Text

Forwards Overall Findings Re Continued Safe Operation of Facility in Response to 810814 Ltr.Current Status of Safety Evaluations Does Not Accurately Reflect Qualification Status of Electrical Equipment.Supporting Documentation Encl
	ML20003H548
Person / Time
Site:	Millstone
Issue date:	04/30/1981
From:	Counsil W; NORTHEAST NUCLEAR ENERGY CO.
To:	Novak T; Office of Nuclear Reactor Regulation
References
	A01203, A1203, TAC-42493, NUDOCS 8105060308
	Download: ML20003H548 (78)
	v • d • e

c: '.

THIS DOCUMENT CONTAINS NORTHEAST UTIEJTIES POOR QUAUTY PAGES C O, % %'?"*Jf" P O. BOX 270 C.T.TrJ."'l,0^ "'~" E Z ,cossec a m i L J rr l:2'=.' "r"

rads gamma provided it is demonstrated that the affects of beta radiation are not significant. It is also of interest to note that the above requirements differ from those delineated in NUREG-0588.

In response to the Staff requirement to provide additional details regarding the determination of the appropriate radiation service conditions, NNECO hereby provides Attachment 1, Radiction Service Conditions Inside Containment, Millstone Unit No. 2. The bases, assumptions, calculations, and results are included to substantiate the validity of the values utilized in the equipnent qualification process. An explicit quantification of the beta dose in air is provided in the Attachment.

8. Regarding ccuponents included in Appendix B, it is not possible to address the specific nature of the Staff concerns in all casen in the absence of the SER. Nonetheless, the following points are relevant:
~~a. Of the 48 components listed in Appendix B, NNECO concludes that full qualification for 8 of those components was provided in Reference (1).~~
b. In the case of some 34 components included in Appendix B, plans for providing full qualification and justification for continued operation was provided in Reference (1). No further justification or action beyond that detailed in Reference (1) is planned for these 34 canponents.
c. Regarding the remaining 6 components included in Appendix B, all are being installed as 'IMI Action Plan requirements, and the documentation requirements are being fulfilled in accordance with NUREG-0737, Reference (4) and Reference (5). For item numbers 15C and 16C, the due date for submitting qualification documentation has not yet passed, and implementation is not required until January 1, 1982. NNECO intends to comply with schedular requirements of NUREG-0737, Reference (4) and Reference (5). Item numbers 17C, 18C, 19C, and 20C all relate to Action Plan requirement II.D.3, Direct Indication of Relief and Safety Valve Position. The qualification status of this equipment was summelzed in Reference (6), and additional details can be found in correspondenco referenced therein.
In light of the above points, NNECO suggests that .;he current status of the Staff's Safety Evaluation Reporte (SER) does 9;t accurately reflect the qualification status of electrical equipnent at Millstone Unit No. 2. It would appear to be in our mutual best interests to have the above concerns evaluated in detail prior to issuance of the SER's, and the Technical Evaluation Report (TER) discussed in Section 3 of Reference (2). Subsequent to receipt of those documents , including the basis for the conclusionr in Appendices B
e o
_9_
and C of the Partial Review, a complete and meaningful reevaluation can be completed. For the interim period, NNECO reiterates its conclusion that tLe Staff's concerns identified in Reference (2) do not alter the determinatioz.
that Millstone Unit No. 2 can continue to be operated safely.
Very truly yours, NORTEEAST NUCLEAR ENERGY COMPANY N(
Senior Vice President '
STATE OF CONNECTICUT )
) ss. Berlin g' q ,q g' /pgf COUNTY OF HARTFORD )
Then personally nppeared before me W. G. Counsil, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, Licensee herein, that he is authorized to execute and file the foregoing infor=ation in che name and on behalf of the Licensee herein and that the statements c 2tained in said information are true and correct to 'che best of his knc , ledge and belief.
Quie e-Notary Public hu bwieu u /@ % O~ N SI f
< o Docket No. 50-336 Millstone Nuclear Power Station, Unit No. 2 Attach =ent 1 Radiation Service Conditions Inside Containment April, 1981
e o MILLSTONE UNIT NO. 2 RADIATION SERVICE CONDITIONS INSIDE CONTAINMENT
! 1. GENERAL 1
By Reference 1, the NRC provided the detailed methodology for determining the radiation service conditions inside containment for the Millstone Unit No. 2. In reviewing the material given in Reference 1, NNECO determined that a detailed radiation environment analysis which was performed in 1979 fulfilled the requirements of Reference 1. The methodology and assumptions used to generate these radiation service conditions and the results are provided below.
2. SCENARIO The shielding analysis was performed assuming design basis accident assumptions. The scenario is based upon the guidance given in NUREG0588 in that an accident occurs which totally depressurizes the reactor vessel and releases substantial fission products into the containment atmosphere.
~~3. SOURCE TERMS AND DISTRIBUTION OF ACTIVITY f~~
NNECO assumed an initial release of 50% of the core iodine and 100%
of the noble gas activity to be uniformly distributed in the contain-t
e 4 ment atmosphere at t=0. For conservatism, no plateout on the containment walls and no removal of iodine from the containment sprays were assumed.
NOTE: It should be emphasized that 50% of the core iodine activity has been assumed to be uniformly mixed in the containment atmosphere (e.g., no plateout assumed) because '
this assumption is conservative and simplifies the modeling by avoiding the necessity of calculating the plateout dose rate. NNECO does not believe that 50% of the core iodines would be available for release from containment in a DBA.
For off-site dose calculations to comply with 10CFR100, we assume that 25% of the iodines are available for release from the containment. This is consistent with criteria given in Regulatory Guide 1.3. Decay of fission products was the only means of removal assumed in the analysis.
l l
l Core activity levels were based on TID-14844 source terms. In order l
l to simulate the band of gamma ray spectra emitted by the nuclides, the source was broken up into a seven energy group gamma source.
The mean energies of the source are: 0.4, 0.8, 1.3, 1.7, 2.2, 2.5 and 3.5 Mev.
I l
l I
~~4. METHOD OF ANALYSIS~~
a. Gamma Dose A modified version of the QAD-P5F (Reference 2) computer code was used to evaluate the gamma dose rates at various locations inside the containment. The airborne source was broken up into four main source regions. The source volume in these regions was represented by a total of 36,000 point sources.
I Shielding which surrounds the steam generators was modeled in the QAD runs. Since the shielding is irregular, the shield wall was modeled as a cylinder in order to simplify the calcu-lation. The inside radius of the cylindrical wall was conservatively taken to be about 35 feet, and the wall thickness was 3.5 feet. The density of the concrete was assumed to be 2.24 gm/cm3 ,
Twelve receptor points were placed throughout the containment and the highest dose rate was used to compute the integrated dose at 30 days after the accident. This receptor location i
corresponds to a receptor point located on the containment centerline. The dose from this receptor location was used to qualify all equipment inside containment except the equipment at the electrical penetrations. Another specific dose calcu-lation was performed at the electrical penetrations using the l
same methodology.
l l
l i
t e a
b. Beta Beta doses inside the containment were obtained using the semi-infinite cloud dose model based on the concentration of fission products which exist inside the containment. This method is appropriate because of the short range of beta particles in air. The beta dose in Table 1 represent the beta dose in air and does not represent the beta dose the equipment will actually receive. Actual beta doses to equipment are expected to be significantly lower due to local shielding considerations (e.g., metal casing around electric motors will absorb most of the betas, etc.).
~~5. SAMPLE CALCULATION~~
~~a. Gamma-LOCA Dose The calculation of the dose at the center of the containment will be used as a sample case.~~
A detailed description of the geometry model of the containment is given in Figure 1. This figure is based on information contained in Drawing #25203-11177 and represents the contain-ment as 5 different regions. Regions 1 throutb 4 represent different source regions, whereas, Region 5 is the concrete shield wall. All boundaries and regions are clearly depicted in the figure.
A seven energy grcup representation of the source was obtained using source data (see Tables 2 and 3) provided by Stone and Webster for the Connecticut Yankee (CY) electrical equipment qualification project. The CY source terms were ratioed by d
core power level and containment volume to obtain appropriate values for MP-2. The appropriate ratio to multiply the CY source terms is:
Ratio = R = V CY PMP-2 X
VMP-2 PCY Where: V *
~~I""* C " "I""*"~~
CY
= 2.23 x 106 gg 3 V = v lume f MP-2 containment MP-2 6 3
= 1.9 x 10 ft P = MP-2 core thermal power level = 2700 Mwt MP-2 P = CY core thermal power level = 1825 Mwt CY
~~00 Mwt R=~~
x = 1.74 1.9 x 10 ft 1825 Mwt The CY source terms (Tables 2 and 3) were s.altiplied by the above ratio in order to obtain the MP-2 source concentrations.
The resulting MP-2 source terms which were used in the QAD code are given in Table 4.
. o In order to simplify the geometry, the containment was modeled as a cylinder of radius 1981 cm and effective height of 4785 cm.
The effective height was based on a total containment volume of 5.853 x 10 IO 3 cm ,
The gross volume of each source region was computed as follows:
Region Volume Calculation Gross Volume 2 2 2 2 cm 3 1 Vi = n(r -r )h = v(1982 -1071.9 ) 467.7 = 4.258 X 109 2 2 V 2=n(r-r3)h=n(1982-1178.6) 2 cm 3 2 (1173.5-487.7) = 5.471 X 109 3 V3 = n r2L = n(1071.9)24785 = 1.727 X 1010 c,3 4 V4 = n(19822 _1071,92 ) (4785-1173.5) = 3.153 X 1010 c,3 10 c,3 Total Volume = V3+V2*v+V4 3 = 5.853 x 10 0 3 Because the free air volume in the containment is 1.9 x 10 ft 10 (5.4 x 10 c,3), the gross volume of each region calculated above must be ratioed by the factor:
5.4 X 10 10 5.853 X 10 in order to account for the actual free air volume of the containment. Therefore, the free air volume of each region is:
a .
Region Free Air Volume 9 3 9 3 1 V = 4.258 x 10 m x .9266 = 3.945 x 10 cm 3
9 3 2 V2 = 5.471 x 109 cm3 x .9266 = 5.069 x 10 .cm M 3 3 V = 1. 27 x 10 cm3 x .9266 = 1.600 x 10 cm 3
10 c ,3 x .9266 = 2.921 x 10 10 c ,3 4
V4 = 3.153 x 10 The entire listing of QAD input parameters which were used in the QAD computer runs are given in Tables 5 to 8. For conserta-3 tism, a density of air has been taken to be 0.0012 gas /cm and 3
for concrete, the assumed density was 2.24 gm/cm . All attenua-tion, build-up, and dose conversion factors which were used in the analysis were obtained from the Stone & Webster RP8A shielding manual (Reference 3). The source was assumed to be t uniformly distributed in each region.
The results of the QAD runs for each source region are given below (see Appendix A for QAD runs).
l l
Source Dose Rate Region (mR/HR)
I 1 1.213 x 10 l 2 8.324 x 10 7 l 3 2.124 x 10 9 4 1.149 x 10 9 3.4 x 10 9 I
l The (0-30) day integrated LOCA dose was calculated by ratioing l results obtained from a Stone & Webster calculation regarding l
l
_g-CY electrical equipment qualification. Figure 2 shows the gamma dose rate and integrated dose as a function of time at an area on the CY charging floor. Since the same mixture of isotopes (e.g. 50% core iodine, 100% core noble gas) was assumed to be released into the containment, the integrated dose for a similar geometric configuration would just be a constant factor multiplied by the initial dose rate. The integrated dose for MP-2 would be obtained using the following equation:
Dy =DR MP-2 MP-2 X D r D CY where: D = 30-day integrated dose in the MP2 IMP-2 containment D = d se rate at t=0 in the MP2 contain-IMP-2 ment 6
= 3.4 x 10 R/HR D = d se rate at t=0 in the CY containment Rg
= 1.2 x 10 6R/HR (see Figure 2)
D = integrated dose in the CY containment 1
at 30 days (see Figure 2)
DI = 3.4 X 106 R/HR MP-2 X 6.6 X -10 6 Rads = 1.9 X 107 rads 1.2 X 106R/HR
9
b. Beta-Dose-LOCA The beta dose calculation was based on a Stone & Webster computer calculation for dose rates and integrated doses inside the CY containment. Since the particular computer run assumeo a 25% iodine and a 100% noble gas release, a correction was made to account for a 50% iodine release.
The computer doses are based on a semi-infinite cloud dose model and are given in Table 9. Corrections to the numbers in these tables are given below to account for MP-2 plant specific parameters.
7 Total Dose (Noble Gas & Halogens) = 6.12 x 10 rads (from Table 9)
Halogen Dose (based on a 25% halogen release) = 1.39 x 10 rads Noble Gas Dose Only = 4.73 x 10 rads Dose from 50% Halogen Release = 1.39 x 10 rads x 2 = 2.78 x 10 rads Total Beta Dose 7
in MP-2 Containment = (4.73 x 10 7 + 2.78 x 10 ) x 1.74 = 1.3 x 100 rads
c. 40-Year Normal Operating Dose The 40-year normal operating dose was calculated based on surveys (Reference 4) performed in the MP-2 containment. The receptor location from the survey was taken to be point N-5 (see Figure 3). It should be recognized that at the time these surveys were taken, MP-2 did not have a neutron shield in the cavity area. A neutron shield was installed during the second
refueling outage and appropriate dose reduction factors have been used to specify the normal operating dose rates.
Neutron Dose Rate = 65,000 mR/KR Gamma Dose Rate = 10,000 mR/HR Assumptions: 1) This neutron dose is a result of streaming
~~2) The neutron shield will reduce the neutron dose by a factor of 20. This assumption is conrervative since Reference 5 specifies tLat the reduc-tion is about a factor of 65.~~
~~3) Quality factor for neutrons = 2.0~~
~~4) No attenuation taken credit for gamma rays because of neutron shield~~
~~5) Plant Capacity factor of .8 assumed The neutron dose rate is, therefore:~~
I '" I #*d 65,000 mrem /HR x x x 1 = 1.6 Rad /HR 103 mRen 2 rem 20 Total Dose Rate = neutron + gamma
= 1.6 R/HR + 10 R/HR = 12 R/HR Total 40-Year Normal Operating Dose 6
40 years x 365 days x 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months x .8 x 12R/HR = 3.4 x 10 Rad year day Assuming 50% higher radiation levels due to changes in core configuration from when surveys were performed would give a dose of:
3.4 x 106 Rads + .5 x 3.4 x 10 6rads = 5.1 x 10 rads 0 Total Gamma Dose = LOCA + Normal Operating 6
= 1.9 x 10 rads + 5.1 x 10 rads = 2.4 x 10 rads A summary of the computed doses is given in Table 1.
t
~~6. COMPARISON WITH NRC NUREG 0588 l~~
l The NRC performed a detailed calculation of the LOCA doses at several locations in a PWR containment. in Appendix D of NUREG 0588.
The sample calculation was performed for a 4000 Mwt reactor housed 6 3
[ in a 2.52 x 10 ft containment. Integrated 30-day doses developed j by the NRC were 1.5 x 10 rads gamma at the containment centerline
! 6 and 9.1 x 10 rads gamma on the containment wali Correcting for MP-2 power level and c,atainment volume we obtain orresponding 7
values for MP-2 o; 1.34 x 10 rads gamma (at the cent.;r of contain-ment) and 8.1 x 106rads gamma (at the containment wall). .
The NNECO calculated LOCA gamma dose at the containment center is 7 6 1.9 x 10 rads and 9.4 x 10 rads at the containment wall. The NNECO LOCA calculation, therefore, bounds the NRC calculation and supports the position that the values used for equipment qualifica-tion are conservative and acceptable.
The Beta LOCA dose calculated by NNECO at the center of contsinnent is 1.3 x 10 rads. After correcting NRC NUREG 0588 Beta LOCA dose of 1.4 x 10 rads for MP*2 power level and containment volume, we 8
obtain a beta dose of 1.25 x 10 rads.
The NNECO calculated LOCA beta dose, therefore, bounds the NRC calculated value.
I
t TABLE 1 30 DAY INTEGRATED DOSE IN MP-2 CONTAINMENT 40 Year Normal
~~P'.ceptor Point. Operating Doses 30 Day Integrated~~
~~Location (Rads) LOCA Doses (Rads) Total Dose (Rads) n+y Gamma Beta Ganuna Beta t~~
i 6 8 7 8 Center of Containment 5.1 x 10 1.9 x 10 7 1.3 x 10 2.4 x 10 1.3 x 10
~
4 6 8 6 8 Containment Wall 1.0 x 10 9.4 x 10 1.3 x 10 9.4 x 10 1.3 x 10 i
4 1
4
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4 1
2
Table 2 emessesene RESULTS asescocare RADIOISOTOPE e e e REVISED 3/14/68 .
CC'24 YA* NEE LCCA STUDY ItGIOE CCriTAIt@ TENT ICDIllES y .
CECAY tit!E (HCURS)= 0.0 PLTIFICATION CC3E trJSER: 0
)
tra'EER CF ISOTCPES HAVIttG INITIAL ACTIVITIES: 5 trJSER OF ALTERATIC:45 TO LIBRARY PART is 0 3
huSER CF ALTERATICNS TO LIBRARY PA'tT 2a 0
)
INITIAL FINAL M ACTIVITY ACTIVITY SPECIFIC ACTIVITY (ItEV/CC-SECS AT ENERGY OtEV) 0F 0.40 0.80 1.30 1.70 2.20 2.50 3.50 ISOTCPE (UC/CCI (UC/CC)
FISSICr4 PRO 7UCTS - N00LE GASES
., c ./'..
-g FISSICM PRODUCTS - HALOGEtiS 0.38BE 07, , 0.947E 06 0.0 0.0 0.0 0.0 0.0
- *' I131 0.3350 03 0.335E 03 .
T 5- 0.50C0 03 0.500E 03 0.0 0.280E 08 0. SSE 07 0.303E 07 0.185E 07 0.0 0.0 1132 I133 0.0180 03 0. alee 03 0.0 0.167E 08 0.0 0.424E 06 0.0 0.0 0.0 0.527E 08 0.193E 08 0.119E 08 0.0 0.0 0.0 1134 0.e69D 03 0.869E 03 0.0
~
113S 0.7740 03 0.774E 03 0.0 0.347E 07 0.272E 08 0.172E 08 0.286E 07 0.0 0.0 J
FISSIQti PRODUCTS - REMAINDER d' CORRCSION/ACTIVATICH PROOUCTS
)
TOTAt. SPECIFIC ACTIVITY (ifEV/CC-SEC) AT ENERGY (HEV) OF v* 0.40 0.40 1.30 1.70 2.20 2.50 3.50 FISSICM PRODUCTS - NOSLE GASES 0.0 0.0 0.0 0.0 0.0 0.0 0.0
.s FISSICr4 VRCDUCTS - HALOGENS 0.388E 07 0.102E 09 0.520E 08 0.325E 08 0.471E 07 0.0 0.0 FISSto'd PRODUCTS - REHAINDER 0.0 0.0 0.0 0.0 0.0 0.0 b.0 r
CORFOSICN/ACTIVATICH PRCOUCTS 0.0 0.0 0.0 0.0 0.0 0.0 0.0
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Table 3 (continued) .
TOTAL SPECIFIC ACTIVITY (MEV/CC-SEC) AT ENERGY EHEV8 CF
). 0.40 0.80 1.30 1.70 2.20 2.50 3.50 FISSIC*I PRCOUCTS - NO3tE GASES 0.15tE 08 0.410E 08 0.117E 07 0.117C 09 0.801E 08 0.475E 08 0.502E 08 3
FISSICH PRODUCTS - HALCGENS 0.0 0.0 0.0 0.0 0.0 0.0 0.0 FISSIC'4 PRLOUCTS - ret:AIN3ER 0.0 0.0 0.0 0.0 0.0 0.0 4.'J J
CCRP05ICTVACTIVATICf4 FRCDUCTS 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Stft CF FISSIC:t At3 CO.720S10tt PRO 3. 0.15tE 08 0.410E 00 0.117E 07 0.117E 09 0.801E 08 0.475E PS 0.SoIE 08 J
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CHAD.By DE TABLE 4: Conservative Source Terms
~~CAT. I cate,,M-Z l *S(,, gj Specific Activity (Mev/cc- rec)~~
SHEET NO. b or_f 9 l Conn Yankee 3
.4 .8 1.3 1.7 2.2 2.5 3.5 :
Energy of Gax.a (Mev) italogensl 3.88 x 10 6 1.02 x 108 5.2 x 10 7 3.25 x 10 7 4.71 x 10 6- 0.0 0.0 8 7 7 7 1.52 x 10 7 4.1 x 10 7 1.17 x 10 6 1.17 x 10 8.01 x 10 4.75 x 10 5.02 x 10 i i Noble Cases 2 Total 1.91 x 10 7 1.43 x 108 5.32 x 107 1.50 x 108 8.50 x 10 7 4.75 x 10 7 5.02 x 107 i i
I MP 2 l
Energy of Gamma (Mev) .4 .8 1.3 1.7 2.2 2.5 3.5 4
Halogensl 6.751 x 10 6 1.775 x 108 9.048 x 107 5.655 x 10 7 8.195 x 106 0.0 0.0 2.645 x 10 7.134 x 10 7 2.036 x 10 6 2.036 x 108 1,394 x 108 8.265 x 107 8.735 x 107
, Soble Cases 2 i e 2.488 x 10 8 9.252 x 10 7 2.602 x 108 1.476 x 108 ,8.265 x 107 8.735 x 107
} i Total l 3.320 x 107 I
I j 1 50% of core inventory of Halogens released to containment atmosphere j 2 100% of core inventory of Noble Gases released to containment atmosphere 2
3 as calculated in S&W Cat. 1 Calc # es PR(B)-008, PR(B)-009. PR(B)-010 (Computer Runs 8024, 8025) t l
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e Table 9 CONrl YAfNEE LCCA STUDY ItGIDE CC :!/.It# TENT r4TEODATED OCSE (RElil It4 VOLUttE 1 ( PRIltARY CCMAlt0 TENT I BETHEEN TIttE T(J-1) Atc Tt JI T(J) HRS 2.00 8.00 24.00 24.00 96.00 720.00 720.00 1440.00 2160.00 BETA CCSE .
I127 2.455-04 1.47E-03 3.93E-03 2.62E-08 1.77E-02 1.53E-01 2.40E-07 1.77E-01 1.77E-01 2131 2.89E+04 1.71E*05 4.39E+05 2.05E+00 1.69E+06 5.13Ee06 2.15Ee00 5.66E*05 4.28E*04 1132 5.56Ee04 1.31E+05 2.52E+04 4.41F403 1.98E+02 6.40E-08 0.0 0.0 0.0 I133 1.30E+05 6.97E+05 1.31E+06 6.6:E*00 1.70E*06 1.74E*05 6.36E-09 1.97E-04 9.38E-15 1134 6.29E+04 5.16E+04 4.50E+02 1.19E-07 1.41E-03 2.40E-08 0.0 0.0 0.0
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I135 1.30E+05 5.53E+05 5.21Ee05 1.36Ee00 1.23E*05 7.25E*01 7.3 E-31 7.26E-27 3.58E-59 1136 6.23E-10 7.87E-23 0.0 0.0 0.0 0.0 0.0 0.0 0.0 LPG 3 5.86E+03 1.40E.04 2.72E+03 7.15E-04 2.22E+01 d.64E-09 0.0 0.0 0.0 FC84 7.23E+03 2.e6E+03 1.49E+00 3.5:E-13 2.6:E-09 1.04E-48 0.0 0.0 0.0 EPS5 3.27E-03 3.1 E-07 2.39E-45 0.0 0.0 0.0 0.0 0.0 0.0 CR87 1.80E-16 1.06E-35 0.0 0.0 0.0 0.0 0.0 0.0 0.0 HF allt 3.31E*03 1.15E 04 3.65E+03 1.34E-03 4.34E+01 1.99E-08 0.0 0.0 0.0 k385tt 2.53E+04 9.07Ee04 5.30E+04 7.7dE-0 4.6 E*03 5.4E-02 1.55E-48 1.01E 44 0.0 kPG5 2.4E+03 1.45E.04 3.80E+04 2.59E-01 1.75Ee05 1.51E*06 2.36E+00 1.73E+06 1.72E*06 HF67 2.04E405 3.05E+05 1.30E*04 1.47E-04 2.50E+00 5.64E-17 0.0 0.0 0.0 H308 1.55E+05 4.26E+05 1.22E*05 6.27E-00 2.37E*03 4.26E-05 7.80E-76 3.2 E-72 0.0 kP89 1.63E-01 3.70E-07 5.06E-41 0.0 0.0 0.0 0.0 0.0 0.0
> XE131tf 4.51E*02 2.7:E.03 7.36Ee03 4.96C-02 3.44E+04 2.30E*05' 2.09E-01 9.01E+04 1.90E+04 XE133!! 6.91E*03 4.14E+04 1.07Ee05 6.91E-01 3.49E+05 2.7:Ee05 1.34E-03 1.09Ee02 1.09E-02 XE133 1.70E*05 1.02E+06 2.67E*06 1.75E*01 1.03E*07 2.18E+07 4.06E+00 7.49E+05 1.50E+04 XE135tt 3.0SE+04 1.31E.05 1.03E+05 3.00E-01 2.91Ee04 1.71E*01 1.73E-31 1.72E-27 8.46E-60
J XE135 1.58E+05 1.10E+06 2.43E+06 1.13E*01 1.69E+ 6 1.0:E404 1.65E-21 2.2:E-17 3.03E-41 4.22E+00 1.15E-04 4.66E-32 0.0 0.0 0.0 0.0 0.0 0.0 XE117 XE138 0.28E+04 2.16E+03 9.00E-04 2.30E-04 8.81E-21 0.0 0.0 0.0 0.0
"' TOTAL 1.22E*06 4.76E*06 7.87E+06 4.11E*01 1.61E+07 2.91E*07 8.79E+00 3.14E*06 1.00E+04 Cult. TOT AL 3.40Ee06 8.16E+06 1.60E+07 1.60E*07 3.21E+07 6.12E+07 6.12E*D7 6.44E+07 4.4 E+07
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REFERENCES
~~1. I&E Bulletin 79-01B Supplement 2.~~
~~2. AKERN, RSCI Computer Code Collection, CCC-190.~~
~~3. Radiation Shielding Design and Analysis Approach for Light Water Reactor Power Plants, RP-8A, Stone & Webster Engineering Corporation, May 1975.~~
4. D. C. Switzer letter to G. Lear dated April 15, 1976, transmitting Millstone Nuclear Power Station Unit 2, "Radistica Survey Results In and Around Millstone Unit 2 Containment Building", Northeast Nuclear Energy Company, Hartford, Connecticut, April 1976.
~~5. W. G. Counsil letter to R. Reid dated November 9, 1979, transmitting, Evaluation of Neutron Shield Effectivenss Report.~~
APPENDIX A Computer Code Output
NCRTHEAST UTILITIES SERVICES COMPANY RADIOLOGICAL ASSESSMENT BERLIN CCNNECTICUT QADPS-F FROGRAM 04/23/81 MP-2 RADIATION QUALIFICATI0H OF ELECTRICAL EQUIPMENT (SOURCEzREGION I)
FROGRAM CONTROL *****
NUMBER OF SOURCE POINTS ALONG THE X AXIS. MAX. 30 8 20 NUMEER OF SOURCE POINTS ALONG THE Z AXIS.NAX. 30 = 20 13.JMBER OF SOURCE FCINTS ALONG THE Y AXIS.HAX. 30 = 20 m 2 NUMBER OF MATERIALS.NAX. 30
= 2 NUMSER OF COMPOSITIONS. MAX. 50
= 5 NUMBER CF ZONES. MAX. 200 NUMBER OF FHOTON ENERGY GROUPS. MAX. 30 8 8 a 7 NUMBER OF BCUNDARIES. MAX. 200 s 0 SOURCE GEOMETRY TYPE CPTICH
= 1
~~10ST FRCSABLE SOURCE ZONE~~
~~= 2 SCURCE CCMPUTATION OPTICH HUMBER OF NEUTRCH SASE MATERIAL ADO ENERGY GROUPS~~
~~O FIRST SOURCE-DETECTOR TRAVERSE FCR GECMETRY FRINTs 0 LAST SOURCE-DETECTOR TRAVERSE FOR GEOMETRY PRINT = 0~~
~~O DEL SOURCE-DETECTOR TRAVERSE FOR GEOMETRY PRINT a 0 GAfftA RAY AND NEUTRON REFERENCE MATERIALS CONVERSION OF GAMMA RAY AND HEUTRCH OUTPUT OPTIONz O~~
~~I GANMA FTOD FACTORS. CARD INPUTS 0. INTERNALS NO.~~
a 4 SUILDUP FACTORS. CARD INPUTz0.INTERNALzNO.
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t NP-2 RADIAT20N QUALIFICATION OF ELECTRICAL EQUIPMENT (SOURCE: REGION 1)
COCRDINATES - X 0.0 Y 0.0 Z 2.3825E+03 RECEIVER HUMBER I ENEPGY FLUX DOSE RATE GRP NEAN ENERGY MEAN BUILDUP DIRECT BEAT 1 WITH BUILDUP DIRECT BEAN WITH EUILDUP ND ENERGY GROUP LINITS FACTCR5 NEV/CNZ-5EC ttREN/NR MEV NEV 1.39410+00 1.75990*08 5.4535D+08 3.69590+05 5.15230+05 1 0.4000 3.5957D*06 2 0.8000 1.2517D*00 1.4363D+09 1.79780*09 2. 872 7D + 06 1.1896D+00 5.6332D+08 6.6954D+0S 1.01400+06 1.2052D+06 3 1.3000 3.22960+06 1.162SD*00 1.633fD*09 1.8993D+09 2.7775D+06 4 1.7000 1.52300+06 1.7435D+06 5 2.2000 1.1448D+00 9.5186D+08 1.08970+09 5.39360+08 6.1320D+0S 8.09040*05 9.19800+05 6 2.5000 1.13690+00 9.24123+05 7 3.5000 1.11870 + 00 5.9004D+08 6.600!D+0S S.26050*05 1.0977D+00 7.0823D-20 7.77440-20 7.79060-23 8.55193-13 8 6.1500 5.8907E+09 6.97550+09 1.0192E+07 1.2133!+07 TOTAL 1.2864 WDBU 1.324i ENERGf FLUX BUILDUP DDSE EUILDUP 1.15410+00 1.190SE+00
~~TIME FOR DETECTOR IN NIN. a 0.0 ese END OF JOS E16103 ==a t~~
s NORTHEAST UTILITIES SERVICE 3 COMPANY RADIOLCGICAL ASSESSMENT BERLIN CONNECTICUT QADFS-F FROGRAM 04/C3/81 MP-2 RADIATION QUALIFICATION OF ELECTRICAL 23JIFMENT (SOURCEsREGION 23 eenen PROCRAM CONTROL emmen NUMBER OF SOURCE ."0INTS ALONG THE X AXIS MAX. 30 a 20 NUMBER OF SOURCE POINTS ALCf G THE Z AXIS.t1AX. 30 a 20 NUMBER OF COURCE POINTS ALONG THE Y AXIS. MAX. 33 = 20 a 2 NUM3ER OF MATERIALS. MAX. 30
= 2 NUMSER OF CONFOSITIONS. MAX. 50 NUMBER OF ZONES. MAX. 200 s 5 NUMBER OF PHOTCH ENERGY GROUPS. MAX. 30 a 8 NUMBER OF BOUNDAKtES. MAX. 200 h 7 a 0 SOURCE GEOMETRY T1FE OPTION a 1 t10ST PROSA8LE SOURCE ZONE a 2 SOURCE COMFUTATION OPTION NUMBER OF NEUTRC*4 BASE MATERIAL AND ENERGY GROUPSs 0 FIRST SOURCE-OF,TiCTOR TRAVERSE FOR GEOMETRY PRINT 8 0 LAST SOURCE-DE1ECTOR TRAVERSE FOR GEOMETRY PRINT a 0 a O DEL SOURCE-DETECTOR TRAVERSE FOR GEOMETRY FRINT GAPT1A RAY AND HEUTRON REFERENCE MATERIALS 8 0 CONVERSION OF GAMMA RAY AND NEUTRON OUTPUT OPTIONS 0 GAft1A FTCO FACTORS. CARO INPUTS 0. INTERNALS NO. = 1 BUILDUP FACTOR $, CARD INPUTS 0.INTERNALsNO. s 4 gal 1PtA MEAT CONVERSION FACTORS (N0s0.YEssil = 0 e
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j RECEIVER HUNGER 1 COOPDINATES - X -0.0 Y 0.0 Z 2.3825E+03 l i i 3 GEOMETRY PRINT FOR PSEU3O SOURCE POINT AT THE COCRDINATE ORIGIN i
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HUMBER OF SOURCE POINTS ALottG THE X AXIS.NAX. 30 a 20 HUMBER OF SOURCE POINTS AL0tiG THE Z AXIS.NAX. 30 s 30 HUMBER OF SOURCE POIt4TS ALot4G THE Y AXIS.NAX. 30 a 20 a 2 HUMBER OF NATERIALS.HAX. 30 a 2 NUMBER OF COMFOSITICitS.HAX. 50 s 5 NUMBER OF ZOt:CS. MAX. C00 HUNBER OF FHOTON ENERGY GROUPS.NAX. 30 8 8 a 7 HUMBER OF BOUNDARIES.NAX. 200 SOURCE GEOMETRY TYPE OPTI0t4 s 0 MOST PROBABLE SOURCE ZONE a 1 s 2 SOURCE CCMP'JTATION OPTIO?4 HUMBER OF 'dEUTRON BASE NATERIAL AND ENERGY GROUPSs. O FIRST SOvRCE-DETECTOR TRAVERSE FOR GEOMETRY FRINTs 0 LAST SOUPCE-DETECTOR TRAVERSE FOR GEOMETRY PRINT a 0 a 0 DEL SOURCE-DETECTOR TRAVERSE FCR GEOttETRT PRINT
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MP-2 RADIATION QUALIFECATION OF ELECTGICAL ECUIPMENT (SOURCE:PEGICt13)
RECEIVER HUMBER 1 COCRDINATES - X 0.0 Y 0,0 Z t.3825E+03 GEOMETRY PRINT FOR PSEUDO SOURCE POINT AT THE COCRDINATE CRIGIN 2CHE BOUtOARY DISTANCE X Y Z 3 0 1.00000+01 0.0 0.0 2.3950+03
.- --- . , . , - - - . . . - - - . - . . . - - . . , . . - ~ - - - - . . ,- - .-- ..- . - . - . . . - - ~ . . , . . _ . - . -
i MP-2 RADI ATIDH QUALIF3CATEDN OF ELECTRICAL EDUIM1Et4T (SDURCE*REGIDH 3)
CDDRDINATES - X 0.0 Y 0.0 Z 2.3S25E+03 RECEIVER HUMBER 1 MEAN BUILOUP Et:ERGY FLUX DDSC RATE '
GRP NEAN ENERGY DIRECT BEAM WITH EUILDUP DIRECT BEA.". WITH BUILDUP t.'D ENERGY GRDUP LIMITS FACfDRS '
HEV/CH2-SEC NREM/tfR MEV NEV I
.3.92260+10 4.33430+10 8.25003607 9.1020D+07 1 0.4000 1.1033D+00 3.0146D+11 3.19280+11 6.0292D+08 6.3354D+08 2 0.8000 1.0591D+00
'1.1365D+11 1.18070+11 2.04570+08 2.1253D*06 ,
3 1.3000 1.03890+00 5.6449D+08-1.03120+00 3.22000+11 3.32060+11 5.47400608 4 1.7000 1.8374D+11 1.88640+11 2.93993+08 3.01S20*08 '
5 2.2000 1.0266D+00 1,5473D+0S t 1.0315D+11 1.05710+11 1.5256D+08 6 2.5000 1.044e0600 1.0975D+11 1.12130+11 1.53660+0S 1.56900+0S ,
7 3.5000 1.02160*00 1.4174D-20 6.1500 1.0174D+00 1.26653-17 1.28850-17 1.39320-20 l 8
1.1730E+12 1.21920+12 2.0398E*09 2.1240E+09 TOTAL 1.2784
. WDBU 1.2919 ,
ENERGY FLUX SUILDUP DDSE BUILDUP 1.03940+00 1.0413E+co t
TIME FOR DETECTOR IN MIN. 0.0 ene Ete OF JDB E16103 *mm ,
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e NORTHEAST UTILITIES SERVICES COMPANY RADIOLCGICAL ASSESSMENT BERLIN CONNECTICUT QAOPS-F PROGRAM 04/23/81 HP-2 RADIATION QUALIFICATION OF ELECTRICAL EQUIPMENT (SOURCE: REGION 4) ename PROGRAN CONTROL *****
NUMBER OF SOURCE POINTS ALONG THE X AXIS.NAX. 30 m a0 NUMBER OF SOURCE POINTS ALONG THE Z AXIS.NAX. 30
20 NUM3ER OF SOURCE POIN13 ALONG THE Y AXIS.NAX. 30 a 20 a t NUMBER OF MATERIALS.NAX. 30 a 2 HUMBER OF COMPOSITIDNS.NtX. 50 NUMSER OF ZONES. MAX. 200 a 5 NUMSER CF FHOTON ENERGY GROUPS.NAX. 30 m a s 7 NUMSER OF BOUNDARIES.NAX. 200
~~0 SOURCE GEOMETRY TYPE OPTION NOST PROBABLE SovRCE ZONE 1~~
a 2 SOURCE COMPUTATIOt3 OPTION NtR*BER OF NEUTRCt1 BASE NATERIAL AND ENERGY CROUPSs 0 FIRST SOURCE-DETECTOR TRAVERSE FOR GECMETRY FRINTs 0 LAST SOURCE-DETECTOR TRAVERSE FOR GEONETRY PRINT a 0 CEL SOURCE-DETECTOR TRAVERSE FOR GEONETRY PRINT s 0 a 0 GAMMA RAY AND NEUTRON REFERENCE NATERI ALS CONVERSION OF GANNA RAY Ato NEUTRCN OUTPUT OPTION
~~O u~~
~~GANNA FT00 FACTORS. CARD INPUTzo. INTERNALS NO. I BUILOUP FACTORS, CARD INPUTa0. INTERNAL *NO. a 4~~
~~0 GANNA HEAT CONVERSION FACTORS (N020.YES*I)~~
0 - 4 1
4 MM MM OO DO DO OO DO DO DO DO DO DO
+ + ++ + + + + + + + +
OO C0 OO DO OC DO to mM DO 00, OO DO 04 kN DO DO DO DO em MG OO DO DO DO Mb 4@ ed h DO DO OO e e e e e e e e e e e e MM e M N4 MM MM MM MM MM CO OO OO OO OO DO OO DO OO OO
+ + ++ ++ + + + + + +
OO DO DO CO DO OD eO Om ou OO DO DO 49 NM On OO DO DO
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~~Mb NN C4 OO OO OO O e e e e e o e e o e o e MM NM M4 HM MM MM I~~
MM MM OO OO DO OO OO DO C0 00 OO 00
+ + + + ++ + + + + + +
00 00 00 00 DO DO am 40 00 00 00 OO en eM OO OO O O' OO O @@ S fd NO DO OO C0 e eJ e Om OO DO 90 O e e e e me M. e e e e e
, HH HM M4 MM MM HM i
MMM Men OOO OOO OOO OOO C00 000 000 000 000 000
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000 000 000 000 000 000 eDO C40 000 000 000 000 MNN mom OOM 000 000 000
~~O mMO P o eJ 200 000 000 000~~
~~+ CMb fJ Q N 000 000 000 O ede e. ?. e e e e .~~
~~e e +~~
~~MMM MM4 MM4* MMM MMM M Haf MMM MMM MOD 000 000 000 000 000 000 000 000 000~~
~~+ + + + + e + + + + + +++ + + +~~
000 000 000 000 000 000 omm NoF C00 000 000 000 GN4 m&4 000 000 000 000 O ONM MMO CON . 000 000 000 000
+ Nme NMe (me e e e 000 e
000 O e e e e e e e o e o e i MMM MM4e PMW MMM MMM MMM MMM MMM MOO OOO OOO OOO COO COO OOO OOO OOD 000
+ + + + + + e + + + + + +++ + + +
000 000 000 000 000 000 POO NMe 400 000 000 000 NhM OOM 000 000 000 000 O @f4@ MNN COM 000 000 000
~~MMC m&4 MMe OOO e~~
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~~HMM MN4 4MM MMM MMM MMM MMM MMM MOO COO OOO OOO 000 000 000 000 000 000~~
+ + + + + + e + + + + + +++ + + +
000 000 000 000 000 000 44m Mhm 000 000 000 000 NMm JOM 000 700 000 000 O Med m@@ cod OOO 000 000
M4C MNN MGM 000 000 000 O e e e e *
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M MMM MMM OO OOO OOO OOO O 000 000 00 000 000 OOO O + + + + + + + + + + + +++ + +
O 000 000 CO 000 000 000 M Pem mMb DO 000 000 000 N MMO MON OO 000 000 000 P NMO hMe OM 000 000 000
C44 Omo 009 000 000 N e o e M@Oe o e e e e e e o e e o e e e MMM MN4 ONm MMM MMM MMM M
U R N M a a a 1 J C Z 4 - - -
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. H H H H tJ N N ts H ts es eJ OOOOOOOOOOOD UC0000000000 000000000000
- O000D0000000 000000000000 N N O M tJ 00000
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OO DO OO DO DO DO 0D OO O. O. O.O. O. O. O. P. P. O. M. M.
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COMP / MAT 1 2 1 1. 2 C .' 03- 0 3 0.0 2 0.0 2.24000+00 MATERIAL REFERENCE HUMBERS 12 13 GAMMA ATTENUATIDH CDEFFICIENT MAT /GRP 1 2 1 9.54000 02 9.63000-02 2 7.07000-02 7.09000-02 3 5.65000-02 5.6600D-02 4 4.89000-02 4.91000-01 5 4.28000-02 4.3100D-02 6 4.02000-C2 4.07000-02 7 3.33000-02 3.42000-02 8 2.52000-02 2.7000D-02
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CDORDINATE TYPE O SDURCE ItiTENSITY OPTIOff 2 e'
R COCRDINATE INTEt451TY COD 90INATE INTENSITY COORDINATE INTENSITY CDDRDINATE INTfH51TY CDORDINATE 1.1401D+03 3.11260+05 1.18560+03 3.236SD+05 1.2311D+03 3.36100+05 1 1.09460+03 2.9884D+05 3.85790 05 1.32210+03 3.6095D+05 1.36760+03 3.73370+05 1.4131D*03 5 1.27660+03 3.4853D + C5 1.59500+03 4.3553D+05 9 1.45c60+03 3.93210+05 1.5042D+03 4.115fD*05 1.5497D*03 4.23083+05 1.66620+03 4.60350+05 1.73170+03 *.72770+0S 1.77750+03 4.90593+05 13 1.64070+03 4.47920+05 5.3432D+05 4.92210+05 1.8682D+03 5.1003D+05 1.91370+03 5.20=5D+05 1.9592D+03 17 1.82300+03 FNI COORDINATE INTEt:3ITY INTENSITY COORDINATE INTENSITY CCORDIHLTE INTENSITY CDOFDIt4 ATE CCORDINATE 4.70000-01 6.40000-01 7.85003-01 6.20000-01 1.09500+00 6.20000-01 1 1.55000-01 6.20000-01 6.00000-01 6.40000-01 .1.7250D+00 6.20000-01 2.04000+00 6.43000-01 2.35000+00 5 1.41000+00 3.65000+00 6.00000-01 9 2.65000+00 6.00000-01 2.95000+00 6.00000-01 3.30000+00 8.00000-01 6.00000-01 4.5500D+00 6.0000D-01 4.66500+00 6.6000D-01 13 3.9500D*00 6.0000D-01 4.25000+00 5.49500+00 6.20000-01 5.81000+00 6.40000-01 6.12650+00 6.26000-01 17 5.18500+00 6.20000-01 Z COCRDINATE INTENSITY INTENSTTY CDDRDINATE INT EttSITY COORDINAT E CODRDINATE INTENSITY CDORDINATE 1.4444D+03 1.3941D+01 1.62490+03 1.3933D+01 1.80550+03 1.3941D*01 1 1.26 380 + 0 3 1.3941D+01 1.39410+01
~~1. 3941D~~
01 2.1667D*03 1.3941D+01 2.3472D+03 1.3933D+01 2.5278D*03 5 1.9061D+03 3.25010403 1.3941D+01 9 2.70840*03 1.3941D*01 2.88900603 1.3941D+01 3.06950+03 1.39330 01 1.39330+01 3.791eD+03 1.3941D*01 3. 9724D
0 3 1.3941D+01 13 3.4307D*03 1.3941D+01 3.6112D+03 4.3335D*03 1.39330+01 4.5141D+03 1.39410+01 4.69470+03 1.39410401 17 4.1530D+03 1.3941D+01 seen CASE SETUP TIME IN NIH. m 0.0
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J l NP-2 RADIATION QUALIFICATI0H OF ELECTRICAL EQUIPMENT ISOURCE*REGICil 4) g i RECEIVEP tRJMBER 1 CCDRDINATES - X 0.0 Y 0.0 Z 2.3825E+03 5 i
.. GRP MEAN ENERGY MEAN BUILDUP ENERGY FLUX CCSE' RATE HD ENERGY GRDUP LIMITS FACTD9S DIRECT BEAN WITH BUILDUP DIRECT BEAM WITH EUILCUP MEV MEV MEV/CM2-SEC MREM /HR i
1 0.4000 1.2314D+00 1.9319D+10 2.37900+10 4.0571D+07 4.99580+07 ,
1 2 0.8000 1.1353D+00 1.52480+11 1.7310D+11 3.0495D+08 3.4620D + 08 [~
i 3 1.3000 1.09260+00 5.84180+10 6.3S25D+10 1.0515D+03 1.1489C+08 j 4 1.7000 1.07460+00 1.6694D+11 1.79390+11 2.83300+08 3.04960+0S 5 2.2000 1.0621D+00 9.5922D+10 1.01870+11 1.53470+08 1.6300D+0S
-6 2.5000 1.0568D+00 5.40070+10 5.7075D+10 8.1011D+07 8.5612D+07 .
}
7 3.5000 1.04520+00 5.7915D+10 6.0533D+10 8.10600+07 .8.4746C+07 j 7.4189D-21 7.65070-21 8 6.1500 1.0313D+00 6.74440-18 6.95520-15 TOTAL ;.2760 6.0500Eill 6.5932D+11 1.0500E+09 1.1494E+09 i WDBU 1.3045 d
ENERGY FLUX EUILDUP DCSE EUILDUP 1.0902D+00 1.0946E+00 5 **** TIME FOR DETECTOR IN MIN. t 0.0 1
j *** END OF JOB E16103 ***
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ML20003H548

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