ML20205E021
Text
.
SAFETY EVALVATION BY THE OFFICE OF WUCLEt9 REACTOR REGULATION RELATING TO ACCEPTANCE OF THE OFFSITE DOSL CALCULATION MANUAL UPDATED THROUGH REVISION 1 BOSTON EDISON PILGRIM GUCLEAR POWER STATION DOCKET NO. 50-293 1.0 1hTR000CT10N On August 30, 1985 the staff issued Amendment No. 89 to Facility Operating License No. OPR-35 for the Pilgrim Nuclear Power Station (Pilgrim). The amendment incorporated the Radiological Effluert Technical Specifications (RETS) into the Pilgrim Technical Specifications (TS). Section 6.9.C.3 of the TS referenced an Offstte Dose Calculation Manual (00CM) and prescribed the rethods for reporting changes.
2.0 EVAlt;ATION The docketed submittal on June 16, 1983, of an ODCM by Boston Edisen (licenste) received NRC approval by letter dated August 30, 1985 from T
H. Leech to the licensee. Recently in their Semi-Annual Racioactive Effluent and Waste Disposal Report for the Period January 1 through June 30, 1987, submitted by letter dated September 1,1987, the licensee submitted revised pages to the ODCM, labeled Rev. 1.
As part of an ongoing review of licensee ODCMr, the Pilgrim ODCM updated through Revi-sion 1. has been reviewed for us in its entirety by EG&G Idaho, Inc.
(EG&G) as part of our technical assistance contract program. The contrac-tor's Technical Evaluation Report (TER), which is enclosed as Appendix 0 EGG-PHY-7725, provides a technical evaluation of the compliance of the licensee's submittal with NRC criteria. The staff has reviewed this
??k027
'Joanco 800926 O'J 90 pg
report and agrees with the evaluation that th'a Pilgrim 00CM updated through Revision 1 uses docun4nted and approved methods that are general-ly consistent with the methodology and guidelines in NUREG-0133. There-fore, we conclude that this 00CM is an acceptable interim reference for 1
l use with the Pilgrim Technicai Specifications. However, the enclosed TER lists a number of discrepancies and suggestions that should be addressed f
1 l
within six months in a new revision to the Pilgrim ODCM.
3.0 CONCLUSION
S i
The Pilgrim ODCM, upcated through Revision 1. is acceptable on an interim basis. Discrepancies noted in the attached TER should be addressed within six months in a revised ODCM submission.
l l
l l
l i
i l
l l
f
EGG-PHY-7725 1
1 i
APPEN0!X 0 Evaluation of Changes to the 00CM, PCP, and Radwaste Treatment Systems (Pilgrim Nuclear Power Station) t l
l l
l l
i I
l l
l l
l l
i l
o
,4 (Blank Page) 02
o O.1 EVALVATION OF CHANGES TO THE ODCM The Boston Edison Company (BEco) prepared an Offsite Oose Calculation Manual (00CM) for the Pilgrim Nuclear Power Station.
The Boston Edison Company submitted this 00CM, Revision 0, dated 6/10/83 to the Nuclear Regulatory Commission (NRC) with letter dated June 16, 1983 E13 The NRC found it to be generally consistent with NRC criteria and an acceptable reference as stated in the NRC letter and SER dated August 30,1985.(2)
The licensee submitted changes labeled Rev.1 to the Revision 0 00CM in the Semiannual Radiological Effluent Release Report isdued for the first 6 i
months of 1987.[33 These changes have been incorporated into the l
Licensee's existing 00CM and, at the request of the NRC, the entire 00CM j
reviewed as a whole. The result of the evaluation is intended to be a stand-alone document, and is given in Supplement 1 to Appendix 0.
D.2 EVALUATION OF CHANGES TO THE PCP i
j No technical specification exists requiring use of a Process Control Program (PCP).
Contequently, it appears that the licensee has rot i
prepared a PCP.
0.3 REPORTED CHANGES TO THE RADWASTE TREATMENT SYSTEMS
'T No technical specification exists requiring the licensee to report to the NRC major changes made to the liquid, gaseous, or solid radwaste j
treatment systems.
Therefore, if changes are made to these systems, they are reported to the NRC in the annual FSAR updates.
l 03 i
l
0.4 REFERENCES
1.
Letter from W. D. Harrington (BEco) to 0. B. Vassallo (NRC)
Subject:
Offsite Oose Calculation Manual, June 16, 1983.
2.
Letter from P. H. Leech (NRC) to W. D. Harrington (BEco),
Subject:
Acceptance of Offsite Dose Calculation Manual (00CM) for Pilgrim
[
Nuclear Power Ste. tion Unit 1. August 30, 1985.
3.
Letter from R. G. Sir 1Eco) to Document Control Oesk (NRC),
Subject:
i Semi-Annual Radioactive Effluent and Waste Disposal Report for the Period January 1 through June 30, 1987, September 1, 1987.
l l
l l
l l
4 i
l l
l i
1 04 i
A
A- - ---
a e
9 SUPPLEMENT 1 to APPEN0!X 0 EVALUATION OF CHANGES TO THE 00CM l
,7 y
-w.y3-
w
-.-m.-
-,-w
l 1
i I
(Blank Pege) l I
1 i
l a
i i
j 4
A I
b 1
01-2 i
1
l 1.
INTRODUCTION l
purpose of Review
]
This document reports the review and evaluation of the latest revised version of the Offsite Ooss Calculation Manual (00CM) submitted by the l
Boston Edison Company (BECo), the licensee for the Pilgrim Nuclear Power I
Station.
The 00CM is a supplementary document for implementing the Radiological Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix ! requirements.Cl3 i
plant-Specific Backaround 5
)
The Boston Edison Company submitted ODCM, Revision 0, dated 6/10/83 I
for the Pilgrim Nuclear Power Station to the Nuclear Regulatory Comission l
f (NRC) with letter dated June 16,1983.[23 Yhe NRC found it to be l
generally consistent with NRC criteria and an acceptable reference as stated in the NRC letter and SER dated August 30,1985.C33 J
The licensee submitted changes labeled Rev.1 to the Revision 0 ODCM j
with the Semianaual Radiological Effluent Release Report issued for the l
first 6 months of 1987.C93 These changes have been incorporated into the Licensee's existing 00CM and, at the request of the NRC. the satire l
00CM reviewed as a whole.
The result of the evaluation is intended to te a stand-alone decoment and is presented in this report.
i 1
1
)
I J
01-3
o 2.
REVIEW CRITERIA i
Review criteria for the 00CM were provided by the NRC ii three documents:
NUREG-0472. RETS for PWR.".43 i
I NUREG-0473 RETS for BWRs[5]
j NUREG-0133, Preparation of RETS for Nuclear Power Plants.(6) f I
The following NRC guidelines were also used in the 00CM review:
"General
)
Contents of the Offsite Oose Calculation Manual," 4evision 1(73, and Regulatory Guide 1.109.(8) l As specified in NUREG-0472 and NUREG-0473, the ODCM is to be developed j
by the licensee to document the methodology and approaches used to l
calculate offsite doses and maintain the operability of the radioactive l
j.
effluent systems. As a minimum, the 00CM should provide equations and l
methodology for the following:
o Alarm and trip setpoints on effluent instrumentation j
o Liquid effluent concentrations in unrestricted areas
]
o Gaseous effluent dose rates at or beyond the site boundary o
Liould and gaseous effluent dose contributions
]
o Liquid and gaseous a.ffh:ent dose projections.
I In addition, the 00CM stould contain flow diagrams, consistent with the systems being used at tht station, defining the treatment paths and the i
componsats of the radicactive liquid, gaseous, and solid waste management
]
systems. A description and the location of samples in support of the l
environmental monitoring program are also needed in the 00CM.
i 4
W 2
4 01-4 4
l 1
3.
EVALVATION 1
The Pilgrim Nuclear Power Station is a single unit nuclear site.
As stated in the introduction of the 00CM, the manual contains information and methodologies to be used by the Pilgrim Nuclear Power Station.
The manual is structured such that it should be unnecessary to refer to otner documents to perform the indicated calculations.
Liould Effluent pathways The Pilgrim Nuclear Power Station is located on the western shore of Cape Cod Bay in the town of Plymouth Plymouth County, Massachusetts.
Liquid effluents are discharged with the once-through condenser cooling water into the bay.
The principal sources of liquid radwaste are the following:
l
)
Clean Waste Tanks Chemical Waste Tanks Miscellaneous Waste Orain Tanks Radwaste, Released Directly to Environment Effluents from the first three sources are processed in the liquid effluent treatment system (LETS).
The fourth source is mentioned in a later paragraph.
The LETS was designed to handle radioactive, chemica',
and miscellaneous liquid wastes.
There is one environmental release point at the site for the processed liquid radwastes.
The system is operated as a batch system and the operating procedures used for all liquid radwaste equipment are based on batch processing throughout the system.
This type of operation allows time to sample and check the effluent batches before and after each process step to present inadvertent discharge of waste having a radioactivity level above the control limit.
Each batch is analyzed prior to release for gross beta / gamma activity, and the resulting specific activity is used to determine the discharge flow rate.
Liquids with radioactivity levels exceeding spec led limits are recycled for further processing.
From the descripti' in 00CM Section 3.2.3, it appears 01-
that discharges from the first three sources are released to a comon header where they are monitored fc
'diation.
The radiation monitor provides alarm and automatic ters, on of release through the discharge valves upon a high radiation con:
There are two waste discharge valves: one is locatad on a one-inch line from the comon header and the other is or. 4 two-inch line from the common header.
The batch release is briefly discussed in 00CM Section 3.1.3.
During liquid releases, the flow rates, and activity levels are continuously recorded.
According to 00CM Section 3.3.2, the radwaste discharge flow is maintained at a predetermined level (not to exceed l
200 gpm). The liquid radwaste effluents are released to the condenser coolirg water discharge canal prior to discharge into Cape Cod bay.
Therefore, the flow control valves and the radistion monitors are the primary methods for controlling discharges from the liquid radwaste system.
In addition to batch releases from the LETS, batch releases from other sources directly to the environment are permitted provided that at least two independent samples are analyzed in accordance with l
Specification 4.8.A.1 as described in 00CM Section 3.1.3.
In addition, independent verifict.tions of the release rate calcul+tions and discharge valving must be performed.
Concentraticns released to the unrestricted areas must also be limited to the values specified in 10 CFR 20.
The j
location of this environmental release point should be identified in the i
CDCM.
1 Liquid Efflusnt Monitor Setpoints 00CM Section 6.1 contains the methodology for determining the setpoint for the liquid radwaste radiation monitor.
The monito* provides alarm and automatic termination of.slease.
The setpoint ensures that the concentration of liquid effluents discharged does n6t increase above the value for which the maximum permissible discharge flo.s was established.
l In other words, the setpoint is set at the level determined from the i
prerelease grab sample.
Since there is no margin allowed, the monitor should be alarming continuously during a release thus 01-6 l
1
preventing the release.
if the setpoint is increased to allow for a i
margin, then the discharge flow (Section 3.3.2) must be decreased accordingly.
In 00CM Section 6.1.3, the documentation for estimating the monitor's efficiency "based on prior release experience" is not referenced' The methodology described in 00CM Section 6.1 for determining the setpoint for the radiation monitor in the liquid radwaste system is, in general, in agreement with the guidelines of NUREG-0133 to provide reasonable assurance that the concentration limits of Technical i
Specificatien 3.8.A.1 will not be exceeded.
However, it is not clear if i
the monitor's actual setpoint is set at the level described in this section.
l l
l Gaseous Effluent pathways _
l There are two monitored environmental gaseous effluent release points at the Pilgrim Nuclear Power Station:
i Main Stack Gas Release I
Reactor Building Exhaust Vent Release The technical specifications identify noble gas monitors and iodine and i
I l
particulate samplers.
Each release point is continuously surveyed during t
releases for noble gases by two monitois.
Each monitor has two upscale trips and one downscale trip.
Each trip initiates an alarm in the main control room, but no automatic termination is provided.
The upscale alarms indicate high radiatinn and the downscale a' arm indicates instrument trouble.
Each release point has iodine and particulates samplers in the gas monitoring stream The samplers are routinely analyzed in accordance with Technical Specification Table 4.8-3.
All gaseous effluent releases from the reactor building exhaust vent are treated as ground level releases and the main stack releases are treated as elevated releases.
01-7
Gaseous Effluent Monitor Setpoints, l
Section 4.2 of the 00CM contains the mechodology used to determine the setpoints for the noble gas radiation monitors.
In items 3) and 4) however, re'ference is made to the equation of Section 4.0 but instead should be made to the equations in Sections 4.F and 4.G.
Simultaneous releases from these two release points are considered when determining each monitor's setpoiat.
This section 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 3.8.0.1 will not be exceeded.
Concentrations in Liouid Effluents Section 4.A of the 00CM contains the methodology for demonstrating that the radionuclide corcentrations in the released liquid effluents are in compliance with the technical specificition.
The option for determining the quantity C g in the concentration equation by "estimates w
based on prior experience" is not acceptable as.his is not permitted in the liquid sampling Table 4.8-1 of the Techtical Scecifications.
The i
methodology however, is in general, within tie guidelines of NUREG-0133 l
(
and should provide reasonable assurance that thi concentrations at the point of release are maintained within the limits of Technical Specification 3.8 A.1.
Dese Rates in Gaseous Effluents i
The equations in Sections 4.0 through 4,L are general equations that are used to determine both the doses and the dose rates due to the gaseous effluents.
It is not clear from the equations if contributions from both the main stack and the reactor building vent are included.
Each individual equation should contain a summation over the stack cod vent contributions.
01-8 i
l Sections 4.F through Section 4.G contain the equations for calculating l
noble gas dose rates to determine compliance with dose rate Technical i
Specification 3.8.0.1.a.
The uose rates due to *.he release of noble gases are assured ti be within the dose rate limits by correctly setting the setpoints for the noble gas monitors as prey!ously described in this report.
l Sections 4.H through 4.L of the 00CM contain the equations for determining dose rates for iodine-l'al, iodine-133, tritium, and all radionuclides b particulate form with half Ilves greater than S days to areas at and beyond the site boundary as specified in Technical l
Specification 3.8.0.1.b.
The titles however identify "Hal gens, I
Particulates and others" instead of "iodine-131, iodine-133, tritium, and i
all radionuclides in particulate form with half lives greater than 8 days". The bases statement for Technical Specification 3.8.0 states that the release rate of these nuclides restricts at all times the thyroid dose j
rate to a infant via the cow-milk-infant pathway to less than or equal to j
1500 mrems/ year.
The more restrictive age group however is the child l
instead of the infant as identified in Revision 3 Oraft 7" of NUREG-0473 datedSeptember1982.(5) The Itcensee should consider changing the
)
bascs statement in their next request for a technical specification
- revision, j
G The equation for C g in Section 4.H contains an "i" in the 7
denominator whereas it should be "Ag".
The constants 1.2x10,
2.2x10, 5.5x10, and 1.1x108 in Sections 4.J through Section 4.L 7
7 are not defined, The definitio% for the quantity Qt should not include the word "annual" since it is already being considered "for the period".
The time unit "hours" has been omitted from the definition for t,.
In f
general, the sections dealing with "Doses for Gaseous Effluents" could include supplementary information to provide guidance and clarity to the user.
l 01-9
i r
l It is uncertain if the maximum organ dose is identified since the f
technical specification bases statement identifies the inf.nt age group instead of the child age group. Nevertheless, the equations in j
Sections 4 H through 4.L are in agreement with NUREG-0133 and Regulatory Guide 1.109 and should provide reasonable assurance that the dose rate f
Ilmit of Technical Spect'fication 3.8.0.1.b will not be exceeded.
Dose Due to I.iould Effluents l
l Sections 4.8 and 4.C of the 00CM contain the method for determining the dose to the maximum exposed member of the public due to radionuclides identified in liquid effluents to demonstrate compliance with the dose limits of Technical Specification 7.2.
The egaations in Sections 4.8 and 4.C are in agreement with those of Regulatory Guide 1.109.
The methods l
include all age groups using the aquatic foods, and shoreline det,osits pathways.
The calculations could be made to only the adult age group since NUREG-0133 identifies the adult as the limiting age group.
.l Condenser cooling is a once-through system and provides dilution water for j
the liquid radwaste releases.
The dilution flow in the equations represented by "F" and "M " should be replaced with the avern e f
p condenser cooling flow for the period to change the dilution flow to the j
average flow of the discharge canal during the reporting period.
The methodology for calculating doses due to the release of radioactivity in liquid effluents is, in general, in agreement with the guidelines of NUREG 0133 and the methodology should provide reasonable assurance that the calculated doses will be within the limits of Technical Specification 3.8.A.1 Dose due to Gaseous Effluents I
Sections 4.0 and 4.E of the 00CM contain the equations for calculating the cumulative dose due to the release of radioactive noble gases in gaseous effluents to demonstrate compilance with the dose limits of 01-10
The values for X/Q in Table 5-1 cre evaluated at and beyond the site bourdary.
The table lists the X/Q values for both the main stack and the reactor building vent.
As mentioned previously, it is not clear from the equations if the dose contributions from 'soth the main stack and the reactor butiding vent are included.
Each individual ecuation should contain a s,umation over the stack and vent f
contributions.
Except for this uncertainty concerning the dose contribution from both release points, the methodology for calculating the l
maximum dose to air due to the release of radioactive noble gases is, in l
general, in agreement with the guidelines of NUREG 0133 to provide reasonab5e assurance that the dose limits of Tecnnical Specification 7.3 l
will not be exceeded.
l l
Sections 4.H through 41, of the 00CM contain the equations for calculating the cumulative dess due to the release of 1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater l
than eight days to deurstrate compliance with the dose limits of l
With the exception of the previously mentioned discrepancies which will be identified in the conclusion j
section, the methodology is, in general, within the guidelines of NUREG-0133 and should provide reasonable assurance that the dose limits of l
Technical Specification 1.5 will not be exceeded.
(
l l
l Dose projections l
Technical Specification 3.8.C.1 require.s that the liquid radwaste treatment system be operated whenever doses due to liquids to be released would exceed certain dose limits. However, the corresponding Surveillance Specification 4.8.C.1 requires that the doses be calculated due to liquids j
that have been released.
Thus, the surveillance specification does not require the dose orojection required by the technical specification.
The 00CM, consequently, does not include a dose projection due to liquid radwaste releases. A dose projection should be incorporated into the 00CM to satisfy the Technical Specification 3.8.C.1 requirements.
01-11
l l
L l
There is no technical specification identifying required use of the ventilation exhaust treatment system.
Consequently, there is no j
l requirem et for projacting doses due to gaseous releases trem the reactor j
building exhaust vent.
l
\\
Dianrams of Effluent pathways I
l i
Simplified diagrams of the 11guld and gasenus radweste treatment systems are contained in Figure 8-1 and Figure 8 2, respectively. A sivlified diagram illustrating the solid waste treatment system is not included in the 00CM.
i Figure F-1 should be modiflod to show the one-inch and the two inch discharge lines, the release pathway to the discharge canal, and the environmental release point for liquid radwastes released without i
treatment.
l Figure 4.8-2 in the technical specifications shows the drywell effluents being released to the main stack whereas Figure S-2 in the 00CM shows these effluents being released to the reactor building vent.
The figures should be consistent.
j Diagrams showing the radiation monitoring systems are shown in Figures 3.1, 3.2, and 3.3.
These radiation monitoring system diagrams are 111egible and should be replaced.
\\
Total Dose There is no separate section in the 00CM addressing the 40 CFR 190 total dose limits of Technical Specification 7.5.
Therefore, there is no expression given for calculating the total dose from the liquid, gaseous.
and direct radiation contributions.
01-12
l l
Environmental Monitorina proaram P
i Table 7-1 in tection 7.0 of the 00CM identifies specific parameters fcr distance and tne direction sector from the site and additional j
information for each and every sample identified in Environmental Monitoring Table 8.1 1 of Technical Specification 7.0.
Tne direction for j
Duxbury appears to be NW of the plant site instead of "$$W SW" as indicated in Table 7 3.
Figures 7.1 through 7.4 are ill6gible and should f
be replaced.
Summary l
In sumary, the licensee's 00CM uses documented and approved methods j
l that are generally consistent with the methodology and guidance in NUREG-0133. However, because of the discrepancies identified in this review, it is recomended chat the NRC request a revision to address the concerns identified in this review, t
1 l
i 01-13
=
I 4
CONCLU$!0N!
t The licensee's 00CM, updated through Revision 1, for the Pilgrim Nuclear Power Station was reviewed.
It was determined that the 00CM uses methods that are, in general, consistent with the guidelines of NURUi 0133. However, it is recommended that a revision to the 00CM be submitted to address the discrepancies identified in the review.
f I
i The following is considered to be a major discrepancy:
f l
o In Section 4.1, it is uncertain if the dose rate to the child's thyroid is identified as the maximum organ dose since the bases statement-in Technical Specification 3.8.0 identifies the infant
(
age group instead of the child age group.
I The following are additional discrepancies:
l l
l o
In Section 3.1.3, the location of the environmental release point l
for liquid radwaste batch releases from sources other than the liquid radwaste treatment system should be ide3tified in the 00CM.
l 1
o Figures 3.1, 3.2, and 3.3 contain diagrams sh v ig the radiation menitoring systems. These radiation monitoring system diagrams are illegible and should be replaced.
j
\\
o in Section 4.A. the option for determining the quantity C,g in the concentration equation by "estimates based on prior exp.rienc.
is not consist.nt witn itauid so.iing Tani.
. -1 of the technical specificetions, o
In the equations of Sections 4.5 and 4.C. the dilution flow is represented by "F" and "M," and should be repiated with the avocage condenser cooling flow for the period to change the dilution flow to the avera p flow of the discharge canal during the reporting period.
01-14
l 1
o In Sections 4.0 through 4.L. it is not clear from the equations that simultaneous dose rate coatributions from the main stack and the rei.: tor building vent are included, j
o l'n Sections 4.H through 4.L., the titles identify "Halogens, Particulates and others" instead of "iodine-131, iodine 133, l
l tritium, and all radionuclides in particulate form with half Ilves greater than 8 days".
f G
o In Section 4.H. the equation for C j contains an "i" in tne denominator tihereas it should be "Aj".
r 7
7 o
In Sections 4.J through 4.L. the constants 1.2x10, 2.2x10,
5.5x10, and 1.1x108 are not defined.
l 7
l o
in Section 4.L., the definition for the quantity Qt should r.ot i
include the word "annual" since the air dose or air dose rate is alrency being considered "for the period".
l f
o In Section 4.L. the time unit "hours" has been omitted fr i the s
definition for t,.
i
?
o In Section 6.1, the setpoint for the liquid radwaste monitor is set to the ievet detereined from the o,ereiease gra sam,ie with no margin allowed, lt is not clear if plant operation is j
consistent with the 00CM description since the monitor should bw alarming continuously during a release, thus preventing the release.
l o
In Section 6.1.3, the documentation for estimating the monitor's efficiency "based on prior release experience" is not referenced.
l i
o in Section 6.2, items 3) and 4) reference the equation of Section 4.0 and should reference the equations in Sections 4.F and 4.G.
i 01-15
r i
i o
A calculation should be included in the 00CM to project doses due to the release of radioactivi*y in liquid effluents to satisfy l
the requirement of Technical Specification 3.8.C.1.
f
{
o A' simplified diagram illustrating the snlid waste treatment system is not included in the 00CM.
l l
l l
o There is no separate section in the 00CM addressing the total dose limits of Technical Specification 7.5 with methodclogy for calculating the total dose from the licuid, gaseous, and dircet radiation contributions, o
The diraction for Dumbury is NW of the plant site instead of l
"SSW-SW" as indicated in Tsble ?-3.
o Figures 7.1 through 7.4 are illegible and snould be replaced.
l o
Figure 8 1 should be modified to show the one-inch ant as two-inch discharge lines, th release pathway to the discharge j
- anal, and the environmental release point for itquid radwestes l
released without treatment.
o Figure 8-2 in the 00CM shows the drywell efflueets being released to the reactor buit( ig vent whereas Figure 4.8 2 in the technical specifications shows these effluents being released to I
I the main stack.
The figures should be correct and consistent, j
i i
i
[
01-16
The following are not discrep.ncies in the 00CM, but are suggestions that should be brought to the attention of the licensee:
l l
o Table 2.1 does not contain reference to Section 4.L.
l o
Section 4 in the Tabve of Contents only contains Sections 4.A through 4.K instead of through 4.L.
1 o
In Sections 4.8 and 4.C. the calculations could be made to only tne adult age group since NUREG-0133 identifies the adult as the limiting age grovo, o
The licensee should consider needifying the bases statement is Technical Specification 3.8,0 to change from the infant to the child age group which is the most restrictive age group for the dose rate calculation.
o The licensee should modify Surveillance Specification 4.8.C.1 in the technical specifications to include a dose projection to saticfy the requirement of Technical Specification 3.8.C.1, 01-17
I
{
\\
l I
I i
(
l s
P f
i i
i r
l
[
r (Blank Page)
\\
l l
t i
l l
I l
i I
l 01-18
4 5.
REFERENCES 1.
Title 10, Code of Federal Regulations, Part 50, Appendix I, "Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion, 'As Low As Is Reasonably Achievable,' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents."
2.
Letter from W. D. Harrington (BECo) to 0. B. Vassallo (NRC).
Subject:
Offsite Oose Calculation Manual, June 16, 1983.
3.
Letter from P. H. Leech (NRC) to W. D. Harrington (BEco),
Subject:
Acceptance of Offsite Oose Calculation Manual (00CM) for Pilgrim Nuclear Power Station Unit 1, August 30, 1985.
4.
"Radiological Effluent Technical Specifications for Pressurized Water Reactors," Rev. 3, Oraft 7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0472, September 1982.
5.
"Radiological Effluent Technical Specifications for Boiling Water Reactors," Rev. 3, Oraft 7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0473, September 1982.
6.
"Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, A Guidance Manual for Users of Standard Technical $pecifications," NUREG-0133, October 1978.
7.
"General Contents of the Offsite Dose Calculation Manual," Revision 1 Branch Technical Position, Radiological Assessment Branch, NRC, February 8, 1979.
8.
"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Regulatory Guide 1.109, Rev. 1, October 1977.
01-19
9.
Letter from R. G. Bird (BEco) to Documsnt Control Desk (NRC),
Subject:
Semi-Annual Radioactive Effluent and Waste Disposal Report for the Period January 1 through June 30, 1987, September 1, 1987.
i e
i
{
I I
I 01-20