ML19326C379
| ML19326C379 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/09/1975 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19326C376 | List: |
| References | |
| NUDOCS 8004220918 | |
| Download: ML19326C379 (36) | |
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- 1 ARKANSAS POER AND LIGirr COMPANY
...: q DOCKET NO. 50-313 a
is q ARKANSAS NUCLEAR ONE - UNIT 1 ii n
AMENDMENT TO FACILITY OPERATING LICENSE
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- Amendment No. 2
- .;Jf.I License No. DPR-51
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The Nucicar Regulatory Comission (the Comission) has found that:
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A.
The application for amendment by Arkansas Power and Light iE
_.9 Cocpany (the licensee) dated January 17, 1975, and March 28, 1975,
--l complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's g.
rules and regulations set forth in 10 CFR Chapter I; i
B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendcient can be conducted without endangering the health and safety of the public, and (ii) that such activities will Ic conducted in compliance with the Comission's h
' regulations; and n
p D.
The issuance of this amendment will not be inimical to the cc:=on defense and security or to the health and safety of A
the public.
e "A 2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachnent to this license amendnent and Paragraph 2.c(2) of Facility 1.icense No. DPR-51 is hereby acended to' read as follows:
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"(2) Technical Specifications The Technical Specifications contained in Appendices
[U A and B, as revised, are hereby incorporated in the
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license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised p~i by issued changes thereto through Change No. 2."
3.
This license a:nendment is effective thirty days after the date Ef its issuance.
FOR T}E PUCLEAR REGULATORY CGf!ISSION y.;.
Dennis L. Zie._ nn, Chief h.
. Operating Reactors Branch #2 Division of Reactor Licensing
Attachment:
Change No. 2 to the Technical Specifications Date of Issuance: MAY 0 9 N
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NITACit"ENT TO LICENSE AMEDMEhT NO. 2_
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CilANGE NO. 2 TO Tile TECHNICAL SPECIFICATIONS _
L:i FACILITY OPERATING LICENSE NO. DPR-51 j.;
DOCKET No. 50-313_
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13, 14, 15, 16, 19, 20, 23, 24, 37, 38, 39, 48, 4Sc, 4Sf, Dalete pages 60, 66, 67a, 68, 71, 72, 73, 73a, 74, 75, 76, 83, 84,100a from the Appendix A Technical Specifications and insert the attached replacement pages 13, 14, 15, 16, 19, 20, 23, 24, 37, 33, 39, 42a, 48, 4Se, zis The 68, 71, 72, 73, 73a, 74, 75, 76, 83, 84 ;- 100a.
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j 48f, 60, 66, 67E, changed areas on the revised pages are shown by a marginal lino.
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- i, The low pressure (1800 psig) and variable low pressure (16.25Tout
-7756) trip setpoint.shown in Figure 2.3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure redrction. (2,3)
Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of i
(16.25 Tout-7796).
D.
Coolant outlet temperature The high reactor coolant outict te'mperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent ex-cessive core coolant temperatures in the operating range.
Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.
E.
Reactor building pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
F.
Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system.
The reactor protection. system segments which can be bypassed are shown in Table 2.3-1.
'nvo conditions are imposed when the bypass is used:
1.
A nuclear overpower trip set point of <5.0 percent of rated power is automatically imposed during reactor shutdown.
2 2.
A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed.
'Ihc purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed.
1his high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be trip' ped before the bypass is initiated.
1hc overpower trip set point of <5.0 percent prevents any significant reactor power from being produced when performing the physics tests.
Sufficient natural circulation (5) would be availabic to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.
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REFERDICES
.(1).,FSAR, Section 1h.1.2.3 (2) FSAR, Section 14.1.2.2 -
-(3) FSAR, Section 14.1.2 7 1
(h)' FSAR, Section'14.1.2.8 (5)' FSAR, Section-14.1.2.6
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Tabic 2.3-1 Reactor Prettetien System Trio S2tting Limits One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in E.ich Loop Operat Wg (Nominal Operating (Nominal (Nominal Operating Shutdown -
Operating Power - 100t)
Operating Pawcr - 75t)
Power - 49*.)
Bypas s 1.
Nuclear power, % of 105.5 105.5 105.5 S.0 (3) rsted, max 2.
Nuclear power based on 1.07 times flow minus 1.07 times flow minus 1.07 times flow minus Bypassed flew (2) and imbalance, reduction due to reduction due to reduction due to J cf rated, max irbalance(s) imbalance (s) imbalance (s)
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..uclear power based on NA NA 55%
pump monitors. % of Bypassed rzted, rax (4) 4 Iligh reactor coolant 2355 2355 2355 1720(3) system pressure, psig, i
Max 5
Low reactor coolant sys-1800' 1800 tem pressure, psig, min 1800 Bypassed 6.
Variable low reactor (16.25 Tout-7756)(I)
(16.25 Tout-7756)(1)
- (16.25 Tout-7756)(1)
Bypassed coolant system pressure, psig, min 7.
Reactor coolant temp, 619 619 619 619 F, max 8.
.gh reactor building 4 (18.1 psia) 4(18.7 psis) 4 (13.7 psia) 4(18.7 ps{
prsssure, psig, max (1) T is in degrees Fahrenheit (F).
(3) Automatically set when other segments of the RPS (as specified) are bypassed.
out (2) Reactor coolant system flow, %.
(,4) The pump monitors also produc6 a trip on: (a) loss of two reactor coolant U
pu=ps in one reactor coolant loop, and (b) loss of one or two reactor coolant 2
pumps during two-pump operation, e
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3.
LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTE4 Applicability
' Applies to the operating status of the reactor coolant system.
1 Objective To specify those limiting conditions for operation of the reactor coolant sys-tem which must be met to ensure safe reactor operations.
3.1.1 Operational Components Specification 3.1.1.1 Reactor Coolant Pumps A.
Pump combinations permissible for given power levels shall be as shown in Table 2 3-1.
B.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat re= oval pump is circulating reactor coolant.
3.1.1.2 Steam Generator A.
One steam generator shall be operable whenever the reactor coolant average temperature is above 280 F.
3.1.1.3 Pressurizer Safety Valves A.
The reactor shall not remain critical unless both pressurizer code safety valves are operabic.
B.
When the reactor is suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accord-ance with ASME Boiler and Pressure Vessel Code,Section III.
Bases A reactor cc.lant pump or decay heat removal pump is required to be in opera-tion before tne boron conce..tration is reduced by dilution with makeup water.
Either pump vill provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reachinc; the reactor.
One decay heat removal pump vill circulat9 the equivalent of the reactor coolant system volume in one half hour or less 11) 1 16
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j loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations.
The_ number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR.
The maximum unit.heatup and cooldown rate of 100 F por hour satisfies stress limits for cyclic opera-tion. (2) The 200 psig pressure. limit for the secondary side of the steam generator at a temperature less than 100 F satisfies stress levels for tem-
'a peratures below the DTr. (3)
The plate material and welds in the core region of the reactor vessel have been tested to verify conformity to specified re-quirements and a maximum NDTP value of 10 F has been determined based on Charpy V-notch tests.
The maximum NDTT value obtained for the steam generator shell material and welds was 40 F.
Figures 3.1.2-1 and 3.1.2-2 contain the limiting reactor coolant system pressure-temperature relationship for operation at DTT(4) and below to assure that. stress levels are low enough to preclude brittle fracture.* These stress levels and their bases are defined in Section 4.3.3 of the FSAR.
As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated nuclear operation.
The predicted maximum NDTT increase for the 40-year exposure is shown on Figure 4-10. (4)
The actual shift in NDTT will be determined periodically during plant operation by testing ef irradiated vessol material samples located in this reactor vessel. (5). The results of the irradiated sampic testing will be evaluated and compared to the design curve (Figure 4-11 of FSAR) being used to predict the incr' ase in transition temperature.
e The design yglue for fast neutron (E > 1 Mov) exposure of the reactor vessel is 3.0 x 10 n/cm2sec at 2568 MWt rated power and an integrated exposure of 3.0 x 1019 n/cm2 for 40 years operation. (6)' The calculated maximum values are 2.*2 x 1010 n/cm2sce and 2.2 x 1019 n/cm2 integrated exposure for 40 years operation at 80 percent load. (4)
Figure 3.1.2-1 is based on the design value which is considerably higher than the calculated value. The DTT value for Figure 3.1.2-1 is based on the projected NDTT at the end of th'e first two years of operation.
During these two years, the energy output has been con-servatively estimated to be 1.7 x 106 thermal megawatt days which is equiva-Icnt to 655 days at 2568 MWt core power.
The projected fast neutron exposure of the reactor vessel for the two years is 1.7 x 1018 n/cm2 which is based on the 1.7 x los thermal megawatt days and the design value for fast neutron exposure.
The actual shift in NDTT will be established periodically during plant opera-tion by testing vessel material samples which are irradiated cumulatively by securing them near the inside wall of the vessel in the core area.
To com-pensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown.
The NDTT shift and the magnitude of the thermal and pressure stresses are sen-sitive to integrated reactor power and not to instantaneous power level.
Figures 3.1.2-1 and 3.1.2-2 are applicable to reactor core thermal ratings up to 2568 MWt.
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The prcssure limit line on Figure 3.1.2-1 has beca selected such that ihe reactor vessel stress rcsulting from internal pressure will not exceed 15 percent yield strength considering the following:
A.
A 25 psi error in measured pressure.
System pressure is measured in either loop.
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C.
Maximum differential pressure between the point of sysEem pressure measurement and reactor vessel inlet for all operating pump combinations.
For adequate conservatism, in lieu of portions of the Fracture Thoughness Testing Requirements of the proposed Appendix G to 10 CFR 50, a maximum pressure of 550 psig and a maximum heatup rate of 50 F/hr has been imposed below 275 P as shown on Figure 3.1.2-1.
The spray temperature difference restriction based on a stress analysis of the spray line nozzle is imposed to maintain the thermal stresses at the pressuri-zer spray line nozzle below the design limit.
Temperature requirements for the steam generator correspond with the measured NDTT for the shell.
The heatup and cooldown rates stated in this specification are intended as the naximum changes in temperature in one direction in a one hour period.
The actual temperature linear ramp rate may exceed the stated limits for a time period provided that the maximum total temperature difference 2
does not exceed the limit and that a temperature hold is observed to prevent the tctal temperature difference from exceeding the limit for the one hour period.
REFERENCES (1)
FSAR, Section 4.3.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3)
FSAR, Section 4.3.1C.5 (4)
USAR, Section 4.3.3 (5)
FSAR, Section 4.4.5 (6)
FSAR, Sections 4.1.2.8 and-4.3.3 20
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- 5.1.4.ReactorCoolantSystem' Activity Specification 3.1.4.1 Whenever the reactor is operating under steady-state conditions,
.the following conditions shall be met.
- a.. The total specific activity of the primary coolant shall not exceed 72/E pCi/gm where E is the sum of the average beta energy and
-average gamma energy per disintegration in MEV/ disintegration.
b.
The I-131 dose equivalent of the radioiodine activity in the primary coolant shall not exceed'3.5 pCi/gm.
c.
If the radioactivity in the primary coolant exceeds the limits given above, corrective action shall be taken immediately to return the coolant activity to within these specifications.
If the specific activity limits given above cannot be achieved within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be brought to a hot shutdown condition using normal operating procedures..If the coolant radioactivity is not reduced to acceptable limits 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 shall be brought to a cold shutdown condition and the cause of the out-of-specification operation ascertained.
. Bases Rupture of a steam generator tube would allow primary coolant activity to enter the secondary coolant. The major portion of this activity is noble gases and would be released to the atmosphere from the condensor vacuum pump or a relief 2
valve.
Activity would continue to be released unt,il the operator could reduce Se primary system pressure below the setpoint of the secondary relief valves and could isolate the faulty steam generator.
The worst credible set of circum-stances is considered to be a double-ended break of a single steam generator tube, followed by isolation of the faulty steam generator within 34 minutes after the tube break.
Assuming the full differential pressure across the steam gener:rtor, no more than one-quarter of the total primary coolant. could be released to the secondary coolant in this period.
The decay heat during this period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for pressure reduction will ge erate steam in the secondary system representing less than 15 wei;ht percent of the secondary system.
The parameters assumed in the dose analysis for the single steam generator tube failure included the following values:
5 1) total primary coolant volume (mass) = 5.2 x 101bs.
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- 2) total secondary coolant volume (mass) = 2 x 101bs.
3) leakage rate from primary to secondary system = 1 gpn.
8 4) fission product decay heat energy for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> = 1.56 x 10 BTU.
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5) steam mass released to environs = 2.84 x 10 1bs,
- 6) primary coolant released to secondary (34 minutes) = 8.7 x 1041bs.
- 7) minimum primary to secondary iodine equilibrium activity ratio =
20 to 1 (for 1 gpm leakage).
8) specific I-131 dose equivalent activity = 3.5 pCi/gm (Primary)
= 0.17 pCi/gm (Secondary).
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9) gross specific activity in primary = 52[E pCi/gm.
10)
X/Q = 7.0 x 10-4 sec/m at limiting point beyond site boundary 3
of 1046 meters for 30 m release height - equivalent to ground Icvel release due to topography including building wake effect for 5 percentile meteorology.
11) total gross radioactivity in primary coolant released to secondary coolant released to environs.
- 12) ten percent of the combined r'dioiodine activity from primary activity in secondary coolant and secondary activity present in steam mass (released to environs) assumed released to environs.
The whole body dose resulting from immersion in the cloud containing the released activity would include both gamma and beta radiation.
The gamma dose is dependent on the finite size and configuration of the cloud.
- However, the analysis employed the simple model of the semi-infinite cloud, which gives 2
an upper limit to the potential gamma dose.
The semi-infinite cloud model is applicable to the beta dose, because of the short range of beta radiation in air.
The resulting whole body dose was determined to be less than 0.5 Rem for this accident.
The thyroid dose from the steam generator tube rupture accident has been analyzed assuming a tube rupture at full load and loss of offsite power at the time of the reactor trip, which results in steam release through the relief valves in the period before the faulty steam generator is is olated and primary system pressure is reduced.
The limiting iodine activitiee for the primary and secondary systems are t. sed in the initial conditions. One-tenth of the iodine contained in the liquid which is converted to steam and passed through the relief valves is assumed to reach the site boundary.
The resulting thyroid dose from the combined primary and secondary iodine activity released to the environs was determined to be 1.5 Rem for this accident.
The limit for secondary iodine activity is consistent with the limits on primary system iodine activity and pri. nary-to-secondary leakage of Igpm.
If the activity should exceed the specified limits following a power transient, the major concern would be whether additional fuel defects had developed bringing the total to above expected levels.
From the observed removal of excess I
activity by decay and cleanup, it should be apparent whether activity is j
returning to a IcVel below the specification limit.
Appropriate acticn to be l
taken to bring the activity within specification include one or more of the following:
gradual decrease in power to a lower base power, increase in letdown flow rate, and venting of the makeup tank gases to the waste gas decay tanks. _
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The engineered safety features valves associated with each of the above systems shall be opcrable or locked in th, ES position.
3.3.2 In addition to 3.3.l 'above, the following FCCS equipment shcIl be
. operable when the reactor coolant system is above 350 F and irradi-ated fuel is in the core:
(A) Two out of three high presourc injection (makeup) pumps shall be maintained operable, pcuered from independert essential busses, to provide redundant and independen', flow paths.
(B) Engineered safety features valves associated with 3.3.2.a above shall be operable or locked in thc 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 f1 coding tanks shall cach contain an indicated minimum of 13 + 0.4 feet (1040 + 30 ft3) of borated water at 600 + 25 psig.
(B)
Core flooding tank boron concentration shall not be less than 22'l0 ppm bcron.
(C) The e.'.cctrically operated discharge valves from the core flood tanks shall be open and breakers Iceked open and toeged.
(D)
One of the two pressure instrument channels and one of the two level instrument channels per core flood tank shall be operable.
3.3.h The reactor shal) not be made cricical unless the following equiptent 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 sprny nozzle headers and four reactor building emergency cooling fans and ussociated cooling units.
(0)
The sodium thiosulfate tank shn11 contain an indicated 31 ft of 30 wt's solution sodium thiosulfate (37,500 lb). The sodium hydroxide tank shall contain an indicated 31 ft. of 20 wtt solution sodium hydroxidn (20,500 lb.).
(C) All manual valves in the main discharge lines of the sodium thiosulfate and sodium hydroxide tanks shall be locked open.
(D)
Engineered safety feature valves and interlocks associated with 3. 3.1, 3.3.2, and 3. 3. 3 chall be operable or locked in the ES position.
3.3.5 Maintenance shall be allowed during power operation on any component (s) in the high pressure injection, low pressure injection, service water, 2
reactor building spray, reactor building cooling and penetration room 37-
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ventilation systems which will not remove more than one train of each system from service.
Maintenance shall not be performed on components which would make the affected system train inoperable for more than 24 consective hours.
Prior to initiating maintenance on any component of a train in any system, the redundant component of that system shall be demonstrated to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the maintenance.
3.3.6 If the conditions of Specifications 3.3.1, 3.3.2, 3.3.3, 3.3.4 and 3.3.5 cannot be met except as noted in 3.3.7 below, reactor shutdown shall be initiated and the reactor shall be in hot shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and, if not corrected, in cold shutdown condition-within an additional 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.3.7 Exceptions to 3.3.6 shall be as follows:
2 (A)
If the conditions of Speci'.ication 3.3.1(G) cannot be met, reactor operation is permissible only during the succeeding seven days unless such components are sooner made operable, provided that during such seven days the other BWST level instrument channel shall be operable.
(B)
If the conditions of Specification 3.3.3(D) cannot be met, reactor operation is permissible only during the succeeding seven days unless such components are sooner made operable, provided that during such seven days the other CFT instrument channel (pressure or level) shall be operable.
Unsen She requi rocents of Specification 3. 3.1 ausure that below 3500F, adequate lonC tern core cooling is provided.
Two low pressure injection pu:Ts are speci fic 1.
llovever, only one is necessary to supply e=crconcy ecolant to the reactor in the event of a loss-of-coolant accident.
The post-accident reactor building cooling and 1cng-term pressure reduction may be accesplished by four ecoling units, by two spray units or by a cenbi-nation of two cooling units and one spray unit.
Post-accident iodine rc=cval may be accomplished by one of the two spray system strings.
T1.e specified requironents assure that the required post-accident co=penents are available for both reactor building cooling and iodine removal.
Specification 3. 3.1 assures that the required equipment is cperational.
The borated water storace tank is used for three purposes:
(A) As a supply of borated water for accident conditions.
(B) As an altygoate supply of borated water for reaching cold shutdown.
I (C) As a supply of borated water for floodin5 the fuel transfer canal during refueling operation. (3).
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6 350,000 gallons of borated water are supplied for emergency core cooling and reactor building spray in the event of a loss-of-coolant accident.
This amount fulfills requirements for ccergency core cooling.
16,000 gallons of borated water are required to recch cold shutdown.
The borated water storage tank capacity of 380,000 gallons is based on refueling volume requirements.
Heaters maintain the borated water supply at a temperature to prevent crystal-lization and local freezing of the boric acid.
The boron concentration is set at a value that vill maintain the core at least 1 percent ak/h suberitical at 70 F vithout any control rods in the core.
The concentration for 1% Ak/k 0
suberitica]ity is 1609 ppm boron in the core, while the minimum value speci-fled in the borated water storage tank is 2270 ppm boron.
Specification 3.3.2 assures that above 350 F two high pressure injection pumps are also available to provide injection water as the energy of the reactor coolant system is increased.
Specification 3.3.3 assures that above 800 psic both core flooding tanks.
are operational.
Since their design pressure is 600 + 25 poig, they are not brought into the operational state.until 800 psig to prevent spurious in-jection of borated water.
Both core flooding tanks are specified as a
)
single core flood tank has insufficient inventory to reflood the core.
Specification 3.3.h assures that prior to going critical the redundant reactor building cooling unit and spray are operational.
The spray system utilizes common suction lines with the low pressure injection system.
If a single train of equipment is removed from either systen, the other train must be assured to be operable in each system.
. When the reactor is critical, maintenance is allowed per Specification 3.3.5.
Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5.
The maintenance period I
of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable if the operability of equipment redundant to that removed from service is demonstrated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to removal.
2 Exceptions to Specification 3.3.6 permit continued operation for seven days if one of two BNST IcVel instrument channels is operable or if either the pressure or level instrument channel in the CFT instrument channel is operable.
In the event that the need for emergency core cooling should occur, function-ing of one train (one high pressure injection pump, one low pressure injec-tion pump, and both core f1 coding tanks) vill protect the core and in the event of a main coolant loop severence, limit the peak clad temperature to less than 2300 F and the metal-water reaction to that representing less than 1 percent of the clad.
The service water system consists of two independent but interconnected, full capacity, 100% redundant systems, to ensure continuous heat removal.(h)
One service water pump is required for normal operation.
The normal operating requirements are greater than the emergency requirements following a loss-of-coolant accident.
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3.5.1.7 The Decay Heat Removal System isolation valve closure setpoints shall be equal to or less than 340 psig for one valve and equal to or less 2
' than 400 psig for the second valve in the suction line. The relief valve setting for the DHR system shall be equal to or less than 450 psig.
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- 3. JExcept fer physics tests or exercising control rods, the control rod withdrawal limits are specified on Figures 3.5.2-1 A and 3.5.2-1B for four pump operation and on Figure 3.5.2-2 for three or two pump operation.
If the control rod position. limits are exceeded, corrective measures 'shall be taken immediately to achieve an acceptable control rod position.
Acceptable control rod posi-tions shall be attained within four hours.
4.
Except for physics tests, power shall not be increased above the power level cutoff (see Figures 3.5.2-1) unless'the xenon reactivity is within 10 percent of the equilibriun value for operation at
, rated power and asymptotically, approaching stability.
3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.
Except for physics tests, imbalance shall be maintained within the envelope defined by Figure 3.5.2-3.
If the imbalance is not within the envelope defined by Figure 3.5.2-3, corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.7 ' 'ihe control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
Bases The power-imbalance envelope defined in Figure 3.5.2-3 is based on 1)
U3CA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-4) such that the maximum clad temperature will not exceed the Interim Acceptance 2
Criteria and 2) the Protective System Maximum Allowable Setpoints (Figure 2.3-2).
C rrective measures will be taken immediately should the indicated o
quadrant tilt, rod position, or imbalance be outside their specified boundary.
Operation in a situation that would cause the interim acceptance criteria to be approached should a LOCA occur is highly improbabic because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and un-certainty factors are also at their limits.* Conservatism is introduced by application of:
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Fuel densification effects d.
Hot rod manufacturing tolerance factors lhe 30 percent overlap between successive control rod groups -is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
- Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors.
The method used to define the operating limits is defined in plant operating procedures.
48 c
r Power Level, 5 RESTRICTE0 102 REGION
+6.1 100 2
(
80
- - = --
-60 PERI.tl SSIBL E OPE" 4 flG REG:
a 40 40 25 20 0
+20
+40 Core imbalance, 5 ARKANSAS P0i!ER & LIGiti C0 OPERATIONAL POWER IMBALANCE FIG.NO.
ARKANSAS NUCLEAR ONE-UNIT 1 ENVELOPE 3.5.2-3 48e e,
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20 18 R
Ec
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16 i
5 5=
14
=
12 0
2 4
6 8
10 12 Axial Location of Peak Power f rom Bottom of Core, it t
ARKANSAS POWER & LIGHT CO.
L O C.% LIMITED MAXIMUM ALLOWABLE FIG.NO.
ARKANSAS nut; LEAR ONE-UNIT 1 LINEAR HEAT RATE 3.5.2-4
- 48f -
l
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~ 3.9 -RADIOACTIVE DISCHARGE This specification has been replaced by specification 2.4 of the environmental technical specifications -(Appendix B to the operating license).
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3.10 SECONDARY SYSTEM ACTIVITY Applicability Applies to the limiting conditions of secondary system activity for operation of the reactor.
Objective To limit the maximum secondary system activity.
Specification The I-131 dose equivalent of the radioiodine activity in the secondary coolant shall not exceed 0.17 pCi/gm.
Bases For the purpose of determining a maximum allowable secondary coolant activity, the activity contained in the mass released following the rupture of a steam generator tube, a steam line break outside containment and a loss of load
~
incident were considered.
The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the atmosphere by the condenser vacuum pumps.
Thus in the event of a loss of load incident or steam line break, there are only small quantities of these gases which would be released.
The dose analysis performed to determine the maximum allowable reactor coolant activity assuming the maximum allowable primary to secondary leakage of 1 gpm as given in the Bases for Specification 3.1.4.1 indicated that the controlling accident to determine the allowable secondary coolant activity would be the rupture of a steam generator tube.
For the loss of load incident with a loss of 2
295,000 pounds of water released to the atmosphere via the relief valves, the resulting thyroid dose at the I-131 dose equivalent activity limit of 0.17 uCi/gm would be 0.6 Rem with the same meteorological and iodine release assump-tions used for the steam generator tube rupture as given in the Bases for Specification 3.1.4.1.
For the less probabl the assumption is made that a loss of 1 x 10g accident of a steam line break, pounds of water or the contents of one loop in the secondary coolant system occurs and is released directly to the atmosphere.
Since the water will flash to steam, the total radioiodine activity is assumed to be released to the atmosphere.
The resulting thyroid dose at the I-131 dose equivalent activity limit of 0.17 pCi/gm would be less than 28 Rem with the same meteorological assumptions used for the steam generator tube rupture and loss of load incident.
Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information.
The nuclear flux (power range) channels shall be cali-brated at least twice weekly (during steady state operatin6 conditions) against a heat balance standard to compensate for instrumentation drift.
Durins non-steady state operation, the nuclear flux channels shall be calibrated daily to compensate for instrumentation drift and changing rod patterns and core physics parameters, espe #
h e
9 9
4 4
6 67a L
~
p3 Oth:r channals cra subject caly to " drift" crrors induced within the
- ins'trumentation itself and, consequently, can~ tolerate longer intervals between calibrations.
Process system instrumentation errors induced r
by drift can be expected to remain within acceptabit tolerances if re-calibration is performed at the intervals of each reiaeling period.
4 Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.
Thus, minimum calibration frequencies for the nuclear flux (power range) channels, and once each refueling period for the process system channels is considered acceptable.
Testing i
On-line testing of reactor protective channels is required once every 4 weeks on a rotational or staggered basis.
The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors
~
j being introduced into each redundant channel.
The rotation schedule for the reactor protective channels is as follows:
~
Channels A, B, C, D Before Startup if shutdown greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel A One Week After Startup Channel B Two Weeks After Startup Channel C Three Weeks After Startup Ch'annel D Four Weeks After Startup The reactor protective system bastrumentation test cycle is continued with one channel's instrumentation tested each week.
Upon detection of a failure that prevents trip action, all instrumentation associated with the protective channels will be tested after which the rotational test cycle is started again.
If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.
The protective channels coincidence logic and control rod drive trip breakers are trip tested every four weeks.
Th,e trip test checks all logic combinations.and is to be performed on a rotational basis.
The logic and breakers of the four protective channels sha11 be trip tested prior to startup and their individual channels trip tested on a cyclic basis.
Discovery of a failure requires the testing of all channel logic and breakers, after which the trip test cycle is started again.
The equipment testing and system sampling frequencies specified in Table 4.1-2 and Table 4.1-3 are considered adequate to maintain the status of l2 the equipment and systems to assure. safe operation.
REFERENCE FSAR Section 7.1.2.3.4 68
cem -
Table h.1-1 (Cont'd)
Channel Description
' Check Test Calibrate Remarks 20.
Reactor Building Spray NA M (1)
NA (1) Including RB spray pump, spray
~
' System Logic Channels valve, and chem. add. valve logic channels.
21.
Reactor Building Spray System Analog Channels 7.
a.
Reactor Building NA _
M R
Pressure Channels 22.
Pressurizer Temperature S
NA R
. Channels
,23.
Control Rod Absolute S(l)
NA R
(1) Compare with Relative Position Position
, Indicator, p,
24.
Control Rod Relative S(l)
NA R
(1)
Check with Absolute Position Position Indicator.
25 Core Flooding Tanks a.
Pressure Channels S
NA R
b.
Level Channels S
NA R
26.
Pressurizer Level Channels S
NA R
27 Makeup Tank Level Channels D
NA R
28.
Radiation Monitoring W'
M(1)
Q(2)
(1) Check functioning of'self-checking Systems feature on each detector.
(2) R for those detectors inaccessible during normal operation
" 29 High and Low Pressure NA NA R
Injection Systems: Flov Channels
Tabic 4.1-1 (Cont'd)
Channel Description ^
Check Test
' Calibrate Remarks 30 Decay lleat Removal S(1)(2)
M(1)(3)
R (1) Includes RCS Pressure Analog System Isolation Valve Chantal Automatic Closure And Interlock System (2) Includes CFT Isolation Valve f.
Position
~
(3) Shall Also Be Tested During Refueling Shutdown Prior to Re-pressuri:ation at a_ pressure greater 2
than 300 but less than 420 psig.
31.
Turbine Overspeed Trip N/A R
N/A
~
Mechanism 32.
Steam Line Break (Later)
Rj Instrumentation And Control 33.
Diesel Generator M
Q N/A Protective Relaying, Starting Interlocks
,And Circuitry i
34.
Off-site Power Under-N R
R voltage And Protective Relaying Interlocks And Circuitry' 35.
Borated Water Storago W
N/A R.
Tank Level Indicator 36.
Boric Acid Ilix Tank
~
a.
Level Channel N/A N/A R
2 b.
Temperature Channel
'M N/A R
N T2ble 4.1-2 Hinimum Equipment Test Frequency Item Test
' Frequency
- 1.. Control Rods Rod Drop Times of All Each Refueling Shutdown Full Length Rods 1/
2.
Control Rod Move' ment Movement of Each Rod Every Tteo Weeks Above Cold Shutdown Conditions 3.
Pressurizer Code Setpoint One Within 2 Weeks Prior Safety Valves to or Following Each Refueling Shutdown 4.
Main Steam Safety Setpoint Four tlithin 2 Neeks Prior Valves to or Following Each Refueling Shutdown 5.
Refueling System Functioning Start of Each Refueling Interlocks Shutdown 6.
Reactor Coolant Evaluate Daily System Leakage 7.
Charcoal and liigh Charcoal and IEPA Fil-Eadt Refueling Period and Efficiency Filters ter for Iodine and at.Any Timo Kork on Filters in Control Room, Particulate Removal Could Alter Their ?ntegrity Penetration Room Efficiencies.
DOP Ventilation System, Test on llEPA Filters.
Ilydrogen Purge Frcon Test on Char-System, and Reactor coal Filter Units 2/
Purge System 8.
Reactor Building Functioning Each Refueling Shutdown Isolation Trip 9.
Service Water Functioning Each Refueling Shutdown
' Systems 10.
Spent Fuel Cooling Functioning Each Refueling Shutdown System Prior to Use 11.
Decay Heat Removal Functioning Each Refueling Shutdown System Isolation Prior to Repressurization Valve Automatic at a pressure greater than Closure and Isolation 30'0 psig but less than 420 2
System psig.
1/ Same as tests listed in section 4.7
[/ Same as tests listed in sections 4.4. 3, 4.5.3, 4.11 and 4.12 73 d
3 Table 4.1-2 (Continued)
Minimum Equipment Test Frequency Item Test Frequency 12.
Flow Limiting Annulus Verify, at normal One year, two years, on Main Feedwater operating conditions, three years, and every Lines at Reactor that a gap of at least five years thereafter Building Penetration 0.025' inches exists measured from date of between the pipe and initial test.
the annulus.
e 4
73a
,n 3
Teblo 4.1-3 MINIMUM SAMPLING AND ANALYSIS FREQUENCY Item Test Frequency 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 Radioiodine Determination c.
Weekly (3)(6)(7) 2 d.
Dissolved Gases d.
Weekly (7) e.
Chemistry (C1, F, and 0 )
c.
3 times / week (8) 2 f.
Boron Concentration f.
3 times / week g.
Radiochemical Analysis for g.
Monthly (7)
E Determination (2)(4) 2.
Borated Water Boron Concentration Weekly and after Storage Tank Water each makeup Sampic 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 Radioiodine Concentration a.
Weekly (5)(7)(10)
Samples b.
Isotopic Radioiodine b.
Monthly (7)(10)
Concentration (4) 6.
Sodium flydroxide Sodium flydroxide Concentration Quarterly and 2
Tank Sample after each makeup 7.
Sodium Thiosulfate Sodium Thiosulfate Concentration Quarterly and Tank Sample after each makeup Notes:
(1) A gross radioactivity analysis shall consist of the quantitative measurement of the total radioactivity of 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 10% of the limit specified in the Specification 3.1.4.1 or increases by 10 pCi/gm from the previous measured level, the frequency of sampling and analyzing shall be increased to a minimum of once/ day until a steady activity level is established..w
4 (2) A radiochemical analysis shall consist of the quantitative measurement of the activity for each radionuclide which is identified in the primary coolant 15 minutes after the primary system is sampled.
The activities for the individual isotopes shall be used in the determination of E.
A radiochemical analysis and calculation of E and iodine isotopic activity shall be performed if the measured gross activity changes by more than 10 pCi/gm from the previous measured level.
The gamma energy per disinte-gration for those radioisotopes determined to be present shall be as given in " Table of Isotopes" (1967) and beta energy per disintegration shall be as given in USNRDL-TR-802 (Part II) or other references using the equivalent values for the radioisotopes.
(3)
In addition to the weekly measurement, the radioiodine concentration shall be determined if the measured gross radioactivity concentration changes by more than 10 pCi/gm from the previous measured level.
(4)
Iodine isotopic activities shall be weighted to give I-131 dose equivalent activity.
(5)
In addition to the weekly measurement, the.radiciodine concentration shall be determined if there are indications that the primary to secondary coolant leakage rate has increased by a factor of 2.
(6)
Whenever the steady state radiciodine or gross radioactivity concentration 2
of prior operation is greater than 1 percent but less than 10 percent of Specification 3.1.4.1, a sample of reactor coolant shall be taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of any reactor criticality and analyzed for radioactive iodines of
'I-131 through I-135 and gross radioactivity as well as the coolant sample and analyses required by the above.
Whenever the steady state radioiodine or gross radioactivity concentration of prior operation is greater than 10 percent of Specification 3.4.1, a sample of reactor coolant shall be taken prior to any reactor criticality and analyzed for radioactive iodines of I-131 through I-135 and gross radioactivity as well as the coolant sample and analyses required by above.
(7)
Not required when plant is in the cold shutdown condition or refueling shutdown cond. tion.
_(8) 02 analysis is not required when plant is in the cold shutdown condition et refueling shutdown condition.
(9)
Required only when fuel is in the pool and prior to transferring fuel to the pool.
(10)
Not required when not generating steam in the steam generators.
4.2 '
REACTOR C00IANT'~'STH4 SURVEILIANCE
^
', Applicability Applies to the surveillance of the reactor coolant system pressure boundary.
Objective To assure the continued integrity of the reactor coolant system pressure boundary.
Specification 4.2.1 Prior to initial unit operation,.an ultrasonic test survey shall be made of reactor coclent system pressure boundary welds as required to establish preoperational integrity and base line data for future inspections.
4.2.2 Post operational inspections of components shall be made in accor-dance with the methods and intervals indicated in IS-242 and IS-261 of Section XI of the ASME Boiler and Pressure Vessel Code,1971, including 1971 Winter addenda, except as follows:
IS-261 Item Component Exception 1.4 Primary Mozzle to 1 RC inlet no::le to be Vessel Welds inspected after approx. 3 1/3 years operation.
All four RC inlet nozzles to be inspected at or near the end of interval.
Both RC outlet no::les vill be inspected after approx. 6 2/3 yrs, operation.
One core flood no::le vill be inspected after 3 1/3 years operation and one core flood nos le inspected near the end of interval
~
33 Safe Ends on Heat Not Applicable Exchanger 4.1 Vessel Safe End Not Applicable Welds 4.2 Valve Pressure Not Applicable Retaining Bolting Larger than 2" 49 Integrally Welded Not Applicable Supports 6.1' Valve Body Welds Not Applicable 6.3 Uslve to Safe End Not Applicable Welds 76
l p
~
4.4.1.2.5 Test Frequency Local Icak detection testa shall be performed at a frequency of at least each refueling period, but in no case at intervals greater than two years except that:
~
(a)
The equipment hatch and fuel transfer tube seals shall be additionally tested after each opening.
(b)
If a personnel hatch or emergency hatch door is opened when reactor' building integrity is required, 2
the affected door seal shall be tested.
In addition, a pressure test shall be performed on the personnel and emergency hatches every six months.
4.4.1 3 Reactor Building Modifications Any major modification or replacement of components affecting the reactor building integrity shall be followed by either an integrated leak rate test or a local leak test, as appropriate, and shall meet the acceptance criteria specified in 4.3.1.1 and 4.3 1.2 respectively.
4.4.1.4 Isolation Valve Functional Tests e Every three months, remotely operated reactor building isolation valves shall be strcked to the position required to fulfill their safety function unless such operation is not practical during plant operation.
The latter valves shall be tested during each refueling period.
4.4.1 5 Visual Inspection A visual examination of the accessible interior and exterior surfaces of the reactor building structure and its components shall be performed during each refueling shutdown and prior to any integrated leak test, to uncover any evidence of deterioration which may affect either the reactor building's structural integrity or leak-tightness.
The discovery of any significant deterioration shall be accompanied by corrective actions in accord with acceptable procedures, nondestructive tests, and inspections, and local testing where practical prior to the conduct of any inte-grated leak test.
Such repairs shall be reported as part of the test results.
Bases (1)
The reactor building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 285 F.
Prior to initial operation, the reactor building will be strength tested at 115% of design pressure and leak rate tested at the design pressure.
The reactor building will also be leak tested prior to iaitial operation at not less than 50% of 83 i
~ - - -
o m
j
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i
)
)
the design pressure. These tests will verify that the leakage rate from d
reactor building pressurization satifies the relationships given in the specification.
The performance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the reactor building in case of an accident that would pressurize the interior of the reactor building.
In order to provide a realistic appraisal of the integrity of the reactor building under accident conditions, the reactor i
building 17olation valves are to be closed in the normal manner.
The test pressure of 30 psig for the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate and it duplicates the pre-operational leakage rate test at 30 psig.
The specification provides a relationship for relating the measured leakage of air at 30 psig to the potential leakace at 59 psig.
The frequency of the periodic integrated leakage rate test is keyed to the refueling i
schedule for the reactor, because these tests can best be performed during refueling shutdowns.
1 The specified frequency of periodic integr'ated leakage rate tests is based on three major considerations.
First is the low probability of Icaks in the liner, because of conformance of the complete reactor building to a 0.20% leakage rate at 59 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation. Second is the core frequent testing, at design
' pressure, of those portions of the reactor building envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the lov value of.60L leakage that is specified a
as acceptable from tested penetrations and isolation valves.
Third is the tendon stress surveillance program which provides assurance that an i
important part of the structural integrity of the reactor building is maintained.
Re ferences (1) ESAR, Sections 5 and 13.
l 4
9
1 i.
4.6.2 ' Station Batteries and Switchyard Batteries 1.
The voltage, temperature and specific gravity of a pilot cell-in each bank and the overall battery voltage of each bank shall be measured and recorded daily.
2.
Measurements shall be made quarterly of voltage of each. cell to the nearest 0.01 volt of the specific gravity of each cell, and of the temperature of every fifth cell in each bank.
The level of the electrolyte shall be checked and adjusted as required.
All data, including the amount of water added to any. cell, shall be recorded.
3.
During each refueling outage, a performance discharge test shall be conducted in accordance with the manufacturer's instructions, for the purpose of determining battery capacity.
4 Any battery charger which has not been loaded while connected to its 125V d-c distribution system for at least 30 minutes during every quarter shall be tested and loaded while connected to its bus for 30 minutes.
The third battery charger, which is capabic of being connected to either of the two 125V d-c 2
distribution systems, shall be loaded while connected to each bus for at least 30 minutes every quarter.
4.6.3 Emergency Lighting The correct functioning of the emergency lighting sys' rem shall be verified at least once each year.
100a i-m-
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resulting from the other two postulated accidents vere determined.
Since we consider. the probability of occurrence for either the steam generator tube rupture or the loss of load incident to be OsEr comparable, the acceptable thyroid dose limit for either incident a
was taken as 1.5 Pen.
We consider the occurrence of a steam line break occident outside containment to be core likely than a
.'~~
loss of coolant accident but less probeble than either the steam 5
generator tube rupture or the ' loss of load incident.
For this reason, we consider the acceptable thy'roid dose limit to be 30 Rem or
~
sienificantly less than the guideline doses of 10 CFR Part 100 b.. =
As stated in the bases for this section of the technical specifications, 7
the resultinn thyroid doses usinF the specified secondary systen activity linit of 0.17 pCi/gm of I-131 dose equivalent are appror.ircately 1.5 ren for the steam generator tube rupture (as stated
{p W=
in the Eases to Specification 3.1.4), 0.6 Rem for the loss of load incident and 28 Ren for the steam line break accident outside t_
con t a int;ent.
All of.these doets are less than the above stated il T" dose guidelines for these accidents and indicate that tne controlling
[
accident for determining the secondary coolant radiciod ine limit L..
is the stece cenerator tube ru pt u re.
An increase in the ratio L.
of radiciodine specific activity for the reactor coolant to the
[
secondary coolant would directly reduce the calculated dose for i.
l the two sceidents involvinF only secondary coolant releases.
[
Eowever, en increese in this ratio vould not significant ly reduce i
the calculated dose for the steam generator tube rupture which I
l re'.ennes both resctor coolant end secondary coolcnt rad ioac t ivit y.
(7) Table 4.1-1 (Ite: 30) and Table 4.1-2 (Iten 11) - 1 ote 3. has been L
c'Edardit3 3efieEt' itie"nediyWL'ab'IUhed'seipoints on the isolation
(-
l valves of DHES given in Specification 3.5.1.7 cnd gives the pressure range within which.the test nust be perforned.
The test will verify the correct setpoints for the isolation valves.
['
1he sc e chaure was r.ade to the test frecuency coluran in Toble 4.1-2 t
for coneistency.
E (F) Tchle
(. 1 3. 'Ihe minitun rampling end anclysis frecuency cnd t ests bcVe been changed for the reactor coolant semples.
The Cross
[
Activity Peterminct ion (previously designated ac Cross Ieta cud
[
Cse-c Activity) f requency has been reduced t ror. 5 tires per week to 3 t ires per week and at least every third day.
This frecuency
)
also 1.ac been desigr,ated f or raensurity the Che-istry and Ioren E.
Conrentrat ien in the Eccctor Coolant.
Irperience has st;oun thct I
such frec,uencies are c.dequate te detect changes in coole.nt chenistry U
on'a timely basis end pereitt reactor operclien over a week-end or holidcy without-the need for a reactor coolent sanple cnalysis.
Cama Isotopic Analyais f recuency hss been increased f ron conthly
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.- TT v. m. oovasa.uswr enownws omets non rec.t es
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to bi-veekly (once every two weeks) and the Radiochemical Analysis E
for E Determination has been increased from cer.i-ennually to conthly.
Both of these changes in f requency are to detect on a tinely basis any change in quality of the pross radioactivity contained in the reactor coolant.
Experience has shown that ruch frequencies
[
uill detect changes in quality of radioactivity due to additional y_
failed fuel or chc.nge in reac'.or operations.
The Cross Radioiodine
'J Detenninet ion has been added to detect radiciodine activity levels h-in the reactor coolant for compliance uith Section 3.1.4 requirecents.
[i The specified frequency for the analysis is weekly but shall be h
nore frequent if the gross activity incresces by a given araount as specified by Note 3.
Experience has shown that a weekly frequency
{?
i
+5" with this condition for more frequent analysis is adequate to detect on a tirely b' asis any changes in reactor coolant radiciodine levels.
s A deterrinction of dissolved gases concentration in the reactor coolant is recuired by Specification 3.1.9.1 which places a linit of 100 std ec per liter of vnter of dissolved rcses in the recetor E
coolant for control rod operation.
The buildup of dissolved gases in the reactor coolant is a slow process and therefore weekly F
detereinstion of this parameter is considered cdequate for tinely j
detect ien of any unusual increases.
t-t The existine require.ments for deternining tritium concentration, Er-t9 end Sr 00 concentrction ccd ?ross alpha activity in the reactor coolart are not required because there are no linits recessary for these radioisot opes for operating of the facility.
=r The ex i s t in.r requireneut for deterr.ining gross beta-gamma activity in the secondcry coolent is not required because ruch activity would not be present and no limits on operation for gross activity f5 cre recuired.
Thm.efore, these recuirerents have been deleted fro-the table.
Analyses for radioactivity levels and dissolved rases are not recuired when the plant is in the cold or refueline snutdovn condition because these parameters do not affect the safety of the plant when in there shutdown conditions.
Ihus, Note 7 provides for thece exclusions.
To determine the level and duration f.
of possible radioact ivity spiking for both gross and iodine activity, aJditional analysis depending upen level of radiocctivity during
[
the previous stecdy state operatice et the time of reactor startup l
end durine reactor startup is required by Eote 6.
This note arrlies only to the gross deterninations of total activity and radioiodinee. _Uote 1 basically recains the sexe as previously stated for thi> table except that the increased frequency of analysis u2 is recuired only until steady state activity level is established.
./s discussed in :ote 1 and Fote 2, the gross activity determination C
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(E+'i:
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vill be based on the activity present 15 nint'tes af ter sannlin.c.
This ti e period is equivalent to the miniciu:n expected decay time r:[i :+i:1 l
. fro' the release point to the nearact site boundary in case of a stecci generator tube rupture.
rote 2, which is associated with the deternination_oi E, specifies the teethod to be used for the p"r d.:terrination of E, the frequency of determinaticn for E and rcdio-(5=
iodine, and the reference (or equivalent) source to be used to 5.....
determine the individual gn=ma and beta enercies per disintegration
[f ~
.7 :f" i
for the rcdioisotopes present in the reactor coolant.
rote a i
to. this same sacple deterrainction indicates that all radiciodine E
act ivit ies (I-131,132,133,134, and 135) are to be weighted to E
determine the 1-131 dose equivalent activity actually present in E:r the reactor coolant for comparison with the limit established in b
Specification 3.1.4.1.b.
Note P states that the 0 2ccalysis is f:;'
- g not ' recuired uhen the plant is in the cold or refuelinF shutdoun
{9 con lition for the scite reason as riven for tiot e ) sccept ability.
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Fote " states that c deterninction of boron cor. centration in
{l ggg.
the Spent Fuel Pool is' required only if fuel is present in the 7
root end prior to fuel being transferred to the pool.
Since tne baron in pool water is to asi.ure that the fuel rercino sub-critical, it is not required when there is no fuel in the pool, 2:
l therefore, I;ote 9 is accepts:,le, rotes 5 and 10 cpply to the cocondary coolant sa:iplinr and ennlysis propron.
Note 5 recuires cdditionci sa:rpline and-analysis it the prir cry to secondary leakaee increases eignificantly.
Fote IC elininates the recuire-rent for sa::pling and analyris of the secondary cysten coolant then stea generation is not occurring.
If stu n is not being L
pencreted, the postulated accidents, which estabt isbed the'
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secondary coolent cet ivity li:-it, could not occur or vould not result in any significant release of radiciocine to the environs p:
fron the recondary coolant.
Fote 4 cleo epi, lie:: to the secondary F
coolant rcdiciedine cct ivity det er:r.innt ion for aesessing the 1-131 h
dose equivalent activity for cocparison with the limit ectsblished r
l in Speciticatien 3.10.
Note 7 also crplies to the secondtry coolant I
scrplire ano cnclysis when the rinnt is ir. the ccid or ref ueling shutdorn concition, j
ic) ~ r;ecificction 4.4.1.2.5(b) - To rohe the test in : recu i reent s E
on'tr H eTriencef hat W add evergency hatch door seals consistent vith Antent'ix.J of 10 CFH 1 art
.0. the first senterce of this l
specificat ion wac c:odified.
The cdd it ion of t he phrase "m-m res.c t or buildicy integrity is.receirec" sctisfies /rpenoir J,10 CFP 50,
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require 2.ents provided the er.istine plo ase, ' tut no e o re frequently P
than dcily durinn r.orral operction' is deleted.
We enist ine r equ ire::ent for weekly testing durine refuelin: or cold shutdouns i?
mit *
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[&:: ::.;.i&l is not necessary for the proposed specification which limits
}E Er ter tinp requirement s to tises then building integrity is required.
7l.sj '
These changes are acceptable and were made.
=
(10) S,recificat ion 4.6.2.4 - The exist in:: testing rerluirement relating to the third battery charper has been expanded to apply to all three bat tery chargers to assure that sdequate surveillance of Q
all bat tery chargers is provided.
The additional surveillance h:::
requirement in considered appropricte and acceptable to the staf f p =::
and should increase the reliability of the station battery syster.
R l.
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(11) Appropriate cl.anges in the Eases were made for clarification, f'
5"~
but they do not affect the specifications governing operation i{. _....
of the fccility.
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Cor.CLUS10::
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7 te have coccluded, bared en the considerat ions discussed soove, that:
m (1) because the change does not involve a significant increcse in the I'
trobt,bility or consequences ot accident s previously considered and i
does not involve a significant decrease in a safety nargin, the change
[
does not involve a sir,nificant hazards considerction, (2) there is
[:
reasonable assurance thct the health and safety of the public vill 1
not be endancered by operation in the proposed v.t.nner, and (3) such I
act ivit ies will be conducted in compliance with the Conrti::sion's F
regulations and the iscunnce et this c~encuent will net be inimical to the.ca ron defenu and security or to' the hecith arc safety of the ptib l i c.
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