ML18227C823
| ML18227C823 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/20/1974 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Case E US Atomic Energy Commission (AEC) |
| References | |
| Download: ML18227C823 (64) | |
Text
AEC DISTRIBUTION FOR PART 50 DOCKET MATERIAL (TEMPORARY FORM)
CONTROL NO:
9790 FILE.
FROM Florida Power
& Light Co Miami, Florida 33101 R
E Uhri TO:
Mr. Case CLASS UNCLASS PROPINFO DESCRIPTION DATE'OF DOC 9>>20-74 ORIG 2 si ned INPUT XXXX 9-23-74 CC OTHER SENT. AEC PDR SENT LOCAL PDR NO CYS REC'D DOCKET NO:
ENCLOSURES:
DATE REC'D LTR TWX RPT OTHER Ltr notarized 9-20<<74, submitting Amdt to Oper Lic., consisting of revised Tech Specs.
M/Attached Tech Specs pages(Appendix A).....
PL"ANTNAME: Turkey Point Unit f$ 3 & 4 FOR ACTION/INFORMATION pp I'lat PctiloU8 9-26-74 BUTLER (L)
W/ Copies
'LARK (L)
W/ Copies PARR (L)
W/ Copies KNIEL (L)
W/ Copies SCHWENCER (L)
W/ Copies STOLZ (L)
W/ Copies VASSALLO (L)
W/ Copies PURPLE (L)
W/ Copies ZIEMANN(L)
REGAN (E)
W/ Copies W/ Copies DICKER (E)
~LEAR (L)
W/ Copies W/9Copies KNIGHTON (E)
W/ Copies W/ Copies YOUNGBLOOD (E)
W/ Copies W/ Copies INTERNALDISTRIBUTION EG F
~AEC DR
~OGC, ROOM P-506A
~ MUNTZING/STAFF CASE G IAMBUSSO BOYD MOORE (L) (BWR)
DEYOUNG (L) (PWR)
SKOVHOLT (L)
~ GOLLER (L)
P. COLLINS DENISE REG OPR
~F I LE 5 R EG ION (3)
MORRIS STEELE TECH REVIEW SCHROEDER MACCARY KNIGHT PAWLICKI SHAO STELLO HOUSTON NOVAK ROSS IPPOLITO TEDESCO LONG LAINAS BENAROYA VOLIMER DENTON G R I MES GAMMILL KASTNER BALLARD SPANGLER ENVIRO MULLER DICKER KNIGHTON YOUNGB LOOD REGAN PROJECT LDR
~LEVHLAND HAR LESS LIC ASST DIGGS (L)
GEAR IN (L)
GOULBOURNE (L)
KREUTZER (E)
LEE (L)
IVIAIGRET (L)
REED (E)
SERVICE (L)
SHEPPARD (L)
SLATER (E)
SMITH (L)
~TEETS (L)
WI LLIAMS(E)
Wl LSON (L)
A/T IND B RAITMAN SALTZMAN B. HURT PLANS MCDONALD CHAPMAN
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~E. COUPE D. THOMPSON (2)
KLECKER EISENHUT EXTERNAL DISTRIBUTION
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~1 TIC (ABERNATHY)
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FLORIDA POWER & LIGHT COMPANY September 20, 1974 Mr. Edson G. Case, Acting Director Directorate of Licensing Office of Regulation U. S. Atomic Energy Commission Washington, D.
C.
20545
Dear Mr. Case:
Re:
Turkey Point Plant Units 3 and 4
Docket Nos. 50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 In accordance with 10 CFR 50.30, Florida Power
& Light Company submits here-with three signed originals and forty (40) conformed copies of a proposed amendment to Facility Operating Licenses DPR-31 and DPR-41.
This submittal replaces our corresponding. June 21,-.1974, request to amend the Facility Operating Licenses.
The changes are as set forth in the attached revised Technical Specification pages (Appendix A to DPR-31 and DPR-41) and are as described below:
~Pa e 1-5 The existing definition of'channel check cannot be applied to the Area and Process Radiation Monitoring Systems because these systems contain only one channel of instrumentation.
Therefore, the definition has been modified to allow an independent comparison of channel behavior by performing a radioactive 1
source check.
~Pa e 1-5 Abnormal occurrence definition (3) has been modified by referencing the specific section in the Technical Specifications which deals with release limits.
This change will make it easier to locate the requirement on release limits.
Abnormal occurrence definitions (4). and (6) have been changed to clarify the definition and bring it into conformance with Regulatory Guide 1.16.
~Pa e 1-6 Subsections 1.18, 1.19, 1.20 and 1.21 have been added to define terms used in the Technical Specifications which were not previously defined.
Subsections 1.18 and 1.19 have been added because of the change we are request-ing to definition (4) of subsection 1.14.
These definitions comply with those of Regulatory Guide 1.16.
HELPING BUILD FLORIDA
0 0
I '
Mr. Edson G. Case,'cting Director Page Two September 20, 1974 Subsection 1.20 has been added because of the changes we have requested to Section 6.
This definition was also taken. from Regulatory Guide 1.16.
Subsection 1.21 was added because of the change to the frequency definitions of Section 4 which we are proposing.
Pa e 2.1-1 The applicability statement and specifications 1 and 2 of Subsection 2.1 have been modified to show that the average reactor coolant temperature is the 1'imiting temperature.
These changes have not changed the intent of this subsection, but provide a consistent and clarified relationship between parameters described in this subsection and those shown in Figures 2.1-1 and 2.1-2.
Pa es 2.3-2 and 2.3-3 The term f(Aq) has been substituted for f(AI) to clarify the relationship
.between the specification and plant instrumentation and parameters.
The term F(AI) is only found in the'echnical Specifications.
It is no longer being used by the reactor vendor.
Therefore, this change will provide a necessary consistency between the Technical Specifications and plant instrumentation and parameters.
Table 3.5-2 The footnote to item 1.4 has been changed to reflect operation at reduced RCS pressure.
On October 12, 1973,'e submitted proposed change number'l which was approved on December 17, 1973.
Among the changers we requested was a reduction in the low pressurizer pressure trip setting.
However, the change in the footnote to item 1.4 which correspondingly reduced the setpoint at which the low pressurizer pressure signal could be bypassed was inadvertently omitted.
Therefore, we are including it in this submittal.
Pa es 3.6-1 and 3.6-2 The wording has been modified in specifications b.3 and c.3 of subsection 3.6 to more clearly define the requirements of the specifications.
The plant has three boric acid tanks which are shared by the two units.
The intent of these specifications is to assure that adequate boric acid is available to any reactor that is critical. It is not necessary to require that all three tanks maintain a continuous minimum inventory of boric acid; only those tanks which are in service should maintain the specified accumulated minimum inventory.
In subsection 3.10, specification 5 has been changed to specify reactor coolant temperature because at low temperature with the Residual Heat Removal System
Mr. Edson G. Case, Acting Director Page Three September 20, 1974 in service there is no meaningful measurement of Tavg.
Coolant temperature is measured by instrumentation in the Residual Heat Removal System.,
Table 4.1-1 A footnote has been added to Items 17A and 17B which eliminates the surveillance
" requirements for containment pressure when the equipment hatch or both" doors of either air lock are open.
The purpose of the check is to assure that the containment pressure instruments would adequately indicate'ny increase in containment pressure.
When the containment.is open to'he'tmosphere contain-ment pressure cannot increase above the atmosphere.
The only time the contain-ment would be open to atmosphere is during-cold shutdown when there is no chance of an accident creating high containment pressure.
The purpose of the test is to assure the containment pressure logic channels would function to isolate containment in case of a significant increase in containment pressure after or during an accident.
The applicability statement of subsection 4.4 has been expanded to include all the surveillance requirements of this subsection.
Subsection 4.4.1 has been rewritten so that the integrated leak rate test will be performed and reported in compliance with 10 CFR 50, Appendix J, which became effective after our Technical Specifications.
Subsection 4.4.2 has been rewritten so that the local penetration tests will comply with 10 CFR 50, Appendix J, except that 1) the, method of leak testing the air locks after each use has been changed from a pressure test to a vacuum test.
The vacuum 'test is an industry approved alternative to the pressure
- test, and will facilitate compliance with the'ntent of 10 CFR 50, Appendix J; and 2), the electrical penetration leak test requirement has been changed to require that the tests be performed prior to the integrated leak tests.
'This is consistent with the existing Technical Specification surveillance require-ments for local penetration tests.
The previous subsection 4.4.3 has been deleted because the test reporting requirements are now contained in 10 CFR 50, Appendix J.
Subsection 4.4.3 (previously designated subsection 4.4.4) has been rewritten so that the containment isolation valve tests will comply with 10 CFR 50, Appendix J, which became effective after our Technical Specifications.
0 0
Mr. Edson G. Case, Acting Diiector Page Pour September 20',
1974 Subsection 4.4.5 has been redesignated, subsection 4.4.4.
Specification (f) has been changed to provide for a test of the Residual Heat Removal System each refueling.
This change eliminates an otherwise unnecessary shutdown and cooldown for both units. It is consistent with other safeguards system test requirements and does not change the intent of this subsection since refueling will be at ~l year intervals in the future.
Pa es 4.4-'4 4.'4-5 and 4.4-6 Subsection 4.4.6 has been redesignated subsection 4.4.5.
Subsection 4.4.7 has been redesignated subsection 4.'4.6.
Since the structural integrity tests of each containment have been completed, the dates have been included in the specifications to clarify inspection dates which are based on the completion dates of the structural integrity tests.
Pa e 4.4-7 Subsection 4.4.8 has been redesignated subsection 4.4.7.
A sentence was also added'o the end of this subsection denoting the fact that liner sur-veillance has been satisfactorily completed.
Pa e 4;5-1 In subsection 4.5, specification l.a has been changed to permit testing of the safety injection system by actually starting a residual heat removal pump.
This change provides an alternate method of testing the RHR without requiring that both residual heat removal pumps be inoperable during the test.
This is a more valid test because the residual heat removal pumps are available for decay heat removal should the need arise.
Pa e'4.6-2 The acceptable level of performance for the emergency containment cooling fans has been modified.
The modification is necessary because the fans are rated for ambient conditions following the Maximum Hypothetical Accident; and these conditions cannot be duplicated for the test.
Actually the test con-ditions provide much lower loads due to lower temperatures and a lower moisture content in the atmosphere preventing the fan motors from reaching their rated current.
The modified wording bases the performance of the fans on the con-tainment atmosphere at the time of the test.
Pa e 4.7-1 In subsection 4.7, specification 4.7.1.1 a typographical error has been corrected.
Pa es 4.8-1 and 4.8-2 In subsection 4.8, specification l.b the second sentence has been corrected.
The intent of this specification is to thoroughly test each diesel generator
Mr. Edson G. Case, Acting Director Page Five September 20, 1974-to assure that the emergency power system will respond promptly and properly.
In subsection 4.8, specification l.e the first sentence has been changed to allow the requirements of subsection 4.8, specification l.a to be exceeded during a test.
This is necessary to meet the requirement that the diesel generator be loaded to 2750 KW. If the'oad is not permitted below 2750 KW, then some operating band above 2750 KW must be allowed to provide for governor swings.
The margin we are requesting above 2750 KW will not adversely affect diesel. operation.
The diesels are rated much higher than 2750 KW for short periods of time; the test lasts approximately 30 minutes.
The second sentence in specification l.e was also changed to clarify the intent of the specification.
During preoperational testing it was found that the loading during a loss of AC power test is less than 2750 KW, however, the monthly test requires a loading of at least 2750 KW.
To eliminate this dis-
- crepancy, the loading listed in FSAR Table 8.2-2 is applicable only during the loss of AC power test.
In subsection 4.8, specifications 2.a and 2.b have been changed to clarify the requirement that each station battery must be checked.
This change eliminates the ambiguity which existed in the original wording.
In subsection 4.10, a statement was added to the end of specification 1 to preclude the requirement to test the auxiliary feedwater pumps during cold shutdown.
It is not possible to comply with this specification during cold shutdown because
- 1) there is no steam available to drive the auxiliary feed-water pumps, and 2) at cold shutdown, the steam generators are either filled with feedwater or drained for maintenance.
In either case, we cannot establish a 600 gpm flow rate as required.
Performing the test during the next startup meets the intent of'the Technical Specifications of assuring Auxiliary feedwater capability during unit operation.
Pa e B2.3-2 The term f(Aq) has replaced f(AI) in the Reactor Coolant Temperature Bases.
This change was made to be consistent with the proposed Technical Specifications.
We have reviewed these changes and have concluded that they do not involve any significant hazards considerations and there is reasonable assurance that the health and safety of the public will not be endangered.
Very truly yours, Robert E. Uhrig Vice President REU/DWR/cpc Attachment cc:
Mr. Jack R.
Newman
STATE OF FLORIDA )
)
ss COUNTY OF DADE
)
Robert E. Uhrig, being first,duly sworn, "deposes and says:
That he is a Vice President of Florida Power a Light.Company, the Licensee herein; That he has executed the foregoing instrument; that the statements made in this said instrument are true and correct to the best of his knowledge, information and belief; and that he is authorized to execute the inst:rument of said Licensee.
Robert E. Uhrig Subscribed and sworn to be ore me o
1974.
Notar Public in and.for the County of Dade, State of Florida My Commission expires'OTARY PUBUC. STATE of FLORIDA at LARGE jg COMMISSION EXPIRES APRIL 2, 1918 hoHDEQ XHRLI BAYBhBQ 305QQlO d04WR
FLORIDA POWER 8I LIGHT COMPANY September 20, 1974 Mr. Edson G. Case, Acting Director Directorate of Licensing Office of Regulation U.
S. Atomic Energy Commission Washington, D. 'C.
20545
Dear Mr. Case:
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Re:
Turkey Point Plant Units 3 and 4
Docket Nos. 50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 In accordance with 10 CFR 50.30, Florida Power
& Light Company submits here-with three signed originals and forty (40) conformed copies of a proposed amendment to Facility Operating Licenses DPR-31 and DPR-41.
This submittal replaces our corresponding June 21, 1974, request to amend the Facility Operating Licenses.
The changes are as set forth in the attached revised Technical Specification pages (Appendix A to DPR-31 and DPR-41) and are as described below:
~Pa e 1-3 The existing definition of channel check cannot be applied to the Area and Process Radiation Monitoring Systems because these systems contain only one channel of instrumentation.
Therefore, the definition has been modified to allow an independent comparison of channel behavior by performing a radioactive source check.
~Pa e 1-5 Abnormal occurrence definition (3) has been modified by referencing the specific section in the Technical Specifications which deals with release limits.
This change will make it easier to locate the requirement on release limits.
Abnormal occurrence definitions (4) and (6) have been changed to clarify the definition and bring it into conformance with Regulatory Guide 1.16.
~Pa e 1-6 Subsections 1.18, 1.19, 1.20 and 1.21 have been added to define terms used in the Technical Specifications which were not previously defined.
Subsections 1.18 and 1.19 have been added because of the change we are request-ing to definition (4) of subsection 1.14.
These definitions comply with those of Regulatory Guide 1.16.
][6788'ELPING BUILD FLORIDA
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(
Mr. Edson G.
- Case, Acting Director Page Two September 20, 1974 Subsection 1.20 has been added because of the changes we have requested to Section 6.
This definition was also taken from Regulatory Guide 1.16.
Subsection 1.21 was added because of the change to the frequency definitions of Section 4 which we are proposing.
Pa e 2.1-1 I
The applicability statement and specifications 1 and 2 of Subsection 2.1 have been modified to show that the average reactor coolant temperature is the limiting temperature.
These changes have not changed the intent of this subsection, but provide a consistent and clarified relationship between parameters described in this subsection and those shown in Figures 2.1-1 and 2.1-2.
Pa es 2.3-2 and 2.3-3 The'erm f(Aq) has been substituted for f(AI) to clarify the relationship between the specification and plant instrumentation and parameters.
The term F(AI) is only found in the Technical Specifications.
It is no longer being used by the reactor vendor.
Therefore, this change will provide a necessary consistency between the Technical Specifications and plant instrumentation and parameters.
Table 3.5-2 The footnote to'tem 1.4 has been changed to reflect operation at reduced RCS pressure.
On October 12, 1973, we submitted proposed change number ll which was approved on December'7, 1973.
Among the changers we requested was a reduction in the low pressurizer pressure trip setting.
However, the change in the footnote to item 1.4 which correspondingly reduced the setpoint at which the low pressurizer pressure signal could be bypassed was inadvertently omitted.
Therefore, we are including it in this submittal.
Pa es 3.6-1 and 3.6-2 The wording has been modified in specifications b.3 and c.3 of subsection 3.6 to more clearly define the requirements of the specifications.
The plant has three boric acid tanks which are shared by the two units.
The intent of these specifications is to assure that adequate boric acid is available to any reactor that is critical. It is not necessary to require that all three tanks maintain a continuous minimum inventory of boric acid; only those tanks which are in service should maintain the specified accumulated minimum inventory.
In subsection 3.10, specification 5 has been changed to specify reactor coolant temperature because at low temperature with the Residual Heat Removal System
Mr. Edson G.
- Case, Acting Director Page Three September 20, 1974 in service there is no meaningful measurement of Tavg.
Coolant temperature is measured by instrumentation in the Residual Heat Removal System.
Table 4.1-1, A footnote has been added to'Items 17A and 17B which eliminates the surveillance requirements for containment pressure when the equipment hatch or both doors of either air lock are open.
The purpose, of the check is to assure that the containment pressure instruments would adequately indicate any increase in containment pressure.
When the'ontainment-is open to the atmosphere contain-ment pressure cannot increase above the atmosphere.
The only time the contain-ment would be open to atmosphere is during cold shutdown when there is no chance of an accident creating high containment pressure.
The'purpose of the test is to assure the containment pressure logic channels would function to isolate containment in case of a significant increase'n containment pressure after or during an accident.
The applicability statement of subsection 4.4 has been expanded to include all the surveillance requirements of this subsection.
Subsection 4.4.1 has been rewritten so that the integrated leak rate test will be performed and reported in compliance with 10 CFR 50, Appendix J, which became effective after our Technical Specifications.
Subsection 4.4.2 has been rewritten so that the local penetration tests will comply with 10 CFR 50, Appendix J, except that 1) the method of leak testing the air locks after each use has been changed from a pressure test to a vacuum test.
"The vacuum test is an industry approved alternative to the pressure
- test, and mill facilitate compliance with the intent of 10 CFR 50, Appendix J; and 2). the electrical penetration leak test requirement has been changed to require that the tests be performed prior to the integrated leak tests.
This is consistent with the existing Technical Specification surveillance require-ments for local penetration tests.
The previous subsection 4.4.3 has been deleted because the test reporting requirements are now contained in 10 CFR 50, Appendix J.
Subsection 4.4.3 (previously designated subsection 4.4.4) has been rewritten so that the containment isolation valve tests will comply with 10 CFR 50, Appendix J, which became effective after our Technical Specifications.
\\
F
Mr. Edson G. Case, Acting Director Page Four September 20',. 1974 Subsection 4.4.5 has been redesignated subsection 4.4.4.
Specification (f) has been changed to provide for a test of the Residual Heat Removal System each refueling.
This change eliminates an otherwise unnecessary shutdown and cooldown for both units. It is consistent with other, safeguards system test requirements and does'ot change the intent of this subsection since refueling will be at ~l yea'r intervals in the future.
Pa es 4.-4-4 4.'4-'5 'and '4.4-6 Subsection 4.4.6 has been redesignated subsection 4.4.5.
Subsection 4.4.7 has been redesignated subsection 4.4.6.
Since the structural integrity tests of each containment have been completed, the dates have been included in the specifications to clarify inspection dates which are based on the completion dates of the structural integrity tests.
Pa e-4.4-7 Subsection 4.4.8 has been redesignated subsection 4.4.7.'
sentence was also added'o the end of this subsection denoting the fact that liner sur-veillance has been satisfactorily completed.
In subsection 4.-5, specification l.a has been changed,to permit testing of the safety injection system by actually starting a residual heat removal pump.
This change provides an alternate'ethod of testing the'HR without requiring that both residual heat removal pumps be inoperable during the test.
This is a more valid test because the residual heat removal pumps are available for decay heat removal should the need arise.
Pa e'4.6-2 The acceptable level of performance for the emergency containment cooling fans has been modified.
The modification is necessary because the fans are rated for ambient conditions following the Maximum Hypothetical Accident; and these conditions cannot be duplicated for the test.
Actually the test con-ditions provi'de much lower loads due to lower temperatures and a lower moisture content in the" atmosphere preventing the fan motors from reaching their rated current.
The modified wording bases the performance of the fans on the con-tainment atmosphere at the'ime of the test.
Pa e'4.7-1 In subsection 4.7, specification 4.7.1.1 a typographical error has been corrected; Pa es 4.8-1 and 4.8-2 In subsection 4.8, specification l.b the second sentence has been corrected.
The intent of this specification is to thoroughly test each diesel generator
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Mr. Edson G. Case, Acting Director Page Five September 20, 1974, to assure that the emergency power system will respond promptly and properly.
In subsection 4.8, specification l.e the first sentence has been changed to allow the requirements of subsection 4.8, specification l.a to be exceeded during a test.
This is necessary to meet the requirement that the diesel generator be loaded to 2750 KW. If the load is not permitted below 2750 KW, then some operating band above 2750 KW must be allowed to provide for governor swings.
The margin we are requesting above 2750 KW will not adversely affect diesel operation.
The diesels are rated much higher than 2750 KW for short periods of time; the test lasts approximately 30'inutes.
The second sentence in specification l.e was also changed to clarify the intent of the specification.
During preoperational testing it was found that the loading during a loss of AC power test is less than 2750 KW, however, the monthly test requires a loading of at least 2750 KW.
To eliminate this dis-
- crepancy, the loading listed in FSAR Table 8.2-2 is applicable only during the loss of AC power test.
In subsection 4.8, specifications 2.a and 2.b have been changed to clarify the requirement that each station battery must be checked.
This change eliminates the ambiguity which existed in the original wording.
In subsection 4.10, a statement was added to the end of specification 1 to preclude the, requirement to test the auxiliary feedwater pumps during cold shutdown.
It is not possible to comply with'his specification during cold shutdown because
- 1) there is no steam "available to drive the auxiliary feed-water pumps, and;2), at cold shutdown,, the steam generators are either filled with feedwater or drained for maintenance.
In either case, we cannot establish a 600 gpm flow rate as required.
Performing the test during the next startup meets the intent of the Technical Specifications of assuring Auxiliary feedwater capability during unit operation.
Pa e'2.3-2 The term f(Aq) has replaced f(AI) in the Reactor Coolant Temperature Bases.
This change was made to be consistent with the proposed Technical Specifications.
We have reviewed these changes and have concluded that they do not involve any significant hazards considerations and there is reasonable assurance that the health and safety of the public will not be endangered.
Very truly yours, Robert E. Uhrig Vice President REU/DWR/cpc Attachment cc:
Mr. Jack R.
Newman
l
STATE OF FLORIDA )
)
ss COUNTY OF DADE
)
Robert E. Uhrig, being first,duly sworn, 'deposes and says:
That he is "a Vice President of Florida-Power 6 Light.Company, the Licensee herein; I
That he has executed the foregoing instrument; that the statements made in this said instrument are true and correct to the best of his knowledge, information and belief; and that he is authorized to execute the instrument of said Licensee.
Robert E. Uhrig Subscribed and sworn to be ore me this 0
dayo'974.
Notar Public in and for the County of Dade, State;o f; Florida s'
My.Commission expires
'OTARY PUBLIC, STATE of FLORIDAat LARGE Jg,COMMISSION EXPIRES APRIL 2, 1978 hoNDEQ XHBQ k!AYNARQ QoBQJ50 hSKHQC
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Case, Acting Diroctox Directorate of Licensing Office of Regulation U. S. Atomic Energy Commission Mashington, D+ C, 20545 Dear Hra Case!
+9llixq
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FLORIDA POWER Ea llGIIT COMPANY Septombor 20, 1974 i;r) I
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Re!
Tuxkoy Point Plant Units 3 and 4 Docket 'gos.~50 25~d 50-251 Proposed Amendment to Facility oratin Licenses DPR-31 and DPR-41 In accordance'ith 10 CFR 50>30, Florida Powex 6 Light Company submits here-with thxee signed oxiginals and forty (40) conformed copies of a proposed amendment to Facility 'Operating Licenses DPR-31 and DPR>>416 This submittal replaces our corresponding Juno 21, 1974, request to amend the Facility Opexating Licenses.
Tho changes are as set foxth in tho attached govised Technical SpeciHcation pages (Appendix A to DPR-31 and DPR-41) and axe ao doscxibed below!
~Pa a 1a6 The existiig deHnition, of channel check cannot'bo applied to tho Area and Px'ocess Radiation Monitoring Systems because these systems contain only ono channel of instrumentation.
,Therefore, the definition has been modified to allow an independent comparison of channel bohavior'y porforming a radioactive source check.
~Pa a 15-H Abnormal Occurrence definition (3).has been modified by roferencing the specific section in tho Technical Specifications which'Goals with release limits.
This change willmako it easier to locate the requirement.
on release limits.
Abnormal occurrence deHnitions (4) and (6) havo been changed to clarify tho definition and bring it into conformance with Regulatory Guide 1.16.
~Pa a 1-6 Subsections 1<18, 1.19, 1.20 and 1.21 have been added to defino terms used in tho Technical SpeciHcations which were not previously deHned.
Subsections 1.18 and 1.19 have been added because of tho change we are request-ing to definition (4) of subsection 1.14.
These doHnitions comply with those of Regulatory Guide 1.16.
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Hx'. Mson Q. Casa, Acting Director Page Two September 20, 1974 Subsection 1.20 has been added because of the changes ve have requested to Section 6 ~
This definition vao also talcon from Regulatory Guide 1.16.
Suboaction 1.21 vao,added because of the change Co the frequency dafinitiono of Soction 4 vhich we axe px'oposing.
The applicability statement and specifications 1 and 2 of Subsection 2,1 have been modified to shov that the average reactor coolant temperature is tha limiting tempex'ature.
Theoe changes have not changed the intent of Chis subsection, but, provide a conoiotent and claxified rolationship between parameters described in this
,subsection and those shown + figures 2.l-l and 2.1-2.
Pa ao 2o3-2 and 2+3-3 The term f(hq) has been substituted for'(AX) to clarify the relationship between Cha specification and plant instrumentation and parameters.
'The term 7(AX) io only found in'he Technical Spacificationo.
Xt, is no longer being used by the reactor vendor.
Therefore,'thio change willprovide a neceosary consistency betvaen the Technical Specifications and plant inotrumentation and parameters.
Table 3.'5-2 The footnote to item 1~4 hao boon changed to x'eflact operation aC reduced RCS pressure.
On October 12, 1973, va submitted proposed change number 11 which vas approved on December 17, 1973.
Among tha changes we requeoted wao a x'eduction in the lov pressuriser pressure txip setting.
However, the chango in the footnote to item 1.4 which correspondingly x'educed the setpoint at vhich the 1ov prossuriser pxessuro signal could be bypassed vas inadvertently omitted.
Therofox'o, ve are including it in this submittal.
Pa eo 3,6-1 and 3.6-2
'gha wording has been modified in specifications b.3 snd c.3 of subsection 3.6 to more clearly define tha requirements of tha specifications.
Tho pLant has three boric acid tanks vhich are shax'ed by the tvo units.
" The intent of these specifications is to assure that adequate boric acid io available Co any reactox'hat is critical.
Xt is not necessary to xequixa that all three tanks maintain a continuous minimum inventory of boxic acid; only those tanks which are in service should maintain the specified accumulated minimum inventory.
Pa e 3.10-2 Xn subsection 3.10, specification 5 has been changed to opocify reactor coolant temperature because at lov temperatuxe with the Residua1 Heat Removal System
0
~
Be. Mson C, Case, Acting Director Page Three September 20, 1974 in sexvice there is no meaningful measurement of Tavg.
Coolant tompexaturo is measured by. instrumentation in, the Residual Heat Removal System, Table 4.1-1 A. footnote has been added to Itcmo 17A and L78 which eliminates the suxveiLLanco requirements for containment prcssure when the equipment hatch or both doors of either air lock are open.
The purpose of the check is Co aosuxe that Cho containment pxeoouxo instruments would adequately indicate any increase in containment pressuxe.
tenon the containment io open to the atmosphere contain<<
ment pressure cannot increase above the atmosphere.
The only time the contain-ment would bo open to atmosphere io during cold shutdown when thero is no chance of an accident croating high containment pressure.
The purpose of the Cost is to assure the containmcnt prcssure logic channels would function to isolate containment in case of a significant increase in containmenC pxessuxe after'r during an accident.
Pa e 4~4<<L The.applicability statement of subsection 4.4 has 'been expanded to include all the surveillance xequix'ements of this subsection.
Subsection 4.4.1 has been rewritten so that the integrated leak rate tesC will be performed and reported in compliance with 10 CPR 50, A>pendix J, which became affective'after our Technical Specifications.
Pa e 4.4-2
, Subsection 4,4.2 has been. xowritten so that the local penetration tests will comply with 10 CFR 50> Appendix J, except that 1) the method of Leak testing the air Locks after oach use has been changed from a pressure test to a vacuum Cest.
The vacuum test is en industry approved alternative to the preosuro Ceot 9 and wi11 faci1itate compliance with the intent of 10 CPR 50, Appendiz J; and 2)
Che electrical peneCxation leak test xequiromont has been changed to require that the tests be performed prior to the integrated Leak teste.
This is consistent with the existing Technical Specification suxveillanco requixo-ments for Local ponetxation tests.
The pxevious subsection 4.4o3 has boon deleted because the Cost, xepoxting xequixements are now contained in 10 CPR 50, Appendix J.
Pa e 4.4-3 Subsection 4.4.3 (previously designated subsection 4.4.4) has been rowritten oo thaC the containment ioolation valve tests wiLL comply with 10 CPR 90, Appondhc J, which became offoctive after our Technical Specificationo.
Hl> Edson Go CQoey ACCing Director Page Pour September 20'974 Subsoction 4<4<5 has boon redesignated ouboection 4.4.4.
Specification (f) has bean changed Co provide for a tost of the Reoidual Heat Removal System each refueling.
This chango el~tea an otherwiaa unnecessary ohutdown and cooldown fox'oth unitoo XC is consistent with othex safcguax'ds system Cest requirements and does not change the intent of this subsection oinca x'efueling will bo at 'C year intervals in the future.
4.4-5 and 4.4-6 Subsoction 4 4.6 haa been redeoignatcd subsection 4,4,5, Subsection 4i4.7 has been redesignated oubsection 4.4,6.
Since Cho structuxal integxity Coats of each contain3Mnt halva beon completed, Che datoo have beon included in Che specificationo Co clarify inspection daCes which axe baaed on the completion dateo of Cho structural hitegrity tests.
Pa o 4.4-7 Subsection 4.4.8 has boon redesignated subsection 4.4.7.
A sentence was also added Co the end of this subsoction denoting the face, Chat. linex sur-vaillance has baen oatiofactorily completed>
~LB-Xn Gubscction 4,5, apacification l.a has been changed Co permit tooting of the safety in)ection ayotom by actually starting G residual heat removal pump.
This change provides Gn alternate method of testing Che RHR without requiring that both residual heat removal pumps be inoperable during the test.
This ia a'more valid test because the residual heat removal pumps are available for decay heat removal Ghould Che nead ariso.
Pa e 4e6-2 The acceptable level of performance for the emergency containment cooling fano has been modified.
Tho modification is necessary becouse the fans are rated for ambient conditiono following the Maximum Hypothetical Accident~ and Chose conditions cannot bo duplicated fox Che test.
Actually the tost con-ditions px'ovide much lower loads due Co lower temperatures and a lower moisture content in the atmoophoro provonting Che fan motors from reaching their rated, current.
The modified wording bases Cho performance of the fans on the con-tainment atmosphex'e at Che time of the Cast.
Pa 0 4.7-1 Xn oubsecCion 4o7y specification'o7alol 6 typographical error haQ beQn correctede Pa es 4.8-1 and 4.8-2 Xn subsoction 4.8, spocification l.b Che second scntcnco has been corrected.
The intent of Chio opocification io to choxoughly toac each diesel gonorator
4
'I I'
F
Hr. Moon G. Caso, Acting Director Page Five September 20, 1974 to assure that tho emergency powox system villrespond promptly and proporly.
In subsection 4.8, spocification l.e the first sontoaco has boon changed to allov the roquixomoats of subsection 4.8, specification. l.a to be ezcooded during a test.
This is necessary to meet the xoquiroment that the diesel generator bo loaded to 2750 KQ. If the load is not permitted bolov 2750 KP, then some opexatiag band abovo 2750 KP must bo allowed to provide for governor wrings.
The margin vo axo xequestiag above 2750 161 villnot advex'sely'affect diesel operation.
The dioaels axe rated much higher than 2750 KH for ohort periods of time; the test lasts approzimately 30 minutes.
The second sentence ia specification l.e vas also changed to clarify tho intent of tho spocification.
During preopoxational tooting it vao found that the loading during a loss of AC power test is loso than 2750 KH, hovevor, tho monthly test requixes a loading of at least 2750 KP.
To eliminate this dis>>
- cropaacy, tho loading listed ia FSAR Table 8.2-2 io applicablo only during tho loss of AC powex test.
Zn subsection 4.8, specifications 2.a and 2.b have boen changed to claxify tho requirement that each station battery must be checked.
This change eliminates the ambiguity which ezisted in the original wording.
~4.
In subsection 4,10, a statemont vas added to tho end of specification 1 to
, preclude tho requirement to tost the auziliary foodvatex pumps during cold shutdown. It is not possible to comply vith this specification during cold shutdown because
- 1) there is no steam available to drive the auzfliary food-vater pumps, and 2) at cold shutdown, the steam gonoratoxs ax'o either ffllod vfth foedvatox ox drained for maintenance, In oithor casa, vo cannot establish a 600 gpm flow rata as roquixod.
Performing tho,tost during the nozt startup moots tho intent of tho Technical Specifications of aoouring Auziliaxy foodvator capability dux'ing unit operation.
Pa o 32.3-2 The term f(hq) has replaced f(AI) in the Reactor Coolant Tompoxatuxo Bases.
This chango vas made to bo consistent vith tho proposed Technical Specifications.
Ve have reviewed those changes and have concluded that thoy do not involve any significant hazards considoratioas and there ia reasonable assurance that the health aad safety of the public villnot bo ondangox'ed.
Very tzuly yours, Robeit E. Uhrig Vice President REU/DRL/opc Attachment ccrc Nr. Jack R. Howmaa
E
~
I n
I
STATE OF FZORXDA )
)
SS COUNTY OF DADE, )
Robert E. Uhr9.g, being first duly sworn, deposes and says~
That he is a Vice Prosident of 'Florida Power a Light Company, the Licensee herein<
That he has executed the foregoing instrument~ that the statements made in this said instrument are true and correct to the boot of his Jmowledge, information and boliefy and that he is authorisad to execute tho instrument of said Licensoo.
Subscribed and sworn to bofo p me this
-6 day of M~ ~, 1974.
No~
lic in and for the County of Dade, State of Florida
'+"~<< "+NBA~t tAHI.=
~'~~ @M""~~3~08 EXPIRES PPgg,g gpg Ny Commission expires P<~-"~ ~u ~~~ ~...
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Cy number of operable channels and the number of channels which when tripped will cause reactor trip.
1.7 INSTRUMENTATION SURVEILLANCE 1)
Channel Check Channel check is a qualitative determination of acceptable operability by observation of channel behavior during operation.
This determination shall include comparison of the channel with other independent channels measuring the same variable or radio-active souice check of the Area and Process Radiation Monitoring
-Systems, for channels.
1 2)
Channel Functional 'Test A channel functional test consists of injecting a,simulated'ignal into the channel to verify that it is operable, in-cluding alarm and/or trip initiating action.
3)
Channel Calibration L
'" P Channel calibration consists of the adjustment of channel output such that it responds, with acceptable range and ac-curacy, to kn'own values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including alarm or trip, and shall be deemed.to include the channel functional test.
- 1. 8 SHUTDOWN 1)
Cold Shutdown F
II The reactor is in the cold shutdown condition when the reactor is subcritical by at least 1X hk/k and T
is less than 200F.
avg 2)
Hot Shutdown The reactor is in the hot shutdown condition when it 1-3 6/21/74
C ~
- 1. 14 ABNORMAL 088fRRENCE An abnormal occurrence is defined as any of the following:
1.
A safety system setting less conservative than the limiting set-ting established in the Technical Specifications.
2.
Violation of a limiting condition for operation established in the Technical Specifications.
3.
An uncontrolled or unplanned release of radioactive material 1
from any plant system designed to act as a boundary for such material in an amount of significance with respect to limits prescribed in Technical Specification 3.9.
4.
Incidents or conditions which prevented or could have prevented the performance of =the intended safety function of an engineered safety feature system or of the reactor protection system.
Jan 5.
Abnormal degradation of one of the several boundaries designed to contain the radioactive materials resulting from the fission process.
6.
Any uncontrolled or unanticipated change in reactivity equal to or greater than 1X.
K 7.
Observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the plant.
8.
.Conditions arising from natural or offsite manmade events that affect or threaten to affect the safe operation of the plant.
1.15 POWER TILT The power tilt is the ratio of the maximum to average of the upper out-of-core normalized detector currents or the lower out-of-core normalized detector currents whichever is greater.
If one out-of-core detector is out of service, the remaining three detectors are to be used to compute the average.
1-'5 6/21/74
1,16 FUEL RESIDENCE TIME LIMIT.
The fuel residence time for cycle 1 shall be limited to 11,600 effective full power hours (EFPH) for Unit 3 and 24,500 EFPH for Unit 4 under low pressure operating conditions.
li17 LOW POWER PHYSICS TESTS Low power physics tests are tests below a nominal 5% of rated power which measure fundamental characteristics of the reactor core and related instrumentation.
r 1.18 ENGINEERED SAFETY FEATURES Features such as containment, emergency core cooling, and containment atmospheric cleanup systems for mitigating the consequences of postulated accidents.
- l. 19 features as necessary.
REACTOR PROTECTION SYSTEM Systems provided to act, if needed, to avoid exceeding a safety limit in anticipated transients and to activate appropriate engineered safety 1.20 SAFETY RELATED SYSTEMS AND COMPONENTS Those plant features necessary to assure the integrity of the reactor coolant pressure
- boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, or the capability to 'prevent or mitigate the consequences of accidents which could result in off-site exposures comparable to the guideline exposures of 10 CFR 100.
1.21 PER ANNE During each calendar year.
1-6 6/21/74
,2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, average coolant temperature and flow during power operation.
~ob ective:
To maintain fuel cladding integrity.
1.,
THREE LOOP OPERATION The combination of'thermal power 1evel, coolant pressure and average coolant temperature shall not exceed the limits shown in Figure 2.1-1 for full flow from three reactor coolant pumps.
2.
TWO LOOP OPERATION The combination of thermal power level, coolant pressure and average coolant temperature shall not exceed the limits shown in Figure 2.1-2 for full flow from two reactor coolant pumps.
3.
ONE LOOP OPERATION c
\\
l v Ve A>e
- The thermal power level shall not exceed
- 20X, coolant pressure shall be maintained in the 1820 2400 psig range, and the average coolant temperature shall not exceed 590 F for full flow from one reactor coolant pump.
v 4.
NATURAL CIRCULATION The thermal power level shall not exceed
- 12K, coolant pressure shall be maintained in the 2135 2400 psig range and the average coolant temperature shall not exceed 602 F, when no reactor coolant pumps are in operation.
- 2. 1-1 6/21/74
Reactor Coolant Tem erature Overtemperature hT
< AT Kl' 0174(T-566.6)
+ 0.000976(P-1885) - f(Aq)
~ 1 I
startup tests such that:
/I hT
~ Indicated hT at.rated power, F
I T
Average temperature, F
P
,, ~ Pressurizer=pressure, psig I
f(Aq).~ a function of the indicated difference between top'nd,,bottom detectors of'the power-range
~-
nuclear ion chambers;
.with gains to be selected based on measured instiument response during For (q - q ) within +10 percent and -14 percent t
b where q
and q
are the percent power in the top and bottom halves of the core respectively, and q
+ q is total. core. power in percent of rated.
power, f(hq) ~ 'O.
For each percent that the magnitude of (q - q )
t b
exceeds
+10 percent, the Delta-T trip set point shall be automatically reduced by 3.5 percent of" its value at, interim power.
4 For each percent that the magnitude of (q - q )
t b
exceeds
-14 percent~
-the Delta-T trip set point shall'be automatically'-reduced by 2 percent of.
its value at interim power.
Kl (Three Loop Operation)
~ 1..120; (Two Loop Operation)
~, 0.88 v
~
~ ~
2 ~ 3 2 6/21/74
C ~
Overpower hT
< hT 1.09 - K K (T T') f (Aq) 1-dt 2
AT
~
Indicated 5T at rated power, F
0 T
Average temperat'ure, F
T' Indicated average temperature at nominal conditions and rated power, F
K 0 for decreasing average temperature, 1
0.2 sec./F for increasing average temperature K2 0.00134 for T equal to or more than T';
0 for T less than T'ate of change of temperature F/sec dT dt 7
f(Aq)
As defined above Pressurizer Low Pressurizer pressure equal to or greater than 1715 psig.
High Pressurizer pressure equal to or less than 2385 psig.
High Pressurizer water level equal to or less than 92X of full scale.
Reactor Coolant Flow Low reactor coolant flow equal to or greater than 90% of normal indicated flow Low reactor coolant pump motor frequency - equal"to or greater than 56.1 Hz Under voltage on reactor coolant pump motor bus equal to or greater than 60X of normal voltage Steam Generators Low-low steam generator water level - equal to or greater than 5X of narrow range instrumen't scale 2 ~ 3 3
6/21/74
TABLE 3,5-2 ENGINEERED SAFETY FEATURES ACTUATION NO.
FUNCTIONAL UNIT MIN.
OPERABLE'HANNELS 2
MIN.
DEGREE OF REDUN-DANCY OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 CANNOT BE MET 1.
SAFETY INJECTION 1.1 Manual 1.2 High Containment Pressure 1.3 High Differential Pressure between any Steam Line and the Steam Line Header 1.4 Pressurizer Low Pressure and Low Level 1.5 'igh Steam Flow in 2/3 Steam Lines with Low Tavg or Low, Steam Line Pressure 0
1/line in 1
each of' lines Cold Shutdown Cold Shutdown Cold Shutdown Cold Shutdown Cold Shutdown 2.
CONTAINMENT SPRAY 2.1 High Containment Pressure and High-High Containment Pressure (Coincidant) 2 per set 1/set Cold Shutdown This signal may be manually bypassed, when the reactor is shut down and pressure is below 1800 psig Each channel has two separate signals
'/21/74
3.6 CHEMICAL AND VOLUME CONTROL SYSTEM Volume Control System.
~0b ective:
To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation.
h>>>>
r flow path to.the core for boron in)ection.
A reactor shall not be-made critical unless the following 1i Chemical.and Volume Control System conditions are met-1.
TWO associated charging pumps shall be 'operable.
2.
TWO boric acid transfer pumps shall be operable.
3.
The boric acid tanks in service shall contain a total of at.least 3,080 gallons of a 20,000 to 22,500 ppm boron solution at a temperature of at least 145 F.
4.
System piping, interlocks and valves shall be operable to the extent of establishing one flow path from the
. boric acid tanks, and one flow path from the refueling water storage tank,. to the
vz v e. <t
(.e'l 5.
TWO channels of heat.tracing shall be operahle. for the flow path from the boric acid tanks.
The'primary water storage tank contains not less than 30,000 gallons of water.
c.
The second reactor shall not be made critical unless the following conditions are met:
3.6-1 6/21/74
1; TWO,associated charging. pumps shall be operable.
2.
THREE boric acid transfer pumps shall. be operable.
3.
The boric acid tanks in service shall contain a total of at least 6160 gallons of a 20,000 to 22,500 ppm boron solution at a temperature, of at least I
145 F.
4.
System piping, interlocks and valves shall be operable to the extent of establishing one flow path from the boric acid tanks, and one flow path from the refueling water storage tank, to each Reactor Coolant System.
5.
TWO channels of heat tracing shall be operable for the flow path from the boric acid tanks.
6.
The primary water storage tank contains not less than 30,000 gallons of water.
d.
During power operation,'he requirements of 3.6.b and c may be modified,to allow one of the following components to be inoperable.
If the system is not restored to meet the requirements of 3.6b and c within the time period specified, the reactor(s) shall be placed in the hot shutdown con-dition. If the requirements'f 3,6.b and c are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor(s) shall be placed in the cold shutdown condition.
1.
One of the two operable charging pumps may be removed from service provided that it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
One boric acid transfer pump may be out of service provided that it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
One channel of heat tracing may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.6-2 6/21/74 I
5.
At least ONE residual heat removal pump shall be in operation, unless reactor coolant temperature is less than 160F.
6.
When the reactor vessel head is removed and fuel is, in the'vessel, the minimum boron'concentration of 1950 ppm shall be maintained in the reactor coolant system and verified daily.
7.
Direct communication between the control room and the refueling cavity manipulator crane shall be available'during refueling operation.
8, The'pent fuel cask shall"not be moved over spent,.
fuel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit.
9.
Fuel which has been discharged from a reactor will not be moved outside the containment in fewer than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.
If any one of the specified limiting conditions for re-fueling is not met; refueling shall cease until specified limits are met, and there shall be no operations which may increase reactivity.
3 10-2 6/21/74
TABLE 4. 1-1'Continued)
- Channel Descri tion Check Calibrate Test Remarks 10.
Rod Position Bank Counters 11.
Steam Generator Level N.A.
N.A.
Mita Analog Rod Position 12.
13.
Charging Flow Residual Heat Removal Pump Flow N.Ae N.A.
R N.A.
N.A'e 14.
Soric Acid Tank Level Refueling Water Storage Tank Level
. N.Ae Jl e
N.A.
16.
Volume Control Tank Level 1?A. Containment Pressure
. 17B. Containment Pressure 18A. Process Radiation 18B. Area Radiation 19; Boric Acid Control N.A.
NiA.
~
R N.A.
N.-A.
Wide Range Narrow Range 20.
Containment Sump Level 21.
Accumulator Level and Pressure 22.
Steam Line Pressure N.A.
N.A.
N.A.
TABLE 4.1-1 (Continued)
Channel Descri tion Check Calibrate Test Remarks 23.
Environmental Radiological Monitors N.A.
A(1)
M(1)
(1) Flow 24.
Logic Channels N.A.
N.A.
25.
Emer. Portable Survey Instruments N.A.
~ M N.A.
26.
Seismograph N.A.
Using moveable in-core detector system.
Frequency only
+** Effluent monitors only.
Calibration shall be as specified in 3.9.
Make trace Test battery (change semi-annually)
D B/W
, R A
N.A.
Ch
. tt Each Shift Daily Weekly Every Two Weeks Monthly Quarterly Prior to each startup if not done previous week Each Refueling Shutdown Annually Not applicable N,A. during cold or refueling shutdowns, The specified tests, however, will be performed within one surveillance interval prior to startup.
N.A. when the equipment hatch or both doors of either air lock are open.
4.4 CONTAINMENT TESTS I
veillance, end anchorage concrete surveillance, and liner surveillance.
~Ob ective:
To verify that potential leakage from the containment and the tendon loading are maintained within specified limits.
4.4.1 INTEGRATED LEAKAGE RATE TEST POST OPERATIONAL Post Qperational Containment Integrated Leak Rate Tests shall be performed-and reported in accordance with 10 CFR 50, Appendix J, (type A tests).
Pa, the'peak calculated containment internal pres-,
sure related to the design basis accident is 49.9 psig.
,* ee Pt, the containment vessel reduced test pressure is 25 psig.
e La, the maximum allowable leakage rate at pressure Pa is 0.25 weight percent of containment atmosphere per day.
5
'4.4-1 o/21/I4
- 4. o. c, CAL PENETRATION TESTS
~i Local penetration tests of the containment purge valves, the personnel and emergency air locks, the equipment access
- opening, the fuel transfer tube flange,'and the electrical pene-trations shall be performed in accordance with 10 CFR 5O Y Appendix J, (type B teste) with the following exceptions:
F 1.
The personnel and emergency air locks shall be tested after each use by applying a 15 inch Hg minimum vacuum between the inner door gaskets.
2.
The electrical penetrations shall be tested prior to each containment integrated leak rate test.
4.4-2 6/21/74
4.4.3 ISOLATION VALVES Containment isolation valves shall be tested in accordance with 10 CFR 50, Appendix J, (type C
tests).
'k k
mr'"
4.4.4 RESIDUAL HEAT REMOVAL SYSTEM k
~ '
0
,b.
k The portion of the Residual Heat Removal Syst'm that is downstream of the first isolation valve outside the containment shall be tested either by use in normal operation or hydrostatically tested at 350 psig at the interval specified below.
Visual inspection shall be made for excessive leakage from components of the system.
Any visual leakage
,that cannot be stopped at test cohditions shall be measured by collection and weighing or by another equivalent method.
Ci The acceptance criterion is that maximum allowable leakage from the Residual Heat Removal System components (which includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.
~ %
Repairs shall be made as required to maintain leakage with the acceptance criterion in (c) above.
k e.. If repairs are not completed within 7 days, the reactor shall be shut down and depressurised until repairs are effected and the acceptance criterion in (c) above is satisfied.
k f.
Tests of the Residual Heat Removal System shall be conducted each refueling, 4.4-3
~
6/21/74
.4.4. 5
'TENDON SURVEILLANCE Lift-off Lift-offreadings will be taken for the following nine (9) tendons'vailable for'nspection:
Unit 3 Unit 4 Horizontal 62H18, 42H70, 64H50 13H15
, 51H50, 35H70 Vertical 23V1, 45V7, 61Vl 12V29, 34V29, 56V29 Dome.1D27, 2D28
, 3D28 1D28, 2D28, 3D28 F
f Wire Ins ection One horizontal, one vertical and one -dome tendon will be relaxed and one wire.will be removed from each as a sample.
(At subsequent inspections different tendons will be used for the sample).
Wires will be visually inspected for corrosion and pitting.
Tensile tests will be performed on three (3) samples cut from each wire (one from each end and.one from Fthe middle) of a length equal to the maximum length acceptable for the test apparatus to be used.
After samples are taken, tendons will be re-tensioned and final lift-offreadings will be taken.
Test Fre uenc Lift-offreadings and wire inspection will take place at the end of the first, third and every fifth year thereafter from the date of the structural integrity test (July 4, 1971, for Unit 3 and February 19, 1972, for Unit 4).
Tendon surveil-lance may be conducted during reactor operation.
4.4-4 6/21/74
Cy Additional Surveillance 'on Unit 3 Dome On Unit 3 dome 12 tendons (including the three listed in the first paragraph under 4.4.6) will be sub)ected to surveillance testing at 6, 12, 24 and 36 months after the structural integrity test (July 4, 1971, for Unit 3).
The ad-ditional tendons are:
- 1D15, 1D18,
- 1D36, 2D24, 2Dll,
- 2D21, 3D4, 3D21 and 3D24.
Lift-offreadings will be taken on. each of the'se tendons.
The.,decrease in prestress force measured from 0.73f'.s 'A will be recorded and compared with I
the'predicted'1'oss; for the period the tendons were stressed.I'I The surveillance tendons will be stressed to 0.8f's, and the elongation recorded, the tendons will then be relaxed and observation will be made at the stressing washer for any indications of a broken wire. '"'The tendons will be re-.
tensioned and lift-offreadings taken.
Wire Ins ection One wire each will be removed from three tendons, not listed in the first paragraph of 4.4,6 (one from each directional group); wires will be visually inspected for corrosion and pitting.
Tensile tests will be performed on three (3) samples cut from each wire (one from each.
end
'nd one from the middle) of a length equal to the maximum length acceptable for the test apparatus to be used.
'After the samples are taken, the tendons will be retensioned and final lift-offreadings taken.
. 4.4-5 6/21/74
-4.4. 6
'END ANCHORAGE CONCRETE SURVEILLANCE The following end anchorages will be subject to surveillance at the 346'uttress on Unit 3 and the 194'uttress on Unit 4:
Elevations 14~ Q<<35~ P<<6P'<<85'<<
110'-0", 152'-0<<
and in the tendon inspection gallery of each unit at tendon numbers:
.12V11, 12V23, 23V9, 23V23, 34V12, 34V28, 45V14, 45V26, 46V24, 56V16, 61V9, 61V26.
The inspection intervals will be approximately one-half year and one year after the structural integrity test (July 4, 1971, for Unit 3 and Feb-ruary 19, 1972, for Unit 4) and shall be chosen such that the inspection occurs during the warmest and coldest part of the year following the test.
The inspections made sha11 include:
1.
Visual inspection of the end anchorage con-crete e'xterior surfaces.
2.
The mapping of the predominant visible con-crete crack patterns.
3.
The measurement of the crack widths, by use of optical comparators or feeler gauges.
The measurements and observations shall be com-pared with those to which prestressed structures have been sub)ected in normal and abnormal load conditions and with those of preceding measurements and observations at the same location on the structures.
4.4-6 6/21/74
( I If the'nspections determine that the conditions are favorable in comparison with experience and predictions, the close inspections will be terminated by the second inspection and a report will be prepared.
If the inspections detect symptoms of greater than normal cracking or movements, an investigation will be made to determine the cause.
- 4. 4. 7 LINER SURVEILLANCE Three representative areas of the liner plate shall be examined and measured for inward deformation'1') prior to the structural in-tegrity test, (2) after the structural integrity
/
test (3) approximately one year after the structural integrity test.
Measurements shall be taken between vertical anchors using a
straight edge to determine liner profile to within a + 0.01 inch accuracy.
If changes are less than 0;25 inches no further tests or action is required other than preparation of records.
Otherwise an investigation and cor-rective action will be taken.
Measurements locations shall be:
Elevation Unit 3 Unit 4 34'0" 62 l Oll 118'0" 70 190'26'0'90'18'hen measurements are made, liner plate and ex-terior concrete surface temperatures in the area of measurement, and inside and outside ambient temp-eratures, will be determined and recorded.
The requirements of this Technical Specification have been met.
6/21/74
- 4. 4-7,
4,5 SAFETY INJECTION
~ob ectfve:
To verify that the subject systems will respond promptly and perform their design functions.
S ecifications:
1.
SYSTEM TESTS a.
System tests shall be performed at each refueling shutdown.
The test shall be
=performed in accordance with. the follow-ing procedure:
With the Reactor Coolant System pressure equal to or less than 350 psig and temperature equal to or less than 350F, a test safety infection signal will be applied to initiate operation of the system.
The breakers for the residual heat r'emoval pump motors willbe tested either in the test position or by actually operating the associa'ted residual heat removal pump motor.
b.
The test will be considered satisfactory if con-trol panel indication and visual observations indicate that all components have received the safety in)ection signal in the proper sequence and timing, appropriate breakers shall open and
- close, and all automatic valves shall complete their travel.
2.
COMPONENT TESTS e.
~Pum e
1.
The safety infection pumps and residual heat removal pumps shall be started at intervals not greater than one month.
4.5-1 6/21/74
2.
COMPONENT TESTS Pum's and Fans The containment spray pumps. and the Emergency Contain-ment Cooling fans shall be started at intervals not greater than one (1) month.
Acceptable levels of performance shall be that the pumps
'I reach their rated shut off heads, the fan motors reach
'heir nominal operating current for the containment at-mosphere during the test, and both operate for at least fifteen minutes'.
Valves The systems motor operated isolation valves will be tested for operation during syst m tests.
4.6-2 6/21/74
4.7 EMERGENCY CONTAINMENT FILTERING AND POST ACCIDENT CONTAINMENT VENT SYSTEMS Applies to the Emergency.Containment Filtering and the Post Accident Containment Vent System components.
To verify that these systems and components will be able to perform their design functions.
4,7,1 EMERGENCY CONTAINMENT FILTERING SYSTEM 1.
OPERATING TESTS.-
I'ystem tests shall be performed at approximately quarterly intervals.
These tests shall consist of visual inspection and pressure drop measurements across each filter bank.
Visual inspection shall
~include inspection of general condition for evidence of:
water, oil, or other foreign material; gasket deterioration; adhesive deterioration in the HEPA units; exces-sive dust cake on the demisters; and unusual or ex-cessive noise or vibration when the fan motor is running.
Pressure drop across any filter bank shall not 'exceed two times the pressure drop when new and shall not be less than the pressure drop when new,.
2.
PERFORMANCE TESTS During each refueling operation, "in-place" DOP and freon tests shall be conducted at design flow on each unit (all flow paths).
99.9X DOP removal and 99.5X freon removal shall constitute acceptable performance.
4.7-1 6/21/74
4.8 Cg t
EMERGENCY POWER SYSTEM PERIODIC TESTS Applies to periodic testing and surveillance require-ments of the emergency power system.,
~Ob ective:
To verify that the emergency power system will respond promptly and properly.
The following tests and surveillance shall be performed as stated:
le DIESEL GENERATORS a.
Each diesel generator shall be manually started and synchronized with normal power sources and loaded to 2750 KW monthly.
b.
Each diesel generator shall be started auto-matically by a simulated loss of all normal A-C power supplies together with a simulated safety injection signal and loaded sequentially with vital loads during each refueling shutdown.
Each diesel shall start and assume loads in 'the time sequence stated in FSAR Table 8.2-3.'he safety injection pumps will be operated using the test lines.
c.
Each diesel generator shall be given a thorough inspection at least annually following the man-ufacturer's recommendations for this class of stand-by'ervice'.
d.
The above tests will be considered satisfactory
if all applicable equipment operates as designed.
e.
Diesel generator electric loads shall not be in-creased beyond.2850 KW during a test.
The 4.8-1 6/21/74
'onnected loads shall not be increased above those listed in FSAR Table 8.2-2 during the test in l.b. above.
f.
The diesel fuel oil transfer pumps shall be tested monthly.
2.
STATION BATTERIES a.
Pilot cell specific gravities. shall.Ee read and.recorded daily.
The pilot cell shall be rotated on a monthly basis.
b.
Monthly each battery shall he'.,given'n equalizing charge, and afterwards specific gravity and voltage readings shall be taken and recorded for each cell.
Water shall be added to restore normal level and total water use shall be recorded.
Complete visual in-spection of batteries shall be made monthly.
c.
Quarterly detailed visual inspection shall be made of chargers.
d.
Annually connections shall be checked for tightness and anti-corrosion coating shall be applied to interconnections.
e.
Perform load test annually.
4.8-2 6/21/74
AUXILIARYFEEDVATER SYSTEM Applies to periodic testing requirements of the auxiliary feedwater system.
~Ob ective:
To verify the operability of the auxiliary feedwater system and its ability to respond properly when required.
S ecifications:
P l.
Each turbine-driven auxiliary feedwater pump shall be started at intervals not greater than one month,.
run for 15 minutes and a flow rate of 600 gpm es-tablished to the steam generators.
The monthly frequency is not intended to require the test while at cold shutdown".
The testing requirement is met by performing this test during startup subsequent to cold shutdown.
2.
The auxiliary feedwater discharge valves shall be tested by operator action during pump tests; 3.
Steam supply and turbine pressure valves shall be tested during pump tests.
4.
These tests shall be considered satisfactory if control panel indication and visual observation of the equipment demonstrate that all components have operated prope'rly.
4.10-1 6/21/74
The f(~q) function in the Overpower hT and Overtemperature 5T protection system setpoints includes effects of fuel densification on core safety limits.
The setpoints will ensure that the safety limit of centerline fuel melt will not be reached and DNBR of 1.30 will not be violated.
(10)
Pressurizer The low pressurizer pressure reactor trip trips the reactor in (6) the unlikely event of a'oss-of-coolant accident.
The high pressurizer pressure, reactor trip is-set below the set pressure of the pressurizer safety valves and limits the reactor operating pressure iange'.
The high pressurizer water level reactor trip protects'the pressurizer safety valves against water relief.
The specified set point allows margin (3) for instrument error and transient level overshoot before the reactor trips.
Reactor Coolant Flow The low flow reactor trip protects the core against DNB in the event of loss of one or mqre reactor coolant pumps.
The set point specified is consistent with the value used in the ac-(7) cident analysis.
The low frequency and under voltage reactor tiips protect against a decrease in flow.
The specified s'et points assure a react'or trip signal before the low flow trip point is reached.
The underfrequency trip set point preserves the coastdown energy of the reactor coolant pumps, in case of a system frequency decrease, so DNB does not occur.
The undervoltage.
trip set point will cause
'a trip before the peak motor torque falls below 100X of rated torque.'team Generators The low-low steam generator ~ater level reactor trip assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting of the auxiliary feed-(s) water'ystem.
B2 ~ 3-2 6/21/74
A