93 Top of overpack, above exit ducts 13.00 2.45 13.42 28.87 photons from neutron interactions are included in the photon dose Dose Rates at 1 meter from the HI-STORM 100S 32,500 MWD/MTU and 5 Year Cooling (mrem/hr)

Location Neutron Photont Cobalt Total Bottom duct 0.34 4.84 5.33 10.50 Fuel midplane 0.35 17.04 0.57 17.95 Top duct 0.37 4.64 5.00 10.02 Top of overpack 0.42 0.39 0.51 1.32 photons from neutron interactions are included in the photon dose Report: HI-2002563 Page: 23

For the operation of the ISFSI, the following dose rates are calculated.

Operation Persons Occupancy Dose Rate Dose (hours/year)

(mrem/hr)

(person-rem/year)

Walk-down 1

122 15.0 1.83 Repairs 2

12 65.0 1.56 Construction of ISFSI pad 7 15 480 6.02 43.3 For other on-site and off-site dose locations, the following dose rates are calculated. Distances specified are distances from the cask surface.

Location Distance Occupancy Dose m

ft hours mrem/hr mremlyear Nuisance 30.48 100 n/a 1.87 n/a

Fence, Front Make-up 67.97 223 2080 5.08E-01 1.06E+3 Water Facility Reactor 243.23 798 2080 2.23E-02 4.65E+01 Site 426.72 1400 2080 2.72E-03 5.65 Boundary I

Nearest 2414 7920 8760 4.03E-08 3.53E-04 Resident I

The hourly and annual dose rates from the cask array as a function of distance are presented in the following table. The annual dose rates are specified for occupancy factors of 8760 hrs/yr and for 2080 hrs/yr. The doses were calculated at locations that were perpendicular to the long side of the array at distances ranging from 40 feet to 600 m from the array. Doses for the distance of 1.5 miles (nearest resident) were determined by extrapolation. Up to 600 m, linear interpolation can be used to determine the dose at intermediate distances.

Report: HI-2002563 Page: 24

Hourly and Annual Dose Rates at Distances from the ISFSI Distance Dose m

ft mremlyr mreml/r mremlhr (8760 (2080 hrs/yr) hrs/yr) 4.57 15.00 1.05E+05 2.50E+04 1.20E+0O 6.10 20.00 8.73E+04 2.07E+04 9.97E+00 15.24 50.00 3.92E+04 9.3 IE+03 4.47E+0O 30.48 100.00 1.64E+04 3.89E+03 1.87E+00 33.53 110.00 1.43E+04 3.40E+03 1.63E+00 42.67 140.00 9.73E+03 2.31E+03

1. IIE+00 45.72 150.00 8.67E+03 2.06E+03 9.90E-01 60.96 200.00 5.35E+03 1.27E+03 6.11E-01 67.97 223.00 4.45E+03 1.06E+03 5.08E-01 79.25 260.00 3.32E+03 7.89E+02 3.79E-01 91.44 300.00 2.46E+03 5.85E+02 2.81E-01 105.16 345.00 1.85E+03 4.39E+02 2.11E-01 135.64 445.00

_.OIE+03 2.41E+02

1. 16E-0 I 243.23 798.00 1.96E+02 4.65E+01 2.23E-02 250.00 820.21 1.79E+02 4.26E+01 2.05E-02 300.00 984.25 9.90E+01 2.35E+01 1.13E-02 350.00 1148.29 5.50E+01 1.31E+01 6.28E-03 400.00 1312.34 3.14E+01 7.45E+00 3.58E-03 426.72 1400.00 2.38E+01 5.65E+00 2.72E-03 450.00 1476.38 1.88E+01 4.46E+00 2.14E-03 500.00 1640.42 1.17E+01 2.77E+00 1.33E-03 550.00 1804.46 7.51E+00 1.78E+00 8.57E-04 600.00 1968.50 4.99E+00 1.18E+00 5.69E-04 7920 2414.02 (1.5 miles) 3.53E-04 8.38E-05 4.03E-08 Report: HI-2002563 Page: 25

The following tables contain total dose rates from effluent release and direct radiation compared to regulatory limits. Dose rates from effluent release were taken from [8]. The first table lists dose rates for normal and off-normal conditions. The second table lists doses for accident conditions for a 30 day period. Both tables are for an individual at the site boundary (1400 ft).

Dose Rate from Direct Dose Total Dose Rate Regulatory Effluent Release Rate (mrem/year)

Limit (mremlyear)

(mrem/year)

Reference 181 This report 8760 hour0.101 days 2.433 hours 0.0145 weeks 0.00333 months s/yr 2080 hour0.0241 days 0.578 hours 0.00344 weeks 7.9144e-4 months s/yr IOCFR72.104(a) - Normal (140 casks)

Whole Body 0.27 5.65 5.92 25 ADE Thyroid ADE 0.043 5.65 5.693 75 Critical Organ 1.46 5.65 7.11 25 ADE (Max) l I OCFR72. 104(a) - Off-Normal (140 casks)

Whole Body 0.75 5.65 6.40 25 ADE Thyroid ADE 0.06 5.65 5.71 75 Critical Organ 5.49 5.65 11.14 25 ADE (Max)

ADE: Annual Dose Equivalent It has to be noted that the dose rates from other Uranium Fuel Cycle operations have to be added to the values listed above to determine whether the regulatory requirements for normal conditions (IOCFR72.104) are met.

Report: HI-2002563 Page: 26

Dose from Direct Dose Total Dose Regulatory Effluent Release Rate (mrem/30 days)

Limit (mrem/30 days)

(mreml30 days)

(mrem/30 days)

Reference 181 (This report) 10CFR72.106(b) - Accident (I cask)

TEDE 0.83 1.96 2.79 5000 TODE = DDE +

6.36 1.96 8.32 50000 CDE (Max)

LDE 0.022 1.96 1.982 15000 SDE 0.026 1.96 1.986 50000 ADE: Annual Dose Equivalent; TEDE: Total Effective Dose Equivalent; TODE: Total Organ Dose Equivalent; DDE: Deep Dose Equivalent; LDE: Lens Dose Equivalent; SDE: Shallow Dose Equivalent Please note that the direct dose under accident conditions is identical to the direct dose under normal conditions since there are no accidents which significantly affect the shielding of the HI-STORM overpack. Therefore, as a conservative estimate, the dose from the complete ISFSI at the site boundary over a period of 30 days (2.72E-03 mrem/hr

30 days

24 hr/day) is used in above table.

Report: HI-2002563 Page: 27

10. Referencest

[I] 0. W. Hermann, C. V. Parks, SAS2H: A Coupled One-Dimensional Depletion and Shielding Analysis Module, NUREG/CR-0200, Revision 5, (0RNL/NUREG/CSD-2N2/R5), Oak Ridge National Laboratory, September 1995.

(2] 0. W. Hermann, R. M. Westfall, ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms, NUREG/CR-0200, Revision 5, (ORNL/NUREG/CSD-2/V2/R5),

Oak Ridge National Laboratory, September 1995.

[3] J. F. Briesmeister, editor, MCNP - A General Monte Carlo N-Particle Transport Code, LA-12625-M, Los Alamos National Laboratory, November 1993.

[4] Final Safety Analysis Reportfor the HI-STORM 100 Cask System, HI-2002444, Rev. 0, Holtec International (US NRC Docket No. 72-1014).

[5] PG&E Specification 10012-N-NPG.

[6] HI-STORM Shielding Design and Analysisfor Storage, HI-971608, Rev. 13, Holtec International.

[7] "Diablo Canyon ISFSI Project - Diablo Canyon Units 1 and 2 - Transmittal of design inputs, site information and reference identification", Letter from R. L. Klimczak (PG&E) to E.

Lewis (Holtec), January 23h", 2001.

[8] Diablo Canyon ISFSI Site Boundary Confinement Analysis, HI-2002513, Rev. 4, Holtec International.

[9] Real Individual, USNRC Interim Staff Guidance (ISG) No. 13, Rev. 0, May 2000.

[10]

HI-STORM 100 License Amendment Request (LAR) 1014-1, Revision 2, July 2001.

Note: This revision status of Holtec documents cited above is subject to updates as the project progresses. This document will be revised if a revision to any of the above-referenced Holtec work products materially affects the instructions, results, conclusions or analyses contained in this document. Otherwise, a revision to this document will not be made and the latest revision of the referenced Holtec documents shall be assumed to supersede the revision numbers cited above. The Holtec Project Manager bears the undivided responsibility to ensure that there is no intra-document conflict with respect to the information contained in all Holtec-generated documents on a safety-significant project. The latest revision number of all documents produced by Holtec International in a safety-significant project is readily available from the company's Document Transmittal Form (DTF) database.

Report: HI-2002563 Page: 28

[I1]

"Diablo Canyon ISFSI Project - Diablo Canyon Units 1 and 2 - Transmittal of PG&E design inputs", Letter from Richard Klimczak (PG&E) to Eric Lewis (Holtec),

November 17, 2000.

[12]

"Diablo Canyon ISFSI Project - Diablo Canyon Units 1 and 2 Transmittal of design inputs, site information and reference identification.", Letter from Richard Klimczak (PG&E) to Eric Lewis (Holtec), February 19, 2001.

[13]

Structural Analysis Of HI-TRA C -Vertical Drops And Tipover In Diablo Canyon Fuel Building, HI-2002506, Rev. 0, Holtec International.

[14]

HI-STAR Shielding Design and Analysisfor Transport and Storage, HI-951322, Rev. 12, Holtec International.

[15]

HI-STORM 100 System Additional Shielding Calculations,HI-2012702, Rev. 4, Holtec International.

10.1 Drawings The following is a list of the applicable drawings and Engineering Change Orders (ECOs) that were used to generate the MCNP models used in this analysis.

The HI-STORM 100S design was enhanced and as a result the analysis in this report was updated to include the new design. The significant differences between the old design and the new design were the removal of the inner shield shell and compensating change in concrete density and minor changes to the lid. From the exterior of the overpack, these changes are not noticeable. All MCNP calculations were rerun in revision 3 of this report. However, all calculations that used the surface source files utilized the older model. This is acceptable because in these MCNP runs the source is being started on the exterior of the overpack and the outer steel shell of the overpack was unaffected in the design enhancements. Therefore, scattering off the exterior of the overpack is unaffected.

Revision 3 of the HI-STORM 100S drawings introduced a variable height for the body of the overpack. Conservatively, the shorter version of the HI-STORM 1 OOS was analyzed. This shorter version of the HI-STORM IOOS was the same overpack configuration that was analyzed in Reference [10] and therefore this analysis is consistent with HI-STORM FSAR and license amendments.

Drawing Number Revision Drawing Number Revision Number Number HI-STORM 100S 3443 3

1 l

HI-STORM 100S - Earlier Design 3067 0

3074 0

3068 0

3075 0

3070 0

BM-3065 0

Report: HI-2002563 Page: 29

Drawing Number Revision Drawing Number Revision Number I

Number 3073 0

BM-3066 0

MPC-24 1395 Sheet I of 4 12 1396 Sheet I of 6 15 1395 Sheet2of4 10 BM 1478 Sheet I of 2 10 1395 Sheet3of4 11 BM 1478 Sheet2of2 13 125-ton HI-TRAC 1880 Sheet I of 10 8

1880 Sheet 7 of 10 8

1880 Sheet 2 of 10 9

1880 Sheet 8 of 10 8

1880 Sheet 3 of 10 8

BM 1880 Sheet I of 2 8

1880 Sheet 4 of 10 9

BM 1880 Sheet 2 of 2 6

1880 Sheet 5 of 10 9

ECO 1025-5 1

1880 Sheet 6 of IO 9

ECO 1025-6 0

Report: HI-2002563 Page: 30

(

( (.

C A

N Pad 7 Pad 6 Pad 5 Pad 4 Pad 3 Pad 2 Pad 1 I

I 0D 0D 0D 0D 0D 0D 0~

0D 0D 0D 0D 0D 0O 0D 0i 0i 0D 0i 0D 07 0i 0D 0i88

8(9

@)(j) (i) 1(i)0 (H) (@~

C1) 1i) 9999 9999

9

9 9

( (i ) I(* * ( @

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(* * ( (

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Figure 1:

Arrangement of the ISFSI at DCPP with assumed cooling time for each cask position.

Report: HI-2002563 Page: 31

PROPRIETARY Figure 2: The three different MCNP models used in the analysis of the ISFSI at DCPP are depicted above.

Report: HI-2002563 Page: 32

Appendix A:

Bounding Burnup and Cooling Time for HI-STORM This appendix determines the bounding MPC with the bounding burnup and cooling time for the HI-STORM analysis.

The following burnup and cooling times were analyzed from the allowable bumup and cooling times. The enrichment used for the analysis is also shown.

MPC Burnup (MWD/MTU)

Cooling Time (years)

Enrichment (wt.% 235U) 24 41,500 5

3.4 24 45,000 6

4.0 24 50,000 8

4.0 24 52,500 10 4.0 24 55,000 12 4.0 32 32,500 5

2.9 32 37,500 7

3.2 32 45,000 8

4.0 The last bumup and cooling time for the MPC-32 has a conservatively higher burnup for the allowable cooling time of 8 years.

The results presented on the pages that follow are summary results from EXCEL for the dose rate on the surface and one meter away from the side and top of the overpack. A brief description of the format of these pages can be found in Appendix D. The dose locations shown in the left hand column are the same locations that are used in the LAR to the HI-STORM FSAR [10]. The segments listed are the segments on the MCNP surfaces that give the highest dose rate.

Appendix D provides a detailed description of the dose locations and presents detailed results for the bounding burnup and cooling time.

The results indicate that the bounding MPC is the MPC-32 with a burnup and cooling time of 32,500 MWD/MTU and 5 year cooling. This combination has the highest dose rate on the radial surface of the cask. The highest dose rate on the top of the cask is achieved with a different burnup and cooling time combination. However, since the contribution to the off-site dose rate from the top of the overpack is a small fraction of the total dose from the overpack, the MPC-32 with 32,500 MWD/MTU and 5 year cooling also produces the highest off-site dose. This is demonstrated in Appendix E.

Report: HI-2002563 Page: A-lI

41.5-5 mpc24 storm-sum-ri.xls 41,500 MWD/MTU burnup - B&W 15xl5 fuel element 3.4 w/o U235 5.0 YR 5.0 YR 5.0 YR 5.0 YR h4nSdl Im h4n5dl Im BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dl Im h4q5dl Im h4cSdl In h4r5dl Im num fuel 24 h4p5eO8m h4qSeO7m h4rfeO7m neutron phot (n,p) phot cobalt total value rel err value

-re err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 4.41 2

7 1.15 3

1 3.87 4

5 1.38 4a 1

19.75 0.03 0.04 0.03 0.02 0.02 0.42 0.03 6.35 0.07 14.66 1.26 0.02 30.30 0.01 0.09 0.17 0.06 9.63 0.06 15.46 0.56 0.02 1.14 0.05 1.24 0.50 0.04 1.96 0.04 11.33 0.01 25.84 0.02 0.41 32.81 0.01 0.01 29.14 0.02 0.04 4.31 0.02 0.02 33.54 0.01 one meter I

I 2

8 3

1 4

6 0.51 0.53 0.57 0.6:'

0.08 0.03 0.07 0.02 0.20 0.61 0.10 0.19 0.17 4A3 0.07 0.02 15.21 0.01 0.13 4.39 0.08 0.03 0.25 0.05 4.59 0.49 4.22 0.43 0.04 9.73 0.04 0.11 16.84 0.01 0.03 9.27 0.04 0.03 1.50 0.(,2 adjacent neutrons fuel gam co-60 total 1

4.41 6.77 14.66 25.84 2

1.15 31.56 0.09 32.81 3

3.87 9.80 15.46 29.14 4

1.38 1.70 1.24 4.31 4a 19.75 2A6 11.33 33.54 one meter 1

0.51 4.63 4.59 9.73 2

0.53 15.82 0.49 16.84 3

0.57 4.49 4.22 9.27 4

0.63 0.44 0.43 1.50 Report: HI-2002563 Page A-2

45-6 mpc24 storm-sum-rI.xIs 45,000 MWD/MTU BW 15x 15 4.0 wt% U235 4.0 wt% U235 6.0 YR 6.0 YR 6.0 YR 6.0 YR h4nSdl Im h4nSd1 Im BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dlI m h4q5dl Im M4c5dl In h4rSdl Im M4p5eO8m h4q5eO7m h4r5eO7m nurn fuel 24 neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

I 4.49 2

7 1.17 3

1 3.94 4

5 1.40 4a 1

20.12 0.03 0.43 0.03 5.38 0.07 12.70 0.04 1.29 0.02 22.72 0.02 0.08 0.03 0.18 0.06 8.42 0.06 13.69 0.02 0.57 0.02 0.94 0.06 1.10 0.02 0.51 0.04 1.50 0.04 10.03 0.01 23.01 0.02 0.41 25.25 0.01 0.01 26.23 0.02 0.04 4.01 0.02 0.02 32.17 0.01 one meter I

1 0.52 2

8 0.54 3

1 0.58 4

1 0.74 0.08 0.20 0.17 3.54 0.08 3.98 0.03 0.62 0.02 11.36 0.01 0.43 0.07 0.10 0.13 3.70 0.09 3.74 0.04 0.29 0.04 0.14 0.04 0.26 0.04 8.24 0.04 0.11 12.94 0.01 0.03 8.12 0.04 0.06 1.44 0.o3 adjacent neutrons fuel gam co-60 total 1

4.49 5.81 12.70 23.01 2

1.17 24.00 0.08 25.25 3

3.94 8.60 13.69 26.23 4

1.40 1.51 1.10 4.01 4a 20.12 2.01 10.03 32.17 one meter 2

3 4

0.52 3.74 0.54 11.98 0.58 3.80 0.74 0.44 3.98 8.24 0.43 12.94 3.74 8.12 0.26 1.44 Report: HI-2002563 Page A-3

S0-8 mpc24 storm-sum-rl.xis 50,000 MWD/MTU BW 15xl5 4.0 wt% U235 4.0 wt% U235 8.0 YR 8.0 YR 8.0 YR 8.0 YR h4nSd1Im h4n5dlI m BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dl1mh4q5dllm b4c5dlln h4r5dl Im h4p5eO8m h4q5eO7m h4r5eO7m num fuel 24 neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 6.35 2

8 1.88 3

1 5.57 4

1 2.89 4a 1

28.44 0.03 0.04 0.03 0.03 0.02 0.61 1.95 0.25 1.37 0.72 0.03 4.27 0.02 15.93 0.06 6.90 0.03 0.01 0.04 1.09 0.08 10.71 0.01 21.94 0.02 0.02 0.04 0.46 19.79 0.01 0.06 11.89 0.01 24.62 0.02 0.07 0.20 0.13 4.47 0.02 0.05 8.71 0.02 38.97 0.02 one meter I

1 0.74 2

8 0.76

3. 1 0.81 4

1 1.04 0.08 0.03 0.07 0.04 0.28 0.87 0.14 0.42 0.17 0.02 0.13 0.04 2.69 8.07 2.98 t.11 0.08 0.01 0.09 0.05 3.35 0.36 3.25 0.23 0.04 7.07 0.04 0.11 10.07 0.01 0.03 7.18 0.04 0.06 1.79 0.03 adjacent neutrons fuel gam co-60 total 1

6.35 4.87 10.71 21.94 2

1.88 17.87 0.04 19.79 3

5.57 7.15 11.89 24.62 4

2.89 1.38 0.20 4.47 4a 28.44 1.82 8.71 38.97 one meter 1

0.74 2.97 3.35 7.07 2

0.76 8.95 0.36 10.07 3

0.81 3.12 3.25 7.18 4

1.04 0.52 0.23 1.79 Report: HI-2002563 Page A-4

52.5-10 mpc24 storm-sum-rl.xls BW 15x15 4.0 wt% U235 52,500 MWID/MTU 10.0 YR 10.OYR 10.OYR l0.OYR h4n5dl Im h4n5d I m BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dl Im h4q5dl lm h4c5dl In h4r5d IIm h4p5eO8m h4q5eO7m h4rfeO7m num fuel 24 neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 7.12 0.03 0.68 0.03 3.50 2

8 2.11 0.04 2.18 0.02 12.60 3

1 6.25 0.03 0.28 0.06 5.77 4

1 3.24 0.03 1.54 0.03 0.01 4a 1

31.89 0.02 0.81 0.04 0.87 0.09 0.02 0.07 0.07 0.05 8.60 0.01 19.90 0.02 0.03 0.46 16.92 0.01 9.98 0.01 22.28 0.02 0.17 0.13 4.95 0.02 7.31 0.02 40.87 0.02 one meter I

1 0.83 0.08 2

8 0.85 0.03 3

1 0.91 0.07 1

1.17 0.04 0.32 0.98 0.16 0.47 0.17 2.19 0.09 2.69 0.02 6.39 0.01 0.29 0.13 2.50 0.09 2.73 0.04 0.09 0.05 0.19 0.04 0.12 0.03 OJ'6 6.03 0.04 8.51 0.01 6.30 0.04 1.91 0.03 adjacent neutrons fuel gami co-60 total 1

7.12 4.18 8.60 19.90 2

2.11 14.78 0.03 16.92 3

6.25 6.05 9.98 22.28 4

3.24 1.55 0.17 4.95 4a 31.89 1.68 7.31 40.87 one meter I

2 3

4 0.83 0.85 0.91 1.17 2.50 2.69 7.37 0.29 2.65 2.73 0.55 0.19 6.03 8.51 6.30 1.91 Report: HI-2002563 Page A-5

55-12 mnpc24 storm-sum-rl.xls BW 15x15 4.0 wt%/ U235 55,000 MWDIMTU 12.0 YR 12.0 YR 12.0 YR 12.0 YR h4n5dllm h4n5dl lm BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dllmh4q5dllm h4c5dlln h4rfdl Im h4p5eO8m h4q5eO7m h4rfeO7m num fuel 24 neutron phot (n,p) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 7.89 0.03 0.76 2

8 2.34 0.04 2.42 3

1 6.92 0.03 0.31 4

1 3.59 0.03 1.70 4a 1

35.34 0.02 0.90 0.03 3.05 0.02 10.69 0.06 5.11 0.03 0.01 0.04 0.73 0.09 0.02 0.07 0.07 0.05 6.88 0.01 18.58 0.02 0.02 0.46 15.46 0.01 8.43 0.01 20.78 0.02 0.14 0.13 5.44 0.02 6.17 0.02 43.15 0.02 one meter 1

1 2

8 3

1 4

1 0.92 0.08 0.35 0.17 1.90 0.94 0.03 1.08 0.02 5.42 1.01 0.07 0.17 0.13 2.22 1.30 0.04 0.52 (1.04 0.07 0.10 0.01 0.09 0l. 0 2.15 0.04 0.24 0.12 2.31 0.03 0.16 0.06 5.32 0.04 7.68 0.01 5.71 0.04 2.05 0.03 adjacent neutrons fuel gain co-60 total 1

7.89 3.81 6.88 18.58 2

2.34 13.10 0.02 15.46 3

6.92 5.42 8.43 20.78 4

3.59 1.71 0.14 5.44 4a 35.34 1.63 6.17 43.15 one meter 2

3 4

0.92 0.94 1.01 1.30 2.25 2.15 5.32 6.50 0.24 7.68 2.39 2.31 5.71 0.59 0.16 2.05 Report: HI-2002563 Page A-6

32.5-5 mpc-32 storm-sum-rl.xls 32,500 MWD/MTU burnup - B&W 15xl5 fuel element 2.9 w/o U235 5.0 YR 5.0 YR 5.0 YR 5.0 YR h4n5dl Im h4n5dlI m BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4pSdllmh4q5dllm h4c5dlln h4rfdl Im num fuel 32 h4pSeO8m h4q5eO7m h4rfeO7m neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 2.90 0.03 0.28 0.03 6.53 2

7 0.76 0.04 0.83 0.02 33.09 3

1 2.55 0.03 0.11 0.06 9.79 4

5 0.91 0.02 0.37 0.02 1.19 4a 1

13.00 0.02 0.33 0.04 2.12 0.07 17.01 0.01 0.11 0.06 18.32 0.05 1.47 0.04 13.42 0.01 26.72 0.02 0.41 34.79 0.01 0.01 30.77 0.02 0.04 3.93 0.02 0.02 28.87 0.01 one meter I

I 2

8 3

1 4

6 0.34 0.08 0.13 0.17 4.71 0.06 0.35 0.03 0.40 0.02 16.64 0.01 0.37 0.07 0.06 0.13 4.58 0.08 0.42 0.02 0.12 0.03 0.26 0.05 5.33 0.57 5.00 0.51 0.04 10.50 0.04 0.11 17.95 0.01 0.03 10.02 0.04 0.03 1.32 0.02 adjacent neutrons fuel gam co-60 total 1

2.90 6.80 17.01 26.72 2

0.76 33.92 0.11 34.79 3

2.55 9.90 18.32 30.77 4

0.91 1.56 1.47 3.93 4a 13.00 2.45 13.42 28.87 one meter 1

0.34 4.84 5.33 10.50 2

0.35 17.04 0.57 17.95 3

0.37 4.64 5.00 10.02 4

OA2 0.39 0.51 1.32 Report: HI-2002563 Page A-7

37.5-7 mpc32 storm-sum-rl.xls 37,500 MWD/MTU burnup - B&W 5x1 5 fuel element 3.2 w/o U235 7.0 YR 7.0 YR 7.0 YR 7.0 YR h4n5dl Im 4n5dl lm BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dI Im h4q5dl Im h4c5dl In h4rfdI Im h4p5eO8m h4q5eO7m h4rMeO7m num fuel 32 neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 4.05 2

7 1.05 3

1 3.55 4

5 1.26 4a 1

18.12 0.03 0.39 0.04 1.16 0.03 0.16 0.02 0.51 0.02 0.46 0.03 4.91 0.02 19.96 0.06 7.81 0.02 0.85 0.04 1.33 0.08 14.14 0.02 0.09 0.06 15.73 0.06 1.26 0.05 11.52 0.01 23.49 0.02 0.41 22.26 0.01 0.01 27.25 0.02 0.04 3.89 0.02 0.02 31.44 0.01 one meter 1

1 2

8 3

1 4

6 0.47 0.48 0.52 0.58 0.08 0.03 0.07 0.02 0.18 0.56 0.09 0.17 0.17 0.02 0.13 0.03 3.19 9.96 3.42 0.18 0.08 4.43 0.01 0.48 0.09 4.30 0.06 0.44 0.04 8.27 0.04 0.11 IIA8 0.01 0.03 8.33 0.04 0.03 1.37 0.02 adjacent neutrons fuel gam co-60 total 1

4.05 5.30 14.14 23.49 2

1.05 21.11 0.09 22.26 3

3.55 7.97 15.73 27.25 4

1.26 1.36 1.26 3.89 4a 18.12 1.79 11.52 31.44 one meter 1

2 3

4 0.47 3.37 0.48 10.51 0.52 3.51 0.58 0.35 4.43 8.27 0.48 11.48 4.30 8.33 0.44 1.37 Report: HI-2002563 Page A-8

45-8 mpc32 storm-surn-rl.xls 45,000 MWD/MTU BW 15x 15 4.0 wt%/o U235 4.0 wt% U235 8.0 YR 8.0 YR 8.0 YR 8.0 YR h4n5dllm h4n5dllm BPRA Curies 331.00 TP Curies 0.00 dwg3443 h4p5dllmb4q5dllm h4c5dlln h4r5dl Im h4p5eO8m h4q5eO7m h4r5eO7m num fuel 32 neutron phot (n,p) phot cobalt total value rel err value rel err value rel err value rel err value rel err TSAR dose locations adjacent segs I

1 5.56 0.03 2

7 1.45 0.04 3

1 4.88 0.03 4

5 1.74 0.02 4a 1

24.90 0.02 0.53 1.59 0.22 0.70 0.63 0.03 5.04 0.08 13.03 0.02 19.26 0.02 0.08 0.06 8.16 0.06 14.73 0.02 0.86 0.06 1.18 0.04 1.29 0.05 10.79 0.01 24.17 0.02 0.41 22.38 0.02 0.01 27.98 0.02 0.03 4.47 0.02 0.02 37.62 0.02 one meter I

1 0.65 0.08 2

8 0.66 0.03 3

1 0.71 0.07 4

1 0.91 0.04 0.25 0.76 0.12 0.36 0.17 0.02 0.13 0.04 3.19 0.09 9.60 0.01 3.53 0.09 0.13 0.05 4.08 0.44 4.03 0.28 0.04 8.17 0.04 0.12 11.47 0.01 0.03 8.39 0.04 0.06 1.69 0.03 adjacent neutrons fuel gam co-60 total 1

5.56 5.57 13.03 24.17 2

1.45 20.85 0.08 22.38 3

4.88 8.38 14.73 27.98 4

1.74 1.56 1.18 4.47 4a 24.90 1.93 10.79 37.62 one meter 1

2 3

4 0.65 3.44 0.66 10.36 0.71 3.65 0.91 0.49 4.08 8.17 OA4 11.47 4.03 8.39 0.28 1.69 Report: HI-2002563 Page A-9

Appendix B:

Bounding Burnup and Cooling Time for the 125-ton HI-TRAC This appendix determines the bounding MPC with the bounding burnup and cooling time for the HI-TRAC analysis.

The following burnup and cooling times were analyzed from the allowable burnup and cooling times. The enrichment used for the analysis is also shown.

MPC Burnup (MWD/MTU)

Cooling Time (years)

Enrichment (wt.% 235U) 24 41,500 5

3.4 24 45,000 6

4.0 24 50,000 8

4.0 24 52,500 10 4.0 24 55,000 12 4.0 24 60,000 15 4.5 32 32,500 5

2.9 32 45,000 8

4.0 32 47,500 15 3.6 The second to last burnup and cooling time for the MPC-32 has a conservatively higher burnup for the allowable cooling time of 8 years.

The results presented on the pages that follow are summary results from EXCEL for the dose rate on the surface and one meter away from the side and top of the overpack. The dose locations are described below. A brief description of the format of these pages can be found in Appendix D. The segments listed are the segments on the MCNP surfaces that give the highest dose rate.

Appendix J provides a detailed description of the dose locations and presents detailed results for the bounding burnup and cooling time.

The design basis MPC and burnup and cooling time was based on a comparison of the radial dose rates along the outer surface of the HI-TRAC. The combination that had the highest dose rate at the midplane of the cask was chosen as the bounding condition. These results indicate that the bounding MPC is the MPC-24 with a burnup and cooling time of 55,000 MWD/MTU and 12 year cooling. This combination had the highest dose rate on the radial surface of the cask at the midplane while a couple of the other dose locations are slightly higher at other burnup and cooling times.

Report: HI-2002563 Page: B-1

Dose Locations Adjacent/ Surface of MIl-TRAC 817 Next to lower water jacket 825 Outer surface of water jacket at peak dose location (830 831)

Next to lifting trunnion in water jacket cutout area 832 Above water jacket and below top forge One Meter from HI-TRAC 829 segment 10 I meter from outer water jacket at axial height of lower water jacket 829 other segment I meter from outer water jacket at peak dose location (1829 2829)

Next to lifting trunnion in water jacket cutout area at distance of 1 meter from water jacket 829 segment 29 1 meter from water jacket at axial height above water jacket I and below top forging Report: HI-2002563 Page: B-2

41.5-5 mpc24 trac -sum.xl s 41,500 MWD/MTU burnup - B&W 15x15 fuel element 3.4 w/o U235 5.0 YR 5.0 YR 5.0 YR 5.0 YR BPRA curies 331.000 u4n5aO2m neutron u4nSaO2m phot (np) sidel25t u4p5aO2m u4q5aO2m u4c5aO2m phot cobalt num fuel 24 total value re] err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 74.68 825 19 49.33 (830 831) 232.37 832 146.10 0.02 13.01 0.02 5.26 0.02 34.97 0.01 69.34 0.04 4.59 0.11 2.18 0.02 2.35 0.07 1.01 0.03 77.77 0.01 170.72 0.01 0.02 0.01 0.24 153.64 0.01 0.12 115.50 0.02 354.64 0.03 0.07 46.45 0.01 195.91 0.01 one meter dose rates 829 10 11.26 0.03 4.76 0.02 9.62 829 18 17.70 0.02 11.20 0.01 32.02 (18292829) 11.18 0.10 2.93 0.09 6.40 829 29 9.42 0.03 3.00 0.03 5.36 0.02 8.67 0.01 34.30 0.01 0.01 0.60 0.01 61.51 0.01 0.07 11.22 0.02 31.73 0.04 0.02 7.30 0.03 25.07 0.02 adjacent neutrons fuel gam co-60 total 817 74.68 18.27 77.77 170.72 825 19 49.33 104.31 0.01 153.64 (830 831) 232.37 6.77 115.50 354.64 832 146.10 3.36 46.45 195.91 one metez 829 10 11.26 14.37 8.67 34.30 829 18 17.70 43.21 0.60 61.51 (18292829) 11.18 9.33 11.22 31.73 829 29 9.42 8.35 7.30 25.07 Report: HI-2002563 Page B-3

45-6 mpc24 trac-sum.xls 45,000 MWD/MTU BW 15x 15 4.0 wt/o U235 K.-

4.0 wt% U235 6.0 YR 6.0 YR 6.0 YR 6.0 YR u4n5aO2m u4n5aO2m BPRA curies 331.000 sidel25t u4p5aO2m u4q5aO2m u4c5aO2m num fuel 24 neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 76.08 0.02 13.25 825 19 50.25 0.02 35.62 (830 831) 236.71 0.04 4.68 832 148.83 0.02 2.40 0.02 3.87 0.04 67.38 0.01 52.18 0.02 0.01 0.11 1.57 0.14 102.17 0.07 0.71 0.08 41.09 0.01 160.58 0.01 0.24 138.05 0.01 0.02 345.13 0.03 0.01 193.03 0.01 one meter dose rates 829 10 11.47 0.03 4.84 829 18 18.03 0.02 11 AO (1829 2829) 11.39 0.10 2.98 829 29 9.59 0.03 3.05 0.02 7.23 0.02 0.01 24.11 0.02 0.09 4.79 0.08 0.03 4.01 0.03 7.51 0.52 9.94 6.47 0.01 31.05 0.01 0.01 54.06 0.01 0.02 29.10 0.04 0.03 23.13 0.02 adjacent neutrons fuel gam co-60 total 817 76.08 17.12 67.38 160.58 825 19 50.25 87.79 0.01 138.05 (830 831) 236.71 6.25 102.17 345.13 832 148.83 3.10 41.09 193 03 one meter 829 10 11.47 12.07 829 18 18.03 35.51 (1829 2829) 11.39 7.77 829 29 9.59 7.07 7.51 31.05 0.52 54.06 9.94 29.10 6.47 23.13 Report: HI-2002563 PageB-4

50-8 mpc24 trac-sumn.xls 50,000 MWD/MTU BW 15xl5 4.0 vt% U235 4.0 wt% U235 8.0 YR 8.0 YR 8.0 YR 8.0 YR BPRA curies 331.000 u4n5aO2m neutron u4n5aO2m pbot (n,p) sidel25t u4p5aO2m u4q5aO2m u4c5aO2m phot cobalt nurn fuel 24 total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 107.52 0.02 18.73 0.02 2.73 825 19 71.02 0.02 50.34 0.01 37.88 (830 831) 334.56 0.04 6.61 0.11 1.06 832 210.35 0.02 3.39 0.07 0.48 0.04 56.81 0.01 185.79 0.01 0.03 0.01 0.24 159.25 0.01 0.16 88.62 0.02 430.85 0.03 0.10 35.64 0.01 249.86 0.01 one meter dose rates 829 10 16.21 0.03 6.85 829 18 25.48 0.02 16.12 (1829 2829) 16.10 0.10 4.21 829 29 13.56 0.03 4.32 0.02 5.24 0.02 6.33 0.01 34.62 0.02 0.01 17.49 0.02 0.44 0.01 59.53 0.01 0.09 3.44 0.09 8.64 0.02 32.40 0.05 0.03 2.89 0.03 5.62 0.03 26.39 0.02 adjacent neutrons fuel gam co-60 total 817 107.52 21.46 56.81 185.79 825 19 71.02 88.22 0.01 159.25 (830 831) 334.56 7.68 88.62 430.85 832 210.35 3.87 35.64 249.86 one meter 829 10 16.21 12.08 6.33 34.62 829 18 25.48 33.61 0.44 59.53 (1829 2829) 16.10 7.66 8.64 32.40 829 29 13.56 7.21 5.62 26.39 Report: HI-2002563 Page B-5

52.5-10 mpc24 trac-sum.xis BW 15x1 54.0 wt% U235 52,500 MWD/MTU 10.0 YR 10.OYR 10.OYR 10.OYR u4n5aO2m u4n5aO2m BPRA curies 331.000 sidel25t u4pSaO2m u4qSaO2m u4cSaO2m num fuel 24 neutron phot (n,p) phot cobalt total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 120.54 0.02 21.00 825 19 79.63 0.02 56.44 (830 831) 375.10 0.04 7A2 832 235.84 0.02 3.80 0.02 2.18 0.01 30.54 0.11 0.82 0.07 0.38 0.04 45.59 0.03 0.00 0.16 74.23 0.10 29.86 0.01 189.32 0.02 0.24 166.61 0.01 0.02 457.57 0.03 0.01 269.88 0.01 one meter dose rates 829 10 18.17 0.03 7.68 0.02 4.22 829 18 28.57 0.02 18.07 0.01 14.08 (1829 2829) 18.05 0.10 4.72 0.09 2.76 829 29 15.20 0.03 4.84 0.03 2.32 0.03 5.08 0.02 0.36 0.09 7.26 0.03 4.73 0.01 35.15 0.02 0.01 61.08 0.01 0.02 32.80 0.06 0.03 27.09 0.02 adjacent neutrons fuel gam co-60 total 817 120.54 23.18 45.59 189.32 825 19 79.63 86.99 0.00 166.61 (830 831) 375.10 8.24 74.23 457.57 832 235.84 4.18 29.86 269.88 one meter 829 10 18.17 11.89 829 18 28.57 32.15 (1829 2829) 18.05 7.48 829 29 15.20 7.16 5.08 35.15 0.36 61.08 7.26 32.80 4.73 27.09 Report: HI-2002563 Page B-6

55-12 mpc24 trac-sum.xis BW l5x15 4.0 wt% U235 55,000 MWD/MTU 12.0 YR 12.0 YR 12.0 YR 12.0 YR BPRA curies 331.000 sidel25t u4p5aO2m u4qSaO2m u4c5aO2m num fuel 24 u4n5aO2m u4n5aO2m neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 133.59 825 19 88.25 (830 831) 415.72 832 261.38 0.02 23.28 0.02 62.56 0.04 8.22 0.02 4.21 0.02 1.86 0.01 26.22 0.11 0.69 0.07 0.33 0.04 36.49 0.03 0.00 0.16 62.56 0.11 25.17 0.01 195.23 0.02 0.24 177.04 0.01 0.02 487.19 0.03 0.01 291.09 0.01 one meter dose rates 829 10 20.14 829 18 31.66 (18292829) 20.00 829 29 16.85 0.03 8.51 0.02 3.62 0.02 20.03 0.01 12.07 0.10 5.24 0.09 2.36 0.03 5.36 0.03 1.99 0.03 0.02 0.09 0.03 4.07 0.01 36.34 0.02 0.29 0.01 64.05 0.01 6.14 0.02 33.74 0.06 4.00 0.03 28.20 0.02 adjacent neutrons fuel gam co-60 total 817 133.59 25.14 36.49 195.23 825 19 88.25 88.78 0.00 177.04 (830 831) 415.72 8.91 62.56 487.19 832 261.38 4.54 25.17 291.09 one meter 829 10 20.14 12.13 4.07 36.34 829 18 31.66 32.10 0.29 64.05 (1829 2829) 20.00 7.60 6.14 33.74 829 29 16.85 7.35 4.00 28.20 Report: Hl-2002563 Page B-7

60-15 mpc24 trac-surn.xls BW 15x15 4.5 wth U235 60,000 MWD/MTU 15.0 YR 15.0 YR 15.0 YR 15.0 YR u4n5aO2m u4nSaO2m BPRA curies 331.000 sidel25t u4p5aO2m u4q5aO2m u4c0aO2m phot cobalt num fuel 24 neutron phot (np) total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 136.11 0.02 23.71 825 19 89.91 0.02 63.73 (830 831) 423.53 0.04 8.37 832 266.29 0.02 4.29 0.02 1.57 0.04 25.00 0.01 22.12 0.03 0.00 0.11 0.57 0.16 47.82 0.07 0.28 0.11 19.25 0.01 186.40 0.02 0.24 175.76 0.01 0.02 480.30 0.04 0.01 290.11 0.02 one meter dose rates 829 10 20.52 0.03 8.67 0.02 3.05 0.03 2.79 829 18 32.26 0.02 20.41 0.01 10.17 0.02 0.20 (1829 2829) 20.38 0.10 5.33 0.09 1.99 0.09 4.73 829 29 17.16 0.03 5.46 0.03 1.68 0.03 3.09 0.01 35.03 0.02 0.01 63.04 0.01 0.02 32.43 0.07 0.03 27.39 0.02 adjacent neutrons fuel gam co-60 total 817 136.11 25.28 25.00 186.40 825 19 89.91 85.85 0.00 175.76 (830 831) 423.53 8.95 47.82 480.30 832 266.29 4.57 19.25 290.11 one meter 829 10 20.52 11.72 2.79 35.03 829 18 32.26 30.58 0.20 63.04 (1829 2829) 20.38 7.32 4.73 32.43 829 29 17.16 7.14 3.09 27.39 Report: HI-2002563 Page B-8

32.5-5 mpc32 trac-sum xIs 32,500 MWD/MTU burnup - B&W 15xl5 fuel element BPRA curies 331.000 2.9 w/o U235 5.0 YR 5.0 YR 5.0 YR 5.0 YR sidel25t num fuel 32 u4n5aO2m u4n5aO2m u4p5aO2m u4q5aO2m u4c5aO2m neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 49.14 0.02 8.56 0.02 5.82 0.03 90.21 0.01 153.73 0.01 825 18 30.74 0.02 23.16 0.01 78.11 0.02 0.01 0.21 132.02 0.01 (830 831) 152.87 0.04 3.02 0.11 2.40 0.12 136.70 0.02 295.00 0.02 832 96.12 0.02 1.55 0.07 1.13 0.07 54.98 0.01 153.78 0.01 one meter dose rates 829 10 7.41 0.03 3.13 0.02 10.59 0.02 10.05 0.01 31.18 0.01 829 18 11.65 0.02 7.36 0.01 35.22 0.01 0.70 0.01 54.92 0.01 (1829 2829) 7.36 0.10 1.92 0.09 7.05 0.07 13.30 0.02 29.64 0.03 829 29 6.20 0.03 1.97 0.03 5.89 0.02 8.65 0.03 22.71 0.01 adjacent neutrons fuel gam co-60 total 817 49.14 14.38 90.21 153.73 825 18 30.74 101.26 0.01 132.02 (830 831) 152.87 5.42 136.70 295.00 832 96.12 2.68 54.98 153.78 one meter 829 10 7.41 13.72 10.05 31.18 829 18 11.65 42.58 0.70 54.92 (1829 2829) 7.36 8.97 13.30 29.64 829 29 6.20 7.87 8.65 22.71 Report: HI-2002563 Page B-9

45-8 mpc32 trac-sum.xls 45,000 MWD/MTU BW lSx 15 4.0 wt% U235 4.0 wt% U235 8.0 YR 8.0 YR 8.0 YR 8.0 YR u4nSaO2m u4nSaO2m BPRA curies 331.000 sidel25t u4p5aO2m u4qSaO2m u4c5aO2m num fuel 32 neutron phot (n,p) phot cobalt total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 94.13 0.02 16.40 825 19 62.18 0.02 44.06 (830 831) 292.87 0.04 5.79 832 184.14 0.02 2.97 0.02 3.25 0.04 69.13 0.01 45.15 0.03 0.01 0.11 1.26 0.16 109.67 0.07 0.57 0.10 44.11 0.01 182.91 0.01 0.24 151.40 0.01 0.02 409.59 0.03 0.01 231.79 0.01 one meter dose rates 829 10 14.19 0.03 5.99 829 18 22.31 0.02 14.11 (18292829) 14.09 0.10 3.69 829 29 11.87 0.03 3.78 0.02 6.24 0.02 7.71 0.01 20.84 0.02 0.54 0.09 4.10 0.09 10.71 0.03 3.45 0.03 6.97 0.01 34.13 0.01 0.01 57.80 0.01 0.02 32.59 0.05 0.03 26.06 0.02 adjacent neutrons fuel gam co-60 total 817 94.13 19.65 69.13 182.91 825 19 62.18 89.21 0.01 151.40 (830 831) 292.87 7.05 109.67 409.59 832 184.14 3.54 44.11 231.79 one meter 829 10 14.19 12.23 7.71 34.13 829 18 22.31 34.95 0.54 57.80 (1829 2829) 14.09 7.79 10.71 32.59 829 29 11.87 7.22 6.97 26.06 Report: HI-2002563 Page B-10

47.5-15 mpc32 tra c-sum. xi s 47,500 MWD/MTU bumup - B&W 15x15 fuel element 3.6 w/o U235 15.0 YR 15.0 YR 15.0 YR 15.O YR u4n5aO2m u4nSaO2m BPRA curies 331.000 sidel25t u4p5aO2m u4q5aO2m u4c5aO2m num fuel 32 neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err adjacent dose rates surf seg 817 106.73 0.02 18.59 0.02 1.72 0.05 30.40 825 19 70.50 0.02 49.96 0.01 24.34 0.03 0.00 (830 831) 332.06 0.04 6.56 0.11 0.63 0.17 59.99 832 208.78 0.02 3.36 0.07 0.30 0.11 24.15 0.01 157.44 0.02 0.24 144.80 0.01 0.02 399.24 0.03 0.01 236.60 0.01 one meter dose rates 829 10 16.09 829 18 25.30 (1829 2829) 15.98 829 29 13.46 0.03 6.80 0.02 3.36 0.03 3.39 0.02 16.00 0.01 11.19 0.02 0.25 0.10 4.18 0.09 2.19 0.09 5.94 0.03 4.28 0.03 1.84 0.03 3.88 0.01 29.63 0.02 0.01 52.73 0.01 0.02 28.29 0.06 0.03 23.46 0.02 adjacent neutrons fuel gam co-60 total 817 106.73 20.31 30.40 157.44 825 19 70.50 74.30 0.00 144.80 (830 831) 332.06 7.19

.59.99 399.24 832 208.78 3.67 24.15 236.60 one meter 829 10 16.09 10.15 3.39 29.63 829 18 25.30 27.19 0.25 52.73 (1829 2829) 15.98 6.37 5.94 28.29 829 29 13.46 6.12 3.88 23.46 Report: HI-2002563 Page B-l I

Appendix C: Source Terms Table I shows the fuel characteristics that were used in generating the source terms [4]. Table 2 shows the characteristics of the BPRA [10]. The calculation of the BPRA source is described in reference [14].

Table 1: Fuel Assembly Parameters for Source Term Calculation Assembly type B&W 15x15 Active fuel length (in.)

144 No. of fuel rods 208 Rod Pitch (in.)

0.568 Cladding material Zircaloy-4 Rod diameter (in.)

0.428 Cladding thickness (in.)

0.0230 Pellet diameter (in.)

0.3742 Pellet material U0 2 Pellet density (gm/cc) 10.412 (95% of theoretical)

Enrichment (wt.% 235 4U) 2.9 and 4.0 Burnup (MWD/MTU) 32,500 and 55,000 Cooling time (years) 5 to 20 Specific power (MW/MTU) 40 Table 2: Physical Characteristics of BPRA Region l

Mass of material (kg)

Upper end fitting (steel) 72.62 Upper end fittingz (inconel) 0.42 Gas plenum spacer (steel) 0.77488 Gas plenum springs (steel) 0.67512 In-core (steel) 13.2 The enrichments that were chosen for this analysis were based on a review of the Diablo Canyon inventory provided in reference [5]. Table 3 shows the average enrichment and median enrichment for different burnup ranges based on the inventory.

Report: HI-2002563 Page: C-l

Table 3: Enrichments for Various Ranges for Fuel in Diablo Canyon Spent Fuel Pool Burnup Average Average Median Number of Range Burnup Enrichment Enrichment Assemblies GWD/MTU MWD/MTU wt% U235 wt0/o U235 15 to 20 17711.10 2.111 2.097 102 20 to 25 22320.33 2.230 2.123 36 25 to 30 27644.82 2.845 2.615 71 30 to 35 32702.12 3.015 3.091 172 35 to 40 37013.26 3.654 3.419 311 40 to 45 42656.28 4.102 4.007 358 45 to 50 46831.56 4.337 4.397 355 50 to 55 51210.40 4.186 4.396 70 55 to 60 55735.82 3.998 3.998 1

Pages C-3 through C-5 show the neutron and gamma source term for 32,500 MWD/MTU.

Pages C-6 through C-8 show the neutron and gamma source term for 55,000 MWD/MTU.

Pages C-9 and C-10 show the cobalt-60 source in the non-fuel hardware for a burnup of 32,500 MWD/MTU.

Pages C-l l and C-12 show the cobalt-60 source in the non-fuel hardware for a burnup of 55,000 MWD/MTU.

Page C-13 shows the cobalt-60 source for the BPP A at a burnup of 40,000 MWD/MTU.

Report: HI-2002563 Page: C-2

32,500 MWD/MTU burnup -

E 2.9 w/o U235 NEUTRON GROUP STRUCTURE 1

6.43E+00 - 2.00E+01 2

3.00E+00 - 6.43E+00 3

1.85E+00 - 3.00E+00 4

1.40E+00 - 1.85E+00 5

9.00E 1.40E+00 6

4.00E 9.OOE-01 7

1.00E 4.00E-01 PHOTON GROUP STRUCTURE 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

1.0000E-02 2.0000E-02 3.0000E-02 4.5000E-02 7.0000E-02 1.0000E-01 1.5000E-01 3.0000E-01 4.5000E-01

7. 0000E-01 1.0000E+00 1.5000E+00 2.0000E+00 2.5000E+00 3.0000E+00 4.0000E+00 6.OOOOE+00 8.0000E+00 TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO I&W 15x15 fuel element 2.0000E-02 3.0000E-02 4.5000E-02 7.0000E-02 1.0000E-01 1.5000E-01 3.0000E-01 4.5000E-01 7.0000E-01 1.0000E+00 1.5000E+00 2.0000E+00 2.5000E+00 3.0000E+00 4.0000E+00 6.0000E+00 8.0000E+00

1. loooE+n NEUTRON SPECTRUM 1

2 3

4 5

6 7

TOT 1

2 3

4 5

6 7

TOT 1

2 3

4 5

6 7

TOT 3.0 YR 3.373E+06 3.826E+07 4.227E+07 2.382E+07 3.225E+07 3.519E+07 6.889E+06

1. 821E+08 10.0 YR 2.574E+06 2.930E+07 3.265E+07 1.829E+07 2.468E+07 2.688E+07 5.260E+06 1.396E+08 17.0 YR 1.980E+06 2.271E+07 2.555E+07 1.419E+07 1.905E+07 2.069E+07 4.050E+06

1. 082E+08 4.0 YR 3.232E+06 3.661E+07 4.051E+07 2.284E+07 3.091E+07 3.372E+07 6.601E+06

1. 744E+08 11.0 YR 2.479E+06 2.825E+07 3.151E+07

1. 764E+07 2.378E+07 2.589E+07 5.067E+06 1.346E+08 18.0 YR

1. 908E+06 2.190E+07 2.468E+07 1.368E+07 1.836E+07 1.994E+07 3.902E+06

1. 044E+08 5.0 YR 3.108E+06 3.523E+07 3.902E+07 2.199E+07 2.975E+07 3.244E+07 6.350E+06 1.679E+08 12.0 YR 2.388E+06 2.723E+07 3.042E+07 1.701E+07 2.291E+07 2.494E+07 4.881E+06 1.298E+08 19.0 YR 1.838E+06 2.113E+07 2.384E+07 1.320E+07

1. 770E+07

1. 921E+07 3.760E+06

1. 007E+08 6.0 YR 2.993E+06 3.394E+07 3.764E+07 2.118E+07 2.865E+07 3.123E+07 6.114E+06 1.617E+08 13.0 YR 2.300E+06 2.626E+07 2.937E+07 1.640E+07 2.208E+07 2.402E+07 4.702E+06 1.251E+08 7.0 YR 2.882E+06 3.271E+07 3.632E+07 2.042E+07 2.760E+07 3.008E+07 5.888E+06

1. 559E+08 14.0 YR 2.215E+06 2.532E+07 2.836E+07 1.581E+07 2.128E+07 2.314E+07 4.529E+06 1.207E+08 8.0 YR 2.775E+06 3.153E+07 3.505E+07 1.968E+07 2.659E+07 2.897E+07 5.670E+06

1. 503E+08 15.0 YR 2.134E+06 2.441E+07 2.739E+07

1. 525E+07 2.051E+07 2.230E+07 4.363E+06 1.164E+08 9.0 YR 2.673E+06 3.039E+07 3.382E+07 1.897E+07 2.561E+07 2.790E+07 5.461E+06

1. 448E+08 16.0 YR 2.056E+06 2.354E+07 2.645E+07 1.471E+07

1. 976E+07 2.148E+07 4.203E+06 1.122E+08 20.0 YR 1.771E+06 2.038E+07 2.304E+07

1. 274E+07

1. 706E+07 1.851E+07 3.623E+06 9.712E+07 GAMMA SPECTRUM Report: Hl-2002563 Page C-3

19 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 1!

10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 3.0 YR 1.374E+15 8.132E+14 9.351E+14 6.621E+14 5.035E+14 6.170E+14 4.614E+14 2.512E+14 3.728E+15 1.037E+15 1.630E+14 1.354E+13 1.547E+13 3.422E+11 4.205E+10 7.848E+06 9.038E+05 1.040E+05 1.058E+16 10.0 YR 3.449E+14

1. 883E+14 2.522E+14 1.628E+14 1.029E+14 1.031E+14 8.945E+13 4.142E+13 1.674E+15 1.212E+14 3.476E+13 1.257E+12 5.192E+10 2.958E+09 3.708E+08 5.989E+06 6.897E+05 7.932E+04 3.117E+15 17.0 YR 2.844E+14 1.478E+14 2.024E+14 1.399E+14 8.237E+13 7.775E+13 7.158E+13 3.036E+13 1.320E+15 2.951E+13

1. 745E+13 7.269E+ll 3.242E+09 1.580E+08 1.378E+07 4.612E+06 5.310E+05 6.107E+04 2.404E+15 4.0 YR 8.482E+14 5.039E+14 5.856E+14 4.030E+14 2.956E+14 3.434E+14 2.692E+14 1.475E+14 3.018E+15 7.347E+14

1. 143E+14 7.279E+12 6.621E+12 1.721E+ll 2.128E+10 7.515E+06 8.655E+05 9.955E+04 7.277E+15 11.0 YR 3.331E+14

1. 798E+14 2.428E+14 1.582E+14 9.882E+13 9.812E+13 8.584E+13 3.986E+13 1.599E+15 9.400E+13 3.084E+13 1.14 0E+12 2.554E+10 1.560E+09 1.94 0E+08

5. 769E+06 6.643E+05 7.64 1E+04 2.961E+15 18.0 YR 2.777E+14 1.438E+14

1. 970E+14 1.374E+14 8.016E+13 7.509E+13 6.965E+13 2.942E+13 1.286E+15 2.580E+13 1.608E+13 6.788E+ll 3.043E+09 1.456E+08

1. 190E+07 4.444E+06 5.116E+05 5.884E+04 2.339E+15 5.0 YR 6.009E+14 3.549E+14 4.237E+14 2.820E+14 1.991E+14 2.202E+14 1.791E+14 9.721E+13 2.562E+15 5.282E+14 8.533E+13 4.294E+12 2.855E+12 8.670E+10 1.078E+10 7.228E+06 8.324E+05 9.575E+04 5.540E+15 12.0 YR 3.232E+14 1.728E+14 2.346E+14 1.544E+14 9.541E+13 9.389E+13 8.287E+13 3 684E+13 1.537E+15 7.412E+13 2.764E+13 1.046E+12 1.356E+10 8.549E+08 1.043E+08 5.557E+06 6.399E+05 7.360E+04 2.834E+15 19.0 YR 2.712E+14 1.400E+14 1.917E+14 1.349E+14 7.804E+13 7.256E+13 6.780E+13 2.855E+13 1.254E+15 2.287E+13 1.485E+13 6.344E+11 2.911E+09 1.388E+08 1.076E+07 4.282E+06 4.930E+05 5.669E+04 2.277E+15 6.0 YR 4.798E+14 2.795E+14 3.449E+14 2.238E+14

1. 527E+14 1.628E+14

1. 355E+14 7.166E+13 2.255E+15 3.844E+14 6.703E+13 2.827E+12 1.242E+12 4.374E+10 5.461E+09 6.960E+06 8.015E+05 9.219E+04 4.562E+15 13.0 YR 3.143E+14 1.668E+14 2.273E+14 1.511E+14 9.240E+13 9.014E+13 8.027E+13 3.518E+13 1.483E+15 5.948E+13 2.496E+13 9.674E+ll 8.049E+09 4.988E+08 5.870E+07 5.354E+06 6.164E+05 7.090E+04 2.726E+15 7.0 YR 4.172E+14 2.388E+14 3.036E+14 1.944E+14

1. 291E+14

1. 344E+14 1.134E+14 5.789E+13 2.041E+15 2.829E+14 5.478E+13 2.074E+12 5.462E+11 2.212E+10 2.771E+09 6.702E+06 7.719E+05 8.878E+04 3.970E+15 14.0 YR 3.062E+14 1.614E+14 2.205E+14 1.480E+14 8.966E+13 8.672E+13 7.790E+13 3.375E+13 1.436E+15 4.860E+13 2.268E+13 8.981E+ll 5.459E+09 3.190E+08 3.541E+07 5.157E+06 5.938E+05 6.830E+04 2.633E+15 9.0 YR 3.820E+14 2.149E+14 2.796E+14

1. 785E+14

1. 162E+14

1. 190E+14

1. 014E+14 4.991E+13

1. 885E+15 2.105E+14 4.611E+13

1. 663E+12 2.436E+ll

1. 122E+10 1.409E+09 6.455E+06 7.434E+05 9.550E+04 3.585E+15 15.0 YR 2.986E+14 1.565E+14 2.142E+14 1.452E+14 8.709E+13 9.354E+13 7.569E+13 3.250E+13 1.394E+15 4.045E+13 2.071E+13 8.358E+ll 4.209E+09 2.280E+08 2.342E+07 4.969E+06 5.721E+05 6.580E+04 2.550E+15 9.0 YR 3.602E+14

1. 994E+14 2.638E+14 1.691E+14 1.083E+14

1. 096E+14 9.426E+13 4.487E+13

1. 767E+15

1. 587E+14 3.970E+13 1.418E+12

1. 108E+ll

5. 729E+09 7.198E+08 6.218E+06 7.160E+05 8.236E+04 3.316E+15 16.0 YR 2.914E+14 1.52 0E+14 2.082E+14 1.425E+14 8.468E+13

8. 056E+13 7.359E+13

.4. 138E+13 1.356E+15 3.427E+13 1.898E+13 7.790E+ll 3.580E+09 1.818E+08

1. 715E+07 4.787E+06 5.512E+05 6.339E+04 2.474E+15 20.0 YR 2.649E+14 1.364E+14

1. 866E+14

1. 326E+14 7.600E+13 7.016E+13 6.601E+13 2.773E+13 1.224E+15 2.050E+13

1. 374E+13 5.932E+11 2.811E+09 1.349E+08 1.002E+07 4.126E+06 4.750E+05 5.463E+04 2.219E+15 Report: HI-2002563 P agre C -4

GAMMA SPECTRUM - MEV/SEC 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

3.0 YR 2.060E+13 2.033E+13 3.507E+13 3.807E+13 4.280E+13 7.712E+13 1.038E+14 9.421E+13 2.144E+15

8. 818E+14 2.038E+14 2.370E+13 3.481E+13 9.411E+ll 1.472E+ll 3.924E+07 6.327E+06 9.876E+05 3.721E+15 10.0 YR 5.174E+12 4.708E+12 9.457E+12

9. 3628+12 8.749R+12 1.289E+13 2.013E+13 1.553E+13 9.626E+14 1.030E+14 4.345E+13 2.200E+12 1.168E+ll 8.135E+09 1.298E+09 2.994E+07 4.828E+06 7.536E+05 1.197E+15 17.0 YR 4.266B+12 3.694E+12 7.591E+12 8.043E+12

7. 001E+12 9.718E+12 1.611E+13 1.139E+13 7.589E+14 2.508E+13 2.181E+13 1.272E+12 7.295E+09 4.346E+08 4.825E+07 2.306E+07 4.0 YR 1.272E+13 1.260E+13 2.196E+13 2.317E+13 2.513E+13 4.292E+13 6.056E+13 5.533E+13 1.735E+15 6.245E+14 1.429E+14

1. 274E+13 1.490E+13 4.734E+ll 7.450E+10 3.758E+07 6.058E+06 9.458E+05 2.785E+15 11.0 YR 4.996E+12 4.495E+12 9.104E+12 9.095E+12 8 400E+12 1 226E+13

1. 931E+13 1.457E+13 9.196E+14 7.990E+13 3.855E+13

1. 994E+12 5.746E+10 4.291E+09 6.790E+08 2.884E+07 4.650E+06 7.258E+05 1.122E+15 18.0 YR 4.165E+12 3.594E+12 7.386E+12 7.899E+12 6.814E+12 9.386E+12 1.567E+13 1.103E+13 7.395E+14 2.193E+13 2.010E+13 1.188E+12 6.847E+09 4.004E+08 4.164E+07 2.222E+07 5.0 YR 9.013E+12 8.873E+12 1.589E+13 1.622E+13 1.692E+13 2.753E+13 4.030E+13 3.645E+13 1.473E+15 4.490E+14 1.067E+14 7.514E+12 6.424E+12 2.384E+ll 3.772E+10 3.614E+07 5.827E+06 9.096E+05 2.214E+15 12.0 YR 4.847E+12 4.320E+12 8.799E+22 8.876E+12 8.110E+12 1.174E+13 1.865E+13 1.382E+13 8.837E+14 6.300E+13 3.455E+13 1.831E+12 3.052E+10 2.351E+09 3.651E+08 2.779E+07 4.479E+06 6.992E+05 1.062E+15 19.0 YR 4.068E+12 3.499E+12 7.189E+12 7.758E+12 6.634E+12 9.070E+12

1. 526E+13 1.071E+13 7.212E+14 1.944E+13 1.856E+13

1. 110E+12 6.549E+09 3.818E+08 3.766E+07 2.141E+07 6.0 YR 7.198E+12 6.989E+12 1.293E+13 1.287E+13 1.298E+13 2.035E+13 3.049E+13 2.687E+13 1.297E+15 3.267E+14 8.379E+13 4.947E+12 2.795E+12 1.203E+11 1.911E+10 3.480E+07 5.610E+06 8.758E+05 1.846E+15 13.0 YR 4.715E+12 4.169E+12 8.524E+12 8.686E+12 7.854E+12 1.127E+13 1.806E+13 1.319E+13 8.530E+14 5.055E+13 3.120E+13 1.693E+12 1.811E+10 1.372E+09 2.055E+08 2.677E+07 4.315E+06 6.735E+05 1.013E+15 7.0 YR 6.258E+12 5.970E+12 1.138E+13

1. 117B+13

1. 097E+13

1. 679E+13 2.552E+13 2.171E+13

1. 173E+15 2.404E+14 6.847E+13 3.630E+12 1.229E+12 6.082E+10 9.697E+09 3.351E+07 5.403E+06 8.434E+05 1.597E+15 14.0 YR 4.593E+12 4.035E+12 8.270E+12 8.512E412 7.621E+12 1.084E+13 1.753E+13 1.266E+13 8.260E+14 4.131E+13 2.835E+13 1.572E+12 1.228E+10 8.772E+08 1.239E+08 2.579E+07 4.157E+06 6.488E+05 9.713E+14 8.0 YR 5.731E+12 5.371E+12 1.048E+13

1. 026E+13 9.875E+12 1.487E+13 2.282E+13 1.871E+13 1.084E+15 1.789E+14 5.764E+13 2.911E+12 5.481E+ll 3.086E+10 4.931E+09 3.228E+07 5.204E+06 8.123E+05 1.422E+15 15.0 YR 4.479E+12 3.913E+12 8.032E+12 0. 348E+12

't.403E+12 1.044E+13 1.703E+13

1. 219E+13 8.017E+14 3.438E+13 2.589E+13 1.463E+12 9.470E+09 6.270E+08 8.197E+07 2.484E+07 4.004E+06 6.250E+05 9.353E+14 9.0 YR 5.403E+12 4.984E+12 9.893E+12 9.721E+12 9.206E+12

1. 370E+13 2.121E+13 1.683E+13 1.016E+15 1.349E+14 4.962E+13 2.482E+12 2.493E+11

1. 575E+10 2.519E+09 3.109E+07 5.012E+06 7.824E+05 1.294E+15 16.0 YR 4.370E+12 3.800E+12 7.806E+12 8.193E+12 7.197F+12

1. 007E+13

1. 656E+13

1. 177E+13 7.795E+14 2.913E+13 2.373E+13 1.363E+12 8.055E+09 4.998E+08 6.002E+07 2.394E+07 3.858E+06 6.022E+05 9.035E+14 20.0 YR 3.973E+12 3.409E+12 6.998E+12 7.622E+12 6.460E+12 8.770E+12 1.485E+13 1.040E+13 7.036E+14 1.742E+13 1.717E+13 1.038E+12 6.325E+09 3.709E+08 3.505E+07 2.063E+07 Report: HI-2002563 Page C-S

2 3.717E+06 3.581E+06 3.451E+06 3.325E+06 1

5.802E+05 5.590E+05 5.386E+05 5.190E+05 TOT 8.749E+14 8.487E+14 8.245E+14 8.017E+14 Report: HI-2002563 Page C-6

BW 15x15 4.0 wt% U235 55,000 MWD/MTU NEUTRON GROUP STRUCTURE 1

6.43E+00 -

2.00E+01 2

3.OOE+00 -

6.43E+00 3

1.85E+00 -

3.00E+00 4

1.40E+00 -

1.85E+00 5

9.00E 1.40E+00 6

4.00E 9.00E-01 7

1.00E 4.00E-01 PHOTON GROUP STRUCTURE 18 17 16 i5 14 13 12 11 10 9

8 7

6 5

4 3

2 1

1.0000E-02 2.0000E-02 3.0000E-02 4.5000E-02 7.0000E-02 1.0000E-01 1.5000E-01 3.0000E-01 4.5000E-01

7. OOOOE-01 1.0000E+00 1.5000E+00 2.0000E+00 2.5000E+00 3.0000E+00 4.0000E+00 6.0000E+00 8.0000E+00 TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO TO 2.0000E-02 3.0000E-02 4.5000E-02 7.0000E-02 1.0000E-01 1.5000E-01 3.0000E-01 4.5000E-01 7.0000E-01 1.0000E+00 1.5000E+00 2.0000E+00 2.5000E+00 3.0000E+00 4.0000E+00 6.0000E+00 8.0000E+00 1.1000E+01 NEUTRON SPECTRUM 1

2 3

4 5

6 7

TOT 1

2 3

4 5

6 7

TOT 1

2 3

4 5

6 7

TOT 3.0 YR 1.612E+07 1.816E+08

1. 993E+08 1.131E+08 1.537E+08 1.680E+08 3.290E+07 8.647E+08 10.0 YR 1.221E+07 1.378E+08 1.517E+08 8.596E+07 1.166B+08 1.274B+08 2.4941+07 6.5661+08 17.0 YR 9.371B+06

1. 060B+08 1.1721+08 6.6161+07 8.9611+07 9.778B+07

1. 914B+07 5.053B+08 4.0 YR

1. 544E+07 1.739B+08 1.9101+08 1.084B+08 1.4731+08 1.610E+08 3.152E+07 8.286E+08 11.0 YR 1.176E+07

1. 327E+08 1.462E+08 8.278E+07

1. 123E+08 1.226E+08 2.401E+07 6.323E+08 18.0 YR 9.026E+06 1.022E+08 1.130E+08 6.375E+07 8.632E+07 9.418E+07 1.844E+07 4.869E+08 5.0 YR 1.483E+07 1.670E+08 1.835E+08 1.042E+08 1.415E+08 1.547E+08 3.028E+07 7.960E+08 12.0 YR 1.132E+07

1. 278E+08 1.408E+08 7.972E+07

1. 081E+08 1.181E+08 2.311E+07 6.090E+08

19. O YR 8.694E+06 9.844E+07 1.089E+08 6.144E+07 8.316E+07 9.072E+07

1. 776E+07 4.691E+08 6.0 YR 1.426E+07

1. 606E+08 1.765E+08 1.002E+08 1.361E+08 1.487E+08 2.911E+07 7.655E+08 13.0 YR 1.090E+07 1.231E+08 1.357E+08 7.679E+07 1.041E+08 1.137E+08 2.225E+07 5.865E+08 7.0 YR 1.371E+07 1.545E+08 1.699E+08 9.639E+07 1.309E+08 1.430E+08 2.799E+07 7.364E+08 14.0 YR 1.049E+07 1.185E+08 1.308E+08 7.397E+07 1.003E+08

1. 095E+08 2.143E+07 5.650E+08 8.0 YR 1.319E+07 1.487E+08 1.636E+08 9.276E+07 1.259E+08 1.376E+08 2.693E+07 7.087E+08 15.0 YR

1. 010E+07

1. 142E+08 1.261E+08 7.126E+07 9.657E+07 1.054E+08 2.063E+07 5.443E+08 9.0 YR

1. 269E+07 1.431E+08

1. 575E+08 8.929E+07 1.212E+08

1. 324E+08 2.591E+07 6.821E+08 16.0 YR 9.730E+06 1.100E+08 1.215E+08 6.866E+07 9.302E+07

1. 015E+08 1.987E+07 5.243E+08 20.0 YR 8.374E+06 9.487E+07 1.050E+08

5. 921E+07 8.013E+07 8.739E+07 1.711E+07 4.521E+08 GAMMA SPECTRUM Report: HI-2002563 Page C-7

18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 S

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 3.0 YR 1.880E+15 1.101E+15 1.271E+15 8.9141+14 6.686E+14 8.028E+14 6.142E+14 3.336E+14 6.597E+15 2.126E+15 2.992E+14 1.888E+13 1.739E+13 4.652E+ll 5.762E+10 3.729E+07 4.295E+06 4.941E+05 1.662E+16 10.0 YR 5.693E+14 3.002E+14 4.160E+14 2.595E+14 1.661E+14 1.760E+14 1.446E+14 6.542E+13 2.832E+15 2.513E+14 7.121E+13 2.473E+12 6.439E+10 4.323E+09 5.544E+08 2.833E+07 3.263E+06 3.753E+05 5.255E+15 17.0 YR 4.695E+14 2.365E+14 3.328E+14 2.213E+14 1.326E+14 1.309E+14 1.156E+14 4.843E+13 2.195E+15 5.899E+13 3.544E+13 1.435E+12 5.150E+09 5.040E+08 5.485E+07 2.175E+07 2.505E+06 2.881E+05 3.979E+15 4.0 YR 1.231E+15 7.175E+14 8.518E+14 5.728E+14 4.158E+14 4.814E+14 3.777E+14 2.046E+14 5.340E+15 1.519E+15 2.185E+14

1. 076E+13 7.504E+12 2.344E+ll 2.919E+10 3.574E+07 4.117E+06 4.736E+05 1.195E+16 11.0 YR 5.503E+14 2.871E+14 4.005E+14 2.520E+14 1.596E+14 1.673E+14 1.389E+14 6.159E+1 2.693E+15 1.942E+14 6.316E+13 2.257E+12 3.270E+10 2.418E+09 3.107E+08 2.727E+07 3.141E+06 3.613E+05 4.970E+15 18.0 YR 4.583E+14 2.301E+14 3.235E+14 2.170E+14 1.290E+14 1.261E+14 1.125E+14 4.695E+13 2.138E+15 5.127E+13 3.261E+13 1.337E+12 4.868E+09 4.850E+08 5.093E+07 2.095E+07 2.413E+06 2.775E+05 3.866E+15 5.0 YR 9.187E+14 5.278E+14 6.516E+14 4.206E+14 2.954E+14 3.329E+14 2.646E+14 1.406E+14 4.508E+15

1. 098E+15 1.680E+14 6.820E+12 3.267E+12 1.183E+ll 1.481E+10 3.434E+07 3.956E+06 4.551E+05 9.336E+15 12.0 YR 5.341E+14 2.762E+14 3.869E+14 2.457E+14 1.540E+14 1.599E+14 1.342E+14 5.852E+13 2.578E+15 1.524E+14 5.654E+13 2.078E+12 1.812E+10 1.458E+09 1.862E+08 2.626E+07 3.024E+06 3.479E+05 4.739E+15 19.0 YR 4.474E+14 2.240E+14 3.147E+14 2.129E+14 1.256E+14

1. 216E+14 1.094E+14 4.556E+13 2.084E+15 4.517E+13

3. 007E+13 1.246E+12 4.672E+09 4.734E+08 4.810E+07 2.018E+07 2.324E+06 2.673E+05 3.761E+15 6.0 YR 7.608E+14 4.289E+14 5.499E+14 3.451E+14 2.357E+14 2.607E+14 2.084E+14

1. 072E+14 3.935E+15 8.007E+14 1.345E+14 4.828E+12 1.437E+12 5.988E+10 7.529E+09 3.303E+07 3.804E+06 4.376E+05 7.774E+15 13.0 YR 5.195E+14 2.667E+14 3.746E+14 2.402E+14 1.491E+14 1.531E+14 1.299E+14 5.596E*13 2.482E+15 1.217E+14 5.100E+13 1.922E+12 1.130E+10 9.724E+08 1.222E+08 2.528E+07 2.912E+06 3.349E+05 4.545E+15 7.0 YR 6.758E+14 3.734E+14 4.935E+14 3.055E+14 2.042E+14 2.230E+14

1. 790E+14 8.873E+13 3.529E+15 5.896E+14

1. 111E+14 3.762E+12 6.401E+ll 3.043E+10 3.845E+09 3.177E+07 3.660E+06 4.210E+05 6.777E+15 14.0 YR 5.060E+14 2.582E+14 3.632E+14 2.350E+14 1.446E+14 1.470E+14 1.260E+14 5.375E+13 2.398E+15 9.892E+13 4.628E+13 1.783E+12 8.042E+09 7.270E+08 8.876E+07 2.435E+07 2.804E+06 3.225E+05 4.379E+15 8.0 YR 6.258E+14 3.395E+14 4.587E+14

2. 832E+14 1.860E+14 2.011E+14

1. 624E+14 7.767E+13 3.231E+15 4.385E+14 9.411E+13 3.145E+12 2.897E+ll

1. 559E+10 1.979E+09 3.058E+07 3.522E+06 4.051E+05 6.102E+15 15.0 YR 4.932E+14 2.504E+14 3.525E+14 2.302E+14 1.404E+14

1. 412E+14 1.224E+14 5.180E+13 2.324E+15 8.184E+13 4.220E+13

1. 657E+12 6.438E+09 6.022E+08 7.085E+07 2.345E+07 2.700E+06 3.106E+05 4.232E+15 9.0 YR 5.931E+14 3.168E+14 4.346E+14 2.692E+14 1.744E+14 1.867E+14 1.519E+14 7.048E+13 3.007E+15 3.298E+14 8.125E+13 2.752E+12

1. 342E+ll 8.099E+09

1. 034E+09 2.943E+07 3.389E+06 3.899E+05 5.618E+15 16.0 YR 4.811E+14 2.432E+14 3.424E+14 2.257E+14 1.364E+14 1.359E+14 1.189E+14 5.004E+13 2.257E+15 6.891E+13 3.861E+13

1. 542E+12 5.610E+09 5.380E+08

6. 083E+07 2.258E+07 2.601E+06 2.991E+05 4.100E+15 20.0 YR 4.369E+14 2.182E+14 3.062E+14 2.090E+14 1.223E+14

1. 173E+14

1. 065E+14 4.426E+13 2.032E+15 4.026E+13 2.776E+13 1.163E+12 4.520E+09 4.654E+08 4.585E+07 1.944E+07 2.239E+06 2.575E+05 3.662E+15 Report: 11-2002563 Page C-8

GAMMA SPECTRUM - MEV/SEC 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

TOT 18 17 16 15 14 13 12 11 10 9

8 7

6 5

4 3

3.0 YR 2.819E+13 2.752E+13 4.765E+13

5. 125E+13 5.684E+13 1.003E+14 1.382E+14 1.251E+14 3.793E+15

1. 807E+15 3.739E+14 3.304E+13 3.912E+13

1. 279E+12 2.017E+11

1. 864E+08 3.007E+07 4.694E+06 6.623E+15 10.0 YR 8.540E+12 7.505E+12

1. 560E+13 1.492E+1sj 1.412E+13 2.200E+13 3.254E+13 2.453E+13 1.629E+15 2.136E+14 8.901E+13 4.328E+12 1.449E+ll 1.189E+10 1.940E+09 1.416E+08 2.284E+07 3.565E+06 2.075E+15 17.0 YR 7.042E+12 5.912E+12 1.248E+13 1.272E+13

1. 127E+13 1.636E+13 2.602E+13

1. 816E+13 1.262E+15 5.014E+13 4.431E+13 2.512E+12 1.159E+10 1.386E+09

1. 920E+08

1. 087E+08 4.0 YR 1.847E+13 1.794E+13 3.194E+13 3.294E+13 3.534E+13 6.018E+13 8.499E+13 7.672E+13 3.071E+15 1.291E+15 2.731E+14 1.883E+13 1.689E+13 6.446E+ll 1.022E+ll 1.787E+08 2.882E+07 4.499E+06 5.030E+15 11.0 YR 8.255E+12 7.178E+12 1.502E+13

1. 449E+13

1. 356E+13 2.092E+13 3.126E+13 2.310E+13 1.548E+15 1.651E+14 7.894E+13 3.950E+12 7.358E+10 6.651E+09 1.089E+09 1.364E+08 2.199E+07 3.432E+06 1.930E+15 18.0 YR 6.874E+12 5.752E+12 1.213E+13 1.248E+13 1.097E+13

1. 576E+13 2.530E+13 1.760E+13 1.229E+15 4.358E+13 4.077E+13 2.340E+12 1.095E+10 1.334E+09 1.783E+08 1.047E+08 5.0 YR

1. 378E+13

1. 320E+13 2.443E+13 2.418E+13 2.511E+13 4.161E+13

5. 953E+13 5.271E+13 2.592E+15 9.331E+14 2.100E+14 1.194E+13 7.352E+12 3.254E+ll 5.183E+10 1.717E+08 2.769E+07 4.323E+06 4.009E+15 12.0 YR 8.011E+12 6.904E+12 1.451E+13 1.413E+13 1.309E+13 1.998E+13 3.019E+13 2.195E+13 1.483E+15 1.296E+14 7.068E+13 3.636E+12 4.077E+10 4.008E+09 6.518E+08 1.313E+08 2.117E+07 3.305E+06 1.815E+15 19.0 YR 6.711E+12 5.600E+12 1.180E+13 1.224E+13 1.067E+13 1.520E+13 2.462E+13 1.709E+13 1.198E+15 3.839E+13 3.758E+13 2.181E+12 1.051E+10 1.302E+09 1.684E+08 1.009E+08 6.0 YR 1.1413+13 1.072E+13 2.062E+13 1.984E+13 2.004E+13 3.259E+13 4.690E+13 4.021E+13 2.263E+15 6.806E+14 1.681E+14 8.449E+12 3.234E+12 1.647E+11 2.635E+10

1. 651E+08 2.663E+07 4.157E+06 3.326E+15 13.0 YR 7.792E+12 6.667E+12 1.40.5E+13 L.381E+13 t.267E313

.1.914E+13 2.923E+13 2.099E+13 1.427E+15 1.034E+14 6.376E+13 3.364E+12 2.542E+10 2.674E+09 4.277E+08 1.264E+08 2.038E+07 3.182E+06

1. 722E+15 7.0 YR

1. 014E+13 9.335E+12

1. 851E+13

1. 757E+13 1.736E+13 2.788E+13 4.028E+13 3.327E+13 2.029E+15 5.011E+14 1.389E+14 6.584E+12 1.440E+12 8.369E+10 1.346E+10 1.589E+08 2.562E+07 3.999E+06 2.852E+15 14.0 YR 7.589E+12 6.454E+12 1.362E+13 1.351E+13 1.229E+13 1.837E+13 2.836E+13 2.016E+13 1.379E+15 8.408E+13 5.785E+13 3.120E+12 1.809E+10

1. 999E+09 3.107E+08 1.217E+08 1.963E+07 3.064E+06

1. 644E+15 8.0 YR 9.386E+12 8.488E+12 1.720E+13 1.628E+13

1. 581E+13 2.514E+13 3.653E+13 2.912E+13

1. 858E+15 3.727E+14 1.176E+14 5.504E+12 6.518E+ll 4.286E+10 6.927E+09

1. 529E+08 2.465E+07 3.849E+06 2.513E+15 15.0 YR 7.399E+12 6.260E+12 1.322E+13 1.324E+13 1.193E+13 1.766E+13 2.754E+13 1.942E+13

1. 336E+15 6.956E+13

5. 274E+13 2.900E+12 1.449E+10 1.656E+09 2.480E+08 1.172E+08 1.890E+07 2.951E+06

1. 578E+15 9.0 YR 9.897E+12 7.920E+12 1.630E+13 1.548E+13 1.482E+13 2.333E+13 3.418E+13 2.643E+13 1.729E+15 2.804E+14 1.016E+14 4.815E+12 3.019E+ll 2.227E+10 3.619E+09 1.471E+08 2.373E+07 3.704E+06 2.263E+15 16.0 YR 7.217E+12 6.080E+12

1. 284E+13

1. 298E+13 1.160E+13

1. 699R+13 2.676E+13 1.876E+13 1.298E+15

5. 857E+13 4.827E+13 2.698E+12

1. 262E+10 1.479E+09 2.129E+08 1.129E+08 1.820E+07 2.842E+06

1. 520E+15 20.0 YR 6.554E+12 5.455E+12 1.148E+13 1.202E+13 1.039E+13 1.467E+13 2.395E+13 1.660E+13 1.168E+15 3.422E+13 3.470E+13 2.035E+12

1. 017E+10 1.280E+09 1.605E+08 9.719E+07 Report: HI-2002563 Page C-9

2 1.753E+07 1.689E+07 1.627E+07 1.567E+07 1

2.737E+06 2.636E+06 2.539E+06 2.446E+06 TOT 1.469E+15 1.423E+15 1.380E+15 1.341E+15 Report: 1H1-2002563 Page C-10O

32,500 MWD/MTU burnup - B&W 15x1 5 fuel element 2.9 wlo U235 - I gm Co59 initially CURIES 3.0 YR 4.0 YR 5.0 YR 6.0 YR 7.0 YR 8.0 YR 9.0 YR CO 60 9.57E+01 8.39E+01 7.36E+01 6.45E+01 5.65E+01 4.96E+01 4.35E+01 I0.OYR ILOYR 12.OYR 13.0 YR 14.OYR l5.0YR 16.OYR CO 60 3.81E+01 3.34E+01 2.93E+01 2.57E+01 2.25E+01 1.97E+01 1.73E+01 17.0 YR 18.0 YR 19.0 YR 20.0 YR CO 60 1.52E+01 1.33E+01 1.17E+01 1.02E+01 Calculated Gamma Sources from Cobalt Mass of Cobalt in source term calc (gm)

Additional fuel source in group 8 'l.0-1.5 MeV) curies Mass (kg) 4.9 Frac tot I

impurity level (gm.kg) 3.0 YR 4.0 YR 5.0 YR 6.0 YR 7.0 YR 8.0 YR 9.0 YR 468.93 411.11 360.64 316.05 276.85 243.04 213.15 10.0 YR 11.0 YR 12.0 YR 13.0 YR 14.0 YR 15.0 YR 16.0 YR 186.69 163.66 143.57 125.93 110.25 96.53 84.77 17.0 YR 18.0 YR 19.0 YR 20.0 YR 74.48 65.17 57.33 49.98 Bottom Nozzle - curies Mass (kg) 9.46 Frac tot 0.2 impurity level (gm.kg) 3.0 YR 4.0 YR 5.0 YR 6.0 YR 7.0 YR 8.0 YR 9.0 YR 181.06 158.74 139.25 122.03 106.90 93.84 82.30 10.0 YR 11.0 YR 12.0 YR 13.0 YR 14.0 YR 15.0 YR 16.0 YR 72.09 63.19 55.44 48.62 42.57 37.27 32.73 17.0 YR 18.0 YR 19.0 YR 20.0 YR 28.76 25.16 22.14 19.30 Report: HI-2002563 Page C-I11

Plenum Springs-curies Mass (kg) 0.72176 Frac tot 3.0 YR 4.0 YR 5.0 YR 13.81 12.11 10.62 I0.OYR I1I.OYR 12.0YR 5.50 4.82 4.23 17.0 YR 18.0 YR 19.0 YR 2.19 1.92 1.69 Plenum Spacer - curies Mass (kg) 0.82824 Frac tot 3.0 YR 4.0 YR 5.0 YR 7.93 6.95 6.10 10.OYR II.OYR 12.0YR 3.16 2.77 2.43 17.0YR 18.0YR 19.0YR

!.26 1.10 0.97 Top End Fitting - curies Mass (kg) 9.28 Frac tot 3.0 YR 4.0 YR 5.0 YR 88.81 77.86 68.30 10.OYR 1l.OYR 12.0YR 35.36 31.00 27.19 17.0YR 18.0YR 19.0YR 14.11 12.34 10.86 0.2 impurity level (gm.kg) 6.0 YR 7.0 YR 8.0 YR 9.0 YR 9.31 8.16 7.16 6.28 13.0YR 14.0YR 15.0YR 16.0YR 3.71 3.25 2.84 2.50 20.0 YR 1.47 0.1 impurity level (gm.kg) 6.0YR 7.0YR 8.0YR 9.0YR 5.34 4.68 4.11 3.60 13.0YR 14.0YR 15.0YR 16.0YR 2.13 1.86 1.63 1.43 20.0 YR 0.84 0.1 impurity level (gm.kg) 6.0 YR 7.0 YR 8.0 YR 9.0 YR 59.86 52.43 46.03 40.37 13.0YR 14.0YR 15.0YR 16.0YR 23.85 20.88 18.28 16.05 20.0 YR

-9.47 Report: 11-2002563 Page C-12

BW 15xl5 4.0 wt%/o U235-1.0 gm Co59 originally 55,000 MWD/MTU CURIES 3.0 YR 4.0 YR 5.0 YR 6.0 YR 7.0 YR 8.0 YR 9.0 YR CO 60 1.30E+02 1.14E+02 9.96E+01 8.73E+01 7.66E+01 6.71E+01 5.89E+01 10.OYR ILOYR 12.OYR 13.OYR 14.OYR IS.OYR 16.OYR CO 60 5.16E+01 4.52E+01 3.97E+01 3.48E+01 3.05E+01 2.67E+01 2.34E+01 17.0 YR 18.0 YR 19.0 YR 20.0 YR CO 60 2.05E+01 1.80E+01 1.58E+01 1.38E+01 Calculated Gamma Sources from Cobalt Mass of Cobalt in source term calc (gm)

Additional fuel source in group 8 (1.0-1.5 MeV) curies I

Mass (kg) 4.9 Frac tot 1

impurity level (gm.kg)

I 3.0 YR 637.00 10.0 YR 252.84 17.0 YR 100.45 Bottom Nozzle - curies Mass (kg) 9.46 3.0 YR 245.96 10.0 YR 97.63 17.0 YR 38.79 4.0 YR 558.60 11.0YR 221.48 18.0 YR 88.20 4.0 YR 215.69 11.0YR 85.52 18.0 YR 34.06 5.0 YR 488.04 12.0 YR 194.53 19.0 YR 77.42 Frac tot 5.0 YR 188.44 12.0 YR 75.11 19.0 YR 29.89 6.0 YR 427.77 13.0 YR 170.52 20.0 YR 67.62 0.2 6.0 YR 165.17 13.0 YR 65.84 20.0 YR 26.11 7.0 YR 375.34 14.0 YR 149.45 7.0 YR 144.93 14.0 YR 57.71 8.0 YR 328.79 15.0 YR 130.83 9.0 YR 288.61 16.0 YR 114.66 impurity level (gm.kg) 8.0 YR 9.0 YR 126.95 111.44 15.0 YR 16.0 YR 50.52 44.27 I

Report: H[-2002563 Page C-13

Plenum Springs - curies Mass (kg) 0.72176 Frac tot 0.2 impurity level (gm.kg)

I 3.0 YR 4.0 YR 5.0 YR 18.77 16.46 14.38 10.OYR I1.OYR 12.0 YR 7.45 6.52 5.73 17.0 YR 18.0 YR 19.0 YR 2.96 2.60 2.28 6.0 YR 7.0 YR 8.0 YR 9.0 YR 12.60 11.06 9.69 8.50 13.0 YR 14.0 YR 15.0 YR 16.0 YR 5.02 4.40 3.85 3.38 20.0 YR 1.99 Plenum Spacer - curies Mass (kg) 0.82824 Frac tot 0.1 impurity level (gm.kg)

I 3.0 YR 4.0 YR 5.0 YR 10.77 9.44 8.25 I0.OYR I1.OYR 12.0YR 4.27 3.74 3.29 17.0 YR 18.0 YR 19.0 YR 1.70 1.49 1.31 6.0 YR 7.0 YR 8.0 YR 9.0 YR 7.23 6.34 5.56 4.88 13.0 YR 14.0 YR 15.0 YR 16.0 YR 2.88 2.53 2.21 1.94 20.0 YR 1.14 Top End Fitting - curies Mass (kg) 9.28 Frac tot 0.1 impurity level (gm.kg)

I 3.0 YR 4.0 YR 5.0 YR 120.64 105.79 92.43 10.OYR I1.OYR 12.0 YR 47.88 41.95 36.84 17.0 YR 18.0 YR 19.0YR 19.02 16.70 14.66 6.0 YR 7.0 YR 8.0 YR 9.0 YR 81.01 71.08 62.27 54.66 13.0 YR 14.0YR 15.0 YR 16.0 YR 32.29 28.30 24.78 21.72 20.0 YR 12.81 Report: 11-2002563 Page C-14

40,000 MWD/MTU burnup - B&W 15xl5 fuel element flux 3.4 w/o U235 - Hybrid of W 17x17 and W 15x15 BPRA CURIES TOTAL 3.0 YR CO 60 1.23E+03 10.0 YR CO 60 4.91E+02 17.0 YR CO 60 1.96E+02 CURIES 3.0 YR CO 60 4.50E+01 10.0 YR CO 60 1.79E+01 17.0 YR CO 60 7.16E+00 4.0 YR 1.08E+03 11.0 YR 4.31E+02 18.0 YR 1.71E+02 TOP ZONE 4.0 YR 3.95E+01 11.0 YR 1.5BE+01 18.0 YR 6.25E+00 5.0 YR 9.48E+02 12.0 YR 3.78E+02 19.0 YR 1.50E+02 Fraction 5.0 YR 3.46E+01 12.0 YR 1.38E+01 19.0 YR 5.48E+00 Fraction 5.0 YR 5.28E+00 12.0 YR 2.10E+00 19.0 YR 8.35E-01 6.0 YR 8.31E+02 13.0 YR 3.31E+02 20.0 YR 1.32E+02 of total:

6.0 YR 3.04E+01 13.0 YR 1.21E+01 20.0 YR 4.82E+00 of to::al:

6.0 YR 4.63E+00 13.0 YR 1.84E+00 20.0 YR 7.35E-01 7.0 YR 7.29E+02 14.0 YR 2.90E+02 0.036545 7.0 YR 2.66E+01 14.0 YR 1.06E+01

0. '05566 7.0 YR 4.06E+00 14.0 YR 1.61E+00 8.0 YR 6.39E+02 15.0 YR 2.54E+02 8.0 YR 2.34E+01 15.0 YR 9.28E+00 8.0 YR 3.56E+00 15.0 YR 1.41E+00 9.0 YR 5.60E+02 16.0 YR 2.23E+02 9.0 YR 2.05E+01 16.0 YR 8.15E+00 9.0 YR 3.12E+00 16.0 YR 1.24E+00 CURIES PLENUM SPACER 3.0 YR 4.0 YR CO 60 6.85E+00 6.01E+00 10.0 YR 11.0 YR CO 60 2.73E+00 2.40E+00 17.0 YR 18.0 YR CO 60 1.09E+00 9.52E-01 CURIES PLENUM SPRING 3.0 YR 4.0 YR CO 60 1.19E+01 1.05E+01 10.0 YR 11.0 YR CO 60 4.76E+00 4.18E+00 17.0 YR 18.0 YR CO 60 1.90E+00 1.66E+00 Fraction of total:

5.0 YR 6.0 YR 9.19E+00 8.06E+00 12.0 YR 13.0 YR 3.67E+00 3.21E+00 19.0 YR 20.0 YR 1.45E+00 1.28E+00 0.009699 7.0 YR 7.07E+00 14.0 YR 2.81E+00 8.0 YR 6.20E+00 15.0 YR 2.46E+00 9.0 YR 5.43E+00 16.0 YR 2.16E+00 CURIES IN CORE 3.0 YR 4.0 YR CO 60 1.17E+03 1.02E+03 10.0 YR 11.0 YR CO 60 4.65E+02 4.09E+02 Fraction 5.0 YR 8.99E+02 12.0 YR 3.58E+02 of total:

6.0 YR

7. 88E+02 13.0 YR 3.14E+02 0.948000 7.0 YR 8.0 YR 9.0 YR 6.91E+02 6.06E+02 5.31E+02 14.0 YR 15.0 YR 16.0 YR 2.75E+02 2.41E+02 2.11E+02 Report: HI-2002563

-Pa-ge C-1 5

17.0 YR 18.0 YR 19.0 YR 20.0 YR CO 60 1.86E+02 1.62E+02 1.42E+02 1.25E+02 Report: HI-2002563 Page C-16

Appendix D: Near Dose Rates for rn-STORM Overpack This appendix presents the dose rates calculated at the surface and 1 meter from the HI-STORM overpack. This output, which is at the end of this section, is for a burnup and cooling time of 32,500 MWD/MTU and 5-year cooling and is presented in the form of tables printed from EXCEL. A more detailed discussion of the calculations is provided in Section 7.

Some pages have labels "sur" on the far left. These refer to MCNP surfaces. A description of the MCNP surfaces and segmentation are provided before the results.

A brief description of the EXCEL output.

The first two lines are title lines indicating the bumup, enrichment and decay time. The next few lines list the runs from which the data is taken. After the runs, the next line describes which dose component each column represents. The phot(n,p) means photons from neutron interactions in the surrounding material. Phot refers to photons coming from decay of fission products in the fuel region and from decay of cobalt-60 in the in-core grid spacers. Cobalt refers to the cobalt source in the end fittings. The line after the dose components titles the columns value or rel err.

The value means dose and the rel err refers to relative error which is defined as the standard deviation over the mean. In order to calculate the standard deviation of a value listed in the value column, one would take the associated rel err number and multiply by the number in the value column. The numbers listed under the total column are total doses. The rel err of the total values are statistically calculated from the standard deviations of the components.

Dose rates from sections of the BPRA located in the active region of the fuel are included in the columns labeled Phot. Dose rates from sections of the BPRA located above the active region are included in the cobalt columns.

Report: HI-2002563 Page: D-1I

Tally locations for Short HI-STORM shrtstorm-locations.xls Axial Tally Segments start finish Segment MCNP pos cm cm 1

-33.02

-15.24 2

-15.24 15.24 3

15.24 45.72 4

45.72 76.2 5

76.2 106.68 6

106.68 137.16 7

137.16 167.64 8

167.64 198.12 9

198.12 228.6 10 228.6 259.08 11 259.08 289.56 12 289.56 320.04 13 320.04 350.52 14 350.52 381 15 381 411.48 16 411.48 441.96 17 441.96 472.44 18 472.44 502.92 19 502.92 533.4 20 533.4 563.88 21 563.88 594.36 22 594.36 624.84 23 624.84 655.32 24 655.32 761.975 start finish pos rel to bottom feet feet 0.00 0.58 bottom of overpack -33.(

0.58 1.58 1.58 2.58 2.58 3.58 3.58 4.58 4.58 5.58 5.58 6.58 6.58 7.58 7.58 8.58 c---

center of bwr fuel 8.58 9.58 <--

centerofpwr fuel 9.58 10.58 10.58 11.58 11.58 12.58 12.58 13.58 13.58 14.58 14.58 15.58 15.58 16.58 16.58 17.58 17.58 18.58 18.58 19.58 <----

top of overpack 553.72 19.58 20.58 20.58 21.58 21.58 22.58 22.58 26.08 223.227 cm 1236.16 cm Radial tally surfaces using segmentation above surface radius distance from edge meters feet edge overpack 720 728 731 724 168.275 168.775 0.005 0.016404 199.255 0.3098 1.016404 229.735 0.6146 2.016404 268.525 1.0025 3.289042 Report: HI-2002563 Page D-2

Tally locations for Short HI-STORM shrtstorm-locations.xls Tally surfaces for ducts - area is same as area of duct surface radius distance from edge meters feet edge overpack 718 1716 726 729 722 719 727 730 723 168.275 168.525 0.0025 0.01 bottom duct 15 inch wide by 10 inch tall 168.175

-0.001 0.00 bottom duct 3.08 in tall by 4.5 in wide in center of duct and DPA 199.005 0.3073 1.01 bottom duct 15 inch wide by 10 inch tall 229.485 0.6121 2.01 bottom duct 15 inch wide by 10 inch tall 268.275 1

3.28 bottom duct 15 inch wide by 10 inch tall 168.525 0.0025 0.01 top duct 6 in tall by 25 inches wide 199.005 0.3073 1.01 top duct 6 in tall by 25 inches wide 229.485 0.6121 2.01 top duct 6 in tall by 25 inches wide 268.275 1

3.28 top duct 6 in tall by 25 inches wide Tally surfaces on top of overpack surface top of overpack 918 919 distance from edge axial loc meters feet 553.72 554.72 653.72 0.01 0.03 3.28 Radial segments on top of overpack start finish width width start finish Segment cm in pos rel to center cm cm feet feet I

15.24 15.24 6.00 0.00 0.50 2

15.24 45.72 30.48 12.00 0.50 1.50 3

45.72 76.2 30.48 12.00 1.50 2.50 4

76.2 86.83625 10.64 4.19 2.50 2.85 5 86.83625 93.3451 6.51 2.56 2.85 3.06 <-

Annulus between 6

93.3451 106.68 13.33 5.25 3.06 3.50 MPC and overpack 7

106.68 137.16 30.48 12.00 3.50 4.50 is between 8

137.16 167.64 30.48 12.00 4.50 5.50 86.83625 and 93.345 cm 9

167.64 168.275 0.64 0.25 5.50 5.52 Cell tallys on lid directly above ducts Cells are 25 inches wide in the following locations ID (cm)

OD (cm) width (cm width (in) 109.22 124.46 15.24 6

109.22 is the OD of the concrete on HI-STORM lid 124.46 139.7 15.24 6

139.7 154.94 15.24 6

Report: HI-2002563 Page D-3

output h24shrt2.xls 32,500 MWD/MTU burnup - B&W 15xl5 fuel element 2.9 w/o U235 BPRA Curies 331.00 5.0 YR 5.0 YR 5.0 YR 5.0 YR dwg3443 TP Curies 0.00 h4n5dllm h4n5dllm h4p5dllmh4q5dllm h4c5dlln Numofassemblies 32 h4rfdl Im h4p5eO8m h4q5eO7m h4r5eO7m neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err sur 720 1

0.611 2

0.316 3

0.163 4

0.330 5

0.557 6

0.778 7

0.756 8

0.859 9

0.857 10 0.848 11 0.770 12 0.775 13 0.593 14 0.487 15 0.309 16 0.143 17 0.063 18 0.104 19 0.634 20 0.825 21 0.844 22 0.617 23 0.445 24 0.258 sur 724.000 1

0.164 2

0.186 3

0.201 4

0.215 5

0.250 6

0.294 7

0.320 8

0.346 9

0.331 10 0.345 11 0.339 12 0.304 13 0.267 14 0.218 15 0.174 16 0.136 17 0.119 18 0.160 0.033 0.033 0.074 0.067 0.052 0.045 0.043 0.043 0.044 0.043 0.041 0.045 0.044 0.051 0.063 0.071 0.069 0.032 0.020 0.016 0.018 0.021 0.024 0.017 0.046 0.034 0.036 0.036 0.034 0.032 0.031 0.031 0.029 0.030 0.030 0.031 0.032 0.035 0.037 0.042 0.040 0.044 0.091 0.129 0.222 0.382 0.555 0.741 0.830 0.889 0.901 0.876 0.826 0.760 0.637 0.503 0.338 0.226 0.127 0.065 0.055 0.048 0.057 0.067 0.065 0.054 0.107 0.122 0.167 0.219 0.260 0.329 0.357 0.399 0.394 0.391 0.371 0.354 0.307 0.248 0.206 0.160 0.112 0.083 0.035 0.036 0.029 0.026 0.022 0.020 0.019 0.017 0.018 0.018 0.019 0.018 0.019 0.022 0.025 0.034 0.035 0.032 0.035 0.019 0.019 0.019 0.017 0.012 0.055 0.035 0.028 0.025 0.024 0.023 0.021 0.021 0.020 0.020 0.020 0.022 0.024 0.024 0.030 0.027 0.034 0.042 1.541 1.675 5.550 15.108 26.534 31.093 33.092 32.598 32.253 32.022 31.530 31.560 29.673 26.745 18.134 8.414 2.661 0.640 1.705 0.156 0.201 0.254 0.203 0.144 2.766 4.009 6.083 8.914 11.993 14.362 15.866 16.637 16.837 16.896 16.399 15.654 14.341 12.191 9.510 6.576 4.223 2.712 0.063 0.022 0.011 0.011 0.011 0.012 0.013 0.013 0.013 0.013 0.013 0.013 0.013 0.012 0.011 0.011 0.11I 0.020 0.052 0.047 0.049 0.081 0.071 0.067 0.018 0.012 0.010 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.010 0.011 0.013 4.205 6.355 12.287 10.904 3.518 0.558 0.110 0.057 0.034 0.013 0.022 0.032 0.117 1.043 4.772 9.370 9.085 4.237 4.346 0.648 0.809 0.872 0.750 0.423 2.616 3.169 3.502 3.174 2.559 1.620 0.854 0.571 0.339 0.297 0.374 0.481 0.946 1.611 2.144 2.597 2.779 2.511 0.010 0.012 0.021 0.030 0.055 0.128 0.405 0.455 0.410 0.495 0.595 0.381 0.279 0.118 0.052 0.027 0.020 0.017 0.009 0.040 0.017 0.016 0.017 0.017 0.023 0.020 0.024 0.031 0.039 0.047 0.063 0.113 0.100 0.112 0.109 0.093 0.073 0.053 0.045 0.033 0.025 0.021 6.448 8.476 18.221 26.724 31.163 33.170 34.788 34.402 34.045 33.759 33.148 33.126 31.019 28.778 23.553 18.154 11.937?

5.045 6.741 1.677 1.912 1.809 1.463 0.878 5.654 7.486 9.953 12.522 15.062 16.605 17.397 17.954 17.901 17.929 17.482 16.793 15.861 14.268 12.035 9.469 7.233 5.466 0.017 0.010 0.014 0.014 0.011 0.012 0.013 0.012 0.013 0.013 0.013 0.013 0.013 0.012 0.013 0.015 0.016 0.015 0.015 0.018 0.012 0.015 0.015 0.014 0.014 0.011 0.010 0.010 0.010 0.009 0.009 0.009 0.008 0.008 0.009 0.009 0.009 0.010 0.011 0.011 0.011 0.012 File: Hl-2002563 Page D-4

output h24shrt2.xIs 32,500 MWD/MTU bumup - B&W 15xlS fuel element 2.9 w/o U235 BPRA Curies 331.00 5.0 YR 5.0 YR 5.0 YR 5.0 YR dwg3443 TP Curies 0.00 h4n5dl Im h4n5dl lm h4p5dl Im h4q5dllim h4cSdl In Num of assemblies 32 h4r5dlI m h4p5eO8m h4q5eO7m h4rfeO7m neutron phot (n,p) phot cobalt total value rel err value rel err value rel err value rel err value rel err 19 0.196 20 0.186 21 0.227 22 0.239 23 0.236 24 0.194 tal 112.000 sur 918.000 1

1.320 2

1.213 3

1.100 4

1.029 5

0.907 6

0.663 7

1.819 8

1.412 9

0.886 sur*

919.000 1

0.477 2

0.459 3

0.436 4

0.409 5

0.424 6

0.415 7

0.401 8

0.381 9

0.397 tal 122.000 sur 718.000 1

2.902 sur 722.000 1

0.338 sur 719.000 1

2.546 sur 723.000 1

0.372 tal 132.000 sur 728.000 1

0.298 2

0.250 3

0.189 4

0.284 5

0.411 6

0.486 0.030 0.022 0.023 0.020 0.022 0.017 0.031 0.014 0.012 0.019 0.023 0.025 0.017 0.017 0.092 0.041 0.017 0.016 0.021 0.031 0.021 0.015 0.017 0.074 0.033 0.075 0.060 0.046 0.035 0.031 0.032 0.028 0.049 0.045 0.037 0.041 0.035 0.015 2.077 0.908 0.585 0.401 0.288 0.179 0.026 0.020 0.028 0.036 0.048 0.035 2.305 1.137 0.744 0.506 0.353 0.270 0.017 0.020 0.020 0.033 0.033 0.019 4.639 2.278 1.591 1.178 0.909 0.672 0.014 0.013 0.014 0.019 0.021 0.013 0.626 0.626 0.543 0.433 0.365 0.265 0.143 0.075 0.054 0.190 0.193 0.168 0.149 0.136 0.124 0.105 0.078 0.073 0.028 0.014 0.011 0.016 0.024 0.024 0.015 0.021 0.176 0.040 0.021 0.016 0.022 0.027 0.025 0.021 0.015 0.088 0.044 0.046 0.173 0.786 1.193 1.027 0.567 0.242

.0.128 0.184 0.195 0.228 0.249 0.251 0.264 0.231 0.172 0.181 0.055 0.030 0.037 0.046 0.054 0.058 0.036 0.035 0.068 0.040 0.041 0.043 0.047 0.051 0.053 0.045 0.033 0.054 0.307 0.262 0.360 0.964 1.467 1.771 2.552 1.130 0.657 0.350 0.361 0.402 0.441 0.465 0.514 0.600 0.614 0.783 0.131 0.061 0.031 0.029 0.035 0.043 0.017 0.016 O.056 0.064 0.033 0.028 0.031 0.031 0.028 0.023 0.016 0.055 2.298 2.147 2.176 3.212 3.931 3.726 5.081 2.859 1.724 1.201 1.208 1.234 1.247 1.277 1.317 1.337 1.244 1.433 0.026 0.012 0.009 0.016 0.022 0.026 0.011 0.011 0.052 0.026 0.014 0.014 0.016 0.018 0.017 0.014 0.010 0.037 0.278

. 0.033 6.526 0.067 17.009 0.015 26.715 0.019 0.129 0.165 0.027 0.114 0.064 0.070 0.064 0.126 4.707 0.065 5.325 0.041 10.500 0.036 9.789 0.061 18.319 0.013 30.769 0.021 4.581 0.080 5.003 0.034 10.019 0.040 0.036 0.033 0.044 0.046 0.040 0.035 0.086 0.049 0.129 0.035 0.201 0.029 0.300 0.027 0.413 0.022 0.524 0.020 1.591 2.705 6.032 12.360 19.103 23.237 0.038 0.016 0.011 0.010 0.009 0.010 3.268 5.317 7.555 6.782 3.685 1.171 0.014 0.015 0.021 0.029 0.040 0.067 5.243 8.401 13.978 19.727 23.612 25.418 0.015 0.011 0.012 0.012 0.010 0.010 File: Hl-2002563 Page D-S

output h24shrt2.xls 32,500 MWD/MTU burnup - B&W 15x15 fuel elemici 5.0 YR 5.0 YR 5.0 YR 5.0 YR h4n5dl Im h4n5dl Im it 2.9 wlo U235 BPRA Curies 331.00 dwg3443 TP Curies 0.00 h4p5dl Im h4q5dl lm h4c5dl In Num of assemblies 32 h4r5dl Im h4p5eO8m h4q5eO7m h4rfeO7m neutron phot (np) phot cobalt value rel err value rel err value rel err value rel err total value rel err 7

0.555 8

0.580 9

0.579 10 0.572 11 0.568 12 0.515 13 0.435 14 0.339 15 0.234 16 0.152 17 0.092 18 0.149 19 0.355 20 0.428 21 0.520 22 0.489 23 0.384 24 0.249 sur 731.000 1

0.209 2

0.215 3

0.214 4

0.254 5

0.336 6

0.366 7

0.415 8

0.462 9

0.449 10 0.456 11 0.440 12 0.404 13 0.351 14 0.265 15 0.206 16 0.146 17 0.110 18 0.158 19 0.259 20 0.273 21 0.342 22 0.354 23 0.323 24 0.230 tal 142.000 0.033 0.032 0.031 0.032 0.032 0.033 0.035 0.039 0.046 0.057 0.061 0.028 0.024 0.018 0.018 0.022 0.021 0.019 0.041 0.033 0.038 0.043 0.038 0.033 0.031 0.030 0.030 0.032 0.032 0.031 0.033 0.039 0.039 0.046 0.049 0.031 0.027 0.020 0.019 0.021 0.022 0.019 0.597 0.649 0.661 0.645 0.616 0.549 0.480 0.369 0.280 0.192 0.114 0.068 0.050 0.041 0.038 0.043 0.047 0.044 0.020 0.018 0.019 0.019 0.020 0.020 0.021 0.022 0.028 0.033 0.034 0.045 0.041 0.034 0.025 0.018 0.022 0.014 25.039 25.312 25.188 25.003 24.426 24.104 22.790 19.663 13.797 7.814 3.424 1.407 1.593 0.342 0.188 0.197 0.210 0.151 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.012 0.018 0.046 0.050 0.062 0.078 0.091 0.056 0.349 0.180 0.078 0.079 0.148 0.233 0.541 1.557 3.631 5.838 5.837 3.820 3.178 1.156 0.594 0.566 0.525 0.440 0.132 0.190 0.279 0.322 0.273 0.200 0.124 0.074 0.043 0.027 0.020 0.019 0.012 0.023 0.091 0.043 0.017 0.024 26.541 26.720 26.506 26.299 25.758 25.401 24.245 21.927 17.942 13.996 9.468 5.443 5.176 1.967 1.339 1.295 1;166 0.884 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.011 0.013 0.013 0.014 0.016 0.017 0.042 0.024 0.019 0.016 0.094 0.128 0.184 0.265 0.325 0.412 0.480 0.505 0.522 0.500 0.497 0.447 0.376 0.314 0.241 0.174 0.121 0.071 0.052 0.042 0.037 0.035 0.035 0.035 0.051 0.034 0.030 0.027 0.022 0.022 0.020 0.020 0.020 0.019 0.020 0.022 0.021 0.025 0.027 0.029 0.038 0.035 0.045 0.053 0.043 0.026 0.021 0.014 2.181 3.511 6.210 10.569 15.176 18.518 20.164 20.740 20.912 20.793 20.255 19.841 18.096 15.588 11.622 7.172 3.988 2.060 1.860 0.546 0.329 0.242 0.205 0.170 0.026 0.013 0.010 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.010 0.010 0.011 0.014 0.034 0.028 0.044 0.064 0.072 0.045 3.028 4.266 5.184 4.624 3.270 1.596 0.760 0.296 0.192 0.195 0.268 0.319 0.770 1.710 2.863 3.885 4.078 3.210 2.711 1.270 0.695 0.422 0.408 0.355 0.018 0.017 0.022 0.029 0.038 0.052 0.087 0.118 0.157 0.168 0.162 0.133 0.092 0.059 0.042 0.029 0.022 0.020 0.014 0.020 0.041 0.035 0.045 0.017 5.511 8.120 11.792 15.711 19.108 20.893 21.818 22.003 22.076 21.944 21.460 21.011 19.593 17.877 14.932 11.376 8.297 5.499 4.881 2.131 1.403 1.053 0.971 0.789 0.014 0.011 0.011 0.011 0.010 0.009 0.009 0.009 0.009 0.009 0.009 0.009 0.010 0.010 0.011 0.012 0.012 0.013 0.015 0.014 0.023 0.021 0.025 0.013 File: HI-2002563 Page D-6

output h24shrt2.xls 32,500 MWD/MTU bumup - B&W 15x I 5 fuel element 2.9 w/o U235 BPRA Curies 331.00 5.0 YR 5.0 YR 5.0 YR 5.0 YR dwg3443 TP Curies 0.00 h4nSdIlm h4n5dlltn h4p5dlImh4q5dllm h4c5dlIn Numofassemblies 32 h4rfdl Im h4p5eO8m h4q5eO7m h4r5eO7m neutron phot (np) phot cobalt total value rel err value rel err value rel err value rel err value rel err sur 1716.000 1

3.017 0.067 0.366 0.112 5.336 0.133 16.481 0.031 25.200 0.036 sur 726.000 1

1.090 0.044 0.142 0.061 4.436 0.072 9.387 0.023 15.056 0.026 sur 729.000 1

0.583 0.064 0.105 0.067 4.459 0.069 7.190 0.031 12.337 0.031 sur 727.000 1

1.024 0.044 0.059 0.089 6.731 0.074 9.841 0.019 17.655 0.030 sur 730.000 1

0.593 0.056 0.056 0.114 5.755 0.080 7.388 0.026 13.793 0.036 tal 104.000 sur inner 6 1

12.997 0.021 0.331 0.035 2.117 0.039 13.421 0.017 28.866 0.013 sur mid 6 1

4.841 0.025 0.151 0.041 0.555 0.040 3.764 0.018 9.311 0.015 sur outer 6 1

2.044 0.035 0.075 0.047 0.190 0.053 1.134 0.030 3.443 0.023 File: HI-2002563 Page D-7

v t e Letter Sequence Supplement
TAC:L23399, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Meeting, Meeting, Meeting Questions 4 September 2002 6 December 2002 3 January 2003 Answers 31 January 2003 15 October 2002 15 October 2002 15 October 2002 15 October 2002 15 October 2002 15 October 2002 15 October 2002 27 March 2003 Results Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: L-02-010, Independent Spent Fuel Storage Installation Submittal of License Application Amendment 1 (TAC L23399), ML020980082, ML020980126, ML022950286, ML022950304, ML030430463, ML030430490, ML031250472, ML031250503, ML031250545, ML031250548, ML031250551, ML031250606, ML031250607, ML032320119, ML032970368, ML032970369, ML032970370, ML040350009, ML062980229
MONTHYEAR2002-08-29 [Table View]

ML030940368

Text

Subject:

Purpose:

Background:

References:

References:

References:

SUBJECT:

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