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| | issue date = 08/03/1977 | | | issue date = 08/03/1977 |
| | title = Letter Forwarding Additional Information Regarding Neutron Shielding in the Reactor Vessel Cavity | | | title = Letter Forwarding Additional Information Regarding Neutron Shielding in the Reactor Vessel Cavity |
| | author name = Uhrig R E | | | author name = Uhrig R |
| | author affiliation = Florida Power & Light Co | | | author affiliation = Florida Power & Light Co |
| | addressee name = Davis D K | | | addressee name = Davis D |
| | addressee affiliation = NRC/NRR | | | addressee affiliation = NRC/NRR |
| | docket = 05000335 | | | docket = 05000335 |
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| | document type = Letter type:L, Report, Technical, Response to Request for Additional Information (RAI) | | | document type = Letter type:L, Report, Technical, Response to Request for Additional Information (RAI) |
| | page count = 27 | | | page count = 27 |
| | | project = |
| | | stage = Other |
| }} | | }} |
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| =Text= | | =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: | | {{#Wiki_filter:U.S NUCLEAR REGULATORY COMM ~ <ON DOCKET NUM BEA NRC FOPV 194 (2.g B) |
| ASSIGNED AD: Ve MOORE LTR BRANCH CHIEF: PROJECT MANAGER: LICENSING ASSISTANT: | | FILE NUMBER NRC DISTRIBUTION FOA PART 50 DOCKET MATERIAL FROM: DATE OF DOCUMENT TO: Light Florida Power & Co 8/3/77 Mr. Don K. Davis Miami, Fla. DATE AECEIVED Ri Ei Uhrig 8/8/77 QNOTORIZEO PAOP INPUT FORM NUMBEA OF COPIES RECEIVED CfLETTE R RIUNCLASSIF IEO CBCOP Y DESCRIPTION ENCI OSURE Consists of requested additional information concerning the installation of additional neutron shielding in the reactor vessel cavity at Unit No 1 ~ .~~~ |
| 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: | | i $ Qjggj( ~L) Il ( ~~egQ~L~Pi (1>>P ) |
| 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
| | PLANT NAME: |
| | I)0 uIT KINK St Lucie Unit No~ 1 R JL 8/8/77 FOR ACTION/INFORMATION ENVIRHNMENTAL ASSIGNED AD: Ve MOORE LTR BRANCH CHIEF: |
| | MANAGER PROJECT MANAGER: |
| | ENSING ASSXSTAiVT: LICENSING ASSISTANT: |
| | Be HARLESS INTERNAL D ISTRI BUTION TFMS SAFETY PLANT SYSTEMS SITE SAFETY & |
| | HEINEMAN TEDESCO ENVIRON ANALYSIS E E BENAROYA DENTON & MULLER ENGINE ERTNG IPPOLITO ENVIRO TECH ERNST OPERATING REACTORS BALLARD B OD BAER B ER GAMMILL 2 CHECK SITE ANALYSIS VOLLMER AT I BUNCH J~ COLLINS A TZMAN ERG KREGER EXTERNAL DISTRIBUTION CONTROL NUMBER TIC NSIC R IV J HAiICHETT 16 CYS ACRS SENT CAT GO NRC FORM 195 I2 70) |
| | : e. 0 N |
| | W t |
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| ==Dear Nr.Davis".RE:==
| | P. O. BOX 013100, MIAMI, FLORIDA 33101 |
| 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: | | , )pigG < |
| 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.
| | FLORIDA POWER 5 LIGHT COMPANY QQC~i |
| 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.
| | ,~gI tq1Q August 3I 1977 X,-77-245 goal Office of Nuclear Reactor Regulation Attention: Mr. Don K. Davis, Acting Chief Operating Reactors Branch N2 Divisi.on of Operating Reactors |
| 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.
| | : 0. S. Nuclear Regulatory Commi.ssion Washington, D. C. 20555 Dear Nr. Davis". |
| The longitudinal break opening tine utilized in your report is significantly greater than 5 milliseconds.
| | 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. |
| There-fore, you should evaluate the effect of a 5 millisecond break time or provide further justification for your originally proposed break time.
| | Very truly, yours, R. E. Uhrig Vice President REU/i41AS/pm Attachment cc: Mr. Norman C. Moseley, Region II Robert Lowenstein, Esquire |
| 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
| | '772200? $ 5 HELPING BUILD FLORIDA |
| '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)'.
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| ~'
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| 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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| 2.(:0 lif.ACT~BLDG 1'.~44~0 PLOR<Ob Po~ER q L1C qr c~ST-Lvc-LE.On)iT wo, 1 2'0 2~c'.0 OBB QA.OUNO P CCC-l-aQ.h.yqg~
| | I( |
| 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
| | 1 "1 |
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| '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)
| | ATTACHMENT RE: Sg. Lucie Unit Docket No. 50-335 l |
| | Neutron Shieldin TABLE OF CONTENTS Ir NRC questions of 4/29/77 II. FPL response to NRC questions III. Schedule |
|
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| 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
| | 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 (See Pigure 17). |
| | 'ates |
| | : 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 |
|
| |
|
| 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
| | 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 Ishielding analysis why additional neutron 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 7 n/cm 2 -sec) and 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. |
|
| |
|
| 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
| | 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 + |
| | the dead load D includes the weight of support structures, E'here 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 |
|
| |
|
| 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. | | 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. |
| tl s~e envue LltN OhlSS3308d f H3l<A000 03AI3038}} | | 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 saved has 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 Avg. Gamma Estimated Exposure Dose Rate Dose Rate Exposure Rate (mRem/hr) (mr/hr) (man-hr/wk) (man-rem/yr) |
| | A. OUTSIDE CONTAINMENT |
| | : l. No Shield 2.5 2.5 1.2 |
| | : 2. Shield 0;5 0.5 .0. 21 B. INSIDE CONTAINMENT |
| | : l. Operating Floor No Shield 2500 500 0.7 109 Shield 0.7 150 100 9.1 |
| | : 2. Other Areas No Shield 100 50 1.4 11. |
| | Shield 25 50 1.4 5.5 3 ~ Refueling, Removal 6 Replacement of Shield 0 33 18 (b) 0.60 4, Stored Activated Support Structure(<) 0 0.5 720 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 |
| | |
| | OB E FL ~ st'EcT.iRR vEBT- FLORIDA PomER >i LlGHl C>. |
| | F 80 REACT. BLQG. |
| | g~ LVC.iE. UPI'O. l Opt= SC.OOhlD A, GC.C i =le. fcoW E3 80 CC 60 CC UJ 40 LLJ 20 CE 00 |
| | ~ |
| | ~ |
| | V I |
| | 80 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. |
| | |
| | PLOR<Ob Po~ER q L1C qr c~ |
| | : 2. (:0 ST- Lvc-LE. On)iT wo, lif.ACT BLDG 1'. 44 0 |
| | ~ ~ ~ |
| | 1 2 '0 OBB QA.OUNO P CCC-l-aQ.h.yqg~ O.pe g 2 c'.0 |
| | ~ |
| | : 2. QQ l . Fl0 1 . (:0 I SSE= oBE x Z CC CC 1 ~ '10 liJ . |
| | l LU 1 ~ 20 (U |
| | CZ 1,0Q |
| | : 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 |
| | ~ ~ ~ |
| | osE, Ft . SPECTRR 1:HGH- ) |
| | : 3. 00 BFfiCT BLDG Fl. 2il 0 |
| | ~ ~ ~ |
| | ST. LUCl'E OP~V ~. 1 2.80 OBt= 6Q-OUZO Ac.C t-l C=Q.4.+ to 4'i066 |
| | : 2. 60 2 ~ 90 2.00 C3 I I |
| | : 1. 8(3 I CC (C ggF =oSE x 2 1 ~ 60. LU LaJ 1 ~ 00 1 '0 (Z 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 L1 |
| | ~ |
| | IV L1 v |
| | u> 610CQ[~6115 K&511 LC 522 L1 522 L1 L1 O~t cs +50 Qo 602 5 603 604 605 0 |
| | 607 |
| | ~A'sss |
| | ~606, 6O9 SO9 508 |
| | ~5 507 506 504 503 8 9S2 502 954 11.5' 5.063'd Q~ |
| | 952 2 |
| | 'sd 959 |
| | ~ 985 QI A |
| | 953 B |
| | 0'0 956 L'I CO 21 L'I eL1 a 222 O O 516 515 00 L1 522 L1 ~ |
| | 518 cs 517 1 L1 L1 L1 L L'I L p 615 616 617 618 619 621 |
| | "~ |
| | 622 623 624 |
| | )0 525 524 523 522 521 Qsdo 519 54? |
| | '(sea NODE NO. |
| | ~O Qns ELEMENT NO. |
| | CCI 0525 Cl 578 634 3.271' 577 628 C) 527 627 2.375'D 901 629 Q522 C522 el O L1 C1 b)9 632 X2GLOBAL, 1 479 ~630 0522 91O Xl <GLOBAL) 908 900 531 07, so 848 901 902 877 74i |
| | ~70 1 702 |
| | ~801 802 8& ~ 72? |
| | &26 802 100" i 702 776 q'l1 803 )0- 4L 0 Quc Q52 5 07~ |
| | , happ 80S 823 - 724 707 gCt Qssd dos 721 710 ~807 705 808 711 707 973 |
| | ( |
| | ~ |
| | 904 809 810 Its qs~ ~+ 4'ots ~ |
| | + ~819 714 8 903 912 |
| | +MBt1 |
| | ~ O~813 cc 814 817 906 721 A 7OS |
| | .0775-718 O e~. |
| | 0 ~+ &713 ~4'~res q", |
| | 716 717 sL PLAN (EL 39.98') |
| | MODEL-NEUTRON SHIELD FIGURE |
| | |
| | 3'ECTION "A-A" 958 ~959 967 966 45 0.75'5 Fi '59.95 591 41.4'05 590 591 A |
| | EL 42.52'L Q690 Q590 EL 39.98''Ec Secs'8 Y A SECTION "B-B" EL 39 98 907 903 900 I |
| | 5.47'L Q930 Q933 Q 34.51'WF31 A 931 e A 933 B A 935 e 830 SEcZ C 0.875'32 16WF 3S 931 SEcT D e |
| | 933 935 6 ~ |
| | 910 912 0 915 I |
| | Q936 Q0 A 937 e a 939 B A941 e 937 939 941 936 3 938 940 Sic%' SKcg SEc.q "RADIALLYOUTWARD DIRECTION" MODEL-NEUTRON SHIELD FIGURE 19 |
| | |
| | Cross Sections 1-H -1 10' 6 |
| | Total |
| | --- - El as t |
| | " (n,y) 1 c 10 8 |
| | )08 1O6 10 10 10 10 10 10 3, Nev 1-H -1 |
| | |
| | III. SCHEDULE Completion of design July 15, 1977 Material purchase August 15, 1977 Material delivery November 15, 1977 Installation First. scheduled unit shutdown of sufficient duration after material delivery. |
| | |
| | tl s ~ e envue LltN OhlSS3308d fH3l<A000 03AI3038}} |
|
---|
Category:Letter type:L
MONTHYEARL-2024-010, Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3)2024-01-25025 January 2024 Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3) L-2024-004, Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 to 20 Years2024-01-18018 January 2024 Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 to 20 Years L-2024-002, Withdrawal of Proposed Alternative to American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump2024-01-0808 January 2024 Withdrawal of Proposed Alternative to American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump L-2023-173, Quality Assurance Topical Report (FPL-1) Revision 30 Update2023-12-15015 December 2023 Quality Assurance Topical Report (FPL-1) Revision 30 Update L-2023-179, Unusual or Important Environmental Event - Turtle Mortality2023-12-14014 December 2023 Unusual or Important Environmental Event - Turtle Mortality L-2023-168, License Amendment Request Supplement to Revision 2 for the Technical Specifications Conversion to NUREG-1432 Revision 52023-12-12012 December 2023 License Amendment Request Supplement to Revision 2 for the Technical Specifications Conversion to NUREG-1432 Revision 5 L-2023-155, Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-06542023-11-28028 November 2023 Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, L-2023-162, Response to 50.69 2nd Round of Rals2023-11-21021 November 2023 Response to 50.69 2nd Round of Rals L-2023-131, Subsequent License Renewal Application - Second Annual Update2023-09-28028 September 2023 Subsequent License Renewal Application - Second Annual Update L-2023-136, Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-26026 September 2023 Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-122, Corrections to the 2022 Annual Radiological Environmental Operating Report2023-09-20020 September 2023 Corrections to the 2022 Annual Radiological Environmental Operating Report L-2023-127, Correction to the 2022 Annual Radioactive Effluent Release Report2023-09-18018 September 2023 Correction to the 2022 Annual Radioactive Effluent Release Report L-2023-113, Correction to the 2020 Annual Radiological Environmental Operating Report2023-09-14014 September 2023 Correction to the 2020 Annual Radiological Environmental Operating Report L-2023-118, Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-11011 September 2023 Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-108, Report of 10 CFR 50.59 Plant Changes2023-09-11011 September 2023 Report of 10 CFR 50.59 Plant Changes L-2023-112, Corrections to the 2021 Annual Radioactive Effluent Release Report2023-09-0606 September 2023 Corrections to the 2021 Annual Radioactive Effluent Release Report L-2023-107, Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.42023-09-0606 September 2023 Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.4 L-2023-114, Proposed Turkey Point Units 6 and 7; Seabrook Station; Point Beach Units 1 and 2 - Official Service List Update2023-08-17017 August 2023 Proposed Turkey Point Units 6 and 7; Seabrook Station; Point Beach Units 1 and 2 - Official Service List Update L-2023-098, and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22023-08-0707 August 2023 and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 L-2023-105, Preparation and Scheduling of Operator Licensing Examinations2023-08-0303 August 2023 Preparation and Scheduling of Operator Licensing Examinations L-2023-099, Pump Relief Request 10 (PR-10), One-Time Request for an Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump2023-07-26026 July 2023 Pump Relief Request 10 (PR-10), One-Time Request for an Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump L-2023-102, Relief Request PSL2-15-RR-01, Proposed Alternative to ASME Section XI Code Examination Requirements for Reactor Vessel Bottom Area and Piping in Covered Trenches2023-07-26026 July 2023 Relief Request PSL2-15-RR-01, Proposed Alternative to ASME Section XI Code Examination Requirements for Reactor Vessel Bottom Area and Piping in Covered Trenches L-2023-097, Subsequent License Renewal Application Revision 1 - Supplement 62023-07-13013 July 2023 Subsequent License Renewal Application Revision 1 - Supplement 6 L-2023-076, In-Service Inspection Program Owner'S Activity Report (OAR-1)2023-07-11011 July 2023 In-Service Inspection Program Owner'S Activity Report (OAR-1) L-2023-087, Florida Power & Light/Nextera Energy, Results of the Safety Culture Program Effectiveness Review, March 20, 2023 (ADAMS Accession No. ML22340A452)2023-06-29029 June 2023 Florida Power & Light/Nextera Energy, Results of the Safety Culture Program Effectiveness Review, March 20, 2023 (ADAMS Accession No. ML22340A452) L-2023-082, Subsequent License Renewal Application Revision 1, Supplement 52023-06-14014 June 2023 Subsequent License Renewal Application Revision 1, Supplement 5 L-2023-074, Addendum to 2021 Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation Ctsfsi) Financial Assurance Update2023-06-0202 June 2023 Addendum to 2021 Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation Ctsfsi) Financial Assurance Update L-2023-071, NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 29 and Florida Power and Light Company Quality Assurance Program Description for 10 CFR Part 52 Licenses (FPL-2) Revision 11, Annual Submittal2023-05-22022 May 2023 NextEra Energy Quality Assurance Topical Report (FPL-1) Revision 29 and Florida Power and Light Company Quality Assurance Program Description for 10 CFR Part 52 Licenses (FPL-2) Revision 11, Annual Submittal L-2023-059, Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 4 Supplemental Response2023-04-21021 April 2023 Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 4 Supplemental Response L-2023-055, 2022 Annual Environmental Operating Report2023-04-12012 April 2023 2022 Annual Environmental Operating Report L-2023-041, Annual Radiological Environmental Operating Report for Calendar Year 20222023-04-0404 April 2023 Annual Radiological Environmental Operating Report for Calendar Year 2022 L-2023-051, Report of 10 CFR 50.59 Plant Changes2023-04-0404 April 2023 Report of 10 CFR 50.59 Plant Changes L-2023-021, Units, 1 and 2, Turkey Point, Units 3 and 4, Seabrook Station and Point Beach, Units 1 and 2 - Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation (ISFSI) Financial Assurance Update2023-03-28028 March 2023 Units, 1 and 2, Turkey Point, Units 3 and 4, Seabrook Station and Point Beach, Units 1 and 2 - Decommissioning Funding Status Reports / Independent Spent Fuel Storage Installation (ISFSI) Financial Assurance Update L-2023-042, Periodic Update of Population Data within 10 and 50 Miles of the Plant2023-03-27027 March 2023 Periodic Update of Population Data within 10 and 50 Miles of the Plant L-2023-026, Subsequent License Renewal Application - Aging Management Requests for Additional Information Set 42023-03-27027 March 2023 Subsequent License Renewal Application - Aging Management Requests for Additional Information Set 4 L-2023-028, and Point Beach Units 1 and 2, 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications2023-03-27027 March 2023 and Point Beach Units 1 and 2, 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications L-2023-025, Fleet Relief Request (Frr) 23-01, Proposed Alternative to ASME Section XI Authorizing Implementation of ASME Code Case N-752-12023-03-15015 March 2023 Fleet Relief Request (Frr) 23-01, Proposed Alternative to ASME Section XI Authorizing Implementation of ASME Code Case N-752-1 L-2023-029, and Point Beach Units 1 and 2 Nuclear Property Insurance - 10 CFR 50.54(w)(3)2023-03-10010 March 2023 and Point Beach Units 1 and 2 Nuclear Property Insurance - 10 CFR 50.54(w)(3) L-2023-039, Cycle 27 Core Operating Limits Report2023-03-0707 March 2023 Cycle 27 Core Operating Limits Report L-2023-032, 2022 Annual Radioactive Effluent Release Report2023-02-28028 February 2023 2022 Annual Radioactive Effluent Release Report L-2023-038, 2022 Annual Operating Report2023-02-28028 February 2023 2022 Annual Operating Report L-2023-016, Radiological Emergency Plan - Revision 74 Report of Changes to Emergency Plan2023-02-15015 February 2023 Radiological Emergency Plan - Revision 74 Report of Changes to Emergency Plan L-2023-019, Annual Summary of Commitment Changes Implemented Without Prior NRC Notification for Calendar Year 20222023-02-15015 February 2023 Annual Summary of Commitment Changes Implemented Without Prior NRC Notification for Calendar Year 2022 L-2023-009, Owner'S Activity Report2023-01-31031 January 2023 Owner'S Activity Report L-2022-188, Unusual or Important Environmental Event - Turtle Mortality2022-12-19019 December 2022 Unusual or Important Environmental Event - Turtle Mortality L-2022-185, Turkey Points, Units 3 & 4; Seabrook Station; and Point Beach, Units 1 and 2 - Supplement to License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22022-12-0909 December 2022 Turkey Points, Units 3 & 4; Seabrook Station; and Point Beach, Units 1 and 2 - Supplement to License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 L-2022-175, Application to Adopt 10 CPR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2022-12-0202 December 2022 Application to Adopt 10 CPR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2022-180, CFR 140.21 Licensee Guarantees of Payment of Deferred Premiums2022-11-0909 November 2022 CFR 140.21 Licensee Guarantees of Payment of Deferred Premiums L-2022-165, Subsequent License Renewal Application - Aging Management Request for Additional Information (RAI) 4.3.1-1a(second Round) - Class 1 Fatigue Response2022-10-26026 October 2022 Subsequent License Renewal Application - Aging Management Request for Additional Information (RAI) 4.3.1-1a(second Round) - Class 1 Fatigue Response L-2022-160, Station,, Point Beach Units 1 and 2, License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22022-10-0404 October 2022 Station,, Point Beach Units 1 and 2, License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 2024-01-08
[Table view] Category:Report
MONTHYEARL-2023-131, Subsequent License Renewal Application - Second Annual Update2023-09-28028 September 2023 Subsequent License Renewal Application - Second Annual Update L-2023-076, In-Service Inspection Program Owner'S Activity Report (OAR-1)2023-07-11011 July 2023 In-Service Inspection Program Owner'S Activity Report (OAR-1) L-2023-028, and Point Beach Units 1 and 2, 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications2023-03-27027 March 2023 and Point Beach Units 1 and 2, 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications L-2023-009, Owner'S Activity Report2023-01-31031 January 2023 Owner'S Activity Report ML22227A0532022-08-15015 August 2022 Biological Opinion for the Continued Operation of St. Lucie Nuclear Power Plant ML22124A0112022-04-30030 April 2022 Scoping Summary Report - Final L-2022-046, Subsequent License Renewal Application Revision - Documents WCAP-18623-P/NP Revision 1 Submittal2022-04-13013 April 2022 Subsequent License Renewal Application Revision - Documents WCAP-18623-P/NP Revision 1 Submittal L-2022-015, Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld - RAI2022-01-14014 January 2022 Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld - RAI L-2022-011, Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld2022-01-12012 January 2022 Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld ML22010A0942022-01-0404 January 2022 Trp 29 St. Lucie SLRA - Tank Breakout L-2021-178, Report of 10 CFR 50.59 Plant Changes2021-11-0808 November 2021 Report of 10 CFR 50.59 Plant Changes L-2021-142, Westinghouse Report LTR-REA-21-1-NP, Revision 1, St. Lucie Nuclear Plant, Units 1 and 2, Subsequent License Renewal: Unit 1 Reactor Vessel, Vessel Support, and Bioshield Concrete Exposure Data, May 26, 20212021-08-0303 August 2021 Westinghouse Report LTR-REA-21-1-NP, Revision 1, St. Lucie Nuclear Plant, Units 1 and 2, Subsequent License Renewal: Unit 1 Reactor Vessel, Vessel Support, and Bioshield Concrete Exposure Data, May 26, 2021 ML19252A4002019-09-0909 September 2019 FPL to NRC, Notification of Smalltooth Sawfish Capture at St. Lucie L-2019-010, Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds2019-03-19019 March 2019 Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds ML18096B3952018-04-0606 April 2018 Exhibit III Estimate of Construction Costs and Exhibit IV Technical Qualifications of Contractors ML18088B1952018-03-29029 March 2018 Hutchinson Island Plant Units 1 and 2 - Chapter 9, Auxiliary Systems and Chapter 10, Steam and Power Conversion System ML18088B1942018-03-29029 March 2018 Hutchinson Island Plant Units 1 and 2 - Chapter 11, Radioactive Waste Management System, Chapter 12, Radiation Protection, and Chapter 13, Conduct of Operations ML18088A0942018-03-29029 March 2018 Unit II Plants ECCS Performance Results L-2017-173, Environmental Protection Plan Report, Unusual or Important Environmental Event - Turtle Mortality - 09/11/2017 Event2017-09-28028 September 2017 Environmental Protection Plan Report, Unusual or Important Environmental Event - Turtle Mortality - 09/11/2017 Event L-2018-081, Kld Engineering, Pc - 2017 Population Update Analysis2017-09-20020 September 2017 Kld Engineering, Pc - 2017 Population Update Analysis L-2017-117, Submittal of SL2-23 Outage, Owner'S Activity Report, Form OAR-12017-06-20020 June 2017 Submittal of SL2-23 Outage, Owner'S Activity Report, Form OAR-1 L-2018-015, Plan of Study 316(b) Implementation2017-04-28028 April 2017 Plan of Study 316(b) Implementation L-2017-015, PWROG-15105-NP PA-MSC-1288 PWR Rv Internals Cold-Work Assessment, Materials Committee.2016-04-30030 April 2016 PWROG-15105-NP PA-MSC-1288 PWR Rv Internals Cold-Work Assessment, Materials Committee. ML16084A6162016-03-24024 March 2016 Submittal of Biological Opinion for the Continued Operation of St. Lucie Nuclear Power Plant, Units 1 and 2 in St. Lucie County, Florida ML16063A0072016-02-26026 February 2016 Participation in Additional Work Under the Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0 PA-MSC-0983 R2 Cafeteria Task 8 and Acceptance Criteria for Measurement Of.. ML15352A0532016-01-0707 January 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f) Seismic Hazard Revaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights L-2015-297, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, and Submittal of Site FLEX Final Integrated Plan2015-12-10010 December 2015 Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, and Submittal of Site FLEX Final Integrated Plan L-2015-300, ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report.2015-11-30030 November 2015 ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report. ML15314A1602015-10-29029 October 2015 St. Lucie, Units 1 and 2 - License Renewal Commitment, Submittal of Pressurizer Surge Line Welds Inspection Program L-2015-221, Report of 10 CFR 50.59 Plant Changes2015-10-16016 October 2015 Report of 10 CFR 50.59 Plant Changes ML15240A1542015-09-0808 September 2015 Staff Observations of Sump Strainer Head Loss Testing at Alden Laboratory for Generic Safety Issue 191 L-2015-206, ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-2062015-07-31031 July 2015 ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-206 L-2015-177, Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 12015-06-30030 June 2015 Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 1 L-2015-143, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2015-05-14014 May 2015 Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events L-2015-272, 1301103.401, Revision 0, Flaw Tolerance Evaluation of St. Lucie Surge Line Welds Using ASME Code Section XI, Appendix L, May 20152015-05-0808 May 2015 1301103.401, Revision 0, Flaw Tolerance Evaluation of St. Lucie Surge Line Welds Using ASME Code Section XI, Appendix L, May 2015 L-2016-052, TN-5696-00-02, Revision 0, Technical Note Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (Foi) Versus Alloys 600 and 182.2015-03-31031 March 2015 TN-5696-00-02, Revision 0, Technical Note Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (Foi) Versus Alloys 600 and 182. L-2015-091, ANP-3396NP, Revision 0, Fuel Transition Supplemental Information to Support the LAR2015-03-31031 March 2015 ANP-3396NP, Revision 0, Fuel Transition Supplemental Information to Support the LAR L-2015-093, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report2015-03-24024 March 2015 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report ML15083A2642015-03-10010 March 2015 St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report L-2015-048, St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report2015-03-10010 March 2015 St. Lucie, Units 1 and 2 - Submittal of Revision 0 to FPL-072-PR-002, Flooding Hazards Reevaluation Report L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Cover Page to Page 1092015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Cover Page to Page 109 L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-92015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-9 L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End2015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End ML15083A2652015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-9 ML15083A2662015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End L-2014-366, ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report2014-12-31031 December 2014 ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report ML14338A5552014-12-0404 December 2014 NRC-2013-TN3079-NRC 2014 St. Lucie License Renewal ML14338A5542014-12-0404 December 2014 NRC-2013- TN2986-NRC 2014 St. Lucie L-2014-125, Report of 10 CFR 50.59 Plant Changes2014-05-0606 May 2014 Report of 10 CFR 50.59 Plant Changes ML13360A2022013-12-12012 December 2013 EPA Echo Report Martin County, Fl 2023-09-28
[Table view] Category:Technical
MONTHYEARL-2022-046, Subsequent License Renewal Application Revision - Documents WCAP-18623-P/NP Revision 1 Submittal2022-04-13013 April 2022 Subsequent License Renewal Application Revision - Documents WCAP-18623-P/NP Revision 1 Submittal L-2022-015, Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld - RAI2022-01-14014 January 2022 Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld - RAI L-2022-011, Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld2022-01-12012 January 2022 Relief Request Number 20 - Request for an Alternative to the Requirements of the ASME Code for Examination of Reactor Vessel Closure Head Control Element Drive Mechanism (CEDM) Housing 27 Canopy Seal Weld L-2021-142, Westinghouse Report LTR-REA-21-1-NP, Revision 1, St. Lucie Nuclear Plant, Units 1 and 2, Subsequent License Renewal: Unit 1 Reactor Vessel, Vessel Support, and Bioshield Concrete Exposure Data, May 26, 20212021-08-0303 August 2021 Westinghouse Report LTR-REA-21-1-NP, Revision 1, St. Lucie Nuclear Plant, Units 1 and 2, Subsequent License Renewal: Unit 1 Reactor Vessel, Vessel Support, and Bioshield Concrete Exposure Data, May 26, 2021 L-2019-010, Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds2019-03-19019 March 2019 Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Welds ML18096B3952018-04-0606 April 2018 Exhibit III Estimate of Construction Costs and Exhibit IV Technical Qualifications of Contractors ML18088B1942018-03-29029 March 2018 Hutchinson Island Plant Units 1 and 2 - Chapter 11, Radioactive Waste Management System, Chapter 12, Radiation Protection, and Chapter 13, Conduct of Operations ML18088B1952018-03-29029 March 2018 Hutchinson Island Plant Units 1 and 2 - Chapter 9, Auxiliary Systems and Chapter 10, Steam and Power Conversion System ML18088A0942018-03-29029 March 2018 Unit II Plants ECCS Performance Results L-2018-081, Kld Engineering, Pc - 2017 Population Update Analysis2017-09-20020 September 2017 Kld Engineering, Pc - 2017 Population Update Analysis L-2018-015, Plan of Study 316(b) Implementation2017-04-28028 April 2017 Plan of Study 316(b) Implementation L-2017-015, PWROG-15105-NP PA-MSC-1288 PWR Rv Internals Cold-Work Assessment, Materials Committee.2016-04-30030 April 2016 PWROG-15105-NP PA-MSC-1288 PWR Rv Internals Cold-Work Assessment, Materials Committee. ML16084A6162016-03-24024 March 2016 Submittal of Biological Opinion for the Continued Operation of St. Lucie Nuclear Power Plant, Units 1 and 2 in St. Lucie County, Florida ML16063A0072016-02-26026 February 2016 Participation in Additional Work Under the Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0 PA-MSC-0983 R2 Cafeteria Task 8 and Acceptance Criteria for Measurement Of.. L-2015-300, ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report.2015-11-30030 November 2015 ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report. L-2015-206, ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-2062015-07-31031 July 2015 ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-206 L-2015-177, Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 12015-06-30030 June 2015 Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 1 L-2015-272, 1301103.401, Revision 0, Flaw Tolerance Evaluation of St. Lucie Surge Line Welds Using ASME Code Section XI, Appendix L, May 20152015-05-0808 May 2015 1301103.401, Revision 0, Flaw Tolerance Evaluation of St. Lucie Surge Line Welds Using ASME Code Section XI, Appendix L, May 2015 L-2016-052, TN-5696-00-02, Revision 0, Technical Note Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (Foi) Versus Alloys 600 and 182.2015-03-31031 March 2015 TN-5696-00-02, Revision 0, Technical Note Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (Foi) Versus Alloys 600 and 182. L-2015-091, ANP-3396NP, Revision 0, Fuel Transition Supplemental Information to Support the LAR2015-03-31031 March 2015 ANP-3396NP, Revision 0, Fuel Transition Supplemental Information to Support the LAR L-2015-093, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report2015-03-24024 March 2015 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Cover Page to Page 1092015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Cover Page to Page 109 L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-92015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-9 L-2015-048, to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End2015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End ML15083A2652015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Table 2-1 Through Figure 4-9 ML15083A2662015-02-0606 February 2015 to FPL-072-PR-002, Flooding Hazards Reevaluation Report for St. Lucie Nuclear Power Plant Units 1 & 2. Figure 4-10 Through the End L-2014-366, ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report2014-12-31031 December 2014 ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report ML14149A1952013-02-0404 February 2013 Pacific Northwest National Laboratory Technical Letter Report for Evaluation of Alternative to 10 CFR 50.55a(G)(6)ll)(F)(4) for Limitations to Volumetric Examination of Dissimilar Metal Welds ML12340A3522012-11-30030 November 2012 St. Lucie, Unit 1, 12Q4116-RPT-001, Rev. 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Force Recommendation 2.3: Seismic L-2012-427, Q4116-R-002, Rev. 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Force Recommendation 2.3: Seismic2012-11-30030 November 2012 Q4116-R-002, Rev. 0, Seismic Walkdown Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Force Recommendation 2.3: Seismic ML12097A2682012-04-17017 April 2012 Biological Assessment for Formal Section 7 Consultation at the St. Lucie Plant, Units 1 and 2 L-2012-072, ANP-2903Q2(NP), Rev 0, St. Lucie, Unit 1 - EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding, Attachment 32012-02-29029 February 2012 ANP-2903Q2(NP), Rev 0, St. Lucie, Unit 1 - EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding, Attachment 3 ML12061A2492012-02-29029 February 2012 ANP-3067, Rev. 1, St. Lucie, Unit 1 EPU - Information to Support NRC Review of RCS Depressurization with Pressurizer Overfill L-2012-072, ANP-3067, Rev. 1, St. Lucie, Unit 1 EPU - Information to Support NRC Review of RCS Depressurization with Pressurizer Overfill2012-02-29029 February 2012 ANP-3067, Rev. 1, St. Lucie, Unit 1 EPU - Information to Support NRC Review of RCS Depressurization with Pressurizer Overfill L-2011-471, ANP-3057(NP), Revision 0, St. Lucie Unit 1 EPU - Responses to NRC Questions SRXB-58, SRXB-59, and SRXB-60.2011-10-31031 October 2011 ANP-3057(NP), Revision 0, St. Lucie Unit 1 EPU - Responses to NRC Questions SRXB-58, SRXB-59, and SRXB-60. L-2011-389, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request2011-09-22022 September 2011 Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request L-2011-311, ANP-3019NP, Revision 0, St. Lucie Unit 1 EPU - Information to Support NRC Review of Steam Generator Tube Rupture, Attachment 22011-08-31031 August 2011 ANP-3019NP, Revision 0, St. Lucie Unit 1 EPU - Information to Support NRC Review of Steam Generator Tube Rupture, Attachment 2 L-2011-342, ANP-3028(NP), Revision 0, St. Lucie Plant, Unit 1 EPU RAIs - Nuclear Performance & Code (Snpb)2011-08-31031 August 2011 ANP-3028(NP), Revision 0, St. Lucie Plant, Unit 1 EPU RAIs - Nuclear Performance & Code (Snpb) L-2011-228, 103-87735, Heated Water Plan of Study2011-06-30030 June 2011 103-87735, Heated Water Plan of Study L-2011-206, ANP-2903(NP), Revision 1, St. Lucie Nuclear Plant, Unit 1 - EPU Cycle Realistic Large Break LOCA Summary Report with ZR-4 Fuel Cladding, Attachment 72011-05-31031 May 2011 ANP-2903(NP), Revision 1, St. Lucie Nuclear Plant, Unit 1 - EPU Cycle Realistic Large Break LOCA Summary Report with ZR-4 Fuel Cladding, Attachment 7 ML11153A0492011-05-31031 May 2011 ANP-3000(NP), Rev. 0, St. Lucie Nuclear, Unit 1 - EPU-Information to Support License Amendment Request, Attachment 6 L-2011-206, ANP-3000(NP), Rev. 0, St. Lucie Nuclear, Unit 1 - EPU-Information to Support License Amendment Request, Attachment 62011-05-31031 May 2011 ANP-3000(NP), Rev. 0, St. Lucie Nuclear, Unit 1 - EPU-Information to Support License Amendment Request, Attachment 6 L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix D, List of Key Acronyms2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix D, List of Key Acronyms L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix B Additional Codes and Methods2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix B Additional Codes and Methods ML1107302992011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Licensing Report L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix a Safety Evaluation Report Compliance2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix a Safety Evaluation Report Compliance L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix F, Camereron Ultrasonics Engineering Reports, Cover2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix F, Camereron Ultrasonics Engineering Reports, Cover L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix E, Supplement to Licensing Report Section 2.4.1 Reactor Protection, Engineered Safety Feature Actuation, and Control Systems2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix E, Supplement to Licensing Report Section 2.4.1 Reactor Protection, Engineered Safety Feature Actuation, and Control Systems L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix G, Holtec Report No. HI-2104753, St. Lucie Unit 2 Criticality Analysis for EPU and Non-EPU2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5; Appendix G, Holtec Report No. HI-2104753, St. Lucie Unit 2 Criticality Analysis for EPU and Non-EPU L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5, Appendix C, Grid Stability Analysis and System Impact Study for St. Lucie Plant with Proposed EPU2011-02-25025 February 2011 St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 5, Appendix C, Grid Stability Analysis and System Impact Study for St. Lucie Plant with Proposed EPU 2022-04-13
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARL-2023-155, Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-06542023-11-28028 November 2023 Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, L-2023-162, Response to 50.69 2nd Round of Rals2023-11-21021 November 2023 Response to 50.69 2nd Round of Rals L-2023-118, Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-11011 September 2023 Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-098, and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22023-08-0707 August 2023 and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 L-2023-059, Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 4 Supplemental Response2023-04-21021 April 2023 Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 4 Supplemental Response L-2023-026, Subsequent License Renewal Application - Aging Management Requests for Additional Information Set 42023-03-27027 March 2023 Subsequent License Renewal Application - Aging Management Requests for Additional Information Set 4 ML23013A2032023-01-13013 January 2023 RAI Set 4 Draft Response L-2022-165, Subsequent License Renewal Application - Aging Management Request for Additional Information (RAI) 4.3.1-1a(second Round) - Class 1 Fatigue Response2022-10-26026 October 2022 Subsequent License Renewal Application - Aging Management Request for Additional Information (RAI) 4.3.1-1a(second Round) - Class 1 Fatigue Response L-2022-156, Correction to Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 1a Response2022-09-19019 September 2022 Correction to Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 1a Response L-2022-143, Subsequent License Renewal Application - Aging Management Requests for Additional Information Set 1A Response2022-09-0808 September 2022 Subsequent License Renewal Application - Aging Management Requests for Additional Information Set 1A Response L-2022-115, Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 3 Response and Submittal of Superseded Response for One Set 2 RAI and One Supplement 1 Attachment2022-08-0909 August 2022 Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 3 Response and Submittal of Superseded Response for One Set 2 RAI and One Supplement 1 Attachment L-2022-108, Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 2 Response2022-07-11011 July 2022 Subsequent License Renewal Application - Aging Management Requests for Additional Information (RAI) Set 2 Response L-2022-075, Subsequent License Renewal Application-Aging Management Requests for Additional Information (RAI) Set 1A Response and Request for Confirmation of Information (RCI) Set 1 Response2022-06-13013 June 2022 Subsequent License Renewal Application-Aging Management Requests for Additional Information (RAI) Set 1A Response and Request for Confirmation of Information (RCI) Set 1 Response L-2022-000, License Amendment Request for the Technical Specifications Conversion to NUREG- 1432 Revision 5 - Request for Supplemental Information (Rsi) Response2022-01-19019 January 2022 License Amendment Request for the Technical Specifications Conversion to NUREG- 1432 Revision 5 - Request for Supplemental Information (Rsi) Response L-2021-105, Response to Request for Additional Information for St. Lucie License Amendment Request to Allow Risk Informed Completion Times (RICT) for the 120-Volt AC Instrument Bus Requirements2021-05-12012 May 2021 Response to Request for Additional Information for St. Lucie License Amendment Request to Allow Risk Informed Completion Times (RICT) for the 120-Volt AC Instrument Bus Requirements L-2021-065, Response to Request for Additional Information. Relief Request Number RR 15, Extension of St. Lucie Unit 2 RPV Welds from 10 to 20 Years2021-04-0101 April 2021 Response to Request for Additional Information. Relief Request Number RR 15, Extension of St. Lucie Unit 2 RPV Welds from 10 to 20 Years L-2020-165, Supplement to Updated Final Response to NRC Generic Letter 2004-022020-12-0404 December 2020 Supplement to Updated Final Response to NRC Generic Letter 2004-02 L-2020-094, Supplemental Response to Request for Additional Information Regarding License Amendment Request to Modify the Reactor Coolant Pump (RCP) Flywheel Inspection Program Requirements2020-06-26026 June 2020 Supplemental Response to Request for Additional Information Regarding License Amendment Request to Modify the Reactor Coolant Pump (RCP) Flywheel Inspection Program Requirements L-2020-061, Response to Request for Additional Information Regarding License Amendment Request to Modify the Reactor Coolant Pump (RCP) Flywheel Inspection Program Requirements2020-04-30030 April 2020 Response to Request for Additional Information Regarding License Amendment Request to Modify the Reactor Coolant Pump (RCP) Flywheel Inspection Program Requirements ML20015A0282020-01-14014 January 2020 NMFS to NRC, Concurrence with Interim Response to Requests for Additional Information for St. Lucie Endangered Species Act Section 7 Consultation L-2019-164, Response to Request for Additional Information Regarding License Amendment Request to Relocate the MOV Thermal Overload Protection Bypass Devices Requirements to Licensee Control2019-09-11011 September 2019 Response to Request for Additional Information Regarding License Amendment Request to Relocate the MOV Thermal Overload Protection Bypass Devices Requirements to Licensee Control L-2019-166, Refueling Outage SL2-24 Steam Generator Tube Inspection Report RAI Response2019-08-21021 August 2019 Refueling Outage SL2-24 Steam Generator Tube Inspection Report RAI Response L-2019-153, Exigent Technical Specification Amendment Request Supplemental RAI Reply One-Time Allowed Outage Time Extension for Inoperable EDG2019-07-25025 July 2019 Exigent Technical Specification Amendment Request Supplemental RAI Reply One-Time Allowed Outage Time Extension for Inoperable EDG L-2019-149, Exigent Technical Specification Amendment Request RAI Reply One-Time Allowed Outage Time Extension for Inoperable EDG2019-07-24024 July 2019 Exigent Technical Specification Amendment Request RAI Reply One-Time Allowed Outage Time Extension for Inoperable EDG L-2019-118, Response to Request for Additional Information Regarding License Amendment Request to Allow Performance of Selected Emergency Diesel Generator (EDG) Surveillance Requirements (Srs) During Power Operation2019-06-28028 June 2019 Response to Request for Additional Information Regarding License Amendment Request to Allow Performance of Selected Emergency Diesel Generator (EDG) Surveillance Requirements (Srs) During Power Operation L-2019-107, License Amendment Request - Iodine Removal System Elimination2019-05-17017 May 2019 License Amendment Request - Iodine Removal System Elimination L-2019-056, Inservice Inspection Plan RAI Reply, Fifth Ten-Year Interval Unit 1 Relief Request2019-03-0707 March 2019 Inservice Inspection Plan RAI Reply, Fifth Ten-Year Interval Unit 1 Relief Request L-2018-205, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies2018-11-15015 November 2018 Response to Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies L-2018-153, Supplemental Information for License Amendment Request to Reduce the Number of Control Element Assemblies2018-08-17017 August 2018 Supplemental Information for License Amendment Request to Reduce the Number of Control Element Assemblies L-16-001, Units. 3 and 4, Response to Request for Supplemental Information Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools.2018-05-24024 May 2018 Units. 3 and 4, Response to Request for Supplemental Information Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools. L-2018-068, Florida Power and Light Co. - Response to Request for Additional Information Re Decommissioning Funding Plan Updates for Independent Spent Fuel Storage Installations2018-04-0303 April 2018 Florida Power and Light Co. - Response to Request for Additional Information Re Decommissioning Funding Plan Updates for Independent Spent Fuel Storage Installations L-77-291, Attachment a to L-77-291 Response to NRC Questions of May 24, 19772018-03-29029 March 2018 Attachment a to L-77-291 Response to NRC Questions of May 24, 1977 L-2018-040, Response to Request for Additional Information Regarding License Amendment Request to Add New Required Actions for an Inoperable Auxiliary Feedwater Pump Steam Supply2018-02-14014 February 2018 Response to Request for Additional Information Regarding License Amendment Request to Add New Required Actions for an Inoperable Auxiliary Feedwater Pump Steam Supply L-2018-006, Third Response to Request for Additional Information Regarding License Amendment Request to Adopt Risk Informed Completion Times TSTF-505, Revision 1, Provide Risk-Risk Informed Extended Completion Times - RITSTF Initiative 4b2018-02-0101 February 2018 Third Response to Request for Additional Information Regarding License Amendment Request to Adopt Risk Informed Completion Times TSTF-505, Revision 1, Provide Risk-Risk Informed Extended Completion Times - RITSTF Initiative 4b L-2017-210, Updated Final Response to NRC Generic Letter 2004-022017-12-20020 December 2017 Updated Final Response to NRC Generic Letter 2004-02 L-2017-216, License Renewal Commitments: Reactor Vessel Internals Aging Management Plan, Clarification of Responses to RAI 1 and RAI 62017-12-19019 December 2017 License Renewal Commitments: Reactor Vessel Internals Aging Management Plan, Clarification of Responses to RAI 1 and RAI 6 L-2017-209, Response to Request for Additional Information Regarding Fifth 10-Year Inservice Testing (IST) Program Interval Relief Request PR-012017-11-30030 November 2017 Response to Request for Additional Information Regarding Fifth 10-Year Inservice Testing (IST) Program Interval Relief Request PR-01 L-2017-159, Response to Request for Additional Information Regarding Inservice Inspection Plan, Fifth Ten-Year Interval Unit 1 Relief Request No. 3, Revision 02017-09-13013 September 2017 Response to Request for Additional Information Regarding Inservice Inspection Plan, Fifth Ten-Year Interval Unit 1 Relief Request No. 3, Revision 0 L-2017-116, Response to Request for Additional Information Regarding License Amendment Request to Revise the Technical Specifications (TS) for the Reactor Protection System (RPS) Power Rate-of-Change Instrumentation and Add New TS 3.0.52017-07-0303 July 2017 Response to Request for Additional Information Regarding License Amendment Request to Revise the Technical Specifications (TS) for the Reactor Protection System (RPS) Power Rate-of-Change Instrumentation and Add New TS 3.0.5 L-2016-220, License Amendment Request EDG Day Tank Fuel Volume Change2016-12-0505 December 2016 License Amendment Request EDG Day Tank Fuel Volume Change L-2016-188, Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools, Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)2016-11-0303 November 2016 Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools, Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) L-2016-153, Response to Request for Additional Information Regarding License Amendment Request for Biological Opinion License Changes2016-08-11011 August 2016 Response to Request for Additional Information Regarding License Amendment Request for Biological Opinion License Changes L-2016-135, Second Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B2016-07-22022 July 2016 Second Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B L-2016-143, Response to Request for Additional Information for the Proposed Technical Specification Change to Remove the 10 Year Sediment Cleaning of the Fuel Oil Storage Tank and Relocate to Licensee-Controlled Documents2016-07-15015 July 2016 Response to Request for Additional Information for the Proposed Technical Specification Change to Remove the 10 Year Sediment Cleaning of the Fuel Oil Storage Tank and Relocate to Licensee-Controlled Documents L-2016-114, Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 48.2016-07-0808 July 2016 Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 48. L-2016-102, RAI Reply - License Amendment Request, Containment Vacuum Gothic Analyses and Conforming Changes2016-05-0606 May 2016 RAI Reply - License Amendment Request, Containment Vacuum Gothic Analyses and Conforming Changes L-2016-104, RAI Reply - Application for Technical Specification Change Regarding Moderator Temperature Coefficient (Mtc) Surveillance Test Elimination at the End of Cycle2016-05-0606 May 2016 RAI Reply - Application for Technical Specification Change Regarding Moderator Temperature Coefficient (Mtc) Surveillance Test Elimination at the End of Cycle L-2016-096, Inservice Inspection Plan, RAI Reply for Fourth Ten-Year Interval Relief Request No. 10, Revision 02016-04-21021 April 2016 Inservice Inspection Plan, RAI Reply for Fourth Ten-Year Interval Relief Request No. 10, Revision 0 L-2016-084, Response to Request for Additional Information Regarding License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-422, Revision 2, Change in Technical Specifications End States.2016-04-20020 April 2016 Response to Request for Additional Information Regarding License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-422, Revision 2, Change in Technical Specifications End States. L-2016-040, License Renewal Commitments, Reactor Vessel Internals Aging Management Plan, Response to Request for Additional Information2016-02-26026 February 2016 License Renewal Commitments, Reactor Vessel Internals Aging Management Plan, Response to Request for Additional Information 2023-09-11
[Table view] |
Text
U.S NUCLEAR REGULATORY COMM ~ <ON DOCKET NUM BEA NRC FOPV 194 (2.g B)
FILE NUMBER NRC DISTRIBUTION FOA PART 50 DOCKET MATERIAL FROM: DATE OF DOCUMENT TO: Light Florida Power & Co 8/3/77 Mr. Don K. Davis Miami, Fla. DATE AECEIVED Ri Ei Uhrig 8/8/77 QNOTORIZEO PAOP INPUT FORM NUMBEA OF COPIES RECEIVED CfLETTE R RIUNCLASSIF IEO CBCOP Y DESCRIPTION ENCI OSURE Consists of requested additional information concerning the installation of additional neutron shielding in the reactor vessel cavity at Unit No 1 ~ .~~~
i $ Qjggj( ~L) Il ( ~~egQ~L~Pi (1>>P )
PLANT NAME:
I)0 uIT KINK St Lucie Unit No~ 1 R JL 8/8/77 FOR ACTION/INFORMATION ENVIRHNMENTAL ASSIGNED AD: Ve MOORE LTR BRANCH CHIEF:
MANAGER PROJECT MANAGER:
ENSING ASSXSTAiVT: LICENSING ASSISTANT:
Be HARLESS INTERNAL D ISTRI BUTION TFMS SAFETY PLANT SYSTEMS SITE SAFETY &
HEINEMAN TEDESCO ENVIRON ANALYSIS E E BENAROYA DENTON & MULLER ENGINE ERTNG IPPOLITO ENVIRO TECH ERNST OPERATING REACTORS BALLARD B OD BAER B ER GAMMILL 2 CHECK SITE ANALYSIS VOLLMER AT I BUNCH J~ COLLINS A TZMAN ERG KREGER EXTERNAL DISTRIBUTION 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 <
FLORIDA POWER 5 LIGHT COMPANY QQC~i
,~gI tq1Q August 3I 1977 X,-77-245 goal 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 Docket No. 50-335 l
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 (See Pigure 17).
'ates
- 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 Ishielding analysis why additional neutron 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 7 n/cm 2 -sec) and 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 +
the dead load D includes the weight of support structures, E'here 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 saved has 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 Avg. Gamma Estimated Exposure Dose Rate Dose Rate Exposure Rate (mRem/hr) (mr/hr) (man-hr/wk) (man-rem/yr)
A. OUTSIDE CONTAINMENT
- l. No Shield 2.5 2.5 1.2
- 2. Shield 0;5 0.5 .0. 21 B. INSIDE CONTAINMENT
- l. Operating Floor No Shield 2500 500 0.7 109 Shield 0.7 150 100 9.1
- 2. Other Areas No Shield 100 50 1.4 11.
Shield 25 50 1.4 5.5 3 ~ Refueling, Removal 6 Replacement of Shield 0 33 18 (b) 0.60 4, Stored Activated Support Structure(<) 0 0.5 720 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
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~
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~5 507 506 504 503 8 9S2 502 954 11.5' 5.063'd Q~
952 2
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'(sea NODE NO.
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+ ~819 714 8 903 912
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716 717 sL PLAN (EL 39.98')
MODEL-NEUTRON SHIELD FIGURE
3'ECTION "A-A" 958 ~959 967 966 45 0.75'5 Fi '59.95 591 41.4'05 590 591 A
EL 42.52'L Q690 Q590 EL 39.98Ec Secs'8 Y A SECTION "B-B" EL 39 98 907 903 900 I
5.47'L Q930 Q933 Q 34.51'WF31 A 931 e A 933 B A 935 e 830 SEcZ C 0.875'32 16WF 3S 931 SEcT D e
933 935 6 ~
910 912 0 915 I
Q936 Q0 A 937 e a 939 B A941 e 937 939 941 936 3 938 940 Sic%' SKcg SEc.q "RADIALLYOUTWARD DIRECTION" MODEL-NEUTRON SHIELD FIGURE 19
Cross Sections 1-H -1 10' 6
Total
--- - El as t
" (n,y) 1 c 10 8
)08 1O6 10 10 10 10 10 10 3, Nev 1-H -1
III. SCHEDULE Completion of design July 15, 1977 Material purchase August 15, 1977 Material delivery November 15, 1977 Installation First. scheduled unit shutdown of sufficient duration after material delivery.
tl s ~ e envue LltN OhlSS3308d fH3l<A000 03AI3038