ML20148A964
| ML20148A964 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/04/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148A955 | List: |
| References | |
| NUDOCS 8001160148 | |
| Download: ML20148A964 (8) | |
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UNITED STATES g...
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?'. NUCLEAR REGULATORY CCMMisstCi.^.
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5 .9 # C W ASHINGToN, D. C. 20555
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% me ,* f ._.w SAFETY EVALUATION BY THE OFFICE OF NUCLEAR ' REACTOR REGutATION SUPPORTING AMENDMENT NOS, 27 AND 12 TO LICENSE NOS.' DPR 53'AND'DPR-69 hM RELATING 'T0 ' MODIFICATION OF THE SPENT' FUEL POOL _
' BALTIMORE GAS & ELECTRIC ' COMPANY . -"
2.7 E. .. _.
-S. CALYERT CLIFFS ' NUCLEAR ' POWER ' PLANT' UNIT NOS; 1 "AND 2
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1.0 INTRODUCTION
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- By letters dated August 5, and September 7,1977, Ba1.timore Gas and Electric Company (BG&E) proposed to change the spert fuel pool (SFP) [.l.]
storage design for Calvert Cliffs Nuclear Power %nt Unit Nos.1 and ~":?:::
2 (CCNPP) from the design which was reviewed :.nc approved in the s:.ge-operating l'icense review and described in the Final Safety Analysis - -j Report (FSAR). The proposed change consists of increasing the "1 existing spent fuel storage capacity for both units from 410, fuel is riid assemblies to 1056 fuel assemblies. In response to our questions.
H BG&E subnitted supplemental infor:::ation by letters dated October 7
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and 19, November 1, 4,16 and 17, and December 7,1977. ,
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2.0 BACKGROUND
The present spent fuel pool at CCNPP has a' nominal 18 inch (
center-to-center distance between ' fuel assemblies with a total capacity _ . . .
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of 410 fuel assemblies. This design was based on storage capacity 3 of nominally 12/3 cores (410 fuel assemblies), adequate storage of the discharge (72 assemblies per year per unit) from each unit fer =+
one year prior to its shipment off-site for reprocessing plus 217 - +
storage locations for either unit core unloading whenever it became N
necessary.
T;=h The CCNPP Unit No.1 and 2 achieved initial criticality on October 7, 1974, and November 30,1976, respectively. CCNPP Unit No. I was x._..
shutdown on December 31,1976,.for a scheduled refueling and mainte- x_ ,
nance outage, at which time 72 fuel assemblies were replaced. The -j refueling schedule for Unit No.1 shows next refueling in January 1973 EE!
and yearly thereafter. The first refueling for Unit No. 2 is =
scheduled for September 1,978. Following this Unit No. 2 refuelir.g J3 outage, there will not be space to offload either entire reactor core should this be necessary cr desirable because of ooerational _
l considerations. Likewise, foll: wing the secor.d refueling of Uni- ..:. . <
No. 2 in late 1979, the existing fuel pool storage capacity will be 2.
used up completely. == H
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Currently, spent fuel is not being reprocessed on a commercial basis a in the United States. Thus, BG&E has requested our approval of the &-
F-SFP modifications for CCNPP Unit Nos.1 and 2 due to a lack of alternatives in the immediate future, for disposal of spent fuel.
.E The proposed fuel pool modification consists of replacing the old fuel racks in the Calvert Cliffs Nuclear Plant SFP with new, higher F capacity fuel storage racks, which are classified as seismic Category 1 %. [
equipment. The modification consists of eleven storage racks in each G side of the pool, each with forty eight (48) storage elements in a E
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13 X 12.5 inch center-to-center spacing. The new racks are to be fabricated from Type 304 stainless steel and will not utilize poison material for neutron absorption. Each storage tube is 8.875- inches square inside, has 3/16 inch walls and is approximately 14 feet 1:ng.
Combining the spacing dimensions with the outer dimension of the fuel region, which is 8.14 inches, results in a fuel region volume fraction [7 l of 0.41 for the nominal storage lattice. The fuel assembly sits on ~I bars across the bottom of each storage tube. The beams which form the .
base structure are supported by legs about seven inches above the pool i i
floor. Each rack in turn sits on support pads on the floor. There are no connections between adjacent racks, or floor nor are there any I supports to the fuel pool walls.
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BG&E states in their August 5,1977 submittal that it is* responsible U
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for the overall modification to the spent fuel storage pool with the !, "
Nuclear Services Corporation being retained to design the spent fuel racks, contract for fabrication, perfonn analysis pertinent to the :.
modif.ication, and provide installation technical assistance and with Bechtel Power Corporation providing engineering assistance in reviewing El F
the spent fuel pool structural considerations.
3.0 DISCUSSION AND EVALUATION In reviewing the SFP modification for CCNPP Unit Nos.1 and 2, we considered various safety aspects of the modification. These aspec s :
) include (1) structural and mechanical design, (2) criticality analysis, E (3) SFP cooling requirements (4) radioactive waste treatuent, (5) caethod i of rack installation, (6) heavy load impact analysis, (7) operational E:.-
radiation exposure and (8) combined fuel shortage. [E
~ l 3.1 'Strictural and_ Mechanical Desf on We find that the BG&E supporting arrangements for the racks, inclucing ~(
their design, fabrication, installation, structural analysis for all !
loads including seismic and impact loadings, lead combinations, structural acceptance criteria, quality assurance requirements f:r design, fabrication and installation, applicable industry codes; were p
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all reviewed in accordance with the relevant parts of Section 3.7 el and 3.8 of the Standard Review Plan. The licensee used seismic input :0
J in the form of floor response spectra as approved for the plant [
r FSAR. The analytical model used for this seismic design is composed -. . .
of members lumped together with the appropriate mass and structural =J
-r properties to maintain the correct stiffness to insure correct :.l structural behavior. The mathematical model interfaces with the. pool floor by means of four beam elements. The fuel assemblies and fuel cell locations were coupled in this analysis. The responses (shears. -m
=,2 moments and inertia forces).. in the vertical direction and the ~~
worst horizontal direction were combined by the square root of the
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sum of the squares (SRSS) to produce the maximum loading on the 4
structures. Although this procedures does not comply with the -
requirements of Regulatory Gujde 1.92, the licensec has conservatively 7 .,
applied a factor of /f to the total stress calculated and shown that ::g
- the resulting stresses will have the same factor of safety as permitted by the code. In addition to this, a detailed nonlinear ::
time history, analysis explicitly including the clearance. gap between the storage cell wall ind the' fuel assembly was performed'resulting in :::2 2EEi support reactions which were compared with those of the simplified linear 4 elastic model with no gap between the storage cell walls and the fuel : '"
assembly. We have concluded that the. analytical techniques used will result in an acceptable design for the fuel storage racks, tp... y= .. .. ..
The use of 300 series stainless steel materials for the fabrication
' of the new storage racks, and its requirements during the service life, were reviewed for consistency with the requirements fr:ntified .
3::s in Section 9.1.2 of the Standard Review Plan. ,
The analysis, design, fabrication, and installation of the new spent h5 fuel storage racks are in accordance with accepted criteria for seismic r ..m Category I equipner.t. We find that the subject modification proposed. . 19 by the licensee is acceptable, and in part satisfies the requirements e cf the General Design Criteria 2, 4, and 61. M l Since the possibility of long term storage of spent fuel exists, we F are investigating the effects of the pool environment on the racks, fuel cladding and pool liner. Based upon our preliminary review and ..
previous operating experience, we have concluded that at the pool lf temperature and the quality of the demineralizeo water, and taking no !=
credit for inservice inspection, there is reasoneble assurance that i::
E no significant corrosion of the racks, the fuel cladding or the pool liner will occur over the lifetime of the plant. However, if the :..l results of the current generic review indicate that additional protective r?
- easures are warranted to ' protect the racks, the fuel cladding and/or the iy
'iner from the effects of corrosion, the necessary steps and/or inspection programs will be recuired to assure that an acceptable level of ih 5
nfety is maintained. Any conceivable problems which cculd be uncovered Il are of a long term nature and warrant no need for immediate concern.
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12J 3.2 Criticality Analyses The Nuclear Services Corporation perfomed the criticality analyses. []
They used the CHEETAH computer program to obtain four energy group g
-j cross sections for diffusion theory :alculations with the CITATION - 1. '
program. The accuracy of this diffusion theory nethod was checked by ,_
comparison with several series of critical experiments. The fuel pool Es criticality calculations are based on no burnable poison or control M i
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.iods 'in the fuel assemblies, fresh, i.e., unf rradiated fuel with M T
3.7 weight percent uranium 235, and no soluble baron in the water.
For the present fuel assemblies, 3.7 percent enrichment corresponds 5 to a fuel loading of 44 grams of uranium 235 per axial centimeter of a l fuel assembly. ,
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i Parametric calculation's were made for the maximus possible reduction
. in storage lattice pitch, eccentric fuel assembly placement, and an %
increase in fuel pool water temperature to 212*F. A calculation was .E also made fdr the inadvertent placement of a fuel assembly adjacent .
to a filled rack, resulting in a maximum neutron multiplication factor #
of 0.925. This result agrees well with results of parametric cal-culations made with other methods for similar fuel pool storage ,
lattices. By assuming new, unirraqiated fuel with no burnable poison . .
' or control rods, these calculationseyield the maximum neutron i;n multiplication factor that could be obtained thr:ughout the life of f' r the nominal fuel assembif es. This includes the effect of the plutonium which is generated during the fuel cycle.
We find that when the number of the fuel assemblies described in the M l BG&E subnittals, having no more than 44 grams of uranium 235 per axial h
centimeter of fuel assembly, are loaded into the proposed racks, the N
[ neutron multiplication factor will he less tnan D.95. ;
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3.3 SFp Coolino Recuirements The maximum heat load for the expanied capacity in the pool was calculated on the assumption of a 314-day refueling cycle for each L unit with Unit No. 2 being refueled sixty days after Unit No.1. A b cooling time of seven days was assumed after reactor shutdown before j the completion of the transfer of both the noma 1, one-third core b
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! refueling and full core offloads into the spent fuel pool. On this i basis, SG&E calculated the maximua heat load for the nonnal refueling to be 13 5x10 6 BTU /hr plus 2.8x10o STU/hr for un:ertainties or 17.3x105 BTU /hr total .
ne c:oling system for the spen: fuel p:01 has two cumps and two heat e.x cha ngers . These are cross conna::ed so that any c:abination of a
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85 pump and heat exchanger can be used to co,1 the spent fuel pool for either Unit Nos.1 or 2. There is also additional cooling available w_ m from valving the shutdown cooling system of either unit to the spent E2d fuel pool cooling system. Each spent fuel cooling pump is designed 71~
C-to pump 1390 gallons of water per minute. With both pumps and heat exchangers in operation, the spent fuel cooling system is designed -[.s.)
to remove 20x100 STU/hr while maintaining the fuel pool outlet water =
- temperature at 127'F with 95'F service water cooling the heat exhangers.
The shutdown cooling system, when connected to the speht fuel pool, is -
= des-igned to remove 27x100 BTU /hr while maintaining the fuel pool outlet
~ temperature at 130'F with 95'F service water cooling the heat exchanger. ,; 7 In its submittal of August 5,1977, BG&E stated that there are alanns 5=f which will annunciate unsatis' factory water levels in the pool or an g excessive fuel pool water temperature. -s
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- Z; Based on a comparison of the spent fuel pool's heat loads, which BG&E reported in its submittal of September 7,1977, with those Es obtained by using the method given on pages 9.2.5-8 through 14 of the NRC Standard Review Plan (with the uncertainty factor, K equal to 0.1), %
e=4 we find the licensee's calculated values for the heat load to be Z~
acceptable.
- Assuming a full array of 1056 stored fuel assemblies, the maximum r-incrementa13 heat load that will be imposed on the plant by this . . . . -
proposed modification will be that due to nine annual refue' .ngs, . ]
all of which will have had more than two years of cooling. This. a maximum incremental heat load will be 2.64x100 BTU /hr. Since this r" is only 1.1 percent of the heat rejection capacity of the Service 6 Water System, which has a total heat removal capability of 240x10 m BTU /hr, we find that the incremental heat lead will have a negligible M effect on the service water temperature and that the capacity of the E present Service Water System is adequate for removing the incremental !Z~
heat load associated with the proposed modification.
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We find that with both spent fuel pool loops operating, the fuel IE.)..
pool outlet temperature for any nonnal refueling will be less than i;y the 127'F design temperature stated in the FSAR. We also find that Z4 in the case of a postulated single failure, which effectively shuts Mil p
down one loop innediately after any nonnal refueling offload, the M l
fuel pool outlet water temperature will not exceed 155'F. For the 5?
full core offload with the safety related shutdown Cooling System - =;
connected to the spent fuel pool, we find that the fuel pool outlet !?d f
water temperature will not' exceed 140*F, which we find to be acceptable. [Es t=;
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3.4 Radioactive Waste Treatment j
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1 The plant contains waste treatser.t systems designed to collect and process the gaseous, liquid and solid wastes that might contain radio-l active material from both units. ~ne waste treatment systems were .
evaluated in the Safety Evaluatien Report (SER) for both units dated u j . August 1972. There will be no change in the waste treatment systems .
or in the conclusions of the evaluation of these systems as described E in Section 3.1.7 of the 5EP. becase of the proposed modification. L
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3.5 Method of Fuel Rack Installation -
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The first refueling for Unit No. 2 is scheduled for September,1978; consequently, the licensee plans to modify the Unit No. 2 side before storing any spent fuel assemblies in it. There will be no movements f.
of racks over the spent fuel curing this portion of the modification. I After the Unit No. 2 side has been modified, the licensee plans to .E g
move the spent fuel stored in the Unit No.1 side to the Unit No. 2 i side by using the fuel pool service plationn in its design mode as [t =-
described in the FSAR. The fuel is passed through the opening in ,k
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l the com=on dividing wall in the :ccl. The gate would then be installed in the slots on the Unit No. 2 side to ccver the opening. As the Unit No.1 side is drained, the differential pressure holds the gate against a rubber gasket providing a leak tight seal. The itcensee ,
can then modify the Unit No.1 side of the pool without moving racks over spent fuel. .-
! By taking advantage of the split-pool concept, the licensee can install ,'
the new racks without having to cove a rack close to or over spent fuel . After the new racks are iis alled, the fuel handling procedures in and around the pool will be the same as those that were in effect prior to the pr: posed modificati:ns. We find this method of fuel rack :
installation to be ac:eptable. ,
3.6 Heavy Load Imoact Analysis The NRC staff has under way a generic review of load handling operations [;;
in the vicinity of spent fuel pcols to detemine the likelihood of a !=
heavy load impacting fuel in the pool and, if necessary, the radio- p logical consequences of such an event. Because the Calvert Cliffs STS F',
for both units prohibit the movenent of loads in excess of 1600 pounds :
over fuel assemblies in the SFP, , e have concluded that the likelihood ?
of a heavy load handling accider: is sufficiently small that the 1 proposed modification is ac:e::abie. No additional restrictions on E" load handlin; ::eratiens in the ti:inity of the SFP are necessary ~l while cur review is under way. I u.1 The consequences of fuel handlir; a :idents in the scent fuel pool are= ~l are not changed fr:m th:se preser. ed in :he SER for both units dated !
August 1972. l
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3.7 Occuoational P.adiation 'Excosure We have estimated the increment in onsite , occupational dose re- E sulting from the proposed increase in stored fuel assemblies on v;;
the basis of information supplied by the licensee and by utilizing .F realistic asst:mptions for occupancy times and for dose rates in t.c the spent fuel area from radionuclide concentrations in the SFP ~
water. The spent fuel assemblies themselves contribute a negli- ==
gible amount to dose rates in the pool area because of the depth
= _. of water shielding the fuel. The occupational radiation exposure M resulting from the proposed action represents a negligible burden. i.li
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Based on'present and projected operations in the spent fuel pool area, we estic. ate that the prooosed modification will add less i8 than one percent to the total annual occupational radiation ex- F posure burden.at this f acili'ty. The small increase in radiation '
exposure will not affect the licensee's ability to maintain inci- h vidual occupational doses to as low as is reasonably achievable and within the limits of 10 CFR 20. Thus, we conclude that storing ad-ditional fuel in the SFP will not result in any significant increase y in doses received by occupational workers. .
3.8 Combined Fuel Storage ..
In the September 7,,1977 submittal', BG&E states that they plan to store l spent fuel frcm either reactor in either side of the Spent F el Storage J Pool. At CCNFP, the SFP is a common pool divided in two siaes by a 7-
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, wall containing a removable gate. This SFP design was presented'in
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the Calvert Cliffs Unit Nos.1 and.2 FSAR as a single shared SFP for q.gy both CCNPP units and approved as such when the operating licenses were issued. We, therefore, conclude that the BG&E proposal to store spent fuel frcm either CCNPP unit in any location of the common ,
SFP continues to be acceptable provided appropriate fuel assembly inventory concrol is maintained. g..
4.0 TECHNICAL SPECIFICATIONS L:i .
As indicated,in the criticality analysis of this safety evaluation, the Standard Technical Specifications (STS) must be modified to ...
2 incorporate a limit of 44 grams of uranium 235 per axial centimeter of fuel assembly. Since the allowed center-to-center distance between fuel assemblies and the acceptable keff is different for spent fuel -
storage and new fuel storage, new Specifications (5.6) for each type of storage were imposed. The requirements are consistent wi-h those used in STS for other plants.
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Specificatien 5.6 as rela ed to the capacity of the combined stor --
is limited to 1056 fuel assemblies. "~""
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A specification required to prevent heavy load i:: pact by prohibit . - =-
the moveent of loads in excess of 1600 pounds over fuel assembitSf;..
in the SFP is already in Sepcifications 3.9.6 for each unit. . . . . . .
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5.0 SAFETY CONCLUSION '
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- .: .c We have concluded, based on the considerations discussed above, ti;~~~
(1) there is reasonable assurance that the health and safety of tli . . . .~~
public will not be endangered by operation in the proposed manner,. .....
and (2) such activities will be conducted in compliance with the ~"
Commissica's regulations and the issuance of this amendment will r?.
be inimical safety of theto public.
the ccamon defense and security or to the health ar:
.f" Date: January 4. I 978 """'
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