View source for L-77-245, Letter Forwarding Additional Information Regarding Neutron Shielding in the Reactor Vessel Cavity

{{Adams
| number = ML18127B237
| issue date = 08/03/1977
| title = Letter Forwarding Additional Information Regarding Neutron Shielding in the Reactor Vessel Cavity
| author name = Uhrig R
| author affiliation = Florida Power &amp; Light Co
| addressee name = Davis D
| addressee affiliation = NRC/NRR
| docket = 05000335
| license number = DPR-067
| contact person = 
| case reference number = L-77-245
| document type = Letter type:L, Report, Technical, Response to Request for Additional Information (RAI)
| page count = 27
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter:NRC FOPV 194 (2.g B)U.S NUCLEAR REGULATORY COMM~<ON DOCKET NUM BEA NRC DISTRIBUTION FOA PART 50 DOCKET MATERIAL FILE NUMBER TO: CfLETTE R CBCOP Y DESCRIPTION QNOTORIZEO RIUNCLASSIF IEO Mr.Don K.Davis PAOP INPUT FORM ENCI OSURE FROM: Florida Power&Light Co Miami, Fla.Ri Ei Uhrig DATE OF DOCUMENT 8/3/77 DATE AECEIVED 8/8/77 NUMBEA OF COPIES RECEIVED Consists of requested additional information concerning the installation of additional neutron shielding in the reactor vessel cavity at Unit No 1~.~~~PLANT NAME: St Lucie Unit No~1 R JL 8/8/77 (1>>P)FOR ACTION/INFORMATION i$Qjggj(~L)Il(~~egQ~L~Pi I)0 uIT KINK ENVIRHNMENTAL MANAGER ENSING ASSXSTAiVT:
ASSIGNED AD: Ve MOORE LTR BRANCH CHIEF: PROJECT MANAGER: LICENSING ASSISTANT:
Be HARLESS INTERNAL D ISTRI BUTION TFMS SAFETY HEINEMAN E E ENGINE ERTNG PLANT SYSTEMS TEDESCO BENAROYA IPPOLITO OPERATING REACTORS SITE SAFETY&ENVIRON ANALYSIS DENTON&MULLER ENVIRO TECH ERNST BALLARD B OD BAER B ER GAMMILL 2 CHECK AT I A TZMAN ERG EXTERNAL DISTRIBUTION SITE ANALYSIS VOLLMER BUNCH J~COLLINS KREGER CONTROL NUMBER TIC NSIC R IV J HAiICHETT 16 CYS ACRS SENT CAT GO NRC FORM 195 I2 70) e.0'N W t P.O.BOX 013100, MIAMI, FLORIDA 33101 ,)pigG<,~gI QQC~i goal tq1Q FLORIDA POWER 5 LIGHT COMPANY August 3I 1977 X,-77-245 Office of Nuclear Reactor Regulation Attention:
Mr.Don K.Davis, Acting Chief Operating Reactors Branch N2 Divisi.on of Operating Reactors 0.S.Nuclear Regulatory Commi.ssion Washington, D.C.20555 

==Dear Nr.Davis".RE:==
St.Lucie Uni.t 1 Docket No.50-335 Neutron Shieldin On November 29, 1976 (L-77-406), we submitted a plan for installing additional neutron shielding in the reactor vessel cavity at St.Lucie Unit l.Your letter of April 29, 1977 requested addi-tional information about our plan.The information you requested is attached.Very truly, yours, R.E.Uhrig Vice President REU/i41AS/pm Attachment cc: Mr.Norman C.Moseley, Region II Robert Lowenstein, Esquire'772200?$5 HELPING BUILD FLORIDA I(1"1 ATTACHMENT RE: Sg.Lucie Unit l Docket No.50-335 Neutron Shieldin TABLE OF CONTENTS Ir NRC questions of 4/29/77 II.FPL response to NRC questions III.Schedule X.NRC QUFSTXONS OF 4/29/77 1.Clarify if the shield support structure has been designed to.with-stand the following load combination:
1.6S=D+E'here D=moments and forces due to dead load of support structures, bags and contained water.2.Clari y if the seismic excitation along three orthogonal directions was imposed simultaneously for the design of the shield support structure.
The peak response rom each direction may be combined bv the square root of'the sum of the squares (SRSS).Provide the vertical and two horizontal floor.response spectra used in the analysis and describe the basis of their development.
3;Providg clear and legible copies of Figures 1 and 2.You may send full size drawings directly to the NRC Project Mi nager.Also, clarify the location of the sections shown in Figure 19.4.Although the pipe break opening time is currently under review by the HRC staff, ion itudinal break opening time of 5 milliseconds for a 30" diameter pipe would be acceptable without further justification.
The longitudinal break opening tine utilized in your report is significantly greater than 5 milliseconds.
There-fore, you should evaluate the effect of a 5 millisecond break time or provide further justification for your originally proposed break time.
5.The report is not clear with respect to neutron and gamma dose rates.'hroughout the report, the unit HR/hr is used for neutron dose rate.This unit is applicable
'only for x and gamma radiation.
The unit for neutron dose'equivalent rate is mrem/hr.To avoid possible confusion, the report should be revised to better characterize the gamma exposure rate, neutron dose equivalent rate and the summation of the two to provide dose equivalent
'ates (See Pigure 17).6.Provide an occupation radiation exposure budget{man-rem)for the proposed shield.The budget should separate the neutron dose from the gamma exposure where applicable and.should include the following: (a)Han-rem doses received outside containment (e.g.streaming'through containment penetrations such as the equipment hatch)(b)Han-rem doses to personnel inside containment during reactor operations consid'ering routine maintenance and inspection pro-~1 cedures.(c)'an-rem exposures to personnel inside containment during re-" fueling after the shield support structure is removed to its storage position.This is needed since the submittal only considered the exposure to personnel during removal and re-placement of the shield.You should address the expected exposure to personnel inside containment from the support structures activation products during refueling operations awhile the structures are in the storage position.7-It is not clear from the shielding analysis why additional neutron I attenuation, provided by a thicker water shield, will not provide a significant dose rate reduction.
The report states that despite I the neutron dose rate attenuation from the proposed one foot thick'hield, the dose rate at the operating level of the containment appears to be dominated by the neutrons wnich stxeam through the cavity depressurization and ventilation openings.The report should quantify this statement.
Therefore," with reference to Figure 17, specify the fraction of the tabulated,"shielded mr/hr" dose rate that is due Ko streaming Eroa the aforementioned openings.8.Provide an analysis of the exposure dose rate (mr/hr)evolved from 2.2 HEV neutron capture gamma-rays formed from neutron capture of the hydrogen in the water shield.The analysis should.address the capture gamma-ray effects from all neutrons incident on the water shield including the incident thermal neutrons (10 n/cm-sec)and 7 2 those fast neutxons interacting in the water shield that are eventually slowed down and captured.9.Provide neutron streaming data taken during the power assention test program.
EI.RESPONSE 20"4/29/77 NRO~UESTZONS''ON NEUTRON SHIELDING 1.The shield support structure will be designed for the load combination 1.6 S=D+E'here the dead load D includes the weight of support structures, bags'nd contained water.2.The shield support structure design will consider seismic excitation along three orthogonal directions imposed simultaneously.
The peak response from each direction will be combined by SRSS.Attached are the vertical and horizontal (OBE)response spectra used in the analysis.The vertical spectra curve applies to all elevations and was used at the support elevation.
In the horizontal directions, spectra curves at the support elevation were not available, so the maximum"g" envelope of the curves from the next upper (El 44.00')and lower (El 24.00')elevation was used.At a given elevation, the same curve applies to both E-W and N-S horizontal, directions.
The magnitude of the DBE response is defined as twice that produced by OBE excitation..The basis of the development of the floor response spectra is described in PSAR Section 3.7.1.3.G-size prints of drawings SK-8770-AS-154 Sh 1 and 2 (Figures 1 and 2)have been transmitted to the NRC Project Manager, E.Reeves, under separate'over.'ttached are marked-up copies of figures 18 and 19.The corrections shown on these figures will clarify the section locations.
4.The time requi'red by, the jet caused by a longitudinal break in the cold 1e'g to reach the bottom of the shield support structure is estimated to be within a 7 or 8 msec range on the basis of a distance of 9 ft of travel.In our opinion the real opening time of the longitudinal break (to full open)will be in the range of 20 msec.Our opinion is based on the Battelle Memorial Institute tests results as stated in our prior submittal.
Nevertheless, were a break opening time of 5 msec to be assumed as computed in CENPD 168 for a smaller break area, it would mean that the source of the jet would be fully open before the jet hits the shield, instead of having an initially smaller jet hitting the shield.In our analysis no credit was claimed for a reduced area of the jet as the breaks develops.The fully developed jet was used, and in this context the choice of the break opening time is immaterial.
However credit was claimed for a reduction in, reservoir pressure prior to the arrival of the fully developed jet at the shield.The choice of a break opening time does influence the time of depressurization of the 30" line.CE has shown that while the time required to depressurize a 30" line from the 2360 psi operating pressure to 1100 psi for a slot break is.reasonably insensitive to the flow area opening time, it is roughly equal to half the opening time for break opening times between 7 and l3 msec.For a 5 msec opening time then it may be safely assumed that the depressurization time would be no longer than that required to depressurize the line for the longer opening times of 7 and 13 msec.These'times arp 4 and 6 msec respectively.
Even for a 20 msec opening break the depressurization time would be of the order of 8,msec.Therefore it can be concluded that the reservoir feeding the jet when the jet hits the shield will be at the saturation pressure.Our choice of a 20 msec opening time results in a conservatism.
A faster opening time would lead to faster depressurization and increased assurance that the jet hitting the shield would be fed by a reservoir at approximately 1100 psi.5.The unit of neutron dose rate used in the calculation is mRem/hr-.6.Table 1 presents the occupation radiation exposure budget (man<<'rem) determined for the proposed shield assuming an 80%plant factor.I The estimated yearly man-rem saved would by itself not be sufficient to warrant the expenditure of capital required for the shield design, fabrication, and installation, particularly as the exposures are very, sensitive to,the occupancy time assumed.for various areas.In fact it is doubtful that as much time would be spent on the containment operating floor as that assumed, particularly with high dose rates, since little activity is required at this level, with most of the jobs being required at the lower levels.Thqs yearly man-rem savedhas probably been overestimated.
The primary reason for installation of the shield is to minimize the neutron dose xates which would otherwise severely hamper potential maintenance and repair operations inside containment, 7.1arger depths of water would indeed provide larger attenuation of neutrons streaming directly upward or scattered upward through the watex bags..Since no occupancy is present directly above the water bags, dose rates directly above them were not computed.The response at the refueling machine detectors (point no.38 of Figure 17)is dominated by the neutrons which are reflected from the shield or miss the shield entirely and stream through the openings between the shield and the concrete walls.For this point 99%of the neutron dose rate is caused by streaming thxough the opening.For the other detectors, the dose rate due to neutron bypassing the shield is somewhat less than 99%but of the same order.This effect had been noted in prior neutronic analyses which employed a similarly configured shield, i.e., roughly the same extent of coverage at the same elevation above the flange, but of different material and thickness (PERMALI in 2-1/2 ft.thicknesses).
This thicker shield, with roughly double the direct neutron attenuation through it, resulted in dose rate reduction at the refueling machine of approximately a factor of 20.
When the opening between the shield and concrete walls were further reduced, the reduction factor increased from 20 to more than 30, signifying that it is the streaming through the openings that dominates the neutron dose rates.8.The total flux measured at St Lucie at a location immediately below the shield, was determined to be less than 107 n/cm sec.The measurement at this location, however, had a large uncertainty associated with it.To conservatively estimate the capture gamma production in the water bags, a conservative flux impinging on the bag has been derived by weighting the average thermal flux (E(0.45 ev)at the seal ring elevation, which was measured with better accuracy, by the solid angle subtended by the shield.The average thermal flux (E<0.45 ev)measured at the seal ring elevation is 1.5x10 n/cm sec.Weighted by the solid angle subtended by the shield, the impinginI thermal flux on the shield is computed to be approximately 1.5x10'(1-cos70o)=5 x 10 n/cm sec.2 The capture gamma source density is then computed utilizing a thermal capture cross section of 0.33 barns (see attached figure).Its value is 1.1 x 10 g/cc sec.The capture gamma dose rate contribution at point 838 of Figure 17 is computed, again conservatively, by assuming a concentrated point source of strength equal to 7.0 x 10 5/sec located at distance of approximat'ely 1500 cm.A 66%reduction is achieved by self attenuation of the capture f's in the water.The resultant dose rate estimated in the very conservative manner outlined above is less than 250 mr/hr.Since the measured total flux is a factor of 5 less'han that conservatively estimated for the thermal flux impinging on the bags, the actual dose rate, including the contribution from the neutrons above 0,45 ev is expected to be less than 50 mr/hr.9.A report of neutron streaming data taken during the power ascension test program was sent to the NRC on April 25, 1977 (FPL letter L-77-126 from R.E.Uhrig to Dennis L.Ziemann)'. 
~'
TABLE 1 OCCUPATION RADIATION EXPOSURE BUDGET Avg.Neutron Dose Rate (mRem/hr)Avg.Gamma Dose Rate (mr/hr)Estimated Exposure Rate (man-hr/wk)
Exposure (man-rem/yr)
A.OUTSIDE CONTAINMENT l.No Shield 2.Shield 2.5 0;5 2.5 0.5 1.2.0.21 B.INSIDE CONTAINMENT l.Operating Floor-No Shield-Shield 2500 150 500 100 0.7 0.7 109 9.1 2.Other Areas-No Shield-Shield 100 25 50 50 1.4 1.4 11.5.5 3~4, Refueling, Removal 6 Replacement of Shield Stored Activated Support Structure(<)
0 0 33 0.5 18 (b)720 0.60 0.72 C.ESTIMATED MAN-HEM SAVED DUE TO SHIELD 105 a)Assumes 4 people present for 15 days b)one operation per year~g F 80 OB E FL~st'EcT.iRR vEBT-REACT.BLQG.FLORIDA PomER>i LlGHl C>.g~LVC.iE.UPI'O.l Opt=SC.OOhlD A, GC.C i=le.fcoW 80 60 40 20 00 80 E3 CC CC UJ LLJ CE~~V I 40 20 0=I~0/y I I.t I J I f Q.QS 0.10 0.15 0 20 0 25 0 30 0.35 0.40 0 45 0 50 0.55 0.60 0 65 0.70 0.75 0.80 0.85 0.90 O.SS 1.00 PER I QD(SECOND 3. 

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O.pe g 2.QQ l.Fl0 1.(:0 1~'10 1~20 1,0Q I-CC CC liJ.l LU (U CZ SSE=oBE x Z 9.AO 3.40).20 0=-1.04 I O.lO 0.20 0.30 0.40 0.50 0.60 0.70 0.90 0.90 1.00 1;lQ 1.20 1.30 1.40 1.'0 1.60 l.'70 1.OO!.90 2.00 PER I QD(SECQNO 3 

'I~~~3.00 osE, Ft.SPECTRR 1:HGH-)BFfiCT~BLDG Fl.~2il~0 ST.LUCl'E OP~V~.1'2.80 2.60 OBt=6Q-OUZO Ac.C t-l C=Q.4.+to 4'i066 2~90 2.00 1.8(3 1~60.1~00 1'0 C3 I I CC (C LU LaJ (Z I ggF=oSE x 2 F 00 0.80 0.60 0.00 0~20 P..1 Oe/0.10 0.20 0 30 0.~10 0.0 0.60 0.70 0.00 0.90!'0.1.10 1.20 1.30 1 00 1.SO 1.60 1 70 1.80 1.90 2 00 PER IODt: SECGND) 

5~v IV L1 L1 u>610CQ[~6115 K&511 LC 522 L1 O~522 L1 L1 cs+50 t Qo 954 11.5'959~602 5 603 604 605 985 A QI L'I 953 CO 21 L1 607~606, 6O9 0~A'sss 5.063'd L'I SO9 Q~952 e a L1 L1 522 L1 L1 L 508 507~5 B 222 O L1 O L'I 504 503 502 506 8 9S2'sd 2 956 0'0~518 cs 517 1 516 p 515 L 615 NODE NO.616 617 618 619 621 622 623 624'(sea"~)0 00 Qsdo~O 525 524 523 522 521 519 54?CCI Cl Qns ELEMENT NO.0525 578 el O L1 C1 634 3.271'628 C)627 2.375'D 629 632 1 479~630 91O X2GLOBAL, Xl<GLOBAL)527 577 901 Q522 0522 908 900 531 C522 b)9 07, 848 877 74i 8&~72?so 901 902~70 1~801 702 802 702 802 803 0&26 4L Q52 5 07~Quc , happ 823-724 721+~819 4'ots~~+817 906 721.0775-718 O 717 sL 716~4'~res q", 39.98')776 100" i q'l1)0-80S 707 gCt Qssd dos 710~807 705 808 711 973~809 707 (810 Its 904+MBt1 qs~cc 714 8 903 9 1 2~O~813 814 A 7OS e~.0~+&713 PLAN (EL MODEL-NEUTRON SHIELD FIGURE 

3'ECTION"A-A" 958 967~959 0.75'5 Fi'59.95 EL 42.52'L 41.4'05 966 45 591 590 A 591 Q690 Q590-EL 39.98''Ec Y A Secs'8 SECTION"B-B" EL 39 98 907 903 900 5.47'L 34.51'WF31 Q930 A 931 e 931 Q933 A 933 B 933 I Q A 935 e 935 830 SEcZ C 16WF 3S 0.875'32 SEcT D e 6~910 912 0 915 Q936 I Q0 a 939 B 939 A 937 e 937 3 936 938 Sic%'SKcg"RADIALLY OUTWARD DIRECTION" MODEL-NEUTRON SHIELD 940 SEc.q A941 e 941 FIGURE 19 

Cross Sections 1-H-1 10'6 Total------El as t 1 c-"-(n,y)10 8)08 3, 1O6 10 10 Nev 10 10 10 10 1-H-1 

III.SCHEDULE Completion of design Material purchase Material delivery Installation July 15, 1977 August 15, 1977 November 15, 1977 First.scheduled unit shutdown of sufficient duration after material delivery.
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