ML20247B585
| ML20247B585 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 07/14/1989 |
| From: | Frantz S FLORIDA POWER & LIGHT CO., NEWMAN & HOLTZINGER |
| To: | Bright G, Cotter B, Harbour J Atomic Safety and Licensing Board Panel |
| References | |
| CON-#389-8919, RTR-NUREG-0452, RTR-NUREG-452 OLA-4, NUDOCS 8907240155 | |
| Download: ML20247B585 (36) | |
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' Jerry Harbour Atomic Safety and Licensing Board. Panel' U.S... Nuclear Regulatory. Commission
~ Washington, D.C.;
20555 Re Docket Nos. 50-250 OLA - 4 and 50-251 OLA - 4; 2In the Matter of Florida Power &' Light Company
-(Turkey Point Plant, Units 3 and 4)
Dear Licensing Board Members:
The purpose'of this letter is-to inform the Licensing Board and the parties of a matter which relates to,- but does not of, the Turkey Point affect the acceptability /T) Limits.
Pressure / Temperature (P On June 5, 1989, Florida Power & Light Company (FPL) submitted an application to amend the operating licenses for Turkey Point.
This amendment would replace the entire set of current technical specifications (including the P/T limits which are the subject of this proceeding) with revised technical specifications.
In general, the revised technical specifications conform with the Standard Technical Specifications for Westinghouse nuclear power plants issued by the NRC in NUREG-4' 0452.
As a result, the revised technical specifications represent a significant improvement in the' format, content, and understandability.of the current technical specifications for Turkey-Point, and are being submitted for NRC' approval as part of the, Turkey Point Performance Enhancement Project (PEP).
The P/T limits in the revised technical specifications
.are substantively identical to the P/T limits that are the subject of this proceeding.
The primary differences between the 8907240155 890714 yp3 s
y NEWMAN & MOLTZINOER, P. C-current and revised technical specifications with respect to the-P/T limits are as follows:
Editorial changes and Changes in Format - For-example, in the revised technical specifications, surveillance requirements are included in the same section as the limiting. conditions for operating (LCOs); in the current technical specifications, the LCOs and the surveillance requirements are in separate sections.
Ernansion of the Description of the Bases for the P/@ Limits - For example, the bases for.the revised technical specifications include an expanded discussion of the provisions of the ASME Code applicable to calculation of'the P/T limits.
Inclusion of Additional Surveillance Requirements - In particular, the revised technical specifications include a new requirement for determining the reactor coolant system temperature and pressure once every thirty minutes during system hentup, cooldown,.and inservice leak and hydrostatic testing operations.
Additionally, the description of the reactor vessel material surveillance program is more complete in the revised technical specifications in zhat, for example, it addresses capsules for shell forgings as well as the capsules for the weld material.
Deletion of an Obsolere Figure - Figure 3.1-2,
" Radiation Induced Increase in Transition Temperature for A302-B Steel," is not referenced in the current technical specifications.and does-not impose any limits-on operation.
Therefore, this figure has been deleted.
None of these differences affects the acceptability of the P/T limits in the current technical specifications.
The application for the revised technical specifications is voluminous -- approximately four inches thick.
Therefore, for the convenience of the Licensing Board and the parties, I am enclosing only those portions of the application that relate to the P/T limits.
The current technical l
specifications will remain effective pending review end approval of the application for the revised technical specifications by the NRC staff.
Sinc,erely,
).flPA Steven P. Frantz l
cc Service List I
e P O Boa 14000 Juno Beacn.FL 33408 0420 c..
FPL JUfE 5 1989 L-89-201 10 CFR 50.90 U..S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.
C.
20555 Gentlemen:
Re:
Turkey Point Units 3 and 4 1
Docket Nos. 50-250 and 50-251 Proposed License Amendment Revised Technical Specifications In accordance with 10 CFR 50.90, Florida Power & Light Company (FPL) requests that Appendix A of Facility Operating Licenses DPR-31 and DPR 41 be amended to replace the current technical specifications with the Revised Technical Specifications..
The Revised Technical Specifications were originally submitted by FFL letters, L-86-393, dated September 29, 1986 and, L-86-475, dated November 28, 1986.
Following NRC Staff review and FPL comment, this Final Draft of the Revised Technical Specifications was issued bf the NRC in letters dated March 14, 1989 and May 12, 1989.
Attachment I provides the proposed Technical Specifications.
This document is a markup of the Final Draft to indicate minor revisions that have been identified since that document was issued.
These revisions have been discussed with the NRC Staff.
Ec certify that, to the best of our knowledge, the attached specifications are consistent with the updated FSAR and with the as-built plant, subject to the following clarifications:
1.
Our review to date has identified several places where the FSAR will have to be revised to be consistent with the RTS.
These items are listed in Attachment III.
Because the l
probable implementation period for the RTS nearly coincides with the January 1990 cutoff date for the next issuance of the FSAR update, we would expect to submit these changes in the 1990 update to the FSAR.
l I
1 en fPL Group compo,
_________D
U. S. Nuclear Regulatory Commission L-89-201-Page two We have not questioned the validity of any current technical 2.
which were transferred to the' RTS.
The specifications setpoints in the current technical specifications were also assumed to be correct.
'3.
Our SSS vendor has reviewed the Limiting Conditions for Operation (LCO) and Bases for the specifications in their responsibility and has indicated that they are scope of Our certification process has taken credit for this correct.
assurance from the NSSS vendor.
of the specifications. use the Standard Technical Most 4.
Specification (STS) wording, LCO times, action statements and surveillance requirements and intervals.
As discussed in the we have no significant hazards evaluation for these items, adopted these STS requirements where we considered Turkey Point to be similar to the plant design which forms the basis for the STS.
STS action times and surveillance intervals cannot be quantitatively justified for Turkey Point with our current level of knowledge of risk contributions.
certification applies to both Units 3 and 4.
Our This 5.
technical specification manuals will continue to be issued in a combined form for ease of'use by the oparations staff.
Our reviews have also indicated that, to the best of our knowledge,
- we have accounted for each specification from the current technical specification to the RTS by:
direct transfer, except for format or editorial changes, 1.
2.
substitution of an appropriate STS, or submittal of justification for other changes or deletions.
3.
II provides the overall safety evaluation and the no Attachment significant hazards evaluations which are specific to the changes from the current technical specifications.
Technical Specifications 3/4.5.2 and 3/4.6.2 added specific residual heat Technical removal and containment spray pump flow requirements.
Specification 3/4.2.5 provides a more restrictive pressurizer pressure DNB limit.
Evaluations performed to certify these parameters resulted in large break LOCA peak clad temperature (PCT) penalties of 7'F and 8'F, respectively.
The current PCT value is now 2144*F, which is below the ' 2200*F PCT limit.
Per the requirements of 10 CFR 50.46, these PCT penalties will be documented in an annual report to the NRC.
1 Attachment IV to this letter provides a listing of open licensing issues that have been tied to this project.
These items can now be closed as a result of this submittal.
l l
w U. S. Nuclear. Regulatory Commission L-89-201 Page three successful implementation'of.the RTS is a key-part of the upgrade We would like to meet with your Staff prior to issuance-effort..
of this license amendment to discuss the implementation method and As previously discussed with the
- Staff, this schedule., ion effort could take six months, or longer.
implementat The discussion of the jnterim nature of the Electrical Power 1989 letter is fully Systems Specifications in the NRC's May 12,The technical-specification changes wh understood by FPL.
plan to submit to support the Emergency Power Enhancement Program will take the form of a markup of the Proof and Review version of the S'fS.
In accordance with 10 CFR 50.91(b) (1), a copy of this proposed license amendment is being forwarded to the State Designee for the State of Florida.
The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the FPL Company Nuclear Review Board.
Should there be any questions on this request, please contact us.
Very truly yours, b
C. O. W Acting ior Vice President - Nuclear COW /PLP/gp Attachments Stewart D. Ebneter, Regional Administrator, Region II, USNRC cc:
Senior Resident-Inspector, USNRC, Turkey Point Plant Mr. Jacob Daniel Nash, Florida Department of Health and Rehabilitative Services i
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STATE OF FI4RIDA
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- c. O. Woody being first' duly sworn, deposes and says:
That he is Actina Senior Vice President - Nuc1 car, of Florida Power ard Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the bast of his knowledge,.information and belief, and that he is authorized to execute the document on behalf of said Licensee.
/
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C. O. Wo Subscribed and sworn to before me this day of TAE
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NOTARY PUBLIC, in and for the County of Dade, State of Florida Dictny Psik. State of bula Ny Commi::en Expes W.:y 30,19M was n la fa% ~a =
My Commission expires l
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l TURKEY POINT UNITS 3 & 4
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REVISED TECHNICAL SPECIFICATIONS AND ASSOCIATED NO SIGNIFICANT HAIARDS EVALUATIONS
.JPN-PTN-SENJ-89-027 REVISION O n
Prepared by: 48#v Date W- #f M. P. Huba, Engineer Date d-$-X9 Reviewed by:.__ N I<0 B. P. Burdick, Supv.
- /F/ff fdI4 M Date Approved by: W. A. Skelley,/ Manager
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J. B. Hosmer, Director-Nuclear Engineering Issue Date: April 3,1989 e
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_JPN-PTN-SENJ-89-027:
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b Page 2'of 4-F ;,
9 NUCLEAR SAFETY RELATED.
4 /,' 2 i.
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' Verified by:
4 L I. l'd>^/ -
Date-
^
W. E. Harvey.
U JPN Interfaces l
.:Non-Lead-L +:
lei lf.q ind Initials /Date
- Discioline U.NA - ' Alii -
X Mechanical.
Mtf "
'X_
Electrical MN4do #Ide Instrumentation &~ Control X-k 1
X' Civil
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Nuclear.
' External' Interfaces
~ No.' External Interfaces Quality Assurance Nuclear. Fuel-General'. Engineering Security Nuclear Plant Nuclear Mutual Limited-(NHL).
Westinghouse:
X 1.
Letter No. FPL-89-619, dated March 15, 1989 2.
Letter No. FPL-89 635, dated April 3, 1989 S"
Nuclear Energy (Chapter 6.0 and associated NSHEs) P. L. Pace Y
f Date
- Signoff does not include Chapter 6.0 review.
- j
_....__,__.._______.__.._____.__U
JPN-PTN-SENJ-89-027 Revision 0 Page 3 of 4 ABSTRACT j
The purpose of this safety evaluation is to document JPN concurrence with Turkey Point Units 3 & 4 Revised Technical Specifications (RTS) and assoc The attached RTS Safety Evaluation /No Significant Hazards Evaluation (NSHEs).
exceptions noted within, reflect the results of FPL/NRC negotiations.
reflect Extensive input and review was provided by Westinghouse, and '.s docum letter reference on the Interface page.
and associated NSHEs were prepared and reviewed by Nuclear Energy - Licen NE signoff on the Interface page takes the place of Discipline signoff for Sectidn 3/4.8, Chapter 6.0, as this chapter was excluded from Discipline review.
is not included as the content is still being Electrical Power Systems,This section will be reviewed in the future.
negotiated with the NRC.
DNB New values have recently been calculated for the following parameters:
1 parameters RTS 3.2.5; RHR pump surveillance, RTS 4.5.2.b; and Containm pump surveillance, RTS 4.6.2.1.b.
These values have been calculated by However, the formal Westinghouse and transmitted for inclusion in this document.
transmittal of the final signed calculations / evaluations ha received.
a Proposed License Amendment until receipt is transmitted to the NRC as documented by letter to the Nuclear Energy Department.
The List of Effective Pages lists the pages of the RTS/SE/NSHEs. These docume If subsequent revisions are have no revision block or other identifier.
necessary, the affected pages will be sidebarred and a " List of Pages Affecte Revisions will be made for all technical by Revision X" will be attached. changes. Editorial / typographical cor review / revision.
Attachments:
Revised Technical Specifications 1.
Safety Evaluation /No Significant Hazards Evaluations RIN.ABSCT-TS
_ _ " - ~ - - - - - - - - _ _ _ _ _ _ _ _
ATTACHMENT I Turkey Point Units 3 and 4 Re:
Docket Nos. 50-250 and 50-251 Proposed License Amendment Revised Technical Specifications d
MARKUP OF FINAL DRAFT OF TECHNICAL SPECIFICATIONS I
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. REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM
. LIMITING CONDITION FOR OPERATION The ' Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit l 3.4.9.1 leak and hydrostatic testing w'ith:
A maximum heatup of 100*F in any 1-hour period, a.
A_ maximum cooldown of 100 F in any 1-hour period, and b.
A maximum temperature change of less than or equal to 5'F in any 1-hour period during inservice hydrostatic and leak testing opera-c.
tions above the heate and cooldown limit curves.
APPLICABILITY:
At all times.
ACTION:
d/or pressure
.With atty of the above limits exceeded,. restore the temperature an to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; deteraine that the Reactor Coolant Syst acceptable for continued operation or be in at least HOT STANDBY w and pressure to less than 200 F and 500 psig, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS Tavg respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during syste 4.4.9.1.1 heatup, cooldown, and inservice leak and hydrostatic testing operations.
The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material proper 4.4.9.1.2 1 accordance with the schedule in required by 10 CFR Part 50, Appendix H i:The results of these examinatio Table 4.4-5.
Figures 3.4-2, 3.4-3 and 3.4-4.
f AMENDMENT N05.
AND 3/4 4-30 TURKEY POINT - UNITS 3 & 4 MA( 05 M
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AND FEB 2 8198!
MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: CIRCUMTEREN11AL %TLD 10*f INITIAL RTNDT:
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AND TURKEY POINT UNITS 3 & 4 3/4 4-33 FEB 2 81989
MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: CIRCUMfERENTIAL WELO 10'T RTNDT *1/4 THICKNESS = 252.5'T INITIAL RTNDT :
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REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL' SCHEDUL UNIT'3 a
CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME U
30 0.49' Standby Specimen withdrawn-
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.0.34 Standby X-50 0.34
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150 0.49 Standby Z
230*
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AND t
TURKEY POINT - UNITS 3 & 4 3/4 4-34 MAY 0 5153 l
REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION The pressurizer temperature shall'be limited to:
3.4.9,2 A maXinum heatup of 100'F in any 1-hour period, a.
A maximum cooldown of 200'F in any 1-hour period, and b.'
A maximum spray water temperature differential of 320*F.
l c.
APPLICABILITY: At'all times.
I ACTION:
With the pressurizer temperature limits in excess of any of the above limits, r
l restore the temperature to within the limits within 30 minutes; perform an l
engineering evaluation to determine the effects of the out-of-limit con on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable.for continued operation or be i in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
i i
SURVEILLANCE REQUIREMENTS I
The pressurizer temperatures shall be determined to be within the The limits at least once per 30 minutes during system heatup or cooldown.
4.4 9.2 spray water temperature differential shall be determined.to be within the j
limit 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 auxiliary spray operation.
i j
l I
l AMEN 0 MENT N05.
AND 3/4 4-35 TURKEY POINT > UNITS 3 & 4 FEBt81ng 1
d
.. r OVERPRESSURE MITIGATING SYSTEMS LIMITING' CONDITION FOR' OPERATION i
3.4.9.3 The high pressure safety injection flow paths to the Reactor Coolant
~
(
n System (RCS) shall be isolated, and below an RCS average coolant temperature l
of 275'F at least one of the following Overpressure Mitigating Systems shall be OPERABLE:
a)
Two power-operated relief' valves (PORVs) with a lift setting of 415 1 15 psig, or b)
The.RCS depressurized with a RCS vent of greater than or equal to y
F 2.20 square inches.
APPLICABILITY: MODES 4, S' and 6 with.the reactor vessel head on.
AETION:
With the high pressure safety injection flow paths to the RCS
' a.
unisolated, restore isolation of these flow paths within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and-In MODE 4 with RCS average coolant temperature iess than or equal to b.
275'F, and in MODE' S or in Mode 6 with the reactor vessel head on:
With one PORV inoperable, perform at least one of the following 1.
within the next 7 days:
4 a)
Restore the inoperable PORV to OPERABLE status, or b)
Depressurize and vent the RCS through at least a 2.20 square inc c)
Depressurize and maintain a RCS vent through at least one open PORV and open ar,sociated block valve.
2.
With both PORVs inoperable, depressurin and vent the RCS through at least a 2.20 square inch vent within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In the event either the PORVs or a 2.20 square inch vent is used 3.
to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-The report shall describe.the
. tion 6.9.2 within 30 days.
circumstances initiating the transient, the effect of the PCWs or RCS vent (s) on the transient, and any corrective act';n necessary to prevent recurrence.
TURKEY POINT - UNITS 3 & 4 3/4 4-36 AMENDMENT N05.
AND FEB 2 81989 a-
REACTOR COOLANT SYSTEM OVERPRESSURE MITIGATING SYSTEMS SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
Performance of an ANALOG CHANNEL' OPERATIONAL TEST on the PORV actua-a.
tio'n channel, but excluding valve operation, within 31 days prior to entering a condition in which the P0i./ is required OPERABLE and at
'least once per 31-days thereafter when the PORV is required OPERABLE.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel b.
at least once per 18 months; and Verifying the PORV block valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> c.
when the PORV is being used for overpressure protection.
'While the PORVs are required to be OPERABLE, the backup air supply
~
~~
d.
shall be verified OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.9.3.2 The 2.20 square inch vent shall be verified to be open 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 vent (s) is being used for overpressure protection.
Verify the high pressure injection flow path to the RCS is isolated j
4.4.9.3.3 at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by closed valves with power removed or by locked l
closed manual valves.
l l
- Except when the vent pathway is provided with a valve which is locked, sealed, f
or otherwise secured in the open position, then verify these valves open at I
least once per 31 days.
TURKEY POINT - UNITS 3 & 4 3/4 4-37 AMENDMENT N05.
AND FEB 28 m
1e 1'
.,:v.
ji;\\'
'-I
4
.,c l
ATTACHNENT II-TURKEY POINT UNITS.3 1 4 SAFETY. EVALUATION FOR REVISED TECHNICAL SPECIFICATIONS APPENDIX A - NO SIGNIFICANT HAZARDS EVALUATION I
y.
I 1
j 1
I
-4 4
SAFETY'EVAtVATION
1.0 BACKGROUND
Turkey Point Units 3 & 4 currently operate with custom technical specifications issued'with the operating licenses in 1972 and 1973. Subsequently, the NRC has issued NUREG 0452, Standard Technical Specifications for Westinghouse Pressurized Water Reactors. The Standard Technical Specifications, which have b'een utilized
.by new licensed plants, are recognized to be more prescriptive and contain an increased number and frequency of surveillance than the custom Turkey Point Technical Specifications.
By letter dated April 11, 1984 to J. P. O'Reilly, NRC Regional Administrator, Region II, FPL formalized commitments to implement the Turkey Point Performance Enhancement Program. It was the intention of.FPL to review and implement, where appropriate, the philosophy and guidance of the Standard Technical Specifications in the development of upgraded plant procedures.
In addition, FPL committed to incorporate the requirements of the Standard Technical Specifications in all future proposed amendments to the Turkey Point Technical Specifications.
NRC Confirmatory. Order EA-84-55 dated July 13, 1984 required implementation of the Turkey Point Performance Enhancement Pronram (Revision I) and the commitments outlined in the April 11, 1984 letter.
In September of 1984, FPL voluntarily expanded the original commitment to include the development of a fully revised and reformatted set of Turkey Point Technical Specifications within certain limitations.
The revised set of Technical Specifications were to be based on Draft Revision 5 of NUREG 0452.
The limitations were that the revised Technical Specifications would not require hardware changes, would reflect the current Turkey Point plant design and analytical basis, and would consider operating hardship or reasonable resource e
additions.
The Technical Specification Revision Project became Performance Enhancement Project (PEP) No. 10, an addition to the original PEP Projects 1 through 9 which were under the Confirmatory Order.
2.0 EVALUATION The proposed amendment is a total replacement of the Turkey Point Units 3 & 4 current Technical Specifications with revised Technical Specifications which include the format and guidance of the Standard Technical Specifications within the limitations discussed above.
Paragraph 50.36 of Part 10 of the Code of Federal Regulations requires that each license authorizing operation of a commercial nuclear power plant shall include Technical Specifications that include the following categories of information:
Safety Limits, limiting Safety Settings and Limiting Control Settings Limiting Conditions for Operation Surveillance Requirements Design Features Administrative Controls including reporting requirements I
1
H Although. the current Technical Specifications include these categories of information, the revised Technical Specifications will allow incorporation' of additional information that has been gained through industry experience.and incorporated in the Standard Technical Specifications.
The revised Technical Specifications also include the format of the Standard Technical Specifications which has gained industry acceptance and will help resolve. minor instances of uncertainty that may exist in the current Technical Specifications.
Numerous new and more restrictive operational and surveillance requirements have been added to the revised Technical Specifications.
These requirements were modelled on the Standard Technical Specifications as they. apply to the Turkey Point plant design.
Because the standard plant model on which the Standard Technical Specifications are based envelopes the Turkey Point plant design bases in these areas, these new and more restrictive requirements are consistent with
. the Standard Technical Specification philosophy, and are, therefore, appropriate L
l-for Turkey Point.
The discussion that follows provides-a general overview of how the required categories of information have improved in content, format and understandability in the revised Technical Specifications.
Appendix A is the No Significant Hazards; Evaluation for each revised Technical Specification and includes - a specific ' summary and justification of changes for each revised Technical Specificari m.
2.1 Safety Limits. Limitina Safety System Settinas and limitina Control Sgttinas Safety limits for nuclear renctors are limits upon important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Key safety limits and safety system settings are found primarily in Chapter 2.0 of both the current and revised Technical Specifications.
Revised Technical Specification improvements in Chapter 2.0 include the addition of explicit ACTION statements for Sections 2.1.1 Safety Limits - Reactor Core and i
Section 2.1.2 Safety Limits - Reactor Coolant System Pressure and Reactor Trip System Instrumentation Setpoints. The revised Chapter 2.0 Technical Specifications have - been revised to clearly indicate applicable modes i
consistent with the Standard Technical Specifications. Revised Technical l
Specification Section 2.2.1 includes a more complete set of trip functions.
All of Section 2.0 revised Technical Specifications are reformatted in accordance with the Standard Technical Specifications.
2.2 Limitina Conditions for Operation 1
The limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
When the Limiting Condition for Operation is not met, the licensee must shut down the reactor 'or follow remedial actions as indicated in the Technical Specifications.
2 i
i I
i The text format of Chapter 3.0 of the current Technical Specifications can make difficult, in some cases, the determination of the Limiting Condition l
for Operation, the required actions and plant operating modes for which 1
conditions and actions are applicable. The revised Technical Specification i
uses the Standard Technical Specification format of explicit LIMITING CONDITIONS FOR OPERATION (LCO), Mode APPLICABILITY and ACTION statements l
for each Chapter 3.0 Technical Specification.
~
l For some Chapter 3 current Technical Specifications, the user must refer i
f to specification 3.0 which provides a generic action requirement in instances where the ACTION statement is not explicitly stated for the individual Technical Specifications. The inclusion of an explicit ACTION statement in each Chapter 2.0 and 3.0 revised Technical Specification will provide the user with improved clarity and direction.
The Chapter 2.0 and 3.0 current Technical Specifications do not consistently use the current industry accepted practice of using the mode applicability numbers to define the plant operating mode for which LIMITING CONDITIONS FOR OPERATION and ACTION statements are applicable. The revised Technical Specifications provide an explicit mode APPLICABILITY statement for each Chapter 2.0 and 3.0 specification in accordance with the format of the Standard Technical Specifications.
In addition to these format changes to improve operator understanding and interpretation, the revised Technical Specifications provide additional limitations, restrictions and controls.
Many of the revised Technical Specifications include more restrictive or additional LIMITING CONDITIONS FOR OPERATION and ACTION statements. Appendix A, No Significant Hazards Evaluations, describes' how certain revised Technical Specifications are more restrictive than the current Technical Specifications because they contain additional or more restrictive LIMITING CONDITIONS FOR OPERATION, Mode APPLICABILITY or ACTION statements.
2.3 Surveillance Requirements The SURVEILLANCE REQUIREMENTS are requirements for tests, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits, and that the limiting conditions for operation will be met.
Chapter 4.0 of the current and revised Technical Specifications provides the surveillance requirements.
In the current Technical Specifications, the Chapter 4 requirements are separated from their associated Chapter 3 system related LC0, APPLICABILITY and ACTION statements. In addition, the surveillance requirements for a particular system may appear in two or more Chapter 4 locations.
These features of the current Technical Specifications can make difficult the locating and identification of all LCO, APPLICABILITY, ACTION and SURVEILLANCE statements for a particular system.
The revised Technical Specifications utilize the Standard Technical Specification format of system Technical Specification which bring together the Chapter 3/4 LCO, APPLICABILITY, ACTION and SURVEILLANCE 3
l
statements. One of the most significant improvements is that a majority of the revised Technical Specifications include added or more restrictive Surveillance Requirements (see Appendix A).
Although most of the added surveillance requirements could previously be found in existing plant logs, t
procedures or testing programs, inclusion in the revised Technical Specifications brings these surveillance requirements into step with current industry requirements and practice.
2.4 Desian Features Design features are those features of the facility such as materials of construction and geometric arrangements, which if altered or modified, would have a significant effect on safety.
Chapter 5 provides a listing of design features in both the current and revised Technical Specifications.
Some new Chapter 5 Technical Specifications have been added to the revised Technical Specifications while other specifications containing outdated or unnecessary information have been deleted consistent with the Standard Technical Specifications.
Several Chapter 5 Technical Specifications are revised to provide additional information (see Appendix A).
.2.5 Administrative Controls and Reoortina Requirements Administrative controls are the provisions relating to organization and management, procedures, and record keeping, review and audit and reporting necessary to assure operation of the facility in a safe manner.
Chapter 6 provides Administrative Controls in both the current and revised Technical Specifications. The Chapter 6 revised Technical Specifications a
are provided in a format consistent with the Standard Technical Specifications (see Appendix A).
3.0 RELAXATIONS Selected revised Technical Specifications contain relaxations from the current Technical Specifications. Many of the relaxations concern changes in the ACTION statement times or surveillance frequencies.
In general, these relaxations bring the revised Technical Specifications in line with industry practice and the Standard Technical Specifications.
The key relaxations are justified in detail in the individual No Significant Hazards Evaluations (Appendix A).
4.0 CONCLUSION
The revised Technical Specifications represent a significant improvement in format, content and understandability for the Turkey Point operators and support groups. The Standard Technical Specification format will provide the NRC onsite inspectors and FPL engineering support groups a set of Technical Specifications that are consistent with the industry and, therefore, easier to use and Mcate information. As indicated in Section 2.0 above and the attached No Significant Hazards Evaluations, there is a significant increase in the content of information.
The added or more restrictive LC0"s, APPLICABILITY modes, ACTION 4
E,.,
.and SURVEILLANCE statements modelled on the Standard Technical Specifications will provide more comprehensive and consistent requirements for system readiness and operation.
The standards used to arrive at a proposed determination that the proposed included in 10 CFR changes involve no significant hazards consideration craThe individual No SE 92.
demonstrate that 'the revised Tecanical Specification (attached Appendix A)
Technical Specifications do not involve a significant hazards consideration in-L that.the proposed amendment would not:.(1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) i
. create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
Given these considerations, there is reasonable assurance that the health and safety of the public will not be endangered by the proposed Turkey Point. Units 3 & 4 Revised Technical Specifications, f
k i
RIN.SAF-E.TS s.
1
_____--________-__a
a
.t i
N0'SIGNIFICANT HAZARDS EVALUATION PROPOSED TECHNICAL SPECIFICATION
-TITLE:
PRES $URE/ TEMPERATURE LIMITS - REACTOR COOLANT SYSTEM N0:
3/4.4.RJ A.
DESCRIPTION OF CHANGES
.1)
Present Condition of License:
As described ' in the current Turkey Point Units 3 and 4 Technical Specification in Specification 3.1.2, B3.1.2,.4.20 and B4.20.
2).
Proposed Condition of License:
a.
The amendment consolidates the current requirements into. this specification and explicitly states the LCO, APPLICABLE MODES, ACTION Limits and SURVEILLANCE REQUIREMENTS.
b.
The revision is more complete than the current Technical Specification as follows:
a 1.
The revision adds a requirement that during heatup, cooldown or l
pressure testing the RCS temperature and pressure be determined-to be within the limits once per 30 minutes.
2..
The revision clearly states the surveillance requirement. that reactor vessel material specimens be removed and examined to determine changes in material properties as specified by 10 CFR50, Appendix H, in accordance witn the schedule in Table 4.4-5.
c.
The revision relaxes the following current requirements:
1.
The revision deletes Figure 3.1-2.
B.
BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION:
The standards used to arrive at a proposed determination that the changes described above involve no significant hazards consideration are included in 10 CFR 50.92.
The regulations state that if operation of the facility in accordance with the proposed amendment would not; (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety, then a no significant hazards determination can be made.
App. A 3/4 4-41
Proposed Tech. Spec. No. 3/4.4.9.1 The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing cartain examples (48 FR 14870) of amendments that are considered. not likely to involve a significant hazards consideration.
Example (i) relates to a purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature.
Example (ii) relates to a change that constitutes an additional, limitation, restriction, or control not presently included in the Technical Specifications for example, a more stringent surveillance requirement.
1)
The proposed change as described in Item 2.a is similar to example (i) of 48 FR 14870 in that 't is an administrative change which consolidates current requirements into a technical specification format consistent with the Standard Technical Specifications and does not involve technical or plant modifications.
2)
The proposed changes as described in Items 2.b.1 and 2.b.2 are similar to example (ii) cf 48 FR 14870 in that they provide additional restriction and controls by including an added surveillance requirement to determine RCS temp'.rature and pressure compliance every 30 minutes during plant status changes plus a reactor vessel material examination schedule.
3)
The proposed change as outlined in Item c.1 above does not involve a significant hazards consideration because this change would not:
a.
Involve a significant increase in the probability of or consequences of an accident previously evaluated.
This figure is obsolete, is not referenced in any current Technical Specification and has no impact on the operation of the plant.
b.
Create the possibility of a new or different kind of accident from any previously analyzed because the proposed change introduces no new mode of plant operation nor involves a physical modification to the plant.
c.
Involve a significant reduction in a margin of safety because this figure has no effect on the accident analysis or the normal operation of the plant.
Based on the above considerations the changes included in the development of proposed Technical Specification 3/4.4.9.1 are considered not to involve a significant hazards consideration as defined in 10 CFR 50.92. Further, there is reasonable assurance that the health and safety r.f the public will not be endangered by the proposed changes.
App. A 3/4 4-42 MM/3-4-4-9.1
J J
NO SIGNIFICANT HAZARDS EVALUATION PROPOSED TECHNICAL SPECIFICATION TITLE:
PRESSURE / TEMPERATURE LIMITS - PRESSURIZER NO:
3/4.4,9.2 A.
DESCRIPTION OF CHANGES 1)
Present Condition of License:
l
]
As described in the current Turkey Point Units 3 and 4 Technical i
Specification in Specifications.3.1.2 and B3.1.2.
I 2)
Proposed Condition of License:
l 1
I a.
The amendment consolidates the-current requirements into this specification and explicitly states the LCO, APPLICABLE MODES, ACTION Limits and SURVEILLANCE REQUIREMENTS.
b.
.The revision is more complete than the current Technical Specification as follows:
1.
The revision adds a requirement thct the PRESSURIZER TEMPERATURES be verified to be within the limits at least once per 30 minutes during heatup or cooldown.
2.
The revision adds a requirement that the spray water temperature differential be determined to be within the limit at least once 1
per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.
B.
BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION:
4 i
The standards used to arrive at a proposed determination that the changes l
described above involve no significant hazards consideration are included in l
i 10 CFR 50.92.
The regulations state that if operation of the facility in accordance with the proposed amendment would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any
{
accident previously evaluated, or (3) involve a significant reduction in a i
margin of safety, then a no significant hazards determinatic. can be made.
1 App. A 3/4 4-43 1
1
______a
Proposed Tech. Spec. No. 3/4.4.9.2 L
i The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve a significant hazards consideration.
Example'(i) relates to a purely administrative change to Technical Specifications:
for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature.
Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications for example, a more stringent surveillance requirement.
1)
The proposed change as described in Item 2.a is similar to example (i) of 48 FR 14870 in that it is an administrative change which consolidates current requirements into a technical specification format consistent with' the Standard Technical Specifications and does not involve technical' or plant-modifications.
l 2)
The proposed changes as described in Item 2.b.1 and 2.b.2 are similar to example (ii) of 48 FR 14870 in that they provide additional restrictions and controls by including an added surveillance requirement to verify pressurizer temperature every 30 minutes during pressurizer heatup or cooldown and to determine a. required temperature differential every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if pressurizer auxiliary spray is in operation.
Based on the above considerations the changes included in the development of proposed Technical Specification 3/4.4.9.2 are considered not to involve a significant hazards consideration as' defined in 10 CFR 50.92.
Further, there is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes.
App. A 3/4 4-44 MM/3-4-4-9-2
lif l
N0 SIGNIFICANT HAZARDS EVALUATION PROPOSED TECHNICAL' SPECIFICATION TITLE:
OVERPRESSURE PROTECTION SYSTEM NO:
3/4.4.9.3
- A.
DESCRIPTION OF CHANGES 1)
Present' Condition of License:
As described. i.n the current Turkey Point Units ' 3 and 4 Technical
~
Specification in Specifications 3.15, 4.16, B3.15 and B4.15.
2)
Proposed Condition'of License:
a.
'The amendment consolidates' the current requirements into this specification and explicitly states the LCO, AP.PLICABLE MODES, ACTION Limits and SURVEILLANCE REQUIREMENTS.
'b.
The revision is more complete than the current Technical Specification as follows:
1.
The revision requires that surveillance of the open.PORV
'Isoiation Valve be performed more. frequently. by reducing the interval from weekly to every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
2.
The revision requires the RCS vent (s) shall be verified to be open 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 vent (s) is being used for overpressure protection.
3.
The revision requires that the high-head safety injection be -
isolated from the RCS within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.
The revision ' changes the surveillance requirement to verify specific valve positions to a requirement to. verify isolation of the high pressure injection capability to the RCS.
App. A 3/4 4-45
~
A
Proposed Tech. Spec. No. 3/4.4.9.3 B.
BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION:
The standards used to arrive at a proposed determination that the changes described above involve no'significant hazards consideration are included in 10 CFR 50.92.
The regulations state that if operation:of the facility in accordance with the proposed amendment would not: -(1) involve a significant.
increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident fram any accident previously evaluated, or- (3) involve a significant reductic in a margin of. safety, then a n'o significant hazards determination can be maue.
The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples-(48 FR 14870) of. amendments that are considered not likely to involve a significant hazards consideration.
Example (i) relates to a purely administrative change to Technical Specifications:
for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a ' change in nomenclature.
Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications for example, a-more stringent surveillance' requirement.
1)
The proposed change as described in Item 2.a is similar to example (i) of 48 FR.14870 in that it is an administrative change which consolidates current requirements into a technical specification format consistent with the Standard Technical Specifications and does not involve technical or l
plant modifications.
I L
2)
The proposed changes as described in Item 2.b.1 and 2.b.2 are similar to
{
l example (ii) of 48 FR 14870 in that they provide additional restrictions and controls by requiring more frequent surveillance on the PORV's and RCS vent valves when used for overpressure protection.
l l
3)
The proposed changes as described in Items 2.b.3 and 2.b.4 do not involve a significant hazards consideration because these changes will not:
a.
Involve a significant increase in the probability of or consequences of an accident previously evaluated.
I l
This is essentially an administrative change rather than a relaxation.
Although the valve numbers are not stated explicitly, the requirement.
to isolate the high-head safety injection still exists. Making this I
change gives flexibility to the operators to use alternate methods / valves to isolate the flow path in the event that the primary method is not available and specifies the maximum time allowed.
l App. A 3/4 4-46 l
l
.1 a
Propos:d Tech. Spec. No. 3/4.4.9.3 b.
Create the possibility of a new or different kind of accident because -
the proposed change introduces no new mode of plant operation nor-involves a physical modification to the plant.
c.
Involve a significant reduction in a margin of safety because the intent of the Technical Specification remains the same, only the
- method of implementation is allowed to change.
Based on the above considerations the changes included in the development of proposed Technical Specification 3/4.4.9.3 are considered not to involve a significant hazards consideration as defined in 10 CFR 50.92.
Further, there-is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes.
App. a 3/4 4-47 MM/3-4-4-9-3
e 3
it ki !il mH UNITED STATES OF AMERICA NUCLEAR' REGULATORY COMMISSION
'89 JLL.18 P3:17 BEFORE THE. ATOMIC SAFETY AND LICENSING BOARD i
([$hig s
nyg In the Matter of
)
M ant h
)
Docket Nos. 50-250 OLA - 4 FLORIDA POWER & LIGHT
)
50-251 OLA - 4 COMPANY
)
)
(Turkey Point Plant, f
(P/T Limits)
Units 3-and'4)
)
CERTIFICATE OF SERVICE I hereby certify.that copies'of a letter dated July-14, 1989, from Steven P. Frantz to Licensing Board Members were served on the following by deposit in the United States. mail, first class, properly stamped and addressed, on the date shown below.
B. Paul Cotter, Chairman Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Glenn O. Bright Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 l
Jerry Harbour Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission We.shington, D.C.
20555 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Office of Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Chief, Docketing and Service Section Attention:
(Original plus two copies) e
g-
~ *,1 :.
3, v;,L e
)
1 j
"r (.Q
[Joette Lorion, Director-Center.for.NucleariResponsibility;.
L 7210 Red' Road #217 Miami,. Florida.;33143 fJanice' Moore:
. Patricia A.1Jehle:
1 Office-~of: General Counsel U.S.' Nuclear-Regulatory Commission E
Washington,t D.C.
.20555 2 Richard Goddard U.S.; Nuclear 7 Regulatory Commission 101)Marietta St., N.W.
- 2900
' Atlanta,. IGA 30323 John T.! Butler-
~
Steel, Hector & Davis
'4000. Southeast Financial' Center Miami, Florida-33131' Dated:thisi 4th day of July 1989.
1 M.
/
- Steven P. Frantz Newman-& Holtzinger, P=.C.
1615 L Street,-N.W.
Suite'1000-Washington,' D.C.
20036 a