ML18052A392

Proposed Tech Specs Re Expansion of Spent Fuel Storage Capacity & Supporting SAR

ML18052A392
Person / Time
Site:	Palisades
Issue date:	04/16/1986
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18052A391	List: ML18052A390 ML18052A392
References
NUDOCS 8604220164
Download: ML18052A392 (12)

Text

ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255

• FINAL TABLES AND FIGURES FOR .THE FEBRUARY 20, 1986 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY.ANALYSIS REPORT (SAR)

April 16, 1986 Table 5.4-1 of the proposed Technical Specifications Change Figure 3-2 of the SAR Figure 3-3a and 3-3b of the SAR (This figure replaces figure 3-3 of the SAR)

Figure 3-4 of the SAR

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~DR ADOCK 050002S5  ! ,;

- .. PDR LJ 11 5 Pages OC0486-0065-NL02

TABLE 5.4-1 Spent Fuel Burnup Requirements for Storage in Region II of the Spent Fuel Pit Initial Discharge Burnup w/o GWD/1!1f.

'1.5 0 1.6 1. 9 1.8 5.2 2.0 8.5 2.2 11.5 2.4 14.1 2.6 16.6 2.8 18.8 3.0 20.9 3.2 22.9 3.26 23.4 Linear interpolation between two consecutive points will yield conservative results.

Proposed 5-4c OC0486-0065-NL02

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CElHER-T 0-<DITER SPAC U<<; ( I1<<>£S)

FIGURE 3-4 Keff as a Function of Storage Cell Center to Center Soacing in a Region of Spent Fuel Storage Rack 1

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ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255

• ~

REVISED PAGE 3-6 OF THE FEBRUARY 20, 1986 SPENT FUEL POOL RERACK SAFETY ANALYSIS REPORT April 16, 1986 5 Pages OC0486-0065-NL02

Palisades Plant 5 Proposed Amendment - Spent Fuel Pool The probability of any of the first four accidents is not affe=_t:_i:_~-------- ______

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by-the- racks -thel!lse-lves ;- thus-*reracking-cannoc in*c-rease the-probability of these accidents. As for an installation accident, Consumers Power Company does not intend to carry any rack directly over the existing stored spent fuel assemblies. All work in the spent fuel pool area will be controlled and performed in strict compliance with specific written procedures. The Spent Fuel Pool Building crane which will be used to access the spent fuel pool area has been addressed in Consumers Power Company's response to the NUREG-0612, "Control.of Heavy Loads at Nuclear Power Plants."

This response demonstrated Palisades Plant's compliance with Phase I of the*NUREG 0612 criteria. In addition, any temporary construction cranes which will be used to move racks within the spent fuel pool will meet the_ design and operational requirements of Section 5.1.1 of NUREG~0612. By letter dated November 9, 1983, the NRG concluded that the control of heavy loads progra~ (Phase 1) at the Palisades Plant is in compliance with the requirements of NUREG-0612. This program provides for the safe handling of heavy loads in the vicinity of the Spent Fuel Pool. Accordingly, the installation of the new racks will not involve a significant increase in the probability of an accident previously evaluated.

The consequences of (1) a spent fuel assembly drop in the spent fuel pool are discussed in the attached Safety Analysis Report.

For this accident conditio~j the criticality acceptance criterion is not violated. The radiological consequences of a fuel assembly drop are not changed from that described in Chapter 14 of the Palisades _FSAR Update. The NRC also conducted an evaluation (as described in the Palisades SER dated- June 30, 1977) of the pot~ntial consequences of a fuel handling accident and found the calculated doses to be less than a small fraction of 10CFRlOO requirements. Thus, the consequences of this type accident will not be increased from previously evaluated spent fuel assembly drops, and have been found acceptable by the NRC.

The consequences of (2) loss of spent fuel pool cooling system flow, are evaluated and are described in Section 3.0 of the Safety

'Analysis Report (Attachment III). As indicated iri Section 3.0, there _is sufficient time to provide an alternate means for cooling (ie, the fire water system or the shutdown cooling system) in the event of a failure in the cooling system. Thus, the consequences of this-type accident will not be significantly increased from previously evaluated loss of cooling system flow ~ccident$.

The consequences of (3) a seismic event, are ev~luat_ed ?Tid are -

described in Secdon 4.0 of the attached Safety Analysis Report.

The new racks are designed and fabricated to meet the requirements of applicable portions of the NRC Regulatory Guides and published standards listed in Section 4.2 of the Safety Analysis Report. The method of support for the free standing new racks is different than that used to support the existing racks which rest on the fuel pool Rev 1 OC0486-0065-NL02

{J 3.1.4.1.1 SPENT FUEL STORAGE

---

The--final -k- ff- for--Region-I-I- with-spent-fuel--ts-*constructed-a*c-corci-ing Eb the - --- ---* - ----- -

following f6rmula:

keff kworst + 8meth + 8 part + [ksmeth 2 + 1/2

+ ks 2 + ks 2]

worst re where k worst case KENO keff that includes centered fuel assembly worst position, material tolerance, and mechanical tolerances which result in spacings between assemblies less than nominal B method bias determined from benchmark critical comparisons meth

= bias to account for poison particle self-shielding ks 95/95 uncertainty in the method bias meth ks 95/95 uncertainty in the worst case KENO keff worst ks re

= 95/95 uncertainty in the reactivity equivalence methodology The final k ff for Region II from this analysis will be less than 0.95, including all uncertainties at a 95/95 probability/confidence level. There-fore, the acceptance criteria for criticality is*met.

3.1.4.1.2 SENSITIVITY ANALYSIS To show the dependence of k ff on fuel storage cell parameters, sensitivity studies are performed in whlcli the poison loading, the fuel enrichment, and the storige cell center-to-center spacing are varied. Figures 3-3a, 3-3b, and 3-4 illustrate the results of these Region II sensitivity studies.

3.1.5 ACCEPTANCE CRITERIA FOR CRtTICALITY The neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions.

Methods for initial and long-term verification of poison material stability and mechanical integrity are discussed in Section 4.8.

3-6 Rev 1 OC0486-0065-NL02

1) 3.0 NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS

-- ----- ---~--------

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3.1 NEUTRON MULTIPLICATION FACTOR Criticality of fuel assemblies in the spent fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing* the minimum separation between assemblies and inserting ne.utron poison between assemblies.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at .a 95 percent confidence level that the effective multiplication factor (k ff) of the fuel*

assembly array will be less than 0. 95 as recommended in ANSie57. 2-1983 and "in .

the NRC guidance, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling*Applications"[l].

The spent fuel rack design will employ two separate and different arrays. The Region I racks have been previously licensed and installed in the Palisades spent fuel pool and are described in Section 2.0. Since these racks are being reused, criticality concerns for them will not be addressed. The new racks in Reg.ion II are designed to maintain k ff < 0. 95 for Combustion Engineering and Exxon fuel which has an initial enri~liment/burnup combination in the

• acc*eptable area of Figure 3-2 *with utilization of every cell committed.

The following are the conditions that are assum~d in meeting this design basis.

3.1.1 NORMAL STORAGE

a. As described in Section 4.1.2.1, spent fuel storage is divided into two regions. The storage cell nominal geometry is shown on Figure 3-1 for Region II.

b. Storage of fuel in R~gion II assumes burnup of.U-235 has occurred.

Suitability for storage of irradiated fuel in Region II is determined utilizing a minimum fuel burnup versus enrichment curve calculated for the rack design. The actual fuel assembly conditions are defined by the zero burnup enrichment (1.5 w/o U-235).

c. The assembly referred to in the preceding paragraph (paragraph b.) is conservatively modeled with water replacing the assembly grid volume and no U-234 or U-236 in the fuel pellet. No U-235 burnup is assumed.

d. The moderator is pure water at the temperature within the design limits of the pool ~hich_ yie_lds the _largest reactivi_ty ._ A conseryative val.ue of.

1.0 gm/cni is used for the density of water. No dissolved boron is included in the water.

e. The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever i's appropriate for the design.

The nominal case calculation is infinite in lateral and axial extent.

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(\

f. Mechanical uncertainties and biases due to mechanical tolerances during construction are treated by either using "worst case" conditions or by

_________per.forming-sens.fti-v-f.t-y-st-udies*-and*--obtaining- appropriat*e-~valtie-s~-- 'Tlie --------- -----.---- -

items included in the analysis are:

- Poison pocket thickness

- Stainless steel thickness

- Cell ID

- .Center-to-center spacing.

The calculated method uncertainty and bias are discussed in Section 3.1.3.

~* Credit is taken for the neutron absorption in full length structural materials and in solid materials added specifically for neutron absorption.

A minimum poison loading is assumed in the poison plates and B c particle self-shielding is included as a bias in the reactivity calculation.

4 Methods for initial and long-term verification of poison material stab.ility

.and mechanical integrity are discussed in Section 4.8.

3.1.2 POSTULATED ACCIDENTS

.Th~ criticality analysis includes postulated accidents so that the double contingency principle of ANSI 8.1-1983 is met and that the effective neutron multiplication factor (keff) is less than or equal to 0. 95 under all conditions.

Most postulated accident conditions will not result in an increase in k ff of the-rack. Examples are the loss of cooling systems (reactivity decreas~s with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than eight inches of water sep~rating it from the active fuel height of stored assemblies which precludes interaction).

However, accidents can be postulated which would increase reactivity. These would include the inadvertent drop of an assembly between the outside periphery of the rack when empty rack modules. are being installed. Ther'efore, for accident conditions, the double contingency principle of ANSI Nl6.l-1975 is applied. This states that one is not required to assume two unlikely, indepen-dent, concurrent events to provide for ptotecti6n against a criticality accident.* Thus, for accident conditions, the.presence of soluble boron in the storage po0-1 water, is a realistic initiai condition.

The presence of approximately 1,720 ppm boron in the pool water will decrease reactivity by about 30 percent ~k. - In perspective, this is more negative reactivity than is present in the poiso'n plates (18 percent ~k), so keff for the rack would be less than 0.95 even if the poison plates were not presen~~ _

Thu~, for pcistulated accidents, -should there be re.act.ivity increa.se, kef:i would still be less than or equal to 0.95 due to the combined effects of the dissolved boron and the poison plates.

The "optimum moderation" accident is not a problem in spent fuel* storage racks because the presence of poison plates removes the conditions necessary for "optimum moderation". T~e k ff continually decreases as moderator density decreases from 1.0 gm/cm inetfie poison rac~ design.

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5.2.5 AIR ANALYSIS

_______ ..Table ...5"°_9..gives- the most...:r;ecen~ -ON-1--~85}--analysis- of--airborne*-radfonuclide*s-in the SFP area.

Figure 5-2 shows the gross beta-gamma activities from SFP airborne samples

,during the normal operating periods from 1979 through 1983. As with the other parameters examined thus far, there is no correlation between the gross airborne activities and the number of fuel bundles in the SFP.

5.2.6 WHOLE BODY DOSE RATE FROM AIRBORNE RADIONUCLIDES The whole body dose rate from a semi-infinite cloud of radionuclides is given

• in Meteorology and Atomic Energy - 1968 as where D dose rate in rads/sec y

E average gamma energy in MeV/disintegration y.

x radionuclide *concentration. in µCi/ml To convert the dose rate to millirad per. hour, the results from the equation above are multiplied by 3.6E6 millirad per hour/l rad per sec.

5.2.6 WHOLE BODY DOSE RATE The whole body dose rate in the SFP area from the radionuclides given in

• Table 5-9 is:

D (0.25)(3.6E+6)[(0.0453)(2.54E-10)+(0.248)(7.06E-11)+(0.598) y (6.53E-ll)]

6.2E-5 millirad per hour.

i Clearly, this dose rate is negligible compared to the background whole body_

dose rate of 0.69 millirem per hour stated above.

The dose rate from airborne. radionuclides at the site boundary is also negligible.

5.2.7 RADIATION PROTECTION DURING RERACK ACTIVITIES The*radiation protection aspects of the.spent fuel pool *modification are the responsibility of the Plant Superintendent of Health Physics. Gamma radiation levels in the pool area are constantly monitored by the station Area Radiation Monitoring System which has a high level alarm feature. Additionally, periodic radiation and contamination surveys are conducted in work areas as necessary. Where there is potential for significant airborne radionucl_ide concentrations, continuous air samplers can be used in addition to periodic Rev 1 OC0486-0065-NL02