Text

Forwards Response to RAI Re License Change Application 237 Requesting Permission to Load Spent Fuel Casks in Trojan Fuel Bldg
	ML20134C873
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	09/23/1996
From:	Yundt C; PORTLAND GENERAL ELECTRIC CO.
To:	Stewartsmith; ENERGY, DEPT. OF
References
	CPY-035-96, CPY-35-96, NUDOCS 9610090264
	Download: ML20134C873 (76)
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Portland General Electric Company Trojan Nuclear Plant 71760 Columbia River liwy.

Rainier, Oregon 97048 (503) 556-3713 September 23,1996 CPY-035-96 Mr. David Stewart-Smith Oregon Department ofEnergy 625 Marion Street NE Salem, OR 97310

Dear Mr. Stewan-Smith,

Resoonse to Reauest for AdditionalInformation On August 20,1996, Mr. Adam Bless sent a Request for Additional Information to Trojan that contained questions about License Change Application 237 (LCA-237), which requests permission to load spent fuel casks in the Trojan Fuel Building. The responses to the questions are provided in Attachment I of this letter.

The responses provided in Attachment I are intended to answer the questions as completely and clearly as possible without divulging information that the cask vendor considers proprietary.

Proprietary information has not been included in any of the responses.

If you have any questions concerning these responses, please contact Harold Chernoff ormy staff j

at 503-556-7480 l

Sincerely, (f

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Yundt General Manager Plant Support and Technical Functions Attachments c:

A. Bless, ODOE L. J. Callan, NRC, Region IV

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L E. Kokajko, NRC, NMSS M. T. Masnik, NRC, NRR R. A. Scarano, NRC, Region IV g.

J. Woessner, TAC

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September 23,1996 ATTACHMENTS to CPY-035-96 I.

Responses to Request for Additional Information II.

Cask Movement Envelope " Safe Load Path" l

III.

Memorandum TOM-022-96, " Dose from Four Failed Fuel Assemblies-ISFSI Accidents" i'

and applicable portions of supponing calculation RPC 93-003.

IV.

Impact Limiter Properties V.

Pages 14-16 from calculation PGE01-10.02.03-18 VI.

Physical Propeny Data on LAST-A-FOAM

FR-3700, seven pages.

1 VII.

Trojan calculation RPC 96-009 l

VIII. Pages 4-4,4-5, and 4-16 of" Topical Repon Seismic Analyses of Structures and l

Equipment for Nuclear Power Plants," BC-TOP-4 Rev 2.1974, Bechtel Power Corporation.

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l IX.

Sections 2.1.1.2 and 2.1.2.1 of the Trojan Defueled Safety Analysis Repon.

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L ATTACHMENTI CPY-035-96 Page1of13 Ouestion 1.

p.9, section 5.1.2: "The thermal analysis considered the basket in the transfer cask with a helium atmosphere and the basket in the transfer cask with a vacuum." Does this bound all scenarios, including the period that the cask is pressurized to 7.3 psig? How is "short term" defined for the i

temperature limit of 1058 F? The limits of 743 F and 851 F were reached within what time l

period in the analysis? Although PNL-4835 states that tests were conducted on dry storage in i

inert gases at temperatures up to 570 C (1058 F), it recommends a maximum temperature of 380 C (716 F), and suggests that evaluations need to be performed to justify a higher guideline temperature. Based on this document, how is a higher temperature limit justified?

Response

The basket in the transfer cask with a vacuum is bounding because it results in the highest fuel cladding temperature. When the basket is pressurized to 7.3 psig, the fuel cladding temperature is lower because the piessurized helium in the basket transfers heat from the fuel to the basket shell more efficiently than a vacuum.

"Short term" is not quantitatively defined but rather used to establish temperature limits for those events which are considered off-normal or infrequent as listed in Table 4.2-12 of the Trojan Irdependent Spent Fuel Storage Installation Safety Analysis Report ('PGE-1069).

The 743 *F(helium) and 851 *F(vacuum) temperature limits are the maximum steady state temperatures which would occur if the basket remained in the transfer cask indefinitely. The times to reach these steady state temperature conditions were not calculated. The time to reach these temperatures can be conservatively estimated by assuming an adiabatic heat up as explained in the answer to question #2.

l 380 C is the temperature associated with long term storage. This long term temperature limit will be satisfied for the Trojan ISFSI. The short term temperature limit applies to off-normal or j

infrequent events as described above. Temperature limits are imposed to minimize accumulated strain which is time and temperature dependent. Shorter durations allow for higher temperature limits.

l Ouestion 2 j

p 9, section 5.1.2; p.35, section 5.3.3: The application states in section 5.3.3 that "The vacuum drying time is conservatively limited to less than 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to ensure that the fuel cladding strain that is accumulated during the vacuum drying process does not exceed 0.19'o. The time is limited as recommended by PNL-6364.. " PNL-6364 presents a range of strain values as a function of temperature and time: " assuming 70 MPa cladding stress for 24 h [ hours], the predicted cladding strain would be 0.19'o at 436 C,0.239'o at 450 C, and 3.39'o at 500 C.. With a 0.19'o strain limit, a temperature of 450 C would be acceptable for an 8-h working shift" What is the basis of the 20 4

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'L Attachment I CPY-035-96 Page 2 of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months ? According to section 5.1.2, the results of a thermal analysis were that cladding temperatures reached 851 F (455 C)in a vacuum. No time was given in this analysis for the cladding to reach this temperature; however, if the limit of 0.1% strain were not exceeded, it would imply that the time limit might be considerably less than 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . In addition to the time limit, will temperature be monitored?

Response

Technical papers on cladding strain indicate that 0.1% cladding strain will not occur for a number of days at temperatures below 760*F(404 C). The 20 hour2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months administrative time limit is imposed to prevent fuel clad temperature from exceeding 760*F(404*C). This time limit was derived by conservatively assuming an adiabatic heat up (i.e., heat transfer only takes place between the fuel assembly and cell wall), heat generation of IKW/ fuel assembly and an initial temperature of 212*F.

The response to Question 1 addresses the tb 7 required to reach 851 F.

The fuel cladding temperature will not be monitored. The 20 hour2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months time limit has been i

conservatively developed to maintain fuel cladding temperature below 760 F(404 C). In addition, heat will be removed from the spent fuel during the vacuum drying process when helium is flushed through the basket between pump downs.

Ouestion 3 p.12, section 5.2.1.1: "The design safety factors, load testing requirements, and administrative controls (i.e. procedures, training, maintenance, inspections) for the fuel handling equipment minimize the possibility of a fuel assembly drop actually occurring." When in the sequence of loading events and how often are the two cranes that will be handling the fuel assemblies and transfer cask load tested? Are both cranes and associated tracks and equipment also inspected for wear and fatigue cracking? We observed numerous crane problems during the Large Component Removal Project (LCRP). Were lessons learned from LCRP incorporated into crane preparations for the ISFSI?

Response

The Fuel Building Crane has been load tested at 125% rated load. Unless the cranes are altered or extensively repaired prior to fuel handling, they will not undergo a rated load test. The cranes, rigging and lifting devices will be tested at maximum anticipated loads using pertinent functions during preoperational testing. Cranes and supports are periodically inspected for degradation and wear in accordance with Trojan Procedure MP-1-20, " Cranes, Hoists, and Winches."

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Attachment I CPY-035-96 Page 3 of13 The crane problems during the LCRP involved alignment problems associated with the circular rail and bearing design of the Reactor Building Polar Crane. Although similar problems are not anticipated due to the different design of the Fuel Building Crane, lessons learned from the LCRP will be incorporated into crane preparations for the ISFSI.

Ouestion 4 p.13, section 5.2.1.2: The application states that the analysis for the basket shield lid drop was not performed because guidance of NUREG-0612 was implemented, specifically paragraphs 5.1.6(b)(ii) and 5.1.l(5). These criteria are only pan of the recommended guidelines in the NUREG. Section 5.1.1 contains specific guidelines on administrative controls, operator training, and equipment testing. Section 5.1.6 additionally specifies upgrades to cranes and interfacing lift points as part of the " single proof failure" criteria. Which of the guidelines in paragraphs 5.1.1, 5.1.2, and 5.1.6 will be followed? Will dual or redundant slings be used, or will other precautions be taken to preclude dropping the shield lid?

Response

PGE has committed to the NRC to implement the guidelines of Section 5.1.1 of NUREG-0612.

As documented in NRC Generic Letter 85-11, the NRC determined that implementation of Sections 5.1.2 thru 5.1.6 was not required to reduce the risks associated with the handling of heavy loads. However, PGE has implemented some of the guidance contained within those sections for the fuel building crane.

Section 5.1.2 provides four separate methods for implementing the guidelines for handling heasy loads in the Spent Fuel Pool area. For the Spent Fuel Pool area PGE has selected to implement the method of Section 5.1.2(2) with two exceptions. The first exception is to Section 5.1.2(2)(a) which recommends mechanical or electrical interlocks be provided to maintain a horizontal separation of 15 feet between heasy loads and the Spent Fuel Pool. Due to the design of the Fuel Building this separation had to be reduced to prevent movement of the hook centerline to no closer than 6 fl. from the Spent Fuel Pool (Refer to response for question 8 for additional information). The second exception is to Section 5.1.2(2)(e) which recommends that an analysis of postulated load drops should be performed in accordance with Appendix A. Since the capacity of the crane significantly exceeds the weight of the shield lid, a shield lid load drop was not considered credible and was therefore not analyzed.

Section 5.1.6 is guidance recommended to satisfy Section 5.1.2(1). Section 5.1.2(1)is another of the four methods acceptable to meet Section 5.1.2. Since PGE selected to implement Section 5.1.2(2) rather than 5.1.2(1), the guidance of Section 5.1.6 is not applicable. For added safety, PGE has implemented the guidance provided in Section 5.1.6(1)(b)(ii) by requiring the lifting-slings be rated for twice that specified by Section 5.1.l(5).

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.L Attachment I CPY-035-96 Page 4 of 13 Ouestion 5 No mention was made in the application about a possible drop of the crane lower load block into a basket loaded with spent fuel. Assuming that it could be postulated that the load block was dropped onto a basket loaded with fuel prior to the shield lid being installed, what would be the radiological consequences, and would they be bounded by one of the existing analyses?

Response

A " drop" of the crane lower load block onto a loaded basket is not considered a credible accident, therefore an analysis was not performed. However if this event were to occur and potentially crush 24 fuel assemblies (maximum allowed in a single basket) the dose consequences would be 0.018 rem (6 times the dose consequences for the failure of 4 fuel assemblies discussed in Section 5.2.1.4), which is only 1.8% of the EPA PAG.

Ouestion 6 Regarding the crane operator training specified in NUREG-0612, how soon prior to load movement willit occur? Does this training adhere to the guidelines in ANSI B30.2-19767 Will the training include movement along the " safe load path"? Will the " safe load path" be defined in procedure and clearly marked on the floor? Please provide a diagram of the safe load path.

Response

Crane operators at Trojan are given training that complies with ASME B30.2. Initial training is provided prior to being qualified to operate the equipment with requalification training required annually. In addition, specifics of crane operation for heavy load handling for ISFSI will be included in prejob briefmgs and in the preoperational setup and testing phase. Safe load paths for ISFSI will be incorporated in Trojan Procedure TPP 14-9 prior to ISFSI load handling. Safe load paths will be marked on the floor where feasible and diagrams will be placed in the crane operators cab. Sacrificial floor coverings (herculite or equivalent) may be placed on the floor between the Cask Loading Pit and Cask Wash Pit to contain drips from the SFP and rinse down.

The floor coverings may cover the floor markings. However, the load path will be designated on those coverings to the extent feasible.

A copy of the safe load path is provided as Attachment II.

Question 7 p 13, section 5.2.1.3: Please provide a copy of the analysis, calcu!ations, and assumptions which show that the shield lid would plastically deform less than 2" and not grossly fail from dropping the lifling yoke onto a basket loaded with fuel.

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.L Attachment I CPY-035-96 Page 5 of 13 j

Response

Sierra Nuclear Corporation has identified this calculation as containing proprietary information.

This calculation is available for ODOE review at the Trojan site. If ODOE review concludes that Trade Secret infonnation is required in order to complete their review, then a Confidentiality Agreement would be required.

Ouestion 8 i

l p.14, section 5.2.1.4: " Mechanical stops and electrical interlocks on the crane used to lid the transfer cass will ensure that sufficient distance from the Spent Fuel Poolis maintained." What is this distance and what is the basis for choosing it? [NUREG-0612: 15 feet in section 5.1.2(2) or 25 feet in section 5.1.2(3)]

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Response

The distance of 6 feet is based on allowing crane operations in the Cask Loading Pit which is only 9 feet by 12 feet and located adjacent to the Spent Fuel Pool. The 6 foot separation distance is sufficient because the maximum lining height of 6 inches prevents a dropped load from tipping and entering the Spent Fuel Pool.

The 25 fl referred to in 5.1.2(3) pertains to heavy loads over the SFP and the distance to be maintained from " hot" fuel. Trojan's fuel does not meet the definition of" hot" fuel.

Heavy loads are not intended to be handled over the SFP as part of the ISFSI fuel transfer operations.

Ouestion 9 Does the lining yoke meet the criteria of a "special lifting device" described in NUREG-0612, section 5.1.6(1)(a)?

I Respoase The lifting yoke is a special lining device as described in ANSI N14.6. The lining yoke does not fully comply with the criteria in NUREG-0612 section 5.1.6 (1) (a), which requires doubled safety factors or load paths for single failure proof handling systems. NUREG-0612 provides the alternative of analyzing consequences of potential drops in lieu of single-failure proofload handling. In accordance with NUREG-0612 Sections 5.1.l(3),5.1.2(4) and 5.1.5(1)(c), the i

effects ofload drops have been analyzed and shown to satisfy the evaluation criteria of Section 5.1 of NUREG-0612.

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L Attachment I CPY-035-96 Page 6 of 13 Ouestion 10 p.14, section 5.2.1.4: Maximum accelerations for intact fuel are given as 82g vertical and 44g horizontal. Maximum accelerations for the closed basket are given as 124g vertical and 44g

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horizontal. What are the bases for these limits? Were the calculations used to get these values i

reviewed by the NRC7 What acceptance criteria were used to determine that these are the limiting accelerations? For example, what is the unacceptable consequence of 82g being exceeded? Furthermore, after sustaining forces in excess of 82g, would the fuel have enough integrity to be offloaded, if necessary?

Response

The storage basket is designed to withstand accelerations of 124g vertical and 44g horizontal without resulting in loss of confinement integrity. Based on studies performed by Lawrence Livermore Laboratories which were distributed to irradiated fuel licensees by the NRC, intact spent fuel assemblies can withstand 82g vertical and 63 g horizontal.

Since no credible or design basis accidents were identified which would exceeded the bounding limits of 82g vertical (intact fuel) or 44g horizontal (storage basket), the consequences of exceeding these limits were not analyzed.

Ouestion 11 p.15, section 5.2.1.4: Visual examinations on failed fuel assemblies are mentioned. Did these visual inspections look at interior pins? Were other techniques like sipping or eddy current testing used?

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Response

Following Cycle 4 fuel failures, Westinghouse performed sipping and sulo probe rod measurements. Failed fuel was then subjected to high and low magnification examinations. In 1988, ultrasonic and visual exams were performed on suspected failed fuel assemblies by PGE/ Westinghouse / Brown Boveri in preparation for reconstitution. Visual and UT inspections identified failed fuel pins in the interior. A Westinghouse fuel repair campaign was conducted in 1989, transferred pins were individually inspected, classified, and failed pins were either left in fuel skeletons or removed to a storage container. Past sipping campaigns, RCS Chemistry trending, and refueling visual examinations have provided Trojan with an extensive data base on the fuel assemblies.

l tM Attachment I CPY-035-96 i

Page 7 of 13 Ouestion 12 p.15, section 5.2.1.4: Do the words " criticality concern" in this section mean keff<[>] 0.957

Response

A criticality concern arises when K,g exceeds 0.95 (K,, > 0.95).

Question 13 p.15, section 5.2.1.4: Please provide a copy of the analysis, calculations, and assumptions which show that the dose at the site boundary would be about 0.003 rem if a transfer cask was dropped prior to the shield and structural lids being welded to the basket. Is there an analogous calculation for dose to personnel in the immediate vicinity? Also, does the term " site boundary" have a consistent meaning throughout this LCA? Is it the Owner Controlled Area, the Industrial Area, or some other boundary?

Response

A copy of the calculation and assumptions for the 0.003 rem dose at the site boundary for a transfer cask drop prior to shield and structural lid closure is provided as Attachment III.

There is no analogous calculation for dose to personnel in the immediate vicinity. Local monitoring of radiation levels and airborne activity would be in effect during fuel handling activities. In the event of an accident personnel would be directed to evacuate the affected area.

The site boundary is described in sections 2.1.1.2 and 2.1.2.1 of the Defueled Safety. Analysis Repon (copy provided as Attachment IX). The site boundary encloses a portion of the Owner Controlled Area (Exclusion Area). The Industrial Area is contained within the site boundary.

Ouestion 14 The consequences of transfer cask drops are diminished by using impact limiters. In order to verify the calculations of the drop a nt scenarios, please describe these devices. Please provide a description of the dimensi material specifications, and energy absorption characteristics of the impact limiters. Are these considered "imponant to safety"? Are they purchased commercial grade? What procedure or process is used to verify their energy absorption characteristics?

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.s Attachment I CPY-035-96 Page 8 of 13

Response

Impact limiter locations, cnish strengths, dimensions and densities are summarized on Attachment IV, and the application methodology is described on Attachment V. Note that energy absorptions characteristics are not anticipated to change, however, dimensions may be revised to accommodate installation and placement. Although calculations indicate that an impact limiter for a side impact at the hoistway is not required, LCA-237 committed to providing j

one. The dimensions for this impact limiter have not yet been determined.

The impact limiters are considered "important to safety." They may be purchased as commercial grade items and dedicated for use, or as basic components under a suitable Quality Assurance program. Vendor certification of material propenies will be required to verify the energy absorption characteristics.

Material properties are described on the manufacturer's data sheets provided as Attachment VI.

Ouestion 15 p.15, section 5.2.1.4.1: Please provide a copy of the analysis, calculations, and assumptions for the accident scenario involving the transfer cask drop into the cask loading pit.

Response

Sierra Nuclear Corporation has identified this calculation as containing proprietary information.

This calculation is available for ODOE review at the Trojan site. If ODOE review concludes that Trade Secret information is required in order to complete their review, then a Confidentiality Agreement would be required.

Ouestion 16 p.16, section 5.2.1.4.1: The application states that "the concrete at the bottom of the Cask Loading Pit was determined to be the limiting component on which the height of the impact limiter is based." What is that height?

Response

The design height of the impact limiter to be used in the Cask Loading Pit is 34 inches.

,a Attachment I CPY-035-96 Page 9 of 13 Ouestion 17 p.16, section 5.2.1.4.1: If the concrete floor were breached, would water be lost from the pool?

Are there any circumstances in which the gate would be open while the cask was being moved into or out of the loading pit? What would be the recovery actions?

Response

An impact limiter will be installed in the bottom of the Cask Loading Pit to preclude damage to the structure and the fuel during a postulated cask drop accident. It is not anticipated for the gate to be open during transfer cask movements. However, breach of the Cask Loading Pit liner, discussed in section 5.2.3 of the License Change Application, addresses the scenario where the gate is damaged, which would result in conditions similar to an open gate.

Ouestion 18 p.16, section 5.2.1.4.2: The analysis assumes a drop from 93'6" What physical controls prevent a higher starting point?

Response

There are no physical controls intended to be used to limit height of the transfer cask above the deck. Administrative controls will be used. Flagging or tape or equivaient may be attached to the load handling equipment to assist the crane operator in maintaining lift heights. Personnel on or near the floor will verify clearance from obstructions and the height of the load.

Ouestion 19 p.17, section 5.2.1.4.3: Please provide a copy of the analysis, calculations, and assumptions for the accident scenario involving the transfer cask drop into Fuel Building Hoist way. Include the determination of the resulting decelerations in both the vertical and horizontal directions. Are the retaining clips used to secure the boral plates designed to remain secure during this drop?

Response

Sierra Nuclear Corporation has identified this calculation as containing proprietary information.

This calculation is available for ODOE review at the Trojan site. If ODOE review concludes that Trade Secret information is required in order to complete their review, then a Confidentiality Agreement would be required.

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Attachment I CPY-035-96 Page 10 of 13 Ouestion 20 l

l p.20, section 5.2.1.8: Please provide a copy of the analysis, calculations, and assumptions for l

the accident scenario involving the basket drop into the concrete cask. Include a copy of the methodology of EPRI Report NP-7551.

Response

Sierra Nuclear Corporation has identified this calculation as containing proprietary information.

This calculation is available for ODOE review at the Trojan site. If ODOE review concludes that Trade Secret information is required in order to complete their review, then a Confidentiality Agreement would be required.

Ouestion 21 p.25, section 5.2.3.1: The application states that water is available from various sources in the event of a breach or tear in the Cask Loading Pit liner. The Trojan DSAR describes several water sources, including Demin Water, Primary Makeup, Senice Water, and the Fire Main.

However, these systems are not subject to traditional Technical Specification operability requirements, and could be out of senice. What steps will PGE take to ensure that makeup is available on a timely basis?

Response

Trojan Procedure, TPP 30-1, " Nuclear Division Defueled Requirements and Defueled Systems List," describes the status and programs for maintenance of certain systems for the defueled plant. Included in TPP 30-1 is the Trojan Defueled Systems List, which lists the plant systems and their status. For those systems identified as Important to Safe Storage ofIrradiated Fuel, which includes those systems identified as makeup sources in the DSAR, administrative controls are imposed.

Ouestion 22 p 26, section 5.2.3.2: Please provide a copy of the analysis, calculations, and assumptions for the accident scenario involving the crane mishandling operation for a horizontal impact.

Response

The crane mishandling operation event is presented in the ISFSI Safety Analysis Report, Section 8.1.1.1.

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,s Attachment I CPY-035-96 Page11 of13 Ouestion 23 p.28, section 5.2.4.2: How fast do the transfer cask bottom doors move? Do they fail open, closed, or as-is?

Response

The Transfer Cask bottom doors open and close in approximately 20-30 seconds. These doors use a three position (open, closed, neutral) valve with a spring return to neutral position. The Transfer Cask bottom doors will fail as-is on loss of hydraulics.

Ouestion 24 p.30, section 5.2.5.1: How does PGE propose to eliminate the possibility of an occurrence similar to the one reported in NRC Information Notice 96-34 where a hydrogen gas ignition occurred during the welding of the shield lid on a spent fuel storage cask at the Point Beach j

Nuclear Plant? Do any of the ISFSI components involved in the spent fuel loading have a i

zinc-based coating similar to the ones mentioned in the Information Notice? Is there any concern of degradation of zircaloy cladding due to the presence of hydrogen?

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Response

Currently, Carbo-Zinc 11, a zinc-based coating, is applied to the Basket internals and to the exposed carbon steel components on the Concrete Cask. Sierra Nuclear Corporation is re-evaluating the use of this coating in light of the chemical reaction problems at Point Beach.

The Trojan design may be modified based on the outcome of this evaluation. Additional information will be provided in a later submittal.

Question 25 p.30, section 5.2.5.1: The application states that "about 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months of heat-up are required for the water in the basket to reach boiling." How much time is required for welding and drying, assuming these evolutions are completed without problems? What are the consequences in the event that boiling is reached? Will the water temperature be monitored if there is a delay in welding the shield or structural lids or in draining and vacuum drying the basket?

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Attachment I CPY-035-96 Page 12 of 13

Response

The time estimated for welding and drying is about 16.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . This estimate is based on data from other facilities that have performed similar operations.

The consequences of the water boiling have not been analyzed because corrective actions will be implemented to prevent boiling of the water if the welding and drying process is not completed within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

The elapsed time for the welding and drying process will be monitored for each basket. A linear heat up rate is estimated for the water, hence, the temperature of the water is directly proportional to the elapsed time and direct measurement of the water temperature is not required.

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Ouestion 26 2

p.32, section 5.2.5.2.2: Where does the 12 pCi/cm come from, and how does this translate into a dose at the site boundary?

Response

A copy of the calculation is provided as Attachment VII.

Ouestion 27 p.33, section 5.2.6.1.3: The analysis for an earthquake determined the kinetic energy input to the transfer cask using the " square root of the sum of the squares" combination of the peak horizontal and peak vertical ground velocities. Why was this method used here but the 100/40/40 method used in the SAR for the ISFSI?

Response

The reference used as a bases for the calculations specified (see Attachment VIII) the " square root of the sum of the squares" combination of peak horizontal and vertical ground velocities.

Ouestion 28 Throughout the application, the use of procedures are a key method of assuring safety. Since it is appears that many of these procedures are not yet drafted, how will PGE ensure that all of the procedures for which credit is taken in the application are satisfactory and ready for use before fuel moves are started?

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1 Attachment I CPY-035-96 Page 13 of13

Response

PGE uses an administrative procedure, TPP 18-2, " License Change Revision," which is a process used for previous Trojan License changes. This process includes reviewing the requirements in new or revised licensing documents to identify existing procedures that require changes and any i

new procedures which are needed.

The procedures for preparation, review, and approval of revised and new procedures will provide assurance that the Trojan LCA-237 requirements are met.

Ouestion 2_9 Will seismic monitors and audible control room alarms be functional? Will there be communication between the control room and the fuel moving crew during loading operations?

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Response

Yes, seismic monitors and control room audible alarms will be maintained in accordance with the Trojan Nuclear Plant DSAR. Communications will be established between fuel handlers and the control room. The on shift supervisor will be able to communicate with the certified fuel handler (floor supervisor) located at the spent fuel pool via independent circuits.

l Ouestion 30 Do procedures preclude fuel handling while the transfer casks is being moved?

Response

Currently they do not. The procedures review process hasjust begun and will be updated to include operational restrictions as more operating design information becomes available. The Fuel Handling Procedures will administratively control fuel movement and cask movements such that during cask movements near the Spent Fuel Pool, no fuel will be moved except to j

mitigate hazardous situations.

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smmmmmu, gl glr- - - - -t-v - - -

's J-

- --T-_.------_FT-_._._.___1 o;f j F,4 vyysg q= =

l l

N h>.

.f j

j!

_. h

.I

~

d.,

A\\

l l

lj fmmm um

, I

'EL.72'-0l j

ij I

l i*

3.___._._.,___~,,_.r-------+

-- r' $,.

L-

i--

p i,

I

(../

l j

EL.52' 6' j!

l, y

C

.I

/

!j l.

I i

lI l

DPERATING FLOOR ji

-TOP OF RACKS I

l EL.93'-O' l

[

EL.67'

ll a

l

[

l'

. -----_-_.-._-._._v._.._._._._._._.

/ (Ij!

Il

!5 CASK LOAD l

PIT, 9'X12' t.l l

ji I'

7 d. '. ! l

/

!ll j

I EL.4 9'-4' j

I

j. _ _ _.

7

_ _ p.p.. _. _ _ _ _,i ;4_,

i f

i..____

.___pp._.___.

p

_ SPENT i

j j FUEL POOL _

I__

m.

~

1

.I

]

i I

/

l s'

/

/,/

/

o c,

O. C l

l i

v pawe m sm n.

C P4 - o 36 -c} 6

)

,s

.t l

l ATTACInfENT III Memorandum TOM-022-96 " Dose from Four 1

Failed Fuel Assemblies-ISFSI Accidents" and 1

applicable portions of supporting calculation RPC 93-003 l

m.

ces-ossW (o Trojan Nuclear Plant Radiation Protection j

\\

TOM-022-96 i

To:

Harold Chernoff From:

Tom Meek @

Date:

May 23, 1996

Subject:

Dose from Four Failed Fuel Assemblies-ISFSI Accidents I have reviewed previous fuel handling calculations completed at Trojan and have concluded that the results of the previously completed / approved calculations can be used to determine the projected dose from ISFSI handling accidents involving failed fuel.

RPC 93-003 ' Dose at Existing and Proposed Site Boundary from Fuel Handling Accident after 6 months decay' can be used to determine the dose at the site boundary due to the projected release from four (4) failed assemblies in one MPC basket.

Method From RPC 93-003 the wholebody dose at the site boundary due to the release of Kr-85 is 0.3268 mrem (page 12 of 19).

To correct for the assumption in RPC 93-003 that only 30% of the Kr-85 activity from the fuel assembly is contained I divided the gap dose by 0.3.

The dose at the site boundary from the release of all the activity in an assembly is 1.09 mrem.

I next corrected the above dose for the decay of Kr-85 (10.74 year half life) to 5 years post shutdown (t=4,5 years).

The decay correction is 0.75 which results in a dose from one assembly with 5 years decay of 0.818 mrem.

Finally I multiplied the dose due to the release of all the Kr-85 activity from one assembly by four (the maximum number of failed assemblies in one basket).

Results The projected dose at the site boundary from the release of all the Kr-85 activity in four fuel assembles with 5 years decay is 3.27 mrem.

References RPC 93-003 ' Dose at existing and proposed site boundary from fuel handling accident after 6 months decay' Radiological Health Handbook, Bureau of Radiological Health, Revised Edition January 1970.

'%m 1-/

/0 6

,,.7.

_ s3 e.

a shee,,

[tJIC M C

'T 3 oo 2 s e tor Joe no, GENERAL COMPUTATION SHEET Z File Project / Job 1

Subjee: Dosas A1-UsH%

h osEb S h bouw enu A4m Fv at M ww.w(, 4 ci. l e,

Chk.By [

S Date

/d /T 7 orig [Rev.Z By

- d N *== =

Date N ~M

U

/

CBJECTOVE The objective of this calculation is to determine the beta + gamma skin dose, gamma whole body dose and thyroid inh'aTation dose at

- ~ - ~ ~ ~

une sace councary, ano at une intaxe auct or une control room

.voncilatiwu.sy-t

., Th. calwulatics includ..

.G scuch;d. coy.

!-the te shulate *be decay *ir- ' frem end of -per-2 eparatiens te l

' possession only license.

The_.s_ource is a si,ngle_ fuel assembly l involved in a fuel handling accident in the Spent Fuel Fool.

The i

jresults of this calculation are intended to be used to justify, 4

emerganuy pAannang assorns.22 v.as nae.eAlmananlon.as osssans l(Aj m__

m =,

=.,,g j

Ac* ion Guid== and Pret=ctiva 1_ctions far * "lamr Inciden*="),* and i

.(2) to orovide the basis for locatina the protected area barrier.

l l l i ! I i l i I I

i i i l l l l ! t I i

l I i i l l l In addition,; thislcalculationiwill estimate ltheigamma!wholel body [ l uano asicae sace uvunuary,, asuur o muuuasiuscay u2me,,as a

iipping cc[k' i

pent ; fd=1 p;;1.

i I i l i l iI l l 1 i

i i i i i i i i ; i ! ;,.

i i i : i I I I I I I I I I I I I i l i I I I I i ! I i l i l l I I I i l I i Ik RESULTS I

' i I ! i

! I i !

I I

I I I I I

'1 2

I ! I I I

1 I I I,! !

! I I I i i

1 i i I i i i i i ! J us

,2...ua = l au. s, ys w v 2w=u a,u p au2, = ; A, t a, w wawu=, u a.

m. u u sayAmau=u,

i.

n m;,,,,

i i i,,

i i i i i

V

' I i l i I

i i i l l l l i i ! l

.,x 1

I i i ! l l I ! I l i l i I 4

Part 1: Existing Site iBoundary Dose i i !

i I l I I I i l I 4

l i l I I I i i i !

i l !

I l I i i ! ! I M

Twg casesi:orinneis1:e Douncary:cose wereiconsiderec using l l l l ; g 3

ag

a. u. -

'% q

,v.. w.

4. < u ww1. y suw.u u

2.

2.u w..v us w.

w.&m..

v...

.u...

u.

i ee us m.w,2

-e'nle'..

w.

w_

4_.

2

....,...a used'thelung ctiidai i I I

raintilti ng I f ri a dame lef o.TR2 mr am.' I t'a me 2 1.25 assumptions ofl104lof the core'inventorylof Iodine!andinoblei G 4 gases (except Kr-85) and 30% of.the' core Kr-85 inventory.l The r'aAM Ia. Die 13.u-D. core anventory was usec.

j

~_ _ _.._ _.

..a._~___.._ _,

.s w.._ _, - -,, - ~. m__,..__ __..__. _.,..__

.m_

y 4

results were then multiplied by the peaking fac*er of 1 65 and 6

the results are reported in Table 1.

iTnereIore, nne gamma wnole cocy cose an :ne site councary arter l6 senth. d cay fe wm th= fu=1 assauly.i h th= h2ghc.;

v e

=2

' density, invc1ved in-a-fuel handling--accident in the cpent f-uc1

peel is 0.s40 mvem.

7 Part 2: Dose at Control Room Ventilation System Intake

?be dos At LL.

utake Of C3

.d. cslwulated u.2av he

.se scurce ten and decay tire.

Eceever, a Chi /C frc= an existing calculation (TNP-88-38) was used for the CB-1 intake case.

I-129

,was. included in:this calculation as'it is a long lived isotope.

The dose from the :1-129 source is small compared to the thyroid,

an eN'w n.an une e po i cecay t time 4,,/ y y ; /Tl sts Irom.caeiA-lal pono 3

--. A n i.

,e m,o nw Se,,1*e ede

%Ie 7 e le m*4me s. ek tm m.=

sk74 a

i i

,1 i !

l [

t i

} ~'

PGE t *37 (Nov 89)

.\\

m 424 712 11 79 Shset of Sheets

.1lC S40207

? Joe No.

& C R 1 ~ C O ~-.*

GENERAL COMPUTATION SHEET

[ File Project / Job Subject Doses A,-

U 5ciN4fblocosES S h bcow en.e J F">L.3 m Fw a t M Auct % 4c$1 e..J5 d Nh--

Date N ~/O' UChk.By.

,N Date 8'// M Y

~

Oris _. Rev.

By Shipping'Cas)t Kccidens Paru a:

I The dess-from-a shippina "=ek accident uae calculated *cr-4__r.he_

av.isting site boundary and B) the CB-1 intake location.

The calculation assumes that 193 fuel assemblies from the recent

,defuel'ing are damaged jand.that the radiai peaJcing factor is:not p.nclucea as nne a A A Ama sueA assemmAAes Irom cae receau care a

defueling are involved in the accident.

6 l

CONCLUSION 1

1 t

! l

Am.s c4AvuAsu vu vvuvauu== una u uum u=uaw amma muun uv==,

samma

'" "*'d^""

^' *' k -

'n N"

'^^

t,^ 01' body dccc and ; thy' cid d^"-

14 6 4 4_ d a,,d DrntactivaIActi'en GuidasIforlth,javisting site; I I I l bo' ndarvi and at :the ICB-2 intakeJ IIn addition; the doses are lessl I i u

than the criteria Ifor Imaintaining ian emergency plan; !

l l l t l l l i

i i i i l I.I l l l l ! I i l I i i I ! I I I i l I I i 1 I I I i w L.y a m m a,w u w l s,u w m u w.. r,,a.

.. u l y y i.ng, ca.k. accid-nt, iuve11 uv l l I tu.

m

, a.,

is,, _ v e..

wt4.,<<,

4.,,,-, a 4 a. 4,

e, i s,,. ',,

w.

I 8

i avisti'nd site boundarv and 87.Id'Ere5 lat the CB-1 intake.

I ' I.'

I I l 1 i i ! ! I i l I l i I I

i i l i i

! I I ! i I i l I ' l

' I l l l !

l I I I l ! l l l i l i i l l 1 l I I I l l l l 1 l I m;urrA.num uxm.aua l l ; l j j l l l ; ; j j j l ; j ; ; ; ; ; y I na 'a nn ! Ja J 6.,* Jr4,;. i

,1+ 4 m A c,,4 a.

en,.!es.Irm,.b IeA.1.I I I i

r a

L g #l a' Nuclear Incident' includes the acceptance criteria of 1 rami I I 5 i

I 1 l l l i i i i ! i i i i : 1 i i i l I ! I 1 ! !

i l i l l l l 1 i ' ' i i ! i l_

l i

l i e

i i i i i I i I l I l l I

i s

l l 1 MW) Y Pi' l l l l l l l l l ! i l I I I I l i l l l l l i

- h *a c:irl e n T n e 4 n n 't.rm a i e n it em the emmnutarl code ;,

\\

o o

u o e r u

'Ph m annehnA s te.d.,*n e

DOSES which is a documented computer co$e per'NDP 200-5 and:is an.

accepted computer code used in RadiatipL Protection at Trojan.

,Th;e code inputs were determined fpym'FSAR data (see references)

,d n u AUGauueG d IJL ua biliuy LaLu, 644a / W, uuvay we anu uuvaAuc iaC.

U[.

~~"

v I

m I

ASSUMPTIONS

_1._

1.

DredLaADQ MdLe m J.ii_

f ai 21/ D e G.

(fbnX_ _ _..AdDie 13.W e).

9 nw i /n = a es r-a cac /rvaca-*Maeaw*woen*1 trgan mahl. 1s_o-o for O-2 Mr release at' site boundarv).

i 3.

Decay Time = 6 Mo. = 180 Day

24 Hr/ Day = 4320 Hr

~

' ~-"-

'hc fuel acccably i-nvcived-i-n-the accident _had--the I

hinhane nankfnc accer

<n tha cara.

Tharafera. us, a 1 i peakinq ' factor'of 1.65 (per Reg.: Guide 1.25).

l l l l l l I I I l l l l 1 l l l I l I I I I i l I i i l I i i l I 5.)

l Ch1/Q = 5.89;E-4.Sec./[ Meter

Meter

Meter] (TNF;88-38) for l l g au aui.aka-vf C5-1 ame.

. g yj

.dv==

g cm,w m m

g et s u 'i2 ti.7s of sheets CJsb Ho.

RPC 93 oc3 s 4 207 GENERAL COMPUTATION SHEET C File J Project /Jcb subject Domes Ar' 645riuc [ h oosEo E h 8 e u w oa u F A N as N A w w w t. 4edle.d

/ d W-cat. 2 -/C - LT chk. By IE, #

l sy f ld M 3 orie W a.,...-

oste DESIGN INPUTS

.i...

F5aR awle 15.U-5 Cvce and Gap,invem ucles, were used in U21

=1=1H =-

i l l i i

l,,,,

i i i i 1

REFERENCES t

i l

i FS?.R,ctspte'r.lf.

f f

1.

2.'

Reg Guide 1.25,.

(Safety Guide 25), " Assumptions Used for.

l l Evaluating,ttleiPotential: Radiological Consequences lof a Fuel, l l 1 sana.ulng Acc2.cenu in it.ne. rue;. nano 12.ng lana ; norage l racamy, ; i

. UYl I l ! l l ! l l l l l l l l 1

5" N 5"

'i 3.I I EPAl500. 'Ma'mia lI of IProt$ective Actilon IGuides And Protectivei I I I i

l l AlctLicns For Nuclear Incidents. I l I i l i l l I I l l l l l l l l I i ! !

i i l i ! I i i ! : ' '

l ! l l

l I I I ! l ! ! ! ' '

I I i i i i i I ; I ! I i 1 ! l 4 l l l l l l

l 1 1 I i ! I I '

i i l i i l I I I I l l I I I ! I i I I I l 1 1 l l l l l i ! l l l 1 I l l l l l l I I l !

I I I ! l I i l l I I I I I I i ! I I I i i i l i i I i l i i i l i i ! i I i ! i i I 1 I i i i I i i l I l l l l l l ! !

l ! I I i I i ! I I ! l ! I i i l i

I 1 I i f I I I :

! I I I I I i ! ;

I I I I I I i l I I i i i I

+

I i l I i i

4 6 I I I i i t i 1 I i i i ! I i I i l ! l l I I I I i l l l l l i I e l I l l I I I I I I I I I I i ! I I i i

I ! l ! I i '

i l i I i i i i 1 i

I i i i 1 l 1 I I i i i

! I I

i i l

! i i

I i i

I l

l 1

i i

i l l 1

i l

4 i

a e

w,e k

)

l l

l l i

l l l l I

1 l

t i

i l i I l I I I I I I I I I l I l I i l I l l l I I i i l l I I _I I I !

i i l l I I I I I l l 1 i l i I I I i I I !

l I I I ! I I !

i l

i I i !, i e I I

i PGE 1737 (Nov 89)

e!

e *,

l' 32 4712 11 79 Sheef of

' Sheet 3 w

4 40 207 Job No.

k b C "T 3 - C C 2 1

GENERAL COMPUTATION SHEET O File j

Protect / Job Subject DesW5 A T' N 5 A ""=

esEo 5 ia bev a en.t a _F% iro y g,,

s,.1, s w (,

A u ;; g,,;-

I d N*- =r ---

N "M" UChk.By

A Cate MdN 3 Orig [Hev._,,,

Date gy rnbarTramTnw l

l l

l '

I i

i !

l I '

i i P a r t! l': Existing Site Boundary Dose I

oww. s.

A-sm usw eaanawleu.

l bab 1'

I

PMID

'T'a bl e T5.0-5 Cap Tnventet-v i

l i

,Use FSAR Table 15.0-5 Gap inventory for 193 l assamoAles j

l l l l l-l

,, i

, l l,:,:i.-ide tetl;e.' W. J W r i r.

e,,'

"'li.

,. j

.ti 100 0; l l l l I l l l l i >'i i ; 1 i I m

.3i.:.

,,,w 3,y l l l l l l l3 i ~l 'i l I i I )

i I I l'

l

' ! I l l l l l l l d:

Casal21 l Reg.! Guide 1.25 Assum':rtions f(10% core inventory IIodine i lj jand Nciblie i as iexcept.Cr-85 wt11ch is ;304)l l j l l l j l l lq G

l l l l 4

i

.' i i t t.._l,.1_ _ :

i l l l.l i

I i i i u_

_.a 3 ! -

%g I l l I I i I s ia -s.....h 114.;. '

+ 1 i ' '

i '

I i i8 i >

%h i l l I le I l biviide Tabl'a 1 5'. 0'-5 v' lues by l'93 tio lobtdin Ciiin ! 1 i

1 a

l l I l l l l the fuel lof one hiel :assestbly ! l l I i l l l l 1 l lpN j l i l i* l l Assume iall gap activl.tyl consists ior 104 orjruel i j g fri A ouwe anu nooAs gn===

ww.ywins os, i ; -

; ; ; ; ; ; myanwry or

- i in 1;

e, n

_.... _, 4 m

w__.__

u.--.,

,l i

i I ()

l l l

I I I I : !

4 i

i,

i i !

! i ;

i i i i i i t

i

i i t i l 7

l l I I l I ! ! I i i i

t i i !

! ASE I J i j i l ! !

l l l Case 2 ) i l l I i J,

I i l l I I i i C

f I I I l_, l. t l_ ! l I i i !

I

' : i I

' l 1 I I I I I i I IN u.

.a way i

u ws.

n.w.

..u y

5.wwwy-.

p., au..e.

i r

i i, _s...

.s._ ewe r_v. s

.s._ewny m_,

l,_o_

.sewmv m_,

. s_ewnr v_,s R

I

r,n,.

e, r,

r, R

l i

s i.

~

3.75 E3-

'8.80 E7 4.56 E4 I-131 i 7.24: E5 I-132 1.20: E5 6.22 E5 13.4 E7 6.94 E4 4

R I-133 5.34.E5 2.77 E3 19.7 E7 1.02 Ed T

N' I-134 1.25 E5 6.63 E1 3.1 E7 1.15 E5

,- v4 r_,,e

-e re 4-r,

- e e-o q

Xe-131m

'O.0669 E5 3.47 El 0.0668 E7 3.46 E2 Xe-133 13'.6:

E5 7.05 E3 20.3 E7 1.05 ES X4-133M 0.225 E5 1.17 E2 0.516 E7 2.67 E3 X=-135 1.00 E3 5.15 Z^

5.55 E7 2.58-E4

%il2EM

^ 155 E5 S.55 El 5.'S E7 2.32 Et v -17a n_ mas rm 7 o, r9 17 o r7 o_97 ra Kr-83d 0.135 E5 6.99 El 1.64 E7 8.50 E3 d.,*.'_ ~f?

? ?? ??

.Y ??."~

.f ?? f.f,

~

-.? ?..--

.".r i

-. ~..

.~

I I

I

'n_. _e n_;.

r_e

_ _ c_,

r_ _e

_ _c a. r.

o._,

v_ 4_

w.L_o,

l a.g E7 I I I I 5.60 l E4 I I I i rr.; ma i l m

_1 rs; I I I s s4 r? I l I 1. b 7!

i 1

1 1Crb89 i !

'O.192 'E5 i i l 9.94 El' l14. 0 E7 -

I ; 7.25'E4 I I l

~

1 l i i ; i ! i i i i

i ! ;.

pos t7s7 ow, es>

.-~

h e3-475a it.7s Sheet of Sheets

_4.as 4-'o 207 Job No.

k b C

T 3 ~ CC ~3 GENERAL COMPUTATION SHEET

.._ File 6

Project / Job Subject Doae5,a r-E s.I s rN 4 T.sosED Siht bCwa On u F43m p v gg y,bau,,w(.,

Aci,1,j-case 2 ~/O *'7 7 chk.sy I E N

~-

/d h

(/lb/i2 orig 7aev.._"

ey Date cau u um un cencinued

i 6

! Fuel Handline ' Accident toneI uel assembly) 3 f

uvses uvde vulpuL (aLLavhud) ausu1L.

[Aue I i

' i i ic,c, s,

case p2

' Skin Dose

.1.539 E+1 mrem '

2.752 E+1 mrem -

fl"l.2h2l l1!md 2$. f I

-O 3 d-ce' 1.224, ;- 1 Li -

n I i l ii i i l l l l l l l 1 I i i I

+

l l

r i

l I

Tn'hmlatien Thyroid DeseI I I 11.3495'E-1 mrem I lJ818!EIOImren J-f !

l l l 1 l l ~l l I l l l l l I I I I I I i I

! ! l l l l l ! I l i i Total lThyroidjDose l l l l l l 3. 319. l E-1 ; mrem i i l24145lEjol mrem f y l

1 1 !

I i i f i i i i i I

i i I. !

. I I Thl. C&5e #2 dG5e esti-o..

th;I

i.~

abGie are ;^altiplied by, h+ = 4 n ' *-h ! <]

..w-

_-a A_ _4'= _T

n. M_ _ _,,.,

_e._ M_ _ _,. i n_ _e m.

_s. _ d_ _c ' i no v.

i

' en 4 A. : s _' e m y i + m non m

m filnsi ldese estimated.I I I I I

' l I i i l ! I I I I I I I I

I !

'M l l f I l l l l l l l l l l i l I l 1 !

I I l l l l l ' l l l$ h l

l l j l l l l l l l l ; Final Existina Site Boundary Dose l l gq) b 4

l l l i l l ! l ! I i^"

- - - l i

l ! l I i i ; I R l

i i

l I I i I lM h cne! n find. I i l I l i I I las;_a mrem.

1 4

! G i l ! I l 1

4 I i l f I I l i i i i i I !

I WholeiBody Dose l l l l l l l l 0.54imrem - i ; l l l l l l l I i i

! i i j l t

i I l I I I l l : I I I l l l 1 i i

8

(/

s e

4uanea u.vgi.uys vid D, v=, =.

a.vv au. =m -

v

- I m,

i

.C m n' +- = 1 mhu4mia!nnem i l I

' ! 8 1_sa mrom

!*************************************************************w***:

'Part 2: Dose at the Intake of CB-1

[

9)

^

waava. alien is the 54s= d5 ids

, aue avuava we&a a v s-ras vs we

.uced in Part 1 (10. Reg Cuide 1.25 Ci inventories)_

EcYever, e

't ha chi /0 f

mm *MD AR-3R van used for the new recenter location.

%g The Doses code results are shown below:

T g

Y k

A Skin Dcce 33.12 =re:

Whole Body Dose 1

0.4523 mrem -

Inna).ation Inyroic pose 2.31.3 mrem -

.a.:,,u,0 a i! i i i

,m 1

I l# l'l' l' l'I l l l 1 I i i I i i i i t

i i i i i i l l I I I I I

I i I I i l !

l l 1 1 I i l i l I t l I l 1 i ' '

l i

i i i i i i

/O y.:

(2 <712 11 75 ef sh is 8*'M87 Jsb No.

k b C. *T 3 ~ CC 3 GENERAL COMPUTATION S.HEET G File J Project / Job Subject Dos #5 A T*

b d 5 FI N (=

PLauc5fD l' h cua en u F% Fv as w a ys.w(, 4ced e,;i osi. S -/O-77 cak. sy T F b 1 cat. h /o/f3 ong M,v._.

/d h sy The above dose estimate:s from above, are multiplied by the radial

~~~

=peaxing ractor or A.oo (per.xeg. wuide A.go) to obcaan nne rinal

[dysmiates.

! I Final CB-1 Intake Dese Estimates i

l Skin Dose 62.9 mrem -

t'[cic Body D00:

3.745 mr:m

,i

! Inhalation Thyroid Dese 4.15 mrem -

l l l l l 1 i I l l l l l l l l l i l t

I l i i l l l ! l l l l 1

Total,Tayrold Dose l l l Jl l 4.59jaren ( [

t j i j j j l l l l l l ;

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' l' ATTACHAENT VI Physical Property Data on Last-a-Foam

FR-3700, seven pages

SNC i

Sierra Nuclear Corporation 1 + (2d - 1)~/ x V' '

~

d 2

Q =

(2d - l)_

_ <gT>

W 1+

d*

where:

d

- allowable ductility ratio for the controlling failure mechanism.

W

- static weight of the missile T

- natural period of the structure g

- gravity acceleration V

- impact velocity This formula is recommended by NRC in NUREG 0800 for missile protection barrier design. It accounts for the dynamic nature of the impacted structure and for the inelastic energy absorption of the structure. This equivalent static force Q can be compared directly to the structure load capacity.

l l

lo Client.' Project: PL _ di Revision l Prepared l Date l Checked l Date l Sheet i

Subject:

Evaluation of the Cask Droo Inside 0

lOpH v.6 fu s Ac. I'/::/% l d'

l Troian Fuel Bui! ding I

l l

l Calculation Number: PGE01-10.02.03 - 18 l

l l

l

SNC Sierra Nudear Corporation (H-h)

W(H - h + A) = [f,, A + W,,)A,

thus, A = IV f,, A - W + W,,

where:

W

- weight of the Transfer Cask H

- drop total height h

- ILP height f,,

- crushing strength of the foam A

- crushed area A

- depth of crush W,

- weight of water (if any) displaced by the Transfer Cask Deceleration is calculated by balancing the forces:

IW' (f,, A

W,,)

a = f,, A e W, thus, a=g

<g)

W where:

g

- gravity acceleration a

- deceleration of Transfer Cask The impact velocity is:

4 V = y/2g (H-h) 3 Transfer Cask H

b)

Niethodology for Hard Surface Impacts For the case where no ILP is used, the Williamson-Ah7 ormula [Ref.14] is used to f

develop the equivalent static force.

I I

l l

15 I Client / Project: PGE-01 Resiston l Preparec l Date l Checked l Date l Sheet Y

Subject:

Evaluation of the Cask Drop Inside 0

l 3u4gx;,1 6,sc, j"/u /% l Troian Fuel Building l

l l

l

,9 Calculation Number: PGE01-10.02.03 - 18 l

l l

l

SNC

.i a

Sierra Nuclear Corporation.

ATPC M E W M where c p q _ o 3 5 - 9 (o R,n = plastic resistance force A = yield displacement at the location ofimpact load y

= ductility ofimpacted slab R, a dead and live load S = yield displacement at location of concentrated dead and live load or y

= 1/2 maximum yield displacement for distributed live and dead load So = dead and live load displacement at the location of concentrated load or

= 1/2 maximum dead and live load displacement for distributed live and dead load For the postulated cask drop into the Cask Loading Pit BC-TOP-9A methodology is used to estimate the velocity at which the cask strikes the submerged impact limiter at the bottom of the pit.

4.4 Estimating Drop Load Two methodologies are used to determine impact loads. Impact loads are required when checking shear capacities in beams and slabs and when designing an impact limiter.

The first methodology is appropriate for an impact limiter (ILP), and is based on the absorption of the drop energy by ILP foam. The second is for a hard impact without ILP, and uses the equivalent static force. In both cases the inelastic response of the structure is anticipated.

a) Methodology with ILP This analysis is based on the assumption that the ILP foam crushes under the cask and absorbs the drop energy. The foam dynamic strength multiplied by the crushed area is the drop load to the structure. The force and energy balances allow calculation of the crush depth.

In parallel with the above. the structure load capacity is calculated. The structure capacity must be found for both flexure and shear to determine which controls the design. Based on this result, the allowable ductility ratio is selected.

The required pad height is calculated. Per vendor information. the foam provides constant dynamic strength as long as it is crushed less than 50% of the total thickness.

Therefore, the crush depth must be kept under 50% of the ILP height.

Crush depth is determined by balancing the drop energy and work produced by the foam.

I I

l l

14 ClienuProject: PGE-01 Revision l Prepared j Date l Checked l Date Sheet

Subject:

Evaluation of the Cask Droo Inside 0

I Qrd.--; y-/ Wl Mc_,

lVn/s N

Trojan Fuel Building l

C l

l l

,,1 Calculation Number:_P_GE01-10.02.03 - 18 l

l l

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3:

l l

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ATTACHMENT V Pages 14-16 from Calculation PGE 01-10.02.03-18 l

l

I Attachment IV to CPY-035-96 Impact Limiter Properties

(

Impact Limiter Properties Description Drop Heig. -

P, h

Length Width Density (psi)

(in)

(in.)

(Ibs/ft')

Drop into Cask Loading Pit From Elevation 865 34 132 96 17 93'8" Tipover At Cask Loading Pit NA 1454 24 216 18 24 l

Drop Into Cask Wash Pit From Elevation 115 125 100 95 5

93'6" Tipover At Cask Wash Pit NA 454 30 108 18 12 Drop Onto Cask Wash Pit 3" above NA NA NA NA NA Wall Elevation 93' Drop onto Shear Wall 3" above NA NA NA NA NA Between Cask Wash Pit and Elevation 93' Iloistway Drop Onto Floor Slab 5" above 48 3.5 108 108 4

Between Load and Wash Pits Elevation 93' Drop Into lloistway From Elevation 675 65 216 216 14.7 93'6"

.t o

l t

i l

i ATTACHMENT IV Impact Limiter Properties l

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t ts. e s k a Pt> Q s fs /s 5

/f 3/2/93 RPC 93-003, REV.1

SUMMARY

OF DOSE RESULTS SFP FUEL HANDLING ACCIDENT MAXIMUM SITE BOUNDARY DOSES (mrem)

TOTAL WHOLE BODY SKIN THYROID All Isotopes Except I-129:

5.398E-1 4.554E+1 5.698E-1 I-129 Dose Contribution :

8.216E-6 2.221E-5 4.901E-2 TOTAL (mrem):

5.398E-1 4.554E+1 6.188E-1 DOSES AT CB-1 INTAKE (MREM)

TOTAL WHOLE BODY SKIN THYROID All Isotopes Except I-129:

6.512E-1 5.493E+ 1 6.874E-1 I-129 Dose Contribution :

9.911E-6 2.680E-6 5.913E-2 TOTAL (mrem):

6.512E-1 5.493E+ 1 7.465E-1 DOSES AT CB-2 INTAKE (MREM)

TOTAL WHOLE BODY SKIN THYROID All Isotopes Except I-129:

1.060E+0 8.942E+ 1 1.119E+0 I-129 Dose Contribution :

1.613E-5 4.362E-5 9.624E-2 TOTAL (mrem):

1.060E+0 8.942E+ 1 1.215E+0

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A1rACMMEMT W

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GENERAL PLASTICS m gx 939,34 MANUFACTURING COMPANY FAX (206) 473 5104 4910 BURLINGTON WAY/P O. BOX 9097 TOLL FREE FROM TACOMA, WASHINGTON 98409 Seanle-Renton Area (206) 473-5000 C06) 623-2795

- w.wr n m.-

PHYSICAL PROPERTY DATA LAST-A-FOAM

FR-3700 Ceneral Plastics Manuf acturing Company's LAST-A-FOAM 8 FR-3700 is a rigid, unicellular polyurethane foam that can be supplied in any density from 3 to 30 lbs./f t. '

The standard sheet size for FR-3700 foam in the 3 to 6 lbs./ft.' density range is 48" *.25" x 96" *.50" with a thickness tolerance of *.015" on sheets up to 2" thick and *.030" on sheets over 2" thick. The standard sheet size for FR-3700 foam in the 8 to 30 lbs./ft.' density range is 18"

.10" x 100"

.50" with a thickness tolerance of i.005" on sheets up to 1" thick, *.015" on sheets 1" to 2"

thick, and *.030" on sheets over 2" thick.

STRENGTH:

LAST-A-FOAM 8 FR-3700 series rigid polyurethane foams exhibit excellent strength to weight ratio because of the high strength polymer and cellular structure.

The strengths shown g the accompanying graphs are nominal ULTIMATE values at which the foam fails to support a higher load. Appropriate safety factors should be incorporated into all designs for structural applications.

DURABILITY:

General Plastics' LAST-A-FOAM 8 rigid polyurethane foam is a very stable material which will not corrode, sustain fungus or attract rodents or insects.

LAST-A-FOAM

has a high chemical resistance and is unaffected by most chemicals and solvents, except for some of the chlorinated solvents.

USES:

LAST-A-FOAM 8 FR-3700 series, rigid polyurethane foan in the 10 to 25 lbs./ft.'

density range is used extensively as high strength framing, or close-out strips, around the perimeter, and at the fastening points in honeycomb core or other similar structural panels.

The LAST-A-FOAM

FR-3700 foam in the 3 to 8 lbs. /f t. ' density range is used extensively as the entire core material for many structural insulated panels.

INSTALLATION:

One method for fastening hardware, or other articles to the foam is to pot a threaded receptacle in the foam.

Threaded fasteners may then be installed and removed many times without damaging the foam.

CLOSED CELL CONTENT:

Tested per ASTM D-2856, Procedure B.

95% minimum @ 3 lbs./ft.' density 987. minimum @ 25 lbs./f t. ' density LAST-A-FOAM 8 has low moisture vapor permeability and high resistance to water absorption because of its closed unicellular structure.

LAST-A-FOAM

Will not crack, split or swell when exposed to moisture.

(1)

MANUFACTURERS AND MOLDERS OF LAST.A. FOAM 8 HIGH DENSITY RlGID AND FLEXIBLE POLYURETHANE FOAMS AND FABRICATORS OF PLASTIC SHEETS FOR AIRCRAFT. INDLSTRIAL. CONSTRL'CTION. MARINE. NUCLEAR. SHIPPING AND MODELING

.t

'd GENERAL PLASTICS m gx 333334 MANUFACTURING COMPANY FAX (206) 473-S104 4910 BURLINGTON WAY/P.O. BOX 9097

'IDLL FREE FROM TACOMA. WASHINGTON 98409 Seattle Renton Area

-,an.wurm.-

(206) 473 5000 (266) 623-2795 LAST-A-FOAM

FR-3700 COEFFICIENT OF LINEAR THERMAL EXPANSION:

LAST-A-FOAM

FR-3700 exhibits a thermal expansion or contraction coef ficient in the range of 3.5 x 10-5 in/in/*F to 5.0 x 10-5 in/in/*F over the temperature range of -310 F to +200*F.

HEAT DISTCRTION:

This test is used to determine the suitability of LAST-A-FOAM

FR-3700 for use as core material in laminated structural panels, using temperatures up to 250*F and vacuum pressure.

Samples of FR-3700,.50" thick are placed on a metal plate under a vacuum bag, with 20 inches of mercury minimum vacuum.

These samples are then put into an oven (preheated to 250*F) for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The samples are then removed and allowed to cool while under vacuum.

The difference in thickness of the samples before and after the test is measured.

LAST-A-FOAM 8 THICKNESS CHANCE ( 7. )

FR-3704

- 1.15 FR-3706

- 0.90 FR-3710

- 0.75 FR-3715

- 0.34 FR-3720 0.25 THERMAL CONDUCTIVITY:

TESTED PER ASTM C-177 AT A MEAN TEMPERATURE OF 75*F.

LAST-A-FOAM 8 k FACTOR R VALUE MATERIAL BTU /hr-ft." *F/in.

Hrs.-ft." *F/ BTU in.

l l

FR-3704 0.184 5.44 FR-3706 0.208 4.81 FR-3710 0.257 3.89 FR-3718 0.355 2.82 FR-3720 0.379 2.64 FR-3725 0.440 2.27 ELECTRICAL PROPERTIES:

LAST-A-FOAM

rigid polyurethane foam offers high electrical resistivity at low weight.

The dielectric constant of LAST-A-FOAM 8 FR-3700 varies linearly with density from 1.05 at 3.0 lbs./ft.' to 1.4 at 20.0 lbs./ft.' when tested at 1.0 MH -

z The change in dielectric constant is negligible within the temperature range of

-50*F through 300 F (2)

MANUFACTURERS AND MOLDERS OF LAST.A FOAM

HIGH DENSITY RIGID AND FLEXIBLE POLYURETHANE FOAMS AND FABRICATORS OF PLASTIC SHEETS FOR AIRCRAFT, INDUSTRIAL. CONSTRUCTION. MARINE. NUCLEAR, SHIPPING AND MODELING

1 d

.g GENERAL PLASTICS m.EX 989134 MANUFACTURING COMPANY FAX (206) 473-5104 4910 BURLINGTON WAY/P.O. BOX 9097 TOLL FREE FROM TACOMA. WASHINGTON 98409 Seattle Renton Area (206) 473-5000 C06) 623-2795

-, m v.m n,*-

LAST-A-FOAM 8 FR-3700 i

FIRE SAFETY:

LAST-A-FOAM 8 rigid polyurethane foam is an organic material which will burn in the presence of sufficient heat and oxygen.

The ASTM D-1692-74 and Federal Aviation Regulation (FAR) 25.853 flame test methods are comparative tests conducted j

under specific laboratory conditions. The sole purpose of these tests is to establish the relative burning characteristics of foam plastics.

'Ihe re sult s of these te st s are not to be considered, or used as fire hazard classifications, and correlation with flammability of LAST-A-FOAM

under actual use conditions is neither intended nor implied.

In some circumstances, if the LAST-A-FOAM

is allowed to remain exposed and unprotected, a fire could spread rapidly under actual fire conditions, creating dense smoke and intense and immediate heat.

LAST-A-FOAM

rigid polyurethane foams have chemical additives which help reduce their flammability and burning rate, and will produce minimum fire contribution in specific building assemblies when properly protected.

It is recommended that adequate automatic sprinklers be incorp-orated into building construction wherever possible.

In order to provide better occupancy protection, and if automatic sprinklers are not feasible, a plaster coating, cement asbestos, gypsum sheetrock paneling, or metal sheeting is necessary to cover exposed foam surfaces.

The ASTM D-1692-74 and FAR 25.853 flame test methods and test results are listed below:

ASTM D-1692-74 FLAME RESISTANCE TEST:

In this test a.5" thick x 2" vide x 6" long foam sample is placed in a horizontal position.

The 2" end of the sample is exposed to the flame of a Bunsen burner with a wing tip attachment for 60 seconds.

The time to flame extinguishment after start of ignition and the length of the sample which was burned are recorded. Average test values for LAST-A-FOAM

FR-3700 foams are given below:

EXTINGUISHMENT, AFTER START LAST-A-FOAMe OF IGNITION, TIME, SECONDS BURN DISTANCE, INCHES FR-3704 77.9 1.65 FR-3706 75.0 1.45 FR-3710 69.0 1.40 FR-3718 65.0 1.30 FR-3720 61.0 1.30 (3)

MANUFACTURERS AND MOLDERS OF LAST A FOAM 8 HIGH DENSITY RIGID AND FLEXIBLE POLYURETHANE FOAMS AND FABRICATORS OF PLASTIC SHEETS FOR AIRCRAFT. INDUSTRIAL. CONSTRUCTION. M ARINE. NUCLEAR. SHIPPING AND MODELING

.i c

GENERAL PLASTICS u tgx 939,34 MANUFACTURING COMPANY FAX (206) 473-5104 l

4910 BURUNGTON WAY/P.O. BOX 9097 Tot 1 FREE FROM TACOMA. WASHINGTON 98409 Seattle Renton Area (206) 473-5000 (206) 623 2795

-,, m m.m n,u.-

LAST-A-FOAM FR-3700

)

I FEDERAL AVIATION RECULATION (FAR) 25.853, FLAME RESISTANCE TEST:

In this test a.5" thick x 3" wide x 12" long foam sample is mounted in a vertical position.

The lower (.5" x 3") end is exposed to a Bunsen burner having a 1.5" high flame. The time of exposure is 12 or 60 seconds. The time to flame extinguish-ment after removal of the Bunsen burner flame and length of the sample which is burned are recorded. Average test values for LAST-A-FOAM 8 FR-3700 foams are given below:

FAR 25.853 (b) 12 SECOND IGNITION LAST-A-FOAM 8 EXTINCUISHMENT, TIME SECONDS BURN DISTANCE, INCHES FR-3704 0.5 5.7 FR-3706 3.0 5.2 FR-3710 2.5 3.8 FR-3718 6.1 2.7 FR-3720 5.5 2.9 FAR 25.853 (a) 60 SECOND IGNITION LAS'f-A-FOAM

EXTINGUISHMENT, TIME SECONDS BURN DISTANCE, INCHES FR-3704 0.7 5.6 FR-3706 0.8 5.4 FR-3710 4.6 FR-3718 4.5 FR-3720 4.7 (4)

MANUFACTURERS AND MOLDERS OF LAST-A FOAM' HIGH DENSITY RIGID AND FLEXIBLE POLYURETHANE FOAMS AND FABRICATORS OF PLASTIC SHEETS FOR AIRCRAFT, INDUSTRIAL. CONSTRUCTION, MARINE, NUCLEAR SHIPPING AND MODELING

i GENERAL PLASTICS m 333334 MANUFACTURING COMPANY FAX (206) 473 5104 4910 BURLINGTON WAY/P.O. BOX 9097 TOLL FREE FROM TACOMA, WASHINGTON 98409 Seattle Renton Area (206) 473 5000 (206) 623 2795

-,, m m..m a.. -

LAST-A-FOAM? FR-3700 PHYSICAL PROPERTY TEST DATA June 1982 95% LIMITS LINE OF OF PROPERTY TESTED BEST FIT CONFIDENCE n*

c**

TEST METHOD 1

COMPRESSIVE STRENGTH (psi)

Parallel to rise 75' F 8.745D.6154 5.75%

6 5.00%

ASTM D-1621 1

@ 250' F 4.537D.6050 1

2 11.30%

6 9.82%

Perpendicular to rise 75' F 3.456D.91 8 1

2 2.27%

6 1.98%

ASTM D-1621 4

l

@ 250' F 2.966D

2 10.40%

6 9.06%

COMPRESSIVE MODULUS (psi)

Parallel to rise

@ 75* F 276.2D.5413 5.90%

6 5.13%

ASTM D-1621 1

@ 250' F 144.0D.6396 1

2 22.06%

6 19.18%

Perpendicular to rise

@ 75' F 74.66D.9995 1

2 9.73%

6 8.46%

ASTM D-1621

@ 250',F 93.03D

2 16.82%

6 14.62%

Wh:n D = Foam density in Ibs./ft.8 n - number of densities tested 1 - 5 specimens each density.

- o.

I[X - I)'

j 6

n 1

I MANUFACTURERS AND MOLDERS OF LAST A. FOAM

HlGH DENSITY RIGlD AND FLEXIBLE POLYURETHANE FOAMS AND FABRICATORS OF PLASTIC SHEETS FOR AIRCRAFT. INDUSTRIAL CONSTRUCTION MARINE. NUCLEAR. SHIPPING AND MODELING I

.i GENERAL PLASTICS

.it a 933334 MANUFACTURING COMPANY FAX (206) 473 5104 4910 BURLINGTON WAY/P.O. BOX 9097 TOLL FRFI FROM TACOHA, WASHINGTON 98409 Seattle Renton Area (206) 473 5000 (206) 623 2795

-, m,..w o r m.. -

LAST-A-FOAM? FR-3700 PHYSICAL PROPERTY TEST DATA June 1982 l

95% LIMITS LINE OF OF PROPERTY TESTED BEST FIT CONFIDENCE n*

c**

TEST METHOD TENSILE STRENGTH (psi)

Parallel to rise 29.68D.1496 3.45%

6 3.00%

ASTM D-1623 1

Type "A" Perpendicular to rise 14.74D.3852 1

1 6.62%

6 5.76%

Specimens TENSILE MODULUS (psi)

Parallel to rise 922.4D '1017 1

7.88%

5 5.68%

ASTM D-1623 Type "B" Perpendicular to rise 286.5D.5324 14.81%

6 12.88%

Specimens 1

1 SHEAR STRENGTH (psi)

Parallel to rise 7.530D.5443 5.99%

5 4.32%

ASTM C-273 i

1 Compression Perpendicular to rise 11.60D.3754 1

2 1.80%

3 0.b2%

Shear SHEAR MODULUS (psi)

Parallel to rise 40.53D.8555 8.30%

5 5.98%

ASTM C-273 1

Compression Perpendicular to rise 133.0D.4209 7.29%

6 6.34%

Shear 1

FLEXURAL STRENGTH (psi) 660 Parallel to rise 8.287D.

9.56%

6 8.32%

ASTM D-790 Method 1-A i

Perpendicular to rise 21.17D.3614 1

1 3.66%

4 1.99%

i FLEXURAL MODULUS (psi)

Parallel to rise 151.2D *O t 11.48%

6 9.98%

ASTM D-790 i

l Perpendicular to rise 545.8D

1 19.78%

4 10.56%

I MANUFACTURERS AND MOLDERS OF LAST A FOAM

HIGH DENSITY RIGlD AND FLEXIBLE POLYURETHANE FOAMS AND FABRICATORS OF PLASTIC SHEETS FOR AIRCRAFT. INDUSTRIAL CONSTRUCTION. MARINE. NUCLEAR, SHIPPING AND MODELING

.g i

GENERAL PLASTICS mEX 989134 MANUFACTURING COMPANY FAX (206) 473 5104 4910 BURLINGTON WAY/P.O BOX 9097 TOLL FREE FROM TACOMA. WASHINGTON 98409 Seattle Renton Area i +.- (206) 473 3000 006) 623-2795

-m uw m

LAST-A-FOAM FR-3700

'000

.s mo

._.2._.

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" =

1 Cxo

.bxg f As4.

= n =

COMPRESSIVE STRENGTH AT 75'F

'I

+-

23 lANN<--* N\\1

+-,

:n=:=

am E

AF AF

-i 1

5 1

DENSITY ccm b

Cirecticn of foam rise LBS./FT.8 STRENGTE STRENGTH I

men psi psi

-a E

m

==

to stress L to stress E ll 3

~

~~

3.0 51.6 28.4 9 l fa 4.0 82.1 49.3 L

+.

.te

. =--4 -4a 5.0 117.7 75.7

~.

[

6.0 158.0 107.4 7.0 202.7 144.3

.f C"

f..

i 8.0 251.5 186.4 iOco

'/ -

9.0 304.3 233.7 O.4

- h --- -

-i -+- e

,'/

10.0 360.7 286.0 I

rm m

, a n

-I

i

~w 11.0 420.8 343.4 c

El.. __, _,,

_-L__... 7,. y' __=_;- 7,;-f6-g 12.0 484.2 405.7 600 v

._.m

.- i_

=

g. _ __. - ;;

~ 12-+~ ~~L? ~~

13.0 551.1 473.0 5

4

~

II M.' _ "_,.--4 = _

-/-/

. - 14.0 621.2 545.3 cm

=

l---

l+.j'"f4,

? 15.0 694.4 622.4 5

i g

+. -

200 m n _,_.

...._..__.4__,._,_

16.0 770.7 704.4 E

>N [fhjz!.~~-'-

Y Y.

O 17.0 850.0 791.3 5

W-rv-r ;

m m

.r-r:

18.0 932.2 882.9 J /+-/ --; l l

19.0 1,017.3 979.4 m

i 1

E 7-4F4 '

20.0 1,105.2 1,080.7 2

S'.t I l O

t I

I

'-E 21.0 1,195.8 1,186.7 I

L w

~e

- e I

22.0 1,289.2 1,297.4 2

YOf C 23.0 1,385.1 1,412.8 300 I/

24.0 1,483.7 1,533.0 E

C' i-

- f /

?-

25.0 1,584.9 1,657.8 INO/~.a/.. -.-.. g' =.. -

=

I 26.0 1,688.5 1,787.4 I

.0

<l;-c_-

w=

_2 T[:t ---f -

~~ 27.0 1,794.7 1,921.5

.5

~

=

28.0 1,903.3 2,060.3 f

4 29.0 2,014.3 2,203.7 E

+'~

m i~

30.0 2,127.7 2,351.8 21 l.j=r---

.I-+ 5:.

w=1

~

c2-- --

. = = =

._._. a=_

--n--

.. an

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20 i~.

n

~-

t I

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DENSITY LBS./FT.s i,

on ii i I i I'll lih ll.

Ili i!.!

i i

?

3 4

6 6

?

4 9 10

'S 2C 30 40 40 80 100 ll Indicates stress is parallel to direction of foam rise.

1 Indicates stress is perpendicular to direction of foam rise

1

' n:

l 1

4 l

l 4

ATTACHMENT VII Trojan Calculation RPC 96-009 i

~.. -.

ANMEuT-M s- :.a ops-- oss-9 g QA RECORD WHEN COMPLETED ***

l RC 2 1 _2_

l CFP GEN ENGR 7-6 1

00R, Admin Services Lettwr Number NA System Number NA Number of Pages 1

Document Date Calc. Reference a TROJAN CALCULATION COVER SHEET Sheet / Cont'd on Sheet d Title b/lC OJ4WMV 00525 -Nudk/A6 TSPSl~ llA5lEb lNltiiAl buMh/

L,yg

/

// '

((C fd"OOf Calculation No.

TrojanNuclearPfant M/A Supersedes Calculation No. A/

Structur EW N 55<//E /45[ I/A/C Quality-Related b /No System

$fS5 hSb' V inal Interim F

Status:

Component References (PMR/DPMR, SPEER, MR, PSC, etc.)

Has Been Changed by Or Revision has been Responsible Affected Identify Change Deferred by (Identify Supervisor /Date Document No. Vehicle: (MR, DPMR, DCP, PCF, Memo, CTL, etc.)

(Deferrals Only)

SPEER, PSC, etc.)

l l

Calculation Objective 6 ENj'fllW56 &l5&

W5 l0 C

n%

/udC

/7 kibuAt S/

C hi -/o ht A sr 2 sr MA 1

l Revision Description l

Rev.

No.

Preparer Date Verified By Late Approved By Date 0

$&, A.

f/b//ft W

5 2ElSi l

i TPP 18-9 Revision 2 Page 1 of 1 Page 11 of 11

_-m

.i e'

- - QA RECORD WHEN COMPLETED ***

[dC CALCULATION / CMR VERIFICATION CHECKLIST Calculation No.

' Of Verified By Date Dial /f4' Title A

G~

0 t/SE 55 5

i M SN / b/Ns

/

// /

Yes No N/A 1.

Are analytical methods appropriate?

I O

O 2.

Can a person technically qualified in

(

0 0

the subject review and understand the calculation without recourse to the originator?

3.

Is the calculation mathematically ET O

O accurate?

4.

Are the calculation objectives and 7

O O

design inputs (including source) clearly defined?

5.

Do calculational parameters comply with 7

O O

design criteria / dimensions?

6.

Does the calculation reference or list O'

O O

major equation sources?

7.

Is input data appropriate?

D' O

O 8.

Are assumptions properly referenced

[

O O

and correct?

9.

Was am applicable and validated computer

[

O O

program used?

10.

If applicable, were program error O

O 3'

notices reviewed?

~

11.

If an unverified program was used, is an O

O c'

alternate check calculation completed?

12.

Do the output results seem reasonable?

[

O C

CFP:

GEN. ENGR 3 OOR:

Engineering Page 1 of 1 NPEP 200-11 Revision 3 Page 13 of 15 4

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l.48E-02 6 86E-04 6.70E 1.058E-02 1.270E-01 '

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II e y [ PORTLAND GENERAL ELECTRIC CALCULATION SHEET ralculation No. (("dof Revision 6 Sheet M (( sarer fJ/nm Date Sydggg o Venfier ~T) 4 _ E Date i _'c,/2gj', g Activity Released to Fuel Building Atmosphere Total Release Activny Activity Activity Fraction Concentration Released Basket Atmosphere Nuchde (Curies) (Cunes) (Ci/m3) (Cunes) H-3 5.37E-02 8.05E-03 ' l.52E 2.60E C-14 7 67E-02 1.ISE-02

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3.03E 519E-02 '

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6. I 4E-01 9.22E-02.

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f/)0 VG- )ff .I udi) %[ 3of[ q/ SITE BOUNDARY DOSE FROM RUPTURE OF llELIUM SUPPLY LINE - VENTILATION ISOLATED g Nuclide CEDE WB CDE Thy CDE RBM CDE Lung CDE BS CDE Gona CDE Breast CDE Remai '7

/2/[9c (mrem)

(mrem) (mrem) (mrem) (mrem) (mrem) (mrem) (mrem) 11-3 3.88E-06 ' 3.88E 3.88E 3.88E 3.88E-06 3.8SE 3.88E 3.88E-06 C-14 1.67E-04 ' l.67E-04 1.67E 1.67E 1.67E-04 ' l.67E-04 ' l.67E 1.67E-04 Sh-125 1.36E 1.34E-03 2.68E 8.95 E 1.13E 1.48E 1.72E 5.99E-03 Cc-144 1.04E 1.94E 2.76E 8.18E 4.69E 2.47E 2.04 E-04 1.06E-02 Mn-54 3.50E 1.43 E 3.2 W 1.28E 4.94E 1.71 E-04 ' l.76E 4.04E-04 Fe-55 4.96E 3. 71 E 3.52E 7.23E-02 3.50E-02 ' 3.58E-02 3.47E 8.26E-02 Co-60 1.48E+01 - 4.04E+00 ' 4.29E + 00 - 8.64E + 01 - 3.37E+00 - 1.19E + 00 - 4.59E+00 8.97E+00 - Ni-63 8.70E 8.70E 8.70E 1.58E-0! - 8.70E 8.70E 8.70E 8.70E-02 Sr-90 8.30E 6.23E-03 ' 7.91 E-Oi - 6.7hE+ 00 - 1.72E+00 - 6.23E 6.23E-03 ' l.35 E-02 Pu-238 2.27E+ 01 2.06E 3.25 E+ 01 - 6.82E + 01 - 4.06E+02 ' 6.01 E+00 - 2.14 E 1.50E + 01 Pu-239/240 2.77E +01 - 2.16E 4.04 E+ 01 - 7.76E + 01-5.05E+02 7.63E+00 - 2.21 E 1.81 E+01 Pu-241 2.87E + 01 - 1.60E 4.32E+ 01 - 4.1 l E+ 01 5.40E+ 02 8.77E+ 00 - 3.93 E 1.69E+ 01 Cm-242 1.45E-04 ' 2.92E-0T 1.21 E-04 4.82E 1.51 E 1.77E 2.93E 7.62E-05 Cm-243 7.65E+ 00

3.53E 1.09E+ 01 1.79E+ 00- 1.36E + 02 - 1.91 E + 00 -

5.80E-OL 5.31 E+ 00 Cm-244 5.86E+ 00-8.83E 8.20E+ 00 - 1.69E+ 00 - 1.02E+ 02 - 1.39E+ 00 - 9.09E 4.18E+ 00 Am-241 3.41 E + 01-4.54E-04 4.94 E+ 01 - 5.22E+00 - .i6E+02 ' 9.22E+ 00 - 7.58E 2.22E+ 01 Pu-242 1.34 E 1.06E 1.95E 3.71E-01 2.43E

00-3.65 E-02 1.14 E 8.68E-02 1.43E+ 02 - 4.17E+00 - 1.90E+ 02 2.89E+ 02 2.31 E+ 03 3.63E + 01 4.73E + 00 9.10E+ 01 4

/ VVCD. igw& A Skd4/ SITE BOUNDARY DOSE FROh! RUPTURE OF llELIUh1 SUPPLY I ME - VENTILATION RUNNING [h,- 5/,1g/f(, 'Q Nuclide CEDE WB CDEThy CDE RBh1 CDE Lung CDE BS CDE Gona CDE Breast CDE Remai f (mrem) (mrem) (mrem) (mrem) (mrem) (mrem) (mrem) (mrem) 11-3 2.36E-05 2.36E 2.36E 2 36E 2.36E-05 2.36E 2.36E 2.36E-05 C-14 3.30E-07 3.30E 3.30E 3.30E-07 3.30E 3.30E-07 ' 3.30E-07 3.30E ' Sh-125 2.69E 2.65E 5.30E-06' 1.77E 2.23 E 2.94E 3.40E-06 1.19E-05 Cc-144 2.07E 3.85E 5.47E-06 1.62E-04 ' 9.29E 4.89E 4.03 E 2.llE-05 hin-54 6.93E 2.83E 6.35E-07 2.54E 9.79E 3.38E 3.50E 7.99E-07 Fe-55 9.82E 7.34E 6.97E 1.43E 6.94 E 7.08E 6.86E 1.64E-04 Co-60 2.93 E 8.00E 8.50E 1.71E 6.68E 2.35E 9.10E-03 1.78E-02 Ni.63 1.72E 1.72E 1.72E-04 3.12E 1.72E 1.72E 1.72E 1.72E-04 Sr-90 1.64 E 1.23 E 1.57E 1.34E 3.40E 1.23 E-05 ' l.23E 2.68E Pu-238 4.49E 4.08E-07 6.43E 1.35E 8.05E-01 1.19E 4.24E 2.98E Pu-239/240 5.50E 4.28E 8.01 E 1.54E-01 ' l.00E + 00 - 1.51 E 4.37E-07 3.59E-02 Pu-241 5.69E 3.16E 8.55E-02 8.14 E-02 ' l.07E + 00 - 1.74 E-02 7.79E 3.34 E Cm-242 2.88E-07 ' 5.79E-11 2.40E 9.55 E-07 ' 3.00E 3.51 E 5.81 E-11 1.51 E-07 Cm-243 1.52E 7.00E 2.16E-02 3.54E 2.69E-01 ' 3.78E 1.15E 1.05E-02 Cm-244 1.16E 1.75E 1.62E-02 3.34E 2.03 E 2.75E 1.80E-07 8.28E-03 ~ Am-241 6.75E 8.99E-07 9.78E-02 1.03 E 1.22E+ 00 - 1.83E 1.50E-06, 4.40E Pu-242 2.66E-04 2.10E-09 ' 3.85E-04. 7.35E 4.81 E 7.23 E 2.26E-09 ' l.72E-04 2.82E-01 8.29E-03 3.76E-01 5.73E-01 4.58E + 00 7.19E-02 9.39E-03 1.80E-01

[/D f6- 'b /f $XC f-DOSE CONVERSION FACTO S - SITE BOUNDARY DOSE RUPTURE OF IIELIUM SUPPLY (/MAttn,k'b Activity k7 Nuclide DCF DCF DCF DCF DCF DCF DCF DCF Released Whole Bod Thyroid RBM Lung Bone Surf. Gonad Breast Remainder (Curies) Il-3 6.40E-02 6.40E 6.40E 6.40E 6.40E 6.40E-02 6.40E 6.40E-02 2.60E. C-14 2.09E+00 ' 2.09E+00 - 2.09E+00 2.09E+00- 2.09E+ 00 - 2.09E+ 00. 2.09E+00 2.09E4 00. 3.71 E-03 Sb-125 1.22E+ 01 1.20E + 00 - 2.40E+ 00 8.03E4 01 - 1.01E + 01 1.33E +00 1.54 E + 00 5.37E+ 00

5.19E Cc-144 3.74 E + 00 6.96E + 00 - 4 RRE+ 01 2.93 E+ 03-1.68E+ 02 8.84 E 7.29E+ 00 - 3.81 E + 02 -

1.30E-03 Mn-54 6.70E + 00- 2 4 E4 00 6.14E FOO - 2 46E+01 - 9.47E+ 00 3.27E + 00 - 3.38E + 00 - 7.73 E + 00-2.43E Fe-55 2.69E + 00' 2.01 E+ 00 - 1.91 E+ 00 3.92in ""- 1.90E+ 00 1.94 E + 00 - 1.88E + 00 - 4.48E+ 00 - 8.58E Co-60 2.19E +0'. - 5.99E+ 01 6.36E+ 01 - 1.28E + 03 5.uud+ 01 1.76E+01 6.81 E + 01 - 1.33 E + 02 - 3.14 E + 00 - Ni-63 6.29E+JO 6.29E4 00 - 6.29E + 00 1.14E+01 - 6.29E + 00 6.29E+ 00 6.29E + 00 - 6.29E + 00 - 6.44E Sr-90 1.30E &O3 9.77E+00 1.24 E+ 03 - 1.06E + 04-2.69E + 03 9.77E + 00 9.77E+ 00 - 2.12E+ 01 - 2.97E-02 Pu-238 3.92'.i+ 05 - 3.56E + 00 5.62E + 05-1.18E + 06 - 7.03 E + 06 1.04 E + 05 - 3.70E + 00 - 2.60E + 05 - 2.69E Pu-239/240 4.2 /E + 05 - 3.34 E+ 00 - 6.25 E + 05 - 1.20E+ 06-7.81 E+ 06 1.18E+05 3.41 E + 00 - 2.80E+05 - 3.01 E Pu-241 8.'.5 E + 03 - 4.59E 1.24 E+ 04 - 1.18E+ 04 - 1.55E+ 05 2.52E+ 03 - 1.13E 4.85 E + 03 - 1.62E-01 Cm-242 173E+04 - 3.48E+00- 1.44 E + 04 ~ 5.74 E+ 04 - 1.80E+05 2.1 l E+ 03 - 3.49E + 00 9.07E+ 03 - 3.91 E Cm-243 '. 07E+05-1.42E+01 4.37E+05 - 7.18E+04 5.44E+06 7.66E + 04 - 2.33E+ 01 - 2.13E +05 - 1.16E Cm-244 L48E+ 05 3.74 E + 00 - 3.47E+ 05 ' 7.14E+ 04 - 4.33E+06 5.88E+ 04 - 3.85 E + 00 - 1.77E+ 05 1.10E-03 Am-241 4.44E+05 - 5.92E+00- 6.44E+ 05 6.81 E + 04 - 8.03E+ 06 1.20E + 05 - 9.88E+ 00 - 2.89E + 05 ' 3.57E Pu-242 4.11 E + 05 - 3.25 E + 00- 5.96E + 05 - 1.14 E+ 06 7.44E + 06 1.12E+ 05 - 3.50E+ 00 - 2.66E+05 - 1.52E.05 -

-. - ~.. -. -. - .. -.. - - - - -.... - -. ~ - t: /[ ~ ~ 82 4712 11 79 cf Sheets s.4o.aor OJo No. O ['OOf GENERAL COMPUTATION SHEET O rsi. Projeci/ Job Sublect W f OS S " / Y / 5.5 4 0 0 ll 5 MM /o. 5/4&, ' cau.sy7/</ o.. ?% r A 8 o,'i, g n o sy i i i ~ l i l i i! 4 i l l I 8 ! I i f f I l ! i i i T l i i A / /- i i li/ r' I' i i i / 'f V V O / ' o dMDd" DMuMbVlM70N a i i r/ // t i \\ i : i I f l l l I l l l i I i r e 8 i i ! ; l l i l I l i l I ! ! i i f 3 4 i i ! l I ! l

I i

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i i

I l l 2

I i

s I i l l I l j I l i j l l ! i I ! l ! l l l I ! i I f 8 i l l I i l ! ! l I I ! ! i l t l i i i i l I I i I i! i i l i ! ! ! : I l l i l i i l l 1 i i i i : l I i i : i i i i I ! I i l l ! lI ! ! I I i i j I I I I i I I i a ! i ' i l i i i l i i Ii I I l I ! I i l l I I i i l ! l l l1 i ! l l l i ! ! l ! ! ! I I I i ; ! I I i l 1 i I ! ! I I I l I f i I : I I I i i i ! l i l l I i i ! I i l l I I I ii l i I I I i ! i i ! l I i l Ii l I l l l l l I I i i l i i I I i i i I i l I i l l I I I I I I I I I i j i i l I ! I i i ! I i ! i l l l ! ! II II I I l l I I II I i i t i i ! 4 i i i i ! l I l l l l l i I l l l 1I i l I 1 2

I !

i i ! i i ! I ! I I I I l l l l l l ! l i I i l ! l I i i i i l l I l l l l l i l I i I

i l 1 i i i i i i ! I i l ! I l I l l l l I l l l I I I i ! I I l ; i l l l l l l I !I i I i ! I I i ! I I i j i i I I i I I i i l i ! I I I I

I j l ! I I I I l i I I i i i I I I I I i l 1 I i !II I I I J I i !! l ! I I l l l l I j l I i !l l I! I I I l l l l l l ll I l l l 1 III PGE 1787 (Nov 89)

' z., l N Y-15-1996 13:00 SIEPRA WM N

8 O

~ fb'.& T SANDIA REPORT ./ pad B A/# / 4 SAND 88-1358

TTC-0811 UC-71 Unlimited Release Printed January 1991 1

Estimate of CRUD Contribution to Shipping Cask Containment Requirements 4 h. f cOk vPv Robert P. Sandoval, Robert E. Einziger, Hans Jordan, Anthony P. Malinauskas, Waher J. Mings fgg i O'a$*' *** " i onn.e sador Camarest DE-AC04-Feopoc733 l l I h 4

'E-I NY-15-L*96 1::c: 5i= w,. tUCLE m COpo 403 435 5206 P.03 f4 -Q)9 Apeh 8 $AJeh impor an radio-For shipmen:s of fuel 5 years old or oldar, the nos:A compilation of mea us the number of nuclide in CRUD is cooal:-60." spot' activity densities and visual o rods, assemblies, or sas;,les examined,The ae:1vity densi:y ranges in Tabits 2 and 3 and BUR fuel, respec:tvely. The reader is referred to Appendix I, A histo 6;as was are distribuced by rod number. Section 1.5 for :he details of this disrribution.2 and 3 to display the cons: rue:sd for FVR and SUR fuel shown in Tablas l Both :ypes of percentage of rods in selected activity density interva by an (see Figures 3 and 4). assantially CRUD-free mode f CRUD In all likelihood, these distribu:Lons are not representative o CRUD esasurements are deposi:s on :hs entire span: fuel rod population. illance programs, normally made during routine fuel qualification and survected to be cau. sed by or when there is an in resetor problem which is suspeThe "spo

activities, and may therefore provide a conservof the a i

n spen: fuel. The CRUD. 5 the control of reactor number of rods in the firs: Rods recently irradiated and modes) are expected :o increase with time because water chemis:ry has improved in recent years.likely fall in the lower part of

hose irradiated in :he fu:ure vill mos:The second mode aneospasses & au the CRUD distribution.

fewer assemblies. but :epresents sitt ificentiv I 4.3 CRUD feallation ra-iological source term during transporta. Spallation CRUD can contribu:e to :hespalls free the red and becomes airborne a the The abill:y of CRUD :s f tion only if 1: depend on the condition of the CRUD atand subsequen: handling.si:e effects of pool s:orage, vill dspend on :he pgr icle ~ become airborne t spalled CRUD. h ion. A There is no published systema:ic study of CRUD adhesion On CRUD can easily be UD was of the 108 samples from Brunswick 2 BUR fuel rods. 65% of the flocculen: CRUD has flocculen: type [AS78). baan observed during handling in' spent fuel puols:In con: ras:, a tenacious CRUD I of silicon carbide is known as the * :omato soup e f f e ct. " Generally some sor usually found on PUR fuel rods. stone or paper is needed to remove th CRUD. affected by the length of time in the f.xaminations of The tenaci:y of the CRUD ney be water in che pool. storaSe pool and the condition of the CRUD which appeared oconee PWR fuel between 1973 and 1982 indica:ed tha:in 1 83, ce will The data, however, are insufficient to decernine if adheren ra:har adheren: J082). degenera:e in every case. dry air There are a number of observations of CRUD behavior in s:ag was obcarved on a steam-CRUD spallation he temperature and nitrogen acnospheres,*anararing heavy vatar reae:or (SCWR) fuel rod wh p%MMMh

.j f: ~ ' MY-13-1996 13:02 SIERRA N.y' FAR CORP 408 4 8 5406 P.04 Dc s-se ~ ~ k N k 0 5 d Y o f it f Tabis 2 Range of Cobalt-60 Maximum " Spot" j Activity in FUR CRED Deposit Activity DensiT *2 Rameter h WAssmab h *Samoles fuC1/es2) Visual Inznection Yankas Rows 10 2 140 Deonee 1 300 5 27 10 Point Sesch 1 280 5 0.1-2.0 very thin aanzau 1 2100 25 0.1 16 7.lon 1 600 16 37-73 g.3. Robinson 5 6 19 seaver valley 45 8 4 213) Troj an 3700 24 lu5 1-220m

memed, d

remainder-CEUD-free 2arita Many 12 Thin 64 1 Not significant Surry I.ittle Calvert Cliffs 93 Turkey Point 320 Very thin L Maine Yankae 27 4 very thin 390-no CRUD. ANO.1 450 90-minimal AND.2 384 0 CRUD

free

^ Thick gg.3 7 4 Light

non Tarley-1 136 2,

Point Beach 112 2 Light (a) At time of fuel discharge. O) CRUD lavais given in pCi/gm were converted to #C1/c=2 using thickness and density seasarements. ~ 11 Gir .rp.

i 'I. P1AY-15-1996 13:02 SIERRA NUCLEAR CORP 408 438 5206 P.05 e /PC %-007 JyaaadrJn Table 3 Range of Cobalt-60 Maximum " Spot ~ Aativity in SQR CRUD Deposic Activitym Density humi S esaccion f ufM /ewJ1

Rods
Asp =bMes WQ-elea CEUD-free 11 Raaetar _

5 L15ht 125 0-350 Peach Bottos 5 4 1 Monticello 140-1250 l 1 5 100 650 Tsuruga 4 M111stens 1 200-40$m 4 Observed 23 Nine Mile ft 19 Vary chin 1 Dresden 1 16 1,ight Big Rock Pt Very thick 30 Oyster Creek 4 9 AISA-660m ~ C1/cm2 u. sing thickness cor. arted to At time of fuel discharge. CRUD levels given in s i/gm v 3 c la) measurements and a density va (t) The red had failed in the ie,a air acnosphare [CA77}.re overheating, and the obser d light-water reactor (LUR) faal ro s.either cuum at reached 100*C in anwith same indic.ations of in-co (MC86j. reactor may not be applicable to41/2 months inA in CRUD characteristics faal assemblias spent 241*C, with no noticsable change d defective M. B. Robinson P d tmosphara ing air and in a static argon air 5.6% of t Eintiger and Cook tasted intact an f the CRUD After 5960 hr in flowins a l Peach Botton EUR fuel rods in f ow lled, and beevean 1.6 and 4.at oi l Further tescing produced only m O 30'C for 5960 hr IEC84). sults However, when relating these rethe CRUD on the Robinson red had spa were allad. on the Peach Roctom rod had sp h(8f llation, it must be rememberad that (less chan it). additional spallation to transportation spastacionary, without vibration. .ry-.

) g:- i 1 G 1 i 3 05 0 70 6 l 80 S'O - 3 4e-g 30 E 2 to 9 h. E 10 0 0 40 80 120 160 200 i Coban-60 ( pClicm2) 1 A Figure 1. w Distributton of FWR Spent-Fuel Rods as a nmet ten of the Maximum g " Spot

CRUD Activity at Fuel Dlacharge

% ggg j A b c m ', + A s w

d. >

l g,- I dm i h. n 5 90 i 80 l'! 9 , 70 f> 3 @3 e so -n

50 E

In f' S no 4 q E30 i no 1 10 0-0 200 400 600 800 1000 1200 l cobalt-so (pcitem2) O u M M g s g FLgure 4. Distribution of BW Spent-Fuel Reds as a Funetton of the Maxinusa p s,et...o. 1.,.t,.e1 1. x 8 s x I

J'!Pf-15-1955 Dot 5157p; m 5pp capp 08 C8 520m, P.08 69 p 4pvkBsUg,fj Table 4 lists a number of transportation conditions that cover the range of thermal and mechanical stresses prescribed in 10 CI1t 71. In the first two thermal cases, es:imates of :he tspallation fractions can be made from existing data IEC84 C1.84 ). Although( che tests by Einziger and Cook [IC84) and Olsen [0L&4) vare a: only 230*C, they should be applicable to a f r because surface dryout due to evaporation should ! y/ [',,, 300*C aur ace tampe a:ure, have occurred in be:h cabes, and no si5nificant claddin5 creep will occur during the relatively shor: period of transporta: ion (EIS4b}. Although 64 for PWR rods and St spallation fractions for these tests are quoted a for 5VR rods. When uncer:41ncies are considered, the upper limit of the spallation fractions in both cases appears to be abou: 86. The similarity be: ween spallation fractions observed during these tests for PWR and BUR fuels may be attributable to spallation cf the loose flocculent CRUD on the 39R fuel rods during handling in the reactor pool prior to testing. N.S Y \\ occur a: A: the higher tempera:ure of 450*C, cladding ballooning migh: incipien: crack sites and cause additional spallation [5786). Tests by I Stahl et al. indicated that when ballooning does occur, deforma:lon is localized in an approximately 0.15-a-long region [ STB 6}. If 1006 CRUD spallation in the 0.15-a-long region is assumed. and 8k spallacion on th( roc, e tota; spallation between 12 and 15% would be re==inder or tne expected for ?WR rods. Less spallation can be expected for BUR rods t because their lower internal gas pressure leads to a lower claddin5 stress, and hence ballooning, than in FWR fuel rods. / The spallation fraction cannot be estimated in the other four cases lia:ed in Table 4 because of insufficient data. 4.4 Particle fire Diser nurian n [ The ability cf CRUD ce transpor: through the cask fill-gas is a strong function of particle size. CRUD particle size may depend oc whe:her the CRUD is flocculent or adheren:, and on whether it is associated with PWR of For ins:ance, the thick, porous CRUD layer on :he rods from the BUR fuel. ~ Dodevaard BUR reactor was observed to bt quite loose, with poor adherence, j and could easily be brushed away as powder [3D83). On the other hand, CRUD j flakes -Ast had been scraped from the Deonee 1 Cycle 1 PWR fuel rods [BA79) vers agglomerates cf smaller particles and were as large as 2 mm3 Quantitative da a on CRUD particle size are summarized in Table 5. t When CRUD was gathered from undervater sampling. it was deposited on a 0.45-um filter paper. ~he particulate size was determined in a few samples by scanning electron microscopie (SEM) examination of the filter paper. The fluffy CRUD on s Brunswick 2 SUR fuel rod was coepesed of amorphous Particles ranging from 0.1 pm to 0.3 um in diameter, and irrerular shaeed par.icles wi h well-defined faces ranging from 1 um to 3 pa in diameter IAN82}. The tenacious CRUD was similar in shape to the fluffy CRUD but consisted of agglomerates of 0.1 pu to 0.3 yn diamets: primary particles. CRUD from the Mon:icello 3 'R fuel rods contained loose clusters of 0.1 ga to 0.3 pu diameter particles and larger particles ran5 ng from 0.5 pm to i A SIM axamination of CRUD from a Nine Mile Point SVR fuel 2.0 pm [Apa2). rod indicatad agglomerates of 0.1 pm diameter primary particles (ST85 J. These CEUD samples were all taken underwater using a scraping stone or s abrainine, the samp,les may have affected the ^-d TOTAL P.08

SNC Sierra Nuclear Corporation .s h fp q 6.0 CALCULATIONS Aard e h}7// 6.1 ACTIVITY CALCULATION // Cobah" Surface Contamination Cobah"is found in a reactors primary system and in its spent fuel pool. As such. it coats the outside of each fuel assembly and control components with a substance called crud. While this crud contains mostly nonradioactive components. it can contain appreciable amounts of Cobalt" 2 From NUREG/CR-3285. crud deposits a maximum 140 pCi/cm of Cobalt" on fuel assemblies and control components. Fuel' inches AU' = n x D x L = = x 0.374 inches x 160.0 inches = 188 rod

Conservativel) assume the rods extent the total lenph of the assemoly to to calculate the end tit:ing surface arts.

inches

rods assv A%' = 188 x (17X17)assy x 24

= 1.30 x 10' inches = 8.4 x 10*cm. rod Basket

lt is conservatne to assume !*N17 rods ompared to the actual number or:6J pCi " (8.4 x 10'em ) = 1176 Ci '

ActU,' = 140 C' C' cm-. Basket ClientProject: BNFL 01-02 l xcuen l Preparea -l o=e lcacekedl o=e t Sace' ,ubject:_TranStor Shinninc Cask Containment 0 ! jf ly.g,2c l /v IJ 7-9( I5 i Analvsis (PW'R) ( l l l 1 of Calculation Number: BNFL 01.10.06.05 i l l l l 29 l

~% T... SNC ~ f* S c'a.*?.,', g' 9 Sierro.Vuclear Corporation 4 x; Ed ff -ggf \\ ~ Reacto ontro onents - ~ W"' ~ jaca4 a sh //0.i, \\ f< .<yy.* 2 tf A *jC = ( x D x Lace x

rods) - ( x D,.m xL,,,)

3 nee .e. 1 A"jC = ( x (0385in) x !50.574in x 36)-( x (1.840in) x 10.375in) = 66163lin; = 42685.8cm-l 1 A"Cru

As x 1.1 = 42685.8cm x 1.1 = 4.7 x 10'em-3
A factor of 10% is conservatively used to estimate the surface area of the RCC's odd geometry's.

Act"jC7,, = 140 h~ x 4.7 x 10' #* s x 24RCC = 158 C# ~ cm-Basket Basket Tne total surface activity. 7 Ci

Ci "

Ci ~ Act "** = Act""CC - Act'^"" = 158 - 1,176 = 1.334 C# C* C 3 s '** Basket Basket Basket Using the exponential law of decay. shown below. the Cobalt activity is decayed 5 years. 6 N(t) = N e-" e I Where Te is the isotopes halflife ). is a unique decay constant for the isotope Client Project: BNFL 01-02 ! Reusion l Preparec l Date l Lne:tec l Dale. l Snect bjeen TranStor' Shinnino Cask Containment 0 l jf l7-29-tcl # II 7-9 C, l 16 l Calculation Number BNFL Analvsis (PWR l l l l 1 of 01.10.06.05 t I I I l L 29

' %'N SNC Sierra Nuclear Corporation i 5.6 Passive Failure of Basket Pressure Test Line i 5.6.1 The Baske free gas volume at 7 psig is converted to an equivalent volume at standard conditions: Vs, = 359,716 in' jydk B SYEl (359,716 in')(14.7 psia + 7 psi) 3 V,IP = = 531,009 in l }4,y,ja p V,,,,m=(V,)-(V,) = 171,293 in' = 99.1 scf 3 ~ 5.6.2 Airbome Particulate Concentration: e l A = 690.2 Ci Co' per basket [Ref 6] nd ] A,, = (0.15)Amo = 103.5 Ci Co' [15% goes airbome] C,, = (103.5 Ci)/(531,009 in') = 1.949 x 10" Ci/in' l = 11.89 pCi/cc 4 5.6.3 Total curies released to fuel building atmosphere: j l Am.m=(1.949 x 10" Ci/in )(171,293 in') = 33.39 Ci Co' i 5.7 Basket Shield Lid Drop onto Basket During Placement

5. 7.1 For a flat drop orientation. the analysis is presented in Appendix C.

5. 7.2 For the edce drop orientation. the maximum number ofimpacted fuel assemblies:

Shield Lid OD = 64.10 in [Ref. 7] l Shield Lid thickness = 8 in [Ref. 7] l Distance to top of sleeve assemblies = 161 in [Ref. 7] Length of fuel assembly with RCCA = 168 in [Ref.10.11] Distance from top of Sleeve assemblies to top of RCCA = 168-161 = 7 in Llient/ Project: PGE-01 ! Revtston l Prepared Date l Checked i Date Sheet

Subject:

TranStormFMEA 0 l sec W/g 9 g/22/g, l Of r Calculation Number: PGE01-10.02.05-05 l i 15

i: ATTACHMENT VIII Pages 4-4,4-5, and 4-16 of" Topical Report Seismic Analyses of Structures and Equipment for Nuclear Power Plants," BC-TOP-4 Rev 2.1974, Bechtel Power Corporation

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BC-TOP-4 i Revision 2 TECHNICAL UBRARY, TNB 2 ,..rsan nucun euur : June 1974 TOPICAL REPORT SEISMIC ANALYSES OF STRUCTURES AND E0.UIPMENT FOR NUCLEAR POWER PLANTS Bechtel Power Corporation San Francisco, California

>k BC-TOP-2 Rev. 2 Two consecutive modes are defined as closely spaced if their frequencies differ from each other by less than 10 percent. For modes that are not closely spaced, { the criterion of "the square root of the sum of the 'f squares" is used. When modes are closely spaced, they f are first divided into groups in such a way that, in each group, the deviation in frequency between the first k and the last mode does not exceed 10 percent of the l lower frequency. The criterion of "the sum of absolute f'. values" is then applied to each group, and the results f' from all the groups are then combined according to the p,) criterion of "the square root of the sum of the squares". d f, Because of the nature of the design spectra and because most structures have a fundamental frequency within the y frequency range of maximum spectral response, which S is 2 to 7 cps, the effect on the response of these h. p; structures due to be possible variation in structural b or foundation material properties would be negligible, W, s,. b) Time History Analysis - Given the ground motion time history as input, the modal equations, Eq. (4-3), ~' are first solved for each mode, and then the modal re-sponses are superimposed according to Eq. (4-1) to ob- 'E tain the total response. O 4.2.2 Method of Direct Integration Equation (3-1) is directly integrated by accetable numerical schemes when this equation cannot be de-( coupled. For this case, the input is the time his-tory motion. 4.3 Total Structural Response From Separate Lateral And Vertical Analyses The total structural response is predicted by combining the applicable collinear responses, say, R R and R 2 y 4-4 vl t

.l BC-TOP-T Rev. 2 calculated respectively from the two lateral and the vertical analyses. The combination is done according to the criterion of "the square root of the sum of the squares" as follows: R = [R2 + R2 +R2 (4_7) 4.4 Structural Overturning And Soil Pressure 4.4.1 Structural Overturning When the combined effect of earthquake ground motion and structural response is strong enough, the structure will undergo a rocking motion pivoting about either edge of the base. When the amplitude of rocking motion becomes so large that the center of structural mass reaches a position right above either edge of the base, the struc-ture becomes unstable and may tip over (see structural position (b) as indicated by dotted lines in Fig. 4-1). The mechanism of such rocking motics is that of an in-verted pendulum, and its natural period ic very long compared with that of the linear, elastic structural response. Hence, so far as overturning evaluation is concerned, the structure can be treated as a rigid body. If the center of mass of the structure experiences a maximum total lateral velocity V and a vertical velo-H city v, the maximum kinetic energy is conservatively y estimated to be: = m, (vj + vj ) /2 (4-8) E g in which m, is the total mass of the structure and fo unda tion. If a spectral response analysis is done for the structure, the total lateral velocity V s H ccmputed as follows: 4-5 ll

__.._..__.._____m. i BC-TOP-f Rev. 2 ^N f j \\ Position (b) Position (a) j N / N / ) / / / / C.M "H / Y / / 'N l/ N I s s_s aywi eh ygru h VV Fig. 4-1 Position of the Structure when Overturning about One Edge 4-16

r: i 1 1 [ ATTACIBENT IX Sections 2.1.1.2 and 2.1.2.1 of the Trojan Defueled Safety Analysis Report i I 4 i

a, ATrAc R MEL3T' I X l <2 P3-o 3 6-9 T 2.1 GEOGRAPHY AND DEMOGRAPHY The Trojan site was originally selected to minimize hazards to the general public. The site environs have low populatica densities and minimal usage for such activities as j farming and recreation. Some of the site characteristics associated with the Trojan site j selection for the operational phase of the facility remain applicable to defueled condition and the storage of irradiated fuel. This chapter provides discussion of those site characteristics applicable to defueled operation. 2.1.1 SITE LOCATION AND DESCRIPTION 2.1.1.1 Soecification of Location The Trojan Plant site is in Columbia County, Oregon, and lies along the bank of the Columbia River at approximately River Mile 72.5,42 miles north of Portland. The specific geographic location of the site is 46 02' 25" N latitude and 122 53' 03" W longitude. In the Universal Transverse Mercator coordinate system, the site location is j 5098352 meters N by 509000 meters E, and in the Oregon North Zone Lambert Coordinate, the site location is 874375 N by 1394615 E. The nearest incorporated communities are Rainier, Oregon, approximately 4-1/2 miles northwest; and across the Columbia River, Kalama,3 miles southeast, and Longview, approximately 6 miles northeast. Within a 5-mile radius of the site are three small unincorporated communities with a total population of less than 2000: Prescott, Oregon, 1/2-mile north; Goble, Oregon,1-1/2 miles southeast; and Carrolls, Washington,2-1/2 miles northeast. Other than the Columbia River and tributaries, there are no nearby natural geographic features of prominence offsite. The Kalama River joins the Columbia at River Mile 73.1, about 1/2-mile upstream on the bank opposite the site. Similarly, the confluence of 2.1 - 1 gTTAC HME MT IM r$9s _o w -9p,

the Cowlitz and Columbia Rivers is about 4-1/2 miles downstream at River Mile 68. Onsite, however, are the 499 feet natural draft cooling tower, which rises 589 feet above mean sea level (MSL), an approximately 26-acre man-made reflecting lake and an approximately 28 acre recreational lake. 2.1.1.2 Site Area Map The Trojan Nuclear Plant site is an approximately 635-acre tract of land owned in fee by Portland General Electric Company (PGE) in Sec. 35 and 36, T. 7 N., R. 2 WWM, and in Sec. I and 2, T. 6 N., R. 2 WWM, Columbia County, Oregon. The tract is all-inclusive of individual and separate parcels as described in the following deed records on file in Columbia County: BK 168, Pages 13 and 14,22,23 to 26 inclusive,81 to 83 inclusive,117 to 121 inclusive; BK.171, Pages 935 and 936; and BK 174, Page 436. The exclusion area, is defined in 10 CFR 100.3(a). The exclusion area boundary coincides with the site boundary on the Oregon side of the river and extends across the Columbia River to the east where the Washington shore of the river forms the eastern boundary. The major physical facilities are grouped approximately in the geographic center of the exclusion area. The center of the containment lies 2172 feet due south of the nearest point on the exclusion area perimeter, approximately 3175 feet from the nearest point to the west, and approximately 4400 feet from the closest southerly point. Because of its irregular shape. the exclusion area comes within about 4200 feet of the containment at two points approximately southwest of the building. The containment center is located about 2200 feet from the nearest point of approach of the eastern boundary of the exclusion area, and 400 feet from the Oregon bank (mean low water) of the Columbia River. 2.1-2

i 2 1_ I _3 Rnnndariet for Establishing Effluent Relence I imitt As pointed out in the preceding section, the site boundary and exclusion area coincide on land t on the Oregon side of the Columbia River. Programmatic requirements for specifications applying to releases of radioactive material in l gaseous effluents are given in the Offsite Dose Calculation Manual. Doses and release limits l have been evaluated at the site boundary and at off-site locations of actual exposure. 112 EXCLUSION _ AREA AIITHORITY AND CONTROI 2I11 Authority The site boundary (owned in fee) extends to mean low water in the southern pan of the site and to mean high water in the nonhern part of the site. By written agreement with the State of Oregon, who is owner of the submerged and remainder of submersible lands in the river at the site, PGE has control of the uses of such areas out to a line at approximately -20 feet MSL. Beyond this line the U. S. Coast Guard has jurisdiction over river operations. The provisions of the tidelands agreement with the State of Oregon include the following conditions: i (1) RESTRICTED USE: Residential use of and overnight camping in the exclusion area shall be prohibited by the State. Nonresidential activities and uses unrelated to the operation of the Company's adjacent reactor shall be permitted only when no significant hazard to public health and safety exists and then only under appropriate limitations as provided in this Agmement. l 2.1-3 Revision 3

{ 2-t (2) USE LIMITATIONS: In managing its lands within the above-described exclusion area and in sales and leases made with respect to such lands, the State will insen in each Deed, Lease, Easement, Permit or other instmment granted a provision to the effect that the lands affected are within an exclusion area as that term is defm' ed by the Nuclear Regulatory Commission; that such lands are subject to safety regulations established by the State, the Nuclear Regulatory Commission, and the Company with which the grantee shall comply; and that the Company has a right to remove or order the removal of all persons and their propeny therefrom in compliance with said safety regulations. t (3) EMERGENCY PROCEDURES: In the event of an emergency or threat thereof, j which may affect public health or safety, the State grants the Company the right to l enter upon its lands within the exclusion area and to remove persons and propeny therefrom. The State also grants the Company the right to proclaim safety regulations affecting persons and propeny occupying State owned lands within the exclusion area, which regulations upon approval by the State shall be made by the State a part of l every Deed, Lease, Easement, Permit or other instrument issued by it as previously provided. Such mies and regulations when so adopted shall also become a pan of the State's management policy for the administration of such lands. f i Similar conditions have been negotiated with the Burlington Nonhern Railroad (1978). Mineral rights not pan of the original land purchase at Trojan have been subsequently bought in fee by PGE. LI_7 7 1:witition.Aren Activities 17nreinteri tn pinnt Onerntion i The basic reference for activities and facilities within the exclusion area is "The Trojan Nuclear Power Plant: Master Plan Prepared by Lawrence Halprin & Associates" In this 2.1-4 . -}}

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