ML20147F148

Response to Intervenor SA Brand & Mid-Hudson Nuclear Opponents.Contention That Seismic Design of Proposed Facility Inadequate Unproven.Ground Acceleration Associated W/Proposed Safe Shutdown Underestimated

ML20147F148
Person / Time
Site:	Green County
Issue date:	10/02/1978
From:	Mcgurren H NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:
Shared Package
ML20147F156	List: ML20147F148 ML20147F159 ML20147F169 ML20147F179 ML20147F191
References
NUDOCS 7810190012
Download: ML20147F148 (7)

Text

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[ 4 UNITED STATES f~ \

g NUCLEAR REGULATORY COMMISsl0N

<a WASHING TON, D. C. 20655 SOUTHERN CALIFORNIA EDISON COMPANY AND SAN DIEGO GAS AND ELECTRIC COMPANY I DOCKET NO. 50-206 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 AMENDMENT'TO PROVISIONAL OPERATING LICENSE Amendment No. 38 License No. DPR-13 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Southern California Edison Company and San Diego Gas and Electric Company (the licensees) dated December 30,1977 (Proposed Change No. 68) complies with the standards and requirements of the Atomic Energy Act of '

1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regula-tions; D.

The issuance of this amendment will not be inimical to the common defense and security or to'the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part S1 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. DPR-13 is hereby amended to read as follows:

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(B) Technical. Specifications 1

The Technical Specifications contained in Appendices A ,'

and B, as revised through Amendment No. 38 , are hereby incorporated in the license. .The licensee shall . operate the facility.in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION ,

Msf A, Dennis L. Ziemann, Chief

/ Operating Reactors Branch #2 ,

Division of Operating Reactors

Attachment:

l Changes to the Technical Specifications Date of Issuance: November 17, 1978 I

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A,TTACHMENT TO LICENSE AMENDMENT N0. 38 f

PROVISIONAL OPERATING LICENSE NO. DPR-13

. 4 DOCKET NO. 50-206 Revise Appendix A Technical Specifications and Bases by removing the following pages' and inserting the enclosed pages. The revised pages are identified _ by the captioned amendment number and contain vertical lines indicating the area of change.

~ REMOVE INSERT 12 12  :

13 13 'o 88 83 89-95 -

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• 22 ' !2* i 25 25*

54 54*  ;

56 66*

58b 58b*

60L 60L*

86 86*

(Appendix B) 3-10 3-10*

5-9 5-9*

5-11 5-11*

• These pages are included to correct administrative errors which occurred with the issuance of previous license amendments. '

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3.1.3. COMBINED HEATUP, COOLDOW, AND PRESSURE LIMITATIONS Applicabilitv: Applies to heatup and cooldown of the reactor-coolant system.

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_Obj ec t ive: .To maintain the structural integrity of the reactor coolant system throughout the lifetime of the plant. - I Specification: l i A. Reactor pressure and heatup and cooldown of the reactor coolant system during the first 6 years of equivalent full power operation shall be limited in accordance with Figures 3.1.3a and 3.1.3b.

Thereafter, limits shall be based on neutron exposure equivalent to not less than 6 years of . full power operation, and Figures 3.1.3a and 3.1.3b shall be updated accordingly.

B. , Figures 3.1.3a and 3.1.3b shall be updated in accordance with the following criteria and procedures:

(1) The methods of Appendix C, " Protection Against Nonductile Fa ilu re", to Section III of the ASME Boiler and Pressure Vessel Code shall be used to obtain the allowable pressure-temperature relationships for the reactor coolant system.

(2) The curves in Figure 3.1.3c shall be used in predicting the reference nil-ductility temperature increase, o RT unless measurements on the irradiation specicens'show *DT' N ARTNDTs greater than those predicted by the curves, in which case a new curve having the same slope as the original shall be constructed.

. C. The pressurizer heatup rate of 100*F/ hour and coaldown rate  !

of 200 F/ hour shall not be exceeded.  !

l i s' D. The reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line as shown in Figures 3.1.3a and 3.1.3b. -

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, . Amendment No. 38 Y

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Basis: The initial Reference Nil Ductility Temperature (RTNDT) for all reactor vessel material based on Charpy V-notch data drop weight tests, and conservative estimates

• is 160*F or less. The RTNDT at the 1/4 thickness location (location of the Appendix G reference flaw tip ) increases as a function of cumulative neutron exposure up to approximately 258'F for the core region of the reactor vessel after 30 years of operation.

A six (6) equivalent full power year service period was chosen for the operational limits given in this specification because at the end of this period the limiting RTNDT of the reactor vessel at the 1/4 thickness location is approximately 207'F in the core region. This provides an approximate 47'F margin over the head region of,the vessel which sees negligible radiation exposure throughout the life of the plant.

The highest RTNDT f the core region material is determined by adding the radiation induced A RTNDT for the applicable time period to the original RTNDT shown in Table 1. The fast neutron (E 2 1 Mev) fluence at 1/4 thickness and 3/4 thickness vessel locations is given as a function of full power service life in Figure 3.1.3d. Using the applicabic fluence at the end of the year period and the copper content of the material in question, the 6 RT NDT is btained from Figure 3.1.3c.

Values of a RTNDT may continue to be determined in this manner unless measurements on the irradiation specimens show 4 RTNDTs greater than those predicted by the curves for the equivalent capsule exposure.

Allowable pressure temperature relationships for various heatup and cooldown rates are calculated using methods derived from Non-Mandatory Appendix G in Section III of the ASME Boiler and .

Pressure Vessel Code, and discussed in detail in Reference 1.

The results of these calculations are provided in Reference 2.

The design heatup and cooldown rates for the pressurizer are 100 F/ hour and 200 F/ hour, respectively.

The criticality limit'given in Figures 3.1.3a and 3.1.3b are based on RTNDT + 160*F and a 40'F margin over the pressure-temperature limit curves for heatup and cooldown. The criticality limit is designed to provide assurance that the vessel will not exceed the allowabic stress intensity factor before yielding under postulated transient conditions.

• " Strong" direction Charpy V-notch data was reduced by 35% to estimate " weak" direction data, and drop weight NDTT of forgings vas estimated as 60'F and that of weldments as 0*F or 30 ft-lb 7' . temperature, whichever is higher.

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f. - Records of in-service ' inspections performed pur-

'suant to these. Technical Specifications. ,

3 Records ~ of Quality Assurance activities as required by the QA Manual.

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h. Records of reviews performed for changes made to ,- ,

, . procedures or equipnent or reviews or tests 'and experiments pursuant to 10CTR50.59.

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1. Records of meetings of the OSRC and the NARC.

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6.10.3 The following records shall be retained for two years.  ; l

a. Records of ' facility radiation and contamination surveys. -

! b. Records of training-of facility personnel.

6.11 .,

RADIATION PROTECTION PROGRAM ,

Procedures for personnel radiation protection shall . I be pre' pared consistent with the requirements of 10CFR Part 20 and shall be approved, maintained and ,

adhered to for all operations. involving personnel

. radiation expecure. .

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.,a Amendment No. M 38 -

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3.1 REACTOR COOLANT SYSTEft, 3.1.1 Maxir um Reactor Cool _a_nt Activity Applicability: Applies to measured maxir:dn activity in the rea: tor coolant system at .any tinie.

Objective: To limit the crinsequences of an accid. ental release of reactor coolart'to the environment.

Speci fica tion: A. The specific activity of the reactor coolant shall be limited to:

1. [.1.0 p Ci/gn dose equivtlent 1-131.

.2. I 100 /E'u Ci/gm, where E is the average (weighted in proportion to the concentration of each radionuclide in the reactor coclant at the tire of sac.pling) of the sur of the average o

' gration (pta andfor in LEV) ganma energies isotopes, other thanper disinte-iodines, with half lives greater than 15 rrin--

utes, making up at least 9E% of the total non-iodine activity in the coolant.

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B. Action

1. With the specific activity of the coolant determined to be >1.0 u Ci/gn but < 6.0 u Ci/

gm dose equivaler.t I-131, the reactor tr.ay be started up or operation may continue for u; to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that operation ~nder u these circumstances does not exceed 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />

- in any consecutive 12 month period. Should the total operating time at a r(acter coolant specific activity > 1.0 u Ci/gran Dose Equivalent 1-131 exceed 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any cen-secutive six month period, the licensee 56 all report the number of hours of operation Ebove this liirit to the NRC within 30 days.

2. With the specific activity of the reactor coolant determined to be > 1 u Ci/gm dose equivalent 1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> dur-ing one continous time interval or > 60 y Ci/gm dose equivalent I-131 or > 100/E u Ci/gm, have the reactor suberitical with the average temperature of the reactor coolant (Tavg) less than 535'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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' Amendment No. 74, 38

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(2) Containment Spray System

a. Two refueling water pumps are operable.

, b. Two hydrazine additive pumps are operable.

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c. Hydrazine tank level and hydrazine concentration comply with Specification 3.3.4.

(3) Valves and interlocks associated with each of the .

above systems are operable.

(4) Effective leakage from the recirculation loop outside the containment shall be less than 625 cc/hr. as calculated from the following formula.

Effective Leakage = a) x L) + a2*L2+a3*L3 where.

L)

= pump and valve leakage which drains to auxiliary building sump L

2

= valve leakage in auxiliary building or doghouse L

3

= valve leakage outside a)

= iodine release factor for leakage in auxiliary building sump a

2

= 1 dine release factor for leakage in auxiliary building or doghouse a

3

= iodine release factor for leakage outside the auxiliary building or doghouse If effective leakage from the recirculation loop outside the containment exceeds 625 cc/hr, make necessary repairs to limit leakage to 625 cc/hr within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

B. During critical operation or when the reactor coolant system temperature is above 200* F, as appropriate per Item A above, maintenance shall be allowed on any one of the following items at any one time:

(1) One motor-operated valve at a time (MOV 1100B or 1100D) in the recirculation loop upstream of the charging pump suction header, for a period of time not longer than 72 consecutive hours.

(2) One refueling water pump and/or its associated discharge valve at a time, for a period not longer than ,

72 consecutive hours.

(3) one hydrazine pump and/or its associqted discharge value (SV600 or 601) at a time, for a period of time not longer than 72 consecutive hours.

AmendmentNo.?/.38

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O 3 .3.3 MINT!TM 1%TER VOLUI(E GD tono'l CO" CENTRATION IN T'4E REFUELIt:C WATER STC'?A0E TAUX Applicabili ty: Applies to the inventory of borated refueling water.

Objective: To insure irre.ediate availability of safety injection and

' containment spray water of required quclity.

Specification: When the Safety Injection System or the Containment Spray System is required to be operable, the refueling water tank shall be filled to at least elevation 50 feet with water having a boron concentration of not less than 3750 ppm and not greater than 4300 ppm.

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Basis: The refueling water s torcge tcnk serves two purposes; namely:

(1) As a reservoir of borated water for accident mitigation purposes, (2') As a reservoir of bora ted water for flooding the refueling cavity during refueling.

Approximately 220,000 gallons of borated water is required to provide adequate post-accident core cooling and containment spray to maintain ca}cylated post-accident doses below the limits of 10 CFR 100\l . The refueling water storage tank l

filled to elevation 50 feet represents in excess of 240,000 gallons.

A boron concentration of 3750 ppm is requirep to meet the requirements of postulated steam line break.121 A maximum bornn concentration of 4300 ppm ensures that the post-accident water is maintained at a pH betweencontainment 7.0 and 7.5(3) sump The refueling tank c pacity of 240,000 gallons is based on refueling volume requirements.

Sustained temperatures belev 32*F do not occur ct San Cnofre.

At 32*F, boric ecid is so'.uble up to approxicately 4650 ppm bo ro n. Therefore, no speial provishns for ter.perature control to avoid either freezing or boron precipitction are necessary.

Re fe renc e: (1) Enclosure 1 " Post-Accident Pressure Reanalysis, San Onofre Unit 1" to letter dated January 19,1977 in O .

(2)

Docket No. 50-206.

" Steam Line Break Accident Reanalysis, San Onofre Nuclear Generating Station, Unit 1, October 1976" submitted by letter dated December 30, 1976 in Docket No. 50-206

/ (3) " Addition information, San Onofre, Unit 1" submitted by Amendment No. 2/5, 38 letter dated March 24, 1977 in Docket No. 50-206.

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'^ ~1 45 RADIOACTIVE I

9 10UID WASTE RELEASE Applicability:

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Applies to release of radioactive liquid waste to the '

Circulating Water System. *

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Obiective:

To verify that discharge of radioactive waste to the Circula- *

ting Water System is maintained below the limits set forth in 40 CFR 20.

Specification: A.

Averaged over a year, the release rates of liquid wastes shall not result in concentrations in the circulatir.g l

vater discharge in excess of Part 20 limits.for unrestric-  ;

ted areas, execpt that the maximem relecse rate over the I period averaged of 11=1t.

one hour shc11 not exceed 10 times the yearly I B. 'At least one circulating water, pump shall be in operation

- whenever rsdiocctive liquid wastes are' released. g C. .

m and determination made of the maximum rate.

e p,

D.

shall be monitored for isotopic or gross act

() discharge.

the following methods:Such monitoring may be accomplished by either of

1.  ;

Continuous monitoring with the in-stream liquid waste

. monitor, is inoperable. channel; or, if the liquid waste monitor channel

' 2.

Analyses of a minimum of three samples of effluentmstrea taken approximately release period." towards the beginning, midpoint, and end of each E.

The liquid waste monitor and the flow rate meter shall be calibrated of the monitor shall be, tested weekly.at a mininum frequ F.

A record of all liquid waste releases shal1' with Specification 6.10. be kept in accordance

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Change }d Amendment No. 3g r, ~ ~ w ,a ,, - ., a . -- - - -. - - . -,. a

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h 4.6 RADIOACTIVE CASEOUS WASTE RELEASE Applicability:, Applies to the release of radioactive gaseous vaste from the plant stack.

Obiective: To verify discharge of radioactive gaseous vaste to the atmosphere will not result in ground level radioactivity -

concentrations outside the plant boundaries in excess of

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- limits established in 10 CFR 20. .

Specification: A. Averaged over a year, release rates of gaseous wastes.

in curies /sec shall not result in a value exceeding that eticulated from the following formula:

ue 1.8 x 105

(.m3 sec X C x

Where Cx is the concentration of any radioisotope X, the values of the concentrations of all isotopes dis-charged shall be such that Cx is icas than TEC) 1.0,(MPC)x as defined above shall he that stated in Colunm 1, Table II of 10 CFR 20. The maximum release rate over any one hour shall not exceed 10 times the yearly averaged limit as stated above.

h B. At least one stack fan shall be in operation delivering normal flow whenever radioactive gaseous vastes are released to the vent stack.

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, C. All radioactive wastes discharged through the stack shd11 i be monitored conrinuously for gross activity. '

D. A record of the above releases shall be kept in accordance with Specification 6.10.

E. The stack gas and particulate monitors shall be calibrated at a mirimum frequency of once every six months, and normal response of each monitor shall be tested weekly.

Basis: Prior to release to the atmosphere, gaseous vas tes from the ratioactive was te disposal svytem are mixed in the stack flow of two 20,000 cfmf fans.(VDilution then occurs in the .

a tmos phe re .

The formula prescribed in specification A takes atmospheric dilution into account and ensure.- that at the point of maximum ground concentration the requirements of 10 CFR 20 will not be

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Amendment No. 38 -

-58b-P. Within 130 days from the date of issuance of addenda to Section XI l of the AStZ Boiler and Pressure Vessel Code, a review for applica- l bility of such addenda shall be made and a proposed change to the j specifications as determined by this review shall be submitted to l the AEC pursuant to Section 50.59 in 10 CFR Part 50. l Basis: Periodic visual inspection of the rod cluster control assemblies will verify that the structural integrity is maintained.

The inservice inspection program specified conforms (as closely as the as-built condition of the plant permits) to the 1971 edition (including the Summer and Winter 1971 Addenda) of Section XI of the ASME Boiler and Pressure Vessel Code, " Rules for Inservice Inspection of Nuclear Reactor Coolant Systems". To the extend applicable, based upon the existing design and construction of the plant, Table 4.7.1 duplicates Table IS-261 of Section XI of the Code for ease of comparison. Table 4.7.2 ensures that the inspections will be made according to the inter-vals rpecified in IS-242 of Section XI of the Code. Specification B includes a provision for future updating as required by outages, changes in refueling dates, etc.

Significant exceptions taken to Table IS-261 of Section XI are as follows:

Item 1.5 - This item has been deleted since the plant design and currently available techniques do not permit a volumetric examina-tion of this type of penetration.

Item 2,.3 - This item has been modified to delete the surface examina-tion since the pressurizer heater penetrations are not accessible to a surface examination.

Item 2.5 - This item has been deleted since no bolting 2-inch and over is used.

Item 3.4 - This itan has been deleted since no bolting 2-inch and over is used.

Item 4.2 - This item has been modified to exclude examination of longitudinal pipe welds since none exist in present plant design.

Item 4.3 - This item has been deleted since no bolting 2-inch and over is used. ,

1 Item 4.5 - This item has been deleted since no integrally-welded supports are used.

s Item 5.1 - This item has been modified to delete volumetric examina- l tions since currently available techniques do not exist.

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Change No. )d Amendment No. 38

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i l4.15 Fire Protection Systems Surveillance

'O Applicability: Applies to the surveillance of fire detection and extinguishing systems and equipment.

Objective: To ensure the operability of fire detection and

. - extinguishing systems and equipment.

l Specifications: A. The Fire Suppression Water Systeml shall be ,

demonstrated to be oper3ble.

(1) With the San Onofre Unit 1 fire water pumps satisfying the pum *eouirements of Technical Specification 3.1 , , at least once pet seven days by verifying the water supply volume in the

' ' San Onofre Unit 1 Service Water Reservoir. With the San Onofre Units 2 and 3 fire water pumps satisfying the pump requirements of Technical Sepcification 3.14.A(1), by initially verifying the water supply volume in the San Onofre linits 2 and 3 service and firewater storage tanks and at least once per.seven days thereafter.

(2) At least once per 31 days on a staggered ' test 1

. basis by starting each pump satisying the pump requirements of Technical Specification 3.14. A(1)

- and operating it for at least fifi. 'en minutes.

(3) At least once per thirty one days by verifying  !

that each valve (tanual, power operated or -

aut'matic o is in its ccrrect position. For vtilves  :

located inside the containment sphere, verifi--

cation sna11 oe made consistent'with the 31-day requirement when possible during available plant  !

outages or during containment entrances for other .

rear.ons. l (4) At least once per 12 months by cycling each testable valve through one complete cycle of full travel.

The isolation valve between the screen wash pumps  ;

and the Fire Suppression Water System shall be tested j only whenetter the screen wash pumps are used to  !

satisfy the pump requirements of Technical Specification l 3.14.A(1). . l (5) At least once per 10 mnths by performing a system I functional test whicn includes simulated actuation of the system,' and:

a. Verifying that each valve in the floit path is in its correct position, j b.- Verifying that each pump develops at least 90% of y

. the flow and head at som

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purnp perfonnance curves,g point on the manufacturer's t

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6.9.3 Unique Reporting Requirements The following special reports shall be submitted as required:

a. Inservice Inspection (Technical Specification 4.7)' .

b. Reactor Vessel Surveillance Program (Technical Specification

. 4.9)

c. Fire Protection Systens (Technical Specification 3.14).

The results of required leak tests performed on sealed source) (Technical Specification 4.12) shall be reported annually if the tests reveal the presence of 0.005pci or more of removable contamination.

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AmendmentNo.7I,38

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Reporting Requirement .

A detailed interpretation and analysis of the results of the.

General' Ecological _ Survey will be presented in the Annual Operating Report, including the' California Depart =ent of Fish and Game annual fish catch statistics. Alte rnatively ,

a su= mary analysis and preliminary interpretation of the General Ecological Survey data may be presented semiannually, in which case a detailed interpretation and analysis of the complete year of data vill be provided annually. .Special reports will be prepared in accordance with ETS 5.6.3.d whenever significant or unusual changes are observed. l Bases i The basis of the general ecological survey will be to assani that

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avery reasonable procedure is taken to monitor any signitts(nt l ef f ects on the matine environment which might be the result of the operation of the Station; and to. identify any significant l changes to the tarine ecology. '

The ecological survey will be coordinated with the chemical ,and .

oceanographic programs. The locations of the plankton, fish,

' . benttc, intertidal and all pnysical monitoring stations are

(~ baseu on 9 years of data from the ongoing monitoring' program.

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Divers will occupy all stations prior to their establishment.

3-10 Amendment No. S4. 38 O

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c. Radiological Environmental Monitoring (1) For each medium sampled during the reporting period, e.g. , air, baybottom, surface water, soil, fish, include:

(a) Number of sampling locations. .

(b) Total number of samples.

(c) Nunber of locations at which levels are I found to be above local backgrounds, and (d) Highest, lowest, and the mean concen-trations or levels of radiation for the sampling point with the highest mean and description of the location of that point with respect to the site.

(2) If levels of radioactive materials in environ-(, ).

mental media as determined by an environmental monitoring program indicate the likelihood of public intakes in excess of 1% of those that could result from continuous exposure to the concentration values listed in Appendix B.

Table II, 10 CFR Part 20, estimates of the likely resultant exposure to individuals '

and to population groups, and assumpti---

' upon which estimates are based shall be <.1o-vided.

(3) If statistically significant variations of offuite environmental concentrations with time are observed, correlation of these results with effluent release shall be provided.

(4) Individual samples which show higher than normal levels (25% above background 'or external dose, or twice background for radionuclide content) shall be noted in the reports.

(5) Results of all radiological samples taken shall be summarized on a quarterly basis following the format of Table 5.6-1 for inclusion in the Annual Operating Report. In the event that some results are not avail-

~' able by March 31 of the followirig year, the report i

' shall be submitted noting anc explaining the reasons

[ for the missing results. The missing data shall be cubmitted as soon as possible ig a supplementary report, but not later than July 1 of the year in which the report was due. ,

5-9 Amendment No. 14, 38 p.

b' d.

s Oceanographic and Biological Environmental Monitoring l

A detailed annual analysis of oceanographic and l t biological monitoring will be sabmitted with each j Annual Operating Report. The analysis report will be l a su= mary'of the monitoring program results and an j assessment of the observed impacts, if any, of the i Station operation on the marine environment. In the l etent that some detailed analyses are not available by i March 31 of the following year, the report shall be [ J submitted noting and explaining the reasons for the missing results, but including at a minimum preliminary analyses and conclusions for the reporting period. The missing data shall be submitted in a supplemental report before July 1 of the year that the report was due.

The first Semiannual Operating Report following implementation of this environmental technical specification will identify 1 the precedures to be used in the monitoring program.

5.6.2 Routine Reports - Semiannual Within 60 days after January 1 and July 1 of each year a

{, report shall be submitted covering the radioactive content of effluents released to unrestricted areas and shipments of solid waste during the previous six months of operation. The data :: hall be suc=srized on a monthly basis and shall include as a minimum the following:

a. Radioactive Effluent Releases ,

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. A statemeht of the quantities of radioactive effluents released from the station with data su=marized on a monthly basis following the format of USAEC Regulatory Guide 1.21.

(1) Gaaeous Effluents .

(a) Gross Radioactivity Releases

1. Total gross radiot.ctivity (in +

curies), primarily noble and activation gases. ,

ii. Maximum gross radioactivity release rate during any one-hour period.

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