ML20199A726

Responds to 971209 RAI Re Design Basis Accident,Dropping Spent Fuel Pool Weir Gate Onto Spent Fuel Assemblies

ML20199A726
Person / Time
Site:	Catawba
Issue date:	01/22/1998
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-MA0135, TAC-MA0136, TAC-MA135, TAC-MA136, NUDOCS 9801280070
Download: ML20199A726 (7)

Text

-_ ___ __ _ _ - _, _ _ _ _ - _ _ - _ - _ _ - _.

m

. c-f Duke Power Company -

r A Me hwy Compey Catawba Nuclear Station

, p'[

.'8' * -

4800 Concord Road York. sc 29745 Go,y R. Freernoa (803) 831-42$1 oma Vice hnident '

(803) 831306 nx January 22,1998-U. IS. Nuclear Regulatory-Commission Attention:

Document. Control Desk Washington, D.C.

20555

Subject:

- Duke Energy Corporation Catawba Nuc. Lear Station,_ Units _1 and 2 Dockets Nos.

50-413 and 50-414 Request for Additional Information Regarding the Design Basis Accident, Dropping of'a Weir Gate 4

onto Spent Fuel (TAC Nos MA0135 and MA0136)

By letter -dated December -

9, 1997 the.NRC requested additional information regarding the Design Basis-Accident, dropping a

spent fuel pool weir gate onto spent fuel-assemblies.

The discussion of-this accident was submitted to-the NRC as _ part

~f the 1997 revision to the Catawba o

Update Safety Analysis-Report (UFSAR) in accordance with 10 CFR -50.71.-

The- -timeliness of this response meets the-f-

requested period of 45 days.

Please review the additional information and contact Martha Purser (803) 831-4015 with any questi'ons.

Sincerely,

..1 s

Av m

Gary R Peterson xc (with attachments):

L.A. Reyes, Regional Administrator, Region II P.S. Tam, Senior Project Manager, ONRR D.J. Roberts, Senior Resident Inspector, CNS

-iG 9901290070 980122 PDR ADOCK 05000413 P

PDR

. (

,j l

3

4 Duke Energy Corporation Catawba Nuclear Station, Units I and 2 January 22,1998 1

Response to NRC Request for Additional Information Design Basis Accident, Dropping of a Weir Gate onto Spent Fuel Request dated December 9,1997 Question No. I What is_ths basis for the x/Q value of 4.78E-47 Is this the same as ground level x/Q values used in other, currently accepted, design basis accidents? If so, please give reference; if not, justify the difference.

Response to Question No.1 i

The value of 4.78E-4 s/m' has been used in Catawba calculations for several years, including the dose analysis work for the recent S/G replacement amendment package.

This submittal marks the first time this specific value has been inserted into the UFSAR.

The value is a ground level release dispersion factor, and is computed from meteorological data from years 1986 to 1990 inclusive. The value is substantiated in a Catawba calculation performed in accordance with internal procedures. The calculation is based on methodology documented in Regulatory Guide 1.145 (Revision 1, Issue Date:

I February 1983) for an accidental ground level release (equations listed in section 1.3.l(a) of Regulatory Guide 1.145). Input parameters are based on stability class G and a wind speed of 0.9 m/s (1986 - 1990 onsite meteorology using 10 meter level winds). Sigma-Y and Sigma-Z values are calculated using the method described in Appendix F of NUREG/CR-5055. Review of meteorological calculations is performed on a routine

- basis to include the most recent meteorological data.

Question No. 2 What is the current licensed maximum fud burn-up limit? Ifit is greater than 40,000 MWD /MT, justify why a 12% gap fraction for I-131 was not used and a value greater than 1200 psig should not be assumed for pin pressure.

Response to Question No. 2 The subject analysis assumed a gap fraction of 0.1 and a power peaking (Fm, enthalpy rise hot channel factor) of 1.65 in accordance with Reference 1. The current licensed fuel burnup limit for Catawba Nuclear Station is 55,000 MWD /MTU per fuel assembly and

60,000 MWD /MTU for a single fuel pin. In Reference 2 the NRC Staff used a gap fraction of 0.12 for iodine in the calculations which support the environmental impact statement for extended burnup fuel. We deem this to be a overly-conservative value when used concurrently with high power peaking, in Reference 3 it is stated that Pacific Northwest Laboratory (PNL) calculated a release fraction of 0.12 for iodine for a peak rod at a bumup level of 60 GWD/t. This burnup would not be applicable to a highly peaked fuel pin for Catawba Nuclear Station. Since isotopic inventory is linearly related to power level (excepting long-lived radionuclides such as Kr-85), engineering judgment indicates that the highest fuel isotopic inventory is present at a burnup for which the gap fraction used by PNL does not apply. While it would be enveloping for Catawba to analyze activity available for release based on a peaking factor of 1.65, a burnup value of 55,000 MWD /MTU, and a gap fraction of 0.12, it is deemed that the high peaking factor is mutually exclusive with the high burnup assumed in the PNL analysis.

Duke Power Company continually re evaluates design basis calculations, and has made plans to perform a rigorous analysis of gap fraction with the ANSI /ANS 5.4 fissio. ;as release model (Reference *).

This analysis will include ANSI /ANS-5.4 and ORIGEN2 analyses at burnup increments through fuel discharge, in which bumup and nodalized fuel pin temperatures are used to calculate total gap fraction, and power history is used to calculate fuel isotopic inventory. The analysis will also be based on an enveloping power history rather than the value for power peaking outlined in Reference 1.

Fuel pin pressure while fuel is part of a core at power is calculated with licensed methodologies by the General Office Nuclear Engineering Department (Reference 5).

The maxi.aum fuel pin pressure in the Spent Fuel Pool has been shown to be below 1200 psig with the TACO 3 fuel performance code, l

(

Question No. 3 What is the basis for the 95% (for inorganic) e.nd 80% (for organic) iodine removal efficiencies assumed in the first paragraph to Section 15.7.4.2.3? If this assumption only applies to Case 1, why was it not just included in the second paragraph of the same section?

Response to Question No. 3 For the values of 95% for inorganic iodine filtration efficiency and 80% for organic iodine filtration efficiency, please see Reference 6. This amendment revised the carbon testing method for carbon samples from Catawba safety related filters. The NRC Staff reassessed Catawba offsite doses using 80% organic iodine filtration efficiency for the annulus, fuel building and auxiliary building ventilation system.

The filtration assumptions apply only to case 1 (as discussed further in the response to question number 4), since no credit is assumed for filtration in case 2. The location of the subject discussion in the UFSAR sections is related to editorial preference, in order to l

. simplify the descriptions of cases I and 2 which follow the paragraph. However, we will conaider an editorial revision to the UFSAR pages to clarify this point.

Question No. 4 What is the application of Case 2 (40 day decay, no filtration)? If no application, why is it discussed in the design basis description?

Response to Question No. 4 The Catawba Technical Specifications require one train of the Spent Fuel Pool Ventilation System to be operable and operating while moving fuel or when moving heavy loads over the storage pool (see Technical Specification Surveillance Requirement 4.9.11.1). Additionally, Limiting Condition for Operation 3.9.7 requires that spent fuel has decayed for 17.5 days prior to movement of a weir gate over fuel in the Spent Fuel -

Pool, in 1993, a set of single failures were identified which could render a single train of the Spent Fuel Pool Ventilation System incapable of filtration of the postulated source term from a fuel handling accident or weir gate drop accident. When describing the Spent Fuel l-Pool Ventilation System, the Final Safety Analysis Report, the Standard Review Plan 1

(Reference 7) Sections 9.4.2 and 9.4.5, and the Catawba Safety Evaluation Report (Reference 8) discuss separate trains of safety related equipment, supplied with class 1E power, thus meeting the single failure criterion. The Catawba Spent Fuel Pool Ventilation System is designed and accepted in accordance with the Standard Review Plan, Section 9.4.2, which states that "a single active failure cannot result in loss of the -

system functional performance capability." Additionally, during licensure of Catawba, the NRC Staff requested a failure modes and effects analysis of ve.-ious plant responses to Chapter 15 accidents. The Duke Power Company response includes the assumption of a single failure during the fuel handling accident (Reference 9).

The corrective actions for these conditions involved revisions to Site Directive 3.1.30 (for the fuel handling accident) and the procedure for movement of the weir gate over spent fuel (for the weir gate drop accident), which now requires that two trains of essential power and two trains of Spent Fuel Pool Ventilation be maintained operable if fuel movement occurs prior to the required number of hours of decay. Beyond the stated number of days of decay (7 days for the fuel handling accident and 30 days for the weir gate drop accident), no filtration is required to ensure that offsite doses are within the appropriate limits. Hence, a single train of essential power or a train of the Spent Fuel Pool Ventilation System may be made unavailable for maintenance. It should be noted that the decay times stated herein pertain to design basis assumptions such as high power peaking, long cycle lengths, and no power coastdown at the EOC, In lieu of using the minimum decay time stated for design basis analyses, Site Directive 3.1.30 allows for a cycle specific evaluation ofinputs such as peaking factor, burnup, etc., to determine enveloping fuel inventory and hence the specific decay time required before a train of A

Spent Fuel Pool Ventilation may be made inoperable while fuel movement is occurring.

This approach may be taken ifit is determined that the refueling schedule supports fuel movement prior to 7 days or necessitates weir gate movement prior to 30 days.

The weir gate drop dose analysis case with no credit for filtration of activity is related to the discussion above. Before the 30 days decay, procedures require that two trains of essential power and two trnins of ventilation be maintained operable. After 30 days decay, procedures allon one train to be removed from service. Also,it has been determined that the value of 40 days stated in the revised pages of the UFSAR submittal is an editorial error (i.e., the calculation actually supports 30 days rather than 40 days). A value of 30 days should actually have been stated in the UFSAR revisions. An editorial revision will be made to the affected UFSAR pages.

Question No. 5 Although two cases are discussed, only one " environmental consequence" is stated (3 rem whole body,35 rem thyroid exclusion area boundary dose). Is this the result for Case 1 or Case 2? Provide the results for the missing case. Provide the consequences at the low population zone boundary and in the control room.

Response to Question No. 5 It has been determined that the dose results reported in the revised UFSAR pages are in error. (An editorial revision will be made to the affected UFSAR pages and submitted with the next update.) A complete and correct tabulation of dose re.;ults for case I and case 2 follows:

Case Thyroid Dose (Rem)**

Whole Body Dose (Rem)

Case I 66.7 0.586 Case 2' 71.8 0.115 No credit assumed for filtration of source term

" The thyroid dose guideline value is "well within Part 100 limits," or 75 rem.

Doses to the low population zone have not been calculated. Since the exclusion area boundary dose analysis is a two hour calculation, and since it is assumed that all of the activity has been released within this two hour time frame, and given that the low pcpulation zone dispersion factor is significantly less than the exclusion area boundary dispersion factor, the exclusion area boundary dose results will envelope the low population zone dose results.

Control room operator dose results for non-LOCA accidents have not previously been calculated for Catawba Nuclear Station. This issue was identified in a Duke Power Company intemal problem report. It is currently planned to calculate doses to the control s

room operator for all accidents, including the fuel handling accident and the weir gate drop accident.

l l

7.

s,.;

I-References

- -. =,,. _..

'.(1]'

Regulatory Guide 1 ^ 25, " Assumptions Usedfor Evaluating the Potential

~

Radiological Consequences oja Fuel Handling Accident in the Fuel Handling -

and Storage Facilityfor Bolling and Pressurl:cd Water Reactors. "

[2]

- Federal Register / Vol. 53 No. 39 / Monday, February 29,1988 / pages 6040 -

6043.-

- (3)

Baker, Bailey, Beyer, Bold & Tawil, NUREG/CR-5009, " Assessment ofthe Use:

. ofExtended Burnup Fuelin Light Water Power Reactors. "

- [4]

ANSllANS-5.4, "Methodfor Calculating the Fractional Release of Volatile Fission Products From n ide Fuel, " and NUREGICR-2507, " Background and x

Derivation ofANS-5.4 Standard Fission Product Release Model. "

(5]

' DPC-NE-2008, " Duke Power Company Fuel Mechanical Reload Analysis Methodologv Using TACO 3. "

.\\6}

Safety Evaluation Report by the Oplce ofNuclear Reactor Regulation Related' o t

Amendment No. 99 to Facility Operating License NPF-35 and Amendment No. 84 to Facility Operating License NPF-52, Duke Power Company, et. al., Catawba

. Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 50-414, ** August 23,

-1991o (7]

NUREG-0800, " Standard Review Planfor the Review ofSafety Analysis Reports for Nuclear Power Plants. "

.\\8]

NUREG-0954, " Safety Evaluation Report Related to the Operation ofCatawba

Nuclear Station. Units 1 & 2. "

i

(9]:

Catawba Final Safety Analysis Report Question & Answer 440.51, Figure Q -

-440.51-24.

,