Text

Forwards Response to 841207 Request for Addl Info Re Proposed Amends to Licenses DPR-67 & NPF-16
	ML17215A885
Person / Time
Site:	Saint Lucie
Issue date:	05/23/1985
From:	Williams J; FLORIDA POWER & LIGHT CO.
To:	John Miller; Office of Nuclear Reactor Regulation
References
	L-85-199, NUDOCS 8505290150
	Download: ML17215A885 (42)
	v • d • e

REGULATORY~NFORMATION DISTRIBUTION S~ EM (R IDS)

.eI ACCESS30N NBR ~ 8505290150 DOC ~ DATE 85/05/23 NOTARIZED NO FACIL:50-335 St, Lucie PlantE Un) t 1< Florida Power 8 Light Co, 50 389 St ~ Lucie PlantE Unit 2E Florida Power 8 Light Co, AUTH ~ NAME AUTHOR AFFILIATION HILLIAMSiJ ~ N.

Flor ida Power 8 Light Co ~

RECIP ~ NAME RECIPIENT AFFILIATION MILLER'RR~

Operating Reactors Branch '3

SUBJECT:

Forwards response to 841207 request for =addi info re proposed amends to Licenses DPR 67 8

NPF 16

'ISTRIBUTION CODE:

ADDIO COPIES'ECEI)ED:LTR 1

ENCL J SIZE:

TITLE:

OR Submittal:

General Distribution NOTES; OL:02/01/76 OL ~ OLI/06/83" DOCKET 05000335 05000389 05000335 05000389 RECIPIENT ID CODE/NAME NRR OR83 BC 01 COPIES LTTR ENCL 7

7 RECIPIENT ID CODE/NAME NRR OR83 BC 01 COPIES LTTR ENCL' 7

INTERNAL: ACRS ELD/HDS2 NRR/DL DIR NRR/DL/TSRG NRR/D S I'/R A8 RGN2 09 6

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'ETB EG FIL 04 1

0 1

1 1

0 1

1.

1 EXTERNAL; EGLG BRUSKEiS NRC PDR 02 1

1 1

1 LPDR NSIC 03 05 1

1' TOTAL NUMBER OF COPIES REQUIRED:

LTTR 34 ENCL 31

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"OX 14000, JUNO BEACH, FL 33408 gyes Ilr FLORIDAPOWER & LIGHTCOMPANY MAY SS l985 L-85-I 99 Office of Nuclear Reactor Regulation Attention:

Mr. James R. Miller, Chief Operating Reactors Branch II3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Miller:

Re: St. Lucie Unit Nos.l 6 2 Docket Nos. 50-335 8 50-.389 Proposed License Amendments Additional information and Clarification Attached is Florida Power & Light Company's response to your December 7, l 984, request for additional information.

Very truly yours, J. W. Willia s, Jr.

Group Vice President

. Nuclear Energy JWW/RJS/cab Attachment 85052'70150 850523 PDR ADOCK 05000335' PDR Pea/

0 PEOPLE... SERVING PEOPL

ATTACHMENT ADDITIONALINFORMATIONAND CLARIFICATION ST. LUCIE UNIT NOS.

I &2 L-84-129 DATED MAY 15, 1984 L-84-130 DATED MAY21, 1984

I.

Unit I proposal:

Florida Power and Light (FPL) has proposed revising the definitions of Containment Vessel Integrity and Shield Building Integrity to be consistent with Unit 2. The proposal, however, failed to include the following in the definitions:

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

Unit 2 technical specifications do include the sealing mechanism in its definitions as does the Combustion Engineering Standard Technical Specifications (STS).

Therefore, due to the importance of the sealing mechanisms, the fact that these mechanisms are addressed in STS, and because Unit "I is being revised to reflect Unit 2, the licensee is requested to submit a supplement to their initial package to include sealing mechanisms in their definitions of Shield Building and Containment Vessel Integrity.

It is recommended the Unit I definitions read exactly like Unit 2 definitions I.7 and l.28.

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'I

': DEFINITIONS CONTAINMENT VESSEL INTEGRITY

/.7 CONTAINMENT VESSEL INTEGR!TY shall exist when:

~1 H

All containment vessel penetrations required to be closed during accident conditions are either:

j a".

Capable of being closed by an OPERABLE containment

~

automatic isolation valve system, or p O'.

Closed by manual valves, blind flanges-, or deactivated automatic valves secured in their closed position except as provided in Table 3.6-2 of Specification 3.6.3.1, 6.

~l A11 containment vessel equipment hatches are closed and

sealed, C

jg C'gynic/ic2ascs mA~ Ai'- <+~ i<>>~~ ~~

~F5 Each containment vessel airlock is.

Specification 3.6.1.3, ~

J.

+l&t-The containment leakage rates are within the limits of Specification 3.6.1.2, a~d c2Qoc'rcHdd &A E'NcA ppnekcahw (E'.g., td@ds CHANNEL CALIBRATION bye(~g A. g i,nqi );s j,c/

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter wi>>ch the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trio functions, and shall include the CHANNEL FUNCTIONAt, TEST.

The CHANNEL CALIBRATION may be performed by any series of sequen-tial, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK

~ 1.5le A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

ST; LUCIE - UNIT 1 1-2

DEFINITIONS STAGGERED TEST BASIS I.32 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

I NIEN N I OTNTION

l. l9 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.Z.

AXIAL SHAPE INDEX

/.2 The AXIAL SHAPE INDEX (Y ) is the power level detected by the lower excore nuclear instrumen( detectors (L) less the power level detected by the upper.excore nuclear instrument detectors (U) divided by the sum of these power levels.

The AAIAL SHAPE (HDEX (Y

) used for the trip and pretrip signals in the reactor protection systsIi is the above, va'lue (Y ) modified by an appropriate multiplier (A) and a constant (B) to dete ine the true core axial power distribution for that channel.

"E =

L+U L-U YI =AYE+B UNRODDED PLANAR RADIAL PEAKING FACTOR - Fx

/.37 The UNRODDED PLANAR RADIAL PEAKING FACTOR:.is the maximum ratio of the peak to average power density of the individual fuel rods in any of the.unrodded horizontal planes.

excluding tilt.

SHIELD BUILDING INTEGRITY I.Z8 SHIELD BUILDING INTEGRITY shall exist when:

a.

Each door is closed except when the access opening is being used for normal transit entry and exit', pntf 6.

ffl C~plsQetCS'N Pl The shield building ventilation system is ~ABBE:

CCsViCC6'~

3. 6. b. I ]

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LUCIE - UNIT 1 1-5 Amendment No.

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T NOPE44rJ'LE

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2.

Units I and 2 proposals:

3.4.8.d:

FPL deleted the reporting requirements from their technical specifications'with respect to Dose Equivalent l-l3l.

This deletion does not meet the intent of Generic Letter 83-43.

Technical staff discussions with Mr. F. Anderson, author of Generic Letter 83-43, have confirmed that a

Special Report in accordance with section 6.9.2 is required.

This Special Report must include the information required in current technical specifications, step 3.4.8.d.

Therefore, in order for the technical specifications to reflect the reporting requirements of Generic Letter 83-43, the licensee is requested to submit a supplement to their initial package to reflect the Special Report requirement.

This Special Report will be required within 30 days.

I 4

Cg 1

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITIHG CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a.

< 1.0 yCi/gram DOSE EQUIVALENT I-131, and b.

< 100/E pCi/gram.

APPLICABILITY:

MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1, 2 and 3*:

a ~

b.

c ~

With the specific activity of the primary coolant

> 1.0 pCi/gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months provided that operation under these circumstances shall not exceed 10 percent of the unit' total yearly operating time.

The provisions of Specification 3.0.4 are not applicable.

With the specific activity of the primary coolant

> 1.0 pCi/gram DOSE EQUIVALENT I-131 for more than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months during one con-tinuous time interval or exceeding the limii line shown on Figure 3.4-1, be in HOT STANDBY with T 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the specific activity of the primary coolant

> 100/E qCi/gram, be in HOT STANDBY with T v 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4 and 5:

d.

With the specific activity of the primary coolant

> 1.0 pCi/gram DOSE EQUIVALFNT I-131 or

> 100/E gCi/gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-.4 until the specific activity,of, th

@mary coolant is restored to witnin its limits.

A shall be prepared and submitted to the Commissionggll.suan%

to Specification 6.9.2,.

This report shall contain the results of the specific activity analyses together with the following information:

1.

Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was

exceeded, With T v

> 500'F.

ST, LUCIE - UNIT 1

3/4 4-17

REACTOR COOLANT SYSTEM ACTION:

Continued 2.

Fuel burnup by core region, 3.

Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was

exceeded, 4.

History of de-gassing operation, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, and 5.

The time duration when the specific activity of the pri-mary coolant exceeded 1.0 uCi/gram OOSE EQUIVALENT I-131.

SURYEILLANCE RE UIREMENTS 4.4.8 The speci,ic activity of the primary coolant shall be determined to be within the limits by performanc of the sampling and analysis pro-gram of Table 4.4-4.

,s ggtl~l C

ST.

LUCIE - UNIT 1

3/4 4-18

REACTOR COOLANT SYSTEM ACTION:

(Continued)

MOOES 1, 2, 3, 4, and 5:

d.

With the specific activity of the primay.y coolant greater than 1 microcurie/gram OOSE EQUIVALENT I-131 or greater than 100/E microcuries/gram, perform the sampling and analysis require-ments of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

A RER~Q= Sp8cIM hpM to Specification 6.9.2.

This report shall contain the results of the specific activity analyses together with the following information:

1.

Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, 2.

Fuel burnup by core region, 3.

Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, 4.

History of degassing operation, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, and 5.

The time duration when the specific activity of the primary coolant exceeded 1 microcurie/gram OOSE EQUIVALENT I-131.

SURVEILLANCE RE UIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

ST.

LUCIE - UNIT 2 3/4 4-26

3.

Vnit I and 2 proposal:

4.6.1.6:

FPL deleted all reporting requirements if any change of appearance or other abnormal degradation was detected in the containment vessel.

This does not meet the intent of Generic Letter 83Q3 since Containment Vessel Structural Integrity is not specifically addressed in 10 CFR 50.73.

In order for the reporting requirements of Generic Letter 83-43 to be met, the licensee is requested to submit a supplement to their initial package to reflect the fact that a Special Report pursuant to specification 6.9.2 shall be submitted to the Commission.

Based on technical staff discussions with Mr. F.

Anderson of NRR, this Special Report shall be submitted within l5 days.

FPL Res onse:

During a telephone conversation on February l4, l985, NRC stated that the reporting requirements are adequately addressed per IO CFR 50.72 and 50.73.

Therefore, the original submittal is correct, and no further revisions are necessary.

4.

Unit I and 2 proposals:

4.6.6.3:

FPL deleted all reporting requirements if any 'apparent changes in appearance of the concrete surfaces or other abnormal degradation was detected in the shield building. This does not meet the intent of Generic Letter 83-43 since Shield Building Structural Integrity is not specifically addressed in IO CFR 50.73.

In order for the reporting requirements of Generic Letter 83-43 to be met, the licensee is requested to submit a supplement to their initial package to reflect the fact that a Special Report pursuant to specification 6.9.2 shall be submitted to the Commission.

Based on technical staff discussions with Mr. F. Anderson of NRR, this Special Report shall be submitted within I 5 days.

~FP

~

During a telephone conversation on February 14, l 985, NRC stated that the reporting requirements are adequately addressed per IO CFR 50.72 and 50.73.

Therefore, the original submittal is correct, and no further revisions are necessary.

5.

Unit 2 proposal:

4.8.l.l.3:

FPL deleted the reporting requirement for all diesel generator failures, valid or non-valid.

The reporting requirement constituted a Special Report and therefore cannot be deleted from technical specifications due to the requirements of Generic Letter 83-43.

It is requested that the licensee submit a supplement to the initial package stating th'at all diesel generator failures, valid or non-valid, shall be reported to the Commission in a Special Report pursuant to specification 6.9.2 within 30 days.

The remainder of the technical specification should read the same as the current 4.8.I.I.3.

This change will result in consistency with other PWR changes currently being approved by the Commission.

In order for Unit I and Unit 2 technical specifications to be consistent, it is also recommended that diesel generator failures and Special Report requirements-be incorporated into Unit I technical specifications.

~FR It should be noted that the Commission approved the deletion of this reporting requirement in Amendment No. 47 to the North Anna - Unit 2 Technical Specifications.

Therefore, we feel that our original submittal is correct, and no additional changes are necessary.

I

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~ 6.

Units I

and 2

proposals:

6.2.2.f:

FPL proposed that administrative procedures shall be developed and implemented to limit the working hours of senior reactor operators and reactor operators.

The NRC had, in the past, allowed the licensee to have an overtime restriction policy for licensed operators and senior operators (Reference NRC letter to FPL dated February 4,

l982 and Unit 2 Safety Evaluation Report Section l3.5.I.I (l)f. dated October l98I). Since these documents were generated, the NRC has further clarified its position with respect to meeting the overtime requirements of NUREG-0737 Action Item I.l.l.3 entitled, "Shift Manning".

Specifically, Generic Letter 82-02, dated February 8, l 982, stated that overtime controls shall apply to the plant staff who perform safety-related functions (e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel).

Subsequently, Generic Letter 82-l0 dated May 5, l 982 was issued.

The enclosure to this Generic Letter stated the following requirement for action item I.A.I.3.I; "Revise administrative procedures to limit overtime in accordance with NRC Policy Statement issued by Generic Letter No. 82-02, dated February 8, I 982." FPL's response to Generic Letter 82-IO was FPL letter L-82-272 dated July 2, l982.

In this letter, FPL continued to refer to Unit 2 Safety Evaluation report, Section l3.5.l. I(l)fand NRC's letter of February 4, I 982, as adequate justification to only limit working hours to licensed operators and senior operators.

Finally, the NRC further clarified its position on shift manning in Generic Letter 82-l2 dated June l5, l982.

Again, this generic letter stated that overtime controls shall apply to plant staff who perform safety related functions (e.g.,

senior reactor operators, reactor operators, auxiliary operators, health physicists, and key maintenance personnel).

FPL responded to Generic Letter 82-.I2 in FPL's letter L-82-4I7 dated September 30, l982.

FPL maintained the same position as previously stated.

On March l4 l983, the Commission issued an Order confirming FPL commitments to implement post-TMI related items.

Attachment'2 to the Order stated the requirement that the licensee revise administrative procedures to limit overtime in accordance with NRC Policy Statement issued by Generic Letter No. 82-I2, dated June l5, l982.

The licensee's status was listed as complete even through 82-l2 was not met.

It has also been noted that Unit 2 technical specification 6.2.2.f states "Adminstrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor oeprators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel."

Contrary to the

above, St.

Lucie Administrative Operating Procedure 00IOI I9, Revision 6, only applies over-time limitations for licensed oeprators.

Therefore, it is recommended the licensee submit a supplement to their initial package changing Unit I proposal Section 6.2.2.f to read exactly like Unit 2 cu'rrent Section 6.2.2.2f. Also, it is recommended that the licensee delete its proposal to change Unit 2 Section 6.2.2.f since it currently fulfillsall NRC requirements.

~

6. 0 AOHINI STRATI VE CONTROLS 6.1 RESPONSIBILITY gn,'f 6.1.1 The Plant Manager shall be responsible for overall @acidly operation and shall delegate in writing the succession to this respon-sibility during his absence.

(poIe g, TI>< ShtA'dodrvisA') 07 ddr<ncf 4e's Chosen~ ~ Wc. ctwkrol reer<, a d siqncu4d indtvidPP, shoQ Le resp' 6)e kr Sic con6et rcew gom~a J Ponc4eon.

Rmanadiee~eM) derec4vi A A<'S ePeck S'qned 6y ge Ie-(Q PiÃSidenfii<C4ae I

OFFSITE Perak4 s st 4e re:ssoed W uQ s'A&i~ Pc~sdnng dn ~

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6aki$ ~

on< t.

6.2.1 The offsite organization for @~I-i'anagement and technical support shall be as shown ~ Figure 6.2-1.

in gdrT HAEC~ STAFF Vnig 6.2.2 The ~H44y organization shall be as shown yrf Figure 6.2-2 and:

'a

~

b.

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

&ado At least one licensedh0perator shall be in the control room when fuel is in the reactor.

2~ ~dCe'A'm, zh;Ie W smack is in NODE i.Z, 8,ee 9, af leos'6 gag /e(enSed ~e,'~ geac+ 4 er~ sLtC bee'n /lie N>kol rani c-'hot CHys".

A Alai% i'hylic,5 +the>>ck~

'en-proteeH~meed~

shall be on site when fuel is in the reactor.

J. g..

11 CORE ALTERATIONS

'hall be ohs~veJ

~C supervised'by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

il'e &<,'n chary~ eF Eve'( Lrndi(e'I n=. inukv. gu~rv'sos'c~ 4c co.~4~pf ro~ a~d has Hr x(eqig~'I Q + ditc.'egg saPc,'<vise aje,'der etc. re'4e(iltci deck nr c ipenk. +~lpoor.

g. P'.

A Pire Brigade of at least~gmembers shall be maintained onsite at all times9 The Fire Brigade shall not include~mR~hcshe<~~W

',,Pc'nP the minimum shift crew necessary for safe shutdown of the tt<~

unit graf any'personnel,required for other essential functions

)he~ 'uring a fire emergency.

Ad~l~'>4ek'Ve p.O<ed~eS ShalL h

dWa4Ped g~s) i~p(gd~e~deJ H I'~'f %e uS'kin9 h0~

gP un<f i4fP who ppcfyrm gaga/- <II'rncfems,'.g.,

S6>>'A'~ac~e

~pay'e~, veacW OpeMacr hgQW p/Yfip/$/s QtjQIJ Qrf per~$

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ST.

LUCIE - UNIT 1

6-1

'Amendment No. g gp'le. Ace+,ohms(G< kchaii)'gg onoI rl~te 0<'iiaJ< <~Ps'>'~ ~+ ~~ I<>< +"+"~'""

re<~;se~dgr ~ a. pereog c4'Anne nof Fo c<ceed 3 h~rs, 4 ~der & acce~ndaA s'~~~p"cled qd pnce rgv,fed )AN'ledh'a4 @dr'~

<s Alkyd A) fi'tl ~ IE'Wired o4tV<'&Is.

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ADMINISTRATIVE CONTROLS

6. 1 RESPONSIBILITV

6. 1. 1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6. 1.2 The Shift Supervisor, or during his absence from the control

room, a

designated individual, shall be responsible for the control room command function.

A management directive to this effect, signed by the Vice President Nuclear ~~ shall be reissued to all station personnel on an annual basis.

Opevak'yn>

6. 2 ORGANIZATION OFF SITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6. 2-1.

UNIT STAFF

6. 2. 2 The unit organization shall be as shown in Figure 6. 2-2 and:

a 4 b.

C.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6. 2-. l.

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

In addition, while the reactor is in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator shall be i'n the control room.

A health-physics technician shall be on site when fuel is in the

¹ reactor.

d.

All CORE ALTERATIONS shall be observed by a licensed operator and supervised by either a licensed Senior Reacto~ Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

The SRO in charge of fuel

'andling normally supervises from the control room and has the flexi-bility to directly supervise at either the refueling deck or the spent fuel pool.

e.

A site Fire Brigade of at least five members shall be maintained onsite at all times.

The Fire Brigade shall not include the Shift

¹ Supervisor, the STA, nor the two other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions;

e. g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

The health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected

absence, provided immediate action is taken to fill the required positions.

ST.-

LUCIE - UNIT 2 6-1

-7.

Units I and 2 proposals:

6.4.2: FPL proposed in their letter to the NRC, L-84-154, dated June 19, 1984, that the Fire Brigade training requirements be deleted from technical specifications since 10 CFR Part 50, Appendix R, Section III.I adequately specifies the requirements for Fire Brigade training.

The technical staff concurs with this proposal, and it is recommended the licensee submit a supplement to their initial package deleting Section 6.4.2.

~FPLtl:

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ADMINISTRATIVE CONTROLS

6. 3 WABBHR'TAFF UALIFICATIONS UNit 6.3.1 Each member of the

'taJ'f shpll me o,egceed tpe mini~mi,~is~<<d>'

the

~ who shall meet or exceed the qualifications of Regulatory Guide 1.8., September'975, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and ~

N transients and accidents.

+8@'a2+ng

+Qgglcfeki5 fi'gf jng/Jd<"nq 6.4 TRAINING un;+

6.4.1 A retraining and replacement training program for the faeHH+y staff shall be maintained under the direction of the Training Supervisor and shall meet

.oj e,

ed the requirements and recomendations of Section 5.5 qf ANSI/

and Appendix "A" of 10 CFR Part 55P, ~We s~~/e~~ >~i<<~~~~

pPeaa'Red&

Si,f;ma f,'n d C a+ ~a'!~spec'of

/Ae &a~ch 28, t980 jQRC /~gr*~ /;CWfddS OAd yak jell~

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gyp8A8+tCa-'.5 REVIEM AND AUDIT

6. 5.1 FACILITY REVIEW GROUP FRG FUNCTION 6.5.1.1 The Facility Review Group shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSITION Plant Manager Operations Superintendent Operations Supervisor Maintenance Superintendent Instrument 8 Control Supervisor Reactor Supervisor Health Physics Supervisor Technical Supervisor Chemistry Supervisor equality Control Supervisor Assistant Plant Supt.

Mechanical Assistant Plant Supt. Electrical

]4 g,~eiaa(A o~ Pie

/-RQ And 44ll rman shall/be designated in writing.

Member:

The ~ Chai 6.5.1.2 The Facility Review Group shall be composed of the:

ST.

LUCIE - UNIT 1

6-5 Amendment No.

~

'C

'I

ADMINISTRATIVE CONTROLS

6. 3 UNIT STAFF UALIFICATIONS (Continued) or exceed the qualifications of Regulatory Guide l. 8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and plant operating characteristics, including transients and accidents.

6. 4 TRAINING 6.4.

1 A retraining and replacement training program for the unit staff shall'e maintained under the direction of the Training Supervisor and shall meet or exceed the requii ements and recommendations of Section

5. 5 of ANSI/ANS 3. 1-1978 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant induptry operational

'experience.

arteA~~Hf+re-Brigade

6. 5 REVIEW AND AUDIT

6. 5.

1 FACILITY REVIEW GROUP FRG FUNCTION 6.5. l. 1 The Facility Review Group shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSITION

6. 5. 1. 2 The Member:

Member:

plc~ber Member:

Member:

Facility Review Group shall be composed of the:

Plant Manager Operations Superintendent Operations Supervisor Tash.v <~.i" ~ ~~'I Sdpgrg.'s~

Reactor Supervi sor Health Physics Supervisor

~;>>. Technical Supervisor guality Control Supervisor Assistantgfupt.

Electrical f)ssisfa Pio& wp~. N~chanic,aL The Chairman shall be a member of the FRG and shall be designated in writing.

ALTERNATES 6.5. 1.3 All alternate members shall be appointed in writing by the FRG Chairman to serve on a temporary basis;

however, no more than two alternates shall participate as voting members in FRG activities at any one time.

ST.

LUCIE - UNIT 2 6-7

'8.

Unit I proposal:

6.5.l.7.a:

This section referred to paragraph "m" which was proposed to be deleted in Section 6.5.l.6.

This appears to be an administrative oversight.

lt is recommended that FPL submit a supplement to their initial package deleting the reference to paragraph "m" in Section 6.5.I.6.

f

ADMINISTRATIVE CONTROLS e.

Investigation of all violations of the Technical Specifications~

including the preparation and forwarding of reports covdering evalua-ri,

<<.~l:<<P d*

V a',d' g,)e@~

e~ces-and to the Chairman of the Company Nuclear Review Board.

aA

&'i=dpi Sm 0

A vn'+

na~/~

g.

Review of -$~~y operations to detect potentialgsafety hazards.

~a~Qys~~

h.

Performance of special r~ei~ews,+a4 investigationsdiand r eports thereon as requested by tlute<

the Company Nuclear Review Board.

i.

Review of the ~net. Security Plan and implementing procedures and ~>>~"~~

Nuclear Review Board.

J

~

k.

Revieu of the Emergency Plan and implementing pr ocedur es and

~~s'<~'nrit'ecommended changes to the

'ompany Hucl ear Review Board.

On-S i4e Review of every unplanned -ons44m release of radioactive material to the environs including the preparation of r epor s covering evaluation, recommendations and disposition of the corrective action to prevent recurr ence and the for warding of these reports to the Vice President Nuclear Energy and to the Company Nuclear Review Board.

~~I~S Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL and RADWASTE TREATMENT SYSTEMS.

AUTHORITY 6.5.1.7 a.

The Facility Review Group shall:

Ih vDia ~a+

Recommendgto the Plant Manager, wri "~ approval or disapproval of items considered under~6.5.1.6fa) through fdj above.

~i'PicdiW 3 Render determinatiogs.>jn.gvqiting with regard to whether or not each item considered undFr~5.%1.6(a) through $e) above constitutes an unrev i ewed safety question.

c.. Prgvi)e,,yr t$qn np)ification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice Presi" nt p/~v/ear vga a,"td

~

Sd the Company Nuclear Review Board of disagree-ment between the FRG and the PlantManager;

however, the Plant Manage shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

W dA'cod'~

ST.

LUCIE - UNIT 1

6-7 Amendment No.

Unit l and 2.proposals:

6.7.l.c:

FPL deleted the period of time a Safety Limit Violation Report must be submitted to the Commission.

lt is recommended the licensee submit a

supplement to their initial change package for Section 6.7.l.c to read as follows: "The Safety Limit Violation Report shall be submitted to the Commission, the CNRB, the Vice President Nuclear Operations, and the Group Vice President Nuclear Energy within l4 days of the violation."

FF~F:

A h

d h

i d

~

s AOflINISTRATIVE CONTROLS 6.7 SAFETY LIfdIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated'.

d. Y.

e.z pic, g e~'g s g~p s~ ~ ~g],'peJ $ y feteplonc as s'cen as pstssg'l(c and r'nake E.'d)es v. ~'"

.The-4

, she Vice President and ~ the CHRBwithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

sJJcfgoy'p res phd hc noh'Pi'ed'afety Limit Violation Report shall be prepared.

The repor shall be reviewed by the FRG.

This repor t 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.

$e submigte~

to,@he...,p,,g<~

C d.

The Safety Limit Violation Repair t sh 1

Commission, the CNRB,-end the I"~days of the violation.

d.

~:4i'ca<

~pe>red<'8 ofAe 6I~:1'Iau no/ Se /ese.~ed u lail Es6skA'ivy stY A ~~<~<l'~.

6.8 PROCEDURES 6.8.1 I'ritten procedures shall be established, implemented and maintained covering the activities referenced below:

a'.

The applicable procedures recommended in Appendix "A" of Regulatory

.&.'>>'-g 0

dA

":~."dJ'A1plg~s'~in/

'f>e-4ego 6 BmAgs pP gg@I=q b.

Refuel ing opera tions.

c.

Surveillance and test activities of safety related equipment.

d.

Securi ty Plan impl ementation.

e.

Emergency Plan ialplementation.

f.

Fire Protection Program implementation.

g.

PROCESS CONTROL PROGRAM implementation.

h.

OFFSITE 00SE CALCULATION MANUAL impl ementation.

F guality Control Program for effluent monitoring, using the guidance in Regulatory Guide 1.21, Revision 1, June 1974.

j.

qua/ity Control Program for environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975.

6.8.2 ach procedure of Specification 6.8.la through i. above, and changes

thereto, shall be reviewed by the FRG and shall be approved by the Plant tianager prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

ST.

LUCIE - UNIT 1

6-13 Amendment No.

yg ~

'l h-W'

ZvzÃrS EE LE EEEEEEEE T

peF4 4i T>>

I I

II LE~

a.

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

.. Sr~4'~

ycA73 go /0 @FR poof',D Fsf'ehJ f b.

Each REPORTABLE shall be reviewed by the FRY and the results of this review<submitted to the CNRB, and-the Vice President Nuclear ~~>j'"~><~z 4>J ge.

@dOOP V'i'CC PCeS'de AgCleCe E~)e~y-

6. 7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The Vice President Nuclear

~~y and the CNRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Q'yt 6essS A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the FRG.

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.

c.

The Safety Limit Vio'lation RePort shall be submitted

$o the o~.idafaEEsf

~J~sP-uEE Commission, the CNRB,+ad the Vice President Nuclear &4Ngy<'within 14 days o> the violation.

d.

Critical operation of the unit shall not be resumed until authorized by the Commission.

6.8 PROCEDURES ANO PROGRAMS 6.8.

1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a.

b.

C.

d.

e.

The applicable procedures-recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February

1978, and those required for implementing tt e requirements of NUREG 0737.

Refueling operations..

Surveillance and test activities of safety-related equipment.

Secur'ity Plan implementation.

Emer gency Plan implementation.

ST.

LUCIE - UNIT 2 6-13

IO. Unit I proposal

6. I0.2.f:

This requirement referred to table 5.7-l.

This appears to be an administrative oversight in that the correct table is actually 5.9-I.

It is recommended the licensee submit a supplement to their initial package with the appropriate correction made.

~FPL~:

A di

~

4>>

1 '1

))

AO)iINISTRATIVE CONTROLS b.

C.

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

<"cucM Qs gs ~J ~eeI~Ws.

Records of

~

d.

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

e.

f.

Records of gaseous and liquid radioactive material relea'sed to the environs.

gn V.

Records of transient or operational cycles for those~~y-components iden tified in Tabl e 5. 9-1.

Records of training and qualification for current members of the

-@4M staff.

Udi+

Records of in-service inspections performed pursuant to these Techni cal Speci fica tions.

Records of Quality Assurance activities required by the QA Manual.

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

k.

Records of meetings of the FRG and the CNRB.

4-i-@earn-whi-eh-are-e

/.

Pf.

m ~

n.

Records. of the service lives of all s nubbers listed Sh Tables 3.7.-2a and 3.7-2b inc'luding the date at which the service life commences and associated installation and maintenance

'records.

ggg&ds of $8cenddt Nc4p Rtmplinq onJ Mc4A cfod4~ /y, Annual Radiological EnvironmentaT Operating Reports; and records of analyses= transmitted to the 'licensee which are used to prepare the Annual Radiological Environmental Monitoring Report.-

o.

Meteorological data, summarized and reported in a format consistent with the recommendation of Regulatory Guides 1.21 and 1.23.

p

&cordf oI d~d 6 poelA~d gqd~ ge ggqgbv~gk gf Specrfieugc~$ 6.~.8.8 coed 6.8.'I 6.11 RAOIATION PROTECTION PROGRAH-Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

ST.

LUCIE -"'NIT 1 6-20 Amendment No. A, -

V o

c~

< I I. Unit I proposal:

6.I4.!:

It appears that an administrative oversight was made in that the proposal stated that the FRG approves the ODCM.

In fact, the Commission approves the ODCM.

It is recommended a supplement be submitted to make the appropriate correction.

~FPLR:

A h

i h

t I

~

g7

AOMINISTRATIVE CONTROLS l9 6.

OFFSITE OOSE CALCULATION MANUAL 00CM 6.+.1 The OOCM shall be approved by the Commission prior to implementation.

6.@Z Licensee initiated changes to the 00CM:

l.

Shall be submitted to the Comnission in the Semiannual. Radioactive Effluent Release Report for the period in which the change(s) was made effective.

This submir tal s hal 1 conta in:

a.

Sufficiently detailed information to totally support the y.

~

rationale for the change without benefit of additional or 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 analyses or evaluations justifying the change(s);

b.

A determination that the change will not reduce the accura'cy or reliability of dose calculations or setpoint determina-tions; and c.

Oocumentation of the fact that the change has been reviewed and found acceptable by the FRG.

2.

Shall become effective upon review and acceptance by the FRG.

ST.

LUCIE - UNIT 1

6-23 Amendment No. 5

C f

k' g ~

p

~

W 4

L-85-199, Forwards Response to 841207 Request for Addl Info Re Proposed Amends to Licenses DPR-67 & NPF-16

Text

SUBJECT:

Dear Mr. Miller:

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L-85-199, Forwards Response to 841207 Request for Addl Info Re Proposed Amends to Licenses DPR-67 & NPF-16

Text

SUBJECT:

Dear Mr. Miller:

Navigation menu

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