ML19269C891
| ML19269C891 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/13/1979 |
| From: | Vandenburgh D VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WVY-79-15, NUDOCS 7902200044 | |
| Download: ML19269C891 (62) | |
Text
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Proposed Change No. 78 VERMONT Y ANimE N UCLEAR POWER COR POR ATI()N SEVENTY SEVEN GROVE STREET PC 78-1 u.2.1 Rtm.no, hm asr onvoi NEPLY TO:
ENGINEERING OFFICE TURNPlKE RO AD WESTDORO M ASS ACHUSETTS 01581 TELEPHONE 617 366-9011 WVY 79-15 February 13, 1979 United States Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Of fice of Nuclear Reactor Regulation
References:
(a) License No. DPR-28 (Docke t No. 50-271)
(b) Letter from B. K. Grimes, Assistant Director for Engineering and Projects, USNRC, to All Power Reactor Licensees, dated July 11, 1978 (c) Letter from B. K. Grimes, Assistant Director for Engineering and Projects, USNRC, to All Boiling Water Reactor Licensees, dated November 15, 1978
Dear Sir:
Subj ect : Vermont Yankee Radiological lf fluent Technical Specifications Pursuant to Section 50.59 of tae Commission's Rules and Regulations, Vermont Yankee Nuclear Power Corporation hereby proposes the following modification to Appendix A of the Operating License.
PROPOSED CHANGE: Reference is made to the Operating License DPR-28 and the Technical Speelfications contained in Appendix A issued to the Vermont Yankee Nuclear Power Corporation for the Vermont Yankee Nuclear Power S ta tio n.
We propose to make the following changes:
(1)
Add to Section 1.0 " Definitions," new definitions for " Ins trument Functional Tes t,"
" Source Check," and "Of f site Dose Calculation Manual," in accordance with the use of these terms in the new specifications outlined below.
(2)
Delete Table 3.2.4, "Of f-gas System Isolation Instrumentation," and replace pages 33, 58, 65 and 65a with revised pages. Updated Section 3.8.I incorporates present criteria on of f-gas monitoring in accordance with NRC Ef fluent Standard Technical Specifications.
(3)
Replace Section 3.8/4.8, " Radioactive Ef fluents," with new Section 3.8/4.8.
The contents of this section dealing with radioactive liquid and gaseous ef flut nts has been revised to reflect the guidance put forth by the NRC in the Ef fluent Standard Technical Specifications.
The updated specifications address both liquid and gaseous ef fluent releases, Steam Jet Air Ejector releases, hydrogen gas concentrations, 7 9 0 2 20 0 0 VV
United States Nuclear Regulatory Commission February 13, 1979 Attn: Office of Nuclear Reactor Regulation Page 2 resulting doses offsite from radioactive effluents, and availability and use of radwaste equipment.
(4)
Replace Section 3.9/4.9, " Radioactive Ef fluent Monitoring Systems,"
with new Section 3.9/4.9.
The contents of this section concerning ef fluent monitoring and surveillance requirements has been revised to reflect the guidance given by the NRC in the Effluent Standard Technical Specification. The new specifications deal with liquid and gaseous cffluent monitoring, including Steam Jet Air Ejector releases and explosive gas mixtures, environmental radiation monitoring, land use census information, and intercomparison program requirements which deal with quality control of laboratory analyses performed as part of the Environmental Radiation Monitoring Program.
(5)
As part of the Administrative Controls, add to Section 6.5.A, " Plant Operating procedures," the requirement to includa the Radiological Environmental Monitoring Program, the implement. ten of the Offsite Dose Calculation Manual, and effluent monitoring instrumentation setpoints in accordance with the new specifications which cover these items.
(6)
As part of the Administrative Controls, add to Section 6.7.A,
" Reporting Requirements," a new Section 6.7.A.4, " Annual Radiological Environmental Operating Report" in accordance with guidance put forth in the Effluent Standard Technical Specification.
(7)
As part of the Administrative Controls, add to Section 6.7.B.2, " Thirty Day Written Reports," the requirement to report the uncontrolled release of radioactive materials above given quantities and the measured levels of radioactivity in environmental sampling media in excess of given limits.
(8)
In Section 6.7.C, " Unique Reporting Requirements," of the Administrative Controls, add new Section 6.7.C.3 for special reports required under Specifications 3.8.B.1, 3.8.E.1, 3.8.F.1 and 3.8.G.I.
Revised Technical Specification pages are provided with this letter.
REASON FOR CHANGE: The proposed changes are in direct response to the USNRC's request (References b and c) that Vermont Yankee Nuclear Power Corporation amend Vermont Yankee's Operating License.
BASIS FOR CHANGE: The proposed Technical Specifications address issues put forth by the USNRC in their Draft Radiological Effluent Standard Technical Specifications (References b and c) and are intended to implement the following Federal Regulations:
10 CFR Part 50, Section 50.36a, Section 50.34a(a), Section 50.34a, 10 CFR Part 20, 10 CFR Part 50, Appendix A, General Design Criteria 60 and 64, and 50 CFR Part 190.
United States Nuclear Regulatory Commission February 13, 1979 Attn: Office of Nuclear Reactor Regulation Page 3 Since the waste disposal system at Vermont Yankee cocplies with the regulations set forth in 10 CFR 50.34a and Appendix A Criterion 60, we believe that further commitments to your requests in Reference (c) cannot be made at this time.
We would be interested in discussing this topic further with your staff, after receipt and review of the generic value-impact and specific value-impact assessment for Vermont Yankee, in support of your proposed regulatory actions. As you are aware, NRC Chairman Joseph Hendrie committed to provide value-impacts in his July 21, 1978, response to Executive Order 12044. This response has been documented in Federal Register Vol. 43, No.
150, dated August 3, 1978, pages 34358 and 34359 which states in part:
"The Policy of tne Nuclear Regulatory Commission is that value-impact analysis be conducted for any proposed regulatory actions that might impose a significant burden on the public... [and) where there are alternative means of realizing equivalent benefits... cost should be a prime consideration."
Certainly the resultant modifications of the waste disposal systems of many nuclear power plants to provide for the intended conversion of radioactive wastes from liquid systems to a homogeneous monolithic, immobilized free standing solid under the process control requirements of Reference (c) will impose a significant economic impact
"...on the nuclear industry and hence on electric consumers." Lastly, we are interested in the documentation upon which your staff concluded that these proposed actions will provide substantial additional protection of the public health and safety as stipulated under 10 CFR 50.109.
We have included with this letter a draft of the "Offsite Dose Calculation Manual (ODCM)" for implementing the dose requirements of the proposed Technical Specifications. Work on this manual is continuing and a final version will be forwarded to you for your review within sixty days.
Along with the ODCM, we have included one-line diagrams indicating the effluent flow paths for both liquid and gaseous radwaste, the location of all effluent monitors, and those equipment items which are included in the liquid and gaseous radwaste treatment system.
SAFETY CONSIDERATIONS: The changes proposed were requested by the USNRC Draft and are not considered to constitute sn unreviewed safety question.
This change has been reviewed by the Nuclear Safety Audit and Review Committee.
FEE DETERMINATION:
The major portion of the proposed change is an extension of the 10 CFR Part 50, Appendix I design study submitted to the USNRC on June 2, 1976, and constitutes completion of the requirenents of Appendix I for the submittal of Technical Specifications. We conclude that this
United States Nuclear Regulatory Commission February 13, 1979 Attn: Office of Nuclear Reactor Regulation Page 4 amendment is exempt from any fees defined in 10 CFR Part 170.12(c) since fees were not applicable when the requirements put forth by Appendix I to 10 CFR Part 50 became effective, and since submittal of this information has been delayed pending guidance from the USNRC which was issued in July, 1978 (Reference b).
SCHEDULE OF CHANGE: These changes will be incorporated into Vermont Yankee's Technical Specifications 90 days after approval by the Commission, but in no case before November 30, 1979.
We trust you will find this submittal satisf actory; however, should you desire additional information, please feel free to contact us.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION
%E\\'
D. E. Vandenburgh Vice President COMMONWEALTH OF MASSACHUSETTS)
)ss.
COUNTY OF WORCESTER
)
Then personally appeared before me, D. E. Vandenburgh, who, being duly sworn, did state that he is Vice President of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to < execute and file the foregoing request in the name and on the behalf c,f Vermont Yankee Nuclear Power Corporation, and that the statements theren.n are true to the best of his knowledge and belief.
Roaert H. Groce Notary Public My Commission Expires September 14, 1984
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e VERM0!;T i A::KEE t4UCLEAR P(NER CORPO:'ATI0:1 1,1 cense !!o. DPR-28 Docket t:o. 50-271 Instructions for enterinr, change p a r,e s to the Vermont Yankee ?;uclear Power Station Technical Specificat ions, Appendix A.
Remove Page_
Add *;ew Page lii 111 Iv iv 9
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33 33 46 46 58 58 65 65 6 5;,
147 thru 160 14 7 t iiru l fiUe 161 thru 172 161 thru 172i 200 200 201 201 210 210 213 213 213a 214 214
P e
VERMONT YANKEE EFFLUENT TECHNICAL SPECIFICATIONS
TABLE OF CONTENTS (CONT)
LIMITING CONDITIONS OF OPERATION Page No.
SURVEILLANCE F.
Automatic Depressurization System --------------------------
92 F
G.
Reactor Core Isolation Cooling System (RCIC) ---------------
93 G
H.
Minimum Core and Containment Cooling System Availability --------------------------------------------
94 H
I.
Maintenance of Filled Discharge Pipe -----------------------
95 I
J.
Average Planar LHGR ----------------------------------------
96 J
K.
Local LHGR -------------------------------------------------
96 K
3.6 REACTOR COOLANT SYSTEM -------------------------- -
105 4.6 A.
T h e rma l L i m i t a t i o n s - -- --- -- -- ------------ -- - --------- -- --- -
105 A
B.
Coolant Chemistry ------------------------------------------
106 B
C.
Coolant Leakage --------------------------------------------
108 C
D.
S a f e t y a n d Re li e f Va l v e s ------------------------------- ---
108 D
E.
Structural I n t e g r i t y - - - -- - ---- -- - - - - - - - - - --- - - - - - - - -- - -- - ---
109 E
F.
JetPumps--------------------------------------------------
109 F
G, Recirculation Pump Flow Mismatch -- ------------------------
110 G
3.7 STATION CONTAINMENT SYSTEMS ------------------------------------
126
.7 A.
Primary Containment 126 A
B.
Standby Gas Treatment System ---------------- --------------
130 B
C.
Secondary Containment System -------------------------------
131 C
D.
Primary Containment Isolation Valves -----------------------
132 D
3.8 RADIOACTIVE EFFLUENTS ------------------------------------------
147 4.8 A.
Liquid Effluents:
Concentration ---------------------------
147 A
B.
Liquid Effluents:
Dose ------------------------------------
148 B
C.
Liquid Waste Treatment ------------
148 C
D.
Gaseous Effluents: Dose Rate -
150 D
E.
Gaseous Effluents:
Dose From Noble Gases -------------
150 E
F.
Gaseous Effluents: Dose From Radiciodines, Radioactive 1laterials in Particulate Form and Radionuclides Other Than Noble Cases -------- --
151 F
G 152 G.
Dose Commitment -----
153 H
H.
Explosive Gas Mixture 153 1
I.
Steam Jet Air Ejector (SJ AE) --- -- -
J.
153 J
154 K
K.
Gaseous Waste Treatment ------ --
iii
TABLE OF CONTENTS (CONT)
LIMITING CONDITIONS OF OPERATION Page No.
SURVEILLANCE 3.9 RADI0 ACTIVE EFFLUENT MONITORING SYSTEMS -------------- --------
161 4.9 A.
Liquid Ef fluent Instrumentation ---------------------------
161 A
B.
Gaseous Effluent Instrumentation -- -------- -
162 B
C.
Radiolcgical Environmental Monitoring Program -------------
162 C
D.
Lan d U s e C e n s u s ------------------------- -------- - -
163 D
E.
Intercomparison Program -----------------------------------
164 E
3.10 AUXILI ARY ELECT RI CAL POWE R SY ST EMS ----------------------------
173 4.10 A.
N o rm a l Op e r a t i o n ------------------------------------- -----
173 A
B.
Operation with Inoperable Components ----------------------
176 B
C.
Diesel Fuel -----------------------------------------------
177 C
3.11 R EACTOR FU EL AS S EMB L I ES - - ----------------- --------------------
180-a 4.11 A.
Average Planar LHGR ------------ -- --------- ---------- --
180-a A
B.
LHGR --------------------------------- --------------------
180-b B
C.
MCPR ----------------------------- ------------------------
180-b C
180-1 D.
Reporting Requirements ---------------
3.12 REFUELINC AND SPENT FUEL HANDLING --------
181 4.12 A.
Refueling Interlocks ------------ -------------------------
181 A
B.
Core Monitoring ------------------------------------- -----
182 B
C.
Fuel Storage Pool Water Level --------------- ---------- --
183 C
D.
Control Rod and Control Rod Drive Maintenance -------------
184 D
E.
Extended Core Maintenance ------------ ---
184 E
185 F
F.
Fuel Movement G.
Crane Operability -----------------------------------------
185 G
H.
Spent Fuel Pool Water Temperature -- - ----------- --------
185a H
iv
VYNPS G.
Instrument Functional Test -
N.
Peaking Factor - The ratio of the fuel rod heat flux to the heat flux of an average rod in an a.
Analog channels - the injection of a simulated identical geometry bundle operating at the signal into the channel as close to the sensor average core power, as practicable to verify operability including alarm and/or trip functions.
b.
Bistable channels - the injection of a simulated O.
Primary Containment Integrity - Primary containment signal into the sensor to verify operability integrity means that the drywell and pressure including alarm and/or trip functions.
suppression chamber are intact and all of the following conditions are satisfied:
H.
Logic System Functional Test - A logic system functional test means a test of all relays and 1.
All manual containment isolation valves on contacts of a logic circuit from sensor to activated lines connecting to the reactor coolant device to insure all components are operable per system or containment which are not required design intent.
Where possible, action will go to to be open during accident conditions are closed.
completic~.,
i.e.,
pumps will be started and valves opened.
2.
At least one door in each airlock is closed and scaled.
I.
Minimum Critical Power Ratio - The Minimum Critical Power Ratio is defined as the ratio 3.
All automatic containment isolation valves are of that power in a fuel assembly which is operable or deactivated in the isolated position.
calculated to cause some point in that assembly to cxperience boiling transition as calculated by 4.
All blind flanges and manways are closed, application of the GEXL correlation to the actual assembly operating power.
P.
Protective Instrumentation Definitions (Reference NEDO-10958) 1.
Instrument Channel - An instrument channel means J.
Mode - The reactor mode is that which is an arrangement of a sensor and auxiliary equipment established by the mode-selector-switch.
required to generate and transmit to a trip system a single trip signal related to the plant K.
Operable - A system or component shall be considered parameter monitored by that instrument channel.
operable when it is capable of performing its intended function in its required manner.
2.
Trip System - A trip system means an arrangemant of instrument channel trip signals and auxiliary L.
Operating - Operating means that a system or equipment required to initiate action to accomplish component is performing its intended functions a protective trip function. A trip system may in its required manner.
require one or more instrument channel trip signals.
M.
Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
2.
VYNPS W.
Shutdown - The reactor is in a shutdown condition Z.
Surveillance Frequency - Unless otherwise stated in when the reactor mode switch is in the shutdown these specifications, periodic surveillance tests, checks, mcde position and no core alterations are being calibrations, and examinations shall be performed, within performed. When the mode switch is placed in the the specified surveillance intervals. These intervals shutdown position a reactor scrcm is initiated, may be adjusted plus 25%.
The total maximum combined power to the control rod J. rives is removed, and interval time for any three consecutive tests shall not the reactor protection system trip systems are exceed 3.25 times the specified interval. The operating de-energized.
cycle interval is considered to be 18 months and the tolerances stated above are applicable.
1.
Hot Shutdown means conditions as above with reactor coolant temperature greater than 212 F.
AA.
Surveillance Interval - The surveillance interval is the calendar time between surveillance test, checks, cali-2.
Cold Shutdown means conditions as above with brations, and examinations to be performed upon an reactor coolant temperature equal to or less instrument or component when it is required to be than 212 F.
These tests, unless otherwise stated in these specifications, may be waived when the instrument, 3.
Shutc)wn ccans conditions as above such that component or system is not requ red to be operable, but the effective multiplication factor (Keff) f these tests shall be performed on the instrument, component the core shall be less than 0.99.
or system prior to being required to be operable.
X.
Simulated Automatic Actuation - Simulated automatic BB.
Vital Fire Suppression Water System - The vital fire accuation means applying a simulated signal to the suppression water system is that part of the fire
~
sensor to actuate circuit in question.
suppression system which protects those instruments, components and systems required to perform a safe shutdown Y.
Transition Boiling - Transition boiling means the of the reactor. The vital fire suppression system includes boiling regime between nucleate and film boiling.
the water supply, pumps and distribution piping with Transition boiling is the regime in which both associated sectionalizing valves, which provide immediate nucleate and film boiling occur intermittently coverage of the Reactor Building, Control Room Building, with neither type being completely stable, and Diesel Generator Rooms.
CC.
Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
DD.
Offsite Dose Calculation Manual (ODCFQ - A manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents.
(Changes to the calculation methodology in the ODCM shall be subject to NRC review and approval prior to implementation.
Any changes in plant related parameters or dose conversion factors shall be submitted to the NRC concurrent with implementa-tion and documented in the subsequent semiannual Radioactive Effluent Release Report.)
4.
VYNPS 3.2 LIMITING CONDITIONS FOR OPERATION 4.2 SURVEILLANCE REQUIREMENT B.
Primary Containment Isolation B.
Primary Containment Isolation When primary contairment integrity is Instrumentation and logic systems shall be required, in accordance with Specification functionally tested and calibrated as 3.7, the instrumentation that initiates indicated in Table 4.2.2.
primary containment isolation shall be operable in accordance with Table 3.2.2.
C.
Reactor Building Ver.tilation Isolation and C.
Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Standby Gas Treatment System Initiation The instrumentation that initiates the Inntrumentation and logic systems shall be isolation of the ; actor building, ventilation functionally tested anu calibrated as system and the actuation of the standby gas indicated in Table 4.2.3.
treatment system shall be opercible in accordance with Table 3.2.3.
D.
Off-Gas System Isolation D.
Off-Gas System Isolation During reactor power operation, the Instrumentation and logic systems shall be instrumentation that initiates isolation functionally tested and calibrated as of the off-gas system shall be operable indicated in Table 4.2.4 and 4.9.2.
In accordance with Table 3.9.2.
E.
Control Rod Block Actuation E.
Control Rod Block Actuation During reactor power operacion the instro-Instrumentation and logic system; shall be mentation that initiates control rod functionally tested and calibrated as block shall be operable in accordance indicated in Table 4.2.5.
with Table 3.2.5.
33
t D
4 4
W M
PW W
A A
W cc Q
4H
VYNPS TABLE 4.2.4 MINIMLSI TEST L CALIBRATION FREQUENCIES OFF-CAS SYSTDI ISOLATION INSTRDIENTATION Trip Function Functional Test (8)
Calibration (81 Instrument Check Air ejector Logie Bus Power Monitor (Note 1) none once each day Trip System Logic (5JAE) every 6 months every 6 months (Note 2)
(Note 3)
Augmented off-gas Trip System Logic ( A0r.)
every 6 months every 6 months (Note 2)
(Note 3)
F 58
VYNPS 3.2 (Continued)
The APRM rod block trip is flow referenced and prevents a significant reduction in MCPR especially during operation at reduced flow.
The APRM provides gross core protection; i.e.,
limits the gross core power increase f rom withdrawal of control rods in the normal withdrawal sequence.
The trips are set so that MCPR is maintained greater than the fuel cladding integrity safety limit.
The IRM rod block function provides local as well as gross core protection.
The scaling arrarnement is such that trip setting is less than a factor of 10 above the indicated level.
Analymis of the worst case accident results in rod block action before MCPR approaches the fuel cladding integrity safety limit.
A downscale indication on an APRM or IRM is an indication the instrument has f ailed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus control rod motion is prevented.
To prevent excessive clad temperatures for the small pipe break, the HPCI or Automatic Depressuri-zation System must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.
The arrangement of the tripping contacts is such as to provide this f unction when necessary and minimize spurious operation.
The trip settings given in the specification are adequate to assure the above criteria are met.
The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration and also minimizes the risk of inadvertent operation; i.e.,
only one instrument channel out of service.
Four radiation monitors are provided which initiate isolation of the reactor building and operation of the standby gas treatment systen.
The monitors are located in the reactor building ventilation duct and on the refueling floor. Any one upscale trip or two downscale trips of either set of monitors will cause the desired action. Trip settings for the monitors on the refueling floor are based upon initiating normal ventilation isolation and 65
VYNPS 3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 3.8 RADI0 ACTIVE EFFLUENT _S 4.8 RADIOACTIVE EFFLUENTS Applicability Applicability Applies to the controlled release of all radioactive Applies to the required surveillance during controlled effluents from the plant.
releases of all radioactive effluents from the plant.
Objective Objective To assure that radioactive effluents are kept "as To ascertain that all radioactive effluents released low as is reasonably achievable" in accordance with from the plant are kept "as low as
's reasonably 10CFR50, Appendix I and, in any event, are within acnievable" in accordance with 10CFR50, Appendix I,
~
the limits specified in 10CFR20.
and in any event, are within the limits specified in 10CFR20.
Specification Specification A.
Liquid Effluents:
Concentration A.
Liquid Effluents:
Concentrati n 1.
The concentration of radioactive material in 1.
The concentration of radioactive material liquid effluents released from the site to in liquid effluents released from the site unrestricted areas (see Figure 3.8.1) shall be shall be monitored in accordance with Table limited to the concentrations specified in 4.8.1.
The results of the analyses shall 10CFR Part 20, Appendix B, Table II, Column be used to assure that the concentrations 2 for radionuclides other than noble gases and at the point of release are Jimited to the 2 x 10-4 ci/ml total activity concentration values in Specificatini 3.8.A.l.
p for all dissolved or entrained noble gases.
2.
With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits, immediately decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits.
147
3.8 LIMITING CONDITION 3 FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS B.
Liquid rffluents: Dose B.
Liquid Effluents: Dose 1.
The dose or dose commitment to an individual from 1.
Cumulative dose contributions shall be determined radioactive materials in liquid effluents released in accordance with the ODCM once per month. A to unrestricted areas (see Figure 3.8.1) shall be cumulative summation of these total body and limited to:
organ doses shall be maintained for each calendar quarter and included in the semiannual Radioactive a.
During any calendar quarter to$_1.5 mrem to the Effluent Release Report, pursuant to Specification total body and to 15 mrem to any organ; 6.7.C.1.
L.
During any calendar year to 13 mrem to the total body and to 110 mrem to any organ.
2.
With the calculated dose from the release of radio-active materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.C.3, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose or dose commitment to an individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ.
C.
Liquid Waste Treatment C.
Liquid Waste Treatment 1.
The Liquid Radwaste Deep Bed Demineralizer 1.
The Liquid Radwaste Deep Bed Demineralizer shall shall be routinely used to reduce the radioactive be demonstrated operable prior to use unless it materials in the liquid radwaste eff'uent has been utilized during the previous three months.
stream prior to its discharge to unt :stricted areas; or 148
VYNPS 3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 2.
With the Liquid Radwaste Deco Bed Demineralizer unavailable for use due to nu intenance or testing, or if the waste stream chemistry is incompatible with the radwaste demineralizer, the release of liquid waste to unrestricted areas may continue without further action provided that the resultant cumulative doses in unrestricted areas does not exceed 50% of the limits of Specification 3.8.B.1.a.
3.
With liquid waste being discharged without treatment and with the resultant doses exceeding 50% of the limits of Specification 3.8.B.l.a, prepare and submit to the Commission within 30 days pursuant to 6.7.C.3, a Special Report which includes the following information:
a.
The reason for inoperability, b.
Action (s) taken to restore the inoperable equipment to operable status.
c.
Summary description of action (s) 'aken to prevent a recurrence.
149
VYNPS 3.8 LIMITING CONDITION FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS D.
Gaseous Effluents: Dose Rate D.
Caseous Effluents: Dose Rate l.
The instantaneous dose rate in unrestricted 1.
The dose rate in unrestricted areas due to areas (see Figure 3.8.1) due to radioactive radioactive materials released in gaseous effluents materials released in gaseous effluents from shall be determined to be within the limits of the site shall be limited to the following:
Specification 3.8.D.1.
The dose rate shall be determined by using the results of the sampling a.
The dose rate limit for noble gases shall be and analysis program (specified in Table 4.8.2,
$500 mrem /yr to the total body, except for noble gases, and Table 3.9.2 for noble gases) in the calculations required by the ODCM.
53000 mrem /yr to the skin 2.
Thc elease rate of radioactive materials released b.
The dose rate limit for all radioiodines and in gaseous effluents from the site shall be radianctive materials in particulate form and determined to be within the required limits by radionuclides other than noble gases with half obtaining representative samples in accordance lives greater than 8 days shall be 51500 mrem /yr with the sampling and anlysis program specified to any organ.
in Table 4.8.2.
2.
With the dose rates exceeding the above limits, immediately take action to decrease the release rate to comply with the limit.
E.
Gaseous Effluents: Dose from Noble Gases E.
Gaseous Effluents: Dose from Noble Gases 1.
The air dose in unrestricted areas (see Figure 1.
Cumulative dose contributions for the total time 3.8.1) due to noble gases released in gaseous period shall be determined in accordance with the ef fluents shall be limited to the following:
ODCM once per month.
a.
During any calendar quarter to; 2.
Calculated dose contributions shall be summed 15 mrad for gamma radiation, quarterly and included in the semiannual
<10 mrad for beta radiation; Radioactive Effluent Release Report pursuant to Specification 6.7.C.l.
b.
During any calendar year to; 110 mrad for gamma radiation, 120 mrad for beta radiation.
150
VYNPS 3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 2.
With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.C.3, a Special Report which identifies the cause(s) foc exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose during this period is within 10 mrad for gamma radiation and 20 mrad for beta radiation.
F.
Gaseous Effluents: Dose from Radiciodines, Radioactive F.
Caseous Effluents: Dose from Radioiodines, Radioactive Material in Particulate Form and Radionuclides Other Material in Particulate Form and Radionuclides Other Than Noble Cases Than Noble Gases 1.
The dose to an individual from radiciodines, 1.
Cumulative dose contributions for the total time radioactive materials in particular form and period shall be determined in accordance with the radionuclides with half-lives greater than 8 days' ODCM once per month.
other than noble gases in gaseous effluents released to unrestricted areas (see Figure 3.8.1) shall be 2.
Calculated dose contributions shall be summed limited to the following:
quarterly and included in the semiannual Radio-active Effluent Release Report pursuant to a.
During any calendar quarter <7.5 mrem, Specification 6.7.C.l.
b.
During any calendar year <l5 mrem 2.
With the calculated dose from the release of radio-iodines, radioactive materials in particular form, or radionuclides other than noble gases in gaseous efflu-ents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.C.3, a Special Report which identifies the cause(si for exceeding the limit and defines the corrective actions to be taken to reduce the releases of radiciodines, radioactive materials in particulate form, and radionuclides with half lives greater than 8 days other than 151
VYNPS 3.8 LIMTING CONDITION FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS noble gases in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose or dose commitment to an individual from such releases during this period is within 15 mrem to any organ.
G.
Dase Commitment G.
Dose Commitment 1.
The dose or dose commitment to a real individual 1.
Cumulative dose contributions from liquid and from all station sources is limited to 5.25 mrem gaseous effluents shall be determined in accordance to the total body or any organ (except the thyroid, with Specifications 4.8. B.1, 4. 8.E.1, 4. 8. F.1, which is limited to 575 mrem) over a period of 12 and in accordance with the ODCM.
consecutive months.
2.
With the calculated dose f rem the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cations 3.8.B.1.a, 3.8.B.l.b, 3.8.E.1.a. 3.8.E.1.b, 3.8.F.1.a or 3.8.F.1.b above, prepare and submit to the Commission within 90 days, pursuant to Specification 6.7.C.3, a Special Report and limit the subsequent releases such that the dose or dose commitments to a real individual from all station sources is limited to 5.25 mrem to the total body or any organ (except thyroid, which is limited to 5.75 mrem) over 12 consecutive months. This Special Report shall include an analysis which demonstrates that radiation exposures to all real individuals from all station sources (including all effluent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard.
152
VYNPS 3.8 LIMITING CONDITIONS FOR OPERATION 4.8 SURVEILLANCE REQUIREMENTS 11.
Explosive Gas Mixture H.
Explosive Gas Mixture 1.
If the hydrogen concentration in the off-gas 1.
The concentration of hydrogen in the off-gas downstream of the operating recombiner reaches system downstream of the recombiners shall be four percent, an immediate transfer to the continuously monitored by the hydrogen standby recombiner shall be made.
monitors required operable by Table 3.9.2.
I.
Steam Jet Air Ejector (SJAE) 1.
Steam Jet Air Ejector (SJAE) 1.
The Air Ejector suction valves shall close 1.
The gross radioactivity rate of noble gases within 1 minute if the gross radioactivity from the SJAE shall be determined to be within exceeds 1.5 Ci/sec (30 minute decay level).
the limit of Specification 3.8.I.1 at the following frequencies by performing an isotopic 2.
The Air Ejector suction valves shall close analysis (for Xe-138, Xe-135, Xe-133, Kr-88, if the gross radioactivity exceeds 0.3 Ci/sec Kr-85m, Kr-87) on a representative sample of (30 minute decay level) unless the level gases taken at the discharge.
decreases to less than 0.3 Ci/sec within 15 minutes.
a.
Once per month.
b.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following an increase of greater than 50% in steady-state activity levels during steady-state reactor operation, as indicated by the SJAE monitor.
J.
Prima ry Containment 1.
If the primary containment is to be vented 1.
The containment shall be sampled prior to purging (excluding pumpback venting), it shall be per Table 4.8.2, and if the results indicate vented / purged through the Standby Gas Treatment radioactivity levels in excess of the limits System whenever the airborne radioactivity levels of Specification 3.8.J.1, the containment shall exceed the levels specified _n 10CFR20, Appendix be aligned for purging through the Standby Gas B. Table I, Column 1 and notes 1-5 thereto.
Treatment System.
153
VYNPS 3.8 LIMITING CONDITIONS FOR OPERATI'N 4.8 SURVEILLANCE REQUIREMENTS K.
Gaseous Waste Treatment K.
Gaseous Waste Treatment l.
During reactor power operation, the Augmented 1.
The A0G system shall be demonstrated operable Off-Gas System (A0G) shall be routinely at least once per quarter unless it has been used to reduce radioactive materials in utilized during the previous three months.
gr'a waste prior to their discharge to e
unrestricted areas; or 2.
With the A0G unavailable for use during reactor power operation due to maintenance or testing, the release of gaseous waste to unrestricted areas may continue without further action provided that the resultant cumulative doses in unrestricted areas does not exceed 50% of the limits of Specification 3.8.E.1.a, or
- 3. 8. F.1. a.
3.
With off-gas being discharged during reactor power operation without treatment and with the resultant doses exceeding 50%
of the limits of Specification 3.8.E.1.a, or 9.F.1.a.
prepare and submit to the Commission within 30 iays pursuant to l
Specification 6.7.C.3, a Special Report which includes the following information:
a.
Identification of equipment of subsystems not operable and the reason for inoperability.
b.
Action (s) taken to restore the inoperable equipment to operable status.
c.
Summary description of action (s) taken to prevent a recurrence.
154
TABLE 4.8.1 RADIOACTIVE LIQUID
,,,ati.
.E PLING AND ANALYSIS PROGRAM Lower Limit M inimutn of Detection Sampling Analysis Type of activity (LLD)
Liquid Release Type Frequency Frequency Analysis (uCi/ml)a Fatch Waste Release Tanks P
P Principal Gamma 5 x 10-7b d
Each Batch Each Batch Emi t t e rs" I-131 1x 10-6 b
Dissolved and 1 x 10-5 P
M One Batch /M Entrained Gases P
M H-3 1 x 10-5 Each Batch l
Composite C Gross alpha 1 x 10-7 P-32 1 x 10-6 P
Q Sr-89, Sr-90 5 x 10-8 l
Composite C Each Batch Fe-55 1 x 10-6 Frequency Notation:
M - once per month Q - once per quarter P - completed prior to release 155
TABLE 4.8.1 TABLE NOTATION a.
The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability or with only 5% probability of falsely concluding that a blank obs e rva t ion represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
b LLD =
2.22 Y - exp(-AAt)
E V
where LLD is the lower limit of detection as defined above (as pCi per unit mass or voluce) sb is the standard devis _ ion of the background counting rate or of the counting rate of a blank sample as a,pr op riate (as counts per minute)
E is the counting e.ficiency (as counts per transformation)
V is the sample s ize (in units of mass or volune) 2.22 is the number of transformation per minute per picocurie Y is the f rac* ional radiochemical yield (when applicable)
A is the radioactive decay constant for the particular radionuclide at is the elapsed time between sample collection (or end of the sampic collection period) and time of counting The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples.
156
TABLE 4.8.1 (Continued)
TABLE NOTATION b.
For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD.
Under these circumstances, the LLD may be increased inversely proportionally to the magnitude of the garra yield (i.e., 5 x 10-7/I, where I is the photon abundance expressed as a decimal f raction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10CFR20, Appendix B, Table II, Colurn 1.
c.
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.
d.
A batch release is the discharge of liquid wastes of a discrete volume.
e.
The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, te-141, and Ce-144.
This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circu ns tances result in LLD's higher than required, the reasons shall be documented in the semiannual Radioactive Effluent Release Report.
157
TABLE 4.8.2 RJ_IOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Sampling Analysis Type of Activity (LLD)a Gaseous Release Type Frequency Frequency Analysis (uci/cc) b A. Main Plant Stack M
M Principal Gamma 1 x 10-4 d
Grab Sample Emitters" H-3 1 x 10-6 Continuous p
I-131 1 x 10-12 d
c I-133 1 x 10-10 d
c Continuous p
b Particulate Principal Gamna 1 x 10_11 Sample Emitters (I-131, Others) d Continuous g
Particulate Gross alpha 1 x 10-11 Sample Continuous Q
Sr-89, Sr-90 1 x 10-11 d
Composite Parti-culate Sample b
B. Containment Purge P
P Principal Gamma 1 x 10-4 Each Purge Each Purge Emitters" Grab Sample 158
TABLE 4.8.2 TABLE NOTATION FREQUENCY NOTATION NOTATION FREQUENCY D.......... Once per day W.......... Once per week M.......... Once per month
~
Q.......... Once per quarter P.......... Completed prior to release a.
The LLD is the smallesc concentration of radioactive material in a sample that will be detected with 95% probability or with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
b LLD =
2.22 Y - exp(-AAt)
E V
where LLD is the lower limit of detection as defined above (as pCi per unit mass or volume) is the standard deviation of the background counting rate or of the counting rate sf a blank sample as appropriate (as counts per minute) o E is the counting efficiency (as counts per transformation)
V is the sample size (in units of mass or volune) 2.22 is the number of transformation per minute per picocurie Y is the fractional radiochemical yield (when applicable) 159
TABLE 4.8. 2 (Cont inued)
TABLE NOTATION A is the rad ioact ive decay constant for the particular radionuclide At is the elapsed time between sample collection and analysis The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a rad ionu c lide de termined by gamma-ray spect romet ry, the background shall include the typical contributions of other radionuclides normally present in the samples.
b.
For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD.
Under these circumstances, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (i.e., 1 x 10-4/I, where I is the photon abundance expressed as a decimal f raction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10CFR20, Appendix B, Table II, Colurn 2.
Analyses s.all also be perf ormed at least once per day when a weekly analysis indicates c.
a release rate which would project a dose rate greater than 50% of that specified in 3.8.F.1.a.
Analysis will revert back to a weekly f requency if ar.d when two consecutive daily analyses show a release rate below 50% of 3.8.F.1.a.
When samples collected for 1 day are analyzed, the corresponding LLD's may be increased by a factor of 10.
d.
The ratio of the sample flow rate to the sampled stream flow rate shall be known.
The principal gamma emitters for which the LLD specification will apply are exclusively e.
the following radionuclides: KR-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, co-58, co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nu c lide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semiannual Radioactive Effluent Release Report.
160
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BASES: 3.8/4.8 Radioactive Effluents A. Liquid Ef fluents: Concentration This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will he less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the International Commission on Radiological Protection (ICRP) Publication 2. B. Linuid Effluents: Dose This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably a ch ievab le". The Surveillance Requirements implement the requirements in Section III.A of Appendix I, i.e., that conf ormance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be subs tantially underestimated. In addition, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in finished drinking water that are in excess of the requirements of 40 CFR 141. No drinking water supplies drawn from the Connecticut River holow the plant have been identified. The appropriate dose equations for implementation through requirements of the Specification are described in the Vermont Yankee Of fsite Dose Calculation Manual. C. Liquid Waste Treatment The maintenance and use of the Liquid Radwaste Deep Bed Demineralizer to perform its design function ensures that this system will be available whenever liquid ef fluents require treatment prior to release to the environment. The requirement that the Deep Bed Demineralizer be used routinely provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". 160b.
D. Gaseous Ef fluents : Dose Rate This specification is provided to ensure that the dose at the restricted area boundary f rom gaseous ef fluents will be within the annual dose limits of 10 CFR Part 20. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits provide reasonable assurance that rad i oa c t ive material discharged in gaseous ef fluents will not result in the exposure of an individual to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be in the restricted area, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the nearest unrestricted area. The specified release rate limits as determined by the procedures in the ODCM, restrict, at all times, the corresponding gamma and beta dose rates above background to an individu il at or beyond the restricted area bounda ry to $500 mrem / year to the total body or to 53000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the grass milk-infant pathway to $1500 mrem / year for the worst case milk animal. E. Gaseous Effluents: Dose from Moble Gases This specification is provided to implement the requirements of Sections IIB, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements tne guides set forth in Section II.B of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous ef fluents will be kept "as low as is reasonably a ch ievab le". The Surveillance Requirements implement the requirements in Section III.A of Appendix I, i.e., that conf ormance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The appropriate dose equations are specified in the ODCM for calculation the doses due to the actual releases of radioactive noble gases in gaseous ef fluents. The ODCM also provides for determining the air doses at the unrestricted area boundary based upon the historical average atmospheric conditions. F. Gasemis Ff fluents: Dase From Radiciodines, Radioactive Material in Particulate Form and Radionuclides Other than Noble Cases This specification is provided to implement the req uirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation are the guides set forth in Section II.C of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous ef fluents will be kept "as low as is reasonably a ch ievab le". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conf ormance with the guides of Appendix I be shown by calculation procedures 160c.
based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The appropriate dose equations are specified in the ODCM for calculating the dose due to the actual releases of the subject materials. These equations also provide for determining the actual doses based upon historical ave raga atmospheric conditions. The release rate specifications for radiciodines, radioactive materi al in particulate form and radionuclides other than noble gases are dependent on the existiu; radionuclide pathways to man, in the unrestricted area. The pathways wh!ch were examined in the development of these specifications were: 1) individual inhalation of airborne radi onu c lide s, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of ma n. G. Dose Commitment Specification 3.8.G.1 is provided to meet the reporting requirements of 40 CFR 190. It is assumed, based on distance, that the sum of all contributions from sources otner than from the immediate plant site will be small compared to this standard and can be ignored. H. Explosive Gas Mixture The hydrogen monitors are used to detect possible hydrogen buildups which could result in a possible hydrogen explosion. Isolation of the off-gas flow would prevent the hydrogen explosion and possible damage to the augmented off-gas system. I. Steam Jet Air Ejector (SJAE) The air ejector of f-gas monitor provides isolation capability on the air ejector suction line. Isolation is initiated when the instrument reaches its upscale trip point. The immediate trip (within 1 minute) set point of 1.5 Ci/sec (30 minute decay) is based upon limitirg the whole body dose at the site boundary to less than 5 Rem in the unlikely event of a boundary failure in the off-gas system concurrent with a spike release of radioactivity from the fuel. The assumption has been made that the rate of radioactivity increase within the 1 minute valve closure time period would be less than a factor of 5 based upon actual experience uith such events. The delayed trip (within 15 minutes) set point of 0.3 Ci/sec (30 minute decay) is based upon limiting the whole body concurrent with an off-gas release from the fuel of a lower value than considered above. 160d.
J. Primary Containment This specification provides reasonable assurance that releases from containment purging operations will not exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas. The effects of releases through the torus pumpback vent were evaluated in the Safety Evaluation Report supporting Amendment No. 50 to DPR-28. As such, releases from the pumpback vent are not considered part of the LCO described in Specification 3.8.J.1. K. Gaseous Waste Treatment The requirement that the AOG be used routinely during reactor power operation provides reasonable assurance that the releases of radioactive materials in the off-gas will be kept "as low as is reasonably achievable." The action level given in Specification 3.8.K.2 and 3.8.K.3 (i.e. 50% of the limits of Specification 3.8.E.1.a, or 3.8.F.1.a) was chosen as a suitable fraction of the quarterly dose limit for gaseous effluents given in 10CFR50, Appendix I, and is equivalent to the annual dose objectives of Appendix I, for gases. 160e.
VYNPS 3.9 LIMITING CONDITIONS FOP. OPERATION 4.9 SURVEILLANCE REQUI' JfENTS 3.9 RADIOACTIVE EFFLUENT MON 1TORING SYSTEMS 4.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS Applicability Applicability I Applies to the monitoring systems or programs Applies to the required surveillance of the monitoring which perform a surveillance, protective systems or programs which perform a surveillance, or controlling function on the release of radio-protective or controlling function on the release of active effluents from the plant and their radioactive effluents from the plant and their iden-identification in the environment. tification in the environment. Objective Objective To assure the operability of the radioactive effluent To specify the type and frequency of surveillance to be monitoring systems and environmental programs. applied to the radioactive effluent monitoring system and environmental programs. Specifications Specifications A. Liquid Effluent Instrumentation A. Liquid Effluent Instrumentation 1. During periods of release through the monitored 1. Each radioactive liquid effluent monitoring pathway, the radioactive liquid effluent instrumentation channel shall be tested and monitoring instrumentation channels shall be calibrated as it;dicated in Table 4.9.1. operable in accordance with Table 3.9.1 with their alarm / trip setpoints set to ensure that the limits of Specification 3.8.A.1 are not exceeded. 2. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that suspendthereleasel the limits of 3.8.A.1 are met, of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable. 161
VYNPS 3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIREMENTS B. Gaseous Effluent Instrumentation B. Gaseous Effluent Instrumentation ~ 1. During periods of release through the 1. Each gaseous process or effluent monitoring monitored pathway, the gaseous process and instrumentation channel shall be tested and effluent monitoring instrumentation calibrated as indicated in Table 4.9.2. channels shall be operable in accordance with Table 3.9.2 with their alarm / trip setpoints set to ensure that the limits of Specification 3.8.D.1 and 3.8.H.1 are not exceeded. 2. With a gaseous process of effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.8.D.1 and 3.8.H.1 are met, delare the channel inoperable. C. Radiological Environmental Monitoring Program C. Radiological Environmental Monitoring Program 1. The radiological environmental monitoring 1. The radiological environmental monitoring program shall be conducted as specified samples shall be collected and analyzed in Table 3.9.3; or pursuant to the requirements of Table 3.9.3. The results of these analyses or devi tions from 2. With the Radiological Environmental Monitoring the program shall be summarized in the Annual Program not being conducted as specified in Jadiological Environmental Operating Report. Table 3.9.3, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. 162
VYNPS 3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIREMENTS 6 ~ 3. With the level of radioactive in an e.nvironmental sampling medium at any sampling location specified in Table 3.9.3 exceeding the limits of Table 3.9.4 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the receipt of the Laboratory analyses, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 3.9.4 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. 4. Any change or permanent modification of any of the sample collection locations which have previously been sampled as part of the routine environmental radiological monitoring program, shall be identified and included in the Annual Radiological Environmental Operating Report. D. Land Use Census D. Land Use Census 1. A land use census shall be conducted to identify 1. The land use census shall be conducted at least the location of the nearest cow and goat, the once per year between the dates of June 1 and nearest residence and the nearest garden
- of October 1 by either a door-to-door survey, aerial greater than 500 square feet producing fresh leafy survey, or by consulting local agriculture vegetables in each of the 16 meteorological authorities.
sectors within a distance of five miles. The land use census shall also identify the locations of 2. The results of the land use census shall be all milk cow and goats and all gardens
- of greater included in the Annual Radiological Environmental Operating Report.
- Broad leaf vegetation sampling may be performed at the point with the highest predicted D/Q in lieu of the garden census.
163
VYNPS 3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIREMENTS than 500 square feet producing fresh leafy ~ vegetables in each of the 16 meteorological sectors within a distance of three miles. 2. With a land use census identifying a location (s) which yields a calculated dose or dose commitment at least 15 percent greater than the values currently being calculated in Specifica-tion 4.8.F.1 prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.C.3, a Special Report which identifies the new location (s). 3. With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) at least lf percent greater than at a location f rom which samples are currently being obtained in accordance with Specification 3.9.C.1 prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.C.3 a Special Report which identifies the new location. If permission from the owner to collect samples can be obtained and suf ficient sample volume is available, then this new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. E. Intercomparison Program E. Intercomparison Program 1. As part of the Radiological Environmental 1. The results of analyses required by Specification Monitoring Program, analyses shall be performed 3.9.E.1 shall be included in the Annual Radiological on referenced radioactive materials supplied as Environmental Operating Report. part of a Intercomparison Program which has been approved by NRC; or 164
VYNPS 3.9 LIMITING CONDITIONS FOR OPERATION 4.9 SURVEILLANCE REQUIREMENTS 2. With analyses not being performed as required in Specification 3.9.E.1, report the corrective actions taken to prevent recurrence to the Commission in the Annual Radiological Environmental Operating Report. l 165
TABLE 3.9.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Mininum Channels In s t ru men t Operable gplicability Action 1. Liquid Radwaste 1 1 Discharge Monitor 2. Service Water Discharge 1 2 Monitor 3. Liquid Radwaste Discharge 1 3 Flow Rate Monitor
- During releases via this pathway.
166
~ TABLE 3.9.1 TABLE NOTATION Action 1 With the number of channels operable less than required by the Minimum Channels Operable requirement, effluent releases may be continued for up to 14 days, provided that prior to initiating a release: 1. At least two independent samples are analyzed in accordance with Specification 4.8.A.1 and; 2. At least bao technically qualified individuals independently verify the release rate calculations and discharge valving; O the rwis e, suspend release of radioactive ef fluents via this pathway. Action 2 With the numbers of channels operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue foi up to 14 days provided that d111y grab Samples are collected and analyzed for gross radioact ivity (beta or gamma). Action 3 With the number of channels operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue for up to 14 days provided the flow rate is estimated at least once per 4 hours during actual releases. 167
TABLE 3.9.2 GASEOUS EFFLUENT MONITORING INSIRUMENTATION ~ Minimum Channels Instrument Operable Applicability Action 1. Advanced Off-gas System a. Noble Gas Activity Monitor 1 6 between the charcoal bed system anti the plant stack b. Flow Rate Monitor 1 5 c. Hydrogen Monitor 1 7 2. Steam Jet Air Ejector (SJAR) a. Noble Gas Activity Monitor 1 4 3. Plant Stack a. Noble Gas Activity Monitor 1 6 b. Iodine Sampler Cartridge 1 8 c. Particulate Sampler Filter 1 8 d. Sampler Flow Integrator 1 5 e. Effluent System Flow Rate 1 5 Monitor
- During releases via this pathway.
168
TABLE 3.9.2 TABLE NOTATION Action 4 With the number of channels operable less than required by the Minimum Channels Operable requirement, gases from the off-gas system may be released to the environment for up to 48 hcurs provided: 1. The A0G system is not bypassed; and 2. The A0G system noble gas activity monitor is operable; Otherwise, be in at least hot standby within 12 hours. Action 5 With the number of channels operable less than required by the Minimum Channels Operable requirement, effluent releases via this nathway may continue f or up to 28 days provided the flow rate is et.imated at leno-once per 4 hours. Action 5 With the number of channels operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue for up to 28 days provided grab samples are taken at least once per 8 hours and these samples are analyzed f or gross act ivity within 24 hours. Action 7 With the number of channels operable less than required by the Minimum Channels Operable requirement, operation of the A0G system may continue for up to 28 days prov ded gas samples are collected at least once per 4 hours and analyzed within the ensuing 4 hours, or an orderly transfer of the off-gas effluents from the operating recombiner to the standby recombiner shall be made. Action 8 With the number of channels operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue for up to 28 days, provided samples are continuously collected with auxiliary sampling equipment for periods on the order of seven (7) days and analyzed within 48 hours after the end of the sampling period. 169
TABLE 3.9.3 RADIOLOGICAL ENVIPONMENTAL MONITORING PROGRAM (1)(3) Exposure Pathway Number of Sampling and TypeandFregtgcy and/or Sanple Sample Locat ions Collection Frequency of Analysis 1. AIRBORNE a. Radiciodine and 5 Continuous operation of sampler with Radiciodine canister. Analyze Pa rt icula t es sample collection as required by once per week for I-131. dust loading but at least once per week. Particulate sampler. Analyze for gross beta radioactivity > 24 hours following filter change. Perform ga mma isotopic analysis on each sampic wheg gross beta activity is >0.5 pCi/m Perf orm gamma isotopic analysis on composite (by location) sample once per quarter. 2. DIRECT RADIATION 8 Once per month. Gamma dose, once per month. 3. WATERBO RNE a. Surface 2 Once per month. Gross beta and I-131 and gamma isotopic analysis of each sample. Tritium analysis of composite sample once per quarter. b. Ground 2 Gnce per quarter. Gamma isotopic triti'm analyses of each sample. c. Sediment from 1 Once per 6 months. Camna isotopic analysis of each Shoreline sample. 170
TABLE 3.9.3 (Continued) PfDIOLOGICAL ENVIRONMENTAL MONITORING PROCPAM(1)(3) Exposure Pathway Number of Sampling and Type and Freque c and/or Samph Sample Locations Collection Frequency of Analalys".s 4. INGESTION a. Milk 4 Once per 2 weeks between Gamna isotepic ar.d I-131 analysis June 1 and October 1 if of each sample. milk animals are on pasture: Once per month at other times. b. Fish and 2 Once per 6 months. One Gamma isotopic analysis on edible Invertebrates sample containing at least two portions. different species; from the following list:
- 1. wh it e pe rch
- 5. large mouth
- 2. yellow pe rch bass
- 3. pickerel
- 6. pumpki m eed
- 4. small mouth bass
- 7. walleye
- 8. American eel c.
Vegetation and 3 At t ime of ha rve s t. One sample of Gamma isotopic analysis on edible Food Products any of the following classes of portion. food products:
- 1. grain
- 2. above-ground vegetable
- 3. tuberous vegetable 1
At time of harvcs t. One sample of any I-131 analysis. broad leaf vegetation. 171
TABLE 3.9.3 NOTATION (1) Specific sample locations for all media are specified in the Offsite Dose Calculation Manual and reported in the Annual Radiological Environmental Operating Report. (2) See Table 4.9.3 for maximum values for the Lower Limits of detection. (3) Deviations are permitted from the required sampling schedule if specinen are unobtainable due to hazardous conditions, seasonal unavailability or to malfunction of sampling equipment. If the latter, every ef fort shall be made to complete corrective action prior to the end of the next sampling period. 172
TABLE 3.9.4 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate l Fish Milk Vegetables Analysis (pCi/1) or Gases (pCi/m3) I (pci/Kg, wet) (pci/1) (pci /Kg, wet ) i l I i l 1 4 1 H-3 3x 10 t i 3 4 Mn-54 1 x 10 3x 10 Fe-59 4 x 10 l 1 x 10 2 4 i 3 4 l Co-58 1 x 10 3 x 10 i 2 4 i co-60 3x 10 1 x 10 1 2 4 Zn-65 3 x 10 2 x 10 2 Zr-95 4 x 10 g 2 I-131 2 0.9 3 1 x 10 v { l 3 3 10 1 x 10 60 1 x 10 Cs-134 l 30 i 3 3 Cs-137 . 50 20 2x 10 70 2 x 10 2 2 2 x 10 3 x 10 Ba-140 i. h 172a.
_ TABLE 4.9.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Ins t ru me nt Source Ins t ru me n t Functional Instrunent Check Check Calibration Test 1. Liquid Radwaste Discharge D* P R(1) Q Monitor 2. Service Water Discharge Monitor D* M R(1) Q 3. Liquid Radwaste Discharge Flow Rate Monitor D* NA NA 0*
- During releases via this pathway.
D... Once per day M... Once per month Q... Once per quarter R... Once per 18 months P Prior to each release NA... Not applicable il) The instrument calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Bureau of Standards) radiation measurement system liquid radioactive source positioned in a reproducible geometry with respect to the sensor. These standards shall pennit calibrating the system over its intended range of energy and rate capabilities. 172b.
TABLE 4.9.2 CASFOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Ins t ru ment Ins t rurie n t Source Ins trumen t Funetional Instrucent Check Check Calibration Test 1. Advanced Offgas System a. Noble Gas Activity D* M R(1) Q Monitor b. Flow Rate D* NA NA NA Monitor c. Hydrogen Monitor D* NA Q M 2. Steam Jet Air Ejector System a. Noble Gas Activity Monitor D* M R(1) 0 3 Plant Stack a. Poble Gas Activity Monitor D* M R(1) 0 b. Sampler Flow Integrator W* NA R NA c. System Flow Rate D* NA NA NA Monitor
- During releases via this pathway 172c.
TABLE 4.9.2 TABLE NOTATION D... Once per day W... Once per week M... Once per month Q... Once per quarter R... Once per 18 months NA... Not Applicable (1) The Instrument Calibration for radioact.vity measurement instrumentation shall include the use of a known (traceable to Nationa Bureau of Standards) Radiation Measurement System radioactive source positioned in
- eproducible geometry with respect to the sensor. These standards should permit 'allorating the system over its intended range of rate capabilities.
172d.
TABLE 4.9.3 RADIOLOGICA ENVIPONMENTAL MONITORING PROGRAM MAXIMint VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)" I Airborne Particulate Analysis Water or Gas Fish Milk Food Products Sediment (pC i/1) (pCi/m3) (pc i /k g, wet) (pCi /1) (pCi/kg, wet) (pC i /k g, dry) b gross beta 4 1 x 10-2 3H 2000 54Mn 15 130 59 30 260 Fe 58,60 15 130 Co 65 30 260 Zn 95 15 Zr b 131 l 7x 10-2 7 1 60C l 134,137 15, 18 1x 10-2 130 15 80 150 Cs d 140 15 d Ba 15 172e.
TABLE 4.9.3 TABLE NOTATION a. The LLD is the smallest concentration of rad ioa c t ive material in a sample that will be detected with 95% probability with anly 5% probability of falsely concluding that a blan; observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): 4.66s b E-V - 2.22 Y exp(-AAt) where LLD is the lower limit of detection as defined above (as pCi per unit mass or volume ) sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute) E is the counting efficiency (as count s per transformat ion) V is the sample size (in units of mass or volume) 2.22 is the nutber of transf ormat ion per minu te per picocurie Y is the f ractional radiochemical yield (when applicable) A is the radioactive decay constant for the particular radionuclide At is the elapsed time between sanple collection (or end of the sample collection period) and time of counting The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the s-iles (e.g., potassium-40 and milk samples). 172f.
TABLE 4.9.3 (Continued) TABLE NOTATION Analyses shall be pe rf ormed in such a manner that the stated LLDs will be achieved under routine conditions. Occasional background fluctuations, unavoidably small samp)e sizes, the presence of interf ering nuclides, or other uncontrollable circumstances may render these I T nr unachievable. In such cases, the contributing factors will be identified and descr oed in the Annual Radiological Environnental Operating Report. b. LLD for surf ace water. c. LLD for leafy vegetables. d. The Ba-140 LLD and concentration can be de t e rmined by the analysis of its short-lived daughter product La-140 subsequent to au 8 day period following collection. The calculation shall be predicted on the normal ingrowth equat ions for a parent-daughter situation and the assumption that any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6 percent of its original value). The ingrowth equations will assure that the supported La-140 activity at the time of collection is zero. e. If the measured concentration minus the 5 sigma counting statistics is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD. 172g.
Bases: o 3.9/4.9 Radioactive Ef fluent Monitoring Systems A. Liquid Effluent In s t ru me nt a t ion The radioactive liquid effluent ins t ru me n t a t ion is provided to monitor the releases of radioactive man rials in liquid ef fluents during actual or potential releases of liquid e'Eluents. These instruments are provided to ensure that the limits of 10 CFR Part 20 will n. be exceeded. B. Gaseous Ef fluent Ins t ru ment at ion The radioactive gaseous ef fluent ins t ru ne n t a t ion is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous ef fluente during actual or potential releases of gaseous ef fluents. These instruments are provided to ensure that the limits of 10 CFR Part 20 will not be exceeded ~ This ins t rumentation also includes provisions f or monitoring and controlling the concentrationc of potentially explosive gas mixturea in the offgas, downs tream of the recombiners. C. Radiological Environmental Monitoring Program The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those rad i onu c lid e s, which lead to the highest potential radiation exposures of individca.ls resulting from the station operation. this monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive ma t e rials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least three yea rs. Following this period, program changes may be initiated based on operational experience. The detection capabilities required by Table 4.9.3 are state-of-the-art for routine environmental measurements in industrial laboratories. 172h.
D. Land Use Census Specification 3.9.D (Land Use Census) is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the sults of this census. This census satisfies the requirements of section IV.B.3 of Apperdix to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet.rovides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assured in Regulatory Guide 1.109 f or consumpt ion by a child. To de termine this minimun garden size, the following assumptions were used, 1) that 20% of the garden was used f or growing broad leaf vegetation (i.e., similar to lettuce and cabbage ), and 2) a vegetation yield of 2 kg/ square meter. In lieu of the garden census, broad leaf vegetation samples f rom the direction sector with the highest D/Q may be substituted. The use of the maximum of f site D/Q value predicted f or gaseous ef fluents from the plant stack (plant stack qualifies for an elevated release as defined in Regulatory Guide 1.111, March 1976) will generate the maximum possible calculated dose and thus no real garden located at any other point could have a greater calculated dose or dose commitment. The addition of new sampling locations to Specification 3.9.C.1 based on the land use census is limited to those locations which yield a calculated dose or dose commitment at least fifteen percent greater than the calculated dose or dose commitment at any location currently being sampled. This eliminates the unnecessary changing of the environmental radiation monitoring program for new locations, which, within the accuracy of the calculation, contributes essentially the same to the dose or dose commitment as the location already scmpled. The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than fifteen percent would not be expected to result in a significant increase in the ability to detect plant effluent related nuclides over the locations already being sampled. E. Intercomparison Program The requirement for an Intercomparison program of Specification 3.9.E is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are perf ormed as part of a quality assurance progran f or environmental monitoring in order t' demonstrate that the results are reasonably valid. 1721.
VYNPS 6.5 PLA'iT OPERATING PROCEDURES A. Detailed writ ,p.ocedures involving nuclear safety, including applicable check-off lists and instructions, covering areas _isted below shall be prepared and approved. All procedures shall be adhered to. 1. Normal startup, operation and shutdown of systems and comnonents of the f acility. 2. Refueling operations. 3. Acti',ns to be taken to correct specific and foreseen potential malfunctions of systems or components, suspected primary system leaks and abnormal reactivity changes. 4. Emergency conditions involving potential or actual release of radioactivity. on the safety of the reactor. 5. Preventive and corrective maintenance operations which could have an affect 6. Surveillance and testing requiremente. 7. Fire protection program implementation including minimum fire brigade requirements and training. The training program shall meet or exceed the requirements of Section 27 of the NFPA Code 1976. Training sessians will be scheduled as plant operations permit but will be completed in specified subjects annually. Initial fire brigade training shall be complete by March 13, 1978. 8. The Radiological Environmental Monitoring Program. 9. Lf f site Dose Calculation Manual implerentation. 10. ".ffluent monitoring instrumentation setpoints. B. Radiation control standarde and nrocedures shall be prepared, approved and maintained and made available to all station personnel. TF tse procedures shall show permissible radiation exposure, and si.cl1 be consistent with the requirements of ~d CFR Part 20. This radiation protection program shall be organized to meet the requirements of 10 CFR Part 20. 1. Paragraph 20.203 " Caution signs, labels, signals, and controls." In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *. Any individual of group of individuals permitted to enter such areas shall be provided with one or more of the following: 200
VYNPS a. A radiation monitoring device which continuously indicates the radiation dose rate in the area. b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them. c. A health physics qualified individual (i.e. qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and who will perform periodic radiation surveillance at the frequency specified in the RWP. The surveillance frequency will be established by the Plant llealth Physicist. The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Plant llealth Physicist.
- Health Physics personuc1 shall be exempt from the RWP issuance requirement during the performancc
,f their assigned radiation protection duties, providing they are following plant radiation protection procedures for entry into high radiation areas. 201
3. Monthly Operating Report Rou t ine reports of operating statistics and shutdown experience shall be submitted on a monthly bacis to the Office of Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Of fice, to arrive no later than the fifteenth of each month following the calendar month covered by the report. These reports shall include a narrative summary of operating experience during the report period which describes the operation of the f acility and any major safety-related maintenance. 4. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The Annual Radiological Environmental Operating Reports shall include summa r i es, int e rp r e t a t ions, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with operat ional controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operat ion on the environment. If ha rmf ul ef f ects or evidence of irreve rsible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the cause. The Annual Radio _ogical Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The Annual Radiological Environmental Operating report shall also include the following: a summa ry descript ion of the radiological environmental monitoring program including a map of all sampling locations keyed to a table giving distances and directions from the reactor; the result of land use censuses required by the Specification 3.9.D 1 and the results of the Intercomparison Program required by Specification 3.9.E.1. B. Reportable Occurrences Reportable occurrences, including correct ive act ions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event rep o rt shall be completed, and reference shall be made to the original report date. Events involving systems or components described in Section 3.13 and 4.13 do not require reporting udder the provisions of this section. Such events will be reported as required in Section 6.7.C. 2 below. 210
o 2. Thirty Dav Written Report s e The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Of fice within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. Reactor protection system or engineered safety feature i ns t ru me n t settings which are a. foun to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the f unct ional requirements of af f ected systems.
- b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition f o r op e ra t ion.
Note-Routine surve illa nce testing, i ns t ru me n t calibratinn, or preventative maintenance which require system configuratiens as described in items 2.a and 2.b. need not be reported except where test results themselves reveal a degraded mode as described above.
- c. Observed inadequacies in the inplementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor pr otect ion systems or engineered safety feature systems.
- d. Abnormal degradation of systems other than those specified 'n item 2.c. above designed ta contain radioact ive material result ing from the fissio process.
Note: Scaled sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the li mi t s for ident i f ic4' leakage set forth in technical specifications need not be reported under this item.
- e. An uncontrolled offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous ef fluents, or 3) more than 0.05 curies of radiciodine in ga s e ou s effluents. The report of an uncontrolled offsite release of radioactive material shall include the following inf ormat ion:
213
1. A description of the event and equipment involved. 2. Cau se (s ) for the uncontrolled release. 3. Actions taken to prevent recurrence. 4. Consequences of the uncontrolled release.
- f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.9.4 when averaged over any calendar qua rter sampling period.
Wh e n mo re than one of the radionuclides in Table 3.4.4 are detected in the sampling nedium, this report shall be submitted if: con cent ra t i on (1) + concentration (?) + limit level (1) limit level (2) -- 1.0 When radionuclides other than those in Table 3.9.4 are detected and are the results of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.8.B.1,
- 3. 8. E.1 a nd 3. 8. F.1.
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. C. Unique Reporting Req u i reme nt s 1. Semiannual Radioactive Effluent Release Report Pa ragraph (a)(2) of subsection 50.36a, " Technical speci f ica t ions On Effluents From Nuclear Power Reactors," of 10 CFR Part 50 requires that a report be made to the Commission within 60 days af ter January 1 and July 1 of each year which specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquids and gaseous ef fluents during the previous 6 months of operation. This inf ormation submitted in the semi-annual ef fluent release report shall be in accordance with Regulatory Guide 1.21 (Revision 1) dat ed June 1974 and Regulatory Guide 4.1 da t ed Ja nu a ry 18, 1973 213a.
2. L'h e re required by Sect ion 3.13, special reports shall be submitted to the Commission following the discovery of certain inop e rab le sensors, ins t ru me n t s, conponents or systems in the vital fire protection system. Note: Routine surveillance test ing or design modif icat ion of sensors, ins t ru ne n t s, components or systems which lead to operation of sensors, ins t ru me nt s, components or sy s t e.as in a degraded mode do not req u i re special reporting except where tests themselve s reveal a degraded mode. 3. Special reports shall be submitted to the Director of the office of Inspection and Enf o rce .t Pegional Office within the time period specified f or each report. These reports .all be submitted covering the activities identified below pursuant to the req u i reme n t s of the applicable reference specification. a) Liquid Effluents: Dose; Specification 3.R.R.l. b) Gaseou s Ef fluents : Dose from Noble Cases; Specification 3.8.v.1. c) Caseous Ef fluents : Dose from Radioiodines, Radioactive varerial in Particulate Form and Radionuclides Other than Noble Cases; Specification 3.8.F.1. d) Dose Commitment Specification 3.8.C.I. 6.8 FIRE PROTFCTION INSPECTION A. An independent fire protection and loss prevention inspection and audit shall be perf ormed annually ut ilizing either qualified of f-site licensee personnel or an ou t s id e fire prot ect ion f i rm. B. An inspect ion and audit by an out s ide qualif ied f ire consultant shall be perforned at int e rva ls no greater than 3 yea rs. 214
9 y u s'" a '$ w g "~ MONT YANKEE NUCLEAR POWER CORPORATION i 0FFSITE DOSE CALCULATION MANifAL
4 f~'I b\\[ ua d TABLE OF CONTENTS Page Section 1
1.0 INTRODUCTION
3 2.0 LIQUID RELEASE DOSE CALCULATIONS 2.1 Technical Specification 3.8 B, Dose to an Individual 3 4 2.1.1 Method I 2.1.1.1 Dose to the Total Body 4 2.1.1.2 Dose to the Critical Organ 4 4 2.1.1.3 Application of Method I 5 2.1.2 Method II 2.1.2.1 Dose to the Total Body 5 2.1.2.2 Dose to the Critical Organ 5 2.1.2.3 Application of Method II 6 6 2.1.3 Method III 7 3.0 GASEOUS RELEASE DOSE CALCULATIONS 3.1 Technical Specification 3.8 D, Dose Rate Limits 7 3.1.1 Method I 7 3.1.1.1 Dose Rate Due to Noble Gases 7 3.1.1.1.1 Dose Rate to the Total Body 7 3.1.1.1.2 Dose Rate to the Skin 8 3.1.1.2 Dose Rate to the Critical Organ Due to 8 Radioiodines and Particulates 3 1.1.3 Application of Method I 9 3.1.2 Method II 9 11 3.2 Technical Specifications 3.8 E, Dose to Air Due to Noble Gases
I [7;s I' ~ 1 l Obs'tt... ua a TABLE OF CONTENTS (continued) Section Page 3.2.1 Method I 11 3.2.1.1 Air Dose Due to Gamma Radiation 11 3.2.1.2 Air Dose Due to Beta Radiation 11 3 2.1.3 Application of Method I 12 3.2.2 Method II 12 33 Technical Specification 3.8 F, Dose to an Individual 14 14 3.3.1 Method I 3.3.1.1 Dose to the Thyroid 14 3.3.1.2 Application of Method I 15 3.3.2 Method II 15 3.3.2.1 Dose to the Thyroid 15 3.3.2.2 Application of Method II 16 3.3.3 Method III 16 17 4.0 METEOROLOGY 5.0 ENVIRONMENTAL MONITORING 20 APPENDIX I - Basis for the Dose Calculation Methods APPENDIX II - Basis for the Atmospheric Dilution Factot4 25 30 REFERENCES
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1.0 INTRODUCTION
The purpose of this manual is to provide methods to insure compliance with the dose requirements of the Technica' Specifications. Each method is based on a plant specific application of the models presented in Regulatory Guide 1.109.(1) Methods are included to calculate the doses to individuals from both gaseous and liquid releases from the plant. Under normal operations, experience has shown that the plant will be operated at a small fraction of the dose limits imposed by the Technical Specifications. For this reason the dose evaluations are presented at dif f erent levels of sophistication. The first' method being the most conservative, but simplest to use, subsequent methods somewhat more entailed and more realistic while the final method requires a full analysis following the guidance presented in Regulatory Guide 1.109. The first is based on a critical organ, critical age group and as such it provide a conservative estimate of the doses required by the Technical Specifications. If the Technical Specifications are met by application of the first method, no further analysis will be required. If, however, it indicates that the Technical Specification limits are being approached a more realistic estimate may be obtained by application of subsequent methods. The final method will calculate the dose to seven organs of four 5 Ib'h j,iL//wN d age groups and is based on measured releases for each nuclide. This method will be used to assess doses for the Semi-annual Radioactive Ef fluent Release Report. The basis for each of the dose calculation methods is described in Appendix I. t Ns_.]l.> 2.0 LIQUID RELEASE DOSE CALCULATIONS 2.1 Technical Specification 3.8 B. Dose to an Individual This section is to be used to irsure compliance with the following Technical Specification: LIMITING CONDITION FOR OPERATION The dose commitment to an individual from radioactive materials in liquid ef fluents released to unrestricted areas shall be limited: During any calendar quarter to 11.5 mrem to the total body a. and to 55 mrem to any organ, and b. During any calendar year to 53 mrem to the total body and to $10 mrem to any organ. The dose commitment to any individual f rom liquid releases is proportional to the quantity (curies) to which that individual is exposed. T5e following equations shall be used to calculate the dose commitment resulting from a liquid release (in curies) from the Vermont Yankee Station. The specification requires a monthly evaluation, however the following equations can be applied for any duration of release. i f= I t' .i, L~ sa ma 2.1.1 Method I 2.1.1.1 Dose to the Total Body The dose to the total body is: Dtb(mrem) = 1.3 Q60Co + 0.001 Q131I + 8.0 Q137Cs where: Q60Co - Cobalt-60 Release (C1) Q131I = I dine-131 Release (C1) Q137Cs= Cesium-137 Release (C1) 2.1 1.2 Dose to the Critical Organ The dose to the critical organ is: Dorgan(mrem) = 6.0 Q60Co + 1.0 Q131I + 10. Q137Cs where: - Cobalt-60 Release (C1) Q60Co = I dine-131 Release (C1) Q131 Q137Cs - Cesium-137 Release (C1) 2.1.1.3 Application of Method I %,. t.D {\\ " m' o Step 1. Determine the number of curies of Cesium 137, Iodine 131 and Cobalt-60 released during the period. Step 2. Perform the above multiplications to )btain the total body and organ doses for the period. Step 3. Record the total body and organ doses and maintain a cumulative dose for the annual and quarterly periods. 2.1.2 Method II 2.1.2.1 Dose to the Total Body The dose to the total body is: tb (mrem) = 2.0 x 10-6 93H + 5.0 x 10-3 g60Co + 0.3 Q65Zn D + 1.0 x 10-3 q + 1.0 Q90Sr + 1.7 x 10-4 Q99g0 + 1.2 x 10-4 Q133I + 5.0 Q134Cs + 3 0 Q137Cs where: Q90Sr = The release, based on the last measurement available, of Strontium-90.(C1). Q311 460Co Q65Zn The respective releases, based on the present Q99go,Q131I = measurement of Tritium, Cobalt-60, Zinc-65, 91331. Q134Cs.9137Cs M lybdenum-99, Iodine-131, Iodine 133, Cesium 134 and Cesium-137.(C1). 2 1.2.2 Dose to the Critical Organ organ (mrem) = 4.0 x 10-2 g60Co + 0.6 Q65Zn + 4.6 Q90Sr D _5_
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+ 0.6 Q1311 + 6.0 x 10-2 q133I + 1.6 x 10-3 Q9930 + 6.0 Q134Cs + 4.6 Q138Cs where: The release values (Q) are defined above. 2.1.2.3 Application of Method II Step 1. Determine the number of curies of each nuclide released during the period. Step 2. Based on the last available measurements of Strontium-90; estimate the number of curies released during the present period. Step 3. Obtain the Total Body and Orgaa doses by performing the calculation. Step 4. Record the total body and organ doses and maintain a cumulative dose for the quarterly and annual periods. 2.1 3 Method III The dose calculated shall be in conformance with Regulatory Guide 1.109 and may be calculated by the computer program IDLE using site specific parameters applicable during the period of release. ,,--n.
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3.0 GASEOUS RELEASE DOSE CALCI 1LATIONS 3.1 Technical Specification 3.8 D. Dose Rate Limits This section is to be used to insure compliance with the following Technical Specification: LIMITING CONDITION FOR OPERATION The instantaneous dose rate in unrestricted areas due to radioactive materials released in gaseous ef fluents f rom the site shall be limited to the following: a. The dose rate limit for noble gases shall be $500 mrem /yr to the total body, 53000 mrem /yr to the skin b. The dose rate limit for all radioiodines and radioactive materials in particulate form and radionuclides other than noble gases with half lives greater than 8 days shall be $1500 mrem /yr to anj organ. 3.1.1 Method I 3.1.1.1 Dose Rate Due to Noble Gases 3.1.1.1.1 Dose Rate to the Total Body The total body dose rate due to noble gases can be determined as 8 4 h h %b. W. follows: rem x 105 E 6tb( yr ) = 4.4 Qi DFBi i where: for each nuclide shown in Table 3.1.( C1) h = Release rate i sec i = Total body dose factor for each nuclide shown in Table 3.1 DFB 3.1.1.1.2 Dose Rate to the Skin The skin dose rate due to noble gases can be determined as follows: tb + 1.0 x 106 I k DFS $ skin ( yr ),f rem i i i where: btb = Dose rate to the total body, as determined in 3.1.1.1.1.(mrem) yr h = Release rate for each nuclide shown in Table 3.1.( C1) i ser DFSi = Skin dose f actor for each nuclide shown in Table 3.1 3.1.1 2 Dose Rate to Critical Organ due to Radioiodines and Particulates The dose rate to the critical organ can be determined as follows: borgan ( yr } 1311 where: Q1311 = Release rate for Iodine-131.( C1) see r. !.bi l J . ML40 d 3.1 1.3 Application of Method I Step 1. Determine the release rate in C1/sec for Iodine-131 and for each noble gas detected. Step 2. Find the dose rate to the total body by multiplying each noble gas release rate by the constant and by its total body dose factor determined f rom Table 3.1 and summing the contributions from each nuclide. Step 3 Find the dose rate to the skin by multiplying each noble gas release rate by the constant and by its skin dose f actor determined from Table 3.1 and summing the coatributions from each nuclide. Add, to this sum, the total body dose rate determined in Step 2 to obtain the total skin dose rate. Step 4. Find the dose rate to the critical organ by multiplying the Iodine-8 131 release rate by 2.2 x 10, 3.1.2 Method 11 The dose rates calculated will follow the guidance presented in Regulatory Guide 1.109. The noble gas dose rates may be calculated by the computer program IDLE and the dose rate due to radiciodines and particulates may be calculated by the computer program ATMODOS. 4 $[' I,, y'kh uM t .l d il Table 3.1 Noble Cases and Their Dose Factors (To be used for Technical Specification 3.3 D) Dose Factor, Total Body Dose Factor, Skin 3 3 DFS (mrem-m ) DFB (mrem-m ) Nuclide i t pCi-yr pCi-yr Kr-83m 7.56E-08* Kr-85m 1.17E-03 1.46E-03 Kr-85 1.61E-05 1.34E-03 Kr-87 5.92E-03 9 73E-03 Kr-88 1.47E-02 2.37E-03 Kr-89 1.66E-02 1.01E-02 Kr-90 1.56E-02 7.29E-03 Xe-131m 9.15E-05 4.76E-04 Xe-133m 2.51E-04 9.94E-04 Xe-133 2.94E-04 3.06E-04 Xe-135m 3.12E-03 7.llE-04 Xe-135 1.81E-03 1.86E-03 Xe-137 1.42E-03 1.22E-02 Xe-138 8.83E-03 4.13E-03 Ar-41 8.84E-03 2.69E-03 7.56E-08 = 7.56 x 10-8 4 4 F-b f, ah ,( y-- 3.2 Technical Specification 3.8 E, Dose to Air Due to Noble Gases This section is to be used to insure compliance with the following Technical Specification: LIMITING CONDITION FCR OPERATION The air dose in unrestricted areas due to noble gases released in gaseous effluents shall be limited to the following: During any calendar quarter to; 35 mrad for gamma radiation, a. and $10 mrad for beta radiation; b. During any calendar year; to $10 mrad for gamma radiation, and 120 mrad for beta radiation. 3.2.1 Method I 3.2.1.1 Air Dose Due to Gamma Radiation D(mrad) = 0.018 E Qi(C1) DFi i where: = Number of curies of noble gas nuclide "1" released. Qi Y DF = Gamma dose factor to air for nuclide "1". See Table 3.2. i 3.2.1.2 Air Dose Due to Beta Radiation 8 D(mrad) = 0.032 E Qi (Ci) DFi k,* 'lyWG }~~ h i4 Q where: Qi = Number of curies of noble gas nuclide "i" released. 8 DF = Beta dose factor to air for nuclide "i". See Table 3.2 i 3.2.1.3 Application of Method I Step 1. Determine the number of curies released during the period for each noble gas detected. Step 2. Find the dose to air due to gamma radiation by multiplying each noble gas released by the constant and by its gamma dose factor determined from Table 3.2. Sum the contributions from each nuclide to find the total air dose from gamma radiation. Step 3. Find the dose to air due to beta radiation by multiplying each noble gas release by the constant and by its beta air dose f actor determined from Table 3.2. Sum the contributions from each nuclide to find the total air dose from beta radiation. 3.2.2 Method II The dose calculated shall follow the guidance of Regulatory Guide 1.109 and may be calculated by the computer program AIRAD using the meteorological dispersion parameters applicable during the periods of release. ? ' i il !), k,' ~,'1 rs c L 4 Li Table 3.2 Noble Cases and Dose Factor to Air (to be used for Technical Specification 3.8 E) Beta Air Dose Gamma Air Dose Factor Factor 3 3 Nuclide DF (mrad-m ) DF (mrad-m } i i pCi-yr pCi-yr Kr-83m 2.88E-04* 1.93E-05 Kr-85m 1.97E-03 1.23E-03 Kr-85 1.95E-03 1.72E-05 Kr-87 1.03E-02 6.17E-03 Kr-88 2.93E-03 1.52E-02 Kr-89 1.06E-02 1.73E-02 Kr-90 7.83E-03 1.63E-02 Xe-131m 1.ltE-03 1.56E-04 Xe-133m 1.48E-03 3.272-04 Xe-133 1.05E-03 3.53E-04 Xe-135m 7.39E-04 3.36E-03 Xe-135 2.46E-03 1.92E-03 Xe-137 1.27E-02 1.51E-03 Xe-138 4.75E-03 9.21E-03 Ar-41 3.28E-03 9.30E-03 2.88E-04 = 2.88 x 10-4
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q 3.3 Technical Specification 3.8 F, Dose to an Ind'vidual This section is to be used to insure compliance with the following Technical Specification: Ifp41 TING CONDITION FOR OPERATION The dose commitment to an individual frou radioiodines, radioactive materials in particulate form and radionuclides with half-lives greater than 8 days other than noble gases in gaseous effluents released to unrestricted areas shall be limited to the following: a. During any calendar quarter <7.5 mrem, b. During any calendar year <l5 mrem 3.3.1 Method I To insure that the dose limit to any organ is met, it is necessary only to calculate the dose to the thyro id. The following equations are applicable: 3.3.1.1 Dose to the Thyroid Dthyroid (mrem) = 10 Q131I where: r- - r, $ ' bi, l\\i1) lI b 2 be i Q1311 = Number of curies of Iodine 131 released.(C1) 3.3.1.2 Application of Method I Step 1. Determine the number of curies of Iodine-131 released during the period. Step 2. Find the thyroid dose by performing the multiplication. Step 3. There will be 0.86 mrem dose to the thyroid from Carbon-14 and yr from unmonitored releases from the turbine building. The correct fraction of this dose must be added to the dose determined in Step 2. Step 4. Record the thyroid dose and maintain a cumulative dose for the quarterly and annual periods. Step 5. If the calculated dose exceeds the Technical Specification, the dose may be calculated by Method II. 3.3.2 Method II 3.3.2.1 Dose to the Thyroid thyroid (mrem) = 7.0 x 10-5 q3H + 7 0 Q1311 D + 0.1 Q133I + 0.01 Q1351 where: Q3He 91311 = Number of curies released during the period for Q1331. Q1351 Tritium, Iodine-131, Iodine-133 and Iodine-135.(C1) \\pF rrr F
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Determine the number of curies of Tritium and the three iodines released during the period. Step 2. Find the thyroid dose by performing the calculation. Step 3. There will be 0.86 mrem dose to the thyroid from Carbon-14 and yr from unmonitored releases from the turbine building. The correct fraction of this dose must be added to the dose determined in step 2. Step 4. Record the thyroid dose and maintain a cumulative dose for the quarterly and annual periods. 3.3.3 Method III The dose calculated shall be in conformance with Regulatory Guide 1.109 and may be calculated by the s mputer program ATMODOS using the meteorological parameters applicable during the periods of release. (_)I b '$u'b b: 1 /. 0 1 4.0 METEOROLOGY Annual average dilution factors based on site meteorological data 2 were computed for routine releases by the AEOLUS computer program. The values used are shown in Table 4.1. [F: I7 Sh I L9 L bit oua 3 TABLE 4.1 Vermont Yankee Dilution Factors Dose to Critical Dose Rate to Individual Dose to Air Organ Total Body Skin Critical Organ Gamma Beta Thyroid X/Q (sec) 1.001E-06(2) 3.851E-07(3) 1.001E-06(2) 3.851E-07(3) m D/Q (-f) 7 390E-10(3) 7.390E-10(3) m [X/Q( (8"C) 5.643E-07(1) 5.643E-07(1) 5.643E-07(1) 3 m (1) Maximum ground level gamma dose location SSE 0.621 miles. (2) Maximum ground point concentration NW l.553 miles. (3) Worst real milk animal concentration NNW l.85 miles. 9 e e l.: 'i ,es d %} h.,, a$'. !x$ k i /m, -19
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- EXPOSURE PATHWAY SAMPLE LOCATION DISTANCE FROM DIRECTION FROM AND/OR SAMPLE AND DESIGNATED CODE THE PLANT (KM)
THE PLANT 1. AIRBORNE
- a. Radioiodine AP/CF-ll Hinsdale, NH 1.1 ENE and AP/CF-12 N. Hinsdale, NH 4.0 NNW Particulate AP/CF-13 Riv<
Station 3.3, VT 1.8 SSE AP/CF-14 Ca al Park, VT 2.5 SSW AP/CF-21
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VT 25.0 WNW 2. DIRECT GM-ll H' lale, NH 1.1 ENE RADIATION GM-12 N. dinsdale, NH 4.0 NNW GM-13 River Station 3.3, VT 1.8 SSE GM-14 Central Park, VT 2.5 SSW GM-15 Guilford Rd., VT 3.2 WNW GM-16 Vernon School 0.563 SW GM-21 Hobgack Mt., VT 25.0 WNW GM-22 Northfield, MA 11.3 SSE 3. WATERBORNE
- a. Surface WR-ll Monitor #3, VT 2.2 Downriver WR-21 Monitor #7, VT 6.4 Upriver
- b. Groundwater WG-ll Plant well Onsite WG-12 Vernon Nursing well 2.0 SSE
- c. Sediment SE-ll Between discharge Discharge pond from structure and Vernon Dam Shoreline
- Sample locations are shown on Figures 5.1 and 5.2.
[ ~}g ph r 'J ' fi e ..i;/ e ,uu a TABLE 5.1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS
- EXPOSURE PATHWAY SAMPLE LOCATION DISTANCE FROM DIRECTION FROM AND/Oi SAMPLE AND DESIGNATED CODE THE PLANT (KM)
THE PLANT 4. INGESTION
- a. Milk TM-ll Miller Farm 0.8 WNW TM-12 Whitaker Farm 2.5 S
TM-13 Bathlon Farm 5.5 SSE TM-21 Brattleboro Dairy 15 NNW
- b. Fish and FH-ll Vernon Pond Discharge pond Inverte-FH-21 Route 9 Bridge 12.8 Upriver brates
- c. Food TF, TV-ll Vernon Garden 1 a
a Products TF-12 Vernon Garden 2 a a TF-21 Five Acre Farm, 7.7 SE Northfield, MA
- Sample locations are shown on Figures 5.1 and 5.2.
a As determined by annual garden survey. __n-, - Q" Qr}i l4 L h ' hfff *' -- p.' ' "% g. Plant Site G d '
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4 'n' T i l. .) l j/ _ C.t " APPENDIX II Annual average dilution f actors based on onsite meteorological data were computed for routine (long-term) releases by the Yankee Atomic Electric Coupany's (YAEC)AEOLUS(2) Computer Code. AEOLUS is based, in part, on the straight-line airflow model as discussed in Regulatory Guide 1.111(3) and includes the following basic features: hourly meteorological data input (wind direction, wind speed, vertical temperature dif ference, and, optionally, direction fluctuation, air temperature, sea water surf ace temperature and solar radiation), ntraight-line air flow model with Gaussian diffusion, plume centerline and sector-average models witl single or split (vertical / horizontal, i.e., split-sigma) atmospheric stabilities, part-time ground level and part-time elevated releases (split-H model), scabreeze option for coastal sites (split-H, split-sigma, trapping and fumigation), multi-energy sector-averaged finite cloud dilution f actors for gamma dose calculations (both normal and seabreeze cases), terrain features, plume rise (buoyant or momentum), h,[k b ,T O ^ .C/U-depletion in transit (2 models), recirculation correction factors (built-in options for flat terrains and river valleys, or user selected values), l deposition rates (2 models), and dose statistical distributions for postulated accidental radioactive releases and exposure intervals (based on dose rate data per unit dilution f actor (X/Q) as input; thyroid, total body beta, total body gamma and skin doses. AEOLUS produces hourly and long-term average of non-depleted dilution f actors for evaluating ground level concentrations of noble gases, tritium, carbon 14 and non-elemental iodines, depleted dilution factor for estimating ground level concentrations of elemental radiciodines and other particulates, effective gamma dilution factors for evaluating gamma dose rates from { a sector-averaged finite cloud (multiple-energy undepleted source), and deposition factors for computing dry deposition of elemental radioiodines and other particulates. A more detailed description of the AEOLUS dif fusion model is provided in section 2.3.5 (long-term diffusion estimates) of the NEP 1&2 PSAR(4)
- hrFU 6).
d 1 !' r L. -. d d d and the AEOLUS computer code manual (2), Annual average non-depleted dilution factors, effective gamma dilution factors and deposition (D/Q) rates for Vermont Yankee were calculated using the following AEOLUS options; Sector-average uodel with temperature dif ference ( AT) atmospheric stabil1: tes, straight-line airflow model with gaussion diffusion, buoyant plume rise, and no recirculation correction factors, y/Q and D/Q values for the restricted area boundry critical sector are provided in Table 5.1. Short term dilution f actors based on on-site meteorological data were computed for intermittent releases by YAEC's SKIRON(5) computer code. SKIRON is based, in part, on Regulatory Guides 1.XXX(6) abd 1.1110) and also includes all the basic features of AEOLUS. Hourly dilution f actors are computed for either ground level or elevated releases using both the plume centerline model and the sector average model. Plume centerline values are for estimating short-term atmospheric dispersion (up to 8 hours) and sector average values are for dispersion during longer periods of time. In the plume centerline model, for ground level releases during neutral and stable etmospheric conditions and low wind speeds, atmospheric dispersion is corrc :ted for either plume meander eff ects or for the additional dispersion of the effluent plume within the wake caused by buildings adjacent to the release point. During all other atmospheric stability and/or wind speed conditions, credit is taken only for building wake effects. In the sector-average model, dispersion is based on Regulatory Guide 1.111 and meander effects are ignored. Both building wake and plume meander are excluded from the equation for elevated C N b ub' mu b releases. Atmospheric dilution factors are computed by the above models for each sequential hour of measured meteorological data and for receptors positioned in the 16 cardinal compass directions around the plant. The hourly dilution factors obtained as described and the corresponding direction in which the wind is blowing during each hour are then stored in sector dependent arrays for sequential processing. This involves the averaging of selected hourly dilution values over successive, overlapping time intervals. For each selected interval size, the processing begins with the first hourly dilution value on record and then repeated for the same interval size starting with each subsequent hour of dispersion data. In the averaging process, the only values within a given interval that are considered in evaluating the mean dilution factor for the interval are those for the specific wind direction being analyzed. Missing data are handled by imposing the condition that at least half of the entries within an averaging interval correspond to valid observations. Missing data points are not included in the averaging. The average dilution f actors computed as described are subsequently classified, for each 22-1/2 degree sector, into groups, and corresponding cumulative probability distributions are prepared. For each sector the dilution factors at a number of percentile points are determined. These points define the percent of time a dilution value is equalled or exceeded. Dilution factors for intermittent releases were obtained by calculating the ratio betwaan the 15% short-term and annual average I/Q's - l1 V.'9,1 i 4 j r. _.. - L1 u; 4 for the site's critical sector. The critical sector was considered that sector with the maximum site boundary annual average X/Q. These ratios were then plotted as a function of duration of total release for use by the plant operators. See Figure 5.1. A more detailed description of the SKIRON dif f usion model is provided in Section 2.3.4 (Short-Term Dif f usion Estimates) of the NEP 1&2 PSAR(4). The SKIRON model has been updated to incorporate those modifications that are provided in the latest verston of Regulatory Guide 1.XXX (August 1978).(6), ~ fl ,'C< ] References 1. Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I", U.S. Nuclear Regulatory Commission, Revision 1, October 1977. 2. Hamawi, J.N., "AEOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power-Plant Effluents and for Computing Statistical Distributions of Dose Intensity from Accidental Releases", Yankee Atomic Electric Company Technical Report, YAEC-ll20, January, 1977. 3. Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents In routine releases from Light Water Cooled Reactors", U.S. Nuclear Regulatory Commission, March 1976. 4. NEP 1&2 Preliminary Safety Analysis Report, New England Power Company, Docket Nos. STN 50-568 and STN 50-569. 5.
- Hamawi, J.N., J. Laznow, "SKIRON: A Computer Code for Determining Atmospheric Dispersion Conditions for Design Basis Accident Evaluations", Yankee Atomic Electric Company, Technical Report YAEC-1138, October, 1977.
6. Regulatory Guide 1.XXX, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", U.S. Nuclear Regulatory Commission, August, 1978. }}