A06007, Responds to 860821 Request for Addl Info Re Util 860521 Request to Amend License DPR-65 to Allow Storage of Consolidated Spent Fuel in Spent Fuel Storage Pool

Responds to 860821 Request for Addl Info Re Util 860521 Request to Amend License DPR-65 to Allow Storage of Consolidated Spent Fuel in Spent Fuel Storage Pool

ML20215L525
Person / Time
Site:	Millstone
Issue date:	10/22/1986
From:	Opeka J, Sears C, Sears J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Thadani A Office of Nuclear Reactor Regulation
References
A06007, A6007, B12297, TAC-61658, NUDOCS 8610290076
Download: ML20215L525 (13)

Text

-

General Offices e Selden Street, Berkn, Connecticut Y EsSNeow=~

• *eam =>=racc*

• P.O. BOX 270 HARTFORD. CONNECTICUT 06141-0270 L L J "Z"" ,0",".%.*co"." (20a ses-sooo October 22,1986 Docket No. 50-336 B12297 A06007 Office of Nuclear Reactor Regulation Attn: Mr. Ashok C. Thadani, Director PWR Project Directorate #3 Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20535 Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Storage of Consolidated Spent Fuel

)

In May,1986,(1) Northeast Nuclear Energy Company (NNECO) submitted to the NRC Staff a request to amend its operating license, No. DPR-65, for Millstone Nuclear Power Station, Unit No. 2., to allow storage of consolidated spent fuel in the Unit No. 2 spent fuel storage pool. As a result of the NRC Staff review of this proposa Information.lI)theThe NRC Staff forwarded purpose to NNECO of this letter a Request is to provide the NRCfor Additional Staff the requested information.

Question #1:

Does the proposed spent fuel storage design include storage of spent fuel assemblies for Millstone Unit 2 only, or will it accommodate fuel from Millstone Units 1 and 3 or other plants?

Response

The proposed spent fuel storage design has been analyzed for and intended for the storage of consolidated spent fuel originating from only the reactor of Millstone Unit No. 2.

(1) J. F. Opeka letter to A. C. Thadani, dated May 21,1986, " Millstone Nuclear Power Station, Unit No. 2 Proposed Change to Technical Specifications Storage of Consolidated Fuel."

(2) D. H. Jaffe letter to 3. F. Opeka, dated August 21,1986, " Request for Additional Information Millstone 2 Storage of Consolidated Fuel in Spent Fuel Pool."

8610290076 861022 PDR p

ADOCK 05000336 PDR 1 h

/

Question #2 Describe the method (procedure) utilized for installing the consolidated fuel assemblies into Region II of the spent fuel pool. Show that the heat load generated by stored spent consolidated fuel assemblies does not exceed the values of 15.2 and 37.8 X 106 BTU /HR for normal and abnormal maximum heat load cases.

Response

The method (procedure) which will be utilized for installing the consolidated fuel assemblics into Region II of the spent fuel pool is essentially the same method used for movement of non-consolidated fuel assemblies. The cover assembly for a consolidated fuel storage box is a spring-loaded self-locking device that is installed on the consolidation box af ter the fuel rods have been loaded into the box. The cover is dimensionally similar to the upper end fitting of a fuel assembly, thereby permitting the consolidated fuel storage box to be transported as would a standard fuel assembly using the fuel handling tool / system. This feature is described on page 4-4 of the license amendment request. Proposed Technical Specification 3/4 9.20 of the license amendment request requires that the blocked cell of the STORAGE PATTERN remain until the entire Region II STORAGE PATTERN of the spent fuel pool racks has been filled. At this time, consolidated fuel can be placed in a previously blocked cell location only if it is completely surrounded by consolidated fuel. In this way, unconsolidated fuel will only be next to consolidated fuel which is stored in a 3 out of 4 pattern. The reactivity of consolidated fuel adjacent to the unconsolidated fuel is less than K-eff 0.90 since it is 3 out of 4, not 4 out of 4.

The decay heat fraction curves for constant power operating times of 1, 2, and 3 years were plotted from values calculated with the ORIGEN point depletion code. The curves were used to calculate the heat generated by the stored spent fuel, taking into account the power operating time experienced by each fuel batch, assuming annual refueling and 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay time for the most recently unloaded batch (normal operation) or full core offload. For abnormal operation, the full core offload is assumed to occur 36 days af ter the most recent refueling shutdown. Intact fuel assemblics must have at least five years decay times before they are consolidated.

Summary of Normal Operation Heat Generation BTU /HR

1. Unconsolidated fuel assemblies, all 3 years 12.59 X 106 operating time:

(5) 1/3-core batches (150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />,1,2,3 and 4 years decay tiine)(Total 362 assemblies)

2. Unconsolidated damaged fuel assemblies, 0.17 X 106 assumed 2 years time: 10 assemblies (1 year decay time)

3. Consolidated fuel, all 3 years operating 2.41 X 106 time: (19) 1/3-are batches, (5 to 23 years decay time)(To:al 1376 assemblics)

Maximum Heat Load 15.17 X 106

l l

Summary of Abnormal Operation Heat Generation BTU /HR

1. Full core offload, all 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay time: 27.94 X 106 (3) 1/3-core batches (3,2, and 1 years operating time) (Total 217 assemblies)

2. Unconsolidated fuel assemblies, all 3 years 7.27 X 106 operating time: (5) 1/3-core batches (36,401, 766,1131, and 1496 days decay time) (Total 362 assemblies)

3. Unconsolidated damaged fuel assemblies, 0.15 X 106 assumed 2 years operating time: 10 assemblics (1 year plus 36 days decay time)

4. Consolidated fuel, all 3 years operating time: 2.41 X 106 (19) 1/3-core batches (5 to 23 years decay time *) (Total 1376 assemblies)

Maximum Heat Load 37.77 X 106

• Effect of additional 36 days is neglible.

The above heat loads are based on the spent fuel stored in the pool at the end of the plant life and therefore will not be exceeded.

Question #3:

Provide the results of CEPOOL calculations including hypotheses (assumptions) used to develop conservative temperature conditions in spent fuel assemblies.

Response

The spent fuel pool holds storage racks comprised of cells for the Millstone Unit No. 2 spent fuel assemblies. The CEPOOL computer code was used in the analysis of the spent fuel pool thermal hydraulics. Using the flow network method, this code predicts temperatures, velocity, and coolant quality within each cell. Each cell has an internal flow resistance and is connected to its neighbors at the cell inlet through cross flow resistance. A constant axial pressure drop is maintained across the flow network by the entire pool. Coolant in the fuel region, upon transferring heat from the fuel, heats up and becomes lighter than the non-fuel region. The difference in coolant densities in these two regions creates a natural circulation loop to direct flow from the non-fuel region (downcomer) to the fuel region (riser), thereby creating a natural circulation flow loop for removal of heat from the fuel.

The conservative method for evaluation of the fuel cooling in the Millstone Unit No. 2 analysis focuses on a row of cells containing the center cell farthest from the pool walls, thus maximizing the hydraulic resistance between the center cell and the downcomer (see attached Figure 1). Furthermore, it is conservatively assumed in the analysis that cooling flow is provided only by the section of the

_4 downcomer adjacent to the row of cells. Hence, thermal hydraulic conditions derived from CEPOOUs flow network are conservative relative to the actual pool conditions where cross flow from adjacent rows of cells in the lower plenum can occur. The row of cells considered in the design calculations is analyzed assuming that it is loaded to capacity with the hottest fuel (i.e., minimum discharge time from the reactor) and including a 1.55 radial peaking factor with a 10% additional uncertainty allowance in the heat rates.

When a box of consolidated fuel is stored in a fuel cell of a rack, a water gap exists between the cell wall and the consolidation box (see attached Figure 2).

Thus, coolant can bypass the fuel rods within the box and flow through the gap.

All heat generated by the fuel within the box is assumed to be removed by the coolant flowing through the box. However, the coolant density change asso_

ciated with this heat removal is also assumed to exist as driving pressure for the bypass flow through the box / cell wall gap. This introduces conservatism into the analysis since no credit is taken for heat removal from the fuel by lateral heat transfer to the coolant in the box / cell wall gap.

The attached Tab!c 1 shows the results of CEPOOL calculations for (3) cases.

These cases are: 1) row of intact fuel, 2) row of consolidated fuel, and 3) worst combination of intact and consolidated fuel (intact fuel in all cell locations except the center cell which contains consolidated fuel).

These cases were analyzed at normal and accident conditions. Normal operation comprises water temperature of 1500F at the base of the racks and a minimum pool depth of 23 feet of water above the fuel. Accident conditions assume that, as a result of loss of external cooling, coolant is evaporated to a minimum pool depth of 10 feet of water above the racks and that the racks are blocked by a dropped fuel consolidated cannister. Water temperature at the base of the racks is assumed to be 2120F under accident conditions.

As can be seen from the results presented in Table 1, the maximum coolant temperature for the cases of all consolidated fuel and mixed storage of consolidated and intact fuel are less than the case of all intact fuel for both normal operation and accident conditions. Therefore, with respect to thermal hydraulic performances, consolidated storage of spent fuel is no worse than the storage of freshly discharged intact fuel assemblics.

Question #4:

I Indicate the maximum fuel pin cladding temperature you determined together with a sample calculation used to derive this temperature. Include any hypotheses (assumptions) used.

Response

Assuming the worst possible scenario, where boiling occurs along the whole length of fuel, the maximum fuel pin cladding temperature is calculated using an i

equation provided by McAdams for low pressure boiling water (Reference 1 below). This equation is for use in the fully developed nucleate boiling:

q/A = 0.074 (AT)3.86 30 <p < 100 psia The maximum local heat flux for freshly discharged fuel is:

q/A = 97.6 kw/ assembly X 3412 BTU /Kw-hr = 1442 BTU /hr-f t2 176 rod / assembly X 136.7 X '?/ X .44 in2 144 in4/f te Thus, substituting and solving for the film A T, AT = 12.90F The maximum fuel pin cladding temperature is 253oF at the base of the fuel racks. This clad temperature is far below temperatures typically encountered by the fuel during residence in the reactor (6530F).

The critical heat flux to cause departure from nucleate boiling can be estimated using the expression developed by Zuber (Reference 1 below). This critical heat flux at spent fuel pool thermal hydraulic conditions is calculated to be 433,000 i BTU /hr-ft2 which is much larger than the maximum heat flux (1442 BTU /hr-f t2) '

from even freshly discharged spent fuel. Therefore, DNB will not occur, and the assumption of nucleate boiling provides a conservative estimate of the maximum clad temperatures.

Reference #1: 3.P. Holman, " Heat Transfer," 3rd Edition, McGraw-Hill Book Company, New York,1972.

Question #5:

Provide the dimensions of the Boroflex insert shown in Figure 4.4a of your May 21, 1986 submittal which appears to have been omitted. What quality control measures will be used to assure the B 10 loading of 0.03 gm/cm27

Response

The dimensions of the Boroflex poison (neutron) material are 141 + 1/4, -1/8 inches long by 8-1/8 3 1/16 inches wide, by 0.110 1 0.007 inches thick. This material is encapsulated within two 0.029 inch thick sheets of stainless steel, of equivalent length and width, spot-welded together to form a composite (i.e.,

sandwich). This composite, in turn, becomes an integral component of the spent fuel poison box assembly.

The manufacturer of the Boroflex material, BISCO Products, Inc., has in place a quality control program which meets the requirements of ANSI N-45.2 and 10 CFR 50 Appendix B. The manufacturing process is controlled by a series of strict internal manufacturing and testing procedures which meet these require-ments. In addition, the vendor has been audited by Combustion Engineering and several other utilities and vendors to ensure their compliance with these requirements.

Question #6 Definition 1.39 on page 1.8 of this submittal describes " Storage Pattern" which is used to limit placement of consolidated fuel assemblies in Technical Specifica-tion LCO 3.9.20, as follows:

" STORAGE PATTERN" "1.39 The Region 11 spent fuel racks contain a cell-blocking device in every 4th rack location for criticality control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and adjacent and diagonal Region Il cell locations surrounding the blocked location."

In order to clarify this definition,it is recommended that the word "all" be added prior to the word " adjacent." Further, it should be noted that the words, "all adjacent and diagonal" include cells in adjacent modules if that is your intention.

Confirm that this is the case.

Response

The words " adjacent and diagonal" were intended to include those cells sharing a cell wall with the cell containing the blocking device and those cells which, while not sharing a cell wall, directly touch the corners of the cell containing the blocking device. The total arrangement includes nine cells, the center cell containing the blocking device and the eight cells surrounding the center cell.

The cells that constitute a STORAGE PATTERN that do not have a blocking device installed may in fact be included in more than one STORAGE PATTERN.

We proposed that Technical Specification Definition 1.39 be modified to read as follows:

"1.39 The Region 11 spent fuel racks contain a cell blocking device in every 4th rack location for criticality control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and all adjacent and diagonal Region 11 cell locations surrounding the blocked loca-tion."

The attached revised page 1-3 reflects incorporation of this change.

Question #9:

Technical Specification 4.9.13 refers to Figure 3.9-1 for fuel enrichment and burnup in order to permit placement of a spent fuel assembly in Region II of the spent fuel pool yet it is not contained in the submittal. Page 3/4 9.23 of the present Technical Specification contains Figure 3.9-1, which is identical to Figure 3.9-3 of the May 21, 1986 submittal. Please make suitable corrections to correct this apparent discrepancy.

i i

_7-1'

Response

Page 3/4 9-23 of the existing Technical Specification for Millstone Unit No. 2 ,

contains Figure 3.9-1, which is titled " MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE I IN REGION 2." Figure 3.9-3 of the May 21, 1986 submittal is titled, " MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL EN-RICHMENT TO PERMIT STORAGE IN REGION 2 AS CONSOLIDATED FUEL."

l The attached Figure 3 is an " overlay" of Figure 3.9-1 and 3.9-3 which serves to demonstrate the differences in the two figures. The inclusion of two figures

rather than one was based on the decision that separate figures dealing with two distinct operations,1) storage of intact spent fuel assemb!!es in Region 2 vs.

1 Region I and 2) storage in Region 2 of spent fuel as consolidated fuel, would 1

cause less confusion for personnel involved in these aspects of the process.

Question //10:

Describe the means utilized to maintain control over the spent fuel subassem-i biles in order to ensure against premature consolidation of spent fuel assemblies,

, and maintain a proper storage condition af ter consolidation has taken place.

Response

Procedures will be written to address the questions and concerns raised. The current plant procedure for placing a fuel assembly into a Region 11 rack will be expanded and utilzed lor identifying the candidate assemblies for the consoll- ,

dation operation. Since the Engineering Forms associated with establishing the

! burnup of the fuel assemblies are retained for the life of the plant; they also 1 permit identification of the candidate fuel assemblics in Region 11 that have

achieved 5-year decay. All candidate assemblies will be taken from Region 11 of ,

the spent fuel pool.

1 Question //11:

} In response to the Staff's safety evaluation report dated November 22,1985, you

. noted that an amendment request, targeted for July,1986, would be prepared to l restrict plant operation in the event both spent fuel pool cooling trains are not

available. Please provide the latest schedule for receipt of this amendment, a

Response

The amendment request to restrict plant operation in the event both spent fuel pool cooling trains are not available was submitted to the NRC Staff September i 26,1936.

Question //12:

What is the magnitude of the difference in calculated reactivity between the 12-group KENO-IV calculations (used for benchmarking) and the 16-group DOT-2W calculations (used for consolidated fuel calculations) and how was this accounted 4

for in the fuel rack calculations? What organization performed these calcula-tions and what is their previous experience?

l

l j Response l Combustion Engineering performed the consolidated fuel calculations, using a

DOT-Il model benchmarked against KENO-IV and KENO-V. The differences in

) calculated reactivity between a 123-group KENO-IV model, and the DOT model, were evaluated by CE, and were considered to be insignificant.

]

The following table for an infinite-array triangular pitch critical experiment (Ref 3-1 of the license amendment request) compares KENO-V, KENO-IV, and DOT results verus number of energy groups:

1 INFINITE ARRAY MULTIPLICATION FACTORS NO. OF GROUPS KENO-V KENO-IV DOT-il

123 -

1.114 1 004 0 -

i j 50 1.115 1 0005 1.111 1 003 0 -

27 1.123 + 0.005 1.112 + 0.004 -

16 1.117 1 0.005 1.110 1 0.004 1.1177 I

i

{ The finite-array KENO-V 16-group model calculated a multplication factor of i

0.996 + .002 for a critical experiment (Ref. 3-1 of the license amendment ,

l request) with a geometry equivalent to the consolidated fuel rack geometry.  !

4 j Very truly yours, j NORTHEAST NUCLEAR ENERGY COMPANY

. .b '

] 3.F. Opeka 1 Senior Vice President

{ By: C. F. Sears l Vice President i

4 i

I i.

- , _ . , _ - , _ . , _ _ - _ _ , , , , _ . . , , _ _ _ _ _ . . - , - . _ , - , . _,=-r_ _ _ . , , , ~ . ~ . .m..-,. . . - - -

j-l -

b i

i TABLE 1 A

i RESULTS OF ANALYSIS 1

I, Max. Coolant Temp (oF)

Configuration Normal Operation Accident a

4 t

TSAT at Top of Rack 238 225

! Row of Intact Fuel 222 228.4 (boiling across top 2.4 f t. of fuel) i Row of Consolidated Fuel 160 222

]

1 i Mixed Row of Consolidated 220 228.2 (boiling across and intact Fuel top 2.1 f t. of fuel)

I

)

4 j

L l

4 l >

i t I

f f

3 J

5 f .

DEFINITIONS 1

A VENTING

1.35 VENTING is the controlled process of discharging air or gas from a confinement

, to maintain temperature, pressure, humidity, concentration or other operating I condition, in such a manner that replacement air or gas is not provided or j required during venting. Vent, used in system names, does not imply a

VENTING process.

l MEMBER (S) OF THE PUBLIC 1

i 1.36 MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this

category are persons who enter the site to service equipment or to make ,

I deliveries. This category does include persons who use portions of the site for i recreational, occupational or other purposes not associated with the plant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is j exposed to existing dose pathways at one particular location.

I SITE BOUNDARY i 1.37 The SITE BOUNDARY shall be that line beyond which the land is not owned, I leased or otherwise controlled by the licensee.

I UNRESTRICTED AREA l 1.38 An UNRESTRICTED AREA shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of l Individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial institutional and/or recreational purposes, i

STORAGE PATTERN l 1.39 The Region 11 spent fuel racks contain a cell blocking device in every 4th rack location for criticality control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and all l adjacent and diagonal Region 11 cell locations surrounding the blocked location.

I 4

MILLSTONE - UNIT 2 1-8 i

B12297 C. s

, \ Qur e .  !

)

i l

l

=

v S &

Y

\ '

N,m,,,o \ $

> ,,,n .

I N - 1 1

=

.at M

~ -

a ie

~

m -- - ,_ ,--

~

B12297 i

ure 9 l lC L Fuel Rods Fuel Cell ,

"# 0*E Consolidated Box

/ /

o w 8.77 in ---- *E YI" I

O O O 175 in :: Call 1 :: Cell 2 :: - - - - *Y Ce11 15 sr Side View l

c::::: c= c=m c= as:s

h. _

{ _ __

Ortfice Figure 2 CEPCOL IS-Cell Model l

4 x

m \

- so-n s-O R ,,_ g %+ b,, (

se-g Figure 3,q 1 ~

a

$ a s- ' --

~  ; s o-g f s4-g ACCEPTAELE FOR g I-v *28-STORACE IN REGION 2 i 88- Pld-r,m 2-Z Q ^ Figur. 3,q.3 96-4 5

an b

ans 9.-

n T.

9*-

• UNACCEPWelf FOR .9 STORAE M REGION 2 2

> 3 m u

• Q

,k,

, s- +

~

v > i M

< I B I I I I 9 e g ... ... ... s.. s.. s..

I I I 3

{ N FUEL ASSEMBLY INITIAL ENRICHMENT, WT. X U-235 }

l 4 FUEL ASSEMBLY INITIAL ENRICHMENT WT. X U-235 h FIGURE 3.9-3 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION l FICURE 3.9-10F INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 AS CONSOLIDATED FUEL l

w -- . - - - - - - - - - - - - - -