ML20079M972

Suppl to Proposed Change 78 to Application to Amend License DPR-28,revising Radiological ETS

ML20079M972
Person / Time
Site:	Vermont Yankee
Issue date:	01/23/1984
From:	Sinclair J VERMONT YANKEE NUCLEAR POWER CORP.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML20079M975	List: ML20079M972 ML20079M989
References
FVY-84-6, NUDOCS 8401270492
Download: ML20079M972 (13)

Text

VERMONT YANKEE NUCLEAR POWER CORPORATION Proposed Change No. 78 Supplement 1

. RD 5, Box 169, Ferry Road, Brattleboro, VT 05301 ,, gly ,g j

ENGINEERING OFFICE January 23, 1984 167 " CE3ERROAU FRAMINGHAM, MASS ACHUSETTS 01701 FVY 84-6 TELEPHONE 617-872-8100 2.C.15.1 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Of fice of Nuclear Reactor Regulation Mr. Darrell C. Eisenhut, Director Division of Licensing Reference s : (a) License No. DPR-28 (Docket No. 50-273)

(b) Letter, VYNPC to USNRC, Proposed Technical Specification Change No. 78, WVY 79-15, dated February 13, 1979 (c) Letter, VYNPC to USNRC, FVY 83-127, dated December 27, 1983 (d) Letter, USNRC to All Operating Licenses, dated July 11,1978 (e) Letter, USNRC to All Boiling Water Reactors (BWRs), dated November 15, 1978 (f) Letter, USNRC to VYNPC, Amendment No. 47 to Facility License No. DPR-28, dated October 10, 1978

Subject:

Revised Vermont Yankee Radiological Effluent Technical Specifications (RETS)

Dear Sir:

Pursuant to Section 50.59 of the Commission's Rules and Regulations, Vermont Yankee Nuclear Power Corporation hereby proposes the following modification to Appendix A of the Operating License:

Proposed Change In accordance with our commitment to provide a formal RETS amendment to our operating license [ Reference (c)), this submittal amends in its entirety Proposed Change No. 78, dated February 13,1979 [ Reference (b)]. Reference is made to the Operating License DPR-28 and the Technical Specifications l contained in Appendix A, issued to Vermont Yankee Nuclear Power Corporation i

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8401270492 840123 PDR ADOCK 05000271 P PDR

United States Nuclear Regulatory Commission . Janua ry 23, 1984 Attention: . Mr. Darrell G. Eisenhut, Director - Page 2 for the Vermont Yankee Nuclear Power Station locaced in Vernon, Vermont'.

" . - Attech= cut: A itami:ce the scjer ch =gce to- the. current Vermont Ycnkee Technical Specifications and indicates where differences occur between our proposal and the NRC's model RETS, as contained in NUREG-0472/0473. .

- Attachment B contains the individual page changes necessary to implement the proposed RETS.

. In addition, Pages 180c and 180h have been revised to reflect removal of the existing provisions for operation action in the event of certain levels of off gas activity measured at the Steam Jet Air Ejector (SJAE).; These levels,

. which were added by Reference (f), are less conservative than those called for in the new RETS (Section 3.8.K) and are deemed no longer to be necessary.

I Reason for Change -

The proposed' changes are in ' direct response to the NRC's request that Vermont Yankee Nuclear Power. Corporation amend its Operatirs License (No.

DPR-28), as discussed in References (d) and (e).

Basis for Change The proposed Technical Specifications address issues put forth by the NRC

.in - their Draf t Radiological Effluent Technical Specifications

1(NUREC-0472/0473), and are intended to implement the following Federal Regulations

10CFR Part 50, Section 50.36a; Section 50.34a; 10CFR Part 20; 10CFR Part 50, Appendix I; General Design Criteria 60 and 64; and 40CFR Pa rt 190.

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In addition, we expect to submit a final Off-Site Dose Calculation Manual

[ (ODCM) for implementing several of the requirements of the proposed Technical Specifications by the end of February, as indicated in Reference (c).

Safety Considerations l The changes proposed were requested by the NRC and are not considered to l constitute an unreviewed safety question as defined in 10CFR50.59(a)(2). This

. change has .been reviewed by the Plant Operations Review Committee and the

! ~ Nuclear Safety Audit and Review Committee.

I Significant Hazards Consideration The NFC has provided guidance concerning the application of the standards for determining whether a "significant hazards" consideration exists by providing certain examples [48FR14870]. One of the examples of actions deemed

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_ not likely to involve a significant hazards consideration relates to changes that constitute additional restrictions or controls not presently included in 4

Technical Specifications.

United States Nuclear Regulatory Commission January 23, 1984 Attention: Mr. Darrell G. Eisenhut, Director Page 3

'The NRC, in a revision to 10CFR Part 50, Appendix I, required Licenrees to .improva -and modify their radialagical ef f1 ment ayarama in a manner that would keep releases-of radioactive material to unrestricted areas during normal operation as' low as is reasonably achievable. . In complying with this re quirement , it became necessary to add additional restrictions and controls to the Technical Specifications to assure compliance.

Based on the above, we have determined that this change does not involve

, a significant hazards consideration, as defined in 10CFR50.92(c), since the change constitutes additional restrictions and controls that are not currently included in Technical Specifications and are being added to conform with requirements of 10CFR Part 50, Appendix I.

Fee Determination The proposed change contained herein is an amendment requested by the NRC to a previously submitted Proposal Change [ Reference (b)] and as such is not subject to any new fees. In addition, it is our position that the major l portion of this proposed change is an extension of the 10CFR Part 50, Appendix I design study submitted to the NRC on June 2,1976 (WVY 76-62), and constitutes completion of the requirements of Appendix I for the submittal of Technical Specifications. We conclude that this amendment should be exempt from any fees defined in 10CFR170.12(c) since fees were not applicable when

.the requirements put forth by Appendix I to 10CFR Part 50 became effective, and since submittal of this information has been delayed pending guidance from the NRC.

Schedule of Change

! These changes will be incorporated into the plant's Technical Specifications on the beginning of the calendar quarter which falls at least 180 days af ter the approval by the Commission of the entire RETS Program (including Technical Specifications, ODCM and PCP), but in no case before January 1, 1985. This implementation schedule reflects the work load implied by the magnitude and scope of these changes and the necessity to begin implementation of a number of changes at the beginning of an annual (calendar)

' period for the purpose of dose calculations, record keeping, and reporting re quirements. If approval from the Ccmmission to implement the RETS is not received prior to 180 days before January 1,1985, appropriate records shall

be maintained as of January 1,1985 so that upon formal implementation of the Technical Specifications af ter this date, all dose limits and reporting

.. requirements for the annual period can be addressed under the RETS requirements.

-United States Nuclear Regulatory Commission January 23, 1984 Attention: Mr. - Darrell G. Eisenhut, Directo r Page 4 We trust that this information is acceptable; however, should you have any questions regarding this matter, please contact us.

Very truly yours, VERM0?iT YANKEE NUCLEAR POWER CORPORATION h lw L. H. Heider Vice President MSS /cir cc: Vermont Department of Public Services 120 State Street Montpelier, VT 05602 Attention: Mr. Richard Saudek, Chaircan COMMONWEALTH OF MASSACHUSETTS)

)ss MIDDLESEX COUNTY )

Then personally appeared before me, L. H. Heider, who, being duly sworn, did state that he is a Vice President of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing document in the name and on the behalf of Vermont Yankee Nuclear Power Corporation and that the statements therein are true to the best of his knowledge and belief.

(f.B.'Sinclair Notary Public My Commission Expires June 1, 1984 spin xrit as..

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a ATTACHMENT A

/ Listing of Major Changes to '

Vermont Yankee Technical Specification

1. The following changes have been made to Section 1.0, Definitions:

New definitions have been added to the Technical Specifications for: i

1. Instrument Functional Test

11. Source Check iii. Dose Equivalent I-131 iv. Solidification

v. Members of the Public vi. Site Boundary vii. Radioactive Materials 4 viii. Contamination ix. Off-Site Dose Calculati, a Manual (ODCM)

x. Process Control Program (PCP) xi. Gaseous Radwaste Treatment System xii. Ventilation Exhaust Treatment System xiii. Purge / Venting These definitions have been added in accordance with the use of these terms in the new specifications and are consistent with the guidance given in NRC's model RETS (NUREG-0472/0473) with several exceptions as follows.

The definition of " Solidification" has been clarified to ensure that dewatered spent resins and filter sludges constitute suitable waste forms for disposal. Current plant design does not include solidification equipment which makes use of binders, such as cement. Due to the use of l filter /demineralizers, there are no liquid waste streams which require immobilization by a solidification binder.

The definition for " Member (s) of the Public" has been altered to include

" casual visitors" to the plant to account for those persons who visit the site because of specific interest in the plant. The statement which specifically includes persons who use portions of the site for recreational, occupational or other purposes not associated with the plant has been removed from the definition since no formal activities of these kind exist on the site within the boundary fence where the off-site limits for gaseous effluents apply (see FSAR Figure 2.2-5 Attached).

l I The definition for " Site Boundary" has been dropped from the RETS.

Reference is now made in Section 5.2 of the Technical Specifications to Fig. 2.2-5 in the FSAR for a description of the Site Boundary (See l

Attachment for reference to. FSAR Fig. 2.2-5 as planned to be amended in l the next revision to the FSAR).- This is in keeping with the format of l

the present Technical Specifications which only provides reference to the FSAR for other Figures called out in Section 5.

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New definitions for " Radioactive Material" and "Contatination" hevo be2n added to c12arly define what constitutes materials which require radiation materials handling control.

i Definitions for " Caseous Radwaste Treatment System" cnd " Ventilation I Exhaust Treatment System" have been clarified to make reference to those specific plants systems to which the use of these terms apply. In the case of the Gaseous Radwaste Treatment System, specific reference is made .

to the Augmented Off-Gas System (A0G), and for Ventilation Exhaust Treatment. System, reference is made to the A0G Building and Radwaste Building HEPA filter banks.

The definitions in the model RETS for " Purge-Purging" and " Venting" have been combined into a single definition of " Venting / Purge". This is to make the use of the term consistent with its application for primary containment venting operations and for continued purging for prolonged period during outages when the primary containment is open. During power

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operations the primary containment is inerted with nitrogen.

It should also be noted that no formal definition for " Applicability" has

[~ been added. It is the current practice in the existing Technical  !

Specifications that " Limiting Conditions for Operation" are assumed to be l applicable at all times unless.specifically stated otherwise. For ,

consistency in the Technical Specifications as a whole, this practice has been extended to the proposed RETS.

2. Table 3.2.4 and 4.2.4 of the existing Technical Specifications have been modified by dropping the steam jet air ejector (SJAE) radiation monitor, along with its isolation functions, from this section. Per NUREG-0473, a new release limit for the SJAE is imposed in Section 3.8.K. The new limitation on the SJ'.E requires corrective action if the release rate l

exceeds 0.16 Ci/sec. or approximately a factor of 2 below the action l 1evel required under the existing Table 3.2.4 requirements. However, the

! existing requirements in Table 3.2.4 force the air ejector suction valves

! to isolate the condenser on exceeding the release rate limit. Under the new Section 3.8.K, the plant has 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to bring the release rate within limits before shutting down, while immediate isolation of the high t radioactivity gas levels in the condenser is avoided. This now allows i radioactive gases to pass into the A0G charcoal absorber vessels for holdup and decay. The A0G system and structure in which it is housed have been designed and constructed to more stringent codes and standards

! than those in effect at the time of construction of the main condenser.

This specific Technical Specification change is deemed a safer action than the existing isolation criteria for the SJAE on Table 3.2.4. In the event of high radiation levels being detected in the main condenser off-gas, the gases are passed into the A0G which provides safer i

conditions for bulk radioactive gas holdup and decay. An additional

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limiting condition beyond those of NUREG-0473 has also been added (see Section 3.8.K.3) which requires the plant to be in Hot Standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the gross radioactivity release rate from the SJAE exceeds 1.5 Ci/sec. This condition requires that corrective actions be bsniemented in an expeditious but controlled manner to limit total air ejector releases to the ADG, and any subsequent releases to the environment, beyond release limit conditions as stated in Section 3.8.K.2.

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3. specific: tie:S 3.8/4.8, "Radioactiva Effluento" hava be:n amendsd in i their entirety to reflect NRC model RETS for effluent control.

4 Liquid effluents are now contained in specifications:

3/4.8.A " Concentration" 3/4.8.5 " Dose" 3/4.8.C "Radwaste Treatment" 3/4.8.D " Holdup Tanks" These new specifications insure that th. requirements of 10CFR50, including Appendix I for the "as low as is reasonably achievable" j criteria, are met with respect to liquid effluent concentrations, ,

resulting doses to members of the public from liquid discharges, the storage of liquid waste in outdoor tanks, sad the use of installed liquid  !

radwaste equipment. All references to requirements which call for the preparation of information to be in special reports or other unique

reports required by Specifications 3.8.B and 3.8.C have been removed in order to be consistent with the format of the balance of the existing Technical Specifications. This change from model RETS format does not l imply that the proposed Vermont Yankee RETS are dropping reporting requirements, only transferring the statement of these requirements to a single section found in Section 6.7 which is concerned with all reporting requirements for the entire Technical Specifications.

In addition, gaseous effluents are now contained in specifications.

3/4.8.E "Do s e Rat e" 3/4.8.F " Dose from Noble Gases" 3/4.8.G " Dose f rom 1-131, 1-133, Tritium, and Particulates 3/4.8.H " Gaseous Waste Treatment" 3/4.8.1 " Ventilation Exhaust Treatment" 3/4.8.J " Explosive Gas Mixtures" 3/4.8.K " Steam Jet Air Ejector" 3/4.8.L " Primary Containment" The contents of these specifications reflect the guidance put forth in the USNRC revised model RETS (Rav. 3 of NUkEG-0472/0473, Draf t 7"). They provide assurance that the requirements of 10CFR20,10CFR50, including j Appendix I, and 40CFR190, with respect to release rate limits and doses to members of the public resulting from plant gaseous ef fluents, as well as the integrity and use of the gaseous radwaste systems are met.

As with liquid effluents, all references to reporting requirements which call for the preparation of information to be in special reports, or other unique reports for Specifications 3.8.F, 3.8.G, 3.8.H. and 3.8.I have been removed in order to be consistent with the format of the balance of the existing Technical Specifications and placed in Section 6 .7.

Please note that Specification 3/4.8.H has been separated into two Specifications, 3/4.8.H and 3/4.8.I. The revised Specification 3/4.8.H relates to gaseous waste treatment of off-gas from the main condenser stema jet air ejector (SJAE) whenever it is ir operation. With the

Augmented Off-Gas System (AOC) out of service for more than seven days with the SJAE in operation, a special report will be filed with the NRC.

The new Specification 3.8.I addressed the use of the A0G and Radwaste Building ventilation filters whenaver certain estimated doses off-site are exceeded. These two specifications are consistent with NRC model RETS.

Also note that the requirement in Specification 4.8.L (Primary Containment) to determine appropriate valve alignment prior to containment venting / purge as called for in the model RETS is not necessary since existing procedural controla which are consistent with NUREG-0737,Section I.C.6 nn independent verification ensure proper valve line-up.

A new subsection 3.8.L.3 has been added so that once the primary containment is opened up for access af ter initial venting, continued purging over the duration of the outage will not require that the containment be purged through the Standby Gas Treatment System (SPGT) since at that time the air being vented is essentially Reactor Building air. This also reduces unnecessary service load on the SBGT and lengthens the time between when the system must be performance tested while still maintaining off-site doses within the ALARA criteria.

In Table 4.8.2, which is referenced by Specification 4.8.E, the tritium activity analysis required by the model RETS for a containment purge is not justified for Vermont Yankee's situation since it is estimated that the maximum H-3 concentration in containment air, assuming the air is saturated with reactor coolant at 1000F is less than the LLD value required by the model REIS.

In footnote b of Table 4.8.2, additional conditions have been added in order to clarify when 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I-131 sampling of the main plant stack is required. As stated in footnote b, sampling shall be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or thermal power change exceeding 25% of rated thermal power in one hour. In addition, footnote b states that the requirement to sample at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for 7 days applies only if: (1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3. However, at relatively low levels of dose equivalent I-131 concentration in the primary coolant, coolant concentration may fluctuate by more than a factor of 3 withcut indicating that a significant I-131 spike has occurred which could noticeably affect off-site dose calculations. We therefore have added the additional condition that the resultant dose equivalent I-131 concentration following an increase of more than a factor of 3 must be at least 1x10-1 uCi/ml. In addition, the change in thermal power by more than 15% of rated thermal power in one hour as a precondition for increasing sampling frequency is raised to 25% cf rated thermal power in one hour since current plant operations require a weekly power reduction of 20% in order to perform turbine valve testing. The above comments also apply to changes in footnote e of Table 4.8.2.

It should also be noted on Table 4.8.2 that no requirements for sampling

- and analysis of waste gas storage tanks has been included as called for by the model RETS since Vermont Yankee plant design does not include vaste gas storage tanks. For a boiling water reactor, waste gas from the main condenser system is collected via the condenser steam jet air ejectors. Table 4.8.2 requires sampling of these collected off-gases up stream of the AOG at the SJAE as well as down stream of the AOC at the plant stack.

Under containment purge requirements on Table 4.8.2, the standard model RETS footnotes which requires sampling and analysis following shutdown, startup or thermal power change exceeding 15 percent rated thermal power within one hour have not been included since Vermont Yankee's primary containment is N2 inerted. The actual mixture of radionuclides present in the primary containment are determined prior to the time of

, containment purge to the environment by Table 4.8.2 and therefore will account for any operational occurrences which could have altered the radionuclide mixture within the containment.

4 A new specification 3/4.8.M, " Total Dose" has been added.

The contents of this section insure that the requirements of 40CFR, Part 190 with respect to doses to member (s) of the public contributed from plant effluents are met. Due to the remote location of our station with respect to other uranium fuel cycle facilities, and the fact that the operation and control of other portions of the uranium fuel cycle are outside our license jurisdiction, conformance with the requirements of 40CFR, Part 190 as applied to Vermont Yankee are considered to be demonstrated by the licensee utilizing plant radioactive sources only.

The reporting requirements for conditions whica exceed the limits of 3.8.M are contained in Section 6.7.C.3 in the Administrative Section.

5. A new Specification 3/4.8.N, " Solid Radioactive Waste" has been added.

The contents of this specification reflect the guidance put forth in the USNRC revised model RETS. It ensures that a suitable Process Control Program will be utilized to meet shipping and burial ground requirements for solid waste.

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6. Specifications 3.9/4.9 " Radioactive Effluent Monitoring Systems" has been amended in its entirety to reflect NRC model RETS for effluent control.

Specification 3/4.9.A references liquid effluent instrumentation for the control of liquid waste discharges as detailed on Table 3.9.1 and 4.9.1.

It should be noted that on Table 3.9.1 the liquid radwaste discharge monitor, as well as the flow rate monitor, are only required to be in service during releases via those pathways, and that no automatic isolation of the discharge pathway is associated with the radioactivity monitor. These deviations from model RETS are based on the infrequent nature of Vermont Yankee's liquid release periods, which occur approximately once per year or less. Due to the limited number of liquid discharges and the administrative controls (such as duplicate sample analysis of batch waste prior to release and two independent verifications of proper valve positions and discharge rate settings)

I coacted to catura prcpir cpsraticn cf dicch rgas whsn thsy occur, the need to modify the plants instrumentation to provide auto isolation of the discharge line at all times is not warranted. In addition, appropriate plant procedures will be revised so that an operator will be present in the Radwaste Control Room whenever liquid radwaste discharges take place so that he can initiate isolation of the discharge line on

. indication of high activity alarm. However, the service water monitor will be required to be available at all times, other than for normal ,

maintenance and repair, since it provides information useful in identifying potential leaks into service water of primary fluids. The same justification holds for Tabic 4.9.1 which indicates the appropriate l time periods when Instrument checks need to be performed. Also for the liquid radwaste discharge monitor, the requirements fer Source Check include a condition that it shall not be performed "more than once each ,

month." This is justified on the grounds that periods of liquid release l occur infrequently, but when they do take place, they may be made up of several discrete batches released within several hours or days of each other. The initial Source Check is sufficient to check monitor response to radioactivity prior to the first release in a series, but subsequent batches in a closely spaced series of releases, along with a daily instrument check, are adequate to insure proper control of discharges.

It should also be noted on Table 4.9.1 that Source Check is "not applicable" for flow rate measurement devices on the liquid radwaste discharge line since an " Instrument Check" is already indicated to be made on a daily basis during periods of release. The source check conditions which are called for in the model RETS add nothing new or different to the condition; addressed under Instrument Check for our case of infrequent releases.

On Table 3.9.2, note 2, it should be noted that the conditions for ,

continued operation of the A0G "for period of up to 7 days provided that at least one of the stack monitoring systems is operable, and the off-gas system temperature and pressure are measured continuously" is an appropriate response to the short term lose of the A0G noble gas monitor. The A0G does not release directly to the environment, but directs its effluents to the plant stack which would require at least a single monitoring channel be operable before discharge to the environment, in addition, continuously monitoring A0G temperature and pressure provides indication of potential problems in A0G or possible source of high activity as might be detected at the stack.

Along these same lines, since the A0G does not constitute a direct release pathway to the environment, but is only an intermediate step before waste gas is discharged though the plant stack, it is not necessary to require that the A0G noble gas monitor itself be operable "st all times", but only during releases via this pathway. It should also be.noted that the A0G is normally out of service when the plant is shutdown, and thus does not constitute a pathway requiring active monitoring when the system is isolated from the plant.

Table 4.9.2 provides surveillance requirements for gaseous effluent instrumentation, including the plant stack monitoring system. This Table indicates that a stack iodine sampler and particulate sampler require weekly " Instrument Checks". Since these two samplers are only grab type composite collectors as opposed to monitors giving a continuous response, l

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thera io so cetual 1.strus=totion to psrform a "Sourco Check" cn othsr than to possibly record weekly if the filter paper or charcoal cartridge had actually been installed during the weekly change out. Since this function is performed weekly during sampler change out, nothing useful to the surveillance program is added by the model RETS requirements for

" Source Check" for the iodine and particulate samplers, but would in effect cause new paper work and record keeping to become necessary. This exception is taken from the model RETS.

On Table 4.9.2, the plant stack sampler flow integrator has no applicable

" Instrument Functional Test" for this type of instrument. The sampler flow integrator is a simple gas meter device with only local read out indicating total flow. There is no capability to inject a " signal" as required by the model RETS definition of " Channel Functional Test". The calibration of the device is equivalent to a functional test for this type of instrument.

Also on Table 4.9.2, Instrument Calibration and Instrument Functional Test requirements have been deleted for the Plant Stack System flow rate monitor since these types of surveillance requirements are not applicable to the type of equipment (Pitot tube) in use.

7. The new Specification 3/4.9.C includen a revised radiological environmental monitoring program as detailed on Table 3.9.3. It should be noted that the requirement for the air monitor sampling frequency has been extended from weekly to semi-monthly. Justification for this proposal is found in Reg. Guide 4.1 (C.2.b) which states in regard to sampling f requency that " sampling and analysis . .... should generally be carried out at intervals no greater than two or three half lives of the nuclide." For Iodine-131, operational experience over more than 10 years at Vermont Yankee has shown that environmental levels are orders of magnitude below the ALARA level, and that a semi-monthly sampling frequency would normally require only three half lives between start of sample collection and laboratory analysis. It should also be noted that no information will be lost by extending the sampling period since continuous samples will still be taken.

4 However, as an added precaution, sample collection will be performed weekly (as noted in footnote h of Table 3.9.3) whenever the main plant

' stack effluent release rate for I 131, as determined by the sampling and analysis program of Table 4.8.2, is equal to or greater than 1x10-1 to the required lower limit uCi/sec. This release rate value correspondg of detection air concentration of 0.07 pCi/m for 1-131 at the maximum predicted air monitoring station (See Bases section 3.9.C). Therefore.

l weekly sampling will be performed whenever there is a positive prospect of detecting it in the environment at the minimes required sensitivity.

Table 3.9.3 has also not included a composite river water sampler upstream of the plant, but relies on grab samples. This practice is sufficient since there are not known users of radioactive materials who discharge into the Connecticut River upstream of the plant, and because of the infrequent nature of liquid releases (once a year or less) from Vermont Yankee which make continuous composite sampling unnecessary.

A new river sediment sampling location at the outfall of the plant's north store drain has been added to the Radiological Environmental Monitoring Program (Table 3.9.3) in order to monitor the concentration and distribution of any radionuclides contained in rain runoff. This location is added to the Program beyond the minimum number required by the NRC's Branch Technical Position so that a particular site-specific phenomenon concerning washout of radioactivity from plant effluents can be studied. As such, the sampling and collection frequency is stated in Table 3.9.3 to be "as specified in the ODCH". The intended collection frequency is semiannually, but the declaration of this is placed in the 1 ODCM so that any future changes to storm drain system which could effect the north outfall can be expeditiously reflected in a change to the sampling frequency of sediments f rom this area.

i For milk sampling on Table 3.9.3, the control location for obtaining samples has not been tied to a specific location in a given distance or direction. Present practice for obtaining a control milk sample is to purchase processed milk from a regional dairy association where the milk from a number of farms has been diluted during processing with each other. The sample obtained in this way is believed to represent a true background condition for the region.

8. A new Specification 3/4.9.D, " Land Use Census" has been added. Included la this specification is the requirement to identify the nearest milk animal and resident in each sector. In addition, because of the plant stack creating an elevated release point, the census requires that the nearest milk animal (within 3 miles of the plant) to the point of predicted highest annual average D/Q value in each of three major meteorological sectors be identified. This action is appropriate in lieu of inventorying all milk animals in all directions out to 3 miles since the valley location of Vermont Yankee has demonstrated a strong annual meteorological persistence for winds up and down the valley. <

The requirement for garden census has not been included in the food product monitoring program due to the substantial and wide spreaa

- occurrence of dairy farming in the surrounding area which dominates the food uptake pathway. It should also be noted that predicted doses calculated at nearest residents, via requirements in the ODCH, conservatively assume the existence of vegetable gardens in the dose analysis.

( 9. A new Specification 3/4.9.E "Intercomparison Program" has been added.

This specification deals with quality control of laboratory analyses performed on samples collected as part of the Environmental Monitoring Program and is in accordance with the USNRC model RETS.

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10. Within Section 5, a reference has been added to a site map in the FSAR l Indicating site boundaries for radioactive ef fluents. This reference to r

the FSAR is consistent with the existing Section 5 Technical Specification which refers to the FSAR for all other figures. Since the FSAR is required to be updated on an annual basis, changes to site map will be made on a timely fashion, while it will not become necessary to make two updates, one in the FSAR and one in the Technical Specification, as would be the case if the site boundary map were to appear in Section 5.

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11. Is Section 6, "Administrativa Contralo"; new ri.tcrial which has besn added to the existing Technical Specifications which reflects the int;nt

• of the model RETS is indicated by vertical lines in the margins opposite the addition. The following major items should be noted with respect to the administrative controls:

a. Under " Responsibilities" of the Plant Operations Review Committee-PORC, the model RETS requirement to include statements concerning the review of any accidental, unplanned or uncontrolled radioactive releases has not been added to the proposed Vermont Yankee RETS since it is felt this requirement is already covered Likewise, the requirement to under items 6.2.A.6.e and 6.2.A.6.f. a review changes to the Process Control Program and Off-Site Dose Calculation Manual have not been added since it is believed this is to be covered under subsection 6.2.A.6.a

b. The model RETS requires that plant operating procedures include detailed written procedures for the " quality assurance program for ef fluent and environmental monitoring, using the guidance in Regulatory Guide 4.1, Revision 1, April 1975." This has not been included under the responsibilities of PORC (Subsection 6.5.A) since no other QA procedures that are part of the Operational Quality Assurance Program are required in the existing Technical Specifications for formal procedure review and approval by them.

Section 6.2.B.5 includes under the responsibilities of the Nuclear Safety Audit and Review Committee that at least once per 12 months, i

I the performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21, Rev. 1, June 1974, and Regulatory Guide 4.1, Rev.1, April 1975 will be determined. As such, implementation of the appropriate requirements of these QA programs will be assured without the need of formal procedures being under the control of PORC.

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