Text

Attachment 4.9, Nine Mile Point Unit 1, Facility Operating License DPR-63, License Amendment Request: Current Technical Specifications Markup and Discussion of Changes, Miscellaneous Page Changes
	ML023390027
Person / Time
Site:	Nine Mile Point
Issue date:	11/22/2002
From:	Conway J; Constellation Energy Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NMP1L 1698, TAC MB2441
	Download: ML023390027 (97)
	v • d • e

ATTACHMENT 4.9 Current Technical Specifications Markup and Discussion of Changes Miscellaneous Page Changes

1.28 (Deleted) 1.29 (Deleted) 1.30 Reactor Coolant Leakage

a.

Identified Leakage (1)

Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered and conducted to a sump or collecting tank, or (2)

Leakage into the primary containment atmosphere from sources that are both specifically located and known not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.

b.

Unidentified Leak*ge All other leakage of reactor coolant into the primary containment area.

1.31 Core Ogeratina 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 Specification Plant operation within these operating limits is addressed in individual specifications.

AMENDMENT NO. Ui,

/

8 1

\\

Par I J 4

J'hScAI evc.cu

SAFETY LIMIT Written procedures will be developed and followed whenever the reactor water level is lowered below the low-low level set point (5 feet below minimum normal water level). The procedures will define the valves that will be used to lower the vessel water level. All other valves that have the potential of lowering the vessel water level will be identified by valve number in the procedures and these valves will be red tagged to precltJde their operation during the major maintenance with the water level below the low-low level set point.

In addition to the irequirements e-

.E E

tp:*.

,,......... there shall be another control room operator present in the control room with no other duties than to monitor tereactor vessel water level.

4h0 F o,+

e... vt ovie-ic.eA.ec4 O-aecr,,c AMEND E fNO conoeol room e*

_vi AMENDMENT NO. 4 U.i /

LIMITING SAFETY SYSTEM SETTING

b.

The IRM scram trip setting shall not exceed 12%

of rated neutron flux for IRM range 9 or lower.

The IRM scram trip setting shall not exceed 38.4% of rated neutron flux for IRM range 10.

c.

The reactor high pressure scram trip setting shall be < 1080 psig.

d.

The reactor water low level scram trip setting shall be no lower than -12 inches (53 inches indicator scale) relative to the minimum normal water level (302'9").

e.

The reactor water low-low level setting for core spray initiation shall be no less than -5 feet (5 inches indicator scale) relative to the minimum normal water level (Elevation 302'9").

f.

The reactor low pressure setting for main-steam line isolation valve closure shall be ?850 psig when the reactor mode switch is in the run position or the IRMs are on range 10.

g.

The main-steam-line isolation valve closure scram setting shall be < 10 percent of valve closure (stem position) from full open.

11 P0r; C~

t.of

(A2 LIMITING CONDITION FOR OPERATION 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Applicability:

Applies to the operating status of the system of isolation valves on lines connected to the reactor coolant system.

Objective:

To assure the capability of the reactor coolant system isolation valves to minimize reactor coolant loss in the event of a rupture of a line connected to the nuclear steam supply system.

Specification:

a.

During power operating conditions whenever the reactor head is on, all reactor coolant system isolation valves on lines connected to the reactor coolant system shall be operable except as specified in "b" below.

b.

In the event any isolation valve becomes inoperable the system shall be considered operable provided at least one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition, except as noted in Specification 3.1.1.e.

AMENDMENT NO. i, i.,

/4 SURVEILLANCE REQUIREMENT ISURVEILLANCE-RE----

EME-4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Applicability:

Applies to the periodic testing requirement for the reactor coolant system isolation valves.

Objective:

To assure the capability of the reactor coolant system isolation valves to minimize reactor coolant loss in the event of a rupture of a line connected to the nuclear steam supply system.

Specification:

The reactor coolant system isolation valves surveillance shall be performed as indicated below.

a.

At least once per operating cycle the operable automatically initiated power-operated isolation valves shall be tested for automatic initiation and closure times.

b.

Additional surveillances shall be performed as required by Specification 1o8J&

),s c-r-1 I a vi e-o,,s

BASES FOR 3.2.7 AND 4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES The list of reactor coolant isolation valves is contained in the procedure governing controlled lists and have been removed from the Technical Specifications per Generic Letter 91-08. Revisions will be processed in accordanccewith w..

Double isolation valves are provided in lines which connect to the reactor coolant system 4o n

minimize reactor coolan loss in the event of a line rupture. The specified valve requirements assure that isolation is already accomplished with one valve shut or provide redundancy in an open line with two operative valves.

Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable of automatic initiation. Valve closure times are selected to minimize coolant losses in the event of the specific line rupturing and are procedurally controlled.

Using the longest closure time on the main-steam-line valves following a main-steam-line break (Section XV C.1.0)(1), the core is still covered by the time the valves close. Following a specific system line break, the cleanup and shutdown cooling closing times will upon initiation from a low-low level signal limit coolant loss such that the core is'not uncovered. Feedwater flow would quickly restore coolant levels to prevent clad damage. Closure times are discussed in Section VI-D. 1 (1).

The valve operability test intervals are based on periods not likely to significantly affect operations, and are consistent with testing of other systems. Results obtained during closure testing are not expected to differ appreciably from closure times under accident conditions as in most cases, flow helps to seal the valve.

-75 2

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 (Fifth Supplement, p. 11 5)(2) that a line will not isolate. Additional surveillances are in accordance with the Inservice Testing Program described in Specificationo (1)

UFSAR (2)

FSAR AMENDMENT NO. 4-42, F-14&

RW.SO 4

115J

(Dj LIMITING CONDITION FOR OPERATION 3.3.3 LEAKAGE RATE APPlicability:

Applies to the allowable leakage rate of the primary containment system.

Obiective:

To assure the capability of the containment in limiting radiation exposure to the public from exceeding values specified in 10 CFR 100 in the event of a loss of-coolant accident accompanied by significant fuel cladding failure and hydrogen generation from a' metal-water reaction.

To assure that periodic surveillances of reactor containment penetrations and isolation valves are performed so that proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating primary containment.

Specification:

Whenever the reactor coolant system temperature is above 215OF and primary containment integrity is required, the primary containment leakage rate shall be limited to:

AMENDMENT NO. Ui. i1,/0' M'IS ~e Ia vi1e LS SURVEILLANCE REQUIREMENT 4.3.3 LEAKAGE RATE Aoolicabilitv:

Applies to the primary containment system leakage rate.

Objective:

To verify that the leakage from the primary containment system is maintained within specified values.

Specification:

a.

The primary containment leakage rates shall be demonstrated at test schedules and in conformance with the criteria specified in the 10 CFR 50 Appendix J Testing Program Plan as described in Specification&

b.

The provisions of Specification 4.0.1 are not applicablej and the surveillance interval extensions are in accordance with the 10 CFR 50 Appendix J Testing Program Plan.

131 I A

MA iscsJ-Ia newws BASES FOR 3.3.4 AND 4.3.4 PRIMARY CONTAINMENT ISOLATION VALVES The list of primary containment isolation valves is contained in the procedure governing controlled lists have been removed from the Technical Specifications per Generic Letter 91-08. Revisions will be processed in accordance with Double isolation valves are provided on lines penetrating the primary containment and open space o t e contament.

osure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable of automatic initiation. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Section VI-D.(1I For allowable leakage rate specification, see Section 3.3.3/4.3.3.

For the design basis loss-of-coolant accident fuel rod perforation would not occur until the fuel temperature reached 1700OF which occurs in approximately 100 seconds.(2) The required closing times for all primary containment isolation valves are established to prevent fission product release through lines connecting to the primary containment.

For reactor coolant system temperatures less than 215 0 F, the containment could not become pressurized due to a loss-of-coolant accident.

The 2150 F limit is based on preventing pressurization of the reactor building and rupture of the blowout panels.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10-7 that a line will not isolate (Fifth Supplement, p. 115).(3) More frequent testing for valve operability results in a more reliable system.

In addition to routine surveillance as outlined in Section VI-D. 1.0(1) each instrument-line flow check valve will be tested for operability. All instruments on a given line will be isolated at each instrument. The line will be purged by isolating the flow check valve, opening the bypass valves, and opening the drain valve to the equipment drain tank. When purging is sufficient to clear the line of non-condensibles and crud the flow-check valve will be cut into service and the bypass valve closed. The main valve will again be opened and the flow check valve allowed to close. The flow-check valve will be reset by closing the drain valve and opening the bypass valve depressurizing part of the system. Instruments will be cut into service after closing the bypass valve. Repressurizing of the individual instruments assures that flow-check valves have reset to the open position.

(1)

UFSAR (2)

Nine Mile Point Nuclear Generation Station Unit 1 Safer/Corecool/GESTR-LOCA Loss of Coolant Accident Analysis, NEDC-31446P, Supplement 3, September, 1990.

(3)

FSAR AMENDMENT NO. ~

44150

)

DISCUSSION OF CHANGES REVISED TS: MISCELLANEOUS PAGE CHANGES ADMINISTRATIVE (A)

A. 1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 Details of the minimum shift crew requirements located in CTS 6.2.2.b are proposed to be relocated to the UFSAR. The reference to CTS 6.2.2.b on TS Page 11 is replaced by stating the CTS 6.2.2.b requirement; i.e., that at least one licensed Operator be in the control room when fuel is in the reactor. Technical changes to minimum shift crew requirements are addressed in the Discussion of Changes for Revised TS 6.2.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" None "Specific" None Page 1 of 1

ATTACHMENT 4.10 Current Technical Specifications Markup and Discussion of Changes CTS 6.4 Training

DISCUSSION OF CHANGES CTS: 6.4 - TRAINING ADMINISTRATIVE (A)

None TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L. LA)

"Generic" LA. 1 CTS 6.4.1 discusses the training and replacement training program for the facility staff.

The proposed change would relocate the details of this training program to the UFSAR.

These training provisions are adequately addressed by other proposed TS Section 6.0 provisions and by regulations. Revised TS 6.3, "Facility Staff Qualifications," provides requirements to assure adequate, competent staff in accordance with ANSI/ANS N18.1 1971 and Regulatory Guide 1.8, September 1975. Revised TS 6.2 details facility staff requirements. Revised TS 6.2.2.a and 10 CFR 50.54 state minimum shift crew requirements. Training and requalification of licensed positions is contained in 10 CFR Part 55. Thus, the relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

LA.2 CTS 6.4.2 discusses the training program for the Fire Brigade. The proposed change would relocate the details of this training program to the Fire Hazards Analysis (UFSAR Appendix 10A). The Fire Protection requirements have previously been relocated to the UFSAR in accordance with Generic Letter 88-12; therefore, the fire brigade requirement with respect to training is not needed in the TS. The relocated requirements will assure an adequate training program is maintained in accordance with NMP1 commitments and regulations. As such, these relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

"Specific" None Page 1 of 1

ATTACHMENT 4.11 Current Technical Specifications Markup and Discussion of Changes CTS 6.5 Review and Audit

Smeet 6.3. FaiiySaff mebroQh ntsafsal rece h

iiu ualifications o

NIN8117 o

oprbe 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; the Manager Radiation Protection who shall meet or exceed the qualifications qf Regulatory Guide 1.8, September 1975; and the Shift Technical Advisor who shall have a bachelor's degree in a physical science or engineering or a professional engineer license issued by examination and shall have received specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the ManagerT-Fr Training and shall meet or exceed the recommendations and requirements of Section 5.5 of ANSI N18.1-1971 and of iOCFR Part 55, and shall include familiarization with relevant industry operational experience.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager Training and Supervisor-Fire Protection, Nuclear and shall meet or exceed the requirements of Appendix R to 10CFR50.

ý_6.5 BeviManagerdi CLA.ist Member:

M n aev w

Rm PtOR S6.5.1.1 The Station Opera

ns Review Committe /shall function to a ise the Plant Manage yon all matters related to

./

nuclear safety.

_ComooSir*

n../*

6...

Th SO shl/ecmoe e

Ch ' n: -

la t a e poLýa I

AMENDMENT NO.

3/5I 351

Qr reAA 5e C*o-6.5 A 1s*......

k AMENDMENT NO. i*,

il1, /135 6.5

.3 All alternate embers shall be appoi ed. in writing by the SOR Chairman or Vice-Chair n to serve on a temporary asis; however, no mor than two alternates shall articipate as voting me ers in SORC activities any one me.

M tin Frequenc 6.5.1.4 T e SORC shall meet a east once per calendar nth and as convened by e SORC Chairman, Vi

-Chairman, or designated alternat 6.1.5 The quorum the. SORC necessary f the performance of the RC responsibility and uhority provisions of these Tech cal Specifications shall onsist of the Chairman, oa Vice-Chairman, and ur members, including alternate R ponsibilities 6.5.1.6 t e SOR shall be respo ible for:

a.

Review of all RE RTABLE EVENTS.

b. Review of u it operations to detect tential safety hazards.

c.

Perform nce of special reviews' vestigations or analyse and reports thereon as quested by the Plant Mana er or the Safety Review nd Audit Board.

d. I estigation of violations f the Technical Specifi tions and shall prepare nd forward a report co ring evaluation and recomm dations to prevent rec rence to the Vice Presi ent - Nuclear Generatio and to the Safety Review and A it Board.

paýe_ 'ý C-P 9

352 LA-i

Curre't'+/- Secc

,.5 V;id

-Nuclear Generation an he.Safety Review dlat Manager; however, e Plant Manager shall pursuant to 6.1.1 above

/

AMENDMENT NO./L A'*I,/'

353 "I e 3 CW-7

p ure 1 ar safetys b dsin by ii 2aqual*

e indivgidual/or nization.

Eac such mod!ifition shall by.feviewe byn nividual//oup other t n the indi rual/group whief designed the *dftonJu

h miyefo h ~'Aoa~iatFn*,*,I Sgroup whih ds*n'=

h oii.nf'rpsed modiffic-ations to structures, systems and components and te

safety-e-valuations shallbe approved prior to implementation by the Plant Manager; or the Manager TechnicalII t

=

Support. as previously designated by the Plant Manager, Lr "V

0_1=v.=_,'

-(,-

I

.5

.4 lndiviidals responsibi for reviews p formed in acc ance with Sp cifications 6.5./.1, 6.5.2.2 and 6.5.2.3 shall be n"mbers of the tation supervi ry staff, previ sly designate by the Plant nager to perfo such reviews Eeh such revie

'shall include a etermination whether or n additional, cro s-disciplinary, view is necess y,

deemed nec sary such revi w shall be per rmed by the a propriate desig ted station r iew personnel.

tfdeemed n sr S-

~

sha ired a, propriate gd edi st*

e/ie re.Yi roposed tests and experiments which affect station nuclear safety and are not addressed in the FSAR or Technical SSpecifications and their safety evaluations shall be reviewed by the Plant Manager, or the Manager Technical

Support as previously designated by the Plant Manager.

-hov_-'

,/

6.5.2.

The Plant anager shall ass the performance opspecial reviews andi vestigations, and the repara ion anh submit f reports thereo;, as requested byetVice President "Nu ar Generation.

.5.2.7 Th acility security p gram, and impleme ing procedures, shal e reviewed at least very 12 months.

commended cha es shall be approv by the Plant Manag and transmitted to e Vice President - N lear Generation and the Chairman of th afety Review and it Board.

6.5..8 The facility ergency plan, and plementing proced s shall be reviewe t least every 12 mo hs.

Recomme ed changes shall approved by the PI t Manager and tran itted to the Vice Pr ident - Nuclear Generat' n and to the Chair an of the Safety Re w and Audit Board LAE" ENT NO,/

of

~4 4

354 AMENDM

" 6.5.2.9 h Pl,,'ant Mvanager ala,,ssure the,,ero. no of areview.y a qualified individual/,gaiz,,o,, of changes 'the, Railoia l wat reatment systems./

6o5..10 Review of an a~ccidental, unplanne, or uncontrolledi adioactive release incl ing the preparation of r orts

/

covering ev/auation, recommend fons and disposi *n of the corrective ac i n to prevent recurreenc /nd the forwardi of these reports toe Vice Pr~esiden Nuclear Generation a to the Safety Review/ai Audit Board.

6.5.2.1 1 Revi* of changes to the/Process Contro! 'PI gram and t~he Offsite

~ose Calculation Manual./

pproval of any

"~c 652*nges shall be made the Plant M~an er or_ his d esignee bfe implementation of suc changes.

/.5.2.13 The; Pa nger sha~llas ateperforma~nce a ireview by a~qualified dvdal/organization o eFr 6.;.3 Saf tvoR t iSafety Review a Audit Board.

6..

Saet Re' ew an ui Ba (

Function

.3. 1. T h e S a f e t e i w a d A d t o r h l u c i n t i e i d p n e t r n

u i f d s g a e a t v t e

/

~in the aas of:

/,/

¢

~a. //uclear power pla~t operations

i.

t~~~hemsry apporate ie asoc iatedwt nqechrceity~f h

ula oe AMENMENTNO, t~ld metllurgy*Q'ie

'*=l*

5

Curr&rt Se&o

65 Meobber:

i~mber:

ember:

Member:

>flview and Audit Board all be comi i: Staff Engineer or nager or Vice Preý Plant Manager o esignee Staff Enginee - Nuclear Staff Engi or - Mechanical o lectric Consult t (See 6.5.3.4)

'the SRAB Chairman xserve on a temporary activities at any orW time.

1.6 The quorum the SRAB necessary f the performance of tIfRAB review and Specificati s shall consist of not I rs than a majority of tji( members, including the pres ce of the Chairman orte Chairman's designp(ed alternate and norp&r shall ve line responsibility f operation of the facj. y.

AMENDMENT NO. X#

pý.c(

3 56 356

Review

.5.3.7 The AB shall review:

The safety eval ions for 1) changes procedures, equipm or systems and 2) sts or experiments completed u

~r the provision of S tion 50.59, 10 CFR, verify that such act' ns did not constitute anue unreviewe.safety question.

b.

Prop *d changes to proce res, equipment or sys ms which involve a nreviewed safety ques *on as d e ad in Section 50.5 10 CFR.

c.

Proposed tests or periments which invol an unreviewed safe question as defined in ction 50.59, 10 CFR.

d.

Proposed nges in Technical S ifications or operatin icense.

a.

Viola ns of codas, regulati s, orders, Technical ecifications, license re rements, or of intern pr edures or instruction aving nuclear safet gnificance.

f.

Significant operatin bnormalitias or devi ons from normal and a ectad performance of ant equipment A

35that affect nu7esafety.

ag. _-All REPORT L

VNS strco urs ny mo opnn 6-C.

XI 0-e--V~j

i.

Reports and metin iue of the

/S

)

Curr-ed-Spec.Accc~a, 6,57 Audits

.5.3.8 Audits o acuity activities s 11 be performed under e cognizance of the AB. These audits s I encompass:

a.

he conformance facility operation to provisions contained ithin the Technical S cifications and applicable licen conditions at least o e per year.

b.

The perfor ance, training and qu ifications of the entire acility staff at least o a per year.

c.

The r ults of actions taken ocorrect deficiencies curring in facility eq ment, structures, syst s or m

od of operation that fact nuclear safety at ast once per six mo

s.

d. The performance of activities required b a Quality Assuranc rogram to meet the cri nia of Appendix "B", "OCFR5O, at ast once per two yea

e.

The Facility ergency Plan and im ementing procedures least once every 12 nths.

f.

The Fac' ty Security Plan and plementing procedure at least once every 12 onths.

g.

Th acility Fire Protectio Program and implame ing procedures at least nce per two years.

h. Any other area of fa ity operation consider appropriate by the S or the Vice President -

uclear Generation.

i.

The radiologi I environmental monit ng program and the r ults thereof at least onc per 12 months.

j.

The Off e Dose Calculation Ma al and implementing7ocedures at least once8r 24 months.

Sk.

T.Process Control Progra and implementing pr edures for processing dpackaging of rdioa ive wastes AMENDMENT NO. ;*,

gt-358

Cair~revcE 5rIQLACJAOV 4*1 AMENDMENT NO. WL, #13 Rap-I dp 7 359

DISCUSSION OF CHANGES CTS: 6.5-REVIEW AND AUDIT ADMINISTRATIVE (A)

A.1 The requirements of CTS 6.5.2.3 and CTS 6.5.2.5 regarding Plant Manager reviews and approvals are proposed to be moved to Revised TS 6.1, "Responsibility." Technical changes to these requirements are addressed in the Discussion of Changes for Revised TS 6.1.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA.1 CTS 6.5.1 describes the Station Operations Review Committee (SORC) review and audit requirements; CTS 6.5.2 describes technical review and control requirements; and CTS 6.5.3 describes the Safety Review and Audit Board (SRAB) review and audit requirements. The proposed change would relocate these requirements to the QATR, which is contained in the UFSAR as Appendix B. These changes are consistent with the guidance of Administrative Letter (AL) 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," dated December 12, 1995, and NUREG-1434, Revision 1. The administrative letter concluded that TS administrative quality assurance-related requirements may be relocated to licensee-controlled quality assurance programs. For NMP1, these requirements would be relocated in their entirety (except as noted below) to the QATR, with only minor wording and formatting changes.

Requirements relating to review and audit activities described in CTS 6.5 are contained in 10 CFR 50.54(p); 10 CFR 50.54(t); 10 CFR 50 Appendix B, Criterion XVIII; 10 CFR 73; ANSI/ANS 3.2-1982; ANSI N18.7-1972; and ANSI/ASME NQAl-1983, including 1983 Addenda. Relocation of these TS provisions to the QATR will provide adequate controls over review and audit activities for NMP1. Thus, the provisions are not necessary to be in the TS to provide adequate protection of the public health and safety. Changes to the QATR are controlled by the provisions of 10 CFR 50.54(a).

Minor Differences Between CTS and the QATR There are several minor differences between the current wording of CTS 6.5 and the existing wording in the QATR, as described below (italics added to highlight differences). These minor differences, which are shown on the marked-up TS pages, will be evaluated in accordance with the NMPC administrative procedures that implement 10 CFR 50.54(a).

1. CTS 6.5.1.8 - The TS states that SORC shall maintain written minutes of each meeting, whereas the QATR specifies that "The SORC shall maintain written minutes Page 1 of 2

DISCUSSION OF CHANGES CTS: 6.5-REVIEW AND AUDIT of each meeting that, at a minimum, document the result of all SORC activities performed under the responsibilities and authority provisions of the Technical Specifications and this section."

2. CTS 6.5.3.7.h -The TS states that SRAB shall review "Any indication of' an unanticipated deficiency, whereas the QATR specifies that SRAB shall review "All recognized indications of" an unanticipated deficiency. Also, the scope of the TS requirement covers "safety related structures, systems, or components," whereas the QATR specifies "structures, systems, or components that could affect nuclear safety."

3. CTS 6.5.3.10.a - The TS requires that minutes of each SRAB meeting shall be prepared, approved and forwarded to the Chief Nuclear Officer within 30 days following each meeting, whereas the QATR requires 14 days for completion of these activities.

4. CTS 6.5.3. 1O.b - The TS requires that reports of certain SRAB reviews be prepared, approved and forwarded to the Chief Nuclear Officer within 14 days following completion of the review. The QATR includes one additional SRAB review within the scope of this requirement, that being: "Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59."

5. CTS 6.5.3.10.c - The TS states that SRAB audit reports shall be forwarded to the Chief Nuclear Officer within 90 days following completion of the review, whereas the QATR requires that SRAB audit reports be forwarded to the Chief Nuclear Officer and to the management positions responsible for the areas audited within 30 days following completion of the audit by the auditing organization."

"Specific" None Page 2 of 2

ATTACHMENT 4.12 Current Technical Specifications Markup and Discussion of Changes CTS 6.6 Reportable Occurrence Action

DISCUSSION OF CHANGES CTS: 6.6 - REPORTABLE OCCURRENCE ACTION ADMINISTRATIVE (A)

A.1 CTS 6.6. l.a delineates NRC notification and report submittal requirements for Reportable Events. The proposed change would delete CTS 6.6.1.a. The notification and report submittal requirements of CTS 6.6.1.a are contained in 10 CFR 50.72 and 10 CFR 50.73. There is no need to repeat these requirements in the TS. Since these requirements are contained in the regulations, and since the NMP1 Operating License requires compliance with 10 CFR 50, deletion of this requirement from the TS is considered administrative in nature.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA. 1 CTS 6.6.1.b describes SORC responsibilities regarding the review of Reportable Events and submittal of the results of the reviews to the SRAB and the Vice President-Nuclear Generation. The proposed change would relocate the requirements of CTS 6.6.1.b to the QATR. The requirements of CTS 6.6.1.b duplicate the SORC responsibilities given in CTS 6.5.1.6.a and CTS 6.5.1.6.d, which are proposed for relocation to the QATR. These activities are required following the event without a specified completion time. As such, the proposed relocated requirements are not necessary to assure operation of the facility in a safe manner, and are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the QATR are controlled by the provisions of 10 CFR 50.54(a).

"Specific" None Page 1 of 1

ATTACHMENT 4.13 Current Technical Specifications Markup and Discussion of Changes CTS 6.7 Safety Limit Violation

6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified and a report submitted pursuant to the requirements of Sections 50.72 and 50.73 to

b. -E -ach R REPORT ABLE EVENT shall b oo reviewed by the SORC and the results of this review submitted to the,

SRAB and the e Vice President - Nuclear Generation.

d 6.8 Procedures 6.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix "A" of USAEC Regulatory Guide 1.33 except as provided in 6.8.2 and 6.8.3 below.

a.

Written procedures' shall be established, implemented, and maintained for activities involving the Fire Protection Program implementation.

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed and approved prior to implementation by the branch manager for the functional area of the procedure or higher levels of management as governed by administrative procedures. Each procedure and administrative policy of 6.8.1 above shall be reviewed periodically asset forth AMENDMENT NO. W, W,

191. /-

All ReA-e/1 S~aj 360 t PoeIc4

/

DISCUSSION OF CHANGES CTS: 6.7 - SAFETY LIMIT VIOLATION ADMINISTRATIVE (A)

A. 1 The proposed change would delete the Safety Limit Violation requirements of CTS 6.7 as they relate to NRC notification (CTS 6.7.1.a, and portions of CTS 6.7.1.b, CTS 6.7.1.c, and CTS 6.7.1.d). These requirements are contained in and based upon the requirements located in 10 CFR 50.36(c)(1), 10 CFR 50.72, and 10 CFR 50.73. Since the NMP1 Operating License requires compliance with 10 CFR 50, there is no need to repeat these requirements in the TS. Deletion of these requirements from the TS is considered administrative in nature.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA. 1 The CTS 6.7. 1.b requirement for notification of the Vice President - Nuclear Generation and the SRAB in the event of a Safety Limit Violation; the CTS 6.7.1.c requirement for SORC to review the Safety Limit Violation Report; and the CTS 6.7.1.d requirement to submit the Safety Limit Violation Report to the SRAB and the Vice President - Nuclear Generation are proposed to be relocated to the QATR. Given that the notification occurs following the Safety Limit Violation and that the Safety Limit Violation Report is an after-the-fact report, the proposed relocated requirements are clearly not necessary to assure operation of the unit in a safe manner and are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the QATR are controlled by the provisions of 10 CFR 50.54(a).

"Specific" None Page 1 of I

ATTACHMENT 4.14 Current Technical Specifications Markup and Discussion of Changes CTS 6.10 Record Retention

6. 10 Record holoRetention F

hl *

=

tyas

/

6.10.1 ing record ained for a

/ a.

Records and. log f-facilit~y operation vering. time interval each power level".*

b.

Records an ogs of principal mai enance activities, in ections, repair and replac ent of principal items of quipment related t uclear safety.

c.

REP TABLE EVENT REP RTS.

d.

ecords of surveilla e activities, inspectio and calibrations required these Technical Speci ations.

e.

Records of rea r tests and experim

s.

f.

Records o hanges made to Op ting Procedures.

"g. Recor-s of radioactive ship nts.

h.

cords of sealed sour leak tests and results.

Records of annual ysical inventory of all so ce material of record.

6.

.2 The following rec s shall be retained for th duration of the Facility perating License:

a.

Record a drawing changes reflec ng facility design modifi tions made to systems d equipment described in t Fi.nal fety Analysis Report.

b.

cords of new and irradia dfuel inventory, fuel tr sfers and assembly bur p h istories.

Records of facility-rad* tion and contamination rveys.

d.

Records of radia n exposure for all indiv uals entering radiation c trol areas.

I AMENDMENT NO.

I 370 LA -/

pag C I --ý

e.

R ords of gaseo and liquid radio tive material rele ad to the environs.

f Records of tr sient or operatio I cycles for those f cility components d igned for a limited nu ber of transients or cycles.

g.

Reco s of training and ualification for curre members of the pl t staff.

h.

cords of in-servi inspections perform pursuant to these chnical Specificati

s.

/

Records of 0 ity Assurance activiti required by the QA anual.

j.

Records reviews performed f changes made to pr edures or equipm t or reviews of test and experiment pursua to 10 CFR 50.59.

k.

R or tings of th SOR and the SRA 6.11 Offsite Dose Calculation Manual (ODCM)

a.

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gases and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b.

The ODCM shall also contain the radioactive'effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 6.9.1.d and Specification 6.9.1.e.

c.

Licensee initiated changes to the ODCM:

1.

Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

(a)

Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b)

A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; AMENDMENT N0. iis

.371

'Pr-Irayn, an.1 MavtxuPnJ3Y

DISCUSSION OF CHANGES CTS: 6.10 - RECORD RETENTION ADMINISTRATIVE (A)

None TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L& LA)

"Generic" LA. 1 CTS 6.10 delineates records retention requirements, including those records that are to be retained for at least five years (CTS 6.10.1) and those records that are to be retained for the duration of the facility operating license (CTS 6.10.2). The proposed change would relocate the requirements of CTS 6.10 to the QATR. These changes are consistent with the guidance of AL 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," dated December 12, 1995, and NUREG-1434, Revision 1. The administrative letter concluded that TS administrative quality assurance related requirements may be relocated to licensee-controlled quality assurance programs.

For NMP1, these requirements would be relocated in their entirety to the QATR, with changes only to the format. Records retention requirements related to activities affecting quality are contained in 10 CFR 50 Appendix B, Criterion XVII, and other sections of 10 CFR 50 that are applicable to NMP1 (e.g., 10 CFR 50.71, 10 CFR 73, etc.). These records retention requirements provide a record of certain activities important to plant safety, but the records themselves do not assure safe operation of the facility since review of these records is a post-compliance review. Relocation of these TS provisions to the QATR will provide adequate controls over records retention requirements for NMP1. As such, the relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the QATR are controlled by the provisions of 10 CFR 50.54(a).

"Specific" None Page 1 of 1

ATTACHMENT 4.15 Current Technical Specifications Markup and Discussion of Changes CTS 6.13 Fire Protection Inspection

6,13 F'

Protection Ins ctiop.

e /

LA.

6.13.1 An in pendent fire protectio and loss prevention i pection and audit sh be performed annu y utilizing either qu ied off-site licensee rsonnel or an outside e protection firm.

6.13,2 An inspection and ait by an outside aua!!fd fire consultant shall e performed at inte s no greater than 3 ars.

6,14 Systems Intearit-RCV ;S -_4

,.S

'Proioi e

aV%

SRyL*'e.4LTS: *,5 "Pr.o*,*

.is L,:/.A~

Procedure shall be established, implemented and maintained to meet or exceed the requirements and recommendations of Section 2.1.6.a of NUREG 0578. The requirements shall apply to the Post Accident Sampling System (PASS) until such time as administrative controls provide for continuous isolation of the associated penetration(s) or a modification eliminates the potential leakage path(s).

6.15 Iodine Monitorin' OF CA,"_

CU$L 1

.y i \\n d m a in e oe o r e x c e e d t h Ae req u ir e >

Procedures shall be established, implemented and maintained to meet or exceed the requirements and recommendations of Section 2.1.8.c of NUREG 0578.

"6.16 10 CPR 50 Appendix J Testingq Progra Plan 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 "Perform ance-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 Topic 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.

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

The maximum allowable primary containment leakage rate (La) at Pac shall be 1.5% of primary containment air weight per day.

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.

g*e-TS I

AMENDMENT NO. 1-42-, 4.-&-

TrS 3

ei~

Mv~cdS PaI

DISCUSSION OF CHANGES CTS: 6.13 - FIRE PROTECTION INSPECTION ADMINISTRATIVE (A)

None TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L. LA)

"Generic" LA.1 CTS 6.13 requires performance of inspections and audits of the fire protection and loss prevention program, to be performed annually utilizing either qualified off-site licensee personnel or an outside fire protection firm (CTS 6.13.1), and at intervals no greater than 3 years by an outside qualified fire consultant (CTS 6.13.2). The proposed change would relocate the requirements of CTS 6.10 to the QATR as activities performed under the cognizance of the Safety Review and Audit Board (SRAB). These changes are consistent with the guidance of AL 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," dated December 12, 1995, and NUREG-1434, Revision 1. The administrative letter concluded that TS administrative quality assurance related requirements may be relocated to licensee-controlled quality assurance programs.

For NMP1, the requirements of CTS 6.13 would be relocated in their entirety the QATR, with changes only to the format. Requirements relating to review and audit activities of the SRAB are contained in 10 CFR 50 Appendix B, Criterion XVIII; ANSI/ANS 3.2 1982; ANSI N18.7-1972; ANSI/ASME NQA1-1983, including 1983 Addenda; and Branch Technical Position ASCSB 9.5-1. Relocation of these TS provisions to the QATR will provide adequate controls over inspection and audit activities relating to the fire protection program for NMP1. As such, the provisions are not necessary to be included in the TS to provide adequate protection of the public health and safety. Changes to the QATR are controlled by the provisions of 10 CFR 50.54(a).

"Specific" None Page 1 of 1

ATTACHMENT 4.16 Current Technical Specifications Markup and Discussion of Changes CTS 6.15 Iodine Monitoring

DISCUSSION OF CHANGES CTS: 6.15 - IODINE MONITORING ADMINISTRATIVE (A)

None TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA.1 CTS 6.15 discusses the iodine monitoring program. The proposed change would relocate the details of this program to the UFSAR. This program is required by the NMP1 commitment to NUREG-0737, Item III.D.3.3 (NUREG-0578, Section 2.1.8.c). This program contains controls to assure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions, and is designed to minimize radiation exposure to plant personnel post-accident. The training aspect of the program is accomplished as part of the continuing training for personnel in the cognizant organizations, as well as during the training for those individuals responsible for implementing the Radiological Emergency Planning procedures. Provisions for monitoring and performing maintenance of the sampling and analysis equipment are addressed in chemistry and radiation protection procedures. Therefore, the relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the UFSAR are controlled by the provisions of 10 CFR 50.59.

"Specific" None Page 1 of 1

ATTACHMENT 5 Improved Standard Technical Specifications (ISTS) (NUREG-1434, Revision 1) Markup and Justifications for Deviation (JFDs)

Note: Changes to the ISTS markup and the JFDs originally submitted by letter dated June 7, 2002 are identified by a vertical bar and an "A" in the right margin.

(er -s>

.0 AI)HINISTRATIVE CONTROLS 0<C rA I>

C4 BWR/6 S.

Th-e Mllant *-s-ha11 be responsible for overall unit operation and shall delegate in writing the succession to this responsibilit during his absence.

The l~ant~gw-ýi-orW'designee shal approve, prior to implementation, each proposed tes~ experimment modification to systems or equipment that affect nuc_ ar safety.

a n COA The hif -Supervisor (tS)Ishall be..responsible fo e control roomtcommand function.

During any abse ce ofthe Sirfrom the control room wh iletthe unit is ' in 2

an individual with an active:Senior Reactor Operator-(SRO)l*ce se s all be

.-~designated to assume the-control room, command function During an absence of the SSfrom-the control "room while the unit is in an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command1function.

~.CAJ S

+

j 4)-& p ower er+AC>I TS Rev 1, 04/07/95

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

Organization

<Cr5s>

e2 ADMINISTRATIVE CONTROLS Organization

.2 SI> #-2 Onsite and Offsite Oraanizations 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

$F-45 be defined and established throughout highest management levels, intermediate levels, and all operating organization,*

(AnSr+ 6.1.- A positions.

These relationships shall be documented and TF-(

updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel posi ions, or in equivalent forms of documentai on.TheeJ shall be documented in the SAR*-i I

b. The lant u in en en Jshall be responsible for overall sa e operation of the p ant and shall have control over s

those onsite activities necessary for safe operation and maintenance of the I C. A r specified corporate shall have

'/ j!corporate responsibility forovera 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 I>

d.

The individuals who train the operating staff, carry out or perform quality assurance functions may r+

report to the appropriate onsite manager; however, these individuals shallihave sufficient organizational freedom to ensure their independence from operating pressures.

Unit Staff

<T(.,2-I>

<T4o.

2-1)

The unit staff organization shall include the following:

a.

fA' n-/licenseddperator sh,1 be assigod to each pactor cpftiainin f

and an #ditional n -licensed merator 1?vse~r+ C.2-2*-A 4-0 (continued)

Rev 1, 04/07/95

[;

<4.2, 1b 16)

BWR/6 STS 5.0-2

INSERT 6.2.1-A E

E-a SIoCo "thep lan specific titles of those personnel fulfilling the responsibilities of the positions

\\LA-I delineated in these Technical Specifications

Organization

<,CTS>

I Organization Unit Staff (continued) m shalj/be assig ed for eacycontrol rgei from wh' h a reac r

is/6peratingh'n MODES 1Z2, or 3.

rTW/

ni with b~otunits shutodwn or defue d

Twqu*n*tetaota, of thr non-li.Aed operator for the two unit

b. Atheast one i hnsed Reactor perator (RO) 4all be preset the contro room when fue/is in the rea tor.

In

/

addition, wh e the unit n MODE 1, 2 or 3, at lea one 1

Z1e(licensed S por Reactor erator (SRO) aall be prest in T the control room.

2.2 Administrative procedures shallibe developed and imp' to limit the working hours of who perform safety related functions (e.g., licensed SROs, licen:

, auxiliary operators, and key main

ýpersonnel).

(continued)

Rev 1, 04/07/95 Shift crew composition may be less than the minimum requiremenof 10 CFR 50.54(m)(2)(i) and

.2.2.a for a period of time nottexceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to 0

accommoda e unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition'to within the minimum requirements.

,Tshall be on site when fuel is Sin e reactor.

Theposition 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 absencef provided immediate action is taken to fill the re uired Sposition.

Li IL'

II I m I*

A*

m i

4 II BWR/6 STS 5.0-3

Organization

.PS.2'Organization Q

2.2 W<T4.2!2.h Unit Staff (continued)

TSTF-58 1.-

An individual should not be.. permitted tp work more than 16 urs straight, e cludin~g shift tu over time;

2.

individual sho d not be permit to work more than 16-hours in any 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period,-

r more than 24 hou in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months eriod, nor mor han 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period all excluding ft turnover time;

3.

A break at least. 8 hou should be allowe etween workp lods, includin hift: turnover tim e

4.

Exc t during exten dshutdown periods the use of o *rtime should b considered on an i ividual basi ajnd otfor the ent e staff on a-i ft Any. deviation e above gu ines al e au orize in advance by. the. -. lant -

or designee, in accordance with ap :ed administrat v roceuures, or, y ig levels 1 manage

" in

-acco ce*n e

shed cedures d wih ocumentation of the oasis for granting tZh

.evi~atiThn.

(rI

Cont s~t shall-b

luded int

'roceus spd that

i.

. nd'

dual overt* e bsha11 ber eiwed monthl~by the [PlAa I

MSTiF-9g S erintendent or his desiee to ensu at exces hay n

Routine deviation from the guidelinesauthorized.

The amount of overtime wo ed by unit staff mbers performi safety relat functions shall e limited and

contro-ed in acdord e with the NRC P icy Statement worJ g:hours (Genqr c Letter 82-12).

U (r

s a V-h*old an SM license.:.

M. 1>

The Shift Technical:Adviso technical support to the I I) of thermal hydraulics, rea analysis with regard to th addition, the STA shall me the Commission Policy Stat BWR/6 STS Ae, 5.0-4 opeAo r (STA)ashall provideadvisor S ervI$,on hift in the areas ctor engineering, and plant e safe operation of the unit.

In et the qualifications specified by ement on Engineering Expertise on Rev 1, 04/07/95 I&~

lw.

<Tcr

< ( 0,3.

KCZTS>

-redC G,2 -,

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

(

<V cO..

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 K

clc-Controls shall be included in the procedures to require a periodic independent review be LA.Z/> conducted to ensure that excessive hours have not been assigned.

CTS -1 F - 2 57ý1 9)

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. (Deleted)

10. The brackets have been removed and the proper plant specific information has been provided. The requirements regarding individuals who must hold an SRO license are consistent with the current licensing basis, as approved by the NRC in License Amendment No. 160 dated February 19, 1998.

Revision A NMP1 1

Unit Staff Qualifications 2JN) INISTRATIVE CONTROLS 4

.3 Unit Staff Qualifications

Reviewer, Note:

Minimum q ificationsý for memb s of the unit :stafshall be spe ýpfied by use of an vealT1 :qualificatio tatement referencg an ANSI Sta frd acceptable to e NRC staff or by s cifying individua osition o ifications.

Gen ally; the first met is preferable; h ever, the econd method is a ptable to those uni staffs requiring ecial "qualification st ements because of que organization structures.

G.*I*

3.1 Each member of the unit lst Sual ificati ons "o egul at Frecen revi slns or ANSI T

lhe aff not-overed.b*/

A exeed the *lnimum*.qu lfi

[,uides* o *ASI St. ~rds

-S --

r G,

3,2 I

^'SLZS aff shall meet or exceed the minimum o Guide l.s, -Ker-sion Z, 1987, o omore andard accep le to the NRC s ff].

Regulatory G ide 1.8] shall me or cations of *egulations, Reg7latory acceptab to NRC staff].

Rev 1, 04/07/95

<C BWR/6 STS 5.0-5

INSERT 6.3.1-A 0 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 (T

6.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed eactor erator (RO) are those individuals who, in addition to meeting the r requireme of *.3.1, perform the functions described in 10 CFR 50.54(m).

Kcrs>

KL3.i>

<40>c.

IL

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 I

Procedures t.0 ADMINISTRATIVE CONTROLS

-04Procedures I

(G.&.? *L4..1 Written procedurestshall be established, im plemented, and maintained coverok~

following activities:.

-/>

a. The applicableures recommended-in Regulatory Guide -1.33, endix.:

<Tco r--

b.

The emergency operating procedurees required to implement the requirements of NUREG-0737 and ('NUREG-0737, Supplement 1, as stated inGeneric Letter 82-33

<c 4.1>

c C.

for effluent an e

monitoring; A

~

Fire fro ec6 io gama

G-Z.~o*)

]

on.*rgramimplementation;

.ana*-'

(,oc

e.

All programs specified in Specification r'4

,3 3

oJ A,Sl' g/8.7-i 7 2a Rev 1, 04/07195 w

A 3 14 5.-0-6 BWR/6 STS

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 I

Programs and Manuals F A0 ADMINISTRATIVE CONTROLS 1

.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b.

The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.2] and Specification (5.6.3].

Licensee initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1.

sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2.

a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b.

Shall become effective after review and acceptance by the (onsite review function] and the approval of the [Plant Superintendent]; and

c.

Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

(continued)

Rev 1, 04/07/95

<(

, 1" fr.

I BWR/6 STS 5.0-7

Programs and Manuals Programs and Manuals

<(oil>

.5.2 Offsite Dose Calculation Manual (ODCM)

(continued)

Each change shall be identified by markings in the margin ofl the affected pages, clearly indicating the area of the page I

Fthat was changed, and shall indicate the date (i.e., month j

and year) the change was implemented.

J Primary Coolant Sources Outside Containment

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 S_~levels as low as practicable.

The systems include. the *xw eslre orepray, Hig ressure re sfray esidua seat o bval, Red aor Core olation

ading, h

oen rge mbiner mrocess su linde ac n Standbytins The or e program shall include the following:

a.

Preventive maintenanceand periodic visual inspection b.-- Proe esfosaplents; and

~~b

b.

leak test requirements for each system at i~

1 intervals 5.5.3 Post AccidentfSamrlin t

ceosmlga nyi S /

This progmprovides contres that ensure th/capability to

~~obtain j~danal.,ze reactor/coolant, radioa iveCgss n

~parti fates in plant g9 eous effluent a*containment atm phere

/*

sam ~les under acciden *conditions.

The ogram shall inc *fe the

b.

Procedes for sampling a analysis; and/

C.

Pr sosfrmit c fsampling al analysis pment.

//j Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of (continued)

Rev 1, 04/07/95 BWR/6 STS 5.0-8

INSERT 6.5.2-A

<4-14>

Core Spray, Containment Spray, Emergency Cooling, Shutdown Cooling, Reactor Cleanup, Vacuum Relief, Reactor Water Sampling, Containment Atmosphere Dilution (CAD) H20 2

//

02_ \\

Monitor, Drywell Containment Atmosphere Monitoring (CAM), Post Accident Sampling,

ýAo* A/31 Radioactive Gaseous Effluent Monitoring (RAGEMS)ft(e program requirements s all apply to provide for continuous isolation of the associated penetration(s) or a modification eliminates the otential leakage path(s)), Offgas Effluent Stack onitoring (OGESMS), and Post Accident Vent to Reactor Building Emergency Ventilation.

INSERT 6.5.2-B

\\A-4l The provisions of Specification 4.0.1 are applicable to the 24 month frequency for performing system leak test activities.

Programs and Manuals S*.5 5

Programs and Manuals

.5.l O. aR adioactive Effluent Controls Program (continued) the public from radioactive effluents as low as reasonably achievable.

The program shall be contained in the ODCM, shall be implemented by 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 to unrestricted areas, conforming to 10 CFR 20, Appendix B, Table 2, Column 2;

c.

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 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 each unit unrestricted areas, conforming to 10 CFR 50, Appendix to 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 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 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% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g.

h.

'i Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table 2, Column 1; Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each (continued)

Rev 1, 04/07/95 (CTs >

(

BWR/6 STS 5.0-9 I

(continued)

I

Programs and Manuals S5 Programs and Manuals N

.S.

9 Radioactive Effluent Controls Program (continued) unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix 1;

i.

Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each 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 due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

Component/:'clic or Trans'*/nt Limit

/

This ogram provides controls to tr the FSAR, Se ion [

],

cyc c and transie occurrences t ensure that co onents are ntained withi he design li

s.

/

-7 0

JfI Ir Inservice Testina Proaram This program provides'controls for inservice testing of ASME Class 1, 2, and 3 components including applicable supports.

program shall include the following:

Code The (continued)

Rev 1, 04/07/95

?I 5V\\

r 5.5.6 Pre-Stressed oncrete Containm t Tendon Surveillan Proqram This pro am provides cont s for monitoring a tendon degrad ion in pre-stres d concrete containme s, including effe iveness of its c rosion protection me um, to ensure co ainment structur integrity.

The pro am shall include seline measureme prior to initial o rations.

The Ten n

Surveillance Pro am, inspection frequ cies, and accepta e

criteria shall e in accordance with egulatory Guide 1 5,

Revisionn3, 9].

The provj ions of SR 3.0.2 and 3.0.3 are applic le to the Tendon urveillance Program i spection frequenci

<6-1-5 V

f V

5.5

.5, BWR/6 STS 5.0-10

Programs and Manuals

.5 Programs and Mantuals

(

Inservice e

a. Testing*

Boiler a TOII ows:

tinc Program (continued) frequencies specified in Section XI of the ASME nd Pressure Vessel Code and applicable Addenda as ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years

b.

The provisions of SR required Frequencies activities; Required Frequencies for performing inservice testing activities At least once per At least once per 7 days 31 days At least once per 92 days At At At least least least once per once per once per 184 days 276 days 366 days At least once per 731 days 3.0.2 are applicable to the above for performing inservice testing

c.

The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

ntilation Filter T stin Proqram VFT A program shall established to i lement the followin required testing of En Ieered Safety Feat e (ESF) filter vent' ation systems at t frequencies specl ied in [Regulatory ide

],

and in acc dance with [Regul ory Guide 1.52, Rev'iion 2; ASME N510-19

and AG-1].

.a.

emonstrate for ea of the ESF systems hat an inplace te of the high e ency. particulate ai (HEPA) filterss a penetration d system bypass <

.05]% when tested ý (continued)

Rev 1, 04/07/95 5.5.8 11-03 SA Iw.

5.0-11 BWR/6 STS

Programs and Manuals 0.9-5

.5 Programs and Manuals 5.5

/

.8 Venti tion Filter Testin Program (VFTP)

(con nued) accordance with [Regulatory Guide 1.52, evision 2, and AS N510-1989] at the stem flowrate spe fied below [+/- 10%

ESF Yen lation System Flowrate LI

b.

Dem strate for each of e ESF systems that an inpiace test o

he charcoal adsor r shows a penetrati and system pass < [0.05]% whe tested in accordan with [Regulatory Guide 1.52, Revisi 2, and ASME N510-9] at the system flowrate specifli below [+/- 10%]:

ESF Ye ilation System Flowrate

c.

Dem strate for each of e ESF systems that a aboratory t t of a sample of th charcoal adsorber, w n obtained as escribed in [Regula ry Guide 1.52, Revisi 2], shows the mt i

espenet tion less than the v ue specified below when tested n accordance with [A 03803-1989] at a temperature ofS

[30*C] and greater t or equal to the relative humid* y specified below:

ESF entilatiob System Penetration Revi er's Note: Allowabl penetration =[100%

-methyl iodide ef ciency for charcoal edited in staff safe evaluation]/

afety factor).

Safety factor = [5 for systems with hea rs.

=

] for systems withou heaters.

(continued)

Rev 1, 04/07/95 I

F BWR/6 STS 5.0-12

/I

Programs and Manuals S 5* Prograi.s and Manuals Explosive Gas and Storage Tank Radioactivity Monitoring Program This program-provides controls for potentially explosive gas mixtures contained in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks].

The gaseous radioactivity quantities shall be determined following thel methodology in (Branch Technical Position (BTP)

ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure"].

The liquid radwaste quantities shall be determined in accordance with [Standard Review Plan, Section 15.7.3, *Postulated Radioactive Release due to Tank Failures"].

i (continued)

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

Programs and Manuals ograms and Manuals

.5 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen inn the

[Waste Gas Holdup System] and a surveillance program to'u'.

ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b.

A surveillance program to ensure that the quantity of radioactivity contained in [each gas storage tank and fed into the offgas treatment system] is less than the amount that would result in a whole body exposure of > 0.5 rem to any individual in an unrestricted area, in the event of [an uncontrolled release of the tanks' contents]; and

c.

A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the

[Liquid Radwaste Treatment System] is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2; Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

)

SN

" Diesel Fuel O1 Testina Pro raf A diesel uel oil testin rogram to imple nt required testi g/of both n fuel oil and s red fuel oil sh 1 be established.

The progr shall include ampling and tes ng requirements, d

ac ptance criteri all in accordan with applicable TM andards.

The p

/

pose of the pro am is to establi the following:

(continued)

Rev 1, 04/07/95 F5.5.10 5.0-14 f

BWR/6 STS

Programs and Manuals y0-4 15

- T$*>

e.5Programs and Manuals 5.5.10 esel Fuel Oil estin Program (c inued)

a.

Accept ility of new fuel o* for use prior to ddition to star e tanks by determin' g that the fuel & has:

"an API gravity o an absolute speci gravity within

limits,

2.

a flash po' t and kinematic v' cosity within limit for ASTh 2D el oil,

3.

a c r and bright appe ance with proper co r;

b.

Othe properties for AS D fuel oil are wit

limits Wi in 31 days followi sampling and addit n to storage nks; and c

Total particulate oncentration of th fuel oil is < 10

/1 when tested ev 31 days in accord ce with ASIM D-22 Method A-2 a -3.

Technical Specifications (TS)

Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

I &N

a.

Changes to the Basc; of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees nay make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

TS1F-3

1.

change in the TS incor oated in the license; or r2.'r{2.

change to the FSAR or Bases that 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the;FSAR.

d.

Proposed changes that meet the criteriaoof &.5.9b a shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without (continued)

Rev 1, 04/07/95

'D'OC-BWRI6 STS 5.0-15 I

Programs and Manuals 5

Programs and Manuals

.5.*

Technical Specifications (TS) Bases Control Program (continued) prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

C (continued)

Rev 1, 04/07/95

<C-TS>

.5.12 Spety Function Detmination Program (SPOP)

T This program en res loss of safety nction is detec d and appropriate a ions taken.

Upon e ry into LCO 3.0

, an

/evaluation all be made to dete ine if loss of fety function alexists.

ditionally, other a ropriatelimita ons and remedial r comp satory actions may identified to taken as a result of th support system inop ability and corr sponding exception to ent ing supported syst ondition and R uired Actions.

This pr ram implements the equirements of 3.0.6.

The SFDP shal

a.

Provisions f r cross division hecks to ensure a los of the capability o perform the s ety function assumed the accident nalysiG does no go undetected;

b.

Prov ions for ensurin the plant is mainta ed in a safe co ition if a loss function condition xists;

c.

rovisions to ens e that an inoperab supported system's Completion Time s not inappropriat y extended as a result of multiple s port system inoper ilities; and

d. Other appr niate limitations nd remedial or compensaty A loss of fety function exi s when, assuming no con rrent single f lure, a safety fu tion assumed in the acc ent analysis cannot e performed. For e purpose of this pro m, a loss of safe function may exi when a support system s inoperable, "A required sy em redundant to syste s) supported by the inoperable pport system is also operable; or

b.

A requir system redundant to ystem(s) in turn sup rted by the noperable.supported stem is also inoper e; or fA-1 cc-\\

14 BWR/6 STS 5.0-16

Programs and Manuals (crs>

grams and Manuals In~Scrt (ai-A Rev 1, 04/07/95 5.0-17 BWR/6 STS

INSERT 6.5.7-A 6.5.7 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 L,, 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.

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 NMP1I Current TS (CTS).

2. Typographical, grammatical, or editorial 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. Revisions to the Offsite Dose Calculation Manual (ODCM) specification that are consistent with ISTS 5.5.1 have been reviewed and approved by the NRC in License Amendment No.

176 dated September 11, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

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. The post accident sampling system (PASS) has been eliminated, as approved by the NRC in License Amendment No. 174 dated August 26, 2002; therefore, the program requirements of ISTS 5.5.3, "Post Accident Sampling," have not been added.

8. Addition of the Radioactive Effluent Controls Program specification, consistent with ISTS 5.5.4, has been reviewed and approved by the NRC in License Amendment No. 176 dated September 11, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

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.

Revision A NMP1 1

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

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

12. Addition of the Inservice Testing Program specification, consistent with ISTS 5.5.7, has been reviewed and approved by the NRC in License Amendment No. 173 dated AA August 5, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

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. Addition of the Explosive Gas and Storage Tank Radioactivity Monitoring Program specification, consistent with ISTS 5.5.9, has been reviewed and approved by the NRC in License Amendment No. 176 dated September 11, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

15. TS requirements for a diesel fuel oil testing program (ISTS 5.5.10) are not part of the NMIP1 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.

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. (Deleted)

Revision A NMP1 2

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

19. The parenthetical phrase relating to the Post Accident Sampling System and RAGEMS (Radioactive Gaseous Effluent Monitoring System) has been incorporated consistent with License Amendment No. 174 dated August 26, 2002, and NRC letter dated September 11, 2002, "Nine Mile Point Nuclear Station, Unit No. 1 - Use of the Offgas Effluent Stack Monitoring System to Meet Regulatory Guide 1.97, Revision 2, and NUREG-0737 Guidance."

Revision A NNWPI 3

Reporting Requirements 0 ADMINISTRATIVE CONTROLS 6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

flcrirnatinn1 RIi2tinn Fynncii,-gi Dann,'+

-S.ff We 6>

NO A single submittal may be made fo submittal should combine sections station.

1L- - - - - - - - - - - - - - - - -

TE r a multiple unit station.

The common to all units at the The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.

The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period.

The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual J

(continued)

Rev 1, 04/07/95 F

NOTE -----------------------

/A single submittal may be made for a multiple unit station.

The I submittal should combine sections common to all units at the station.

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures

> 100 mrem/yr and their associated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance

[describe maintenance], waste processing, and refueling).

This tabulation supplements the requirements of 10 CFR 20.2206.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD),

or film badge measurements.

Small exposures totalling < 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.

The report shall be submitted by April 30 of each year.

[The initial report shall be submitted by April 30 of the year following initial criticality.]

Annual Radiological Environmental Operatinq Report@Y~

< (6.9 1.

.6.2

.ci>

BWR/6 STS 5.0-18

Requirements

.6 Reporting Requirements'

.6.2 Annual Radiolog Id>

.6.3 K (,,eq.

ical Environmental Operating Report (continued)

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted 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 ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section UV.B.I.

Monthly tOnPatina Rpnnrtsz

<.q.-. 1>

Routine reports of operating statistics and shutdown experience

I Z Z1 I; 1

-4 (continued)

Rev 1, 04/07/95 (ODCH),

and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements (in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979].

[The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.]

In the event that some individual 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 in a supplementary L report as soon as possible.

-fUl A single submittal may be made for a multiple unit station.

The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

Reporting

<C-TS>

-12!;8 BWR/6 STS 5.0-19

Reporting Requirements 6.6 Reporting Requirements'

.L6.4 Monthly Operating Reports (continued)

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

.6.5 CORE OPERATING LIMITS REPORT (COLR)

4ALLf>

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

iV' 3

jb.

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

n~e_

IIentý** thee TopiA Ruemportt(s)/ by number, jrtle, d~ate, dw*

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS)]limits, nuclear limits such~as,0Dl',

transient (3

analysis limits, and accident analysis limits)rf the safety o-Y analysis are met.

$kuo+4ogy R

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the s

NRC.

r5.6.6

-- /Reactor cOOlait System (RCS)

WESSURE AND TEMPERA;KRE LIMITS

/

~~REPORT (PTIL) 40_

Sa.

R pressure and t perature limits f

~heatup, co/oldown/

SvX'~ow temperature o ration, criticalt'y, and hydrostati/

S*~testing as well ds heatup and coo *own rates shall bi/

Sestablished a/d documented in t PTLR for the fol wing:

(continued)

Rev 1, 04/07/95 K

jiŽ

BWR/6 STS 5.0-20

C-Ts,>

INSERT 6.6.5-A

q..;> 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.

INSERT 6.6.5-B t

1. NEDE-240 11-P-A, "General Electric Standard Application For Reactor Fuel," U.S.

Supplement, (NRC approved version specified in the COLR).

Oc~c-A 2>

Reporting Requirements porting Requirement6 5.6.6 React r Coolant System MS) PRESSURE AND TEHPOATURE LIMITSq (The individ specifications that ddress RCS pressure d

S/

temperature imits must be referen d here.]

b.

The ana ical methods used to etermine the RCS pre ure and t erature limits shall those previously r iewed and proved by the NRC, sp ifically those desc bed in the fol owing documents: [Ide ify the NRC staff a roval d ument by date.]

C. The PTIR shall be pr ided to the NRC up issuance for each reactor vessel flu ce period and for yrevision or S

supplement theret Reviewer's Notes:

he methodology for he calculation of t P-T limits for NRC a roval should inclu the following provi ons:

1. The me ology shall descri ehow the neutron fl nce is calcul ed (reference new gulatory Guide when ssued).

2.

Th Reactor Vessel Mat ial Surveillance Pro am shall mply with Appendix to 10 CFR 50. The actor vessel material irradiatio surveillance specim removal schedule shall be provided along with how the ec~imen, examinations shall be used t update the PTLR cur s.

3.

Low Temperat e Overpressure Prot tion (LTOP)

System ]if setting 1its for the Power Op ated Relief Valves (P0. s),

develope sing NRC-approved thodologes may b2 inc ded in the *LR...

/-"i

4.

The dusted reference t peraturb (ART) for ea reactor b

line material shah be calculated, accoun ng for diation embrittlemt inaccordance with eguhatory Guide 1.9g, Revision 2.//

The limiting AR shall be incorporate into the calculatio of the pressu and temperature lim curves in accordan with NUREG-0 Standard Review P n 5.3.2, Pressure Temperatur Limits.

(continued)

Rev 1, 04/07/95 BWR/6 STS 5.0-21

Reporting Requirements Ti6oReporting Requirements 5.6.6 eactor Coolant System (RCS)

PRES E AND TEMPERATURE IIMITS

6.

T minimum temperatur-requirements of Ap ndix G to 10 CFR

art 50 shall be incorporated into the p ssure and

/temperature limit prves.

Licensees who ve removed two or re capsules should compare for ch surveillance ma rial the measured inc ase in referen temperature (RT, o the predicted incr se in RT T; re the predicted i rease in RT is base on the mean s Ift in RTV*,

plus t two standardmJ~eviati value (2 j pecified in Regula ry Guide 1.99, Revisi

2. If the me red value exceeds e predicted value (i rease in RT

+ o the licensee ould provide a supple nt to the PTIN/

o demonstrate how e results affect the pproved methodology.

When apecial Report s required by ondition B or G LCO

.3. [3.1],

Pos Accident Moni ring (PAM)

Instr rep6rt shall be s itted within e following 14 y:

port shall out ne the prepl ned alternate me od monitoring, th cause of the operability, an /the p schedule for estoring the 4i/strumentation chnnels o0 Function t OPERABLE stat Rev 1, 04/07/95 BWR/6 STS 5.0-22

<eCf S>

4.6 eporting Requirements Reporting Requirements

~4y.6 5.6.9 Tendon Surveillnce Report (co tinued)

-Containment endon Surveill ce Program shal be.reported to t

-]

NRC withi 30 days.

The port shall inclu e a description o the tendon ndition, the c dition of the co rete (especially t

nchorages), t inspection proc ures, the tolera es on c.crac ng, and the co ective action t en.

I viewer's Note:

hese reports ma be required covying inspection, tes, and maintenanc activities. T" e reports determined on n individualtba.'s for each unit d their preparation nd submittal are esignated in t Technical

_Specificat ons.

A Rev 1, 04/07/95 BWR/6 STS 5.0-23 I't

INSERT 6.6.6-A a 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. (Deleted)

c. (Deleted)

d. (Deleted)

e. (Deleted)

f. (Deleted)

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

h. Accident Monitoring Instrumentation Report, Specification 3.6.1ta (Table 3.6.11-2 Action 3 or 4) (Within 14 days following the event).

&N

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. (Deleted)

3. Revisions to the Occupational Radiation Exposure Report specification that are consistent with ISTS 5.6.1 have been reviewed and approved by the NRC in License Amendment No.

176 dated September 11, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

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

5. (Deleted)

6. Revisions to the Annual Radiological Environmental Operating Report specification that are consistent with ISTS 5.6.2 have been reviewed and approved by the NRC in License Amendment No. 176 dated September 11, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

7. (Deleted)

8. Revisions to the Radioactive Effluent Release Report specification that are consistent with ISTS 5.6.3 have been reviewed and approved by the NRC in License Amendment No. 176 dated September 11, 2002. Any deviations from the ISTS are addressed as part of that license amendment.

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.

Revision A NMP 1 1

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

11. The requirements for submitting a Special Report to the NRC in the event that accident monitoring instrumentation is inoperable, including both the specified time period and the report contents, currently reside in TS 3.6.11 (Table 3.6.11-2, Action 3 or 4). 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 special report requirements are retained in CTS 3.6.11 rather than relocating them to or duplicating them in the Administrative Controls portion of the TS. As an alternative to ISTS 5.6.8, an "Accident Monitoring Instrumentation Report" item is added to the Special Reports specification (CTS 6.9.3), with the proper reference made to Specification 3.6.11.a, and the applicable time period for the report specified (consistent with Specification 3.6.11.a). The term "Accident Monitoring Report" replaces the ISTS term "PAM" to be consistent with the title of CTS 3.6.11.

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 (including the changes in License Amendment No. 173, issued on August 5, 2002, and License Amendment No. 176, issued on September 11, 2002).

14. The term "shutdown margin" is inserted since the "SDM" acronym is not defined elsewhere in TS Section 6.0.

Revision A 2

NMP1

<CTS>I Rev 1, 04/07/95 "p.0 ADMINISTRATIVE CONTROLS 0.7 High Radiation Areab 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

> 100 mrem/hr but < 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 (RWP).

Individuals qualified in radiation protection procedures (e.g., (Health Physics Technicians]) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates

1000 mrem/hr, provided they are otherwise 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 or accompanied by one or more of the following:

a.

A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b.

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the

[Radiation Protection Manager] in the RWP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation levels Ž 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision.

Doors shall remain locked except during periods of access by personnel (continued)

BWR/6 STS 5.0-24

igh RadiatonAreat;&

Radiation Areal I

Rev 1, 04/07/95 5.7.2 (continued) under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas.

In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.7.3 For individual high radiation areas with radiation levels of

> 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed a-ound the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

BWR/6 STS 5.0-25

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. The plant specific information has been reviewed and approved by the NRC in License Amendment No. 176 dated September 11, 2002. The wording from NUREG-1433, Revision 2, has been adopted for the "High Radiation Area" specification.

Revision A NMP1 1

ATTACHMENT 6 Tables Summarizing the Proposed Chanzes The following four tables summarize the proposed changes to NMP1 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 Note: Changes to the four summary tables originally provided by letter dated June 7, 2002 are identified by a vertical bar and an "A" in the right margin.

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).

I I

_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 4.2.7.b, 4.3.3.a, 6.9.1.e, 6.11, 6.12, 6.14, 6.16, 6.17,6.18,6.19 A.2 Incorporates wording changes consistent with the changes to 10 CFR 50.59 published in 6.5.6 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 A.5 Incorporates a note indicating that the specification requirements apply to the 6.5.2 6.14 Radioactive Gaseous Effluent Monitoring System (RAGEMS) as long as it is a potential leakage path, consistent with an NRC-approved change.

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

6.6 1.31, 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 A.4 Adds an item for Accident Monitoring Instrumentation Reports.

6.6.6 3.6.11.a 6.7, High Radiation Area A. 1 Editorial changes, reformatting, and revised numbering.

6.7 6.12 Current Specification 6.4, Training None None I

None I

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 I

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

SUMMARY

REVISED TS CTS SECTION SECTION Current Specification 6.6, Reportable Occurrence Action A. 1 Removes Reportable Event notification requirements from the Technical Specifications, N/A 6.6.1.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, 6.7.1.d Current Specification 6.10, Record Retention None None I

None None Current Specification 6.13, Fire Protection Inspection None None None None Current Specification 6.15, Iodine Monitoring None None aNone 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.

6.5.6 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 Page 2 of 2 DOC #

SUMMARY

REVISED TS CTS SECTION SECTION Current Specification 6.5, Review and Audit None None None None Current Specification 6.6, Reportable Occurrence Action None INone I

None None Current Specification 6.7, Safety Limit Violation None I None None None Current Specification 6.10, Record Retention None I 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 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 S.....None INone None I

None I

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

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 T

None Current Specification 6.6, Reportable Occurrence Action None I None None None None None::

=Nn Current Specification 6.7, Safety Limit Violation None INone I

None I

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 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, Organization 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 INone I

None I

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 None None None None None None 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 I ZŽ

TABLE R - RELOCATED SPECIFICATIONS AND REMOVAL OF DETAILS MATRIX REVISED TS SECTION AND DOC#

CTS SECTION CHANGE CONTROL None - LA. 1 6.6.1.b Current Specification 6.(

The requirements of CTS 6.6.1.b; Reportable Events reviews by SORC and submittal of the results of the reviews to the SRAB and the Vice President - Nuclear Generation.

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

Si I

I 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 SAppendix 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 2

Current Snecification 6.(

SUMMARY

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

ATTACHMENT 7 List of Regulatorv Commitments The following table identifies those actions committed to by Nine Mile Point Nuclear Station, LLC (NMPNS) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE NONE Page 1 of 1

v t e Letter Sequence Request
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]

ML023390027

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