ML20205T472

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Marked-up Tech Specs Pages Re Suppl to 990201 Request for Amend to License DPR-28,revising Portions of Proposed Change 208
ML20205T472
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 04/19/1999
From:
VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20205T466 List:
References
NUDOCS 9904270261
Download: ML20205T472 (32)
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  • Regulatory Guide 1.109, Revision 1, October 1977.

DD. Solidification - Solidification shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. Suitable forms include dewatered resins and filter sludges. EE. Deleted FF. Site Boundary - The site boundary is shown in Figure 2.2-5 in the FSAR. GG. Deleted HH. Deleted II. Off-Site Dese Calculation Manual (ODCM) - A manual corataining the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduction of the environmental radiological monitoring program. JJ. Process Control Program (PCP) - A process control program shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste from liquid systems is assured to be converted to a form suitable for off-site disposal. KK. Gaseous Radwaste Treatment System - The Augmented Off-Gas System (AOG) is the gaseous radwaste treatment system which has been designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity ) j pri?r to release to the environment. l 110 LL. Ventilation Exhaust Treatment Svstem - The Radwaste Building and AOG N Building ventilation HEPA filters are ventilation exhaust treatment 

           $4          systems which have been designed and installed to reduce radioactive material in particulate form in gaseous effluents by passing              ,

O ventilation air through HEPA filters for the purpose of removing l [ radioactive particulates from the gaseous exhaust stream prior to release to the environment. Engineered safety fer.ture a'anospheric I cleanup systems, such as the Standby Gas Treatment (SBGT) System, are { not considered to be ventilation exhaust treatment system components. l 1

  • W "

MM. Vent / Purging - Vent / Purging is the controlled process of discharging air or gas from the primary containment to control temperature, l pressure, humidity, concentration or other operating conditions. l 5 NN. Core Ooerating Limits Report - The Core Operating Limits Report is the  ! unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating , limits shall be determined for each reload cycle in accordance with 6 Specification C.' .' O Plant operation within these operating limits is addressed in incivid specifications. Amendment No. 44, M4r M4,168 6.0.C "- 3 

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VYNPS l nt Health

5. Th diation Prote t or ion Supervisor exceed the qu or Pfications of sicist shall .8, Revision 1 (Se ember 1975). . /3] ,

egulatory Guid chelor's degree or  ! 

                        . The Shift nineer     a shall have a insering discipli d scientific    or with analysis equivalen        raining in plant de gn, and response ant for transients .d accidents.

specific of the l j the ope ions Supe ser does noterationsossess aSu sen r rvisor' 

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7. conse, the an Assistant erator hat does pos ss a Senior erator shall designate to the shift rews invol ng ated - -{/S]

Licen .f instructio then be ap ved by des icensed etivities sh Assist t operation upervisor. _ e operating staf~ i f 2.. . .: J The individuals who train qualityassuranh'functiosmay - l carry out health physic report to the appropria on-site manager; however, c eir IN 

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nizational freedom to ensurv [ shall have sufficient o j independence from opera ing pressures.  ! l I 1 N S QW l g pfAfd4M gfd l l l i 

       ~l 256 Amendment No. G , G , M , M , M . M , m , m , 168

     . . . . . . . .._ y _.m u n , -                     _ . _ _      . . . . _ . . . . . . .   .....s_.

(... VWPS

f. Investigat reported instances of vi lations of hh Technica Specifications, such inv tigations to includ reporting, evaluation, a recommendations to prev t recurrence, to the Man er of Operations.
g. rform special reviews and vestigations and render reports thereon as rer;uest by the Chairman of the l Nuclear Safety Audit and eview Committee.

l h. Review of the Fire Pr ection Program and implementing procedu s, and submittal of recommended I changes to the Nuc ar Safety Audit and Review l' Committee.

7. Authority
a. The Pitnt ration Review Committee shall be advisory,
b. The Pla Operation Review Committee sha' recot: mend J 1 to th Plant Manager approval or disapp val of I

prop als under Items 6 (a) through ( above. 1 In the event of disagreement tween the 

                   ,                               recommendations of the Plan Operation Review                                 :

Committee and the actions entemplated by the l Plant Manager, the cours determined by the i Plant Manager to be th more conservative will be followed with imm late notification to the Manager of Operatio .

c. The Plant Operation iew Committee shall make tentative detemina* ons as to whether or not ,

proposals consider d by the Committee involve ' unreviewed safet questions. This determination shall be subje to review by the Nuclear Safety Audit and Rev w Committee.

8. Records Minutes shal be kept at the plant of all meetings o the Plant Oper ion Review Cc=mittee and copies shall sent-to the Ma ger of Operations and the Nuclear Safe y Audit and Rev w Committee.
                                    ~

B. Nuclear Sa ety Audit and Review Committee

1. Th Committee shall consist of at least si (6) persons:
                                        . Chairman
b. Vice Chairman
c. Four technically qualified p sons who are not members of the plant staff.

l

d. No more than three membee shall be selected from the organization reporting o the Manager of Operations.
e. The Committee will o ain advice and counsel from scientific or techt cal personnel e= ployed by the l Ccmpany or other .ganizations whenever the Committee considers it nec ssary to obtain further scientific or technical a istance in carrying out its I responsibilit s.

Amendment No. M, 4M,16 8 259 

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e VYNPS 

                                                                                                                                  *{Gk Any repor           le occurrence shall                reported to the Mana            of Operatio          and shall be review              by the Plant operatic            Review Commit e. This Committee sh 1 prepare a separate, equentially numbe d, report for each r ortable occurrence.                                 ch report sha     describe the circums nces leading up to an resulting from th occurrence, the corr tive action taken by                               e shift, an tempt to define the e se of the occurrence, nd shall recomm .d ppropriate action to revent or reduce the p obability of a repetition of the oc rrence.

Copies of all sue reports shall be submi ted to the Chai n of the Nuclear Saf y Audit and Review Co . ttee for revie and to the Manager of erations for review a d approval of a recommendatic . 6 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED ,,,,,,, 

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Applies to administrative action to be followed in the event a 3 safety limit is exceeded. If a safety limit is exceeded, the_ reactor shall be shutdown immediate p An imme - ate report shall made to the Mana er of perar.1 .s. A comp 1 e analysis of the ircumstances lear

  • ng up to an resulting fr the situation to ether with recomm .dations by t Plant Opera ions Review i.'ommi ee shall also be repared.

Thi report shall be submitted to t M.tager of Oper ions and t Chairman of .he Nuclear Safety udi md Review =mittee. eactor opera .on shall not be r sumed .til authe 12ed by the U.S. Nuclear egulatory Commis on. 6 {L'2- Of Ris- Ng PROCEDtmES 

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A. De . led written ocedures, inv ving both nuclear nd n ..-nuclear saf y, including licable check-of' lists and nstructions, overing areas .sted below shall e prepared ,fg and approved ( .J All proc ures shall be hered to. Normal startup, operation and shutdown of systems and components of the facility, g ', h. Refueling operations. 

                               -[.          Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, suspected Primary System leaks and abnormal reactivity changes.
f. Emergency conditions involving potential or actual release of radioactivity.

h  % h. Preventive and corrective maintenance operations which could have an effect on the safety of the reactor. h i h. Surveillance and testing requirements. h  % h. Fire protection program implementation. A endment No. n , 6 . H , 1 H 262

VYNPS Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, 

                           " Supplemental Information to VYNPC April 19, 1990 Response Regarding FROSSTEY-2 Fuel Performance Code,"

BVY 90-054, dated May 10, 1990 (Approved by NRC SER, dated September 24, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, c,, 

                           " Responses to Request for Additional Information on           9 FROSSTEY-2 Fuel Performance Code," BVY 91-024, dated March 6,    1991 (Approved by NRC SER, dated September 24, 1992).

Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, "LOCA-Related Responses to Open Issues on FROSSTEY-2 Fuel fC Performance Code," BVY 92-39, dated March 27, 1992 yn[N (Approved by NRC SER, dated September 24, 1992). Letter from L. A. Tremblay, Jr. y'lY"EO, to USNRC, ' ' ' - {S( 

                           "FROSSTEY-2 Fuel Performance Code - vermont Yankee Response to Remaining Concerns," BVY 92-54, dated May 15, 1992 (Approved by NRC SER, dated September 24, 1992).

Report, " Loss-of-Coolant Accident Analysis for Vermont Yankee Nuclear Power Station," NEDO-2'.697, August 1977, as amended (Approved by NRC SER, dated November 30, 1977). Report, " General Electric Standard Application for Rasetor Fuel (GESTARII)," NEDE-240ll-P-A, GE Company Proprietary (the latest NRC-approved version will be listed in the COLR). Report, General Electric Nuclear Energy, "BWR Owner's Group Long-Term Solutions Licensing Methodology," NEDO-31960, June 1991 (Approved by NRC SER, dated July 12, 1993). Report, General Electric Nuclear Energy, "BWR Owner's Group Long-Term Solutions Licensing Methodology," NEDO-31960, Supplement 1, March 1992 (Approved by NRC SER, dated July 12, 1993). Report, N. Fujita, et al., " Method for Power / Flow Exclusion Region Calculation Using the LAPUR5 Computer Code," YAEC-1926-A (Approved by NRC SER, dated November 5, 1996). Report, Yankee Atomic Electric Company, " Application of the FIBWR2 Core Hydraulics Code to BWR Reload Analysis," YAEC-1339-A, January 31, 1997. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The COLR, including any mid-cycle revisions or supplements thereto, shall be provided ueen issuance, for each reload cycle, to the NRCfbecur^"* """*"01 Derk 4:i;;, ;rri r tr '.h; ncgi;nr1 3 % i;;.oLu and sniucai- 

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L Iuaywuiu l Amendment No, M, M, M, W, W, W, W, W,W 167 270

VYNPS B. eportable D6currence V f(.} This on del 

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(C./ Uniqu/ Reportird Requi/e - - 69] t , [_- ._D Radioactive Effluent Release Report 

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a. W hin yo day after January of each yea a g eport shal e submitted vering the r ioactive content of ffluents rol sed to unres icted areas during t previous ca ndar year of peration.
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l l l l l Amendment No. W 167 270a 

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e VYNPS e I e. IAndUseCensus,Ioecification3./D With a land u census not be g conducted required by ecification 3 .D. prepare a submit to the com ssion within 30 days a specia report 

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which ide ifies the reas s why the sur ey was not C+"~ * "*"*** '"* **** **** ***"' ** "* - correct he situation. I J Qnvirenmentadadiolooieag7 nit rin:p@pewgrNs ,grAmer)-- (r.2) 

                                          "- ** rl:;ic;1 2n;irr cr : 1 E 8                                                             ri;1:n:: ==rrrt ;/ covering' %63) the operation ~or the unit _during previous calendar year shall be submitted @rirr :: :N % of each year.                                   ,

I My ma ic)-~ ~ The 5 .u:1 ". di;1:;i :1 Cn~;ir;.-_ :nt:2 Er crill:nrO 

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port shall include summaries, interpretations, _and an h 5] 7 analysis of trends of the results of the radiological environmental surveillance activities for the repcrt periogjinc' ding a compari n with operati .a1 controls  ; p(a n w ro inte), and pre .ous environme

  • 1 urveil nee reports ar an assessment gg1 the observed L J impac of the plant erarien m *ka nvironment.

Th .nual Radiological Environmental [Erren mer ~ gpgAATN i ort shall include summarized and tabulated results of QG7] all radiological environmental samples taken during the report period pursuant to the table and figuies in the i ODCM. 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.

I With the 1 el of radioacti .ty in an environment sampling,.edia at one or in Tab 3.9.3 exceedin re of the location

  • the reporting levels ecified
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2 Table .9.4, the condi son shall be describe in the nex* annual Radiologd al Environmental Surv 111ance Re ort only if the easured level of radi etivity was t the result of lar.t ef fluents . Wit the adiclogical env'ronmental monitoring .ogram not being conducted as s cified in Table 3.9.3 a description of the reasons f r not conducting the .ogram as required / and the plar for preventing a rec rrence shall be included i the next annual Radi .ogical Environmental Surveill ce Report. Amendment No. H M,168 25 

                                                                                 )

VERMONT YANKEE NUCLEAR POWER CORPORATION Docket No. 50-271 BW 99-54 1 i Attachment D Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 208 Section 6 - Administrative Controls Revised Technical Specifications Pages (The attached revised Technical Specifications pages completely replace those previously provided as part of BW 99-20, dated February 1,1999.)

7-. VYNPS TABLE OF CONTENTS GENERAL Page No. 1.0 DEFINITIONS..................................................... 1 5.0 DESIGN FEATURES....'.......................................... . 253 6.0 ADMINISTRATIVE CONTROLS........................................ 255 6.1 . RESPONSIBILITY.................................................. 255 6.2 ORGANIZATION.................................................... 255 6.31 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED............... 257 

    '6.4      .

PROCEDURES..................................................... 257 

   . 6. 5 . HIGH RADIATION AREA............................................         257 6.6        REPORTING REQUIREMENTS......................................... 258 6.7        PROGRAMS AND MANUALS........................................... 262 l

l 1 4 I i 1 J l l l l 

                                                -i-Amendment No. R, %

l VYNPS 1.0 DEFINITIONS factors used for this calculation shall be those listed in NRC l Regulatory Guide 1.109, Revision 1, October 1977 1 DD Solidification - Solidification shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. Suitable forms include dewatered resins and filter sludges. EE. Deleted l FF. Site Boundary - The site boundary is shown in Figure 2.2-5 in the FSAR. GG. Deleted HH. Deleted II. Off-Site Dose Calculation Manual (ODCM) - A manual containing the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduction of the environmental radiological monitoring program. l JJ. Process Control Program (PCP) - A process control program shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste from liquid systems is acsured to be converted to a form suitable for off-site disposal. KK. Gaseous Radwaste Treatment System - The Augmented Off-Gas System (AOG) is the gaseous radwaste treatment system which has been designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. LL. Ventilation Exhaust Treatment System - The Radwaste Building and AOG Building ventilation HEPA filters are ventilation exhaust treatment systems which have been designed and installed to reduce radioactive material in particulate form in gaseous effluents by passing ventilation air through HEPA filters for the purpose of removing radioactive particulates from the gaseous exhaust stream prior to release to the environment. Engineered safety feature atmospheric cleanup systems, such as the Standby Gas Treatment (SBGT) System, are I not considered to be ventilation exhaust treatment system components. j i MM. Vent / Purging - Vent / Purging is the controlled process of discharging j air or gas from the primary containment to control temperature, j pressure, humidity, concentration or other operating conditions. ' NN. Core Operating Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with l Specification 6.6.C. Plant operation within these operating limits is addressed in individual specifications. i Amendment No. 43, 4Mr 4M, MB 5

m VYNPS TABLE 3.2.6 NOTES Note 1 - From and after the date that a parameter is reduced to one indication, operation is permissible for 30 days. If a parameter is not indicated in the Control Room, continued operation is permissible during the next seven days. If indication cannot be restored within the next six hours, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following 18 hours. Note 2 - Deleted. Note 3 - From and after the date that this parameter is reduced to one indication in_the Control haom, continued reactor operation is permissible during the next 0 days. If both channels are inoperable and indication cannot be restored in six hours, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following 18 hours. Note 4 - From and after the date that safety / relief valve position from pressure switches is unavailable, reactor operation may continue provided safety / relief valve position can be determined from Recorder #2-166 (steam temperature in SRVs, 0-600*F) and Meter 16-19-33A or C (torus water temperature, 0-250*F) . If both parameters are not available, the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following 18 hours. Note 5 --From and after the date that safety valve position from the acoustic monitor is unavailable, reactor operation may continue provided safety valve position can be determined from Recorder #2-166 (thermocouple, 0-600*F) and Meter #16-19-12A or B (containment pressure (-15) - (+260) psig). If both indications are not available, the reactor shall be in a hot shutdown condition in six hours and in a cold shutdown condition in the following 18 hours. Note 6 - Within 30 days following the loss of one indication, or seven days following the loss of both indications, restore the inoperable channel (s) to an operable status or a special report to the Commission must be prepared and submitted within the subsequent 14 days, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status. Note 7 - From and after the date that this parameter is unavailable by Control Room indication, within 72 hours ensure that local sampling capability is available. If the Control Room indication is not restored within 7 days, prepare and submit a special report to the NRC within 14 days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status. Amendment No. 60, 43, M, M, 444, 4M, 44, MB 55

7 VYNPS BASES: 3.6 and 4.6 (Cont'd)- J. Thermal Hydraulic Stability The reactor design criteria is such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without i exceeding specified fuel design limits. To minimize the likelihood of an instability,.a power / flow exclusion region to'be avoided during normal operation is calculated using the approved methodology as stated in Specification 6.6.C. Since the. exclusion region may change'each fuel cycle, the limits are contained in the Core operating Limits Report. Specific directions are provided to avoid operation in this region and to immediately exit upon.an entry. Entries into the-exclusion region are not part of normal operation. An entry mayLoccur as a result of an abnormal event, such as a single recirculation pump trip. In these events, operation in the' exclusion region may be needed to prevent equipment deras;2, but actual time spent inside the exclusion region is minimized. Though each operator action can prevent the occurrence and protect-the reactor from an instability, the APRM flow-biased scram function is designed to suppress global oscillations, the most likely mode of oscillation, prior to exceeding the. fuel safety limit. While global oscillat' ,./ are the most likely mode, protection from out-of-phase oscillations a' provided through avoidance of the exclusion region and administrk ,1e controls on reactor conditions which are primary factors affecting reactor stability l l l i l l l l l i I l Amendment No. - 444 14Sa i

VYNPS l TABLE 4.8.1 NOTATION: (Cont'd) l

c. A composite sample is one in wnich the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the 4 method of sampling employed results in a specimen which is representative )

of the liquids released. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release,

d. 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, Ce-141, and Ce-144. This list does not mean i that only these nuclides are to be detected and reported. Other peaks l which are measurable and identifiable, together with the above nuclides, i 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, ) but as "not detected". When unusual circumstances result in LLDs higher ' than required, the reasons shall be documented in the Radioactive Effluent  ! Release Report. I J Amendment No. &B, 444 182

VYNPS TABLE 4.8.2 NOTATION:

a. See footnote a. of Table 4.8.1.
b. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after removal from samplers. Sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or thermal power change exceeding 25% of rated thermal power in one hour, and analyses shall be completed within 48 hours of changing the samples. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement to sample at least once per 24 hours 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 the resultant ccncentration is at least 1 x 104 pCi/ml; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.
c. Sampling and analyses shall also be performed following shutdown, startup, or a thermal power change exceeding 25% of rated thermal power per hour unless: (a) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3 and the resultant concentration is at least 1 x 104 pCi/ml; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
d. The principal gamma emitters for which the LLD specifice ton will apply are exclusively the following radionuclides: Kr-87, Kr-88 ta- 133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions, and Mn-54, Fe-39, Co-58, Co-60, Zn-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. Nu;11 des whien are below LLD for the analyses should not be reported as being present at the LLD level for that nuclide, but as "not detected". When unusual circumstances result in LLDs higher than required, the reasons shall be l documented in the Radioactive Effluent Release Report.

e. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specification
  • 3.8.E.1, 3.8.F.1 and 3.8.G.I.
f. The gaseous waste samplin? and analysis program does not explicitly require sampling and analysis at c specified LLD to determine the I-133 release.

Estimates of I-133 releases shall be determined by counting the weekly charcoal sample for I-133 (as well as I-131) and assume a constant release rate for the release period.

g. Lower Limit of Detection (LLD) applies only to particulate form radionuclides identified in Table Notation d. above.

Amendment No. 43, 444, 4&& 184 i

VYNPS , 3.9 LIMITING CONDITIONS FOR 4.9 SURVEILLANCE REQUIREMENTS OPERATION C. Radiological Environmental C. Radiological Environmental Monitoring Program Monitoring Program

1. The radiological 1. The radiological environmental monitoring environmental monitoring program shs11 be samples shall be conducted as specified in collected pursuant to Table 3.9.3. Table 3.9.3 from the locations given in the ODCM and shall'be analyzed pursuant to the requirements of Table 3.9.3 and the detection capabilities required by Table 4.9.3.

D. Land Use Cc.tsus D. Land Use Census

1. A land use census shall 1. The land use census shall be conducted to identify be conducted at least the location of the once per year between the nearest milk animal and dates of June 1 and the nearest residence in October 1 by either a each of the door-to-door _ survey, 16 meteorological sectors aerial survey, or by within a dintance of five consulting local miles. The survey shall agricultural authorities. -

also identify the nearest The results of the land milk animal (within use census shall be 3 miles of the plant) to included in the Annual the point of predicted Radiological highest annual average Environmental Operating D/Q value in each of the Report pursuant to three major Specification 6.6.E. meteorological sectors due to elevated releases from the plant stack. g

2. With a land use census identifying one or more locations which yield a calculated dose or dose commitment (via the same exposure pathway) at least 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.9.C.1, .I add the new location (s) to the radiological j environmental monitoring program within 30 days if  ;

permission from the owner  ; to collect samples can be obtained, and sufficient sample volume is available. The sampling location (s), excluding the control Amendment No. 84 191

VYNPS 3.9 LIMITING CONDITIONS FOR 4.9 SURVEILLANCE REQUIREMENTS OPERATION station 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 1and use census was , conducted. E. Intercomparison Program E. Intercomparison Program l j

1. Analyses shall be 1. A summary of the results 2 performed on referenced of analyses performed as radioactive materials part of the above supplied as part of an required Intercomparison Intercomparison Program Program shall be included which has been approved in the Annual i by NRC. Radiological l Environmental Operating l )

Report. The identification of the NRC approved Intercomparison Program which is being part( :ipated in shall be i stated in the ODCM. 1 i l 1 i l l l 1 I Amendment No. 84 192

VYNPS TABLE 3.9.3 NOTATION a Specific parameters of distance and direction sector from the centerline of the reactor and additional descriptions where pertinent, shall be provided for each and every sample location in Table 3.9.3 in a table and figure (s) in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every reasonable effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmente.1 Operating Report pursuant to Specification 6.6.2. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.6.D, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s) . b One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a Thermoluminescent Dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading. c Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is grcater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. d Gamma isotopic analyris means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. e The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone. f Composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. 9 Each meteorological sector shall have an established " inner" and an " outer" monitoring location based on ease of recovery (i.e., response time) and year-round accessibility. h Sample collection will be performed weekly whenever the main plant stack effluent release rate of I-131, as determined by the sampling and analysis program of Table 4.8.2, is equal to or greater than 1 x 10~1 uCi/sec. Sample collection will revert back to semimonthly no sooner than at least two weeks after the plant stack effluent release rate of I-131 falls and remains below 1 x 10-1 uCi/sec. Amendment No. 83, 444 201

VYNPS TABLE 4.9.3 NOTATION (a) See Footnote (a) of Table 4.8.1. (b) Parent only. (c) 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. (d) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the listed nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.6.E. (e) The Ba-140 LLD and concentration can be determined by the analysis of its short-lived daughter product La-140 subsequent to an 8 day period following collection. The calculation shall be predicted on the normal ingrowth equations 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 assume that the supported La-140 activity at the time of the collection is zero. (f) Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD, but as "not detected". For purposes of J averaging, the LLD will be assumed to be zero. 1 Amendment No. 83 208

i VYNPS BASES: 3.11 FUEL RODS 

            -A. Average Planar Linear Heat Generation Rate QPLHGR) l            Refer to the appropriate topical reports listed in Specification 6.6.C for analyses methods.

(Note: All exposure increments in this Technical Specification section are expressed in terms of megawatt-days per short ton.) The MAPLHG3 reduction factor for single recirculation loop operation is based on the assumption that the coastdown flow from the unbroken recirculation loop would not be available during a postulated large break-in the active recirculation loop. See Core Operating Limits Report for the cycle-specific reduction factor. B. Linear Heat Generation Rate (LHGR) l Refer to the appropriate topical reports listed in Specification 6.6.C for analyses methods. C. Minimum Critical Power Ratio (MCPR) Operating Limit MCPR l

1. The MCPR operating limit is a cycle-dependent parameter which can be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Cladding Integrity Safety Limit (FClSL) for potential abnormal occurrences. The MCPR operating limits are justified by the analyses, the results of which are presented in the current cycle's Supplemental Reload Licensing Report. Refer to the appropriate topical reports listed l in Specification 6.6.C for analysis methods. The increase in MCPR operating limits for single loop operation accounts for increased core flow measurement and TIP reading uncertainties.

i Amendment No. M, 4-7, M, M, 44, M, MG, M6, MG, S"Y 99-55 227 

                                                                                        ~

i VYMPS 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY A. The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during absences. . B. The Plant Manager or designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect nuclear safety. C. The Shift Supervisor shall be responsible for the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in plant startup or normal operation, an individual with an active Senior Reactor Operator I (SRO) license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in cold shutdown or refueling with fuel in the reactor, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. 6.2 ORGANIZATION A. Onsite and Offsite Organizations Organizations shall be established for unit operation and corporate management. These organizations shall include the positions for activities affecting safety of the nuclear power plant.

1. Lines of authority, responsibility, and com*unication m shall be established and defined for the highest management levels through intermediate levels to and including all operating organizational positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be l documented in the Vermont Yankee Operational Quality Assurance Manual.

l 2. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those on-site activities necessary for safe operation and maintenance of the plant.

3. The corporate executive with direct responsibility for the plant shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

Amendment No. 66, M, 4M 255 l

i 7, VYMPS 6.2 ORGANIZATION (Cont' d)

4. The individuals who train'the operating staff, carry out
                  ' health physics, or perform quality assurance functions may report to the appropriate on-site manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

B. Unit Staff The unit staff organization shall include the following:

1. A non-licensed operator shall be assigned when the reactor contains fuel and an additional non-licensed operator shall be assigned during Plant Startup and Normal Operation.
2. At least one licensed Reactor Operator (RO) or one licensed Senior Reactor Operator (SRO) shall be present in the control room when fuel is in the reacter.
3. When the unit is in Plant Startup or. Normal Operation, at least one licensed Senior Reactor Operator (SRO) and one licensed Reac'.or Operator (RO), or two licensed Senior Reactor Operators, shall be present in the control room.
4. Shift crew composition shall meet the requirements stipulated herein and in 10 CFR 50.54(m). Shift crew composition may be less than the minimum requirement of 10 CFR 50.54 (m) (2) (i) and Specifications 6.2.B.1 and 6.2.B.8 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift' crew members, provided immediate action is taken to restore the shift crew composition to within the minim 9m requirements.
5. An individual qualified in radiation protection procedures shall be present on-site when there is fuel in the reactor.

The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

6. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, radiation protection technicians, auxiliary operators, and key maintenance personnel).
7. The operations manager or an assistant operations manager chall hold an SRO license.
8. While the unit is in Plant Startup or Normal Operation, the Shift Engineer shall provide advisory technical support to the Shift Supervisor (SS).

I Amendment No. G, 4, M, M, M, M1, 4-M, MB 256 3

VYNPS 6.2 ORGANIf,ATION (Cont'd) C. f_,it Staff Qualifications Each member of the unit staff shall meet or exceed the minimum qualifications of the American National Standards Institute N-18.1-1971, " Selection and Training of Personnel for Nuclear Power Plants," except for the radiation protection manager who shall meet the qualifications of Regulatory Guide 1.8, Revision 1 (September 1975) and the Shift Engineer, who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS F.XCEEDED Applies to administrative action to be followed in the event a safety limit is exceeded. If a safety limit is exceeded, the reactor shall be shutdown immediately. 6.4 PROCEDURES Written procedures shall be estsblished, implemented, and maintained covering the following activitie s: A. Normal startup, operation and shutdown of systems and components of the facility. B. Refueling operations. C. Actions to be taken to correct specific and fereseen potential malfunctions of systems or compor.ents, suspected Primary System leaks and abnormal reactivity changes. D. Emergency conditions involving potential or actual release of radioactivity. E. Preventive and corrective maintenance operations which could have an effect on the safety of the reactor. l F. Surveillance and testing requirements. ] G. Fire protection program implementation. . H. Process Control Program in-plant implementation. I. Off-Site Dose Calculation Manual in-plant implementation. 6.5 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraphs 20.1601(a) and 20.1601(b) of 10 CFR 20: Amendment No. M, 44, 43, 84, -1M , M B 257 i i

VYMPS A. Paragraph 20.1601, " Control of Access to High Rad,iation Areas. In lieu of the " control device" or " alarm signal" required by Paragraph 20.1601(a), each high radiation area in which the intensity of radiation is greater than 100 mrem /hr at 30 cm, but less than 1000 mrem /hr at 30 cm, 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 (RWP) . Radiation Protection personnel qualified in radiation protection procedures (e.g., radiation protection technicians) may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas, provided they are othervise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:

1. A radiation monitoring device which continuously indicates 1 the radiation dose rate in the area.
2. A radiation monitoring device which continuously integrates I 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.
3. A Radiation Protection individual qualified in radiation protection procedures (e.g., radiation protection technicians) with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and who will perform direct or remote (such as closed circuit TV cameras) periodic I radiation surveillance at the frequency specified in the RWP. The surveillance frequency will be established by the Radiation Protection Manager.

B. The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr at 30 cm, but less than 500 rad /hr at 1 meter. In addition, locked or continuously guarded entryways shall be provided to I 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 Radiation Protection Manager. 6.6 REPORTING REQUIREMENTS The following reports shall be submitted in accordance with 10 CFR 50.4. A. Occupational Radiation Exposure Report An annual rr. port covering the previous calendar year shall be submitted prior to April 30 of each year. The annual report shall l Amendment No. 44, 44 258 ) l l l 1

VYMPS include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, 1/ e.g., reactor operations and surveillance, inservice inspection, routine snaintenance, special maintenance (describe maintenance), waste , processing, and refueling. l The dose assignment to various duty functions may be estimates based on Self-Reading Dosimeter (SRD), TLD or film badge , measurement. Small exposures totaling less than 20% of the l individual total dose need not be accounted for. In the  ! aggregate, at least 80% of the total whole body dose received from l external sources should be assigned to specific major work l functions. B. Monthly Operating Reports 

                                                                                       ]

Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the fifteenth I of each month following the calendar month covered by the report. The.se reports shall include a narrative summary of operating experience during the report period which describes the operation of the facility. C. Core Operating Limits Report l The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

1. The Average Planar Linear Heet Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la,
2. The Kr core flow adjustment factor for Specification 3.11.C.,
3. The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.la,
4. The Linear Heat Generation Rates (LHGR) for Specifications 2.1.A.la, 2.1.B.1, and 3.11.B, and
5. The Power / Flow Exclusion Region for Specifications 3.6.J.la and 3.6.J.lb.

The analytical methods used to determine the core, operating limits shall be those previously reviewed and approved by the NRC in: Report, E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982). 1 

     / This tabulation supplements the requirements of 20.2206 of 10 CFR Part 20.

Amendment No. 42, 444, 444, M 259

VYNPS Report, D. M. VerPlanck, " Methods for the Analysis of Boiling Water Reactors Steady State Core Physics," YAEC-1238, March 1981 (Approved by NRC, SER, dated September 15, 1982). Report, J. M. Holzer, " Methods for the Analysis of Boiling Water Reactors Transient Core Physics," YAEC-1239P, August 1981 (Approved by NRC SER, dated September 15, 1982). Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Guide Fuel Rod Steady-State Thermal Effects (FROSSTEY): 

          -Code /Model Description Manual," YAEC-1249P, April 1981 (Approved by NRC SER, dated September 27, 1985).

Report, A. A. F. Ansari, " Methods for tha Analysis of Boiling Water Reactors: Steady-State Core Flow Distribution Code (FIBWR) , " YAEC-12 34, December 1980 (Approved by NRC SER, dated September 15, 1982). Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Oxide Fuel Rod Steady-State. Thermal Effects (FROSSTEY): Code Qualification and Application," YAEC-1265P, June 1981 (Approved by NRC SER, dated September 27, 1985). Report, A. A. F. Ansari and J. T. Cronin, " Methods for the Analysis of Boiling Water Reactors: A System Transient Analysis Model (RETRAN)," YAEC-1233, April 1981. (Approved by NRC SERs, dated November 27, _1981 and September 4, 1984). Report, A. A. F. Ansari, K. J. Burns and D. K. Beller, " Methods for the Analysis of Boiling Water Reactors: Transient Critical Power Ratio Analysis (RETRAN-TCPYA01)," YAEC-1299P, March 1982 (Approved by NRC SER, dated September 15, 1982). Report, A. S. DiGiovine, et al., "CASMO-3G Validation," YAEC-1363-A, April 1988. Report, A. S. DiGiovine, J. P. Gorski, and M. A. Tremblay, 

           " SIMULATE-3 Validation and Verification," YAEC-1659-A, September 1988.

Report, R. A. Woehlke, et al., 

           "MICBURN-3/CASMO-3/ TABLES-3/ SIMULATE-3 Benchmarking of Vermont Yankee Cycles 9 through 13," YAEC-1683-A, March 1989.

Report, J. T. Cronin, " Method for Generation of One-Dimensional Kinetics Data for RETRAN-02," YAEC-1694-A, June 1989. Report, V. Chandola, M. P. LeFrancois, and J. D. Robichaud, 

           " Application of One-Dimensional Kinetics to Boiling Water Reactor Transient Analysis Methods," YAEC-1693-A, Revision 1, November 1989.

Report, L. H. Steves, et. al, "HUXY: A Generalized Multirod Heatup Code with 10CFR50, Appendix K Heatup Option: User's Manual," XN-CC-33(A), Revision 1, dated November 14, 1973 (Approved by NRC SER, dated March 6, 1975). Amendment No. 444, 4G4, 446 260

r VYNPS Report, "RELAPSYA, A Computer Program for Light-Water Reactor System Thermal-Hydraulic Analysis," YAEC-1300P, October 1982 (Approved by NRC SERs, dated August 25, 1987 and October 21, 1992). Report, R. T. Fernandez and H. C. daSilva, Jr., " Vermont Yankee BWR Loss-of-Coolant Accident Licensing Analysis Method," YAEC-1547, June 1986 (Approved by NRC SER, dated October 21, 1992). Letter from R. W. Capstick (VYNPC) to USNRC, "HUXY Computer Code information for the Vermont Yankee BWR LOCA Licensing Analysis Method," FVY 87-63, dated June 4, 1987 (Approved by NRC SER, dated Feoruary 27, 1991). Letter from R. W. Capstick (VYNPC) to USNRC, " Request for Supplemental Safety Evaluation Report Supporting the Use of I RELAPSYA for Vermont Yankee Nuclear Power Station," FVY 88-006, dated January 26, 1988 (Approved by NRC SERs, dated February 27, j 1991 and October 21, 1992). I 1 Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Supplementary Information Regarding NRC LOCA Analysis Review Effort," BVY 89-91, 4 dated October 6, 1989 (Approved by NRC SER, dated October 21, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Supplementary Information Regarding NRC LOCA Analyses Review Effort," BVY 90- I 028, dated March 9, 1990 (Approved by NRC SER, dated October 21, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Response to Second Request for Additional Information on the Use of RELAP5YA," BVY 90-067, dated June 8, 1990 (Aporoved by NRC SER, dated February 27, 1991). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Response to Request for Additional Information on the Use of RELAP5YA," BVY 90-087, dated August 28, 1990 (Approved by NRC SER, dated October 21, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Response to Second Request for Additional Information on the Use of RELAPSYA," BVY 91-05, dated January 9, 1991 (Approved by NRC SER, dated October 21, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Response to l Third Request for Additional Information on the Use of RELAPSYA," l BVY 91-41, dated April 19, 1991 (Approved by NRC SER, dated j October 21, 1992). i Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Supplementary  ; Information Regarding the Use of RELAP5YA," BVY 92-12, dated February 7, 1992 (Approved by NRC SER, dated October 21, 1992). Amendment No. 4Fs 261

l l VYNPS , I l I Letter from R. W. Capstick (VYNPC) to USNRC, " Vermont Yankee LOCA Analysis Method FROSSTEY Fuel Performance Code (FROSSTEY-2)," FVY 87-116, dated December 16, 1987 (Approved by NRC SER, dated September 24, 1992). 1 Letter from R. W. Capstick (VYNPC) to USNRC, " Response to NRC Request for Additional Information on the FROSSTEY-2 Fuel l Performance Code," BVY 89-65, dated July 14, 1989 (Approved by NRC i SER, dated September 24, 1992). l l Letter from R. W. Capstick (VYNPC) to USNRC, " Supplemental l Information on the FROSSTEY-2 Fuel Performance Code," BVY 39-74, l dated August 4, 1989 (Approved by NRC SER, dated September 24, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Responses to Request for Additional Information on FROSSTEY-2 Fuel Performance l Code," BVY 90-045, dated April 19, 1990 (Approved by NRC SER, ) dated September 24, 1992). 1 Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Supplemental Information to VYNPC April 19, 1990 Response Regarding FROSSTEY-2  ! Fuel Performance Code," BVY 90-054, dated May 10, 1990 (Approved by NRC SER, dated September 24, 1992). . l Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, " Responses to Request for Additional Information on FROSSTEY-2 Fuel Performance Code," BVY 91-024, dated March 6, 1991 (Approved by NRC SER, dated September 24, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, "LOCA-Related Responses to Open Issues on FROSSTEY-2 Fuel Performance Code," BVY 92-39, dated March 27, 1992 (Approved by NRC SER, dated September 24, 1992). Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC,."FROSSTEY-2 Fuel I Performance Code - Vermont Yankee Response to Remaining Concerns," BVY 92-54, dated May 15, 1992 (Approved by NRC SER, dated September 24, 1992). Report, " Loss-of-Coolant Accident Analysis for Vermont Yankee Nuclear Power Station," NEDO-21697, August 1977, as amended (Approved by NRC SER, dated November 30, 1977). Report, " General Electric Standard Application for Reactor Fuel (GESTARII)," NEDE-24011-P-A, GE Company Proprietary (the latest NRC-approved version will be listed in the COLR). Report, General Electric Nuclear Energy, "BWR Owner's Group Long-Term Solutions Licensing Methodology," NEDO-31960, June 1991 (Approved by NRC SER, dated July 12, 1993). Report, General Electric Nuclear Energy, "BWR Owner's Group Long-Term Solutions Licensing Methodology," NEDO-31960, Supplement 1, March 1992 (Approved by NRC SER, dated July 12, 1993). Amendment No. 444, 446, m, 446 262

VYMPS Report, N. Fujita, et al., " Method for Power / Flow Exclusion Region Calculation Using the LAPURS Computer Code," YAEC-1926-A (Approved by NRC SER, dated November 5, 1996). Report, Yankee Atomic Electric Company, " Application of the FIBWR2 Core Hydraulics Code to BWR Reload Analysis," YAEC-1339-A, January 31, 1997. The core operating limits shall be determined so that all applicable limits (e.g., . fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the-safety analysis are met. The COLR, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC. D. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted by May 15 of each year and in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix 1, Section IV.B.l. E. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering j the operation of the unit during the previous calendar year shall  ! l be submitted by May 15 of each year. The report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, . IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall j include summarized and. tabulated results of all radiological environmental samples taken during the report period pursuant to the table and figures in the ODCM. In the event that some results are not available for inclusion with the report,.the report shall i 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. 6.7 PROGRAMS AND MANUALS , The following programs shall be established, implemented and maintained: A. INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT A program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels will be Amendment No. G, M, M6, M6, 444, M4, M4, MB 263

VYNPS l l implemented. This program shall include the following: I

1. Provisions establishing preventive maintenance and periodic visual inspection requirements.

l

2. System leakage inspections, to the extent permitted by system design and radiological conditions, for each system at a frequency not to exceed refueling cycle intervals. The systems subject to this testing are: (1) Residual Feat Removal, (2) Core Spray, (3) Reactor Water Cleanup, (4) HPCI, (5) RCIC, and (6) Sampling Systems.

B. OFF-SITE DOSE CALCULATION MANUAL (ODCM) . An Off-Site Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents for the i purpose of demonstrating compliance with 10 CFR 50, Appendix I,.in

l. the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program.

The ODCM shall also contain the radioactive effluent controls and I radiological environmental monitoring activities and descriptions of the information that should be included in the Radioactive Effluent Release Report and the Annual Radiological Environmental Operating Report required by Specification 6.6.D and Specification 

             !6.6.E, respectively.
1. Licensee initiated changes to the ODCM: .

I

a. Shall be submitted to the Commission in the Radioactive l Effluent Release Report for-the period in which the 1 change (s) was made effective. This submittal shall 4 contain:

l 1. Sufficient information to support the change together with appropriate analyses,or-evaluations justifying the change (s) and

11. A determination that the change will maintain the l l level of radioactive effluent control required by l 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50, and do not adversely ,

impact the accuracy or reliability of effluent dose l or setpoint calculations.

b. Shall become effective upon review by PORC and approved by the Plant Manager,
c. Shall be submitted to the Commission in the form of a legible copy of the affected pages of the ODCM as a part of or conce rent with the Radioactive Effluent Release Amendment No. 43, 83, 443, 444, 444 264

VYNPS Report for the period of the report in which any change to the ODCM was made. Each change shall be identified l by markings in the margin of the affected pages, clearly I indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented. C. PRIMARY CONTAINMENT LEAK RATE TESTING PROGRAM A program shall be established to implement the leak rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled " Performance Based Containment Leak-Test Program," dated September 1995. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44 psig. The maximum allowable primary containment leak rate, La, at Pa, shall be 0.8% of primary containment air weight per day. , l Leak rate acceptance criteria are: j

1. Primary containment leak rate acceptance criterion 5 1.0 La.
2. The as-left primary containment integrated leak rate test (Type A test) acceptance criterion is 5 0.75 La.
3. The combined local leak rate test (Type B and C tests) acceptance criterion is 5 0.60 La, calculated on a maximum pathway basis, prior to entering a mode of operation where containment integrity is required.
4. The combined local leak rate test (Type B and C tests)  ;

acceptance criterion is 5 0.60 La, calculated on a minimum ' pathway basis, at all times when primary containment integrity is required.

5. Airlock overall leak rate acceptance criterion is 5 0.10 La when tested at > Pa.

The provision of the Definition (1.0.Y) for Surveillance Frequency does not apply to the test frequencies specified in the Primary Containment Leak Rate Testing Program. D. Radioactive Effluent Controls Program This program conforming to 10 CFR 50.36a provides for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably Amendment No. 141, 163 265 1 1

m VYNPS achievable. The program shall be contained in the ODCM, shall be implemented by operating procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional. capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint. determination in accordance with the methodology in the ODCM;
b. Limitations on the concentrations of radioactive material released in liquid effluents from the site to unrestricted areas, conforming to 10 times the concentration values in I Appendix B, Table 2, Column 2, to 10 CFR 20.1001 - 20.2402;
c. Monitoring, sampling, and analysis of radioa'ctive liquid and gaseous effluents pursuant to 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;
e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology i and parameters in the ODCM at least every 31 days; ]
f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that j appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2 percent of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; j
g. Limitations on the dose rate resulting from radioactive  !

material released in gaseous effluents from the site to i areas at or beyond the site boundary shall be limited to the following: *

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to i a dose rate of 3000 mrims/yr to the skin, and
2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives greater than 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ; Amendment No. 266

VYNPS

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the unit to areas at or beyond the site boundary, conforming to 10 CFR 50, Appendix I;
1. Limitations on the annual and quarterly doses to a member of the pdblic from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from the unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and
j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

l l l I 1 l i l Amendment No. 26'}}

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