ML19317E064
| ML19317E064 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/05/1977 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 7912120780 | |
| Download: ML19317E064 (17) | |
Text
s MAY 5 1977 Y0f\\g
\\
j Occkets Nos.
in uu and S0-237 Duke Power Company ATT!1: Mr. William O. Parker, Jr.
Vice President - Steam Production Post Office Box 2178 422 Sout!. Church Street Charlotte,florth Carolina 28242 Gentlemen:
Gy letter dated Novenber 30, 1976, you requested an acendment to the license for the Oconee fluclear Station to incorporate requirements concerning the operability and inservice inspections of steam generators.
Your request was in response to our letter dated Septenber 21, 1976, which provided nodel Technical Specifications to be adapted to the Oconce Technical Specifications. We find that your response does not address all of the requirements covered in our model Technical Specifi-cations.
For those requirements you have addressed, we request additional information.
In our letter of Septe::ter 21, 1976, wa provided model Technical Speci-fications which included a reactor coolant leakage limit of 1 GPit through the steam generator tubes. Your lette* of Noverber 30, 1976, stated without justification, that a 1 GPM Icakage ate is overly restrictive and that the 10 GPM now allowed by the Cconee Te6nical Specifications is adequate.
It is our position that this leakage rate ::hould be limited to 1 GPM, particularly in view of the steam generator tube leaks which have been I
occuring at Gconee and other PWR facilities.
It is requested that you submit a request for chan;e to the Oconee Technical Specifications that limits the leakage rate to 1 GPH or provide detailed justification for not doing so.
An analysis perfonned by us shows that with a 1 GPM reactor coolant-to-secondary leakage rate, dose rates from postulated accidents would he well below the limits of 10 CFR Part 100. This analysis assumed that the reactor coolant activity was 1.0 uci/g and the secondary coolant activity was 0.1 uci/g.
1 OFFICE W auftNAME W oars >
NRC PORM 318 (946) NRCM 0240 W u. e. aavsRNMENT PmMTWO OFFICES le78-ese ead a91212 0780
[
Duke Power Capany MAY 5 1977 We have reviewed the Oconee Technical Specifications and find that they do not include iodine activity limit for the reacter coolant and the iodine limit for the secondary coolant is well above that assumed in our analysis.
It is requested that you submit a request for a change to the Oconee Technical Specifications that would limit the iodine coolants, respective $.and 0.1 pci/ m in the reactor and secondary activity to 1.0 pei/
Enclosure $ is a copy of the 8&W Standard Technical Specifications which you should use for guidance.
Enc' osure 2 contains coments we have on your proposed Steam Generator Ins'<rvice Inspection Technical Specifications.
It is requested that you respond to these coments by modifying your proposed Technical SpecificatMas to conform with the model Technical Specifications provided in our letter of Septenber 21, 1976, or by providing justification for any deviations.
It is requested thet you respond to the requests herein within 45 days of receipt of this letter.
Sincerely, Original signed by A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Enclosures:
1.
D&W Standard Technical Specifications 2.
NP.C Comments on Steam Generator Inservice Inspection o
cc w/ encl:
see next page DISTRIBUTION Docket Files OI*E (3)
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Duke Power Company May 5, 1977 cc: Mr. William L. Porter Duke Power Company P. O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 J. Michael McGarry, III, Esquire DeBevoise & Liberman 700 Shoreham Building 806-15th Street, NW.,
Washington, D.C.
20005 Oconee Public Library 201 South Spring Street Walhalla South Carolina 29691 l
-ENCLOSURE 2 l
COMMENTS ON OCONEE UNITS 1, 2, AND 2 PROPOSED TECHNICAL SPECIFICATIONS FOR STEAM GENERATOR INSERVICE INSPECTION 1.
Table 4.17-1 which specifies steam generator tube sample size, inspection result classification, and the corresponding acticn required for each sample inspection has several discrepancies with the standard technical specifications:
If the results of the first sample inspection fall in the C-2 a.
category, the corrective action should be the plugging of the defective tubes and inspection of h% additional tubes in that steam generator. Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection.
b.
If the results of the first sample inspection of a steam gener-ator fall in the C-3 category, the corrective action should include inspection of all tubes in the steam generator, plugging of defective tubes, and inspection of an additional 6"% of the I
tubes in each other steam generator.
Corrective actions corresponding to results of second and third c.
sample inspectiors have not been specified in the table. These actions should be specified in accordance with Table 4.4-2 of the standard technica1 specifications. Guidance for subsequent (second and third) sample inspections as stated in paragraph 4.17.1,b of the proposed technical specifications is unacceptable.
The sample sizes required during each sample inspection are clearly specified in Table 4.4-2 of the standard technical specifications.
Deviation from the specified sample sizes will require a statistical analysis justifying the proposed sample sizes.
2.
Paragraphs 4.4.5.2,b, and c of the standard technical specifications should be included under paragraph 4.17.1,b of the proposed technical specifications.
. I l
i Paragraph 4.17.2,b should read the same as paragraph 4.4.5.3,b in 3.
the standard technical specification.
Paragraph 4.17.2.c should also specify that unscheduled inservice inspections shall be performed on each steam generator in accordan 4.
with the first sample inspection during the shutdown subsequent to:
A seismic occurrence greater than the Operating Basis Earthquake, a.
A loss-of-coolant accident requiring actuation of the engineered b.
safeguards, or A main steam line or feedwater line break.
c.
Justificatic, for the proposed 40% plugging limit must be provided 5.
in accordance with Regulatory Guide 1.121.
The tenn " unserviceable" used in the definition of plugging limit 6.
must be defined.
The definition of defect should read as follows:
7.
Defect means an imperfection of such severity that it equals or exc adefectisdefectide.
the plugging limit. A tube containing 8.
The basis should state:
Cracks having a primary-to-secondary leakage less than the speci-fied limit during operation will have an adequate marg'in of safety a.
~
l to withstand the loads imposed during nonnal operation and by postulated accidents. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Cases when the results of any steam generator tubing inservice b.
inspection fall into category C-3 will be considered by the Commission on a case-by-case basis and may result in a requiremen for analyses, laboratory examina' tion, tests, additional eddy current inspection, and revision of the Technical Specification, if necessary.
l l
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9 9.
The subject of operability of steam generators as discussed in sections 3.4.5 and 4.4.5.6 of the standard technical specifications should be addressed.
/ ~ %,}
UNITED STATES NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555 g
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May 5, 1977 Dockets Nos. 50-269 50-270 and 50-287 Duke Power Company ATTN: Mr. William 0. Parker, Jr.
Vice President - Steam Production Post Office Box 2178 422 South Church Street Charlotte, North Carolina 28242 Gentlemen:
By letter dated November 30, 1976, you requested an amendment to the license for the Oconee Nuclear Station to incorporate requirements concerning the operability and inservice inspections of steam generators.
Your request was in response to our letter dated September 21, 1976, which provided model Technical Specifications to be adapted to the Oconee Technical Specifications. We find that your response does not address all of the requirements covered in our model Technical Specifi-cations.
For those requirements you have addressed, we request additional information.
In our letter of September 21, 1976, we provided model Technical Speci-fications which included a reactor coolant leakage limit of 1 GPM through the steam generator tubes. Your letter of November 30, 1976, stated without justification, that a 1 GPM leakage rate is overly restrictive and that the 10 GPM now allowed by the Oconee Technical Specifications is adequate.
It is our position that this leakage rate should be limited to 1 GPM, particularly in view of the steam generator tube leaks which have been occuring at Oconee and other PWR facilities.
It is requested that you submit a request for change to the Oconee Technical Specifications that limits the leakage rate to 1 GPM or provide detailed justification for not doing so.
An analysis perfonned by us shows that with a 1 GPM reactor coolant-to-secondary leakage rate, dose rates from postulated accidents would be well below the limits of 10 CFR Part 100. This analysis assumed that the reactor coolant activity was 1.0 pci/g and the secondary coolant activity was 0.1 pci/g.
A b
Duke Power Company May 5, 1977 We have reviewed the Oconee Technical Specifications and find that they do not include iodine activity limit for the reactor coolant ana the iodine limit for the secondary coolant is well above that assumed in our analysis.
It is requested that you submit a request for a change to the Oconee Technical Specifications that would limit the iodine activity to 1.0 uci/am and 0.1 uci/cm in the reactor and secondary coolants, respectiveTy. Enclosure T is a copy of the B&W Standard Technical Specifications which you should use for guidance. contains comments we have on your proposed Steam Generator Inservice Inspection Technical Specifications.
It is requested that you respond to these connents by modifying your proposed Technical Snecifications to conform with the model Technical Specifications provided in our letter of September 21, 1976, or by providing justification for any deviations.
It is requested that you respond to the requests herein within 45 days of receipt of this letter.
Sincerely, J
< deem A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Enclosures:
1.
B&W Standard Technical Specifications 2.
NRC Comments on Steam Generator Inservice Inspection cc w/ encl:
See next page l
l l
l l
lb i
ENCLOSURE 1 DEFINITIONS c.
Reactor coolant system leakage through a steam generator to the secondary system.
UNIDENTi ED LEAKAGE 1.15 UNIDEN IED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE o ONTROLLED LEAKAGE.
PRESSURE BOUNDARY lex. AGE 1.16 PRESSURE BOUNDARY L n "GE shall be leakege (except steam generator tube leakage) through a non-1 lable fault in a Reactor Coolant System component body, pipe wall or vess 1 wall.
CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that sea water flow supplico to the reactor coolant pump seals.
QUADRANT POWER TILT b
1.18 QUADRANT POWER TILT is defined by the following e tion and is se expressed in percent.
j QUADRANT POWER TILT =
j Power in any core cuadrant
.) )
100 (Average power of all quaarants DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (uCi/ gram) which alone would oroduce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Tablu III of TID-14844, " Calculation of Distance Factors for Power and Test heactor Sites."
E - AVERAGE DISINTEGRATION ENERGY 1.20 E-AVERAGE DISINTEGRATION ENERGY iaail be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies B&W-STS 1-4 January 1, 1977
DEFINITIMS per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
STABOERED TEST BASIS 1.21 A TAGGERED TEST BASIS shall consist of:
test schedule for n systems, subsystems, trains or designated a.
mponents obtained by dividing the specified test interval I
in n equal subintervals.
b.
The t ting of one system, subsystem, train or designated compon ts at the beginning of each subinterval.
FREQUENCY NOTATION 1.22 The FREQUENCY NO ATION specified for the performance of Surveillance Requirements shall corre and to the intervals defined in Table 1.2.
^
AXIAL POWER IMBALANCE l.23 AXIAL POWER IMBALANCE shall be the THERMAL POWER in the top half of the core expressed as a perc'entage of RATED THERMAL POWER minus the THERMAL POWER in the bottom half f the core expressed as a percentage I
of RATED THERMAL POWER.
SHIELD BUILDING INTEGRITY 1.24 SHIELD BUILDING INTEGRITY shall ex'st when:
Each door in each access opening 's closed except when the a.
access opening is being used for n6rmal transit entry and exit, then at least one door shall b closed.
i b.
The shield building filtration system 1 OPERABLE.
l The sealing mechanism associated with eac' penetration (e.g.
c.
welds, bellows or 0-rings) is OPERABLE.
REACTOR PROTECTION SYSTEM RESPONSE TIME 1.25 The REACTOR PROTECTION SYSTEM RESPONSE TIME shall be hat time interval from when the monitored parameter exceeds its trip tpoint at l
the channel sensor until power interruption at the control ro drive breakers.
l B&W-STS 1-5 January 1, 1977 l
REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall Le limited to:
I a.
1 1.0 pCi/ gram DOSE EQUIVALENT I-131.
b.
1 100/T pCi/ gram.
l APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION:
MODES 1, 2 and 3*.
With the specific activity of the primary coolant > 1.0 a.
pCi/ gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the lef t of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that operation under these circumstances shall not exceed 10% of fg the unit's total yearly operating time.
The provisions of
(
V Specification 3.0.4 are not applicable.
b.
With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T avg (500)"F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With the specific activity of the primary coolant > 100/f c.
pCi/ gram, be in at least HOT STANDBY with T
< (500)"F avg within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />'.
MODES 1, 2, 3, 4 and 5:
With the specific activity of the primarv_ coolant > 1.0 a.
pCi/ gram DOSE EQUIVALENT I-131 or > 100/E uCi/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE shal:
be prepareu a..o submitted to the Commission pursuant to Specification 6.9.1.
This report shall contain the results of the specific activity analyses together with the following information:
m
- With Iavg - (500)"F.
B&W-STS 3/4 4-20 January 1, 1977
REACTOR COOLANT SYSTEM AC110N:
(Continued) 1.
Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the i
first sample in which the limit was exceeded.
I 2.
Fuel burnup by core region.
3.
Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the 3
first sample in which the limit was exceeded.
4.
History of de-gassing operations, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit I
was exceeded.
5.
The time duration when the specific activity of the primary coolant exceeded 1.0 pCi/ gram DOSE EQUIVALENT I-131.
\\
i SURVEILLANCE REOUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
l BW-STS 3/4 4-21 January 1, 1977
4 TABLE 4.4-4 PRIMARY CO'0LANT SPECIFIC ACTIVITY SAMPLE h,
AND ANALYSIS PROGRAM v
TYPE OF MEASUREMENT SAMPLE AND MODES IN WHICH SAMPLE AND ANALYSIS ANALYSIS FREQUENCY AND ANALYSIS REQUIRED 1.
Gross Activity Determination At least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4 2.
Isotopic Analysis for DOSE 1 per 14 days 1
EQUIVALENT I-131 Concent' ration 3.
Radiochemical for E Determination 1 per 6 months
- 1 4.
Isotopic Analysis for Iodine a)
Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever l,2,3,4,5 In.'.luding I-131, I-133, and I ~135 the specific activity exceeds 1.0 pCi/ gram DOSE EQUIVALENT I.131 or 100/E pCi/ gram.
b)
One umple between 2 and 6 1, 2, 3
~
hours following a THERMAL POWER change exceeding 15 per-cent of the RATE's THERMAL POWER within a one hour period.
g.
~
j' Until the specific activity of the primary coolant system is restored within its limits.
I
- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
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0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Peccent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0gCi/ gram Dose Equivalent 1131 B&W-STS 3/4 4-23 June 1, 1976 l
i PLANT SYSTEMS ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.S The specific activity of the secondary coolant system shall be
< 0.10 uCf/ gram 00SE EQUIVALENT I-131.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the specific activity of the secondary coolant system > 0.10 uCi/ gram I
DOSE EQUIVALENT I-131, be in at least HOT STANCBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
I SURVEILLANCE REOUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of-the sampling and analysis program of Table 4.7-2.
l B&W-STS 3/477 June 1, 1976 l
l 21
--- =
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l l
l TABLE 4.7-2 SECONDAP.Y COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM t
SAMPLE AND l
j
?YPE OF MEASUREMENT ANALYSIS FREQUENCY I
AND ANALYSIS i
1.
Gross Activity Detennination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1
2.
Isotopic Analysis for DOSE a) 1 per 31 days, whenever EQUIVALENT I-131 Concentration the gross activity determina-tion indicates iodine concen-trations greater than 10%
of the allowable limit.
l b) l per 6 months, whenever i
the gross activity determination indicates iodine concentrations below 10% of the allowable limit.
e t
4 B&W-STS 3/4 7-8 June 1, 1976
-w-4
-7 w
ties:
! $f,qN fx. W ev:v BASES 3/4.4.7 EMISTRY The limit
'ons on Reactor Coolant System chemistry ensure that corrosion of the ctor Coolant System is minimized and reduce the potential for Reactor colant System leakage or failure due to stress corrosion. Maintaining e chemistry within the Steady State Limits shown on Table 3.4-1 provi adequate corrosion protection to ensure the structural integrity of the Re +or Coolant System over the life of the plant. The associated effects of ceeding the oxygen, chloride and fluoride limits are time and tempera e dependent.
Corrosion studies show that operation may be continued wi, contaminant concentration levels in excess of the Steady State Limits, p to the Transient Limits.
for the specified limited time intervals witho having a significant effect on the structural integrity of the Reactor aclant System. The time interval permitting continued operation within o restrictions of the Transient Limits provides time for taking correctiv ctions to restore the contaminant concentrations to within the Stead tate Limits.
The surveillance requirements provide adequate assurance tha. con-centrations in excess of the limits will be detected in sufficient t e
to take corrective action.
i 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site Loundary will not exceed an appropriately small fraction of the Part 100. limit following a steam generator tube ruptre accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site l
These values are conservative in the specific site para-locations.
l meters of the site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finali:-
ing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.
B&W-STS B 3/4 4-5 June 1, 1976
BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > l.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accomodates possible iodine spiking phenomenon which may occur following changes in TfiERMAL POWER. Operation with specific ac-tivity levels exceeding 1.0 pCi/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 10 percent of the units yearly operating time since the activity levels allowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.
Reducing T to < (500)*F prevents the release of activity should a steam generator dbe rupture since the saturation pressure of the primary a
coolant is below the lift pressure of the atmospheric steam relief valver.
The surveillance reclirements provide adequate assurance that excessive specific activity levels in the primary coolant will be de-tected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with
~
spiking phenomena. A reduction in frequency of isotopic analyses follow-ing power changes may be permi"4ble if justified by the data obtained.
s 3/
PRESSURE / TEMPERATURE LIMITS All
.ents in the Reactor Coolant System are designed to with-stand the effecth of cyclic loads due to system temperature ano pressure changes.
These cyc14c loads are introduced by normal load transients, reactor trips, and stahyp and shutdown operations.
The various categories of load cyclesised for design purposes are provided in Section (
) of the FSAR.*.puring heatup and cooldown, the rates of temperature and pressure changesN re limited so that the maximum speci-fied heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclics peration.
During heatup, the thermal gradients in reactor vessel wall produce thermal stresses which vary from compresshe at the inner wall to tensile at the outer wall. These thermal induced com essive stresses tend to alleviate the tensile stresses induced by the i rnal pressure.
Therefore, a pressure-temperature curve based on steady stal onditions (i.e., no thermal stresses) represents a lower bound of all s1 r
curves for finite heatup rates when the inner wall of the vessel treated as the governing location.
her B&W-STS B 3/4 4-6 September 1, 1976
r.
PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEE 0 WATER SYSTEMS The OPERABILITY of the auxiliary feedwater systems ensures that the Reactor Coolant System can be cooled down to less than (305)*F from normal operating conditions in the event of a total loss of offsite power.
Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of (350) gpm at a pressure of (1133) osig to the entrance of the steam generators. Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of (700) gpm at a pressure of (1133) psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than (305)*F where the Decay Heat Removal System may be placed into operation.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than (305)*F in the event of a total loss of offsite power or of the main feedwater system. The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for (
) hours with steam discharge to atmosphere concurrent with loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 ACTIVITY i
The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction l
of 10 CFR part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.
l 3/4.7.1.5
-fN STEAM LINE ISCLATION VALVES i
The OPERABILITY of the main steam line isolation valves ensures l
that no more than one steam generator will blcwdown in the event of a l
steam line rupture. This restriction is required to 1) minimize the B&W-STS B 3/4 7-2 June 1, 1976
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