ML20155C556

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Summary of 981008 Meeting with Portland General Electric in Rockville,Md Re Util 10CFR72 License Application.Attendance List,Agenda & Slide Presentation Matl,Encl
ML20155C556
Person / Time
Site: Trojan  File:Portland General Electric icon.png
Issue date: 10/28/1998
From: Kobetz T
NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To: Shankman S
NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
TAC-L22102, NUDOCS 9811020267
Download: ML20155C556 (53)
v • d • e


Text

WWoMW' W, TWJB MEMORANDUM TO: Susan F. Shinkman, D:puty Dir:ctor Lic nsing and Insp ction Dir:ctorate Spint Fu:1 Proj:ct Office Office of Nuclear Material Safety and Safeguards FROM: Timothy J. Kobetz, Project Manager Spent Fuel Licensing Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards

SUBJECT:

SUMMARY

OF THE OCTOBER 8,1998, MEETING BETWEEN THE NUCLEAR REGULATORY COMMISSION AND PORTLAND GENERAL ELECTRIC TO DISCUSS PGE'S 10 CFR PART 72 LICENSE APPLICATION (L22102)

On October 8,1998, a meeting was held between the Nuclear Regulatory Commission (NRC) staff and representatives of Portland General Electric (PGE) to discuss severalissues that require resolution before NRC can issue a license to PGE to operate an independent spent fuel storage installation (ISFSI) at Trojan Nuclear Plant. An attendance list is included (Attachment 1), and the agenda of issues discussed is included (Attachment 2). Notification of this meeting was published on September 24,1998.

The meeting opened with Spent Fuel Project Office (SFPO) management reviewing the status and schedule of PGE's 10 CFR Part 72 license application. The slides presented by SFPO are included (Attachment 3). SFPO management stated that sufficient information had been submitted by PGE for the staff to write a safety evaluation report. However, several technical issues still required resolution before NRC could issue PGE a license to operate an ISFSt.

1 Representatives of PGE addressed the company's proposed resolution to each of the issues l noted in the agenda. The slides presented by PGE are included (Attachment 4). The staff believes that this meeting led to the resolution of several areas of concern. PGE staff committed to provide NRC with its final resolutions to the issues no later than October 29,1998. ,

PGE plans to begin operating the ISFSI at Trojan in April 1999. During the course of the meeting, no regulatory decisions were requested or made.

Please contact me if you wish to further discuss these issues.

Attachments: 1. Attendance List

2. Agenda
3. NRC Slides
4. PGE Slides Docket 72-0017 Distribution:

Dockets PUBLIC NRC File Center NMSS R/F SFPO R/F SFLS R/F WFKane j PEng FLyon, NRR EEaston LKokajko SGagner, OPA Dried .

VEverette, RIV BSpitzburg,RIV NRC Attendees LThonus, NRR SO' Conner SAIC Attendees

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OFC SFPQM C SFP[ h SFPO C \

NAME TJKobetz:jc VNrpe EJLee[

DATE 10fN/98 10/ /98 10/M/98 C = COVER E = COVER & ENCLOSURE N = NO COPY 0:\ TROJAN \THOJ1003. SUM OFFICIAL RECORD COPY 9811020267 981028

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1 ATTENDANCE LIST October 8,1998, Meeting between Nuclear Regulatory Commission Staff and Portland General Electric Name Affiliation Tim Kobetz NRC/NMSS/SFPO Wayne Hodges NRC/NMSS/SFPO Eric Leeds NRC/NMSS/SFPO Susan Shankman NRC/NMSS/SFPO Mary Jane Ross Lee NRC/NMSS/SFPO Charles interrante NRC/NMSS/SFPO  !

Daniel Huang NRC/NMSS/SFPO I Fritz Sturz NRC/NMSS/SFPO  ;

Steve O' Conner NRC/NMSS/SFPO Charles Gaskin NRC/NMSS/FCSS Sue Gagner . NRC/OPA Steven Mirsky SAIC John Stokely SAIC David Williamson SAIC Roy Karimi SAIC Gary Zimmerman Portland General Electric Chris M. Dieterle Portland General Electric Steve Schneider Portland General Electric Tom Meek Portland General Electric Ted Bushnell Portland General Electric l Dan Gildow Portland General Electric Joel Westvold Portland General Electric John Vesik Portland General Electric David Snedeker Sierra Nuclear Corporation / BNFL Wayne Massey Sierra Nuclear Corporation / BNFL )

James Nestell- Sierra Nuclear Corporation / BNFL Jim Woessner Oregon Office of Energy Rita Bowser Westinghouse .

Steve Schulin IBEX Group  !

Bill Hollaway Shaw Pittman/PFS l l

Attachment 1 f

a

+.

AGENDA Meeting Between NRC and PGE to discuss Outstanding Technical issues involvir.g The Trojan ISFSI Application Technical issues (9:00 am to 12:00 cm. T-7-A-1)

1. Structural Lid Weld Inspection
2. Heavy Loads issues
a. Use of Mobile Crane in association with the Transfer Station
b. Analysis of Drop Accident Scenarios
c. Transportation and Movement of Loaded Storage Casks
3. Radiological Analysis
a. Off Site Dose Calcula6ns
b. Accident Release Calculations
4. Coatings Qualification Report
5. Peak Cladding Temperature
6. Unloading Procedures
7. Changes to the Transfer Cask Design
8. Use of the 10 CFR 72.48 Safety Evaluation Process before issuance of the Part 72 License.
9. Security Plan.

Technical Soecifications (1:00 om to 5:00 cm. O-6-B-11)

PGE will present proposed revisions to the technical specifications submitted with its SAR update on August 25,1998.

NRC staff will present its position on content of the proposed technical specifications revisions with respect to the standard technical specification format.

Attachment 2

g-Attachment 3 APPROACH TO LICENSING REVIEWS j>P""*%<g e

, s s ' n( o g illf!11 gg 4$O o k, * * * *

  • ERIC J. LEEDS, SECTION CHIEF SPENT FUEL LICENSING SECTION LICENSING & lNSPECTION DIRECTORATE SPENT FUEL PROJECT OFFICE OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS US NUCLEAR REGULATORY COMMISSION

l' l OVERALL STRATEGY l

l l

REPROGRAMMED RESOURCES WITHIN SFPO i i

  • WORKLOAD BASED ON INDUSTRY NEEDS l
  • STRICT RULES OF ENGAGEMENT WITH APPLICANTS
  • DEDICATED TEAMS FOR REVIEWS
  • DISCIPLINED STAFF REVIEWS
  • PUBLISHED SCHEDULES WITH CLEAR ACCOUNTABILITY FOR ANY FAILURE TO MEET THEM .

i

  • PROMPT DECISIONS ON TECHNICAL DIFFERENCES OF .

OPINION i

APPROACH To LICENSING REVIEWS PARTIAL OR INCOMPLETE APPLICATIONS WILL BE

RETURNED TO THE APPLICANT
  • REVIEW OF RAI RESPONSE WILL NOT START UNTIL COMPLETE RESPONSE IS RECEIVED
  • APPLICANT'S FAILURE TO MEET SCHEDULE WILL CAUSE RESCHEDULING OF ENTIRE REVIEW
  • APPLICANT TO IDENTIFY AND EYoLAIN THE REASONS FOR ALL DEVIATIONS FROM THE STANDARD REVIEW 4 PLAN

t APPROACH To LICENSING REVIEWS (CONT'D)

WITH SRP IN PLACE, NRC GOAL IS NO RAl' FOR ANY NEW APPLICATION OR AMENDMENT

  • ONE RAI (PERHAPS TWO) WILL BE CONSIDERED ACCEPTABLE, BUT STAFF WILL:

> EXPECT RESPONSE ON SCHEDULE FROM APPLICANT

> PERFORM 2 WEEK REVIEW OF RESPONSES TO DETERMINE WHETHER AND HOw REVIEW SHOULD PROCEED

> SLIP OVERALL SCHEDULE ACCORDINGLY IF RESPONSES ARE NOT RECElVED ON SCHEDULE 1

STAFF GUIDANCE FOR APPLICANTS

-

  • WITH SRP IN PLACE, NRC GOAL IS NO RAI FOR ANY NEW APPLICATION OR AMENDMENT
  • ONE RAI (PERHAPS TWO) WILL BE CONSIDERED ACCEPTABLE, BUT STAFF WILL:

- Expect Response on Schedule from Applicant

- Perform 2 week Review of Responses to Determine Whether and How Review Should Proceed

- Slip Overall Schedule Accordingly if Responses Not '

Received on Schedule

STAFF GUIDANCE FOR APPLICANTS (CONT' D)

  • lF MORE THAN TWO RAIS ARE NEEDED, STAFF WILL:

- Identify its Positions and Concerns

- Suspend Further Technical Review Pending Certification of Application Sufficiency by the Respective Owners l Group or Other Independent Third Party Review Group

!

  • RAIS WILL BE DISCUSSED IN A PUBLIC MEETING i

IF APPLICANT IS UNABLE TO MEET THE NRC PUBLISHED SCHEDULE FOR ANY MILESTONE, A Lt.I i ER MUST BE SUBMITTED AT LEAST TWO WEEKS IN ADVANCE OF THE .

MILESTONE PROVIDING THE NEW SUBMITTAL DATE AND THE REASONS FOR THE REQUESTED CHANGE

  • NRC WILL ASSESS THE lMPACT AND PUBLISH THE i REVISED SCHEDULE

__ _ . _ _ _ _ _ - _ _ _ __-__. _ _ _ _ _ . _ _ - _ .- _ ___. m

CAN WE WRITE AN SER?

Two WEEK REVIEW To DECIDE FOLLOWING RECEIPT OF RAI RESPONSES

  • CRITERIA

> HAS THE APPLICANT lDENTIFIED AND JUSTlFlED ALL DEVIATIONS FROM THE SRP ?

i > HAS THE APPLICANT FULLY RESPONDED TO ALL RAIS ?

> IS THE APPLICATION INTERNALLY CONSISTENT ?

> NOTwlTHSTANDING THE ABOVE, CAN CONDITIONS TO THE CERTIFICATE BE WRITTEN TO ADDRESS IDENTIFIED DEFICIENCIES ?

CAN WE WRITE AN SER?

(CONT' D . )

TELEPHONE CALL TO VENDOR AND OWNERS GROUP PARTICIPANTS TO lDENTIFY ANY LIMITATIONS ON THE SCOPE OF OUR APPROVAL AND FUTURE ACTIONS IF SER IS TO BE WRITTEN

  • PUBLIC MEETING WITH VENDOR AND OWNERS GROUP PARTICIPANTS IF SER CAN NOT BE WRITTEN

.' e PREPARATION OF SER PUBLIC MEETINGS ON 24 HOUR NOTICE TO RESOLVE MINOR OPEN lSSUES 'l VENDORS WILL PROVIDE Lt.i i ERS WITHIN TWO WORKING DAYS OF MEETINGS TO DOCUMENT COMMITMENTS AND/OR PROVIDE INFORMATION NEEDED FOR SER FINAL CLEANUP AMENDMENT TO APPLICATION PRIOR TO SER 2

  • SER WILL BE ISSUED ON SCHEDULE BUTMAY CONTAIN UNRESOLVED ISSUES; IF SO, CERTIFICATE WILL NOT BE ISSUED
  • WHEN TEAM RE-ENGAGES, THE SAME PROCESS WILL BE USED TO PREPARE SER SUPPLEMENT

.k UTILITY lNVOLVEMENT

  • VITAL TO SUCCESSFUL LICENSING REVIEW OF .

VENDORS APPLICATIONS '

  • lNVOLVEMENT NEEDS TO START EARLY AND BE SUSTAINED i
  • CRITICAL MILESTONES

> INITIAL APPLICATION

> RAI RESPONSES i

  • TECHNICAL SPECIFICATIONS ,

> SER DEVELOPMENT MEETINGS

> COMMITMENT LETTERS

> CLEANUP AMENDMENT

> CERTIFICATE

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. i l AGENDA  !

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+ introduction J. Westvold/G. Zimmerman

. Closure Weld NDE J. Vesik / J. Nestell (MPR)

+ Heavy Load issues T. Bushnell

. FuelIntegrity during basket drops and cask tipovers

. Mobile Crane

. Transportation / Movement of Concrete Casks '

+ Radiological Analyses T. Meek

. Shielding Analysis [ Occupancy Factor (Normal Operations)].

. Radionuclides and Duration (Accident)

+ Thermal Analysis C. Dieterle

. Overpack Peak Clad Temperature

. Emissivity Values

+ Changes to Transfer Cask C. Dieterle -

+ Control of Field Changes J. Westvoid

  • Security Plan G. Zimmerman l

+ Technical Specification issues  !,

. Basket Unloading Procedure S. Schneider  !;

. Functional and Operating Limits G. Zim. merman j

. LCO's G. Zimmerman

+ Summary of Actions and Schedule J. Westvoid

i 1

. NDE OF BASKET LID STRUCTdRAL l .

1 WELD - . . .. ..-....- -,.. ._. . ._.. . . . . ...- . .. . --.. ..-s l

+ ASME code requires radiography of basket closure welds.

+ Radiography is not possible due to lack of access to the inside of 1

j the basket after welding.

+ Ultrasonic examination of the weld is acceptable, but the technique is limited due to the weld geometry and the use of stainless steel material. l l + The use of double closure welds and multiple surface exams (PT) has been accepted as an alternative to volumetric examination for ,j i dry fuel storage baskets. 4

! + The critical flaw size was calculated in accordance with ASME Section XI for use in specifying the NDE for the structural lid weld.

~- . ~ . .-

I STRUCTURAL LID WELD NDE l

- . - . . . . . . . - - . .. . . . . . . . ~ - - - . . - - ..... .. -. ..-. - ...

NRC Previously Accepted Guidance Revised Trojan SAR

1. PT may only be used on austenitic stainless steels. 1. The TranStor baskets are ASTM 240, Type 304 stainless steel.  ;
2. PT should be done in accordance with ASME Sec V, Art 6. 2. Table 4.2-1 requires PT to be performed per ASME Sec V, Article 6.
3. PT alone must include the root and final layers and 3. The Trojan SAR is being revised to require a three step, sufficient intermediate !ayers to detect critical flaws. multi layer PT exam for the 3/4" thick structural tid weld at approximately 1/4" intervals of weld deposit (after the first 1/4" for the root PT, after the second 1/4" for the mid-layer PT, and after the 3rd 1/4" for the cap PT).
4. PT of welds must be performed by qualified persons 4. Table 4.2.1 requires PT personnel to be qualified to i and meet the requirements of ASME III, NB-5350. SNT-TC-1 A and for weld to meet the acceptance *i criteria of NC-5300 for PT.
5. For PT alone, a design stress reduction of 0.8 must be 5. A preliminary calculation has been performed to demonstrate that applied to the weld design. a stress reduction factor of 0.8 has been applied to the weld design. ,
6. PT results, including all relevant indications, shall be a 6. The Trojan SAR will be revised to require that a permanent video ,

part of the permanent record by video, photograph or photographic record be made of each PT exam. The Ii or other means. Trojan welding system is already equipped with video ,

cameras and a recording system.

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l OTC MODEL DR-600 ARC WELDING ROBOT >

l + 500 amp MIG / MAG power supply

+ Custom mounting bridge fits transfer and 05T  ; storage casks l !'&d'

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+ Weld data monitoring system ,

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ADVANTAGES

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+ Accurate weld parameter control and repeatability from pass to Pass .

+ Accurate placement of weld passes (laser tracking compensates for basket out of round) 9

+ Minimizes the number of weld stops and starts (where many
weld problems occur)

+ Remote operation and monitoring reduces dose to welders

+ Quick setup - bridge bolts onto existing holes in transfer or storage cask

+ FCAW process has high deposition rates to minimize weld times i

+ Fully programmable system with 6 axes of motion provides maximum versatility l

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HEAVY LOAD ISSUES I i

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+ Fuel integrity during basket drops and cask tipovers

+ Mobile crane i

+ Transportation of concrete casks ,

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BG&E NOTICE OF VIOLATION I j . . . ...-.... . . . .- .

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Original calculations performed to support the USAR noted that during a vertical top end drop accident the l

clip angles, which attach the guide sleeves to the bottom spacer disk plate, would fail in shear at an

! i acceleration of 35 g. p i

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I BG&E determined through analysis and testing that the

! clips actually failed by bending (not shearing) at approximately 43 g and that the guide sleeve would pinch the fuel assemblies Design: A y = 51 g A n = 31 g l

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BOUNDING CONSEQUENC5 'l 4

EVALUATION

+ The question of criticality as a result of an impact with extensive fuel rod damage was postulated in which clad rupture and fuel pellet release would occur.

+ The evaluation concluded that the number of fuel pellets can not be arranged in any configuration that is ~

as reactive as the configuration assumed to be inside the canisters (in the corner fuel sleeves) in the TranStor basket criticality analyses. >

i

+ The most reactive configuration for the assumed number of pellets results in a calculated k o,, value of O.8738

11 MOBILE CRANE l

! i s _. _. _ _ _.. .. ._ .- .... . .._ _.. _ _ _ ._ . ... _ ..._ _..._ _.. ___.___.___../ g l

+ Equivalent requirements to NUREG 0612

+ Crane operated, maintained, inspected and tested in accordance with ANSI B30.5

+ State and Federal OSHA requirements specify B30.5 t

+ Vertical lift only. Transfer station restrains against t

lateral movements 1 .

i

+ Drop analyses within transfer station

i  !

COMPARISON i u

NUREG 0612 TROJAN ISFSI Safe load paths

  • Restrained in transfer station Procedures
  • Trojan ISFSI program, TIP 10 Crane operators qualified & trained
  • Mobile cranes are inspected, tested [

tested & maintained in accordance & maintained in accordance with y with ANSI B30.2 ANSI B30.5 1 The crane should be designed in

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i AIR PAD SYSTEM -

+ Lift height <4" (normally 2-3 inches)

+ Approximately 450 cfm required to maintain gap

+ Emergency stop button

+ Pads deflate if moved off pad surface

+ Previous experience at Trojan on other components

! + Speed limited to <2 ft/sec.

+ 10 ton front end loader prime mover ,

i

+ 1% roadway grade i

i

'i, RADIOLOGICAL ANALYSES

+ lssue

. The Trojan shielding analysis assumed 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> at 100 meters whereas the SRP guidance is to use 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> at the site boundary (e.g., closest point of .

public access).

+ PGE Response / Position

~

. The closest point to the ISFSI perimeter that could be continuously occupied is 100 meters. Since it could only be occupied by workers,2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> was used to calculate the dose.

..+.w. . . .. p.= , .

RADIOLOGICAL ANALYSES i i .

+ lssue l

. The radionuclides used in the Trojan ISFSI release model are l different than the those listed in the SRP. i

+ PGE Response / Position

. PGE used the radionuclides that are most volatile and could be l released from the basket as a result of the design basis event ,

(non-mechanistic failure of basket and fuel assemblies within). ll

li t

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+ +

=

w r * ~-..,e ,e THERMAL ANALYSIS )

l Overpack Peak Clad Temperature

+ Calculation TI-031 (PGE01-10.02.04-03, Rev 6).,10% ,

. Rod Failure Case -

714 F vs. 705 F at 26 kw .

26 kw vs. 24 k' '

i I

f i

f f

f i

THERMAL ANALYSIS l g t i . . . _ _ . . . _

Emissivity Values . _ ._.. ._.._. _.. _ _ ___ ._..., .__.___.__..... __. _.-

+ Calculation TI-033 (PGE01-10.02.04-05, Rev 7) l

. Keeler & Long on basket exterior 0.9 vs. 0.94 l = Sensitivity Study AT of 1 F l

. Overpack to concrete cask gap 0.887 vs. 0.688

+ Used in calculation TI-031 (PGE01-10.02.04-03, Rev 6)

+ Revised Calculations ,

. TI-031 (PGE01-10.02.04-03, Rev 6)

. TI-033 (PGE01~-10.02.04-05, Rev 7)

+ SAR Revision

i I

TRANSFER CASK  !.

+ Design Changes

. Increase ID - decrease RX-244 thickness e Change in door material

. Misc fabrication clarifications -!

. Decrease in hydrogen w/o 2.8 vs. 4.2 ,

+ Impact on Transfer Cask Design

. Revised drawings

. Revised shielding calculation TI-004 (PGE01-10.02.01-05, Rev 3) i i

ISFSI .i t Changes Requiring Evaluation i

+ Drawing changes (both revisions and field changes)

+ Calculation changes (both SNC and PGE calculations)

+ PGE procedure changes (new procedures, revised .

l i

procedure, temporary changes)

+ Use "as is" and " repair" dispositioned ,,

nonconformances (PGE, SNC or fabricator initiated) li h

l!

~

i

+ Fabrication specification changes i

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _+

r.

79c, 3 r,7 s,

,3 IS 72.48 SCREENING NO g IS 50.59 SCREENING NECESSARY?  ? l 5 . GO i .<-

NECESSARY?

L 2 L ' J s J - a i RYES Att 11 ,,

r. - g ..m9 DO PERFORM SCREEN; - ,,..r. .

SCREENING DOES CHANGE ?CFECT NO /IMD .GO;' ~i.1M-

UCENSING  :

DOCUMENTS?

t a w J Each "Yes" pods unust be foWowed to compledon prior to implement >g a change.

YES YES YES YES Att 12 o Question 14 Att 13 o Questkm 713 " "-^ ^ ^23 18

r. , ,., r;, , rg. " Questkm 14,15 r.; .

.7 3 3

3 ' t- .

PERFORM SAFETY f- PERFORM

  • MATERIAL fFROM W N DOES CHANGE WPAbT.

SIGNIFICANCE NO DETERMINATION";IS STANDPOINT, DO WE

  • YES P ANY 10CFR71 UCENSING NO GO DETERMINATION; ANY IMPACT ON LICENSING WANT TO MAKE THE .

.- e

'YES* ANSWERS? DOCUMENTS MATERIAL 7 + CHANGE 7 7 L  ; L  ; L J t  ;

s J s J s J s J

" NO 1 RYES DYES r n.,

. ,3 CONSULT SNC N"~'5 JgiTres. W M C N AND NO DO WE WANT TO MAK'E ES

jfT ,GOT
-A; INFORM E
2 L CHANGE 7 2,

NO s3@, . DO NOT "W i4.A IMPLEMENT'hNr -

YES QCHANGEfjf[if

. g, ,,:s

k UNLOADING PROCEDURE -

.........,-_.....__-._...............J

+ lssue

. PGE has not committed in the ISFSI SAR to develop and implement I a basket unloading procedure as a contingency during loading l operations.

+ PGE Response / Position l

l . PGE will develop a basket unloading procedure and supporting ll analysis / analyses. This will support added technical specification i actions.

l i

l

. . I 1

TECHNICAL SPECIFICATIONS }.

l ISSUES

. . . _ _ . _ _ _ - . . _ . _ _ _ _ _ _ . . . . . _ . . _ . ~ . _ _ . . _ . . . _ . _ .

~

l

+ lssue

. Proposed ISFSI TS LCO 3.1.1, PWR BASKET Vacuum Drying Pressure, and 3.1.3, PWR BASKET Shield Lid Weld and Helium Leak Rate, do not have ACTIONS that return the APPLICABILITY to a l previous MODE, or defined state. The ACTIONS for both TS are open-ended which is contrary to NRC TS philosophy.

+ PGE Response / Position

. PGE will revise both TS to include ACTIONS that require the PWR BASKET to be returned to the Spent Fuel Pool (i.e., Cask Loading Pit) and unloaded if the LCO cannot be met in an established time period.

t

~

i TECHNICAL SPECIFICATI NS j

ISSUES

. _ . ....__....__.___..__._..J

+ lssue

. Calculation PGE01-10.02.04-04, TranStor Transfer Cask Thermal Analysis, assumes an initial ambient temperature of 100 F to determine the maximum design temperatures of the Transfer Cask (TC). Since this calculation is used to determine the limiting temperatures for the design of the TC, a LCO is needed to ensure that the TC is not operated at temperatures >100 F.

+ PGE Response / Position

. PGE agrees that a LCO is needed to restrict movement and operations of the TC at the Transfer Station when the ambient temperature is >100 F.

t l  !

TECHNICAL SPECIFICATIONS l :

ISSUES

+ issue.

. Proposed Technical Specification (TS) Limiting Condition for Operation (LCO) 3.1.2, Helium Backfill Pressure, and associated Surveillance, is not a practical LCO and should be moved to 4.0, Design Features.

+ PGE Response / Position

~

. PGE will revise TS to delete currently proposed LCO 3.1.2 and relocate the requirement to backfill the PWR BASKETS with helium 14.0 to 14.5 psia to Section 4.0, Design Features. Currently numbered LCO 3.1.3, Shield Lid Helium Leak Rate Test, will be renumbered 3.1.2.

__________ _ ____ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ __ _______________-____-________________-_--_____-____\

s TECHNICAL SPECIFICATIONS ,

l

?r ISSUES _ . _ _ . . _ . _ _ . . _ . . _ . _ _ . . _ . _ _ . _

i r

1 -

+ lssue

=

f . Proposed Functional and Operating Limit 2.1.1.d. states that,

[ ' Fuel debris, i.e., loose fuel pellets, fuel pellet fragments, and i fuel assembly fragments ... and shall not exceed 10 kg of ,

v

? fissile material per PWR BASKET and 20 Curies of Plutonium t

per PWR BASKET.' Calculation PGE01-10.02.02-02, Most

} Reactive Loose Pellet Configuration, assumes a maximum of a

7.5 kg of fissile material per PWR BASKET. The proposed d Functional and Operating Limit 2.1.1.d. of 10 kg is not e

1 supported by a calculation.

s

a

]! + PGE Response / Position

{

j . PGE will submit a revised Functional and Operating Limit n 2.1.1.d. of 7.5 kg of fissile material per PWR BASKET.

I i

TECHNICAL SPECIFICATIONS l ISSUES

. . _ .. _._ . _. . . _.__.~ _._._- _. __ .._ . _ .. _.

+ lssue

. PGE did not propose a TS LCO for Concrete Cask Differential Temperature. The NRC feels that such a limit is necessary.

+ PGE Response / Position

. Concrete Cask differential temperature (delta T) is not a meaningful or accurate measurement of any safety parameter.

. During the startup test, the correct loading, shielding, and heat transfer capability of the Concrete Cask will be verified.

. The inlet and outlet mesh screens are visually inspected every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the outlet temperature is measured every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

. The concrete of the Concrete Cask is the limiting safety parameter so  ;

it is more important to ensure it does not approach and exceed a limit i that would make it less effective. 'l

l i

SUMMARY

OF ACTIONS AND SCHEDULE lh

, t

. . . . _ . _ . _ . _ . . . . . . . . . . . _ . . _ . . _ . _ _ . . _ . . . - t L

1. Submit ISFSI SAR Changes . .. .... .. ........ .... .. .......... .... .... .... .. .. .. .. .... .. 10f29/98 '
1. Overpack Therma! Limit - change to 24 kw Table 4.2.12 l
2. Identify specific coatings Chapter 4
3. Change statement, "no fuel damage' to "no significant fuel damag' Section 5.2 t
4. Describe process and systems involved in providing cooling to PWR Baskets IAW Tech Specs Chapters 5 and 8
5. Revise description of NDE of closure welds Section 3.3.2.2 t
6. Reflect impact of revised calculations. Various sections II. Submit Technical Specifications Changes ..... .. .... .... .. .... .. .. ........ .... .... .............. .... ............ .... .. 10/29/98 I
1. Relocate LCO 3.1.2, Helium Backfill Pressure, to Section 4.0, Design Features
2. Add Action Stater,1ent to LCOs 3.1.1, Vacuum Drying, and 3.1.3, Shield Lid Helium Leak Rate, to unload PWR Basket
3. Prepare new LCO 3.1.4, Transfer Cask Ambient Temperature Limits (> -3* F and < 100*F) >
4. Change Functional and Operating Limit of 10 kg of fissile material pe- Basket to 7.5 kg (Section 2.1.1.d)
5. Specify materials in Fuel Assembly inserts (revise Table 2-2) i, Ill. Sub m it re vised calc ulatio ns .. .. ..............

.. .. .......... .......... ........ .......... .... ...... ........ .... ......10/29/98

1. TI-031 (PGE01-10.02.04-03, Rev 6), Thermal Analysis inr TranStor PWR Basket'for Storage Conditions
2. TI-033 (PGE01-10.02.04-05, Rev 7), TranStor Concrete Cask Thermal Analysis  :
3. TI-004 (PGE01-10.02.01-05, Rev 3), TranStor Storage System Shielding Analysis IV. Submit revised Basket and Transfer Cask Drawings ..... ...... .................... .... ..... ...... ........ .... .. .............. ...... 10/29/98 I t

i b

i l

___-_-__ _ __-_ _ ____________ - __________ - - - - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ - - _ - _ - _ _ _ _ - _ _ _ -!

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