Text

Application for Amendment to Technical Specifications Concerning Section 6.0, Administrative Controls - Response to Request for Additional Information
	ML021710054
Person / Time
Site:	Nine Mile Point
Issue date:	06/07/2002
From:	Conway J; Constellation Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NMPIL 1668, TAC MB2441
	Download: ML021710054 (68)
	v • d • e

P.O. Box 63 Lycoming, New York 13093 0

Constellation Nuclear Nine Mile Point Nuclear Station June 7, 2002 A Member of the Constellation Energy Group NMP 1L 1668 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE:

Nine Mile Point Unit 1 Docket No. 50-220 DPR-63 TAC No. MB2441

Subject:

Application for Amendment to the Technical Specifications Concerning Section

6. 0, Administrative Controls - Response to Request for Additional Information Gentlemen:

Nine Mile Point Nuclear Station, LLC (NMPNS) hereby transmits supplemental information requested by the NRC in support of a previously submitted application for amendment to Nine Mile Point Unit 1 (NMP 1) Operating License DPR-63. The initial application, dated October 26, 2001, proposed to revise the format and content of Technical Specification (TS) Section 6.0, Administrative Controls, in a manner similar to the Nine Mile Point Unit 2 (NMP2)

Administrative Controls section. The NMP2 TS were converted to the Improved Standard Technical Specifications (ISTS) format in License Amendment No. 91.

The supplemental information is provided in Attachments A and B to this letter to respond to the request for additional information documented in the NRC's letter dated April 4, 2002. In addition, to aid the NRC staff in completing their review of the proposed changes, Attachment C provides four tables that group the proposed TS changes by change category (i.e., Administrative (A), More Restrictive (M), Less Restrictive (L), and Relocated (R)). The change categories were previously defined in the October 26, 2001 submittal. These tables provide a summary description of the proposed changes to the Current TS (CTS), the specific CTS that are being changed, and the specific Revised TS that incorporate the changes. This supplemental information does not affect the No Significant Hazards Consideration analysis provided in the October 26, 2001 submittal.

Pursuant to 10 CFR 50.91(b)(1), NMPNS has provided a copy of this supplemental information to the appropriate state representative.

Page 2 NIP1L 1668 I declare under penalty of perjury that the foregoing is true and correct. Executed on June 7, 2002.

Very truly yours,

? nT. Conway ite Vice President JTC/DEV/jm Attachments cc:

Mr. H. J. Miller, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR (2 copies)

Mr. J. P. Spath NYSERDA 286 Washington Avenue Ext.

Albany, NY 12203-6399 Records Management

ATTACHMENT A NINE MILE POINT NUCLEAR STATION, LLC LICENSE NO. DPR-63 DOCKET NO. 50-220 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI)

DOCUMENTED IN NRC LETTER DATED APRIL 4,2002 RAI The licensee's application proposed to revise Section 6. 0, "Administrative Controls, " of the Nine Mile Point Nuclear Station, Unit No. 1, Technical Specifications (TSs). In the NRC staff's judgment, the proposed amendment is in fact a conversion of Section 6 0 of the TSs to the Improved Standard Technical Specifications (ISTS) format.

In accordance with Nuclear Energy Institute (NEI)-9606, "NEI Improved Technical Specifications Conversion Guidance, " datedAugust 1996, the attachments to an ISTS conversion application for each chapter should include the following:

1. A reprinted copy of the proposed TS in the ISTS format;

2. Marked-up pages of the current Technical Specifications to show the proposed changes;

3. Discussion of the proposed changes of the current TS;

4. Marked-up pages of the ISTS and Bases to show the proposed changes;

5. Justification for differences between the proposed changes and the ISTS;

6. Proposed no significant hazards consideration determination for the changes.

The October 26, 2001, application is lacking Items 4 (excluding Bases) and 5. These items need to be provided before the review can be completed In addition, NEI-9606 specifies that electronic files be provided in WordPerfect 5.1 on diskettes which contain discussions of the changes and justification for differences between the ISTS and the proposed new improved TS.

Response

Attachment B provides annotated pages of Section 5.0 of the ISTS to indicate deviations between the ISTS and the proposed NMP1 Revised TS. For consistency with the October 26, 2001 license amendment application and with Nine Mile Point Unit 2 TS Section 5.0, the ISTS version used as the basis for the annotation is NUREG-1434, Revision 1. Justifications for each Page 1 of 2

of the deviations are provided for each individual subsection of Revised TS 6.0. The annotated ISTS pages and the discussion of the deviations are cross-referenced by "clouds" that are numbered sequentially for each subsection.

Each line item in the annotated copy of the ISTS also contains a cross-reference to the equivalent NMP1 Current TS (CTS) requirement and/or discussion of change (DOC), as appropriate. This cross-reference is intended to provide reviewers with a quick reference to the equivalent CTS section.

Electronic files that contain the discussions of the changes and the justifications for differences between the ISTS and the proposed revised TS are provided separately.

Page 2 of 2

ATTACHMENT B NINE MILE POINT NUCLEAR STATION, LLC LICENSE NO. DPR-63 DOCKET NO. 50-220 REVISED TECHNICAL SPECIFICATION SECTION 6.0 ADMINISTRATIVE CONTROLS ISTS (NUREG-1434, REVISION 1) MARKUP AND JUSTIFICATION FOR DEVIATIONS

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.0 ADMINISTRATIVE CONTROLS

<'. 1 1 >

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BWR/6 eh lant 0**

(

all be responsible for overall unit opertion an shall delegate in writing the succession to this i

responsibility during his absence./)

The I-ant Fria.*Jpolr-designee shal approve, prior to impl entation, each proposed tes experiment modification to systems or equipment that affect nuc j]ar safety.

3 1ý;~h hf ýuerio th~mo~i atot ST~eiuperviso~r (S'shall be responsible fo e control room command function.

During any absence of thetShfrom e

control room while the unit is in:

an individual with an active Senior Reactor Operator (SRO) lice se s a e

L designated to assume the control room command function During any absence of the S* from the control room while the unit is in an individual with an active license or Reactor Operator license shall be designated to assume the control room command-function.

Se--*-

-c rTS

c Rev is 007n5 STS Rev 1, 04/07195

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.1 - RESPONSIBILITY

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. The brackets have been removed and the proper plant specific information has been provided.

3. Typographical or grammatical correction/revision.

4. The text "not addressed in the UFSAR or Technical Specifications" has been added to Revised TS 6.1.1 to clarify when this approval is required. If a test or experiment is already defined in the UFSAR or TS, it is not necessary to obtain the plant manager's approval since the safety concerns have already been addressed. This is consistent with current licensing basis.

5. The reference to "MODE 1, 2, or 3" is replaced with "the power operating or hot shutdown conditions," and the reference to "MODE 4 or 5" is replaced with "the cold shutdown or refueling conditions." These changes are consistent with the reactor operating conditions defined in NMP1 CTS 1.0, "Definitions."

6. The TSTF-65 reviewer's note has been deleted. This information is for the NRC reviewer to be keyed in to what is needed to meet the TSTF-65 allowance. This is not meant to be retained in the final version of the plant-specific submittal.

7. The NMP1 CTS page numbering is retained. The Administrative Controls portion of the Revised TS begins on Page 347, with subsequent pages numbered sequentially.

NMP1 1

ADMINISTRATIVE CONTROLS M

ý.2 Oroanization TkefuAc~na JS~U~4Ok S f

((oi2e I>

2.2.

<-'T a

- I>\\

Unit Staff The unit:

organization shall the following:

a.

A non icensed oper r shall be a igned to each eactor co aining fuel pan additionWnon-licensed perator (continued)

Rev 1, 04/07/95 i

<4'Z I (7

2.1 Onsite and Offsite Or anizationw&

Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.

The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall STF-G(

be defined and established throughout highest management AdJ F

levels, intermediate levels, and all operating organization Tflser4

. I-A*

positions.

These relationships shall be documented and STKF-*

/

updated, as appropriate, in organization charts, functiona

)

escriptions of departmental responsibilities and

/

Q(*_*l~

ationships, and job-descriptions for key personnel.S

O:'*'**-i~siios~*in equivalent fo~rms of document io Th;i shl

-dc**

umentedin theSA u

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)

v safe nperation of the plant and shall have control over

**l*those-onsite activities necessary for safe operation and maintenance of the plann S

c-2 I -A p

specified corporate sosi**shall bhave dscrporate responsibility foi overall plant nuclear safety rand 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 S>d.

The indivi uals who train the operating staff, carry out or perform quality assurance functions may

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-.5 ( reporthto appropriate onsite manager; however, these safe individtals shall have sufficient organizational freedom to pefensure their independence from operating pressures.

BWR/6 STS 5.0-2

<6Ts1- >

an.

3 INSERT 6.2.1-A

-- T S

I the planspecific titles of those personnel fulfilling the responsibilities of the positions LA>

delineated in these Technical Specifications

Organization a_-zs r.2 Organization Unit S2.2

" o Staff (continued) 4 shalle assigned/for each con ol room fr iwhich a r torr:

is erating iniODES 1, 2,...S3.

Twonit sites wt both units *iutdown or de eled r uire a tota of three non-censed oper rs for the At*eas one l~icenssZl Reactor Oper.o (RO) shall b resent the control r m when fuel i

'in the reactor."

/A'n addition, whil/the unit is ip -ODE I, 2, or 3, t least one licensed Se *or Reactor Ope tor (SRO) shall present in the controT room.

rrwt omposition may be'l--ss thank mrn um

[iremienl4of 10 CFR 50.54(m)(2)(i) and*..2.a kperiod of time notjgB exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in-orderlto mnodate unexpected absence of on-duty. shift crjw members ided immediate action is taken to restore the !hift trew osition to within the minimum requirements.

""&ý 14W4ý0s shall be on site when fuel is in the reactor.

The posltfo may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absenc provided immediate action is taken to fill the required position.

2 S.ar 0e Administrative procedures shal be developed and implemented to limit the working hours of" wh perform safety related functions (e.g., licensed l,

icensed (,

auxiliary operators~rand key maintenancel personnel).

I 46

,4..'

Adequate heavy usi

't coverage sha 1 be-maintained wihout routii overtime... T' 5dobjective shall]Y to have sonnel wor an [8 or 12] hour ay, nominal

while t unit is operatij.

However, in iforese problems requir sIubstantial am nt

.o be sed, or during e ended periods rWeling, major mai enance, or maj plant on a temporary bassthe followin uidelin owed:

(continued)

Rev 1, 04/07/95 BWR/6 STS 5.0-3

Organization Organization 4.2 Unit Staff (continued)

< (.2.2, ~>

1.

An individual should not be permitted t work more ti 16 urs straight, e cluding shift u _over time;

2.

individual sho d not be permit'd to work more th 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period, r more than 24 houp in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months eriod, nor mor than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period all excluding s ift turnover time;

3.

A break

'at leasth8 hou should be allowe etween workp iods, includin hift turnover ti

4.

Exc t during exten shutdown periodsthe use of o

rtime should b considered on an i ividual basiX<

ion

.above gul e ines mall e au orize by the-lant -

or designee, in with oved administrative proce ures, o

_lsf

,m in -accor ce wi es is h

it staff mbers shall e limite(

tC P icy Statemf I,.

sna ijoli a

&=Wl notcenase.

The Shift Technical: Advisor (STA shall provideadvisor S

technical support to the jhift n the areas of thermal hydraulics, re'actor engineeringl and plant analysis with regard to the safe operation of the unit.

In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

Rev 1, 04/07/95

<,,CTS>

J.2 '

T57TF- ;2!58

'Pc-r M.1I>

BWR/6 STS

KcTs>

4 IJ

(:1 INSERT 6.2.2-A At least two non-licensed operators shall be assigned when the unit is in the power operating condition; and at least one non-licensed operator shall be assigned when the unit is in the hot shutdown, cold shutdown, or refueling conditions. In addition, if the process computer is out of service for greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , at least three non-licensed operators shall be assigned when the unit is in the power operating, hot shutdown, cold shutdown, or refueling conditions.

INSERT 6.2.2-B IU_

Oc The controls shall include guidelines on working hours that ensure adequate shift coverage shall LA.-(/

be maintained without routine heavy use of overtime.

INSERT 6.2.2-C

'Doc-\\ Controls shall be included in the procedures to require a periodic independent review be LA >

conducted to ensure that excessive hours have not been assigned.

rT__STF__2_5_ý78

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.2 - ORGANIZATION

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. The brackets have been removed and the proper plant specific information has been provided.

3. Typographical or grammatical correction/revision.

4. Since NMP1 is a single unit, the non-licensed operator requirements have been revised for clarity. Also, the bracketed information regarding dual units has been deleted.

5. The referenced requirement is a Specification, not a CFR requirement; therefore, the word "Specification" has been added to clearly state that "6.2.2.a" is a Specification. In addition, the ISTS reference to Specification 5.2.2.g has been deleted since ISTS 5.2.2.g only describes the Shift Technical Advisor (STA) qualifications, not that an STA is part of the shift crew composition or the reactor operating conditions when the STA is required.

6. The words "of on-duty personnel" have been added to Revised TS 6.2.2.c for consistency with a similar statement in Revised TS 6.2.2.b.

7. The STA provides advisory technical support to all members of the shift crew, including the Station Shift Supervisor (SSS) and the Assistant Station Shift Supervisor (ASSS) (i.e., the NUREG-1434 Shift Supervisor position). In addition, the STA position may be filled by the ASSS (provided the ASSS meets the appropriate requirements). To provide a more generic, but technically accurate, statement as to whom the STA provides technical support, the term "Shift Supervisor" has been replaced with "shift supervision."

8. The proper plant-specific department description has been provided in Revised TS 6.2.1.d and Revised TS 6.2.2.d.

9. Plant-specific wording preference or item numbering that is consistent with the NMP1 CTS.

NMP1 1

Unit Staff Qualifications

(<TS>

?N 0 AMINISTRATIVE CONTROLS

.3 Unit Staff Qualifications Reviewer' Note:

Minimum q ifications for memb s of the unit staf shall bsp.

e fied by use of -an verall qualificati.

atement referenc.

an ANSI Sta rd acceptable to e NRC staff or by s cifying individua osition q

ifications, Gen eally, the first met* is preferable; h ever, the

/econd method is a Pptable to those uni s/*taffs requiring ecial 1

qual*if tbf q o s

Each staff shall meet or exceed the mini latorfuide i.e, Ke slon Z, 1981,

I andard accep,le to the NRC s

[Regulatory Gide 1.8] shall me ifications of egulations, Re~ at(

Is acceDtab 7to NRC staff].

<oc,4.,>

Rev 1, 04/07/95 5.0-5 BWR/6 STS

r SINSERT 6.3.1-A

<6-.3 I> ANSI N18.1-1971 for comparable positions, except for; the Manager Operations who, in lieu of meeting the senior reactor operator license requirements of ANSI N18.1-1971, shall 1) hold a senior reactor operator license at the time of appointment, or 2) have held a senior reactor operator license at Nine Mile Point Nuclear Station Unit 1 or at a similar unit, or 3) have been certified for equivalent senior reactor operator knowledge; and the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

INSERT 6.3.2-A@

/'*oc-\\

6.3.2 For the purpose of 10 CF:R55*.4),a licensed Senior Reactor Operatoro and a license eacto perator are those individuals who, in addition to meeting the requirements of v.3.1, perform the functions described in 10 CFR 50.54(m).

4e

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.3 - UNIT STAFF QUALIFICATIONS

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. The bracketed "Reviewer's Note" has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This is not meant to be retained in the final version of the plant-specific submittal.

3. The brackets have been removed and the proper plant specific information has been provided. The requirements stated for the Manager Operations are consistent with the current licensing basis, as approved by the NRC in License Amendment No. 160.

4. Plant-specific wording preference that is consistent with the NMP1 CTS.

NMP1 1

Procedures INISTRA cedures

cr*

..4AONI 4 Proc

<'Pcc A.2>

<Dcc Mc

c.

d.

TIVE CONTROLS

..Written proceduresf shall be established, implemented, and

-maintained cover t e following activities:

a.

The applicable erocedures recommended in Regulatory Guide 1.33, Appendix A, d I

b.

The emergency operating procedures required to implement th requirements of NUREG-0737 and NUREG-0737, Supplement 1, ZQuaiT efor effluent and 6epy rFi re KMtiel on M rogram implementation; monitoring; All programs specified in Specification 5.

ry',3-or eAc-ce-i A"-g.

4-ej

.w

,ev Rev 1, 04/07/95 BWR/6 STS r

5.0-6

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.4 - PROCEDURES

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. The brackets have been removed and the proper plant specific information has been provided.

3. Typographical or grammatical correction/revision.

4. These words have been added for clarity to ensure that this program is not confused with the environmental monitoring program.

5. The term "administrative policies" and the requirement to "meet or exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1972" have been added consistent with the current licensing basis.

6. The referenced version of Regulatory Guide 1.33 has been changed consistent with the current licensing basis.

NMP1 1

Programs and Manuals

<cs>

L

.0 ADMINISTRATIVE CONTROLS Programs and Manuals The following programs shall be established, implemented, and maintained.

<

SOffsite Dose Calculation Manual (ODCM) 5

a.

The ODC! shall contain t methodology and p ameters used in t calculation of fsite doses result' g from ra oactive gaseous d liquid effluents; in the calculation gaseous and 1ii d effluent monitor g alarm and trip setpoints, and J therconduct of the adiological environmental nitoring program; d

b.

The ODCM all als contain th radioactive.efflue control andradiological en renmental monitori activ ies and description of the information at should be cluded in the Annua Radiological Envir mental 0 rating, and Radioac ye Effluent Releas eports requi y Specification [5..2] and Specificati

[5.6.3].

hanges to the ODCM0

,i-4.

a.

S all be documentedand records of reviews performed shall e retained.

is documentation hall contain:

S1.

suffi ent information t support the chan (s) together wit the appropriate a lyses or evaluati ns justifying t

change(s), and

2.

a determination t t the change(s) intain the levels of radioactive fluent control re ired by 10 CFR 20.130, 40:CFR 190, 10 C*

50.36a, and 10 CFR 50 endix I, and not dversely impact e

accuracy o reliability of ef uent, dose, or tpoint calculat' ns;

b.

Shall bec effective after eview and.accep nce by the

[onsite eview function] a the approval of he [Plant Superi endent]; and

c.

Sh be submitted to he NRC in the f m of a complete, 1 ible copy of the ntire ODCM as a art of, or concur ent ith, the Radioac ye Effluent Rel se Report for th period of the report i which any change in the ODCM was de.

(continued)

Rev 1, 04/07/95

BWR/6 STS 5.0-7

INSERT 6.5.1-A g shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the Offsite Dose Calculation Manual to be changed, together with appropriate analyses or evaluations justifying the change(s);

b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. Documentation of the fact that the change has been reviewed and found acceptable.

/ e--s >

4-9, 1 -e..>

Programs and Manuals 5

.5 Programs and Manuals

,_5 1

lffsite Dose Calculation Manual (ODCV4)

(continued)

<6.14 > S.2

'p Each c nge shall b identified by m kings in the in o he fected page clearly indic ng the area o he page was change and shallindate the date a

e., month and year) the hange was impla nted.

I Primary Coolant Sources Outside-Containment

<Toc-A. 3>

This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to Ilevel s as low as practi cable.

The systems includef(the

-ý Pres re ore pr, ig res e Core Spra, esiual #at R oval, React

-Core Isolat *onCooling,-

p n recoiner Aroce sam n

and Sta Gas Treatu nt The program shall inc ude the o owing:

a.

Preventive maintenance and periodic visual inspection rl._

requirements; and

b.

leak test requirements for each system at S*

interva~ls (continued)

Rev 1, 04/07/95

ý..9.e-,

5.0-8 BWR/6 STS

<T.S>

INSERT 6.5.2-A Core Spray, Containment Spray, Emergency Cooling, Shutdown Cooling, Reactor Cleanup, Vacuum Relief, Reactor Water Sampling, Containment Atmosphere Dilution (CAD) H20 2 Monitor, Drywell Containment Atmosphere Monitoring (CAM), Post Accident Sampling, Radioactive Gaseous Effluent Monitoring (RAGEMS), Offgas Effluent Stack Monitoring (OGESMS), and Post Accident Vent to Reactor Building Emergency Ventilation.

INSERT 6.5.2-B

1)

The provisions of Specification 4.0.1 are applicable to the 24 month frequency for performing system leak test activities.

<\\A >

Programs and Manuals f.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

/ he public from radio ctive effluents as low reasonably achievable.

The pr ram shall be contained '

the ODCM, shall be implemented by pro dures, and shall includ remedial actions ty

.be taken whenever the program limits are ceeded.

The progra cshall include t following elements:

SLimitat ns on the functional cap bility of radioacti S:a.

moniorin liquid rnd gaseous monitoring i trumentation includng surv lance tests and setpoin determination in a ordance wit the methodology in the M;

b.

mitations on the concent-ations of radioactiv material released in liquid efflu ts to unrestricted a eas, conforming to 10 CFR 20 Appendix B, Table 2, Coluimn 2; Monitoring, sampling and analysis of radi ctive liquid and gaseous effluents i accordance with 10 20.1302 and with the methodology a parameters in the 0 M;

d. Limitations on e annual and quarter y doses or dose commitment to member of the publi from radioactive materials in iquid effluents rele ed from each unit to unrestrict areas, conforming 0 CFR 50, Appendix I;

e.

Determit ion of cumulative and projected dose contri tions from r ioactive effluents fo the current calendar arter and c rent calendar year in ccordance with the me odology and arameters in the 0CM least every 31odays;

f.

mitations.on the functi nal capability and us of the iqand and gaseous effl nt treatment systemst ensure that appropriate portions o these systems are use to reduce releases of radioacti ity when the projected doses in a period of 31 days w ld exceedf2% oftht g delines for the annual dose or dos. :.coitment, conformin to 10 CFR 50, Appendix 1;

g.

Limitations on he dose rate resultin from radioactive material rel sed in gaseous effluen to areas beyond the site bounda conforming to the do associuated with 10 CFR 20 Appendix B, Table 2, C umn 1;

h.

Limitatons on the annual and arterly air doses r ulting from n ble gases released in aseous effluents fro each (continued)

Rev 1, 04/07/95 BWR/6 STS 5.0-9

Programs and Manuals Programs and Manuals 5.5.4 Radioactive Efflue t Controls Program

( ntinued) unit to a as beyond the site b ndary, conforming t 10 CFR

,Appendix 1;

i.

Li ations on the annual d quarterly doses t a member of epublic from iodine-i, iodine-133, triti

, and all adionuclides in parti late form with half ives > 8 days in gaseous effluents: eleased from each un*

to areas beyond the site boundary, onforming to 10 CFRO, Appendix I; and

j.

Limitations on he annual dose or do

-commitment to any member of th public due to releass of radioactivity a to radiation om uranium fuel cyc sources, conforming o

40 CFR 5.5.5 Comionent clic or Transienttimit 10 This pr*gram provides co ols to track the-FR, Section [

cyc*l" and transient o rrences to ensure at components e

ma tained within the esign limits.

5.5.6/

Pre-Stress Concrete Containme2 Tendon Surveillance PP 4ram This p gram provides contr s for monitoring any t don S:degr ation in pre-stress concrete containments including eff ctiveness of its co osion protection mediu, to ensure ntainment structura integrity.

The progra shall include aseline measuremen prior to initial oper

/

ions.

The Tendon Surveillance Progr,, inspection frequenc s,* and acceptance Scriteria shall b in accordance with [R ulatory Guide 1.35, Revision 3, 19

].

The provisi s of SR 3.0.2 and SR

.0.3 are. applicable o the Tendon Su eillance Program ins tion frequencies.

5.5.7 Inservice Ustinq Program This pr gram provides co ols for inservice esting of ASME Co Class

, 2, and 3,compo nts including app cable supports.

e proam shall inclu.duhe following:)

(continued)

Rev 1, 04/07/95 BWR/6 STS 5.0-10

Programs and Manuals

.5

S 5.5.In ervice Testin Pro am (continued)

a.

Testing freq ncies specified in -Se ion XI of the ASME Boiler and ressure Vessel Code a applicable Addenda as

/

~

follows:"

ASH *oiler and Pressure Ve el Code and plicable Addenda terminology for Required Frequen *es inservice testing for performing i service activities testinq activiies

.Weekly, At least ce per 7 days Monthly -At least nce per 31 days S~~Quarterly~r

-every-*.i.-/

S3 mon s

At 1 st once per 92 days Semian ually or

/

eevvy 6 months least once per 184 day EvJ g months t least once per 276 d s

,arly or annually At least once per 366 ays iennially or every S Zyears At least once per 31 days The.provisions of SR

.0.2are appli.cable to e above requied Fequenici for performing inserv*e testing activities

c.

The provision of.SR 3.0.3 are oappli le to inservice testing actjities; and

d.

Nothing the ASME.Boiler and ressure Vessel Code all be constr d to supersede the r uirements of any TS.

Vt r.5-.8' Ventilation Filtep/Testinq Program (Vp)

A program sha be established to mplement the following equired testing of gineered Safety Fe6 re (ESF) filter ventil ion systems a the frequencies."spe fied in [Regulatory Gufe

],

and in cordance with [Regu tory Guide 1.52, Revis-' n 2; ASME N510-1 9; and AG-I].

a.

Demonstrate for ea of the ESF systems t an inplace test of the high effi ency particulate air PA) filters shows a penetration d system bypass < [0.

% when tested i nn (continued)

Rev 1, 04/07/95 5.0-11

.BWR/6 STS programs and Manuals 1;L J

Programs and Manuals KCTS>..s Programs and Manuals 5.5

/

.8 Venti ation Filter Testin Program

,VFTP

( ntinued) accordance with egulatory Guide 1.

Revision 2, and ASME N510-1989] at t system flowrate ecified below [+/- 10%]:

ESF ntilation System Flowrate

b.

De nstrate for each the ESF systems that an place test the charcoal adso er shows a penetration a system ypass,< [0.051% w n tested in accordance wi

[Regulatory Guide 1.52, Revison 2, and ASME N510-1989]

t the system flowrate specif d below [+/- 10%]:

ESF V tilation System Flowrate

/

I]

c.

Dem strate for each of the SF systems that a labor ory t't of a sample of the c rbcoal adsorber, when obt ned as scribed in [Regulatory uide

1.52, Revision 2],

hows the "methyl iodide penetratoin less than the valUe s cified below when tested in ccordance with [ASTM D3

-1989] at a temperature of < [ *C] and greater than or qual to the relative humidit specified below:

ESF Ve ilation System Pen ration RH

]

E Reviewe s Note:

Allowable pe ration =-[100% - methyl d ide

--effic nncy. for charcoal credi ed in,staff safety evalua 'on]/

(saf y factor).

fety factor

[5] for systems with heaters.

[ (7] f r systems without heaters (continued)

Rev 1, 04/07/95 k

BWR/6 STS 5.0-12

Programs and Manuals

<-T'5Programs and Manuals 55.5.9 x losive Gas andAtoraqe Tank Radioact it Monitoring Program This program povides controls for tentially explosive gas mixtures con ined in the [Waste s Holdup System], [the ntity of radioac vity contained in g storage tanks or fed in the offgas t atment system, and t e quantity of radioactiv y

contai d in unprotected out or liquid storage.tank].

The gaseo radioactivity quan ties shall be determin Ifollowing the met dology in [Branch T nical Position (BTP).rSB 11-5, "P stulated Radioacti/v elease due to Waste G System Leak or ahe liqu radwaste quantities s 1 be determined in accordance with [S ndard Review Plan, Sec on 15.7.3, "Postulat Radioactive Relea, due to Tank Failures" (continued)

Rev 1, 04/07/95 BWR/6 STS 5.0-13

" 1

Programs and Manuals

<TS Programs and Manuals 5.5.9

£xp !,sive Gas and Storaqe Tank Radioactivit Monitorin Pro ram (ntinued)

The program shal include:

a.

The limi' for concentrations lhydrogen and oxygen i he (Wastee as Holdup System] an a surveillance programo ensur the limits are main 1 ed. Such-limits shal be a

priate to the syste design criteria (i.e.

whether or not the system is de gned to withstand a hy ogen plosion);

b.

A surveillance pro am to ensure that the antity of radioactivity co ained in [each gas sto ge tank and fed into the offgaa treatment system] is I s than the amount that would r ult in a whole body ex sure of 2 0.5 rem to

.any individ al in an unrestricted ea, in the event of [an uncontrol d release-of the tanks' contents]; and

c.

A survilance program to ens e that the quantity of radi activity contained

,i outdoor liquid radwaste t s

th are not surrounded by/ iners, dikes, or walls, cap le holding the tanks' coents and that do not have t k

overflows and surroundi g area drains connected to e

[Liquid Radwaste Trea ent System] is less than t amount that would result i concentrations less than t limits of 10 CFR 20, Append' 'B,.Table 2, Column 2, at e nearest potable water s ply and the nearest surfaewater supply in an unrestricte area, in the event of an ncontrolled release of t anks' contents.

The provision f SR 3.0.2 and SR 3.0.3 re applicable to the Explosive Ga andStorage Tank Radio ivity Monitoring Prog m surveillan frequencies.:

(continued)

Rev 1, 04/07/95 5.0-14 BWR/6 STS

"Programs and Manuals 5Programs and Manuals

,5 3>D Tpr¢hnical Snecifications (TS) Bases-Control Proqiram This program provides a means for processing changes to the Bases of these Technical Specifications.,

a.-

Changes to the -Bases of the -TS. shall be,-made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the fol owing:

1.

change.in the T I ncor orated in the license; or

2. 2_ cchange to the 9

FSAR or Bases that.

U-10 CFR 50.59.

c. The Bases Control Program,.shall ' contain provisions to enssure that the Bases are maintained consist t

$FthISAR.

d.

Proposed changes that.meet the criteria of.Gb.

a'ove shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without (continued)

Rev 1, 04/07/95 (c[s>

'.5. 10 Die 1 Fuel Oil Testing Pr ran.(continued)

Acceptability of ew fuel oil -for use pror to addition to storage tanks

.determining that the uel oil has:

1.

an AP gravity or an absolut specific gravity withi

2.

a flash point and kin atic viscosity within li its for ASTM 2D fuel oil,

3.

a clear and bri t :appearance with prope color; Other properties or ASTM 2D fuel oi.1ar ithin limits within 31-days-oll~owing sampling and dition to storage tanks; and

c.

Total pa iculate concentration the fuel oil is

/I

..when t ted -every 31 days -in ordance with ASTM D-76, Meth A--2 or.A-3.

Technical Snecifications (TSI Bases.Control Program 15' 5.0-15 OWR-/6. STS

Programs and Manuals

.5 Programs and Manuals Technical Specifications (TS) Bases Control Proqram (continued)

<*>'/.

~prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.12 Saf Function Determoition Program (SFDP) is program ensures oss of safety functi n is detected and appropriate action taken.

-Upon entry jto LCO 3.0.6, an evaluation shall e made to determine f loss of safety funct n

.exists. Addit* nally, other appropr ate limitations and re dial or compensat y actions may be id Hified to be taken as result of the sup rt system inoperabi*' y and corresponding e eption to entering pported system Cond* ion and Required Actio This program mplements the requi ments of LCO 3.0.6.

T SFDP shall conta* the following:

a. :Provisions-for c sivision checks to en-re a loss of the capability to pe orm the safety function ssumed in the accident analy does not go undetecte.,

b. Provisions-or ensuring the plant-i s aintained in a safe conditio if a loss of function co ition exists; C.

Provi ons to ensure that an in erable supported syste s Coi etion Time is notjinappr riately extended as a result o multiple support system i perabilities; and

d.

Other apprupriate limita ons and remedial or com ensatory actions.

A loss of safety function exists when, assuming n concurrent single failure, a safet function assumed in th accident analysis cannot be performed.

or the purpose of this rogram, a loss of safety function may xist when a support sys em is inoperable, and:

a.

Asystem redundant to sy em(s) supported by the inope r e support system is al inoerable; or

b.

A ed system redundant o system(s) in turn su orted b

the inoperable -support system is also inoper e; or (continued)

Rev 1, 04/07/95

-5.0-16 IBWR/6 STS

Programs and Manuals Programs and Manuals 5.5.12 Safety Function ermination Pro ra ffSFDP)

(continued)

c.

A requi d system redundantto support system(s) fo, the supp ed systems (a) an:d

) above is also ino able.

The SF identifies where a oss of safety functi 'exists.

If a los f safety function i determined to exist this program, t

appropriate Conditis -and Required Actio

. of the LCO in ich the loss of saf yjfunction exists ar required to be

/

~entered.

f..

r-A Rev 1, 04/07/95 BWR/6 STS 5.0-17

INSERT 6.5.4-A 6.5.4 10 CFR 50 Appendix J Testing Program Plan

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B.

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 with the following exceptions:

1. Type A tests will be conducted in accordance with ANSI/ANS 56.8-1994 and/or Bechtel Topical Report BN-TOP-1, and

2. The first Type A test following approval of this Specification will be a full pressure test conducted approximately 70, rather than 48, months since the last low pressure Type A test.

b. The peak calculated containment internal pressure (Pac) for the design basis loss of coolant accident is 35 psig.

c. The maximum allowable primary containment leakage rate (La) at Pac shall be 1.5%

of primary containment air weight per day.

d. Leakage Rate Surveillance Test acceptance criteria are:

1. The as-found Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than 1.0 La.

2. The as-left Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than or equal to 0.75 La, prior to entering a mode of operation where containment integrity is required.

3. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 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 & C Tests including airlocks) acceptance criteria is less than 0.6 La, calculated on a minimum pathway basis, at all times when containment integrity is required.

e. The provisions of Specification 4.0.1 do not apply to the test frequencies specified in the 10 CFR 50 Appendix J Testing Program Plan.

CrTS>

INSERT 6.5.5-A

<G. I I

6.5.5 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.5 - PROGRAMS AND MANUALS

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. Typographical or grammatical correction/revision.

3. The brackets have been removed and the proper plant specific information has been provided.

4. The surveillance frequency has been changed from "refueling cycle intervals" to "24 months," consistent with the current NMP1 refueling interval. In addition, since normal Surveillance Requirements in the LCO Sections allow a 25% extension of the frequency per CTS 4.0.1, this allowance has also been added for this Surveillance Requirement. In addition, the term "or less" is unnecessary and has been deleted for consistency.

5. Paragraphs a and b of ISTS 5.5.1 that describe the contents of the Offsite Dose Calculation Manual (ODCM) have not been incorporated as part of this submittal. Paragraphs a, b, and c of ISTS 5.5.1 that describe requirements for changes to the ODCM have been replaced with the plant specific information from NMP1 CTS 6.9.1.e. Proposed revisions to the ODCM specification that are consistent with ISTS 5.5.1 are described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

6. The word "integrated" has been replaced with the word "system" as related to leak test requirements. This terminology is consistent with the NMP1 response to NUREG-0578 Item 2.1.6.a that was documented in a Niagara Mohawk Power Corporation (NMPC) letter dated December 31, 1979 and accepted by the NRC in the safety evaluation for License Amendment No. 42 (NRC letter dated April 13, 1981).

7. TS requirements for a post accident sampling program (ISTS 5.5.3) are not part of the NMP1 current licensing basis, and this submittal does not propose to add such requirements.

Confirmatory Orders issued by the NRC on March 14, 1983 and on June 12, 1984 contain requirements pertaining to NMP1 post accident sampling capability. Controls that ensure the capability to obtain and analyze liquid and gaseous samples under accident conditions are contained in plant procedures. Topics covered in these procedures include sampling and analysis, maintenance of sampling and analysis equipment, and training of personnel.

Established change control processes will provide sufficient control of changes to these procedures. Therefore, administrative controls relating to the post accident sampling program are not required to be added to the NMP1 TS to provide adequate protection of the public health and safety.

8. The Radioactive Effluent Controls Program specification of ISTS 5.5.4 has not been adopted as part of this submittal. The proposed addition of this program to the NMP1 TS is described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

NMP1 1

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.5 - PROGRAMS AND MANUALS

9. The proper plant-specific information/nomenclature has been provided.

10. TS requirements for a component cyclic or transient limit program (ISTS 5.5.5) are not part of the NMP1 current licensing basis, and this submittal does not propose to add such requirements. Controls to track the UFSAR Table V-2 cyclic and transient occurrences are contained in plant procedures. Established change control processes will provide sufficient control of changes to these procedures. Therefore, administrative controls relating to the component cyclic or transient limit program are not required to be added to the NMP1 TS to provide adequate protection of the public health and safety.

11. This bracketed requirement has been deleted because it is not applicable to NMP1 (NMP1 does not have a prestressed concrete containment).

12. The Inservice Testing Program specification of ISTS 5.5.7 has not been adopted as part of this submittal. Inservice inspection and testing requirements currently reside in TS 3/4.2.6.

The proposed deletion of TS 3/4.2.6, and the addition of an Inservice Testing Program specification to the Administrative Controls portion of the NMP1 TS, are described and evaluated in a separate license amendment request (letter no. NMP1L 1628 dated November 26, 2001).

13. The Ventilation Filter Testing Program (VFTP) specification of ISTS 5.5.8 has not been adopted as part of this submittal. The requirements for testing of engineered safeguards ventilation filter systems currently reside in TS 3/4.4.4 for the Reactor Building Emergency Ventilation System (RBEVS) and in TS 3/4.4.5 for the Control Room Air Treatment System (CRATS). This submittal is not converting either the Limiting Conditions for Operation or the Surveillance Requirements portions of the NMP1 TS to the ITS format and content; therefore, the ventilation filter testing requirements are retained in their existing TS sections rather than relocating them to the Administrative Controls portion of the TS.

14. The Explosive Gas and Storage Tank Radioactivity Monitoring Program specification of ISTS 5.5.9 has not been adopted as part of this submittal. The proposed addition of this program to the NMP1 TS is described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

15. TS requirements for a diesel fuel oil testing program (ISTS 5.5.10) are not part of the NMP 1 current licensing basis, and this submittal does not propose to add such requirements.

Requirements for testing of both new diesel fuel oil and stored fuel oil are contained in plant procedures. These procedures include sampling and testing requirements and acceptance criteria that are in accordance with applicable ASTM standards. Established change control processes will provide sufficient control of changes to these procedures. Therefore, administrative controls relating to the diesel fuel oil testing program are not required to be added to the NMP1 TS to provide adequate protection of the public health and safety.

NMP1 2

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.5 - PROGRAMS AND MANUALS

16. The Safety Function Determination Program (SFDP) specification of ISTS 5.5.12 has not been adopted as part of this submittal. An evaluation in accordance with the SFDP is to be performed upon entry into ISTS LCO 3.0.6. The NMP1 TS do not currently contain a specification analogous to ISTS LCO 3.0.6. This submittal is not converting either the Limiting Conditions for Operation or the Surveillance Requirements portions of the TS to the ITS format and content; therefore, a specification analogous to ISTS LCO 3.0.6 and the associated SFDP specification are not proposed to be added to the NMP1 TS.

17. The 10 CFR 50 Appendix J Testing Program (CTS 6.16) has been added to be consistent with the current licensing basis and TSTF-52.

18. The Radiation Protection Program (CTS 6.11) has been added to be consistent with the current licensing basis. The proposed relocation of these program requirements to the UFSAR is described and evaluated in a separate license amendment request (letter no.

NMP1L 1617 dated October 19, 2001).

NMP1 3

0 ADIMINISTRATIVE CONTROLS Reporting Requirements Reporting Requi rements The following reports shall be submitted in accordance with 10 CFR 50.4.

.6.2

/,*.*.~a>

Reporting Requirements eporting Requirements Annual Radiological-Environmental Operating Report (continued)

ý6-q.I Rev 1, 04/07/95":d'*'

-)

<6TS>

6 R

..,.6.2 (ODC

, and inn CFR 50, App ix I, Sectio.

IV.B.2, IV.

a W.-C.

The Annual Radiological Envir nmental Operating Reporrshal1 (a*

include the results of analy ll radiological environmental samples and of all environmental radiation measurements taken o~si:

4JJ:se R,

durin the period pursuant to the locations specified in the table SC*I Ures in the as well as summarized and tabulated resu ts o ese analyses and measurements 5in-the format of the table in:the Radiological Assessment Branch Technical Position Revision 1, November 1979.

[The r eort-snhal ide *ify he LD.

/* sulsxnaT reres~r c011ocateýN/osqimeters in. ~lation to th*

S... :

NRCL'Oprogram a.the axpbsu

~period.:assoc* ted with each s e-

^e 2r t]

In t

e evnt that some ind-ividal results are not Sr*

.*-3 avai able for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing Serf 6a, (o results.

The missin data shall be submitted in a supplementary report arssoon as possibl*

1 anyniuQ 5.0-19 BWRI6 STS

<C.TS>

rd

( 0 R. iA)

INSERT 6.6.2-A D including a comparison with operational controls as appropriate, and with environmental surveillance reports from the previous 5 years, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.6.22.

INSERT 6.6.2-B The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps ** covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.6.21; discussion of all deviations from the sampling schedule of Table 3.6.20-1; and discussion of all analyses in which the LLD required in Table 4.6.20-1 was not achievable.

INSERT 6.6.2-C @

(Footnote for Revised TS 6.6.2)

One map shall cover stations near the site boundary; a second shall include the more distant stations.

CIS'>

INSERT 6.6.3-A

<(0.coI

> as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summery of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, and atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wing direction, and atmospheric stability.

This same report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports.

The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the Offsite Dose Calculation Manual.

The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in the Offsite Dose Calculation Manual.

The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f. Solidification agent or absorbent (e.g., cement)

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP) and to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.6.20.

INSERT 6.6.3-B (Footnote for Revised TS 6.6.3)

K(0!i.

Le:>

In lieu of submission with the Semi-annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

<C-TS >

.6.

I'-Cn 7.4 Reporting Requirements Reporting Requirements Monthly Operating Reports (continued)

[TS-F -Z

-shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

PIIEPE -vC-24i-P-A,1' &~tJ2KAL ELE'C7A~ek%

.STA)PA-R1 A90L~c-A71W FrA gE4LCT=9 CORE OPERATING LIMITS;REPORT (COLR)

Fu

_ (Lo-esf aeeprav res;'n aoS 5 ec~~

)A LOL).

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload 4

cycle, and shall be documented in the COLR for-the following::

(1. 6 A

The in idual spe ifications tlt address c e opera)*ng j/

S:

l*_I~mi us

-erenced herjK

/./

I 4b.

The analytical methods used to determine the core operating limits shall be those previously. reviewed and approved by the NRC, specifically those described in.E

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal -mechanical' imits, core thermal hydraulic limits,

-Or CCS limits, nuclear limits such as transient J ianairsis rliits, and accident analysis limits) of the safet analysis are met.

lown ma r*

d. The COLR,Jincluding any mi cycl revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Reactor Coolant 2stem (RCS)

PRESSUJt AND TEMPERATURE LIMTS REPORT (PTLR)

a.

RCS p essure and tempera re limits for heatu, cooldown, low emperature operatf n, criticality, and ydrostatic t ting as well as h.tup and cooldown ra s shall be

/

stablished and do mented in the PTLR r the followin -

(continued)

Rev 1, 04/07/95 L

w

BWRI6 STS 5.0-20

/.c-r s>

(oq1,- >

INSERT 6.6.5-A

1) The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specifications 3.1.7.a and 3.1.7.e.

2) The Kf core flow adjustment factor for Specification 3.1.7.c.

3) The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.1.7.c and 3.1.7.e.

4) The LINEAR HEAT GENERATION RATE for Specification 3.1.7.b.

5) The Power/Flow relationship for Specifications 3.1.7.d and 3.1.7.e.

Reporting Requirements eporting Requirements 5.6.6 Reaotor.:Coolant System (RCS)

ESSURE AND TEMPERATURE IMITS RE RT PTLR (continued)

-[The individual spe fications that address S pressure and temperature limit must be referenced here

b.

The analytical thods used to determin the RCS pressure and temperat e limits shall be those reviously reviewed and approve by the NRC, specificall those described in the following ocuments: [Identify the RC staff approval documen -y date.]

c.

The LR shall be provided to he NRC upon issuance for eac re tor vessel fluence peri and for any revision or s pplement thereto.

Re ewer's Notes:

The metho ology for the calculation of he P-T li its for NRC approval sh ld include the following pr isions:

1.

The methodoloy sh I describe how-the neutron uence is calculated (ref nce new Regulatory Guide wh issued).

2. The Reactor V sel Material Surveillance Pogram shall comply with pendix H to 10 CFR 50. Th reactor vessel

material ir adiation surveillance spec en removal schedule shall. be rovided, along with how th specimen examinations shall b used to update the PTLR / yes.

3. Low mperature Overpressure Pr ection (ITOP) System lift set ing limits for the Power erated Relief Valves (POR~s d eloped using NRC-approve ethodologies may b2 includ I the PTLR.

4.

The adjusted reference emperature (ART) for each r ctor beltline material sha be calculated, accounting.or

.ýradiation embrittle nt, in accordance with Reg atory Guide 1.99, Revision 2.

5. The limiting A shall.-be incorporated it the calculation of the press e and temperature limit cu es in accordance with NUREG- 00 Standard Review Plan 5.2, Pressure L_ Temperatu, Limits.

(continued)

Rev 1, 04/07/95

< e--Ts>

6 R(

.BWR/6 STS 5.,0-21

Reporting Requirements porting Requirements 5.6.6 eactor Coolant System (RCS)

PRESSURE AND TEMPFATURE LIMITS REPORT (PTLR)

(contin d)

6.

The minimum te erature requirements* f Appendix G to 10 CFR Part 50 shal be incorporated into e pressure and temperatur limit curves.

7.

License s who have removed tw or more capsules shoul compa for each surveillan material the measured ncrease in ference temperature,..6) to-the predicted i crease in RT; where the predic d increase in RT." is ased on the ean shift in RTPT p s the two standard de ation value (2a) specified in gulatory Guide 1.99, ision 2. If the measured value ex eds the predicted val (increase in R

+ 2a A), the lice ee should provide a pplement to thePtT to demonstrat ow the results affec the approved methodology, 45.6-.7 DG Failure Re ort 10 If an individ emergency diesel g erator (EDG) experie es four or more val' failures in the las 25 demands, these fa' ures and any nonva d failures experienc by that EDG in that ime period shall b reported within 30 d

s.

Reports on EDG f lures shall inclu the information rec ended in Regulatory uide 1.9, Revjion 3, Regulatory Po Ition C.5, or existi Regulatory

_G de 1.108 reporting re uirement.

.6.8

'PAM P ort en a Special Repor is required by Cond ion B or G of LCO 3.3.[3.1], "Po Accident Monitorin (PAM)

Instrumentati a

report shall be bmitted within the llowing 14 days.

T report shall o0a line the preplanned.lternate method of monitoring, ie cause of the inop,ability, and the pl s and schedule fX restoring the insumentation channels the Functi orto OPERABLE status. V11 5.6.9 endon SurveillanceApnort Any abnormal d radation of the cont nment structure detecd

-/

L rduring the t ts required by the P e-Stressed Concrete

_1 (continued)

Rev 1, 04/07/95

.BWR/6 STS 5.0-22

Reporting Requirements

<- CRTeS ui e

6.6 Reporting Requirements K5.6.9 Ten n Surveillance ReDore (continued) ontainment Tendon S eillance Program s 11 be reported to NRC within 30 days.

The report shall i ude a descriptionl

/

tendon condition, he cOndition of th /concrete (especial.

tendon anchorag

), the inspection

.ocedures, the tole nceý

-cracking, and e -corrective acti taken.

FReviewer's ote:

These.repor may be required c bring Iinspecti, test, and maint ance activities.

se reports I determ ed on an individ

. basis for each un and their IPre ationand it are designated in e Technical LS ifications.

/-

L(-

4r-A Rev 1, 04/07/95

.BWR/6 STS

ý5.0-23

KCTS>

INSERT 6.6.6-A C 6.6.6 Special Reports Special reports shall be submitted within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Reactor Vessel Material Surveillance Specimen Examination, Specification 4.2.2.(b)

(12 months).

b. Safety Class 1 Inservice Inspection, Specification 4.2.6 (Three months).

c. Safety Class 2 Inservice Inspections, Specification 4.2.6 (Three months).

d. Safety Class 3 Inservice Inspections, Specification 4.2.6 (Three months).

e. Primary Containment Leakage Testing, Specification 3.3.3 (Three months).

f. Secondary Containment Leakage Testing, Specification 3.4.1 (Three months).

g. Sealed Source Leakage in Excess Of Limits, Specification 3.6.5.2 (Three months).

h. Calculate Dose from Liquid Effluent in Excess of Limits, Specification 3.6.15.a.(2)(b) (30 days from the end of the affected calendar quarter).

i.

Calculate Air Dose from Noble Gases Effluent in Excess of Limits, Specification 3.6. 15.b.(2)(b) (30 days from the end of the affected calendar quarter).

j. Calculate Dose from 1-131, H-3 and Radioactive Particulates with half lives greater than eight days in Excess of Limits, Specification 3.6.15.b.(3)(b) (30 days from the end of the affected calendar quarter).

k. Calculated Doses from Uranium Fuel Cycle Source in Excess of Limits, Specification3.6.15.d (30days from the end of the affected calendar year).

1. Inoperable Gaseous Radwaste Treatment System, Specification 3.6.16.b (30 days from the event).

m. Environmental Radiological Reports. With the level of radioactivity (as the result of plant effluents) in an environmental sampling medium exceeding the reporting level of Table 6.6.6-1, when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within thirty (30) days from the end of the calendar quarter a special report identifying the cause(s) for exceeding the limits, and define the corrective action to be taken.

INSERT 6.6.6-A (Continued) a qc-r3>

TABLE 6.6.6-1 REPORTING LEVEL FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS Analysis H-3 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Zr-95, Nb-95 1-131 Cs-134 Cs-137 Ba/La-140 Water (pCi/l) 20,000*

1,000 400 1,000 300 300 400 2**

30 50 200 Airborne Particulate Or Gases (nCi/m3)

Fish (DCi/kQ. wet)

Milk (nCi/)

Food Products (nCi/k2. wet) 30,000 10,000 30,000 10,000 20,000 0.9 10.0 20.0 1,000 2,000 3

60 70 300 100 1,000 2,000 If no drinking water pathway exists, a value of 20 pCi/liter may be used.

For drinking water samples. This is a 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/liter may be used.

Or Gases ffiCi/W)

Fish (D i/kLy wet)

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.6 - REPORTING REQUIREMENTS

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. The bracketed Note has been deleted since it is not applicable to NMP1 (individual reports are submitted for NMP1 and NMP2).

3. The Occupational Radiation Exposure Report specification of ISTS 5.6.1 has not been adopted as part of this submittal. The NMP1 CTS 6.9.1.b wording has been retained.

Proposed revisions that are consistent with ISTS 5.6.1 are described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

4. The brackets have been removed and the proper plant-specific information has been provided.

5. The initial report requirement in ISTS 5.6.1 is deleted since this initial report has been submitted on a one-time basis.

6. The Annual Radiological Environmental Operating Report specification of ISTS 5.6.2 has not been adopted as part of this submittal. The NMP1 CTS 6.9.1.d wording has been retained. Proposed revisions that are consistent with ISTS 5.6.2 are described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

7. Typographical or grammatical correction/revision.

8. The Radioactive Effluent Release Report specification of ISTS 5.6.3 has not been adopted as part of this submittal. The NMP1 CTS 6.9.1.e wording has been retained. Proposed revisions that are consistent with ISTS 5.6.3 are described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

9. The utilization of a Pressure and Temperature Limits Report (PTLR) requires the development and NRC approval of detailed methodologies for future revisions to pressure/temperature (P/T) limits. At this time NMPNS does not have the necessary methodologies submitted to the NRC for review and approval; therefore, references to the PTLR are deleted. The NMP1 specific limits and curves are provided in the P/T limits specification (CTS 3/4.2.2).

10. TS requirements for EDG Failure Reports (ISTS 5.6.7) are not part of the NMP1 current licensing basis, and this submittal does not propose to add such requirements. NMP1 has implemented a maintenance program for monitoring and maintaining diesel generator performance in accordance with the provisions of the maintenance rule and consistent with the guidance of Regulatory Guide 1.160. Therefore, in accordance with the guidance of Generic Letter 94-01 and consistent with TSTF-37, this ISTS section has been deleted.

NMP1 1

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.6 - REPORTING REQUIREMENTS

11. The PAM Report specification of ISTS 5.6.8 has not been adopted as part of this submittal.

The requirements for submitting a Special Report to the NRC in the event that accident monitoring instrumentation is inoperable currently reside in TS 3/4.6.11. This submittal is not converting either the Limiting Conditions for Operation or the Surveillance Requirements portions of the TS to the ITS format and content; therefore, the accident monitoring instrumentation requirements, including related special reports, are retained in CTS 3/4.6.11 rather than relocating them to the Administrative Controls portion of the TS.

12. This bracketed requirement (ISTS 5.6.9) has been deleted because it is not applicable to NMP1 (NMP1 does not have a prestressed concrete containment).

13. The Special Reports specification (CTS 6.9.3) has been added to be consistent with the current licensing basis. The proposed deletion of the inservice inspection related special reports (CTS 6.9.3.b through d), and the proposed deletion of the primary and secondary containment leakage testing special reports (CTS 6.9.3.e and f) are described and evaluated in a separate license amendment request (letter no. NMIP1L 1628 dated November 26, 2001).

The proposed relocation of the radiological effluent technical specification (RETS) related special reports (CTS 6.9.3.h through m, including Table 6.9.3-1) to the ODCM is described and evaluated in another separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001).

14. Plant-specific wording preference that is consistent with the NMP1 CTS.

NMP1 2

< CTS~

ADMINISTRATIVE CONTROLS High Radiation Areas' I

ýfflJe-t Ins (continued)

Rev 1, 04/07/95 5.7.1 Pursupnt to 10 CFR 20, paragra 20.1601(c), in lie of the req rements of 10..CFR-`20.460, each high radiatio area, as ede ned in 10 CFR20, in whi h thed intensity of diation is

>/00 mrem/hr but < 1000-1 em/hr, shall be barr caded and nonspicuously posted as high ýradiation area

ýnd entrance thereto ashall be controlled by equiring issuance o a Radiation Work ePermit (RWP).

lndivi, als qualified in.r iation protection

..procedures (e.g., [

aith Physics Techni ans]) or-personnel continuously esco dby ch hindividU s'

may be exempt from t PRWP. issuance re rement :during :the formance of their assi ned duties in high adiation areas with xposure rates 1000 em/hr,

.provided they are otherwise follo ng plant radiation pro ction procedure4 or entry into such hI radiation areas.

Any md* idual or group of i ividuals permitted to nter such areas all be provided wit or accompanied by one or more of the foil ing:

a A radiation monit ing device that cont' uously indicates the radiation d e rate in the area.

b.

A radiation nitoring device that ontinuously integrates the radiat n dose rate in the a a and alarms when a pres integrat dose is received.

try into such areas with this mo itoring device may be ade after the dose rate level in the area have bee established and personne are awar of them.

C.

individual qualifie in radiation protection rocedures with a radiation dos rate monitoring dev ice, ois responsiblewforipro iding positive control o r the activities wihin he area and shall perfo periodic radiation surveipance at the frequency ecified by the

[Radiation Pro ýction Manager] in the P.

5.7 In addition to he requirements of Spe fication 5.7.1, areas wiv radiation le Is Ž:

1000 mrem/hr shal be provided with locked continuousl guarded doors to prey t unauthorized entry and he keys shal be maintained under t administrative control the Shift F eman on duty or healt physics supervision.

Do s shall remai locked except during priods of access by perso el L

I BWR/6 STS 5.0-24

e>.7 High Radiation Areat7 5.7.2 (co inued) under an approv RWP that shall spe ify the dose rate le els in the immediate ork areas and the m imum allowable stay imes for individuals N those areas.

In Ieu of the stay time specifica on of the RWP, direc or remote (such as c osed circuit TV camer s) continuous surveil nce may be made by rsonnel qualifj in radiation prote ion procedures to pr ide positive expo re control over the ivities being perfo d within the are.

5.7.3 For individual high iation areas with r iation levels of

> 1000 mrem/hr, acc ssible to personnel, at are located with large areas such reactor containment here no enclosure ists for purposes of ocking, or that canno be continuously gu ded, and where no e closure can be reason y constructed a-ou the indi vidual a a, that individual ar shall be barricad 'and conspi cuou y-posted,- and a flashj g light shall be a ivated as a warning d ice.

Rev 1, 04/07/95 BWR/6 STS 5.0-25

<&rS>

INSERT 6.7

<e,,.12. I> 6.7.1 LA.- I>

2, I 6.7.2

<voc-LA. I>

In lieu of the "control device" or "alarm signal" required by Paragraph 20.203(c)(2) of 10CFR20, each high radiation area normally accessible* by personnel in which the intensity of radiation is greater than 100 mrem/hr** but less than 1000 mrem/hr** 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 in accordance with site approved procedures. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rates in the area have been established and personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection, with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the radiation protection manager or designate in the Radiation Work Permit.

In addition to the requirements of 6.7.1 areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem** shall be provided with locked doors to prevent unauthorized entry, and the hard keys or access provided by magnetic keycard shall be maintained under the administrative control of the Station Shift Supervisor or designate on duty and/or the radiation protection manager or designate. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify in accordance with site approved procedures accordingly, the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area.

In lieu of the stay time specification of the RWP, continuous surveillance, direct or remote, such as use of closed circuit TV cameras, may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem** that are located within large areas, such as the drywell, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device.

by accessible passage and permanently fixed ladders measurement made at 18" from source of radioactivity

JUSTIFICATION FOR DEVIATIONS FROM NUREG-1434, REVISION 1 REVISED TS 6.7 - HIGH RADIATION AREA

1. The Administrative Controls portion of the NMP1 Revised Technical Specifications (TS) continues to be identified as Section 6.0, and all subsection numbers begin with "6" rather than "5," to be compatible with the remainder of the NMP1 Current TS (CTS).

2. The brackets have been removed and the plant specific information from NMP1 CTS 6.12 has been provided. The High Radiation Area specification of ISTS 5.7 has not been adopted as part of this submittal. Proposed revisions to the High Radiation Area specification that are consistent with ISTS 5.7 are described and evaluated in a separate license amendment request (letter no. NMP1L 1617 dated October 19, 2001 and supplement).

NMP1 I

ATTACHMENT C NINE MILE POINT NUCLEAR STATION, LLC LICENSE NO. DPR-63 DOCKET NO. 50-220 The following four tables summarize the proposed changes to NMP 1 Current Technical Specification (CTS) 6.0, Administrative Controls. The change categories were previously defined in the October 26, 2001 submittal.

"* Table A, Administrative Changes Matrix

"* Table M, More Restrictive Changes Matrix

"* Table L, Less Restrictive Changes Matrix

"* Table R, Relocated Specifications and Removal of Details Matrix

TABLE A - ADMINISTRATIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION SECTION Table of Contents A. 1 Editorial changes, reformatting, and revised numbering.

N/A N/A 6.1, Responsibility A. 1 Editorial changes, reformatting, and revised numbering.

6.1 6.1,6.5 A.2 Moves the requirements of CTS 6.5.2.3 and 6.5.2.5 to Revised TS 6.1.1. Removes the 6.1.1 6.5.2.3, phrase "and their safety evaluations" from the CTS requirements regarding Plant 6.5.2.5 Manager reviews and approvals of proposed tests, experiments, and modifications to systems or equipment that affect nuclear safety, since approval of the safety evaluation is inherent in the approval of the modification, test, or experiment.

A.3 Adds the acronym "SSS" for the Station Shift Supervisor-Nuclear position title.

6.1.2 6.1.2 A.4 Deletes the requirement for a management directive to be reissued annually to all N/A 6.1.2 personnel stating that the Station Shift Supervisor - Nuclear is responsible for the control room command function.

6.2, Organization A. 1 Editorial changes, reformatting, and revised numbering.

6.2 6.2 A.2 Replaces the phrase "qualified in" with "qualified to implement" as it relates to 6.2.2.c 6.2.2.d radiation protection procedures.

A.3 Replaces the term "health physics" with "radiation protection," and replaces the term 6.2.1.d, 6.2.1.d, "health physicists" with "key radiation protection personnel."

6.2.2.d 6.2.2.h A.4 Moves the requirements for unlicensed operating personnel from CTS Table 6.2-1 to 6.2.2.a Table 6.2-1 Revised TS 6.2.2.a, clarifies the requirements for unlicensed operators when the process including Notes computer is out of service for greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and replaces the term "unlicensed" (2) and (3) with "non-licensed.'

A.5 Moves the requirement that allows the shift crew composition to be less than the 6.2.2.b Table 6.2-1 minimum requirements from CTS Table 6.2-1 to Revised TS 6.2.2.b, and replaces including references to Table 6.2-1 with references to Revised TS 6.2.2.a and 10 CFR Note (6) 50.54(m)(2)(i).

Page 1 of 4

TABLE A - ADMINISTRATIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION SECTION A.6 Deletes note that specifically disallows any shift crew position to be unmanned upon 6.2.2.b Table 6.2-1 shift change because an oncoming shift crewman scheduled to come on duty is late or including absent, since the requirement of this note is covered by the wording of Revised TS Note (6) 6.2.2.b.

A.7 Deletes statement that more operators can be assigned if needed, since the requirements N/A Table 6.2-1 of the minimum shift crew composition are specified and thus it is not necessary to including specify whether the requirements may be exceeded.

Note (1)

A.8 Incorporates the qualification requirements of the Shift Technical Advisor from CTS 6.2.2.f 6.3.1 6.3.1, and modifies those requirements to reference the Commission Policy Statement on Engineering Expertise on Shift.

A.9 Replaces the person to whom the STA provides advisory technical support with a more 6.2.2.f N/A generic statement; i.e., the term "Shift Supervisor" has been replaced with "shift supervision."

6.3, Unit Staff Qualifications A. 1 Editorial changes and reformatting.

6.3 6.3 A.2 Moves the qualification requirements for the Shift Technical Advisor to Revised TS 6.2.

6.2.2.f 6.3.1 6.4, Procedures A. 1 Editorial changes, reformatting, and revised numbering.

6.4 6.8 A.2 Moves the requirement relating to Regulatory Guide 1.33 to a separate sub-item within 6.4.1.a 6.8.1 Revised TS 6.4.1, and identifies the specific revision of the regulatory guide.

6.5, Programs and Manuals A. 1 Editorial changes, reformatting, and revised numbering.

6.5 4.3.3.a, 6.9.1.e, 6.11, 6.14, 6.16 A.2 Incorporates wording changes consistent with the changes to 10 CFR 50.59 published in 6.5.3 N/A the Federal Register (Volume 64, Number 191) dated October 4, 1999.

Page 2 of 4

TABLE A - ADMINISTRATIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION SECTION A.3 Provides a more descriptive paragraph for the Primary Coolant Sources Outside 6.5.2 6.14 Containment program (previously CTS 6.14, Systems Integrity) that outlines program elements and identifies applicable systems.

A.4 Adds a statement of applicability of TS 4.0.1 to CTS 6.14 (Revised TS 6.5.2).

6.5.2 6.14 6.6, Reporting Requirements A. 1 Editorial changes, reformatting, and revised numbering.

6.6 1.31, 3.6.15.a, 3.6.15.b, Table 4.6.15-2, 3.6.15.d, 3.6.16.b, 3.6.20, 3.6.22, Table 4.6.20-1, 6.9.1, 6.9.2, 6.9.3 A.2 Delete the references to three topical reports, since all of the methods reviewed and 6.6.5.b 6.9.1.f approved by the NRC for Loss of Coolant Accident analysis and Stability analysis are now contained in a single report, NEDE-2401 1-P-A.

A.3 Deletes duplicate statements and unnecessary details regarding submittal of reports in 6.6 6.9, 6.9.1.c, accordance with 10 CFR 50.4.

6.9.1.f, 6.9.3 6.7, High Radiation Area A. 1 Editorial changes, reformatting, and revised numbering.

6.7 6.12 Current Specification 6.4, Training None None None None Current Specification 6.5, Review and Audit A. 1 Moves the requirements of CTS 6.5.2.3 and 6.5.2.5 to Revised TS 6.1.1.

6.1.1 6.5.2.3, 6.5.2.5 Page 3 of 4

TABLE A - ADMINISTRATIVE CHANGES MATRIX Page 4 of 4 DOC #

SUMMARY

REVISED TS CTS SECTION SECTION Current Specification 6.6, Reportable Event Action A. 1 Removes Reportable Event notification requirements from the Technical Specifications, N/A 6.6. L.a since these requirements are contained in 10 CFR 50.72 and 10 CFR 50.73.

Current Specification 6.7, Safety Limit Violation A. 1 Removes the Safety Limit Violation requirements as they relate to NRC notification, N/A 6.7. 1.a, since the requirements are contained in and based upon the requirements located in 10 6.7.1.b, CFR 50.36(c)(1), 10 CFR 50.72, and 10 CFR 50.73.

6.7.1.c, 6.7. l.d Current Specification 6.10, Record Retention None None None None Current Specification 6.13, Fire Protection Inspection None None None None Current Specification 6.15, Iodine Monitoring None None None None

TABLE M - MORE RESTRICTIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION SECTION 6.1, Responsibility M. 1 More clearly specifies the qualifications of the individual designated to assume the 6.1.2 N/A control room command function in the absence of the Station Shift Supervisor-Nuclear.

6.2, Organization M. 1 Add description of the duties of the Shift Technical Advisor.

6.2.2.f N/A 6.3, Unit Staff Qualifications M. 1 Clarifies the qualification requirements for licensed Senior Reactor Operators and 6.3.2 N/A licensed Reactor Operators to ensure that there is no misunderstanding when complying with 10 CFR 55.4 requirements.

6.4, Procedures M. 1 Adds requirement that there be written procedures for activities involving the 6.4.1.b, N/A emergency operating procedures, quality assurance for radioactive effluent and 6.4. 1.c, radiological environmental monitoring, and the programs listed in Revised TS 6.5.

6.4.1.e 6.5, Programs and Manuals M. 1 Adds a new program, the Technical Specifications Bases Control Program.

6.5.3 N/A 6.6, Reporting Requirements None None None None 6.7, High Radiation Area None None None None Current Specification 6.4, Training None None None None Page 1 of 2

TABLE M - MORE RESTRICTIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION SECTION Current Specification 6.5, Review and Audit None None None None Current Specification 6.6, Reportable Event Action None None None None Current Specification 6.7, Safety Limit Violation None None None None Current Specification 6.10, Record Retention None None None None Current Specification 6.13, Fire Protection Inspection None None None None Current Specification 6.15, Iodine Monitoring None None None None Page 2 of 2

TABLE L - LESS RESTRICTIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION CHANGE SECTION TYPE 6.1, Responsibility L. 1 CTS provides the title of the individual designated by the Plant Manager 6.1.1 6.5.2.3, 1

to approve modifications to structures, systems, and components, and 6.5.2.5 approve proposed tests and experiments. The Revised TS will not specify this individual, but will require the person to be designated by the plant manager.

6.2, Organization L. 1 CTS provides a description of the individuals who can be designated by 6.2.2.d 6.2.2.h 1

the Plant Manager to approve modifications to overtime requirements.

The Revised TS will not provide this description, but will require the person to be designated by the plant manager.

6.3, Unit Staff Qualifications None None None None None 6.4, Procedures None None None None None 6.5, Programs and Manuals None None None None None 6.6, Reporting Requirements L. 1 Removes the requirement to include documentation of challenges to the N/A 6.9. 1.c 2

safety relief valves or safety valves in the monthly operating report.

6.7, High Radiation Area None None None None None Page 1 of 2 I

TABLE L - LESS RESTRICTIVE CHANGES MATRIX DOC #

SUMMARY

REVISED TS CTS SECTION CHANGE SECTION TYPE Current Specification 6.4, Training None None None None None Current Specification 6.5, Review and Audit None None None None None Current Specification 6.6, Reportable Event Action None None None None None Current Specification 6.7, Safety Limit Violation None None None None None Current Specification 6.10, Record Retention None None None None None Current Specification 6.13, Fire Protection Inspection None None None None None Current Specification 6.15, Iodine Monitoring None None None None I

None CHANGE TYPE

1. Relaxation of the administrative requirement.

2. Elimination of CTS reporting requirement.

Page 2 of 2

TABLE R - RELOCATED SPECIFICATIONS AND REMOVAL OF DETAILS MATRIX REVISED TS CTS

SUMMARY

SECTION SECTION AND DOC#

6.1, Responsibility 6.1 - LA. 1 6.1.1, Replaces the specific title "Plant Manager" with the 6.5.2.3, generic title "plant manager" and relocates the specific 6.5.2.5 title.

6.2. Orgranization 6.2 - LA. 1 6.2. 1.a, Replaces the specific title "Plant Manager" with the UFSAR 10 CFR 50 2

6.2. 1.b, generic title "plant manager," replaces the specific title Appendix B 6.2.1.c, "Chief Nuclear Officer" with the generic title "a programs 6.2.2.h specified corporate officer," and relocates the specific titles.

6.2 - LA.2 6.2.2.a, Details of the minimum shift crew requirements.

UFSAR 10 CFR 50 2

6.2.2.b, Appendix B 6.2.2.e, programs Table 6.2-1 6.2 - LA.3 6.2.2.c, Requirements for at least two licensed Operators in the UFSAR 10 CFR 50 2

Table 6.2-1 control room during reactor startup, scheduled reactor Appendix B including shutdown, and during recovery from reactor trips; two programs Note (4) licensed Operators in hot shutdown; and only one Senior Operator and one Operator for cold shutdown and refueling conditions.

6.2 - LA.4 6.2.2.e, Staffing requirements during power operations or hot Site Emergency 10 CFR 50.54(q) 2 Table 6.2-1 shutdown and when the emergency plan is activated.

Plan Note (7)

Page 1 of 5

TABLE R - RELOCATED SPECIFICATIONS AND REMOVAL OF DETAILS MATRIX REVISED TS CTS

SUMMARY

LOCATION CHANGE CHANGE SECTION SECTION CONTROL TYPE AND DOC #

PROCESS 6.2 - LA.5 6.2.2.f Details that require all Core Alterations to be UFSAR 10 CFR 50 2

supervised by either a licensed Senior Reactor Operator Appendix B or Senior Reactor Operator Limited to Fuel Handling; programs and the details that require all fuel moves be directly monitored by a member of the reactor analyst group.

6.2 - LA.6 6.2.2.h Details of working hour limits for personnel who Administrative 10 CFR 50 2

perform safety-related functions.

Procedures Appendix B programs 6.2 - LA.7 6.2.2.i Details of the operator license requirements for the UFSAR 10 CFR 50 2

specific positions of Station Shift Supervisor Nuclear Appendix B and Assistant Station Shift Supervisor Nuclear, and the programs CTS requirement that only licensed individuals may direct licensed activities.

6.3, Unit Staff Qualifications 6.3 - LA. 1 6.3.1 Replaces the specific title "Manager Radiation UFSAR 10 CFR 50 2

Protection" with the generic title "radiation protection Appendix B manager" and relocates the specific title.

programs 6.4, Procedures 6.4 - LA. 1 6.8.1, The details of procedure reviews and approvals Quality 10 CFR 50.54(a) 1 6.8.2, including temporary changes.

Assurance 6.8.3 Topical Report (UFSAR Appendix B) 6.5, Programs and Manuals None None None None None None Page 2 of 5

TABLE R - RELOCATED SPECIFICATIONS AND REMOVAL OF DETAILS MATRIX REVISED TS CTS

SUMMARY

LOCATION CHANGE CHANGE SECTION SECTION CONTROL TYPE AND DOC #

PROCESS 6.6, Reporting Requirements 6.6 - LA. 1 6.9.1.a The details associated with the Startup Report UFSAR 10 CFR 50.59 1

specification.

6.6 - LA.2 6.9.1.e The details regarding changes to the Process Control UFSAR 10 CFR 50.59 1

Program.

6.6 - LA.3 6.9.2 The details contained in CTS 6.9.2, "Fire Protection UFSAR Operating License 1

Program Reports."

Paragraph 2.D(7) 6.7, High Radiation Area 6.7 - LA. 1 6.12.1, Replaces the specific title "Manager Radiation UFSAR 10 CFR 50 2

6.12.2 Protection" with the generic title "radiation protection Appendix B manager" and relocates the specific title.

programs Current Specification 6.4, Training None - LA. 1 6.4.1 The details on training and replacement training for the UFSAR 10 CFR 50 2

facility staff.

Appendix B programs None - LA.2 6.4.2 The details of the Fire Brigade training program.

UFSAR 10 CFR 50 2

Appendix B programs Current Specification 6.5, Review and Audit None - LA. 1 6.5 The details of the Review and Audit specification.

Quality 10 CFR 50.54(a) 2 Assurance Topical Report (UFSAR Appendix B)

Page 3 of 5

TABLE R - RELOCATED SPECIFICATIONS AND REMOVAL OF DETAILS MATRIX REVISED TS CTS

SUMMARY

SECTION SECTION AND DOC#

(7irrmnt Snecification 6.6. Renortable Event Action None - LA. 1 6.6.1.b The requirements of CTS 6.6. 1.b; Reportable Events Quality 10 CFR 50.54(a) 2 reviews by SORC and submittal of the results of the Assurance reviews to the SRAB and the Vice President - Nuclear Topical Report Generation.

(UFSAR Appendix B)

Current Specification 6.7, Safety Limit Violation None - LA. 1 6.7.1.b, The requirement for notification of the Vice President -

Quality 10 CFR 50.54(a) 2 6.7.1.c, Nuclear Generation and the SRAB in the event of a Assurance 6.7.1.d Safety Limit violation, the requirement for SORC to Topical Report review the Safety Limit Violation Report, and the (UFSAR requirement to submit the Safety Limit Violation Appendix B)

Report to the SRAB and the Vice President - Nuclear Generation.

Current Specification 6.10, Record Retention None - LA. 1 6.10 The details contained in the Record Retention Quality 10 CFR 50.54(a) 2 specification.

Assurance Topical Report (UFSAR Appendix B)

Current Specification 6.13, Fire Protection Ins ection None - LA. 1 6.13 The details contained in the Fire Protection Inspection Quality 10 CFR 50.54(a) 2 specification.

Assurance Topical Report (UFSAR Appendix B)

Page 4 of 5

TABLE R - RELOCATED SPECIFICATIONS AND REMOVAL OF DETAILS MATRIX CHANGE TYPE

1. Procedural details for meeting TS requirements and related reporting requirements.

2. Relocated administrative controls requirement.

Page 5 of 5 REVISED TS CTS

SUMMARY

LOCATION CHANGE CHANGE SECTION SECTION CONTROL TYPE AND DOC#

PROCESS Current Specification 6.15, Iodine Monitoring None - LA. 1 6.15 The details contained in the Iodine Monitoring UFSAR 10 CFR 50.59 2

specification.

v t e Letter Sequence Response to RAI
TAC:MB2441, TSTF-65, Implement 10 CFR 50, Appendix J, Option B, TSTF-37 (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions 4 April 2002 30 August 2002 Answers 7 June 2002 Results Approval Other: ML020740208, ML020870527, ML022210090, ML031140140
MONTHYEAR2002-06-07 [Table View]

ML021710054

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Subject:

Response

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