ML20245F903
| ML20245F903 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/25/1989 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245F906 | List: |
| References | |
| NUDOCS 8906280287 | |
| Download: ML20245F903 (14) | |
Text
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UNITED STATES
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e...e APKANSAS F0WER AND LIGHT COMPANY DCCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT I AMENDMEfiT TO FACILITY OPERATIllG LICENSE Amendment No.121 License No. OPR.51 1.
The fluclear Regulatory Comission (the Comission) has four.d that:
A.
The application for amendment by Arkansas Power and Light Company (thelicensec)datedDecember 12, 1986, complies with the standards ar.d requirements of the Atcmic Energy Act of 1954, as amended (the Act), and the Corriission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will orcrate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance: (1)thattheactivitiesauthorized by this amendment can be conducted withcut endangering the health and safety of the public, and (ii) that such activities will be conducted in compliarce with the Comission's regulations; D.
The issuar.ce of this licerse amendment will nct be inimical to the concon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2.
. Accordingly, the license is arended by changes to the, Technical 7'
Specificaticos 'as indicated in the attachmer:t to this license amendment, and Paragraph 2.C.(2) of facility Operating License No.
DPR-51 is hereby amended to read as follows:
(2) Technical Specifications The' Technical Specifications contained in Appendices A and B, as revised thrcugh Amendment No.121, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSIM JD A
Freder J. Hebdon, irector Project Directorate - IV' Division cf Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 25, 1989
4 ATTACHMENT TO LICENSE AtiEf1Dl4ENT f10.121-FACILITY OPERATING LICENSE NO. DPR-51 DOCKCT NO. 50-313 Revise'the following pages of the Appendix "A" Technical Specifications
- with the attached pages. The revised pages are identified by Amendment number 'and contain vertical lines. indicating' the area of. change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE PAGES INSERT PAGES 36 36 37 37 74~
74 75a 79 79 80 80 82-82 105 105 106 106 110t 110t 110z 1102 t'
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1 9
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A 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING AND REACTOR BUILDING SPRAY SYSTEMS Applicability Applies to the emergency core cooling, reactor building cooling and reactor building spray systims.
Objectivity To define the conditions necessory to assure immediate availability of the emergency core cooling, reactor building cooling and reactor building spray systems.
Specification 3.3.1 The following equipment shall be operable whenever containment integrity is established as required by Specification 3.6.1:
(A) One reactor building spray pump and its associated spray nozzle header.
(B) One reactor building cooling fan and itt inciated cooling unit.
(C) Two out of three service water pumps shall be operable, powered from independent essential buses, to provide redundant and independent flow paths.
(D) Two engineered safety feature actuated low pressure injection pumps shall be operable.
(E) Both low pressure injection coolers and their cooling water supplies shall be operable.
(F) Two BWST level instrument channels shall be operable.
(G) Theboratedwaterstoragetankshallcontainalevelof37.5[f.5 ft.
(362,000 2 13,000 gallons) of water having a concentration of 2470 2 200 ppm boron at a temperature not less than 40F.
The manual valve on the discharge line from the borated water storage tank shall be locked open.
(H) The four reactor building emergency sump isolation valves to the LPI system shall be either manually or remote-manually operable.
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l Amendment No. 25, 25, 121 36
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(I) The angineered safety features valves associated with each of the above systems shall be operable or locked in the ES position.
3.3.2 In addition to 3.3.1 above, the following ECC5 equipment shall be operable when the reactor coolant syste : is above 350F and irradiated fuel is in the core:
(A) Two out of three high pressure injection (makeup) pumps shall be maintained operable, powered from independent essential buses, to provide redundant and independent flow paths.
(B) Engineered safety features valves associated with 3.3.2.a above shall be operable or locked in the ES position.
3.3.3 In addition to 3.3.1 and 3.3.2 above, the following ECCS equipment shall be operable when the reactor coolant system is above 800 psig:
(A) The two core flooding tanks shall each contain an indicated minimum of 13 2 0.4 feet (1040 2 30 ft3) of borated water at 600 2 25 psig.
(B) Core flooding tank boron concentration shall not be less than 2270 ppm boron.
(C) The electrically operated discharge valves from the core flood tanks shall be open and breakers locked open and tagged.
(D) One of the two p,ressure instrument channels and one of the two level instrument channels per core flood tank shall be operable.
3.3.4 The reactor shall not be made critical unless the following equipment in addition to 3.3.1, 3.3.2, and 3.3.3 above is operable.
(A) Two reactor building spray pumps and their associated spray nozzle headers and two reactor building emergency cooling fans and associated cooling units powered from oparable independent emergency buses.
(B) The sodium hydroxide tank shall contain an indicated 34[
ft.of18[
wt % solution sodium hydroxide (19,500 lb. 2 2500 lb.).
(C) All manual valves in the main discharge lines of the sodium hydroxide tanks shall be locked open.
Amendment No. 25, 25, 121 37
Ie Table 4.1-3 MINIMUM SAMPLING AND ANALYSfS' FREQUENCY Item Test Frecuency 1.' Reactor Coolant
- a. Gamma Isotopic Analysis
- a. Bi-weekly (7)-
Samples
- b. Gross Activity Determination
- b. 3 times / week and at least every-third day (1)(6)(7)
- c. Gross Radiciodine
- c. Weekly (3)(6)(7)
Determination
- d. Dissolved Gases
- d. Weekly (7)'
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- e. Chemistry (C1, F, and 0 )
- e. 3 times / week (8) 2
- f. Boron Concentration
- f. 3 times / week
- g. Radiochemical Analysis for
- g. Monthly (7) determination (2)(4) l
- 2. Borated Water Boron Concentration Weekly and after Storage Tank Water each makeup Sample
- 3. Core Flooding Tank Boron Concentration Monthly and after Sample each makeup
- 4. Spent Fuel Pool Boron Concentration Monthly and after Water Sample each makeup (9)
- 5. Secondary Coolant
- a. Gross Radiciodine
- a. Weekly (5)(7)(10)
Samples Concentration
- b. Isotopic Radiofodine
- b. Monthly (7)(10)
Concentration (4)
- 6. Sodium Hydroxide Sodium Hydroxide Quarterly and after Tank Sample Concentration each makeup Notesi (1)
A gross radioactivity analysis shall co' Ost of the quantitative measurement of the total radioactive *),, the primary coolant in units of pCi/gm. The total primary coolant activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities 15 minutes after the primary system is sampled. Whenever the gross radioactivity concentration exceeds ID% of the limit specified in the Specification 3.1.4.1 or increases by 10 pCi/gm from the previous mcasured level, the frequency of sampling and analyzing shall be increased to a minimum of once/ day until a steady activity level is established.
Amendment No. If, 25, 121..
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4.,4 REACTOR BUILDING 4.4.1-Reactor Buildino Leakace Tests
' Applicability I
Applies to the reactor building.
Objective To verify that leakage from the reactor building is maintained within allowable limits.
' Specification 4.4.1.1' Integrated Leakage Rate Tests 4.4.1.1.1-Design' Pressure Leakage Rate The maximum allowable integrated leakage rate, L, from the reactor building at the 59 psig design pressure, P, shail not exceed 0.20 weightpercentofthebuildingatmosphereEtthatpressureper24 hours.
4.4.1.1.2 Testing at Reduced Pressure The periodic integrated leak rate test may be performed at a test-pressure, P, of 30 psig provided the resultant leakage rate, L,
doesnotexbeedapre-establishedfractionofL,determinedas g
follows:
Prior to reactor operation the initial value of the integrated a.
leakage rate of the reactor building shall be measured at design pressure and at the reduced pressure to be used in the periodic integrated leakage rate tests.
The leakage rates thus measured.
shall be identified as L,,
and L respectively.
tm b.
L shall not exceed L' L
for values of L bklow0.7 a tm tm
'am
'am L, shall not exceed L, P for values of L c.
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a5cve 0.7 p
t a
am d.
If L /L is less than 0.3, the initial integrated test results shalNbe,IubjecttoreviewbytheNRCtoestablishanacceptable j
value of L
- t 79 Amendment No. 121
4 Whare (L,)
D sign B2 sis Accidant Leakcge Rate at Prassure P, (L )
Maximum Allowable Test Leakage' Rate at t
Reduced Test Pressure P Under Test t
Condition (L,,)
Maximum allowable operational leakage rate at pressure P, (Lto)
Maximum allowable leakage rate at pressure Pt (L,,)
Initial Measured Leakage Rate at Pressure a
(Ltm)
Initial Measured Leakage Rate at Pressure t
(P,)
Peak Test Pressure of 59 psig (P )
Reduced Test Pressure of 30 psig t
4.4.1.1.3 Conduct of Tests a.
Leakage rate tests should not be started until essential temperature equilibrium has been attained. Containment test conditions should stabilize for a period of about four hours prior to the start of a leakage rate test.
b.
The leakage rate test period shall extend to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of retained internal pressure.
If it can be demonstrated to the satisfaction of.those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test period, the agreed upon shorter period may be used.
c.
Test accuracy shall be verified by supplementary means, such as measuring the quantity of air required to return to the starting point or.by imposing a known leak rate to demonstrate the validity of measurements.
d.
Closure of reactor building isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves without preliminary exercised or adjustment.
4.4.1.1.4 Frequency of Test After the initial preoperational leakage rate test, a set of three integrated leak rate tests shall be performed at approximately l
equal intervals during each 10 year service period, with the third test of each set coinciding with the and of each 10 year service i
period.
The test may coincide with the plant inservice inspection l
shut down periods.
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4 4.1.1.5'
_ Conditions for Return to Criticality If L*I is less than L (L,
75% L )
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or
'I f L is less than L (L,,"I 75% L,)
4.4.1.1.6 Corrective Action Retest If L is greater than L local leak tests will then be perfbYmedandtherequirI8,repairsmade. The integrated leakage test need not be repeated provided local measured leakage reduction achieved by repairs of individual leaks reduces the reactor building's overall measured leakage rate sufficiently such that L,is less than L,.
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4.4.1.1.7 Report of Test Results The initial test report shall include a schematic arrangement of the leakage rate measurement system, the instrumentation used, the supplemental test method and the test program selected as applicable to the initial test and all subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test data to the extent necessary to demonstrate the acceptability of the reactor buildings leakage rate in meeting the acceptance criteria.
4.4.1.2 Local Leakage Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for components in the following categories:'
Reactor building penetrations whose design a.
incorporates resilient seals, gaskets, or sealant compounds; piping penetrations fitte with expansion bellows.
b.
Air lock door seals, including operating mechanism and penetrations' with resilient seals which are part of the reactor building pressure boundary in the air lock structures.
Equipment and access doors with resilient seals or c.
gaskets (seal-welded doors are excluded).
d.
Components other than those listed in items a,b, and c abovs which develop leaks inservice and 81 L_____----.-_---
e
.. require repairs to meet the acceptance criterion of specification 4.4.1.1.5.
e.
Reactor building isolation valves which' provide a direct.
connection with the inside atmosphere of the reactor building.
f.
Reactor building isolation valves which in the event of valve leakage or-valve malfunction upon a reactor building -
l
. isolation signal. may extend (outside of the reactor building) the boundary of the leakage-limiting barrier of the reactor primary containment beyond that included during the conduct of the tests required by specification 4.4.1.1
- m (includes instrument valves in lines connected to the reactor i
u coolant pressure boundary).
g.
Reactor building isolation valves in engineered safety systems penetrating the reactor building which, under post-accident conditions, are required to close following the termination of the safety function.
4.4.1.2.2 Conduct of Tests a.
Local leak rate tests shall be performed at a pressure of 59 psig. '
b.
Acceptable methods of testing are halogen gas detection, soap bubbles, pressure decay, hydrostatic flow or equivalent.
4.4.1.2.3 Acceptance Criteria The total leakage from all tested penetrations and isolation valves shall not exceed 60% L,.
4.4.1.2.4 Corrective Action a.
If at any time during operation it is determined that specification 4.4.1.2.3 is exceeded, repairs shall be initiated immediately.
b.
If conformance with specification 4.4.1.2.3 is not demonstrated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following detection of excessive local leakage, the reactor shall be shutdown and placed in a condition such that i
reactor building integrity is not required.
(Specification 3.6.1) 82 Amendment No.121 A
8, i
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.,4.8-EMERGEN'CY FEEDWATER PUMP TESTING l
Applicability Applies to the periodic testing of the turbine and electric motor driven emergency feedwater pumps.
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Objective i
k To verify that the emergency feedwater pump and associated valves are operable.
Specification 4.8.1 Each EFW train shall be demonstrated operable:
a)
By verifying on a STAGGERED TEST BASIS:
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1.
at least once per 31 days or upon achieving hot shutdown following a plant heatup and prior to criticality, that the turbine-driven pump starts, operates for a minimum 1 1200 psig at a flow of 1 500 gpm through the test loop flow path.
2.
at least once per 31 days by verifying that the motor driven EFW pump starts, operates for a minimum of 5 minutes and develops a discharge pressure of 1 1200 psig at a flow of 1 500 gpm thorough the test loop. flow path, b)
At least once per 31 days by verifying that each valve (manual, power ' operated or automatic) in each EFW flowpath that is not locked, sealed, or otherwise secured in position, is in its correct position.
c)
Prior to exceeding 280'F reactor coolant temperature and after any EFW flowpath manual valve alterations by verifying that nach manual valve in each EFW flowpath which, if mispositioned may degrade EFW operation, is locked in its correct position, d)
At least once per 92 days by cycling each motor-operated valve in each flowpath through at least one complete cycle.
e)
At least once per 18 months by functionally testing each EFW train and:
1)
Verifying that each automatic valve in each flowpath actuates automatically to its correct position on receipt of an actuation signal, l
Amendment No. 25, 59, 51, 121 105
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4.9 REACTIVITY ANOMALIES Applicability Applies to potential reactivity anomalies.
Objective To require the evaluation of reactivity anomalies.of a specified magnitude occurring during the operation of the unit.
Specification Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be periodically compared with the predicted value.
If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one percent in reactivity, an evaluation of this abnormal occurrance will be made to determine the cause of the discrepancy.
Bases To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point.
As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve l
relating burnup and reactivity is compared with that predicted.
This process of normalization should be completed after about 10 percent cf the total core burnup.
Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1 percent Ak/k would be unexpected, and its occurrence would be thoroughly investigated and evaluated.
The value of 1 percent Ak/k is considered a safe limit since a shutdown margin of at least 1 percent Ak/k with the taost reactive rod in the fully withdrawn position is always maintained.
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Amendment No. 25, 121 106 4
b
'4.21 SPR2NKLER SYSTEMS l
1 Applicability s
lf Applies to surveillance of the sprinkler' systems which are required to be operable by Specification 3.18.
Objective To assure that the various sprinkler systems are available and operable when needed.
Specification n
,4.21.1 The cable spreeding room sprinkler system shall be demonstrated operable at least once per 31 days by verifying that the system is aligned to the fire pumps.
4.21.2 The sprinkler systems located in the four reactor building cable penetration areas and four cable penetration rooms shall be demonstrated operable:
a.
At least once per 31 days by verifying that each system is aligned to the fire pumps; b.
At least once per 12 months by cycling each testable valve in the flow paths through at least one complete cycle of full travel.
4.21.3 The sprinkler systems located in the two emergency diesel generator rooms and two diesel generator fuel vaults shall be demonstrated operable:
a.
At least once per 31 days by verifying that each system is aligned to the fire pumps; b.
At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel; c.
At least once per 18 months by inspection'of the spray headers to verify their. integrity.
Bases The cable spreading room sprinkler system is a wet pipe system actuated only by heat action on the fusible head sprinkler. The only maintainable aspect of the system is the verification of system alignment.
The sprinkler systems in the four cable penetration areas and four cable penetration rooms are manually-operated closed-head (fusible head) systems which are remotely operated from the control room.
The sprinkler systems in the two emergency diesel generato= rooms and two diesel generator fuel vaults are manually-operated open-head systems which are remotely operated from the control room.
The required inspections will assure availability of the various sprinkler systems when they are needed.
Amendment No. 35, 121 110t
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I 4.26 REACTOR BUILDING PURGE VALVES Applicability This specification applies to the reactor building purge supply and exhaust isolation valves.
.i Obj,ecti ve To assure reactor building integrity.
Specification 4.26.1 The reactor building purge supply and exhaust isolation valves shall be determined closed at least once per 31 days when containment integrity is required by TS 3.6.1.
4.26.2 Prior to exceeding conditions which require establishment of reactor building integrity per TS 3.6.1, the leak rate of the purge supply and exhaust isolation valves shall be verified to be within acceptable limits per TS 4.4.1, unless the test has been successfully completed within the last three months.
Bases Determination of reactor building purge valvo closure will ensure that reactor building integrity is not unintentionally breached.
l As a result of Generic Issue B-20, " Containment Leakage Due to Seal Deterioration," it was concluded that excess leakage past valve resilient seals is typically caused by severe environmental conditions and/or wear due to use.
Recommended leak test frequencies of three months are deemed to be adequate to detect seal degradation of resilient seals.
The three month test need not be conducted with the precision of the Type C 10CFR50, Appendix J criteria, however the test must be sufficient to detect degradation.
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Amendment No. 55, 75, 121 110Z
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SAFETY EVALL'ATirt: PY THE OFFICE OF NUCLEAP PEACTOP FEGULAT:Of:
RELATED TO At'Ef DidENT NO.121 TO FACILITY OPERATING LICENSE NO. DPR-51 ARKAliSAS POWER At:0 LIGHT COMPANY
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ARKANSAS NUCLEAR ONE, UNIT NO. 1 DOCKET NO. 50-313 INTRODUCTION By letter dated December 12, 1986, Arkansas Power and Light Corpany (AP&L or the licensee) reques:2d an amendment to the Technical Specifications (TSs) appended to Tacility Operating License Nc. DPR-51 for Arkansas ruclear One, Unit 1(Ah0-1). The amendraent removes the text of several temporary specifi-cations which are no longer applicable, regarding gross iodine determination, the sodium thiosulfate system, and the borated water storage tank. The amendment also makes seteral changes to correct typographical errors, and revise wording to provide eersistent terminology.
EVALUATION The staff has reviewed the set of changes requested by the licensee and has determined that they are purely administrative in nature and result in irpreving clarity of information in the Technical Specifications. The changes do not decrease or otherwise modify existing requirements. Therefore, they are acceptable. A number of typo so no changes were necessary. graphical errors were fuund to no longer exist These were on pages 11, 7, 58, 61, 66c, 79 (TS 4.4.1.1.2.a), and 81.
ENVIRCf;t' ENTAL CONSIDERATION The amendment related to changes in recordkeeping, reporting, or administrative i
procedures or requirements. Accordingly, the amendment meets the eli ibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(g)(10).
c Pursuant to 10 CFR 51.22(b), no environmental impact statement or environ-i mental assessment need be prepared in connection with the issuance of the amendment.
CONCLUSION The staff has concluded, based on the considerations discussed above, that:
(1) there is reasor,able assurance that the health and safety of the will not be endargered by operation ir the proposed manner, and (2) public such ectivities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical tc the common defense and security or to the health and safety of the public.
Date: May 25, 1989 Principal Contributor:
C. Craig Harbuck AJ4fa GUIS ff
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