ML19319C792

Tech Spec Change Request 2 Re Tech Specs 6-1 Through 6-13, Revising Offsite Organization for Technical Support & Facility Mgt

ML19319C792
Person / Time
Site:	Crystal River
Issue date:	05/20/1977
From:	FLORIDA POWER CORP.
To:
Shared Package
ML19319C793	List: ML19319C792 ML19319C969
References
NUDOCS 8003040765
Download: ML19319C792 (23)

Text

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.(~J , . Technical Specificatirn Change R quest No. 2a Replace pages 6-1, 6-2, 6-3, 6-5, 6-6, 6-7, 6-8, 6-9, 6-10, 6-11, 6-12, 6-13 with the attar.hed replacement pages- 6-1, 6-2, 6-3, 6-5, 6-6, 6-7, 6- 8, 6-9, 6-10, 6-11, 6-12, 6-13.

Reasons for Proposed Changes Revisions in the offsite organization for technical support and facility management are not reflected in the existing pages.

The scope of activity of the Nuclear General Review Connitttee has been revised to more closely meet the requirements of ANSI N18.7 1976.

Safety Analysis Justifying Proposed Change This change coes not modify any limits, or diminish any requirement applicable to safety analysis; therefore, no unreviewed safety question is involved.

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6.0 ADMINISTRATIVE' CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Nuclear Plant Manager shall be responsible for overall l facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 '0RGANIZATION 0FFSITE 6.2.1 The offsite organization for facility management and technical j support shall be shown on Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in table 6.2-1.

b. At least one licensed Operator shall be in the control room when fuel is in the reactor.

c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.

d. An individual qualified in radiation protec' tion procedures shall be on site when fuel is in the reactor.

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e. All CORE ALTERATIONS after the fritial fuel loading shall be i directly supervised by either a itcensed Senior Retetor Operator or Senior Reactor Operator Limited to fuel Handling 4

who has no other concurrent responsibilities durin; this operation.

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c ENGINEEfl ENGINEEll Tile GUARD MATJAGE R FOflMAN 5 I I I W COfAPLIANCE CLERK & UUILDING AUDITOll GUARDS STENOGRAfflER OEllVICEMEN I  ; I i ASST. CllEM. & I!EALTil OPERATIONS MAINTEfJANCE TECil SUPPOllT TilAINING R AD. PflOTECTION PilYSICS SUPE RVISOll ENGIN3 Ell ENGINEEll COOllDifJATOR sol ENGINEER SUPERVISOR 1 f-I p Sillf T COMPUTER & - " RESULTS T ECittJICI AfJS TECilNICIANS SO! CONTflOLS ENGR. ENGINEER I

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PLANT OL ENGINEER ENGirJEER I 1 LEGEND: MECllANICAL E LECTftl CAL l TECil SUPPORT ENGINEERING SUPE RVISOR SUPERVISOR , OUPERVISOil ASSISTANT SOL - SENIOR OPERATOR LICENSE g OL - OPERATOR LICENSE -l l MECif ANICS E LECTillCI ATJS TECilNICIANS FACILITY ORGAtlIZATI0ft FIGURE 6.2-2

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.31. Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Protection Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September,1975.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Assistant Nuclear Plant Manager and shall meet or exceed the requirements and recommendations l of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.5 REVIEW AND AUDIT 6.5.1 PLANT REVIEW COMMITTEE (PRC)_

FUNCTION 6.5.1.1 The Plant Review Committee shall function to advise the Nuclear Plant Manager on all matters related to nuclear safety. l COMPOSITION 6.5.1.2 The Plant Review.Comittee shall be composed of the:

Chairman: Assistant Nuclear Plant Manager l Member: Operations Supervisor Member: Technical Support Engineer Member: Maintenance Engineer Member: Chemistry and Radiation Protection Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PRC activities at any one time.

MEETING FREQUENCY 6.5.1.4 The PRC shall meet at least once per calendar month and as convened by the PRC Chairman or his designated alternate.

CRYSTAL RIVER - UNIT 3. 6-5

,i ADMINISTRATIVE CONTROLS QUORUM

- 6.5.1.5 A quorum of the PRC shall consist of the Chairman or his designated alternate and four members including alternates.

RESPONSIBILITIES 6.5.1.6 The Plant Review Committee shallEbe responsible for:

a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Nuclear Plant fianager to affect l nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear Safety.

c. Review of all proposed changes to the Appendix "A" Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Director-Power Production and to the Chairman of the Nuclear l General Review Connittee.

! f. Review of 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.

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g. Review of facility operations to detect potential nuclear safety hazards.

h. Psrformance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Nuclear General Review Connittee.

i. Review of the Plant Security Plan and implementing procedures and shall submit reconnended changes to the Chairnan of the 2

fluclear General Review Connittee.

j. Review of the Emergency Plan and implementing procedures and-shall subnit recommended changes to the Chairman of the Nuclear General Revicw Committee.

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CRYSTAL RIVER - UNIT 3 6-6

s AOMINISTRATIVE CONTROLS AUTHORITY 6.5.1.7 The Plant Review Comittee shall:

a. Recomend to the Nuclear Plant Manager written approval l or disapproval of items considered under 6.5.l(a) through (d) abc'e.

b. Render determinations-in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c. 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 Director-Power Production and the Nuclear General Review Comittee of disagreement between the PRC and the Nuclear clant fianager; however, the Nuclear Plant Manager shall have responsibility for resolution of such disagreements pursu&nt to 6.1.1 above.

RECORDS 6.5.1.8 The Plant Review Comittee shall maintain written minutes of each meeting and copies shall oe provided to the Director-Power

Production and Chairman of the Nuclear General Review Comittee. l 6.5.2 NUCLEAR GENERAL REVIEW C0f41ITTEE (NGRC)

FUNCTION 6.5.2.1 The Nuclear General Review Comittee shall function to provide independent review and audit of designated activities in the areas of: ,

a. Nuclear power plant operations

b. Nuclear engineering

c. Chemistry and radiochemistry

d. Metallurgy

e. Instrumentation and control

f. Radiological safety

g. ftechanical and electrical engineering i

h. Quality assurance practices CRYSTAL RIVER - UNIT 3 6-7

ADMINISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The NGRC shall be composed of the Chairman, Vice Chairman, and at least 5 members. No more than a minority of the members shall have line responsibility for operation of the facility. The committee shall collectively have the experience and competence required to review problems in the following areas:

a. Nuclear power plant operations

b. Nuclear engineering

c. Chemistry and radiochemistry

d. Metallurgy

e. Nondestructive testing l

f. Instrumentation and control

g. Radiological safety

h. Mechanical and electrical engineering

1. Administrative controls J. Environmental

k. Quality assurance practices QUALIFICATIONS 6.5.2.3 The following minimum experience requirements shall be established for those persons involved in the independent off-site safety review and audit program:

a. Chairman and Vice-Chairman-Bachelor of Science in engineering or related field and ten years related experience including five years involvement with operation and/or design of nuclear power plants,

b. Member-Bachelor of Science in engineering or related field and five years related experience including three years involverent with operation and/or design of nuclear power plants.

ALTERNATES 6.5.2.4 All alternate members shall be appointed in writing by the NGRC Chairman to serve on a temporary basis; however, no more than two alternates shall participne as voting members in NGRC activities at any one time.

CRYSTAL RIVER - UNIT 3 6-8

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CONSULTANTSL Consultants shall be utilized as determined by the NGRC Chairman 6.5.2.5 tc provide ~ expert advice to the NGRC.

MEETING FREQUENCY 6.5.2.6 The NGRC shall meet at least once per_ calendar quarter during the initial year of the facility operation following fuel loading and at least once per six months thereafter.

QUORUM 6.5.2.7 A quorum of NGRC shall consist of the Chairman or his designated.

alternate and five additional NGRC members, including alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility.

REVIEW 6.5.2.8 The NGRC shall review:

a. The safety evaluations for 1) changes to procedures equipment '

or systems as described in the Safety Analysis Report and 2) tests or experiments not described in the Safety Analysis Report which are completed without prior NRC approval under the provision of Section 50.59, 10 CFR, to verify that such actions did not cons +.itute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems any of which involve a change in Technical Specifications or an unreviewed safety question as defined in Section 50.59, 10 CFR.

c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59,10 CFR.

d. Proposed changes in Technical Specifications or this Operating License.

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

f. Significant operating abnormalities or deviations from normal and expected performance of plant eg)1pment that affect nuclear safety.

i . CRYSTAL RIVER - UNIT 3 6-9

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ADMINISTRATIVE' CONTROLS REVIEW (CONTINUED) g.- 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.

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures,

• systems, or components.

1 AUDITS 6.5.2.9 Audits of facility activities shall be perfonned under the cognizance of the NGRC. These audits shall encompass:

a. The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire facility staff at least once per 12 months. ,

c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.

d. - The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.

e. The Facility Emergency Plan and implementing procedures at least once per 24 months.

f. The Facility Security Plan and implementing procedures at least once per 24 months.

g. Any other area of facility o)eration considered appropriate by the NGRC or the Senior Vice President-Engineering and Construction. l l

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6.5.2.10 The NGRC shall report to and advise the Senior Vice President- I Engineering and Construction on those areas of responsibility specified in Sections 6.5.2.8 and 6.5.2.9.

l CRYSTAL RIVER - UNIT 3 6-10

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ADMINISTRATIVE CONTROLS RECORDS 6.5.2.11 Records of NGRC activities shall be prepared, approved and distributed as indicated below:-

a. Minutes of each NGRC meeting shall be prepared, approved and forwarded to the Senior Vice President-Engineering and Construction within 30 days following each meeting.

b. - Reports of reviews encompassed by Section 6.5.2.8 above, shall be prepared, approved and forwarded to the Senior Vice President-Engineering and Construction within 30 days following completion of the review, s

, c. Audit reports encompassed by Section 6.5.2.9 aboy', shall be forwarded to the Senior Vice President-Engineering and Construc-tion and to the management positions responsible for the areas l audited within 30 days after completion of the audit.

6.6 REPORTABLE OCCURRENCE ACTION  !

1 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES: l

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

I b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shall be reviewed by the PRC and subnitted to the NGRC and the Director - Power Production. l CRYSTAL RIVER - UNIT 3

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ADMINISTRATIVE CONTROLS 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 STANDBY within one hour.

b.- The Safety Limit violation shall be reported to the Commission, the Director-Power Production and.to the NGRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l

c. A Safety. Limit Violation Report shall be prepared. .The report shall be reviewed by the PRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects-of the violation upon f acility 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 NRCC anc the Director-Power Production within l 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and main- i tained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November, 1972.

b. Refueling operations'

c. . Surveillance and test activities of safety _related equipment. l

d. Security Plan implementation.

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e. Emergency Plan implementation. l 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PRC and approved by the Nuclear Plant fianager prior to implementation and reviewed periodically as l i set forth in adninistrative proccdures.

CRYSTAL RIVER - UNIT 3 6-12 i

s ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made-provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the PRC and approved by the Nuclear Plant Manager within 14 days of implementation.

l 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCE REPORTS 6.9.1 Information to be reported to the Commission, in addition to the reports required by Title 10, Code of Federal Regulations, shall be in accordance with the Regulatory Position in Revision 4 of Regulatory Guide 1.16, " Reporting of Operating Information - Appendix "A" Technic'al Specifications."

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enfcrcement, Region II, 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. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

c. Inoperable Meteorological Monitoring Instrumentation, Specifi-cation 3.3.3.4.

d. Seismic event analysis, Specification 4.3.3.3.2.

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CRYSTAL RIVER - UNIT 3 6-13 I-

z s, TECHNICAL SPECIFICATION CHANGE REQUEST NO. 2h

' Replace page 3/41-16 with the attached replacement page 3/4-1-16.

REASONS FOR PROPOSED CHANGE The existing page 3/41-16 contains a typographical error. Section 3.1.2.9 a.2 shows a minimum of.12500 ppm of baron. Section 3.1.2.8 a.2 of Technical Specifications shows a minimum of 12250 ppm.

Bases page B 3/4 1-2 refers to 12250 ppm.

The correct value is-12250 ppm.

SAFETY ANALYSIS JUSTIFYING PROPOSED CHANGE This change request does not modify any limits used in safety analysis, therefore no unreviewed safety question is involved.

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. REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.9 Each of the following borated water sources shall be OPERABLE:

, a. The concentrated boric acid storage system and associated

! heat tracing with:

1. A minimum contained borated water volume of 5000 gallons,

2. Between 12,250 and 14,000 ppm of baron, and l

3. A minimum solution temperature of 105 F.

b. The barated water storage tank (BWST) with:

1. A contained borated water volume of between 415,200 and 449,000 gallons,

2. Between 2270 and 2450 ppm of boron, and

3. A minimum solution temperature of 40*F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With the concentrated boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1". Ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; re-store the concentrated boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

CRYSTAL RIVER - UNIT 3 3/4 1-16

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TECHNICAL SPECIFICATION CHANGE REQUEST NO 2c Replace page 3/4 3-35 table 3.3-9 with attached replacemer.t page 3/4 3-35.

REASONS FOR PROPOSED CHANGE It is the Decay Heat Removal Temperature and not the Decay Heat Closed Cycle Cooling Temperature that needs to be monitored during remote shutdown. The DHR Temperature gives the operator better information than the DHCCC temperature and DHCCC temperature readout location is in the Decay Heat Pits, which could have a high radiation field. Also, the B&W Standard Technical Specifications, upon which ours are modeled, call for the Decay Heat Removal Temperature to be monitored during remote shutdown. (The Readout Location and Measurement Range are correct for the Decay Heat Removal temperature.)

SAFETY ANALYSIS JUSTIFYING PROPOSED CHANGE The Final Safety Analysis Report for Crystal River 3 lists the decay heat removal temperature in the list of instrumentation required for cold shutdown from out:,ide the control room, therefore no unreviewed safety question is involved.

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TABLE 3.3-9 n

$! REMOTE SiluTDOWN MONITORING INSTRUMENTATION ,

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.:o MINIMUM 2 READ 0UT MEASUREMENT CHANNELS 9 INSTRUMENT LOCATION RANGE OPERABLE c- 1. Reactor Trip Breaker CR0 switchgear room open-close 1 per trip breaker and Indication 124 foot elevation '

1 per secondary trip breaker

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" 4160ES-B switchgear

2. Reactor Coolant --3 Temperature - Th room 108 foot elevation 520-620 F 1 per loop  ;

3. Reactor Coolant Pressure 4160ES-B switchgear rooih 108 foot elevation 0-2500 psig 1

4. Pressurizer Level 4160ES-B switchgear room 108 foot elevation 0-320" 1120 1

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i' 5. Steam Generator Pressure 4160ES-B switchgear l;l room 108 foot elevation 0-1200 psig 1 per steam generator

6. Steam Generator Level 4160ES-B switchgear

. room 108 foot elevation 0-250" H 2O 1 per steam generator

7. Decay lleat Removal 4160ES-B switchgear Temperature room 108 foot elevation 0-300 F 1 per cooler l

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8. Motor Driven Emergency Intennediate Building Feedwater Pressure 95 foot elevation 0-2000 psig '1 per pump

9. Nuclear Se vices Closed Auxiliary Building Cycle Cooling Pumps 95 foot elevation 0-300 psig I Discharge Pressure

10. Nuclear Services Closed Auxiliary Building Cycle Cooling Cooler 95 foot elevation Outlet Temperature 0-250 F 1 per cooler

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i l TECHNICAL SPECIFICATION CHANGE REQUEST NO. 2d Replace page 3/4 3-38 table 3.3-10 with attached replacement page 3/4 3-38 table 3.3-10.

REASONS FOR PROPOSED CHANGE There is only one 0-600 psig pressure transmitter on the reactor coolant system (RC 131 PT) and there is only one 0-600 psig instrument in the control room from this pressure transmitter. Two channels for the 0-600 psig range was inadvertantly determined during the drafting of the CR #3 Technical Specifications.

SAFETY ANALYSIS JUSTIFYING PROPOSED CHANGE A single low-range pressure transmitter is described in the Crystal River 3 Final Safety Analysis Report.Redu;idancy is provided by the 0-2500 psig instruments, therefore no unreviewed safety question is involved.

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n TABLE 3.3-10 4

y POST-ACCIDENT MONITORING INSTRUMENTATION ,'

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$ MINIMUM g MEASUREMENT CHANNELS

, INSTRUMENT RANGE OPERABLE E

l. Power Range Nuclear Flux 0-125% 2

2. Reactor Building Pressure 0-70 psia 2

3. Source Range Nuclear Flux 10-1 to 106 cps 2 ,

4. Reactor Coolant Outlet Temperature 520-620 F 2 per loop

5. Reactor Coolant Total Flow 0-110% full flow 1 w 6. RC Loop Pressure 0-2500 psig 2 2 0-600 psig 1 m 1700-2500 psig 2

7. Pressurizer Level 0-320 inches 2

8. Steam Generator Outlet Pressure 0-1200 psig 2/ steam generator

9. Steam Generator Operating Range Level 0-100% 2/ steam generator

10. Borated Water Storage Tank Level 0-50 feet 2 6

11. Startup Feedwater Flow 0-1.5x10 lb/hr. 2

, TECHNICAL SPECIFICATION CHANGE REQUEST NO. 2e Replace page 3/4 7-19 with the attached replacement page 3/4 7-19.

REASONS FOR PROPOSED CHANGE The applicability of this specification is "At all times". If we were in Mode 6 and the action statements became effective, we would have 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to go to Mode 5. Because watertight doors provide flood pro-tection for safety related systems, components, and structures, it should not be necessary to put the head on the reactor for further protection. The words "at least" are used throughout the Technical Specifications to denote a less important or more safe Mode (i.e. Mode 3 less important than Made 2, etc.). Therefore, if we are in Mode 6, t;te proposed wording would allow us to stay there.

SAFETY ANALYSIS JUSTIFYING PROPOSED CHANGE This change request provides a means of . insuring that the reactor is maintained in a more safe mode if it is already in that mode, therefore no unreviewed safety question is involved.

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.. j PLANT SYSTEMS-3/4.7.6 FLOOD PROTECTIO _N N LIMITING CONDITION FOR OPERATION 3.7.6.1 Flood protection shall be provided for all safety relate.

systems, components and structures when

a. The water level of the Gulf of Mexico exceeds 98 fet t Plant Datum, at the intake structure, and

b. A Hurricane Warning is in effect.

APPLICABILITY: At all times.

ACTION:

With the water level at the intake' structure above elevation 98 feet Plant Datum and with a Hurricane Warning in effect:

a. Close all watertight doors providing flood protection for safety related systems, components and structures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

b. Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; be in at i least COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. j SURVEILLANCE REQUIREMENTS 4.7.6.1.1 When a Hurricane Watch or Warning is in effect, the water level at the intake structure shall be determined to be within the limit by:

a. Measurement at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the water level is below elevation 92 feet Plant Datum,

b. Measurement at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the water level is equal to or above elevation 92 feet Plant Datum, and

c. Measurement at least once per 30 minutes when the water level 1

is equal to or above elevation 94 feet Plant Datum.

4.7.6.1.2 Meteorological fcrecasts shall be obtained from the National Hurricane Center in Miami, Florida, at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when'a Hurris Warning is in effect.

CRYSTAL RIVER . UNIT 3 3/4 7-19

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TECHNICAL SPECIFICATIuN CHANGE REQUEST NO. 2f

., Replace page'3/4 9-2 with the attached replacc. ment page 3/4 9-2.

REASONS FOR PROPOSED CHANGE During the insertion of the final fuel assembly during the Initial Fuel Load, it was difficult to obtain an ~ audible signal from the auxiliary source range neutron flux monitor while an audible signal could have been obtained from the source neutron flux monitors in the Control Center.

This change will allow us to obtain the audible signal required in the containment in one of three ways: 1) From an auxiliary source range neutron flux monitor located f a the containment; 2) From a source range neutron flux monitor and hardwired to the containment; or 3) From a source range neutron flux monitor and broadcasted over the public address system.

There is no Regulatory Guide requirement to use an Auxiliary Detector after Initial Fuel Loading, but being able to use an Auxiliary Detector or a source range monitor to supply the required audible indication would add operational flexibility and make for safer operation.

SAFETY ANALYSIS JUSTIFYING PROPOSED CHANGE This change request provides a means for readily obtaining an audible signal from a neutron flux monitor during insertion of every assembly, therefore no unreviewed safety question is involved.

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REFUELING OPERATIONS ,

INSTRUNENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, the following instrumentation shall be OPERABLE:

a. Two source range neutron flux monitors, each with visual indication in the control room and one.with audible indica-tion in the control room, and 3

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b. One auxiliary source range neutron flux monitor with audible indication in the containment or one source range neutron flux monitor with audible indication'in the containment.

APPLICABILITY: MODE 6.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9,2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and

b. A CHANNEL FUNCTIONAL TEST wit::in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial l start of CORE ALTERATIONS, and  !

c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE AL TERATIONS.

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