ML20150C641

Technical Evaluation Rept for Evaluation of Offsite Dose Calculation Manual,Rev 8

ML20150C641
Person / Time
Site:	Hope Creek
Issue date:	12/31/1987
From:	Amaro C, Duce S, Serrano W EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20150C630	List: ML20150C628 ML20150C641
References
CON-FIN-A-6034 EGG-PHY-7897, NUDOCS 8807120572
Download: ML20150C641 (24)

Text

[f1 CLOSURE

• O EGG-PHY-7897

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TECHNICAL EVALUATION REPORT for the EVALVATION OF ODCM REVISION 8 .

PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK GENERATING STATION UNIT NO. 1

- NRC Docket NO. 50-354 NRC LICENSE NO. NPF-57 S. W. Duce

. W. Serrano C. R. Amaro Published December 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls,, Idaho 83415

. Prepared for the U. S. Nuclear Regulatory Commission Washington, D.C. 20555- -

Under DOE Contract No. DE-AC07-761 DOI 570 .

FIN No. D6034 go71$$$$ck P 1 _ -

l ABSTRACT The Offsite Dose Calculation Manual for the Hope Creek Generating )

Station Unit No. I contains current methodology and parameters used in the calculation of offsite doses due to radioactive liquid and gaseo"us effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmental' radiological monitoring program. The entire ODCM manual up through Revision 2 dated January 1986 (incorporating flRC comments to Revision 1) was submitted to and approved by the NRC. -Subsequently, Revision 6 dated January 9, 1987, an entire ODCM document, containing changes from all previous', revisions up through Revisien 6 was submitted to the NRC in the Semiannual Radioactive Effluent Report for July-December 1986. The NRC transmitted Revision 6 to the Idaho Natior,al Engineering Laboratory (INEL) for review. Revision 6 l

was reviewed by EG&G Idaho and reported to the NRC in Technical Evaluation 5 Report EGG-PHY-7815 dated September 1987. With no prior knowledge of the concerns identified in the review of Revision 6, the Licensee submitted an entire ODCM document incorporating Revisions 7 and 8 to the NRC in the  !

Semiannual Radioactive Effluent Report for January-June 1987. The NRC transmitted the document to EG&G Idaho for review of the changes affected j by Revisions 7 and 8. The review of Revisions 7 and 8 combined with the l previous review of Revision 6 constitues a review of the entire ODCM document up through and including Revision 8. The combined results of both reviews are presented in this report. It is determined that the ODCM document up through Revision 8 uses methods that are, in general, in agreement with the guidelines of NUREG-0133. However, it is recommended that another revision to the ODCM be submitted to address and correct the discrepancies identified in the review.

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FOREWORD This report is submitted as partial fulfillment of the "Review of RadiologicalIssuesforBWRs"projectbeingconductedbytheIdjho National Engineering Laboratory for the the U. S. Nuclear Re'g ulatory Commission, Office of Nuclear Reactor Regulation. The U. S. Nuclear Regulatory Commission funded the work under FIN 06034 and NRC B&R Number 20 19 10 12 2. . l This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warrant, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information,

,. apparatus, product or process disclosed in this report, or reprasents that

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its use by such third party would not infringe privately-owned rights.

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. CONTENTS Pag;e Abstract. . . . . . . . . . . . . . . . . . . . . . . . . . . - ri .

Foreword. . . . . . . . . . . . . . . . . . . . . . . . . . .- ii

1. Introduction. . . . . . . . . . . ... . . . . . . . . . . 1-

2. Review Criteria . . . . . .' . c . . . . . . . . . . . . . 2

3. Evaluation. . . . . . . . . . . . . . . . . . . . . . . . 3

4. Concl u s i on s . . . . . . . . . . . . . . . . . . , . . . . 14

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5. References. . . . . . . . . . . . . . . . . . . . . . . . 20 ,

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1. INTRODUCTION 1.1 Purcose of Review l This Technical Evaluation Report (TER) contains the review Md evaluation of the Offsite Dose Calculation Manual (00CH) up through Revision 8 submitted by the Public Service Electric and Gas Company (PSE&G), the Licensee for the Hope Creek Generating Station (HCGS)

Unit 1. The ODCM is a supplementary document for implementing the.

Radiological Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix I requireme'nts.Ill ., i

. . . l 1.2 Plant-Soecific Backaroun_d 1

The PSE&G submitted ODCM Revisic.) 2 dated January 1986 for HCGS Unit 1

- to the Nuclear Regulatory Commission (NRC) with letter dated January 31, 1986.(2) The NRC reviewed the ODCM, and provided notification of approval to the Licensee with letter dated February 25,1986.[3]

Subsequently, an entire ODCM manual up through Revision 6 dated January 9, 1987 containing changes from all previous revisions was submitted to the NRC in the Semiannual Radioactive Effluent Report for July-December 1986.[43 Revision 6 of the ODCM was transmitted by the NRC to the Idaho National Engineering Laboratory (INEL) for review by EG&G Idaho, an independent review team. The complete ODCM up through Revision 6 was reviewed by EG&G Idaho and the results and conclusions were l reported to the NRC with TER dated September 1987.(5) j 1

With no prior knowledge of the Revision 6 review, the Licensee submitted to the NRC in the Semiannual Radioactive Effluent Report for the first half of 1987 with letter dated August 28,1987[6] an entire ODCM manual incorporating Revisions 7 and 8 and all previous revisions. The i

. document was transmitted by the NRC to EG&G Idaho for review of the changes affected by Revisions 7 and 8. This review combined with the previous review of the entire manual up through Revision 6 results in a complete review of the entire ODCM document up through and including 1

Revision 8. The results and conclusions of the combined review are l presented in this report. '

2. REVIEW CRITERIA

. .c Review criteria for the ODCM were provided by the NRC in'three documents: ,

NUREG-0472, RETS for PWRsI73 .

NUREG-0473, RETS for BWRs(8]

NUREG-0133, Preparation of'RETS for Nuclear Power Plantsd93 The following NRC guidelines were also used in the ODCM review: "General Contents of the Offsite Dose Calculation Manual," Revision 1[10], and Regulatory Guide 1.109.Illl As specified in NUREG-0472 and NUREG-0473, the ODCM is to be developed by the Licensee to document the methodology and approaches used to calculate offsite doses and maintain the operability of the radioactive effluent systems. As a minimum, the ODCM should provide equations and methodology for the following: .

o Alarm and trip setpoints on effluent instrumentation o Liquid effluent concentrations in unrestricted areas o Gaseous effluent dose rates at or beyond the site boun6 ,

o Liquid and gaseous effluent dose contributions o Liquid and gaseous effluent dose projections.

In addition, the ODCM should contain flow diagrams, consistent with the systems being used at the station, defining the treatment paths and the components of the radioactive liquid, gaseous, and solid waste management systems. A description and the location of samples in support of the environmental monitoring program are also needed in the ODCM. - -

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3. EVALUATION The HCGS Unit 1 is presently operational sharing Artificial Island,

. the plant site, with Salem Generating Station Units I and 2. At stated in the introduction of the ODCM, the manual contains information and metnodologies to be used by the HCGS Unit 1. Revision 8 addresses deficiencies identified in Revision 6 by utility personnel. -

3.1 Liouid Effluent Pathways Condenser cooling for HCGS Unit 1 is provided by water circulated through a natura! draft cooling tower. The HCGS is located on the shore of the Delaware River which supplies makeup water to the circulating water system and receives decant from the cooling tower via the cooling tower

, blowdown line. According to Unit 1 Technical Specification 3.3.7.9, there

is one monitored liquid effluent pathway for Unit 1; the liquid radwaste treatment system discharge.

. Liquid radwastes are discharged as batch releases from one of seven tanks: the two detergent drain tanks, the two floor drain sample tanks, the two equipment drain sample tanks, and the condensate storage tank.

The detergent drain tank liquid effluents are processed through a filter

. after leaving a detergent drain tank. All tank discharges are into the radwaste treatment system discharge line which empties into the cooling tower blowdown line. Thus, radioactivity released from the liquid radwaste system is the primary concern when assuring compliance to the concentration and the dose limit technical specifications. A block diagram description of the liquid radwaste treatment system and effluent pathways is shown in Figure 1.

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3.2 Liouid Effluent Monitor Setooints For compliance with HCGS Unit 1 Technical Specification 3.3.7.9,

, Section 1.2 of the ODCM contains the methodology used to determke the setpoints for the radiation monitors in the liquid radwaste syst'em using flow rates of the liquid radwaste discharge line and the cooling tower blowdown line. Discharge flow from the liquid radwaste discharge line and cooling tower blowdown line are monitored. It appears from Equation 1.1 that both flow rates can be adjusted to ensure the limits of Technical Specification 3.11.1 are not exceeded. However, this is not explicitly stated.

There are two radiation monitoring systems: a Liquid Radwaste Discharge Line Monitor which provides alarm and automatic termination of release; and a Cooling-Tower Blowdown Effluent Monitor which provides 5 alarm only. Sections 1.2.1 and 1.2.2 present the methodology for determining setpoints for the liquid radwaste radiation monitor and the cooling tower blowdown radiation monitor. The setpoints are calculated using an effective MPC value for the mixture of radionuclides in the effluent stream, discharge flow rates and the monitor's existing background. It is uncertain how the value of Ci in Expression 1.3 is obtained prior to release as this is a diluted value. The setpoints are determined for a high alarm only which is the value at which the concentration limits'will be exceeded. A low-level alarm should be included with a setpoint set slightly above the concentration in the tank to be discharged to prevent spurious alarms. A low level alarm would provide assurance that the correct tank is being discharged and would provide earl,y warning to the operators in the event of an unusual release

, situation. With these exceptions, the methodology described in Sections 1.2.1 and 1.2.2 to determine the setpoints for the radiation monitors in the licuid radwaste system is, in general, in agreement with the

, guidelines of NUREG-0133 to provide reasonable assurance that th,e concentration limits of Technical Specification 3.11.1.1 will not be exceeded. -

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I 3c3 Gaseous Eff' rothways According to HCGS Unit 1 Technical Specification 3.3.7.10, there are three monitored environmental gaseous effluent release points: .

a Filtration, Recirculation and Ventilation System e South Plant Vent j North Plant Vent - '

All gaseous effluent releases from HCGS are continuous releases with the I exception of the filtration, recirculation and ventilation system which is i used as an effluent treatment s'ystem for the reactor building. l containment /drywell purges and vents. Block diagram descripti'ons of the gaseous radwaste treatment system and the gaseous effluent pathways are shown in Figures 2 and 3, respectively. In addition to the three release points for HCGS there are two gaseous effluent release points for Salem

Units 1 and 2 which consists of the plant vent for each unit. Total gaseous releases from HCGS and the two Salem units are administratively allocated so that gaseous releases from all three units stay within the gaseous dose rates of Technical Specification 3.11.2.1.

3.4 Gaseous Effluent Monitor Setooints

. Section 2.2 of the ODCM contains the methodology used to determine the setpoints for the noble gas radiation effluent monitors. The methodology uses the highest annual average relative dispersion factor for areas at the site boundary and includes conservative factors to allow for simultaneous releases. The methodology is, in general, in agreement with the guidelines of NUREG-0133 to provide reasonable assurance that the noble gas dose rate limits of Technical Specification 3.11.2.1.a will not be exceeded.

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I 3.5 Concentrations in Liouid Effluents l

Section 1.3 of the ODCM contains the methodology for demonstrating l

. that radionuclide concentrations in liquid effluents are in compliance with 10 CFR 20.[12] The calculation includes all actual dilution during the release from the cooling tower blowdown. The methodology is, in general, within the guidelines of NUREG-0133 and should provide reasonable ,

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assurance that the concentrations at the point of release are maintained I within the limits of Technical Specification 3.11.1.1.

3.6 Dose Rates in Gaseous Effitents Section 2.3.1 of the ODCM contains the methodology for demonstrating compliance with the noble gas dose rate limits. Calculations will be made i in accordance with this methodology in the event any gaseous releases l result in an alarm setpoint being exceeded. The methodology is, in general, in greement with the guidelines of NUREG-0133 to provide reasonable assurance that the noble gas release rates are maintained within the limits of Technical Specification 3.ll.2.1.a. )

l Section 2.3.2 of the ODCM contains the methodology for determining the  !

dose rate due to the release of I-131, tritium, and all radionuclides in )

particulate form with half-lives greater than eight days to demonstrate compliance with Technical Specification 3.ll.2.1.b. Table 2.4, referenced in this section, does not include I-133 as required by Technical Specification 3.11.2.1.b. With this exception the methodology in Section 2.3.2 of the ODCM for determining the dose rate to a child's thyroid via j the inhalation pathway is, in general, within the guidelines of NUREG-0133  !

to provide reasonable assurance that the release rates are maintained within the limits of Technical Specification 3.ll.2.1.b.

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l 3.7 Dose Due to Liouid Effluents Section 1.4 of the ODCM contains the method for determining the dose ,

to the maximum exposed member of the public due to radionuclides l identified in liquid effluents to demonstrate compliance with the limits l of Technical Specification 3.11.1.2. NUREG-0133 states that.nortnally the adult is the maximum exposed individual and the exposure is via the potable water and aquatic food pathways only. The doses for HCGS are calculated based on two dose pathways: saltwater fish and invertebrates. I A simplified liquid effiuent dose calculation method is presented in j Section 1.4.2 and Appendix B. .The method is acceptable with the aception I that the constant 54.0, identified in Section 1.4.2, is calculated using fresh water bioaccumulation factors and is a factor of 10 too ' low. This is not consistent with the salt water pathways used in Equation 1.6 and J Table 1-3. Also, the phosphorus bioaccumulation factor of 3.0E+04 in l Table 1-3 for saltwater invertebrates could be changed to 6.0E+02 as  !

,_- allowed by the NRC.

With these identified discrepancies it is uncertain the methodology presented in Section 1.4 will provide reasonable assurance that the calculated doses due to the release of radioactivity in liquid effluents ,

will not exceed the limits of Technical Specification 3.11.1.2. It is certain that the methodology in Section 1.4.2 will over estimate the total

- body dose.

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3.8 Dose due to Gaseous Effluents Section 2.4 of the ODCM contains the methodology for calculating the gamma air dose and the beta air dose to demonstrate compliance with the air dose limits of Technical Specification 3.11.2.2. Themethodhlogyfor calculating the maximum air dose due to the release of radioactive noble gases is, in general, in agreement with the guidelines of NUREG-0133 to provide reasonable assurance that the dose limits of Technical Specification 3.11.2.2 will not be exceeded. .

Section 2.5 contains the methodology for calculating the. dose due to the release of I-131, tritium, and radionuclides in particulate form with half-lives greater than eight days. This section does not include I-133 as required by Technical Specification 3.11.2.3. Use of the SFp correction factor of 0.5 for milk and vegetation in Section 2.5.1, is only

- appropriate for the annual dose calculation. The value of 0.5 is not correct for a quarterly dose chiculation. The SFp factor for the quarterly dose calculations must be adjusted to the fraction of the quarter that these exposure pathways are applicable. With these exceptions, the methodology presented is, in general, in agreement with the guidelines of NUREG-0133 to provide reasonable assurance that the dose t

limits of Technical Specification 3.11.2.3 will not be exceeded.

3.9 Dose Pro.iections Section 1.5 of the ODCM describes the method used to project doses due to the anticipated release of radioactive liquids to determine required use of the liquid radwaste treatment system as required in Technical Specification 3.11.1.3. The equations as presented will work, however it is not stated that the dose projections will take into account liquid releases expected during the next 31 days.

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Section 2.6 of the ODCM describes the method used to project doses due to l the anticipated release of radioactive gaseous effluents to determine required use of the ventilation exhaust treatment system as reqttired in , l Technical Specification 3.11.2.5. For consistency with the reqtrirements of Technical Specification 4.11.2.5.1, it should be stated in the ODCM l that the dose projection is only required when the ventilation exhaust I treatment system is not being utilized. The equations as presented will work, however, it is not stated that the dose project, ions will takh into account gaseous releases expected during the next 31 days.  !

3.10 Diaarama of Effluent Pathways A simplified diagram illustrating the discharge pathway for the radioactive liquid waste system is included as Figure 1-1. Simplified

diagrams illustrating the discharge ;;thwa.

s for the radioactive gaseous l waste systems are included as Figi.:es 2-1 and 2-2. A simplified diagram illustrating the solid waste treatment rystem is not included in the ODCM.

i 3.11 Total Dose Section 3.2 of the ODCM describes the method used to demonstrate

- compliance with 40 CFR 190 including direct radiation. The method fer )

l determining the total dose is an acceptable method for demonstrating compliance with the dose limits of Technical Specification 3.11.4.

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3.12 Environmental Monitorino Procram Appendix Table E-1 of the ODCM identifies specific parameters of distance and the direction sector from the site and additional i) formation for most samples identified in Environmental Mcnitoring Table 3.12.1-1 of Technical Specification 3.12.1. Table E-1 of the ODCM does not include 43 direct radiation stations as locations 10F2 and 10G1 are listed twice.

Table 2.3 of the ODCM states the highest X/Q and D/Q are 0.5 miles N and 4.9 miles W, respectively. These locations are not included in the airborne sampling program listed in Table E-1. Table E-1 does not include any drinking water locations nor food product sample stations for crops ~

grown with water affected by the liquid effluents which is inc'onsistent with Table 3.12.1-1 of the Technical Specifications. If these pathways do not exist, then the Technical Specifications should be modified. Figures E-1 and E-2 are illegible.

e 3.13 Summary In summary, the Licensee's ODCM as revised uses documented and approved methods that are generally consistent with the methodology and guidance in NUREG-0133. However, it is recommended that the NRC request another revision to address the discrepancies identified in this review, I

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4. CONCLUSIONS The Licensee's ODCM, including all revisions through Revision 8, for the Hope Creek Generating Station was reviewed. It was determined that '

the ODCM uses methods that are, in general, consistent with thequidelines of NUREG-0133. However, it is recommended that another revision" to the ODCM be submitted to address ed correct the numerous discrepancies identified in the review.

l The following are considered to be major discrepancies:

. 1 o In Section 1.4.2 and Appendix B, salt water bioaccumdlation factors must be used instead of fresh water factors in the calculation of the totti body default term.

1 o In Sections 1.5 and 2.6 it is not stated that the dose  ;

_I projections will take into account effluent releases expected in l l

the next 31 days. 1 i

o I-133 is missing throughout Tabic 2 4 and is not mentioned in Section 2.5.1. . 1 o In Equation 2.11 in Section 2.5.1, use of the SFp correction factor of 0.5 for milk and vegetation it correct only for the l annual dose calculation. The value of 0.5 is not correct for a l quarterly dose calculation. The SF p factor for the quarterly dose calculations must be adjusted to agree with that fraction of the quarter that these exposure pathways are applicable.

The following are additional discrepancies:

o In Section 1.2, 10 CFR 30 should be changed to 10 CFR 20.

o In Section 1.2.1, the background term must be in the same units as the setpoint, i.e., pCi/ml, instead of cpm. .

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o In Sections 1.4.1 and 1.4.2, the definition of term CTBD must have "release period" replaced with "calculation period" to state the flowrate is averaged over the entire calculational period, e

o In Section 2.2.1, the units of "SP" and "bkg" should be pCi/s. .

o In Section 2.2.2, the radionuclide distribution should be br. sed on historic plant aver, ages instead of data from the ANSI .

Standard. .

o In Section 2.4.2, the definition of Meff, the value should be 8.1E3 (as calculated in Appendix C) instead of 8.1E2.

,- o In Section 2.5.1, the last paragraph identifies the Child-Vegetation pathway as the limiting pathway for the child which is inconsistent with the information in Table 2-3.

o In Section 2.5.2, the RI-131 value is 1.67E12 whereas Table 2-4  ;

lists the value as 1.05E12. l l

o In Section 3.2, the wording of the second paragraph is I confusing. .It appears that words "Hope Creek" should be replaced with "Salem".

o In Table 1-1 there are several values that are different from those presented in Rev. 6. It can not be determined which values are the correct values.

o In Table 1-2 there are a few values that have been changed from those listed in Rev. 6. The value for Y-91m for bone has been i*

changed to 5.72E+2 from 5.72E-2, and C-14 for the GI-LLI has been

!. changed to 2.90E+5 from 2.90E+3. It is not clear which value is

, CorrtCt.

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o In Table 1-3, the value for I for saltwater invertebrates should l be 2.5E+01 instead of 2.5E+02.

o In Table 1-3, the phosphorus bioaccumulation factor of 3.0E+04 for saltwater invertebrates could be replaced with 6.0E+02.

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c o Figures 2-1 and 2-2 reference Figures 4.1-2 and 4.1-3. Should these references be to Figure 2-2 and Figure 2-1, respectively?

o A simplified diagram illustrating the solid waste treatment system is not included,in the ODCM.

o In Table 2-1, the units of Mi and Ni should be mrad /yr per pCi/m3 . Also the Mi value for Xe-133 should be 3.53E+02 instead of 3.35E+02.

o Table 2-2 references Table 3.1-1. This reference should be changed to be Table 2.1-1. The units of Ci should be pCi/cm3 and the units of Mi should be mrad /yr per pCi/m3 ,

o Table 2-4 must have I-133 included in all pages of the table to be consistent with Technical Snecification 3.11.2.3. Within

. Table 2-4 there are several values that need to be checked for

) correctness. The nuclides and organs are identified in the following list:

Teen Inhalation: Ag-110m lung, Te-125m lung Child Inhalation: Rb-86 liver Infant Inhalation: Fe-59 lung Adult Cow-Milk: Rb-86 all organs, Zr-95 all organs, Nb-95 all organs.

Teen Cow Milk: I-131 all organs, Ba-140 liver, Ce-141 all .

organs

• Child Cow-Hilk: Y-91 all organs Child Vegetation: Ce-141 all organs.

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o In Table A-1, the MPC values for P-32 and Cr-51 should be 2E-5 and 2E-3, respectively. In Table A-1 under the expected activity released column, the values for Rh-106 through Ba-139 are

- incorrect. 7 o In Table A-4 the Na-24 MPC value should be 2E-4. In. addition the expected activities released for Sr-92 and Nb 95 are much larger than those used in Rev. 6. Are those values listed in Rev. 8 the correct values? .

o Table B-1 does not include values for Co-60 and Co-58 which are required according to the text of Appendix B.

o On page C-3, the table referenced should be Table C-1 instead of B-1. Also on pages C-3 and C-4 the term D x should be Dg .

o Table C 1 has incorrect Heff and Neff values for Xe-135, Xe-135m and Kr-88.

o On page D-3 the definition of Qi must include H-3.

o Table E-1 of the ODCM does not include 43 direct radiation stations as, required in the Technical Specifications as locations )

10F2 and 10G1 are listed twice. .

l l

o Table 2.3 of the ODCM states the highest X/Q and D/Q are 0.5 l miles N and 4.9 miles W, respectively. These locations are not l included in the airborne sampling program listed in Table E-1.

o Table E-1 does not include any drinking wate. locations nor food product sample stations for crops grown with water affected by the liquid effluents which are required by Table 3.12.1-1 of the Technical Specifications. If these pathways do not exist, then

, the Technical Specifications should be modified. -

o Figures E-1 and E-2 are illegible.

17

The following are not discrepancies in the ODCM, but are suggestions that should be brought to the attention of the Licensee:

o It is recommended that use of Ci be limited to the accepted use as the unit for curie and that another symbol be used for concentrations throughout the document. Also the use of the l i

terms e and C should be reviewed to determine if better terms could be used to avoid confusion, e.g., see Equation 1.1. l I

o In Section 1.2.1 the clefinition for Ci should state that it is the diluted concentration. , l

.- l l

o On pages C-2 and C-3 the meaning of "ds" is not clear in the 1 definition of the terms fi and Hi.

o A low level alarm setpoint should be considered for the liquid effluent monitor.

o For consistency with the requirements of Technical Specification 4.11.2.5.1, it should be stated in the ODCM that the dose projection for anticipated gaseous releases is only required when .

the ventilation exhaus treatment system is not being utilized.

1 18 I

l

.Fw-eI12Sim 6 -h- - ,"ph m- m -

5. REFERENCES

1. Title 10, code of Federal Reaulations, Part 50, Appendix I, "Numerical GuidesforDesignObjectivesandLimitingConditionsforOppationto Meet the Criterion, 'As Low As Is Reasonably Achievable,' for s Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents."

2. Letter from C. A. McNeill (PSE&G) to E. Adensam (NRC), Subject Offsite Dose Calculation Manual Revision 2 Hope Creek Generating -

Station, Docket No. 50-354, January 31, 1986. .-

3. Letter from E. Adensam (NRC) to C. A. McNeill (PSE&G),

Subject:

Hope Creek - Offsite Dose Calculation Manual, February 25, 1986.

4. Letter from C. A. McNeill (PSE&G) to NRC,

Subject:

Radioactive Effluent Release Report - 2 Hope Creek Generating Station, i February 27, 1987. ,

5. S. W. Duce, W. Serrano, C. R. Amaro, "Technical Evaluation Report for the Evaluation of ODCM Revision 6 - Public Service Electric and Gas Company, Hope Creek Generating Station, Unit No. l' , EGG-PHY-7815, September 1987.

6. Letter from C. A. McNeill (PSE&G) to NRC,

Subject:

Radioactive Effluent Release Report - 3 Hope Creek Generating Station, August 28, 1987.

7. "Radiological Effluent Technical Specifications for Pressurized Water Reactors," Rev. 3, Draft 7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0472, September 1982. .

19

4 &

8. "Radiological Effluent Technical Specifications for Boiling Water Reactors," Rev. 3, Draft 7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0473,

. September 1982. =

r

• l l

9. "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, A Guidance Manual for Users of Standard Technical Specifications," NUREG-0133, October 1978. ,

j l

10. "General Contents of the Offsite Dose Calculation Manual.," Revision 1 )

Branch Technical Position, Radiological Assessment Branch,' NRC, February 8, 1979. l

11. Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, l e Appendix I," Regulatory Guide 1.109, Rev. 1, October 1977. l

12. Title 10, Code of Federal Reaulations, Part 20, "Standards for Protection Against Radiation."

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