ML20094B225

Proposed Tech Specs Incorporating New Reporting Requirements as Defined by Commission in Generic Ltr 83-43

ML20094B225
Person / Time
Site:	Brunswick
Issue date:	10/29/1984
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20094B203	List: ML20094B198 ML20094B225
References
GL-83-43, NUDOCS 8411070068
Download: ML20094B225 (62)

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BRUNSWICK STEAM ELECTRIC' PLANT PROPOSED TECHNICAL SPECIFICATION PAGES - UNIT 1

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SUMMARY

LIST & REVISIONS-

~ BRUNSWICK UNIT 1 1

l P_ age.

Comments

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" Reportable &currence" changed to " Reportable Event" XV ~

Incorporates change to reporting requirements

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XVI

~ Revised to reflect repagination 1-6^

. Reportable Event change incorporated

^

~3/4 3-62.

Specification 6.9.1.14.b changed to 6.6.1

-3/4 3 Specification 6.9.1.14.b changed to 6.6.1 3/4 4-7 Reference.to Specification 6.9.1.12 deleted 3/4 4 Mathematical Symbols have been put into words Revised to reflect Special Report changes

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- Typographical errors correcred ("of the" added )

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'3/4 6-6 Revised to reflect reporting of abnormal primary.

f! n containment degradation pursuant to Specification 6.9.2 -

3/4 11-22 Specification 6.9.1.'14.b changed to 6.6.1 s

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16-3' Revised to reflect organizational changes

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6-4 Revised to reflect organizational changes v...

rf 6-5 Revised -to reflect organizational changes 0:

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" Manager - Operations" chnged to " Director - Training" N,'

't 11

" Manager - Plant Operations" eliminated from Section 6.5.3.3, PNSC membership

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6i12 Reportable Events change incorporated 6-13

" Reportable &currence" changed to " Reportable Event"

' 6-14 --

Comma addal in Seetion 6.5.4.6 6-15 Revisai to reflect title changes

- "Bi-monthly" changed to "once every two months"

^ 6-17

" Principal QA Specialist

. Performance Evaluation Unit" changed to " Manager - Quality Assurance Service Section"

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6-18 Reportable Events change incorporated

^i 6-20 Reportable Events change incorporated 6-25 Sections 6. 9.1.12, 6.9.1.13, and 6. 9.1.14 d eleted Item 6.9.2.c added Item 6.9. 2.q aided Section 6.9.2 re-ordered Repaginated 2

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Comments 6-26

'." Reportable Occurrences" changed to " Reportable Events" Repaginated 6-27 Reference to snubber Table 3.7.5-1 revised. to reflect table

- deletion per oer ' submittal datal May 7, 1984

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. Repaginated I

6-2,8

" Dose" addal to "Of fsite Calculation Manual" Typographical error corrected ("toally" changed to

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(BSEP-1-12) i.

e-m INDE_X_

DEFINITIONS SECTION 1.0- DEFINITIONS (Continued)

PAGE PROCESS CONTROL PROGRAM (PCP).........................

1-5 PURGE - PURGING.......................................

1-6 RATED THERMAL POWER...................................

1-6 REACTOR PROTECTION SYSTEM RESPONSE TIME...............

1-6 REFERENCE LEVEL ZT.R0..................................

1-6 l

REPORTABLE EVENT......................................

1-6 ROD DENSITY...........................................

1-6 SECONDARY CONTAINMENT INTEGRITY.......................

1-6 g SHUTDOWN MARGIN.......................................

1-7 SITE BOUNDARY.........................................

1-7 SOLIDIFICATION........................................

1-7

' l-7 SOURCE CHECK.........................'.................

~

M SPIRAL' RELOAD

...................(.....................

1-7 SPIRAL UNLOAD..................'.......................

1-7 STAGGERED TEST BASIS..................................

1-7 1

THERMAL POWER.........................................

1-8 TOTAL PEAKING FACTOR..................................

1-8 UNIDENTIFIED LEAKAGE..............................

4..

1-8 UNRESTRICTED AREA.....................................

1-8 VENTILATION EXHAUST TREATMENT SYSTEM..................

1-8 V5NTING...............................................

1-8 FREQUENCY NOTATION, TABLE 1.1.........................

1-9 OPERATIONAL CONDITIONS, TABLE 1.2.....................

1-10 BRUNSWICK - UNIT 1 II Amendment No.

m

~

(BSEP-1-12)

INDEX ADMINISTRATIVE CONTROLS'

~'

~~~~

SECTION PAGE J

6.5.4 CORPORATE NUCLEAR SAFETY SECTION Function...............................................

6-13 Organization...........................................

6-13 Review.................................................

6-14 3

Records................................................

6-15 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Function...............................................

6-16 Audits.................................................

6-16 Records................................................

6-17 Authority..............................................

6-17 Personnel..............'................................

6-17 6.5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM............

,6-18 3,

J-6.6 REPORTABLE EVENT ACTION....................................

6-18

)

6.7 SAFETY LIMIT VIOLATION.....................................

6-18

' a 6.8 PROCEDURES AND PR0 GRAMS....................................

6-19 6.9 REPORTING REQUIREMENTS

,n Routine Reports...............................'........

6-20 Startup Reports........................................

6-20 Annual Reports.........................................

6-21 Personnel Exposure and Monitoring Report...............

6-21 Annual Radiological Environmental Operating Report.....

6-22 Semiannual Radioactive Effluent Release Report.........

6-23 Monthly Operating Reports..............................

6-24 Special Reports........................................

6-25 BRUNSUICK - UNIT 1 XV Amendment No.

a-

~

~

(BSEP-1-12)

INDEX p.

s ADMINISTRATIVE CONTROLS SECT 70N' PAGE

'6.10 RECORD = RETENTION..........................................

6-25 6.11 RADIATION PROTECTION PR0 GRAM..............................

5-27 6.12 HIGH RADIATION AREA.........

6-27 6.13 0FFSITE DOSE CALCULATION MANUAL (0DCM)....................

6-28 6.14-PROCESS CONTROL PROGRAM (PCP).............................-

6-28 6.15 MAJOR CHANGES TO LIQUID, GASECLS, AND SOLID WASTE TREATMENT SYSTEMS..........................

6-29 c3 l

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BRUNSWICK UNIT 1 XVI Amendment No.

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- i (BSEP-1-12) 1 DEFINITIONS PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a

. confinement to' maintain. temperature, pressure, humidity, concentration or other operating condition, in'such a manner that' replacement air or gas is required to purify the containment.

. RATED THERHAL POWER

--RATED. THERMAL' POWER shall be total reactor core heat transfer rate to the reactor. coolant of 2436 MWt.

~

REACTOR PROTECTION SYSTEM RESPONSE TIME

-REACTOR PROTECTION SYSTEM RESPONSE TIME.shall be the time interval from when the monitored parameter exceeds its tcip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO

.., J{ht" The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the

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vessel zero' point. This REFERENCE LEVEL ZERO is approximately mid-point or.

• y the top fuel' guide and is the single reference for all specifications of

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-vessel water level.

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REPORTABLE EVENT

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. A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 ;

to 10 CFR Part 50.

' ROD DENSITY.

e ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of notches.- All rods fully inserted-is equivalent to 100% ROD DENSITY.

~

SECONDARY CONTAINMENT INTEGRITY SECONDARY CONTAINMENT INTEGRITY shall exist when:

a.

All automatic reactor building ventilation system isolation valves or dampers are OPERABLE or secured in the isolated-position.

b.

The standby gas treatment system is OPERABLE pursuant to

. Specification 3.6.6.1.

c.

At least one door in each access to the reactor building is closed.

~

-d.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERAELE.

BRUNSWICK - UNIT 1 1-6 Amendment No.

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(BSEP-1-12)

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION

-3.3.5.8, The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.5.8-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: As shown in Table 3.3.5.8-1.

ACTION:

With a radioactive liquid effluent monitoring instrumentation channel a.

alarm / trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, declare the channel-inoperable,'or change _the setpoint so it is acceptably c y' t.

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conservative.

b.

With less than one radioactive liquid effluent monitoring instrumen-tation channel in each release pathway OPERABLE, take the ACTION shown in Table 3.3.5.8-1.

Return the instruments to OPERABLE status 7;

.V within 30 days or, if unsuccessful, explain in the next Semiannual.

Radioactive Ef fluent Release Report why the inoperability was not corrected in a timely manner.

• 1he provisions of Specifications 3.0.3, 3.0.4, and 6.6.1 are not l

c.

applicable.

SURVEILLANCE REQUIREMENTS 4.3.5.8 Each radioactive liquid effluent nonitoring instrumentation channel shall be demonstrated OPERABLE by performance of ' the CHANNEL CHECK,, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the

~ frequencies shown in Table 4.3.5.8-1.

p NOTE:

See Base

• 3/4.3.5.8.

l t

i BRUNSWICK - UNIT 1 3/4 3-62 Amendment No.

s 4

(BSEP-1-12)

INSTRQ4ENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.5.9-1 shall be OPERABLE with'their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The

'setpoints shall-be determined in accordance with the methodology as described in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY:

As shown in Table 3.3.5.9-1.

ACTION:

~

a.

With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably iff conservative.

b.

With less than one' radioactive gaseous effluent monitoring instrumentation. channel OPERABLE, take the ACTION shown-in Table 3-3.3.5.9-1.

Return the instruments to OPERABLE status within 30 days or, if unsuccessful, explain-in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a u

timely manner.

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The provisions of Specifications 3.0.3, 3.0.4, and 6.6.1 are not l

c.

applicable.

SURVEILLANCE REOUIREMENTS

'4.3.5.9.Each radioactive gaseous effluent monitoring instrumen'tation channel

.shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3.5.9-1.

NOTE: See Bases 3/4.3.5.9.

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- BRUNSWICK - UNIT 1.

3/4 3-68 Amendment No.

g.;

t (BSEP-1-12)

' REACTOR COOLANT SYSTEM 3/4.4.4 CHEMISTRY LIMITING CONDITION FOR OPEi% TION 3.4.4 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.4-1.

APPLICABILITY: At all times.

ACTION:

a".

In OPERATIONAL CONDITION 1. 2, and 3:

1..

With the conductivity or chloride concentration exceeding the olimits specified in Table 3.4.4-1, but less than 10 pmho/cm at 25*C and less than 0.5 ppm, respectively, operation may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and this condition need not be reported to the Commission provided, that operation under these conditionsi l

shall not exceed 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> per year. The provisions of

'RF

_ Specification.3.0.4 are not applicable.

2.

With the conductivity or chloride concentration exceeding the linics specified in Table 3.4.4-1 for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during (a

-one continuous time interval or.with the conductivity exceeding

?"f 10 paho/cm at 25*C or chloride exceeding 0.5 ppm, be in at least jj HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

At all other times.with the conductivity and/or chloride concentration of the reactor coolant in excess of the limit specified in Table 3.4.4-1, restore the conductivity and/or chloride

- ~ -

'-concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

L s

BRUNSWICK - UNIT 1 3/4 4-7 Amendment No.

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(BSEP-1-12)

C s '

E-REACTOR COOLANT SYSTEM S

3/4.4.5 -SPECIFIC ACTIVITY LIMITING CONDITION'FOR OPERATION

~~-

- 3.4.5 The specific activity of'the reactor coolant shall be limited to:

less than or equal to 0.2 pCi/ gram DOSE EQUIVALENT I-131, and l

a.-

.q,3 less than or equal to 100/E' pCi/ gram.

l b.

1 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

5

' ACTION:

a.

In OPERATIONAL CONDITION 1, 2, and 3, with the specific activity of the reactor coolant; 1.

Greater than 0.2 pCi/ gram DOSE EQUIVALENT I-131 but less than or l

x equal to 4.0 pCi/ gram, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that. operation under these conditions shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

2.

Greater than 0.2 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 pCi/ gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

s; 3.

Creater than 100/E' pCi/ gram, be in at least HOT SHUTDOWN with l

the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

. %,7 L

b..

' In OPERATIONAL CONDITION 1, 2, 3, or 4, 1.

With the specific activity of the primary coolant greater than 4

0.2 Jfi/ gram DOSE EQUIVALENT I-131 or greater than 100/E pCi/ gram, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1. at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> until the specific activity of the primary coolant is restored to within its limits. In-lieu of a License Event Report, prepare and submit.to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the results of the specific activity analyses and the time duration when'the l

specific activity of the coolant exceeded 0.2 pCi/ gram DOSE EQUIVALENT I-131 together with the below additional information.

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- BRUNSWICK - UNIT t 3/4 4-10 Amendment No.

- _m (BSEP-1-12) 7 w

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'CONTAIMENT SYSTEMS ~

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PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION

> 3.6.ls4. The structura2 integrity of the primary containment shall be,,,

f maintained at a level coesistent with the acceptance / criteria in Specification

4. 6.1. 4..

2

' APPLICABILITY: OPERATIONAL' CONDITIONS 1, 2, and 3.

A_CTION:

With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212*F.

7 SURVEILLANCE REQUIREMENTS

^

l 4.6.1.4.1 The structural integrity of the exposed accessible interior and-exterior surfaces of the primary containment, including the liner plate shall i

be determined during the shutdown for each Type A containment leakage rate

- test by a visual inspection of-those surfaces. This inspection shall be

_4 performed prior to the Type A containment leakage rate test to verify no e

apparent changes in appearance or other abnormal degradation.

4. 6.' 1. 4. 2 Reports Any abnormal degradation of the primary containment

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structure detect.ed during the above required inspections shall be reported to the Commission pursuant to Specification 6.9.2.

Thi Special Report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

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BRUNSWICK - UNIT 1 3/4 6-6 Amendment No.

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s RADI0 ACTIVE EFFLUENTS 3/4.11.3-SOLID RADIOACTIVE WASTE-LIMITING CONDITION FOR OPERATION

-3.11.3 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive "astes to meet shipping and burial ground' requirements.

APPLICA3ILITY: At all times.

ACTION:

a.

With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

b.

The provisions of Specifications 3.0.3, 3.0.4, and 6.6.1 are not l

applicable.

SURVEILLANCC REQUIREMENTS g.

'4.11.3 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at~1 east every tenth batch g;

of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, and sodium sulfate solutions).

a.

If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-TION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alte rnative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL

- g""; PROGRAM, and a subsequent test verifies SOLIDIFICATION.

SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

b.

If the initial test specimen from a batch of waste fails to verify-SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection of testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial' test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.14, to assure SOLIDIFICATION of subsequent batches of waste.

NOTE:

See Bases 3/4.11.3 BRUNSWICK - UNIT 1 3/4 11-22 Amendment No.

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i VICE PRESIDENT L

SRUNSWICK l-NUCLEAR' PROJECT f

I i

GENERAL MANAGER I~

BRUNSWICK PLANT l

l MANAGER OPERATIONS

,z.

I OPERATIONS SUPERINTENDENT.

i I

PRINCIPAL ENGINEER SHWT OPERATING PRINCIPAL ENGINEER PRINCIPAL ENGINEER OPERATIONS (UNIT 1)

SUPERVISOR (SRO) OPERATIONS (UNIT 2)

OPERATIONS I

I SHIFT FIRE BRIGADE 4

44 SENIOR SPECIALIST-(EMERGENCY FIRE PROTECTION CO O RDIN ATIO N) l I

I SHIFT FIRE BRIGADE l

SUPPORT GROUP i

LEGEND i

4 Number of Brigade Fire Chief s varies with Shif t Organization 44 One Specialist is Assigned The Duties of The Plant Fire Chief Amendmerit No.

BRUNSWICK STEAM A.

. ELE CTRIC - PL A NT PLAN T FIRE PROTECTION ORGANIZATION Carolina

.., Power & Light Company

^

Figure 6.2.2-2 Page No. 6-5

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( BSEP-1 -12)

ADMINISTRATIVE CONTROLS 6.2.3 ONSITE NUCLEAR SAFETY (ONS)

FUNCTION 6.2.3.1 The ONS Unit shall function to examine facility operating characterisitics, NRC issues, industry advisories, and other sources which may indicate areas for improving facility safety.

RESPONSIBILITIES 6.2.3.2 The ONS Unit shall be responsible for maintaining surveillance of facility activities to provide independent verification

• that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.3 The ONS Unit shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving facility safety to the Manager-Corporate Nuclear Safety Section.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shif t Technical Advisor shall serve in an advisory capacity to the Shift Operating Supervisor on matters pertaining to the engineering aspects assuring safe operation of the unit.

6.3 FACILITY STAFF QUALIFICATION 6.3.1 Each member of the f acility staff defined in Figure 6.2.2-1 shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Manager - Environmental & Radiation Control who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shif t Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or. engineering discipline with specific training in plant design, and response and analysis of the plant during transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Director - Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and AppendL< "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Director - Training and shall meet or exceed the requirements l

of Section 27 of the NFPA Code-1975.

• No t responsible for sign-of f function.

BRUNSWICK - UNIT 1 6 -8 ppendment No.

(BSEP-1-12)

ADMINISTRATIVE CONTROLS 6.5.3 PLANT NUCLEAR SAFETY COMMITTEE (PNSC)

FUNCTION 6.5.3.1 As an effective means for the regular review, overview, evaluation, and maintenance of plant operational safety, a Plant Nuclear Safety Committee (PNSC) shall be established.

6.5.3.2 The PNSC shall function through the utilization of subcoamittees, audits, investigations, reports, and/or performance of reviews as a group.

COMPOSITION 6.5.3.3 The PNSC shall be composed of the:

Chairman:

General Manager - Brunswick Plant

• Member:

Manager - Technical & Administrative Support Member:

Manager - Technical Support Member:

Manager - Operations Member:

Manager - Maintenance Member:

Manager - Environmental & Radiation Control Member:

Assistant to Plant General Manager Member:

Director - QA/QC Member:

Director - Regulatory Compliance Member:

Director - Administrative Support ALTERNATES 6.5.3.4 All alternate members shall be appointed in writing by the PNSC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as members at any one time.

6.5.3.5 All alternates, shall as a minimum, meet equivalent qualification

. criteria as specified for professional-technical personnel in Section 4.4 of ANSI N18.1-1971.

MEETING FREQUENCY 6.5.3.6 The PNSC shall meet at least once per calendar month and as convened by the PNSC Chairman or his designated alternate.

QUORUM 6.5.3.7 The minimum quo' rum of the PNSC necessary for the performance of the PNSC activities of the Technical Specifications shall consist of the PNSC Chairman or his designated alternate and five members including alternates.

No more than two alternates shall be counted toward meeting the minimum quorum requirement.

l

• Or designated alternate.

BRUNSUICK - UNIT 1 6-11 Amendment No.

W

.4,,.

p-(BSEP-1-12)

' ADMINISTRATIVE CONTROLS-

' ACTIVITIES 6.5.3.8' The PNSC activities shall' include the following:

a.'

Review of_.all procedures required by Specification 6.8 and changes E the'reto (and any other procedures and changes thereto), any. of which

~

constitute an unreviewed safety question or involve a change to the Technical Specifications,; prior to implementation.

b.

' Review of all proposed tests or experiments that constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

c. -

Review of all proposed modifications that constitute an unreviewed safety question.or involve a change to the Technical Specifications, prior to implementation.

d.

Review of all proposed changes' to the Technical Specifications or

' Operating License, prior to implementation.

Review of reports on violations of Technical Specifications including e.

reports covering evaluation and recommendations to prevent recurrencc

- to the Vice President - Brunswick Nuclear Project and to the Manager

- Corporate Nuclear Safety Section.

f.-

Performance of special reviews, investigations (or analyses), and reports thereon as requested by the tianager - Corporate Nuclear Safety Section, g.

Review of all REPORTABLE EVENTS.

h.

Review of facility operations to detect potential nuclear safety hazards.

1.

Annual review of.the. Security Plan.

j.

Annual review of the Emergency Plan.

k.

Review of accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vica President - Brunswick Nuclear Project and the !!anager - Corporate Nuclear Safety Section.

1.

Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE 2

CALCULATION MANUAL.

BRUNSWICK - UNIT 1 6-12 Amendment No.

g m

3.

3

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(BSEP-1-12) 1 ADMINISTRATIVE' CONTROLS AUTHORITY 6.5.3.9 : If there is a disagreement between recommendations of a majority of the PNSC and thecactions' contemplated by the General Manager - Brunswick Plant, the

'PNSC shall provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President -

' Brunswick Nuclear Project and the Vice President _ Corporate Nuclear Safety and '

-Research. The course determined by the General Manager - Brunswick Plant to be the most conservative shall be followed.

RECORDS 6.5.3.10. The PNSC shall maintain written minutes of each PNSC meeting that, at a minimum,' document.the results of all PNSC activities performed under the provisions of these Technical Specifications. Copies shall be provided to the Vice President - Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Caction.

6.5.4 CORPORATE NUCLEAR SAFETY SECTION FUNCTION 6.5.4.1 The Corporate Nuclear Safety Section (CNSS) of the Corporate Nuclear Safety & Research Department shall function to provide independent review of significant plant changes, tests, and procedures; verify that REPORTABLE EVENTS are investigated in a timely manner and corrected in a manner that reduces the 1

probability of recurrence of such events; and detect trends that may not be

~

apparent to a day-to-day observer.

ORGANIZATION 6.5.4.2 The individuals assigned responsibility for independent reviews shall be specified in technical disciplines. These individuals shall collectively have the experience and competence required to review activities in the following areas:

a.

nuclear power plant operations b.

nuclear engineering

~

- c.

' chemistry and radiochemistry d.

v.etallurgy e.

.non-destructive testind f.

instrumentation and control g.

radiological safety h.

mechanical and electrical engineering i.

administrative controls BRUNSWICK - UNIT 1-6-13 Amendment No.

g; -

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+

I (BSEP-1-12)

L s

' ADMINISTRATIVE CONTROLS s

0'GANIZATION'(Continued)

R

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seismic.and environmental

.k.

quality assurance practices 1.

Other. appropriate fields associated with the unique characteristics of the nuclear power plant..

e 6.5.4.3. The Manager - Corporate Nuclear Safety.Section shall have an academic

~

degree in an engineering or related field and, in addition, shall have a minimum of ten years related experience, of which a minimum of five years

~

shall be in :the operation and/or -design of nuclear power plants.

6.5.4.4 The independent safety review pcogram reviewers shall have an academic degree in an engineering or related field or equivalent and,- in addition, shall have a minimus' of :five years related experience.

6.5.4.5 An individual may possess competence in more than one specialty area. If.suf ficient expertise.is not available within the Corporate Nuclear-

~

Safety Sectica,Leompetent individuals from other Carolina Power & Light

, Company organizations or outside consultants shall be utilized in performing

+

-independent reviews. and investigations.

6.5.4.61 At -least three individuals, qualified - as ' discussed in 6.5.4.4 above,

[

?ci

. shall review - each item submitted under the requirements of Section 6.5.4.9.

7 f^:

6.5.4.7 sIndependent safety reviews shall be performed by individuals not

~

- directly involved with the activity under. review or responsible for the l

activity under review.

6.5.4.8 The Corporate Nuclear Safety Section independent. safety review

. program shall'.be conducted in accordance with written, approved procedures.

REVIEW-6.5.4.9 The Corporate Nuclear Safety Section shall perform reviews of the following:

4 The safety evaluations for'1) changes to procedures' required by D a.

Specification 6.8, 2) modifications of equipment or systems, and

' 3) tests or experiments that constitute a change to the safety analysis report to verify that such actions did not constitute an

~ unreviewed safety question or involve a change to 'the Technical' Specifications.

Implementation may proceed prior to completion of-

~^

this' review.

b.

fProposed changes to procedures required by Specification 6.8, and proposed modifications that constitute an unreviewed safety question

.as defined in 10 CFR 50.59 or a change to the Technical

' Specifications, prior to implementation.

BRUNSWICK. - UNIT 1 6-14 Amendment No.

m

~w~m

,w _

(BSEP-1 -12)

ADMINISTRATIVE CONTROLS REVIEW (Continued) c.

Proposed tests or experiments that involve an unreviewed safety question as defined in 10 CFR 50.59 or a change to the Technical Specifications, prior to implementation.

Proposed changes to the Technical Specifications and Operating d.

License.

e.

Violations, deviations, and events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Commission, such as:

1.

Violations of applicable codes, regulations, orders, Technical Specifications, license requirements, and internal procedures or instructions having nuclear safety significance.

2.

Significant operating abnormalities or deviations from normal and expected performance of plant safety-related structures, systems, or components.

f.

Reports and minutes of the PNSC.

g.

Any other matter involving safe operation of the nuclear power plant that the Manager - Corporate Nuclear Safety Section deems appropriate for consideration or which is referred to the Manager - Corporate Nuclear Safety Section by the on-site operating organization or other functional organizational units within Carolina Power & Light Company.

6.5.4.10 Review of items considered under 6.5.4.9(e) through (g) above shall include the results of any investigations made and the recommendations resulting from these investigations to prevent or reduce the probability of recurrence of the event.

RECORDS 6.5.4.11 Records of Corporate Nuclear Safety Section reviews, including recommencstions and concerns, shall be prepared and distributed as indicated below:

a.

Copies or documented reviews shall be retained in the CNSS files.

b.

Recommendations and concerns shall be submitted to the General Hanager - Brunswick Plant and Vice President - Brunswick Nuclear j

Proj ect, within 14 days of completion of the review, c.

A summation of Corporate Nuclear Safety recommendations and concerns shall be submitted to the Chairman / President and Chief Executive Officer; Executive Vice President - Power Sapply and Engineering and Construction; Vice President - Corporate Nuclear Safety and Research; Vice President - Brunswick Nuclear Project; General Manager -

Brunswick Plant; and others, appropriate, at least once every two months.

BRUNSWICK - UNIT 1 6-15 Amendment No.

7__

( BSEP-1 -12)

ADMINISTRATIVE CONTROLS

3. -

AUDITS-(Continued)"

k.

The performance of activities required by the Quality Assurance Program to meet the provisions of. Regulatory Guide 1.21, Revision 1, June 1974, and Regulatory Guide 4.1, Revision 1, April 1975, at least

~once per.12 months.

m 1.

Any other area of facility operation considered appropriate by the Manager - Quality _ Assurance Services Section.

6.5.5.3 Personnel performing the quality assurance audits shall have access to the plant operating records.

RECORDS 6.5.5.4 - Records of audits shall be prepared and retained.

6.5.5.5 Audit reports encompassed by 6.5.5.2 above shall be prepared, approved by the Manager - Quality Assurance Service Section, and forwarded to l

the Executive' Vice President - Power _ Supply and Engineering and Construction; Vice President - Brunswich Nuclear Project; Vice President - Corporate Nuclear Safety and Research; General Manager - Brunswick Plant; and others, as appropriate, within 30 days af ter completion of the audit.

AUTHORITY 6.5.5.6 The Manager - Quality Assurance Services Section under the Manager -

Corporate Quality Assurance shall be responsible for the following:

-a.

The administering of the Corporate Quality Assurance Audit Program.

7 b.

The approval of the individual (s) selected to conduct quality assurance audits.

PERSONNEL 6.5.5.7 Audit personnel shall be independent of the area audited.

6.5.5.8 Selection of personnel for auditing assignments shall be based on experiet.Se or training that establishes that their qualifications are commensurate with the complexity or special nature of the activities to be audited.

In selecting audit personnel, consideration shall be given to special abilities, specialized technical training, prior pertinent experience,

. personal characteristics, and education.

6.5.5.9 Qualified outside consultants or other individuals independent from those personnel directly involved in plant operation shall be used to augment the audit teams when necessary.

Amendment No.

BRUNSWICK - UNIT 1 6-17

(BSEP-1-12)

ADMINISTRATIVE CONTROLS 6.5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM 6.5.6.1 An independent fire protection and loss prevention inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

~

6.5.6.2 An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 36 months.

6.6 REPORTABLE EVENT ACTION 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 Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the Plant Nuclear Safety Ccmmittee - Brunswick Plant and shall be submitted to the Manager -

Corporate Nuclear Safety Section and the Vice President - Brunswick Nuclear Project.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The facility shall be placed in at least HOT SHUTDOWN within two hours.

b..

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President -

~

Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Section shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, c.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the General Manager - Brunswick Plant. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Commission, the Vice President - Brunswick Nuclear Project, and the Manager - Corporate Nuclear Safety Section within 14 days of the violation.

BRUNSWICK - UNIT 1 6-18 Amendment No.

r...

( BSEP-1-12) i ADMINISTRATIVE CONTROLS

! PROCEDURES AND PROGRAMS (Continued) 1.-

Preventive maintenance and periodic visual inspection requirements, and L

'2.

Integrated leak test requirements for each system at refueling

. cycle intervals or. less.

b.-

In-Plant Radiation Monitoring A program which will-ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident

~

conditions. This program shall include the following:

1.

Training of personnel,

' Procedures for monitoring, and 2.

3.

Provisions for maintenance of sampling and analysis equipment.

c.-

Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant -

gaseous effluents, and' containment atmosphere samples under accident-

' conditions. The program shall include the following:

" 1. -

Training-of personnel, 2.

Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysis, equipment.

6.9 REPORTING REQUIRE!iENTS ROUTINE REPORTS 6.9.1 =In addition to the~ applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office unless otherwise noted.

STARTUP REPORTS 6.9.1.1 -A summary report of plant startup and power escalation testing shall

~

be submitted following-(1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of

' fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

BRUNSWICK - UNIT 1 6-20 Amendment No.

~

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.o

.[~,

es g + ~ -.gc (BSEP-1-12) y

~ ADMINISTRATIVE ~ CONTROLS s

>SPECIAL-REPORTS-

.6'9.2s Special reports shall be submitted to the Regional Administrator of the Regional Office 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.

3 t

.m

.a.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.5.1.

b.

Seismic event analysis,- Specification 4.3.5.1.2.

Jc.-

Accident. Monitoring Instrumentation, Specification 3.3.5.3.

d.

Fire. detection instrumentation, Specification 3.3.5.7.-

.O e.

-Reactor coolant specific activity analysis, Specification 3.4.5.

[

~

f.'

'ECCS actuation, Specifications 3.5.3.1 and 3.5.3.2.

N g.'.

Fire suppression systems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3, and 3.7.7.5.

h.

Fire barrier penetration,' Specification 3.7.8.

i..

Liquid Effluents Dose, Specification 3.11.1.2.

j.

Liquid Radwaste Treatment, Specification 3.11.1.3.

k.

Dose:- Noble Gases,. Specification 3.11.2.2.

i 1.

Dose - Iodine-131 Iodine-133 Tritium, and -Radionuclides in Particulate Form, Specification 3.11'.2.3.

m.' -

Gaseous-Radwaste Treatment, Specification 3.11.2.4.

n.

Ventilation Exhaust Treatment, Specification 3.11.2.5.

3 o..

~ Total Dose, Specification 3.11.4.

p..

, Monitoring Program, Specification-3.12.1.b.

q.

~ Primary Containment Structural Integrity, Specification 4.6.1.4.2 6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-:145.2.9-1974.

. 6.10.1 The following records shall be retained for at 1 cast five years:

a.

Records and logs of facility operation covering time interval at each power level.

BRUNSWICK - UNIT l 6-25 Amendment No.

N (BSEP-1-12) p

' ADMINISTRATIVE CONTROLS

. RECORDS RETENTION (Continued)

E

[b.

Records and logs of principal maintenance activities, inspections,

_ repair and replacement of principal items of equipment related to nuclear safety.

I c.

All REPORTABLE EVENTS.

-d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

e.

Records of changes made to Operating Procedures.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

h.

Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a.

Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c.

Records cf facility radiation and contamination surveys.

d.

Records or radiation exposure for all individuals entering radiation control areas.

e.

Records of gaseous and liquid radioactive material released ta the environs.

f.

Records of transient or operational cycles for those facility components identified in Table 5.7.1-1.

g.

Records of reactor tests and experiments.

h.

Records of training and qualification for current members of the plant staff..

1.

Records of inservice inspections performed pursuant to these Technical Specifications.

J.

Records of Quality Assurance activities required by the QA tianual.

BRUNSWICK - UNIT 1 6-26 Amendment No.

w-

e

' ~

(BSEP-1-12)

ADMINISTRATIVE CONTROLS RECORDS RETENTION (Continued) k.

Records of reviews performed for changes cade to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

1.

Records of the service lives of all hydraulic and mechanical snubbers referenced in Section 3.7.5 including the data at which the service l

life commences and associated installation and maintenance records.

Records of analyses required by the radiological environmental m.

monitoring program.

n.

Records of (1) meetings of the PNSC, (2) meetings of the previous off-site review organization, the Company Nuclear Safety Committee (CNSC), (3) the independent reviews performed by the Corporate Nuclear Safety Section, and (4) the independent reviews performal by the Corporate Quality Assurance Audit Program, Performance Evaluation Unit.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved,

maintained and adhered to for all operations involving per> nnel radiation ex posure.

6.12 HIGR RADIATION AREA 2

6.12.1 In lieu of the " Control Device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is 1000 mrem /hr or less 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)*.

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 radiation monitoring device which continuously indicates the a.

radiation dose rate in the area.

b.

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

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt f rom the RWP issuance requirement during the performance of their assigned ra11ation protection duties, provided they comply with approved raliation protection procedures for entry into high radiation areas.

BRUNSWICK - UNIT 1 6-27 Amend ment No.

y;:

3; c -- -

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, BSEP-1-12)-

(

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LADMINISTRATIVE CONTROLS:

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[~

HIGH' RADIATION AREA 1(Continued)

~

c.

- An individual qualified 'in radiation protection procedures who is

^

equipped with a.. radiation dose. rate monitoring. device. This l individual.shall be responsible for providing positive control over -

f['

J-

- the - activities within the area and shall perform periodic radiation a

surveillance at the frequency specified.by the facility Health Physicist: sin the Radiation Work Permit.-

'6.12.211The_ requirements of.6.12.1 above shall also apply to each high radiation area in which the intensity of radiation is greater.than e'

.1000 area /hr.. - In addition,. locked doors shall be provided to prevent unauthorized entry.into such areas and the keys shall be maintained under the-

-

• administrative control of the Operations Shif t Foreman on duty and/or the

' Radiation Control Supervisor.

,7 e-

-6.13 0FFSITE' DOSE CALCULATION MANUAL'(ODCM) m 6.13.11 The OFFSITE' DOSE CALCULATION MANUAL (00CM) shall be approved by'the Commission prior to implementation.

(

.6.13;2 Licensee initisted changes to the ODCH:

['~

a.'-

Shall be submitted to' the Corssission in the Semiannual Radioactive Effluent Release Report for the period lin which the change (s) was' A.

ande ef fective. This submittal shall contain:

1..

Sufficient 1y' detailed information to totally support rationale

]

wie.out benefit of additional of supplemental information.

Information submitted should consist of a package of those pages

^,

of the ODCM to be changed with.each page numbered and provided with an approval'and date box,'together with appropriate anal' ses or evaluations justifying the change (s);

y

-2.

A determination that' the change will not reduce the accuracy or

^

_ reliability of dose calculacions or setpoint determinations;

.and,.

^

3.

Documentation of the. fact that the change has been reviewed and found acceptable by_the'PNSC.'

u.

b..

Shall become: ef fective upon review 'and acceptance by the PNSC.

6.14 PROCESS CONTROL PROGRAM (PCP)'

6 L6'.14.1.'The' PROCESS CONIROL PROGRAM (PCP) shall be approved by the Commission ptfor to implementation.

6.14.2 Licenson initiated changes to the PCP:

hendment No.

BRUNSWICK - UNIT 1 6-28 1

(BSEP-1-12)

ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAli (PCP) (Continued) a a.

Shall be submitted to the Commission in the Semiannual Radioactive Ef fluent Release Report for the period in which the change (s) was made. This submittal shall contain:

i 1.

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; 2.

A determination that the change did not reduce the overall conformance of the solidification waste product to existing criteria for solid wastes; and u

3.

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

b.

Shall become effective upon review and acceptance by the PNSC.

6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID WASTE TREATtENT SYSTEMS 6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

a.

Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

l 1.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59; 2.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 3.

A detailed description of the equipment, components, and processes involved and the interf aces with other plant systems; 4.

An evaluation of the change that shows the predicted release of radioactive materials in liquid and gaseous ef fluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5.

An evaluation of the change that shows the expected maximum exposure to an individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 7/ Licensees may choose to submit the information called for in this

~~

Specification as part of the annual FSAR update.

BRUNSWICK - UNIT L 6-29 Amendment No.

(BSEP-1-12)

ADMINISTRATIVE CONTROLS

- MMOR CHANCES TO LIQUID,' GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS (Continued) r 6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

^

7.

An estimate of the exposure to plant operating personnel as a result of the change; and

~

8.

Documentation of the fact that the change was reviewed and found acceptable-to the PNSC.

b.

Shall become effective upon review and acceptance by the PNSC.

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4 E-1.

BRUNSWICK - UNIT 1 6-30 Amendment No.

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.J ATTACHMENT 2 4J. -

SERIAL: NLS-84-204 1'

BRUNSWICK STEAM ELECTRIC PLANT-PROPOSED TECHNICAL SPECIFICATION PAGES - UNIT 2 i

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SUMMARY

LIST & REVISIONS-BRUNSWICK UNIT 2

-M Comments

_II

- " Reportable Occurrence" changed to " Reportable Event" XV

-Incorporates change to reporting requirements Revised to reflect repagination XVI-Revised.to reflect repagination 1-7 Reportable Event. change incorporated 3/4-3-62 Specification 6.9.1.14.b changel to 6.6.1 3/413-68 Specification 6.9.1.14.b changed to 6.6.1 3/4 4-7 Reference to Specification 6.9.1.12 deleted p,-

3/4 4-10 Revised to reflect Special Report changes M'

Mathematical symbols have been put into words

. Typographical ' error corrected ("of the" 'added )

-3/4 6-6 Revises to reflect reporting of abnormal primary containment degradation pursuant to Specification 6.9.2 3/4 11-22 Specification 6' 9.1.14.b changed to 6.6.1 7

6-3 Revised to reflect organizational changes A-6-4 Revised to reflect organizational changes u.

".[I 6-5 Revisai to reflect organizational changes e;>0 6-6 Typographical error corrected (" Unit 2" changel to e

" Unit 1")

N.

6-8

" Manager - Operations" changed to "Direc tor - Training" "D

" Manager - Plant' Operations" eliminated from

,.q

'6-11 Section 6.5.3.3, PNSC membership 6-12 Reportable Events change incorporatai 6-13

" Reportable Occurrence" changal to " Reportable Events"

. 6-14 -'

Comma aidai in Section 6.5.4.6

.A

• i 6-15 Revised to reflect title changes

[. E -

"Bi-monthly" changel to "once every two months" 6-17

'" Principal QA Specialist - Performance Evaluation Unit" -

changel to " Manager - Quality Assurance Service Section" 6-18' Reportable Events change incorporatal

^

4 h-6-20 Reportable Events change incorporatal

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.P,,a, ge Comments 6-25 Sections 6.9.1.12, 6.9.1.13, and 6. 9.1.14 d eleted Item 6.9.2.c added Item 6.9.2.q added Section 6.9.2. re-ordered 4

Repaginated 6-26

" Reportable Q:currences" changed to ' "Reportab.e Events" Repaginated 6 Reference to snubber Table 3.7.5-1 revised to reflect table deletion per our submittal of May 7,1984 Repaginated

'6-28 "Dosa" aided to "Of fsite Caleulation Manual" Typographical error corrected ("toally" changed to

" totally")

Repaginated 6-29 Repaginated 6-30 Repaginatai

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~ (9974 MAT /c fr)

c (BSEP-2-II)

-INDEX DEFINITIONS SECTION 1.0 DEFINITIONS (Continued)

PAGE PHYSICS TESTS..................................................

1-5 PRESSURE BOUNDARY LEAKAGE......................................

1-5 PRIKARY CONTAINMENT INTEGRITY..................................

1-5 a

PROCESS CONTROL PROGRAM (PCP)..................................

1-6 PURGE - PURGING................................................

1-6 RATED THERMAL POWER............................................

1-6' REACTOR PROTECTION SYSTEM RESPONSE TIME........................

1-6

1. e REFERENCE LEVEL ZERO...........................................

1-ik ~

l REPORTABLE EVENT................................................

1-7 ROD DENSITY....................................................

1-7 SECONDARY CONTAINMENT INTEGRITY................................

1-7 SHUTDOWN MARGIN................................................

1-7 b

SITE BOUNDARY..................................................

1-7 SOLIDIFICATION.................................................

1-7 SOURCE CHECK...................................................

1-7 SPIRAL RELOAD..................................................

1-8 SPIRAL UNLOAD..................................................

1-8 STAGGERED TEST BASIS......................................'.....

1-8 THERMAL POWER..................................................

1-8 TOTAL PEAKING FACTOR...........................................

1-8 UNIDENTIFIED LEAKAGE...........................................

1-8 UNRESTRICTED AREA..............................................

1-8 VENTILATION EXHAUST TREATMENT SYSTEM...........................

1-9 VENTING........................................................

1-9 FRE QUENCY NOTATION, TAB LE 1.1..................................

1-10' OPERATIONAL CONDITIONS, TABLE 1.2..............................

1-11 BRUNSWICK - UNIT 2 II Amendment No.

(BSEP-2-ll)

INDEX

[

ADMINISTRATIVE CONTROLS SECTION.

PAGE 6.5.4 CORPORATE NUCLEAR SAFETY SECTION.

Function................................................

6-13 Organization............................................

6-13 Review..................................................

6-14 Records.................................................

6-15 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Function................................................

6-16 Audits..................................................

6-16 Records.................................................

'6-17 Authoricy...............................................

6-17 Personnel...............................................

6-17 4

6.5.6 OUTSIDE AGEhCY INSPECTION AND AUDIT PROGRAM.............

6-18 6.6 REPORTABLE EVENT ACTI0N....................................

6-18 l'

e 1

6.7 S AFETY LIMIT V IO LATIO N.....................................

'6-18 s

6.8 PROCEDURES AND PR0 GRAMS....................................

6-19 6.9 REPORTING REQUIREMENTS Ro u t ine Re p o r t s........................................

6-20 Startup Reports.........................................

6-20 Annual Reports..........................................

6-21

~

Personnel Exposure and Monitoring Report................

6-21 Annual Radiological Environmental Operating keport......

6-22 Semiannual Radioactive Ef fluent Release Report..........

6-23 Monthly Operating Reports...............................

6-24 Special Reports.........................................

6-25 4

BRUNSUICK - UNIT 2 XV' Amendment No.

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(BSEP-2-11)

INDEX

' ADMINISTRATIVE CONTROLS

_SECTION PAGE 6.10 RECORD RETENTION..........................................

6-25 6.11 RADIATION PROTECTIO N PR0 GRAM........'......................

6-27 6.12-HIGH RADIATION AREA.......................................

6-27 6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM)....................

6-28 6.14 PROCESS CONTROL PROGRAM (PCP).............................

6-28 6.15 MAJOR CHANGES TO LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS..........................

6-29 r

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h BRUNSWICK - UNIT 2 XVI Amendment No.

- - ~

(BSEP-2-11)

DEFINITIONS REPORTABLE EVENT A REPORTABLE EVENT shall be any of those1 conditions specified in Section 50.73 Jto 10 CFR Part 50.

ROD DENSITY J

' ROD' DENSITY shall be the number of control rod notches inserted ~as a fraction sof the total number of notches. All rods fully inserted are equivalent to 100%. ROD DENSITY.

SECONDARY CONTAINMENT' INTEGRITY

' SECONDARY CONTAINMENT INTEGERITY shall exist when:

All automatic Reactor Building ventilation system isolation valves or a.

dampers are OPERABLE or secured in the isolated position.

b.

.The standby gas treatment system is OPERABLE pursuant to Specification 3.6.6.1.

c.;

At least-one door in each access to the Reactor BuildinS is closed.

d.

'The sealing mech nism' associated with each penetration (e.g.,-welds,

~

g bellows, or 0-rings) is.0PERABLE.

SHUTDOWN MARGIN

. SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be suberitical~ assuing that all control rods capable of insertion are fully.

inser,ted except for the analytically determined highest worth rod which is assued$$to be fully withdrawn, and the reactor is in the shutdown condition, cold,-68'F, and Xenon-free.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased. nor otherwise controlled by the licensee, as defined by Figure 5.1.3-1.

SOLIDIFICATION' SOLIDIPICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK

'~

A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

. BRUNSWICK - UNIT 2 1-7 Amendment No.

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\\ - 4 *1 (BSEP-2-11) 9 INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRLMENTATION

. LIMITING CONDITION FOR OPERATION 3.3.5.8.The radioactive liquid effluent monitoring instrumentation channels shown in Table ~ 3.3.5.8-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alare/ trip setpoints_shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

F APPLICABILITY: As shown in Table 3.3.5.8-1.

ACTION:-

With a radioactive liquid effluent monitoring instrumentation channel a.

alarm / trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, declare the

-channel inoperable, or change the setpoint so it is acceptably

~

conservative.

b.-

With less than one radioactive liquid effluent monitoring instrumen-tation channel in each release pathway OPERABLE, take the ACTION shown in Table 3.3.5.8-1.

Return the instruments to OPERABLE status

~

within 30 days or, if unsuccessful, explain in the next Semiannual Radioactive Ef fluent Release. Report why the inoperability was not corrected in a timely manner.

The provisions of Specifications 3.0.3, 3.0.4, and 6.6.1 are not l

c.

applicable.

SUR&LANCEREQUIREMENTS

-4.3.5.8 Each radioactive liquid effluent monitoring instrumentation channel shall be ' demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CRANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.5.8-1.

NOTE: See Bases 3/4.3.5.8.

1 BRUNSWICK - UNIT 2 3/4 3-62 Amendment No.

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-r (BSEP-2-11) 1 INSTRUMENTATION RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION

/

LIMITING CONDITION FOR OPERATION 3.3.5.9 The radioactive gaseous effluent monitoring instrumentation channdia shown in Table 3.3.5.9-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The setpoints shall be determined in accordance with the methodology as described in the OFFSITE DOSZ CALCULATION MANUAL (ODCM).

APPLICABILITY:

As shown in Table 3.3.5.9-1.

ACTION:

a.

With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, without delay suspend the release _of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably ' U conservative.

f

'l b.

Withlessthanoneradioactivegaseouseffluentmonitoribg' instrumentation channel OPERABLE, take the ACTION shown in Table 3.3.5.9-1.

Return the instruments to OPERABLE status within 30 days or, if unsuccessful, explain in the next Semiannualr3adioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c.

The provisio s of Specifications 3.0.3, 3.0.4, and 6.6.1 are not applicable.

t:

SURVEILLANCE REQUIREMENTS

~

4.3.5.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CRANNEL CHECK,. SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST. at the frlquencies d

shown in Table 4.3.5.9-1.

NOTE: See Bases 3/4.3.5.9.

f BRUNSWICK - UNIT 2 3/4 3-68 Amendnent No.

u:

c=

i (BSEP-2-11) p REACTOR COOLANT SYSTEM 3/4.4.4 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.4 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.4-1.

APPLICABILITY: At all times.

/

ACTION:

a.

In OPEP.ATIONAL CONDITION 1, 2, and 3:

1.

With the conductivity or chloride concentration exceeding the limits specified in Table 3.4.4-1, but less than 10 unho/cm at 25'c and less than 0.5 ppa, respectively, operation may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and this condition need not be reported to the Commission provided that operation under these conditions shall not exceed 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> per year. The provisions of '

vr.'

Specification 3.0.4 are not applicable.

C 2.

With the conductivity or chloride concentration exceeding the limits specified in Table 3.4.4-1 for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during one continuous time interval or with the conductivity exceeding 10 paho/cm at 25'c or chloride exceeding 0.5 ppm, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN

.f.

within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

At all' other times' with the conductivity and/or chloride concentration of the reactor coolant in excess of the limit specified in Table 3.4.4-1, restore the conductivity and/or chloride fgp,concentrationtowithinthelimitwithin48 hours.

c.

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I BRUNSWICK - UNIT 2 3/4 4-7 Amendment No.

s

(BSEP-2-11)

~ REACTOR COOLANT SYSTEM L-3/4.4.5 ' SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION

-3.4.5 The specific' activity of the reactor coolant shall be limited to:

~

less than or equal to 0.2 pCi/ gram DOSE EQUIVALENT I-131, and a.

b.

less than or equal to 100/E' pC1/ gram.

' APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

In OPERATIONAL CONDITION 1, 2, and 3, with the specific activity of a.

the reactor coolant; 1.

Greater than.0.2 pCi/ gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 pCi/ gram, operation may continuerfor up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that operation under these conditiohs shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

y 2.

Greater than 0.2 pCi/ gram DOSE EQUIVALENT I-131 for more than, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 pCi/ gram, be in at least HOT SHUTDOWN with the main steam c

line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.

Greater than 100/E' pCi/ gram, be;,in at least HOT SHUTDOWN with l

the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

In OPERATIONAL CONDITION 1, 2, 3, or 4, 1.

With the specific activity of the primary coolant greater-than 0.2 JCi/ gram DOSE EQUIVALENT I-131 or greater than 100/E pCi/ gram, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 at least once per-4 hours until the

. specific activity of the primary coolant is restored to within its limits.

In lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report-that defines the results l

of'the specific activity analyses and the time duration when the specific activity of the coolant exceeded 0.2 pCi/ gram DOSE l

~

~

EQUIVALENT I-131 together with the below additional information.

t Y

BRUNSWICK - bn'IT 2 3/4 4-10 Amendment No.

n a

(BSEP-2-11)

~

~CONTAIMIENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY

~ LIMITING CONDITION FOR OPERATION 3.6.1.4 The structural integrity of the primary containment shall.be maintained:st a level consistent with the acceptance criteria in Specification 4.6.1.4.

APPLICABILITY: OPERATIONAL' CONDITIONS 1, 2, and 3.

ACTION:'

'With the structural-integrity of the primary containment not conforming to the i., -

'above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212*F.

SURVEILLANCE REQUIREMENTS h

j 4

s 4.6.1.4.1 The structural-integrity of the exposed accessible interior and'

~

exterior surfaces of the primary containment, including the liner plate, shall.

be. determined during the shutdown for each Type A containment leakage rate 6

test by a visual inspection of those surfaces. This inspection shall be

x,

performed prior,to the Type-A containment leakage rate test to verify no 9

apparent changes in appearance or other abnormal degradation.

,. J.,

+

4.6.1.4.2 Reports Any abnormal degradation of the primary containment

~ structure detected during the:above required inspections shall be reported to

~

the Commission pursuant to Specification 6.9.2.. This Special Report shall include'a: description of the condition of the concrete, the inspection

,tocedure, the tolerances on cracking, and the corrective actions taken.

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BRUNSWICK - UNIT 2.

3/4 6-6 Amendment No.

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~7 (BSEP-2-11).

^

RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE

-LIMITING CONDITION FOR OPERATION-F

'3.11.3 The solid'radwaste system shall.be used in accordance with a PROCESS CONTROL 'PROGRM( to process wet radioactive wastes' to meet shipping and burial ground requirements.'

APPLICABILITY: At all times..

' ACTION:

ca.

With the provisions'of.the PROCESS CONTROL. PROGRAM not satisfied, y

suspend shipments of defectively processed or defectively packaged

~-

-solid radioactive wastes from the site.

b.

The provisions of Specifications 3.0.3, 3.0.4, and 6.6.1 are not "I

applicable.

.v u. -

SURVEILLANCE REQUIREMENTS 5"

7 4'11.3 The PROCESS CONTROL PROGRAM s, hall be used to verify the SOLIDIFICATION

??_.

of at least one, representative test specimen'from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins,

~

{ j(-

evaporator _ bottoms, and sodium sulfate solutions).

If any test specimen fails to verify SOLIDIFICATION,.the SOLIDIFICA-s.

-TION of the, batch under test shall be suspended until such time as

~ ~ _

additional test specimens can be obtained. alternative SOLIDIFICATION parameters can be determined 'in accordance with the PROCESS CONTROL

- PROGRAM, and a subsequent test verifies SOLIDIFICATION.~

~

1&hE.iSOLIDIFICATION of the batch may then be resumed using the _ alternative

. SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

t b.

IfLthe initial t'est specimen from a batch of. waste fails to verify 1

SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the.

collection of testing of representative test specimens from each

~

consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION..The s? PROCESS CONTROL PROGRAM shall.be modified as required, as provided in Specification' 6.14, - to assure SOLIDIFICATION of subsequent batches of waste. -

. NOTE: ' See, Bases 3/4.11.3

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LBRUNSWICK - UNIT 2 :

3/4 11-22 Amendment No.

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BRUNSWICK STEAM &.

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. ELECTRIC PLANT PLANT FIRE PROTECTION ORGANIZATION Carolina

. Power & Lignt Co,mpany Figure 6.2.2'-2 Page No. 8-5

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~ TAB LE 6.2. 2-1 MINIMUM FACILITY SHIFT CREW COMPOSITION f

WITH UNIT.1 IN CONDITION 1, 2, OR.3

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POSITION NUNER & INDIVIDUALS REQUIRED TO FILL POSITION.

CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5' 1

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WITH UNIT-1 IN CONDITION 4 OR 5

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- POSITION NUEER T INDIVIDUALS REQUIRED TO FILL POSITION

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CONDITIONS 1, ~ 2, '& 3 CONDITIONS 4 & 5

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l-WITH UNIT 1 DE-FUELED POSITION NUEER & INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS-1, 2, & 3

' CONDITIONS 4 & 5 1((b ) -

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BRUNSWICK - UNIT 2 ~

6-6 Amendment No.

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.m (BSEP-2-11) y ADMINISTRATIVE CONTROLS 6.2.3 ONSITE NUCLEAR SAFETY (ONS)

FUNCTION

- 6.2.3.1 ' The ONS Unit shall function to examine facility operating characterisitics, NRC issues, industry advisories, and other sources which may e

indicate areas for improving facility safety.

RESPONSIBILITIES 4

6.2.3.2 The ONS Unit shall be responsible for maintaining surveillance of

- f acility activities to provide independent verification

• that these-activities

'are performed correctly and that human errors are reduced as much as practical.

AUTHORITY

'6.2.3.3 The ONS Unit shall make detailed recommendations for revised procedures,; equipment modifications, or other means of improving facility safety to the Hanager-Corporate. Nuclear Safety Section.

6.2.4 -SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shif t Technical Advisor 'shall' serve in an advisory capacity to the Shif t Operating Supervisor on matters pertaining to the engineering aspcets assuring safe operation of the unit.

6.3 FACILITY STAFF QUALIFICATION 4'

. 6.3.1 Each member of the facility staff defined in Figure 6.2.2-1 shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable

. positions, except for (1). the Manager - Environmental & Radiation Control who

.shall meet'or exceed the qualifications of Regulatory Guide 1.8, September.1975 and. (2).the Shif t Technic'al Advisor whof shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant during transients and accidents.

6,.,4 TRAINING 6.4.1" A~ retraining and replacement training program for the facility staff Q

shall be maintained under the direction of the Director - Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 -A training program for the Fire Brigade shall ne maintained under the direction of the. Director - Training and shall meet or exceed the requirements l

. of' Section 27 of the NFPA Code-1975.

• Not. responsible for sign-of f function.

BRUNSWICv - UNIT 2 6-8 Amendment No.

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(BSEP-2-ll)

ADMINISTRATIVE CONTROLS 6~S.3 ~ PLANT NUCLEAR SAFETY COMMITTEE (PNSC)

-FUNCTION'

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-6.5.3.1 As an effective means for the regular review, overview, evaluation,

= and maintenance of plant operational safety, a Plant Nuclear Safety Committee

-(PNSC) shall be established..

6.5.3.2 oThe PNSC shall function through the utilization of subcommittees, audits, investigations, reports, and/or performance of reviews as a group.

_ COMPOSITION 6.5.3.3s The PNSC shall be composed.of the: -

Chairman:

General Manager - Brunswick Plant

• f",f-Member:

Manager - Technical & Administrative Support 3

,g.

Member:

Manager - Technical Support Member:

Manager - Operations Member:

Manager - Maintenance Member:

Manager.- Environmental & Radiation Control

^

Member:

Assistant to Plant General Manager Member:

~ Director - QA/QC ~

-Member:

Director - Regulatory Compliance

' Member:

Director - Administrative Support

' ALTERNATES

<3 6.5.3.4 All* alternate members. shall be appointed in writing by the PNSC

. Chairman. to serve on a temporary basis;. however, no more than two alternates shall participate as members at.any one time.

44 '

.6.5.355 All alternates, shall as a minimum, meet equivalent qualification criteria as specified for professional-technical personnel in Section 4.4 of ANSI N18.1-1971.

1HEETING FREQUENCY 3-6.5.3.6 The PNSC shall meet at least once per calendar month and as convened by.the PNSC Chairman or his designated alternate.

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QUORUM

'f 6.5.3.7 The minimum quorum of the PNSC necessary for the performance of th'e PNSC activities of the Technical Specifications shall consist of the PNSC.

~ Chairman or his designated alternate and five members including alternates.

No

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more than two alternates shall be counted toward meeting the mininum quorum requirement.

O,

~*-Or designated alternate.

BRUNSWICK - UNIT 2 6-11 Amendment No.

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(BSEP-2-11)

ADMINISTRATIVE CONTROLS ACTIVITIES-6.5.3.8 llue PNSC activitiesf shall include the following:

Review..of all procedures required by Specification 6.8 and changes la. <

u thereto (and any other' procedures and changes thereto), any of which constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

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b.

Review of all proposed _ tests or experiments that constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

c.

Review of all proposed modifications that constitute an unreviewed safety question er involve a change to the Technical Specifications, prior to implementation.

d.

Review of all proposed changes. to the Technical Specifications or Operating License, prior to implementation.

Review of reports on violations of Technical Specifications including e.

L reports covering. evaluation and recommendations to prevent recurrence to the Vice President - Brunswick Nuclear Project and to the Manager -

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Corporate Ruclear Safety Section.

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f.

Performance of special' reviews,. investigations (or _ analyses), and

_ reports thereon as requested.by the Manager - Corporate Nuclear Safety f>3 Section.

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Review of all' REPORTABLE' EVENTS.

h..

Review of facility operations to detect potential nuclear safety i

hazards..

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Annual review of the. Security Plan.

j.

Annual review of the Emergancy Plan.

k.

Review of accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President -

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Brunswick Nuclear Project and the lboager - Ccrporate !belear Safety Section.

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Review of changes to the PROCESS CONTROL PROGRA!' and the OFFSITE DOSE CALCULATION MANUAL.

BRUNSUICK - UNIT 2 6-12 Amendment No.

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( BSE'. 11)

' ADMINISTRATIVE CONTROLS AUTHORITY 6.5.3.9 If there is a disagreement between recommendations of a majority of the PNSC and the actions contemplated by the General Manager - Brunswick Plant, the PNSC shall provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President -

Brunswidk Nuclear Project and the.Vice President - Corporate Nuclear Safety and nResearch.: 1Bne course determined by the General Manager - Brunswick Plant to be the most conservative -shall be followed.

RECORDS 6.5.3.10 The PNSC shall maintain written minutes of each PNSC meeting that, at a minimum, document,the results of all PNSC activities performed under the

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L provisions of these Technical Specifications.

Copies shall be provided to the

.Vice President - Brunswick Nuclear Proj ect and the Manager - Corporate Nuclear Safety Section.

6.5.4 CORPORATE NUCLEAR ~ SAFETY SECTION

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FUNCTION-6.5.4.1 The Corporate' Nuclear Safety Section (CNSS) of the Corporate Nuclear Safety.& Research Department shall function to provide independent review of significant plant changes, - tests, and procedures; verif y that REPORTABLE EVENTS l

are investigated in a timely manner and corrected in a manner that reduces the probability of recurrence of such events; and detect tr>.nds that may not -be apparent to a day-to-day observer.

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ORGANIZATION

-6.5.4.2 The individuals assigned responsibility for independent reviews shall be specified in technical disciplines. These individuals shall collectively have the

' experience and competence required to review activities in the following areas:

p q-a.

nuclear power plant operations m

b.

nuclear engineering

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chemistry and radiochemistry

.c.

d.

metallergy e.

non-destructive testing f.

-instrumentation and control g.

radiological safety h.

mechanical and electrical engineering 1.

administrative controls BRUNSWICK - UNIT 2 6-13 Amendment No.

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( BSEP-2-11) f-4 ADMINISTRATIVE' CONTROLS J

ORGANIZATION (Continued)

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. seismic 'and environmental k.

quality assurance practices

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1.. Otheri. appropriate fields associated with the unique characteristics -

of'the' nuclear power plant.

X 6'.5.4.3.

The - Manager Corporate Nuclear Safety Section shall have an academic

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degree in an engineering or related field and, in addition, shall have a

-minimum of ' ten years related experience, of which 'a minimum of _ five years

shall'be'in the operation and/or" design of nuclear power plants.

6;5.4.41 The independent safety review program reviewers shall have an -

academic-degree in an engineering or related field or equivalent and, in addition, shall have a minimus of five years related experience.

'An individual may: possess. competence in more than 'one specialty 6.5.4.5 c

-area.- If: sufficient' expertise is not available within the Corporate Nuclear Safety Section, competent.indtviduals from other Carolina Power & Light.

, Company organizations or outside consultants shall be utilized in performing

", --independent reviews and ~investigatiod4.

9' yi 6.5.416 At least three individuals, qualified as discussed in. 6.5.4.4.above, l.

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shall: review each' item submitted under the requirements of Section 6.5.4.9.

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L6.5.4.7. Independent' safety reviews shall be performed-by individuals not directly involved with the activity under review or responsible for the l activity under. review.

6.5.4.8 The Corporate Nuclear Safety Section independent safety review ce

-(program shall be conducted in accordance with written, approved procedures.

REVIEW

- 6.5.4.9'.The Corporate Nuclear Safety Sectica shall perform reviews of the following-

-.The 1 safety evalcations. for 1) enanges to procedures required by-a.

j Specification 6.8, 2) modifications =of equipment or systems, and

3) tests or experiments that constitute a change to the safety analysis report to verify that sudi actions did not constitute an 3

'unreviewed safety question or involve a change to the Technical Specifications. ' Implementation may proceed prior to completion of-

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.this review, b.

. Proposed changes to procedures required by Specification 6.8, and proposed modifications that constitute an unreviewed safety question as defined in 10 CFR 50.59 or a change to the Technical Specifications, prior to implementation.

BRUNSWICK - UNI'T 2 6-14 Amendment No.

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05 SEP-2-11)-

' ADMINISTRATIVE CONTROLS REVIEW.(Continued);

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k Proposed tests or experiments that involve an unreviewed safety c..

. question as. defined -in -10 CFR 50.59 or a change to the. Technical

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Specifications, prior to implementation.

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d.

~ Proposed changes to the Technical Specifications. and Operating

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License.-

-Violations,. deviations, and events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written e.

notification to the Commission, such as:

1.-

. Violations of applicable codes,' regulations,-orders, Technical Specifications,- license requirements, and internal procedures or instructions having nuclear safety significance.

2.

.Significant operating abnormalities.or deviations from normal land ' expected performance of plant safety-related structures, systems, or components.

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Reports and minutes of the PNSC.

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Any other-matter-inuelving safe operation of the nuclear power plant that the Manager - Corporate Nuclear Safety Section deems appropriate

9; 44 for consideration or which is referred to the' Manager -_ Corporate Nuclear Safety Section by the on-site operating organization or other 2J functional organizational units within Carolina Power & Light

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' Company.

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6.5.4.10 Review of items considered under 6.5.4.9(e) through (g' above shall

' include the results of any investigations made and the recommendations 1resulting from. these investigations-to prevent or reduce the probability of recurrence of the event.'

RECORDS 6.5.4.11 Records of Corporate Nuclear Safety Section reviews, including recommendations and concerns, shall be prepared and distributed as indicated below:

Copies of documented reviews shall be retained in the CNSS files.

.a.-

b.

Recommendations and concerns shall be submitted to the General Manager - Brunswick Plant and Vice President - Brunswick Nuclear 35 Project,-within 14 days of completion of the review.

A summation of Corporate Nuclear Safety recommendations and concerns c.

s shall be submitted to the Chairman / President and Chief Executive Officer; Executive -Vice President - Power Supply and Engineering and LConstruction; Vice President - Corporate Nuclear Safety and Research; l

Vice President - Brunswick Nuclear Project; General Manager - Brunswick Plant; and others, appropriate, at least once every two months.

BRUNSWICK - UNIT 2 6-15 Amendment No.

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(BSEP-2-11) -

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CADMINISTRATIVE CONTROLS J _^

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? AUDITS (Continued) s

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TheLperformance of. activities required by, the; Quality Assurance

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[ Program to me,et' the provisions of Regulatory Guide :1.21, Revision _1,.

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7 June:1974', and. Regulatory Guide 4.1,' Revision 1, April 1975, at least.

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once'per'12: months.

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Any other, area Jof facility operation considered appropriate by the Manager - Quality Assurance Services Section.

r 6.5.5.3.- Personnel performing the' quality ' assurance audits shall have access

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,to.the plant operating records.'

RECORDS

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6.5.5.4 Records of audit's shall be prepared:and retained.

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i6.525.5 Audit reports -encompassed by 6.5.5.2 ' above shall be prepared,

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approved by the Manager Quality Assurance : Service Section, and forwarded to l

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7 tha. Executive;Vice President - Power Supply a'nd Engineering and Construction; y

q Vice - PresidentT-Brunswick Nuclear ' Project; Vice President - Corporate Nuclear

Safety and Research;. General Manager - Brunswick Plant; and others,;as y

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" appropriate, 'within 30 days af ter completionTfTe audit.;

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. AUTHORITY-1, s

s K 6.'5;5.6 ; The Manager - Quality Assurance Services.Section under the Manager -

Corporate Quality Assurance:shall:be responsible.for the following:

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The administering of the Corporate' Quality ' Assurance Audit - Program.

a.;

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'b.

The approval of ' the individual (s) : selected. to' conduct quality

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-assursace audits.

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,Q, iPERSONNCL'

. Audit peisonnel shall be independent of the.trea audited.

6.5.5.7 6.5.5.8 Selection of personnel.for auditing assignments.shall be based on

experience or training that establishes that their qualifications are JC L eoamensurate with the complexity.or special nature of the activities to be audited.. In selecting audit personnel, consideration shall be given to

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1special' abilities, specialized' technical training, prior pertinent experience,.

personal characteristics, and education.

I, 6.5.5.9 LQualified outside consultants or other ' individuals ' independent from p"

.'those personnel directly involved in plant operation shall be used to' augment the audit teams when necessary.

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BRUNSWICK - UNIT'. 2 6-17 Amendment No..

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(BSEP-2-11) s ADMINISTRATIVE CONTROLS 6.5.6 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM

'6.5.6.1 -An independent fire protection and loss prevention. inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

6.5.6.2.An inspection and audit of the fire protection and loss prevention

j

. program shall be performed by an outside qualified fire consultant at intervals no greater than 36 months.

6.'6 REPORTABLE EVENT ACTION s.

6.6.1 The following actions shall be taken-for REPORTABLE EVENTS:

a.

The Commission shall be notified and a report submitted pursuant to P._,

the requirements of Section.50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENr shall be reviewed by the Plant Nuclear Safety Committee - Brunswick Plant and shall be submitted to the Manager -

Corporate Nuclear Safety Section and the Vice President - Brunswick, Nuclear Project.

6.7 SAFETY LIMIT VIOLATION

~ ~

6.7.1. ; The following actions shall be taken. in the event a Safety Limit is lS violated:

' a. :

The facility shall be placed in at least HOT SHUTDOWN within two hours.

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President -

' Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Section shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

A Safety Limit Violation Reporr shall be prepared.

The. report shall be reviewed by the General Manac.r - Brunswick Plant. This repor shall describe (1) applicable circumstances preceding the violation, (2) -ef f ects of the violation upcu facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Linit-Violation Report shall be submitted to the Commission, the' Vice President - Brunswick Nuclear Project, and the Manager - Corporate Nuclear Safety Section within 14 days of the violation.

BRUNSWICK - UNIT 2 6-18 Amendment No.

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_ (3SEP-2-11)'

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.-ADMINISTRATIVE CONTROLS'

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. PROCEDURES-AND PROGRAMS (Continued)-

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11. L-( Preventive maintenance and periodih visual inspection E

. requirements,,and.

2.'

Intiegrated leak?sest. requirements for each system atirefueling

. cycle ~ intervals or less.-

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b.In-PlantRadiationMonitorin[

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fA program which will ensure the capability to accurately determine -

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the airborne iodihe concentration in' vital areas under accident

. conditions.' This' program shall include:the following:

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Training of personnel,-

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Procedures for monitoring, and.

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3.

- Provisions for maintenance of sampling and analysis equipment.

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Post-Accident Sampling

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" A program which.will ensure the capability to - obtain and analyze reactor coolant, radioactive iodines, and particulates. in plant-i.

gaseous effluents', and.containtent atmosphere samples.under accident' y'

conditions. -The program.shall include the following:

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1.

Training of. personnel, y,-

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Procedures-for sampling and analysis, and.

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3.1 Provisions for maintenance of sampling and' analysis equipment.

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'6.9' REPORIING REQUIREMEYrS r$

'ROOTINE REPORTS

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' 6.9.1:' In addition to the applicable reporting _ requirements of Title 10, Code'

-of Federal Regulations, the following reports shall be submitted to the l Regional Administrator' of the Regional Of fice ' unless otherwise noted. --

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~STARTUP REPORTS _

9 4

J 6.9.1.'If A summary report of plant startup and pouer es'calation testing shall i,

Ebe' submitted lfollowing (1) receipt of an operating-license, (2) amendment to "the license ' involving a planned increase in power level, (3) installation of

' fuel that has:a different design or has been manufactured by a different fuel

-supplier, and (4) modifications that may have significantly altered the

_ nuclear, thermal, or hydraulic performance of the plant.

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= BRUNSWICK - UNIT-2 6-20 Amendment No.

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A ADMINISTRATIVE CONTROLS i

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office 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.

-Inoperable Seismic Monitoring Instrumentation, Specification 3.3.5.1.

Seismic event analysis, Specification.4.3.5.1.2.

b.

'c.

Accident Monitoring Instrumentation, Specification 3.3.5.3.

d..

Fire detection instrumentation, Specification 3.3.5.7.

e.

Reactor coolant specific activity analysis, Specification 3.4.5.

f.-

ECCS actuacion, Specifications 3.5.3.1 and 3.5.3.2.

o 3

Fire suppression systems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3, and 3.7.7.5.

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Fire' barrier penetration, Specification 3.7.8.

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Liquid Effluents Dose, Specification 3.11.1.2.

'j.

Liquid Radwaste Treatment, Specification 3.11.1.3 k.

Dose - Noble Gases, Specification 3.11.2.2.

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Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form, Specification 3.11.2.3.

4

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Gaseous Radwaste Treatment, Specification 3.11.2.4.

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Ventilation Ethaus: Treatment, Specification 3.11.2.5.

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Total Doso, Specification 3.11.4.

p.

Monitoring Program, Specification 3.12.1.b.

q.

Primary Containment Structural Integrity, Specification 4.6.1.4.2.

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974.

6.10.1 The following records shall be retained for at least five years:

a.-

Records and logs of facility operation covering time interval at each power level. -

BRUNSWICK - UNIT 2 6-25 Amendment No.

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(BSEP-2-ll)

~ ADMINISTRATIVE CONTROLS RECORDS RETENTION -(Continued)

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b..

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related-to nuclear-safety..

c.

All ~ REFORTABLE EVENTS.

d.

Records oof surveillance activities, inspections, and calibrations required by these Technical Specifications.

e.

Records of changes made to Operatin's Procedures.

f.

' Records of radioactive shipments.

g.-

Records of sealed source and fission detector leak tests and results.

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h.

Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall' be -retained for the duration of the

' Facility Operating License:

a.

Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis 4-Report.

b.

Records of new and ' irradiated fuel inventory, fuel transfers and

. f assembly burnup histories.

c..

Records of facility radiation and contamination surveys.

r d.

Records or radiation exposure for all individuals entering radiation control areae, e.

' Records of gaseous and liquid radioactive material released to the

-environs.

f.

Records of transient or operational cycles for those facility ecmponents identified in Table 5.7.1-1.

g.

Records of reactor tests and experiments.

h.

Records of trainind and qualification for current membecs of the plant staff.

1.

Records of inservice inspections performed pursuant to these Technical Specifications.

J.

. Records of Quality Assurance activities required by the QA Itanual.

BRUNSWICK - UNIT 2 6-26 Amendment No.

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. ADMINISTRATIVE CONTROLS RECORDS RETENTION (Continued)

'k.-

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

1.

Records of the service lives of all hydraulic and mechanical snubbers referenced in Section 3.7.5 including the data at which the service l

life commences and associated installation and maintenance records.

Records of analyses required by the radiological-environmental m.

monitoring program.

n.

Records of (1) meetings of the PNSC, (2) meetings of the previous

~

off-site review organization, the Company Nuclear Safety Committee (CNSC), (3) the independent reviews performed by the Corporate Nuclear Safety Section, and (4) the iulependent reviews performed by the Corporate Quality Assurance Audit Program, Performance Evaluation Unit.

6.11 RADIATION ~ PROTECTION PROGRAM -

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-6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved,

- maintained and adhered to for all operations involving personnel radiation exposure.

h 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " Control Device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be a

. controlled by requiring issuance of a Radiation Work Permit (RWP)*.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

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

b.

A radiation monitoring device wnich continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is re:eived.

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

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the perforaance of their assigned. radiation protection duties, provided they comply with approved radiation protectioa procedures for entry into high radiation areas.

'BRUNSUICK - UNIT 2 6-27 Amendment No.

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(BSEP-2-11) 4 ADMINISTRATIVE CONTROLS

[HIGH RADIATION AREA (Continued) 4 '

c.' ' An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.. This individual. shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation

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2..

surveillance at' the frequency specified.by the facility Health Physicist'in the Radiation Work Permit.

6.12.2 l The' requirements of 6.12.1 above shd11 also apply to each high 1

radiation area in which the intensity of. radiation is greater than

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1000~arem/hr. In addition, locked doors shall be provided to prevent.

unauthorized entry into such areas and the keys shall be maintained under the 7

administrative control of the Operations Shif t Foreman on duty and/or the

' Radiation Control Supervisor.

'6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM)

' 6.13.1' The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be approved'by the l

A Commission; prior to implementation.

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6.13.2, Licensee initiated changes to the ODCM:

a.

Shall-be submitted.to the Commission in the Semiannual Radioactive

).j.

Effluent Release Report for the period in-which the change (s) w.as ed '

made ef fective. ' This submittal shall. contain:

MA 1.

Sufficiently-detailed information to totally support rationale l-

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without benefit of additional of supplemental information.

Information submitted should consist of a package of those pages-of the ODCM.to be changed with each page numbered' and provided with an ayoroval and-date box, together with appropriate analyses 'or evaluations justifying the change (s);

2.

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

and, 3.

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

b..

Shall becoma ef fective upon review and acceptance by the PNSC.

6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 The PROCESS CONTROL PROGRAM (PCP) shall be approved by the Commission prior.to-implementation..

~6.14.2 - Licensee initiated changes to the PCP:

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BRUNSUICK - UNIT 2-6-28 Amendment No.

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(BSEP-2-11)

ADMINISTRATIVE CONTROLS L

PROCESS CONTROL PROGRAM (PCP) (Continued) a.

Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:

Sufficiently detailed information to totally support the 1.

rationale for the change without benefit of additional or supplemental information; 2.

A determination that the change did not reduce the overall conformance of the solidification waste product to existing criteria for solid wastes; and 3.

Documentation of the fact that the change has bean reviewed and found acceptable by the PNSC.

b.

Shall become effective upon review and acceptance by the PNSC.

6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID UASTE TREATMENT SYSTEMS 6.15.1 Licensee initiated major changes to the radicactive waste systems (liquid, gaseous, and solid):

a.

Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion cf each change shall contain:

1.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Fart 50.59; 2.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplenental information; 3.

A detailed descripcion of the equipcent, components, and processes involved and the interf aces with other plant systems; 4.

An evaluation of the change thac shows the predicted release cf radioactive caterials in liquid and gaseous ef fluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5.

An evaluation of the change that shows the expected maximum exposure to an individual in the UNRESTRICTED AREA and to the-general population that differ from those previously estimated in the license application and amendments thereto; Z/ Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.

BRUNSWICK - UNIT 2 6-29 Amendment No.

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ADMINISTRATIVE CONTROLS MAJOR CHANGES TO LIQUID, GASEQUS, AND SOLID WASTE TREATMENT SYSTEMS (Continued) 6.-

- A ' comparison of the predicted releases of radioactive materials, in-liquid and gaseous effluents and in solid waste, to the octut.1 releases for the. period-prior to when the changes are to

-be made;

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7.

.An estimate of the. exposure co plant operating personnel as a result of the change; and

.8.

Documentation of the fact that the change was reviewed and found acceptable to the PNSC.

b..-

Shall become effective upon review and acceptance by the PNSC.

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.g BRUNSUICK, UNIT 2 6-30 Amendment No.