ML20044H026
| ML20044H026 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 05/21/1993 |
| From: | Pickett D Office of Nuclear Reactor Regulation |
| To: | Spangenberg F ILLINOIS POWER CO. |
| References | |
| TAC-M77807, NUDOCS 9306070236 | |
| Download: ML20044H026 (45) | |
Text
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t UNITED STATES
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j NUCLEAR REGULATORY COMMISSION N
WASHINGTON, D.C. 20506-0001 J
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May 21, 1993 I
Docket No. 50-461 l
Mr. Frank A. Spangenberg i
Manager - Licensing and Safety j
Clinton Power Station P. O. Box 678 Mail Code V920 Clinton, Illinois 61727 r
Dear Mr. Spangenberg:
t
SUBJECT:
BASES CHANGES TO TECHNICAL SPECIFICATIONS - CLINTON POWER STATION,.
l UNIT NO. 1 (TAC NO. M77807)
Your letter of June 19, 1990 (U-601650), requested a number of changes to the Bases section of the Clinton Power Station Technical Specifications.
In.
accordance with 10 CFR 50.36, the Bases section is not part of the Technical i
Specifications. Therefore, these Bases changes were not submitted as an application for amendment to Facility Operating License NPF-62 since the requirements of 10 CFR 50.90 are not applicable. The proposed changes were categorized to accomplish the following:
(1) correct typographical errors and references to the Final Safety Analysis Report (FSAR) and testing to be performed during initial plant startup; (2) incorporate a revised operating power / flow map for-two reactor recirculation loop operation; (3)- delete information that is no longer applicable to the current ACTION statements l
associated with reactor coolant specific activity; (4) update guidance on the use of the containment purge and exhaust lines; (5) clarify the basis for drywell internal pressure limits; (6) provide additional: guidance'on OPERABILITY of the reactor core isolation cooling system relative to the
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ACTION statements of related Technical Specifications; (7) provide additional i
guidance concerning entry into and compliance with the ACTION statements associated with the Technical Specifications for diesel generators; and (8) clarify the Bases for station battery specific gravity limits to support the use of acceptable replacement battery cells having a different nominal specific gravity.
We have found items (1), (2), (3), (4), (5), and (8) above, acceptable as they
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can be made without affecting the Technical Specifications. Following discussion with your staff, we understand that you wish to withdraw items (6) and (7). The concerns necessitating the proposed wording for items (6) and (7) are resolved with the Improved Standard Technical Specifications (ISTS) for BWR6 facilities.
Since you are planning a full' conversion to the 9
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l Mr. Frank A. Spangenberg !
ISTS (with a submittal scheduled for October 1993), the staff concurs that further review would not be productive.
Our evaluation of these proposed changes is enclosed.
l Sincerely,.
OrigintJ Signed By:
Douglas V. Pickett, Senior Project Manager Project Directorate III-2 l
Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
l 1.
Bases Pages l
2.
Evaluation cc w/ enclosures:
See next page DISTRIBUTION Docket File GHill (2)
HRC & Local PDRs Wanda Jones PDIll-2 p/f CGrimes JRoe ACRS (10)
JZwolinski OPA JDyer OC/LFDCB CMoore BClayton, RIII DPickett OGC DHagan
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Mr. Frank A. Spangenberg ISTS (with a submittal scheduled for October 1993), the staff concurs that further review would not be productive.
Our evaluation of these proposed changes is enclosed.
Sincerely, Origir.rj s:gned By:
t Douglas V. Pickett, Senior Project Manager Project Directorate 111-2 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Enclosures:
1.
Bases Pages 2.
Evaluation cc w/ enclosures:
See next page DISTRIBUTION Docket File GHill (2)
NRC & Local PDRs Wanda Jones PDill-2 p/f CGrimes j
JRoe ACRS (10)
JZwolinski OPA JDyer OC/LFDCB CMoore BClayton, Rll!
DPickett OGC DHagan i
l
- See previous concurrence go
- LA:PD32:DRPW PM:PD32:DRPW
- D:PD32:DRPW
- 0GC l
CMoore DPickett/rc JDyer CMarco 05/12/93
(/2 \\ /93 05/12/93 05/14/93 l
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l Mr. Frank A. Spangenberg l l
ISTS (with a submittal scheduled for October 1993), the staff concurs that j
further review would not be productive.
Our evaluation of these proposed changes is enclosed.
Sincerely, T
v9 l
Douglas V. Pickett, Senior Project Manager Project Directorate III-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
1.
Bases Pages 2.
Evaluation l
cc w/ enclosures:
See next page i
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i Mr. Frank A. Spangenberg Clinton Power Station lilinois Power Company Unit No. I j
CC:
i Mr. J. S. Perry Illinois Department Vice President of Nuclear Safety _
Clinton Power Station Office of Nuclear Facility Safety i
Fost Office Box 678 1035 Outer Park Drive Clinton, Illinois 61727 Springfield, Illinois 62704 Mr. J. A. Miller Mr. Donald Schopfer
'i Manager Nuclear Station Project Manager L
Engineering Department Sargent & Lundy Engineers Clinton Power Station 55 East Monroe Street Post Office Box 678 Chicago,-Illinois; 60603 i
Clinton, Illinois ~ 61727 i
Sheldon Zabel, Esquire Schiff, Hardin & Waite 7200 Sears Tower 233 Wacker Drive i
Chicago, Illinois 60606 Resident Inspector U.S. Nuclear Regulatory Commission i
RR*3. Box 229 A Clinton, Illinois 61727 i
Ms. K. K. Berry Licensing Services Manager General Electric Company 175 Curtner Avenue, M/C 382 7'
San Jose, California 95125
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Regional Administrator, Region III 799 Roosevelt Road, Building 4 j
Glen Ellyn, Illinois 60D.
Chairman of DeWitt County c/o County Clerk's Office DeWitt County Courthouse Clinton, Illinois 61727 i
Mr. Robert Neumann j
Office of Public Counsel State of Illinois Center 100 W. Randolph, Suite'll-300 Chicago, Illinois 60601 l
p u
l' ENCLOSURE 2 l
EVALUATION RELATED TO l
BASES CHANGE RE00EST (U-601650)-
j This evaluation addresses the proposed changes in the categories as presented-by the licensee in their submittal.
q 1.
Correction of Editorial Errors and References to FSAR and Initial Plant -
Startuo Testino The original version of the Bases had a large number of references to-the'.
Final Safety Analysis Report-(FSAR). Title 10 of the Code of Federal.
1 Regulations, Section 50.71(e), requires licensees to update the FSAR and this t
updated version of the Safety Analysis Report is referred to as the USAR. The j
licensee proposes to change all references to the FSAR to reference the.
i Updated. Safety Analysis Report-(USAR).
In addition, the licensee has made several minor editorial, changes to the Bases..The staff finds' these changes acceptable.
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The original version of Bases Section 3/4.2.3 described how the licensee would perform initial startup testing to demonstrate that a large Minimum. Critical Power Ratio (MCPR) margin.would exist below 25 percent of. rated: thermal power.
The licensee has stated that.such testing was performed and that the results were acceptable. Therefore, the licensee.has. proposed to revise this section to reflect that this confirmatory testing is complete. Since this testing has been performed as required, the staff finds these changes acceptable.
l 2.
Incorporation of Revised power / Flow mad for Two Reactor Recirculation'Looo Operation Bases Figure 3/4.2.3-1, " Reactor Operating Map for Two' Recirculation Loop-Operation," was revised by Amendment No.18 to reflect the first refueling t
l with new fuel types and operation in the Maximum Extended Operating Domain.
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i The figure provided in Amendment No. 18 was based on a~ generic BWR/6 design.
The licensee has now proposed a revised figure that is based.on Clinton Power Station's characteristics of_the 105 percent rod line.
In addition, the 40 and 60 percent rod lines have been added.
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.The proposed changes to the Power / Flow map reflect Clinton Power Station's characteristics and improves the accuracy of this Bases'section.
Therefore, i
the staff finds these changes acceptable.
j 3.
Deletion of Information No Loncer Applicable Generic Letter 85-19, " Reporting Requirements on Primary Coolant Iodine l
Spikes," eliminated certain reporting requirements and ACTION requirements for-i limiting plant operation when specific reactor coolant activity levels i
exceeded limits. These cumulative time limitsLand reporting requirements were removed from the Clinton Power Station Technical Specifications prior to issuance of the Low Power Operating License.
However, the discussion of these
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items in the Bases was not deleted. Therefore, the licensee has proposed deletion of these additional items so that the Bases will be consistent with the current ACTION of Technical Specification 3.4.5.
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The staff concludes that these additional operating limits and reporting-requirements are no longer applicable and that their inclusion in the Bases i
section was an oversight. Therefore, the staff finds these deletions acceptable.
l 4.
Update Guidance on Use of Containment Purae and Exhaust Lines l
l The Clinton Power Station has both a 36-inch diameter containment ventilation building purge system and a 12-inch diameter continuous containment purge system.
Supplement 5 to the Safety Evaluation Report (NUREG-0853) described three interim programs implemented at the Clinton Power Station to assess the need for operation of the containment purge systems during Operational Conditions 1, 2, and 3, and to investigate methods to minimize their use consistent with as low as reasonably achievable (ALARA) considerations. -The three programs (i.e., the Containment Purge Operational Data Gathering Program, the Containment Access Management Program, and the Interim Guidelines for Containment Purge Operations) were intended to be utilized during the first operating cycle to gather operational data.
Following the first cycle of operation, data from these programs was to be reported so that the staff could reevaluate the need for containment purging along with the need for any restrictions.
By letter dated April 4, 1989, the licensee submitted the reouired information from these programs and proposed the following containment purging guidelines for Operational Conditions 1, 2, and 3:
1.
The 36-inch and 12-inch containment purge systems shall be isolated while venting the drywell, 2.
the 36-inch and 12-inch containment purge systems shall not be operated l
simultaneously, l
3.
operation of the 36-inch containment purge system shall be limited to 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> per year, 4.
either the 36-inch or 12-inch containment purge system may be operated to maintain the containment building air pressure within the limits of the Technical Specifications, to reduce airborne radioactivity levels in the l
containment building to maintain worker dose ALARA, or for the duration of any welding task or any task in which chemicals requiring ventilation are used, and 5.
continuous operation of the 12-inch containment purge system is allowed except while venting the drywell or while operating the 36-inch containment purge system.
The staff reviewed the licensee's submittal and approved the above purging criteria in a letter dated May 18, 1992.
. The Bases section of the Technical Specifications includes ~a discussion of the interim programs-as described in Supplement 5 to the Safety Evaluation Report :
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along with restrictions on the use of the 12-inch containment purge system.
i The licensee's letter of. June 19,--1990, proposed to delete'these discussions-l and to update this section to include the revised purging criteria. _ Since the proposed changes merely reflect what the staff had previously reviewed and approved, the proposed changes are acceptable.
5.
Clarify Bases for Drywell Internal ' Pressure limits.
The original design basis accident analysis which produced the calculated. peak positive drywell pressure of 18.9 psig was the main steam line break.
Amendment No. 18 to the Clinton Power Station, which; approved operation'in the.
Maximum Extended Operating Domain, identified a higher. calculated peak-positive drywell pressure of 19.7 psig resulting from a recirculation line break.
Although the Bases section for Technical Specification 3/4.6.2.5 were revised in Amendment No.18 to reflect the higher calculated peak drywell--
pressure, they were not revised to reflect that the limiting design basis accident had changed from the main steam line break.to the recirculation line i
break. The licensee's letter proposed revising the Bases section to correctly
-l reflect that the calculated peak positive.drywell pressure of 19.7 psig occurs i
from the recirculation line break.
The staff finds. these changes acceptable.
y 6.
Provide Additional Guidance on OPERABILITY of the Reactor Core Isolation l
Coolina (RCIC) System The licensee's letter of June 19, 1990, identified a-potential startup problem that has resulted from a mismatch in operability requirements of the automatic depressurization system (ADS), the high pressure core. spray (HPCS) system, and l
the reactor core isolation cooling (RCIC) system.
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Technical Specification 3.5.1 states that the HPCS system must be operable in Operational ~ Conditions 1. 2, and 3.
The HPCS system is required to be restored to OPERABLE status within 14 days provided that the RCIC system is l
The specifications do not address operation when both HPCS and RCIC are inoperable.
Such a condition would require entry into' Specification 3.0.3.
Technical Specification 3.3.3 states that the ADS trains must be operable in
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Operational Conditions 1,.2, and 3 when reactor steam dome pressure is greater than 100 psig.
With one train of ADS instrumentation out'of service, a 7 day I
allowed outage time is permitted provided that both RCIC and HPCS are i
If either the HPCS or RCIC is inoperable coincident with an inoperable ADS train, the allowed outage time is reduced to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Technical Specification 3.7.3 states that the RCIC-system must be operable in Operational Conditions 1, 2, and 3 when reactor steam dome pressure is greater than 150 psig.
The RCIC system is required to be restored to OPERABLE status within 14 days when the HPCS system is OPERABLE.
Similar to Specification 3.5.1, Specification 3.7.3 does not address operation with both the HPCS and i
i RCIC system inoperable.
Such a condition would-require entry into Specification 3.0.3.
The RCIC is a steam driven system and the turbint pump requires a minimum steam pressure of 150 psig to be operable.
Due to the.above mismatch in operability requirements, the licensee has identified potential situations when ACTION statements for the ADS or HPCS system rely on the. operability of the RCIC system. An example would be declaring the HPCS system _ inoperable.
during Operational Condition 1, 2, or 3 with a reactor steam dome pressure l
l less than 150 psig. Since the minimum steam pressure did not exist, the RCIC system-would be considered inoperable, thus,. resulting in a plant shutdown.
The licensee has argued that the situation postulated.above would cause unnecessary restrictions on plant operation and unnecessary entry into i
Technical Specification 3.0.3.
The licensee also states that Division I and II low pressure systems will be available and capable of injection if ~needed.
With tnis in mind, the licensee proposed modifying the Bases section of the i
ADS, HPCS, and RCIC systems such that the RCIC system be considered operable
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when the reactor steam dome pressure is less.than 150 psig.
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l The situation described by the licensee is generic to BWR/5 and BWR/6 facilities. The staff notes that alternative operability requirements of the ADS, HPCS and RCIC have been proposed for the Improved BWR/6 Standard Technical Specifications (ISTS) which would avoid the type of operability problems postulated.
While the staff believes that.a 10 CFR 50.90 license i
amendment application revising the operability requirements would be a.more i
appropriate solution available to the licensee, the staff notes that the licensee has committed to make a full conversion to the ISTS (with submittal scheduled for October 1993).
With this in mind, the licensee.has requested to withdraw this portion of their submittal. The staff concurs with the licensee's request in that further review would not be productive.
7.
Provide Additional Guidance Concernina Entry into and Compliance'with the ACTION Reauirements Associated with the Technical Specifications for j
Diesel Generators Current technical specifications state that if any dissel generator becomes inoperable for any reason other than " preplanned preventive maintenance or testing," the operability of the remaining diesel generators must be demonstrated by starting and loading them for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This requirement is a standard item in technical specifications and has been included to identify potential common mode failures.
.Both industry and the NRC have recognized that unnecessary testing of diesel generators contributes to increased wear and potentially reduces their reli ability.
In this regard, the staff has approved technical specificatioi, changes that can reduce unnecessary testing In particular, the staff has approved changes that permit licensees to determine that an individual diesel generator failure is unit specific and does not represent a common mode
- failure.
This allows the licensee to forego starting and running the remaining diesel generators.
i The licensee's letter of June 19, 1990, discussed the conflict between making.
periodic minor repairs (e.g., minor leaks) and " preplanned preventive-maintenance.
The licensee argued that the performance of work to correct a condition, wnich by itself would not' cause the associated diesel generator. to l
be inoperable, should not result in a requirement to demonstrate the:
operability of the remaining diesel _ generators
'Therefore, the licensee proposed to modify the Bases section of the technical specifications to define
" preventive maintenance" as the correction of minor leaks that do not otherwise make the diesel generator inoperable.
The intent and activities to be covered by " preplanned preventive maintenance"-
i has not been well defined.
The staff has questioned whether repair of leaks should be considered to be " preplanned preventive maintenance" or corrective actions (the latter designation requiring operability demonstrations by the remaining diesel generators). However, the staff notes that this' question will be resolved with the licensee's conversion to the ISTS.
The ISTS does.
not include 'such wording, instead...it requires the licensee to make a.
determination regarding the likelihood of common mode. failures. With this in i
mind, the licensee has requested to withdraw this portion of their submittal.
The staff concurs with the licensee's request in that further review would not be productive.
8.
Clarify Bases for Specific Gravity Limits for Station Batteries Technical Specification Table 4.8.2.1-1 identifies the float voltages and specific gravity limits for the battery cells of the safety-related' Division I, II, III, and IV station batteries.
Surveillance requirements periodically verify these parameters for the pilot cells and connected cells. While. Table 4.8.2.1-1 states the specific gravity for each surveillance, the Bases section of the technical specifications references tolerances to.within "the manufacturer's full charge specific gravity."
l The licensee's letter of June 19, 1990, stated that.the safety-related l
Division I, II, III, and IV station batteries are no longer being manufactured by the original supplier.
As loads have been increased since initial site l
operation, the licensee has had to obtain slightly different replacement cells. While calculations demonstrate the acceptability of the replacement cells, they have a slightly different specific gravity. Due to the different specific gravity of the r'eplacement batteries, the wording of the current Bases section is inappropriate. The licensee's proposed changes delete references to "the manufacturer's full charge specific gravity" and replaces it with the specific gravity values of Table 4.8.2.1-1.
In addition, the Bases will include assurances that the values verified during surveillances are those recommended by the manufacturer.
The proposed changes to this Bases section supports the use of acceptable replacement battery cells having a different nominal specific gravity.
The l
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. specific gravity values referenced in the Bases are identical to those identified in the technical specifications.
Since the proposed change represents a more accurate Bases description of the replacement battery cells, the staff finds the changes acceptable, l
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LIMITING SAFETY SYSTEM SETTINGS l
BASES l
2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) i 9.
Scram Discharoe Volume Water level-Hiah The scram discharge volume receives the water displaced by the_ motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod. insertion would be hindered. The reactor is, therefore, tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig-when.they are tripped.
The trip setpoint for each scram discharge volume is t
equivalent to a contained volume of 19.6 gallons of water.
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- 10. Turbine StoD Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.
With a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient.
- 11. Turbine Control Valve Fast Closure. Valve Trio System Oil-Pressure-Low l
The turbine control valve fast closure trip anticipates the pressure, neutron I
flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident failure of the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than'20 msec after the start of control valve fast l
closure. This is achieved by the action of the fast actingLsolenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is' sensed.by pressure switches whose contacts form the two-out-of-four logic input to the Reactor Protection System. This trip setting, a slower closure time, and a different valve characteristic from that of the turbine stop valve combine to produce transients which are very similar to that for the stop valve.
Relevant transient analyses are discussed in Section 15.2.2 of'the Updated Safety Analysis Report (USAR).
12.
Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position provides additional manual reactor i
l trip capability to the manual scram pushbutton switches.
- 13. Manual Scram The manual scram pushbutton switches provide a diverse means for initiating a reactor shutdown (scram) to the automatic protective instrumentation channels and provides manual reactor trip capability.
CLINTON - UNIT I B 2-9 revised by letter dated 5/21/93
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made sub-critical from all operating conditions, with postulated accident conditions are c(2) the reactivity transients associated ontrollable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude i
inadvertent criticality in the shutdown condition.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be per-formed in the cold, xenon-free condition and shall show the core to be sub-critical by at least R + 0.38% a k/k or R + 0.28% a k/k, as appropriate.
The value of R in units of % a k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated begin-ning-of-life core reactivity.
The value of R must be positive or zero and must be determined for each fuel loading cycle.
Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.
The highest worth rod may be determined analytically or by test.
The SHUTOOWN MARGIN is demonstrated by an insequence control rod withdrawal at the begin-ning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure.
Observation of subcriticality in this condition assures subcriticality with the most reactive control rod fully withdrawn.
This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small a careful check on actual conditions to the predicted conditions is necessary,,and the changes in reactivity can be inferred from these comparisons of rod patterns.
Since the comparisons are easily done, frequent checks are not an imposition on normal operations.
A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.
A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.
3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the safety analyses, and (3) limit the potential effects of the rod drop accident.
The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation.
CLINTON - UNIT I B 3/4 1-1
1
-i REACTIVITY CONTROL SYSTEMS r
BASES 3/4.1.3 CONTROL RODS (Continued)
A limitation on inoperable rods is set such that the resultant effect on total.
rod worth and scram shape will be kept to-a minimum.
The requirements for.the.
various scram time measurements ensure that any indication of systematic pro-blems with rod drives will be investigated on a timely basis.
Damage within the control rod drive mechanism could be a generic problem,
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therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the. reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable 'for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are con-sistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shut down for investigation and resolution of the problem.
The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the limiting power transient analyzed. in Section 15.4 of the USAR. This analysis shows that the negative reactivity rates resulting l
from the scram with the average response of all the' drives as given in the specifications, provide the required protection and MCPR remains greater than the fuel cladding safety limit.
The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.
The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.
Control rods with inoperable accumulators are declared inoperable and Specifi-l cation 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may t
still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.
Control rod coupling integrity is required to ensure compliance with the analysis j
of the rod drop accident in the USAR. The overtravel position feature provides l
l the only positive means of determining that a rod is properly coupled and there-
' fore this check must be performed prior to achieving criticality after completing l
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l CLINTON - UNIT 1 B 3/4 1-2 revised by letter 5/21/93
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REACTIVITY CONTROL SYSTEMS l
t i
i BASES 4
t 3/4.1.3-CONTROL RODS (Continued)
CORE ALTERATIONS that could have affected the control rod coupling integrity.
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The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indica-I tion system must be OPERABLE.
s ine wou ve 2
to less than 3 inches in the event of a housing. failure.roo nousing support re i
The amount of rod j
reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure j
to act as a driving force to rapidly eject a drive housing.
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The required surveillance intervals are adequate to determine that the rods are i
OPERABLE and not so frequent as to cause excessive wear on the system components i
3,4.1.4 CONTROL ROD PROGRAM CONTROLS l
The rod withdra-al limiter system input power signal'orginates from the first stage turbine pressure.
When operating with the steam bypass valves open, this signal indicates a core power level which is less than the true core power.
Consequently, near the low power setpoint and high power setpoint of the rod pattern control system, the potential exists for nonconservative control rod J t MrE d s.
Therefore, when operating at a sufficiently high power level, i
basis rod withdrawal error transient.there is a small probability of violating fuel To ensure that fuel Safety Limits are not violated, this specification prohibits control rod withdrawal when a biased i
power signal exists and core power exceeds the specified level.
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Control rod withdrawal and insertion sequences are established to assure that i
the ra ir m insecuence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod i
j drop accident.
The specified sequences are characterized by homogeneous, scat-tered patterns of control rod withdrawal.
When THERMAL POWER is greater than 1
20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Thus requiring the RPCS to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.
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The RPCS provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
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CLINTON - UNIT 1 B 3/4 1-3
l REACTIVITY CONTROL' SYSTEMS
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L BASES i
3/4.1.4 CONTROL R0D PROGRAM CONTROLS (Continued) l
[
The analysis ~of the rod drop accident is presen'ted in Section 15'.4 of the USAR i
and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
i The RPCS is also designed to automatically prevent fuel damage'in the event of erroneous rod withdrawal from locations 'of high power density during higher l
power operation.
l A dual channel system is provided that, above the low power setpoint, restricts the withdra'wal distances of all non-peripheral' control rods.
This
(
restriction is greatest at highest power levels.-
I l
3/J ~.5 STANDBY LIOUID CONTROL SYSTEM i
i i
Tt e standby liquid control system provides a backup capability for bringing-
-i
'ne reactor from full power to a cold, xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated' power pattern.
To meet this i
objective it is necessary to inject a quantity of sodium pentaborate solution which produces a concentration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. :To allow for potential leakage and j
imperfect mixing this concentration is increased by 25%.
The required concentration is achieved by having a minimum available quantity of_
l 3574 gallons of sodium pentaborate solution containing a minimum of 4246 lbs.
of sodium pentaborate.
This quantity of solution is a net amount which is_
above the pump suction, thus allowing for the portion which cannot be' injected.
The pumping rate.of 41.2 gpm per pump provides.a negative i
reactivity injection rate over the permissible pentaborate solution volume l
range, which adequately compensates for the positive reactivity effects due to temperature and xenon during shutdown.
The temperature' requirement is l
necessary to ensure that the sodium pentaborate remains in solution.
i With redundant pumps and explosive injection valves and with a highl.y reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of.
time with one of the redundant components inoperable.
t Surveillance requirements are established on a frequency that assures a high i
reliability of the system. Once the solution is established, sodium pentaborate i
6 1.
C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NED0-10527, March 1972 2.
C. J. Paone, R. C. Stirn and R. M. Young, Supplement I to NED0-10527, July 1972 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cores,"
Supplement 2 to NED0-10527, January 1973 CLINTON - UNIT 1 B 3/4 1-4 revised by letter dated 5/21/93
BASES TABLE-B 3.2.1-1 i
SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS
- Plant Parameters:
[
Core THERMAL POWER....................
3015 MWt** which corresponds to 105% of rated steam flow Vessel Steam Output...................
13.08 x ' 106 lb,/hr which l
corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.............. 1060 psia Design Basis Recirculation Line Break Area for:
a.
Large Breaks 2.2 ft2 b.
Small Breaks.
0.09 fit.
l Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM j
LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft)
FACTOR RATIO Initial and 8x8 a
1.4 1.17***
Reload Cores
- A more detailed listing of input of each model and its source is presented in Section 11 of Reference 1 and Section 6.3 of the USAR.
l
- This power level meets the Appendix K requirement of 102%.. The core heatup j
calculation assumes a bundle power consistent with operation of the highest i
powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION l
RATE limit.
For core flows less than'8S% of rated, _the initial MCPR is ~ taken from the MCPR, l
Curve specified in the CORE OPERATING LIMITS REPORT.
- This value is specified in the CORE OPERATING LIMITS REPORT.
CLINTON - UNIT I B 3/4 2-3 revised by letter dated 5/21/93
?0WER DISTRIBUTION LIMITS GASES 3 /4. 2. 3 MINIMUM CRITICAL POWER RATIO (Continued)
The MCPR,s are established to protect the core from plant transients other than core flow increases, including the localized event such. *nd withdrawal error. The MCPR s were calculated based upon the most limit! y,osient at the given core p,ower level, including feedwater controller and 3,
" ejection 4
transients. For core power below 40% of RATED THERMAL POWER'where the EOC-RPT and reactor scram on turbine stop valve closure and turbine control valve fast closure are bypassed, separate sets of MCPR limits are provided for high and low core flows to account for the sensitivily to initial core-flows. For core power above 40% of RATED THERMAL POWER, bounding MCPR, limits were developed.
At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum'retirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR.
evaluation was made below 25% of RATED THERMAL POWER level with minimum recirculation pump speed which demonstrated that future MCPR evaluation below this power level is unnecessary.
The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power dis-tribution to stabilize. The requirement for calculating MCPR after initially i
determining a LIMITING CONTROL R0D PATTERN exists ensures that MCPR will be l
known following a change in THERMAL PbWER or power shape that could place operation exceeding a thermal limit.
1 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat _ generation even if fuel pellet j
densification is postulated.
The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shift. Calculating LHGR after initially determining a LIMITING CONTROL R0D PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.
1 CLINTON - UNIT 1 B 3/4 2-5 revised by letter dated 5/21/93
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i BASES FIGURE B 3/4.2.3-1 REACTOR OPERATING MAP FOR TWO RECIRCULATION LOOP OPERATION CLINTON - UNIT I B 3/4 2-7 revised by letter dated 5/21/93 i
i l
i INSTRUMENTATION BASES 3/4.3.2 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM (Continued) each case which in turn determines the valve speed in conjunction with the 13 second delay.
It follows that checking the valve speeds and the 13 second time for emergency power establishment will establish the response time for the isolation functions.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The Trip Setpoint and Allowable Value also contain additional margin for instrument accuracy and calibration capability.
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.
This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection.
Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one sysicm at the same time.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The Trip Setpoint and Allowable Value also contain additional margin for instrument accuracy and calibration capability.
The emergency core cooling system (ECCS) pump minimum flow instruments-are provided to ensure that ECCS pump minimum flow paths are preserved to prevent pump damage in the event that ECCS pumps are started without reactor or test line flow paths. The minimum flow instruments are not part of ECCS actuation instrumentation.
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.
The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971 and NEDO-24222, dated December 1979, and Section 15.8 of the USAR.
The end-of-cycle recirculation pump trip (E0C-RPT) system is an essential safety supplement to the Reactor Protection System.
The purpose of the E0C-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a CLINTON - UNIT 1 B 3/4 3-3 revised by letter dated 5/21/93
INSTRUMENTATION I
l BASES l
t 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued) faster rate than the control rods add negative scram reactivity.
Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events.
The two events for which the EOC-RPT protective _ feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
i l
A fast closure sensor from each of the turbine control valves provides input to the four RPS logic divisions of the EOC-RPT system.
Similarly, a position switch for each of the turbine stop valves provides input to the four logic divisions of the EOC-RPT system.
For each EOC-RPT system, the_ sensor relay con-tacts are arranged to form a 2-out-of-4 logic for the fast closure of turbine I
control valves and a 2-out-of-4 logic for.the turbine stop valves.
The opera-tion of either logic will actuate the EOC-RPT system snd trip both recirculation pumps.
Each EOC-RPT system may be manually bypassed by use of a keyswitch which is adTini stratively controlled.
The manual bypasses and the automatic Operating Bypass at less than 40% of RATED THERMAL POWER are annunciated in the control room.
The EOC-RPT system response time is the time' assumed in the analysis between i
initiation of valve motion and complete suppression of the electric arc, i.e.,
140 ms.
Included in this time are:
the time from initial valve movement to reaching the trip setpoint; the response time of the sensor; the response time of the system logic and the time alloted for breaker arc suppression.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The Trip Setpoint and Allowable Value also contain additional margin for instrument accurac', and calibration capability.
3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isola-tion from its primary heat sink and the loss of _ feedwater flow to the reactor vessel.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The Trip Setpoint and Allowable Value also contain additional margin for instrument accuracy and calibration.
)
CLINTON - UNIT 1 B 3/4 3-4
3 /4. 4 REACTOR COOLANT SYSTEM l
BASES t
3/4.4.1 REC 1RCULATION SYSTEM i
The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively, MAPLHGR limits are decreased in accordance with the values specified in the CORE OPERATING LIMITS REPORT, and MCPR operating i
limits are adjusted in accordance with the values specified in the CORE j
OPERATING LIMITS REPORT.
Additionally, surveillance on the volumetric flow rate of the operating recir-4 culation loop is imposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below 30%*
i THERMAL POWER or 30%* rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump, and vessel bottom head during the extended operation of the single recirculation loop mode.
An inoperable jet pump is not, in itself, a sufficient reason-to declare a re-i circulation loop inoperable, but it does, in case of a design-basis-accident, l,
increase the blowdown area and reduce the capability of reflooding the core';
thus, the requirement for shutdown of the' facility with a jet pump inoperable.
Jet pump failure can be detected by monitoring jet pump performance on a pre -
scribed schedule for significant degradation.
Significant; degradation is
}
indicated if more than one of three specified surveillances performed confirms unacceptable deviations from established patterns or relationships. The t
surveillances, including the associated acceptance criteria, are in accordance with General Electric Service Information Letter No. 330, the recommendations i
of which are considered acceptable for verifying jet pump operability according to NUREG/CR-3052, "Closecut of IE Bulletin 80-07:
BWR Jet Pump Assembly Failure." Performance of the specified surveillances, however, is not required when thermal power is less than 25% RATED THERMAL POWER because flow oscillations and jet pump noise precludes the collection of repeatable meaningful data during low flow conditions approaching the threshold response of the associated flow instrumentation.
2 Recirculation loop flow mismatch limits are in compliance with ECCSLOCA i
analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
In the case where the mismatch limits cannot be maintained 3
1 during two loop operation, continued operation is perditted in a single recirculation loop mode.
In order to prevent-undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop.
The loop temperature must also be within 50*F of-the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
Sudden equilization of a temperature difference > 100*F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.
- The threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.
CLINTON - UNIT 1 B 3/4 4-1 revised by letter dated 5/21/93-
REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECTRCULATION SYSTEM (Continued)
The objective of GE BWR plant and fuel design is to provide stable operation with margin over the normal operating domain. However, at the high power / low flow corner of the operating domain, a small. probability of neutron flux limit cycle oscillations exists depending on combinations. of-operating conditions (e.g., rod pattern, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron.
flux noise levels should be monitored while operating in this ~ region.
Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.
Plant specific-calculations can be performed to determine an applicable region for monitoring neutron flux noise levels.
In this case the degree of conservatism can be reduced since plant to plant variability would be eliminated.
In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.
Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations.
BWR cores typically operate with neutron flux noise caused by random boiling and flow noise.
Typical neutron flux noise levels of 1-12% of rated power- (peak-to-peak) have been reported l
for the range of low to high recirculation loop flow during both si_ngle and dual recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence.
In addition, stability tests at operating BWRs have demonstrated that when stability related neutron flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5-10 times the typical valbes.
Therefore, actions taken to reduce neutron flux noise 1.evels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux osciliations.
Typically, neutron flux noise levels show a gradual ihcrease in absolute magnitude as core flow is increased (constant control rod pattern) with two reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows.
To maintain a reasonable variation between the low flow and high flow end of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation. Data from tests and operating plants indicate-that a range of 20% of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two. recirculation loops. Baseline data should be taken near the maximum rod line at which the majority of operation will occur. !iowever, baseline data taken at low rod lines (i.e. lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow.
CLINTON - UNIT 1 B 3/4 4-2 revised by letter dated 5/21/93
l i
REACTOR COOLANT SYSTER BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)
The recirculation flow control valves provide regulation of individual recir :
culation loop drive flows; w*ich, in turn, will vary the flow rate of coolant through the reactor core ovei a range consistent with the rod pattern and re-circulation pump speed. The recirculation flow control system consists of the electronic and hydraulic components necessary for the positioning of the two'-
hydraulically actuated flow control valves.. Solid state control logic will genera.te a flow control. valve " motion inhibit" signal in response to any one of several hydraulic power unit or analog control circuit failure signals.
The " motion inhibit" signal causes hydraulic power unit shutdown and hydraulic isolation such that' the flow control valve fails "as is."
This design feature insures that the flow control valves do not respond to potentially erroneous control signals.
Electronic limiters exist in the position control; loop of each flow control valve to limit the flow control valve stroking rate to'101% per.second in opening and closing directions on a control signal failure.
The analysis of the recirculation flow control failures on. increasing and decreasing flow are presented in Sections 15.3 and 15.4 of the USAR respectively.
l The required surveillance interval is adequate-to ensure that the flow control' valves remain OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig in accordance with the ASME Code.
A total of 11 OPERABLE safety-relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient..Any combination of 5 SRVs operating in the relief mode and 6 SRVs operating in the safety mode is acceptable.
Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 4.0.5.
The low-low set system ensures that safety / relief valve discharges are minimized for a second opening of'these valves, following any overpressure transient. This is achieved by automatically lowering the closing setpoint of 5 valves and lowering the opening setpoint of 2 valves following the initial opening.
In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced.
Sufficient redundancy is provided for the low-low set. system.such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis.
I CLINTON - UNIT I B 3/4 4-3 revised by letter dated 5/21/93 1
- = -
REACTOR COOLANT SYSTEM BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE i
3/4.4.3.1 LEAKAGE DETECTION SYSTEMS i
The RCS leakage detection systems required by this specification are provided to' monitor and detect leakage from the reactor coolant pressure boundary. With t
certain exceptions as noted in the Clinton. Power Station Updated Safety Analysis Report, these detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Presscre Boundary Leakage Detection Systems," May 1973.
Except for.the drywell particulate and gaseous radioactivity monitors,. the systems provide the ability to measure leakage from fluid systems in the drywell. The drywell sump flow monitoring system consists of the drywell floor drain sump flow monitoring subsystem and the drywell equipment drain sump- -
flow monitoring sub:ystem. OPERABILITY of each of these subsystems requires-that the applicable portion of the monitoring subsystem associated with the v-notched -
weir box be OPERABLE. Other portions of the subsystem, including the sump pump control circuit and the associated timer, cycle counter and level switches, may-
'3 be utilized as appropriate to provide an alternate.means of monitoring and determining UNIDENTIFIED or IDENTIFIED leakage under the provisions of the associated ACTION statements for the respective subsystem.
-3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on i
the predicted and experimentally observed behavior of cracks in' pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also con-sidered. The evidence obtained from experiments suggests that_ for leakage some-what greater than that specified for UNIDENTIFIED LEAKAGE the probability is i
small that the imperfection or crack associated with such leakage would grow l
rapidly. With respect to IGSCC-related cracks in service sensitive-austenitic stainless steel piping however, an additional limit on the allowed increase in UNIDENTIFIED LEAKAGE (within a 24-hour period or less) is imposed in accordance with Generic Letter 88-01, "NRC Postion on IGSCC in BWR Austenitic Stainless Steel Piping," since an abrupt increase in the UNIDENTIFIED LEAKAGE could be indicative of leakage -from such a source.
In all cases, if the-leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action. The reactor will also be shut' down if an increase in f
UNIDENTIFIED LEAKAGE exceeds the specified limit and the source of increased leakage cannot be isolated or it cannot be determined within a short period of time that the source of increased leakage is not associated with austenitic stainless steel.-
l The Surveillance Requirements for RCS pressure isolation. valves provide added.
i assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
CLINTON - UNIT I B 3/4 4-4 Amendment No. M,65 AUG21199g f
i
l l
l i
BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent interim limits based upon a parametric j
evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation.
l The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 micro-curies per gram DOSE EQUIVALENT I-131, but less than or equal to l
4.0 microcuries per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking which may occur following changes in THERMAL POWER.
Closing the main steam line isolation valves prevents the release of activity i
l to the environs should a steam line rupture occur outside containment.
The surveillance requirements provida adequate assurance that excessive l
specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section 3.9 of the USAR.
During startup l
and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASRE Code Section III, Appendix G.
The curves are based on the RT and stress intensity factor information for the reactor vessel componen,ts.
Fracture toughness limits and j
the basis of compliance are more fully discussed in USAR subsection 5.3.1.5 l
entitled " Fracture Toughness."
l CLINTON - UNIT 1 B 3/4 4-5 revised by letter dated 5/21/93
REACTOR COOLANT SYSTEM BASES 3 /4.4,6 PRESSURE / TEMPERATURE LIMITS (Continued) i The reactor vessel materials have been tested to determine their initial RT,.
The results of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT., of the core beltline region. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.
1 The pressure / temperature limit curvs Figure 3.4.6.1-1, curves A, B, and C, includes an assumed shift in RT., for the conditions at 12 Effective Full Power Years. The actual shift in RT of the vessel material will be established periodically during operation @ removing and evaluating, in accordance with ASTM E185 and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens can be used to predict reactor vessel material transition temperature shift. Flux wires which were removed after the first fuel cycle and will be removed at later intervals with the surveillance specimens are analyzed and provide an improved neutron fluence estimate for the reactor vessel. This data is then used to modify Bases Figure B 3/4.4.6-1 and predictions of reactor vessel material transition temperature shift per Regulatory Guide 1.99, Revision 2. The operating limit curves of Figure 3.4.6.1-1 have been and will be adjusted, as re: Wired, on the basis of the specimen data and the recommendations of Regulatory Guide 1.99, Revision 2.
The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C and A for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.
3/4.4,7 MAIN STEAM LINE ISOLATION VALVES l
Double isolation valves are provided on each of the main steam lines to minimize l
the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment; however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve.
l The maximu:a closure time has been selected to contain fission products and r
to ensure the core is not uncovered following line breaks. The minimum closure l
time is consistent with the assumptions in the safety analyses to prevent pressure surges.
i P
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l CLINTON - UNIT 1 B 3/4 4-6 Amen g t Ng. g 68
3/4.5 EMERGENCY CORE C00LINC SYSTEM l
BASES 3/4.5.1 AND 3/4.5.2 ECCS - OPERATING AND SHUTDOWN ECCS division 1 consists of. the.1ow pressure core spray ;ystem and low pressure coolant injection subsystem."A" of the RHR system and the automatic depressuriza-=
tion system (ADS) as actuated by ADS trip system "1".
ECCS division 2 consists j
of low pressure coolant injection subsystems "B" and "C" of-the RHR system and '
the automatic depressurization system as actuated by ADS trip. system "2".
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The low pressure core spray'(LPCS). system is provided to assure that the core
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is adequately cooled folowing a loss-of-coolant accident and, together with the-
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LPCI system, provides adequate core cooling capacity for all break sizes up toj j
and including the double ended reactor recirculation line break, and for~ smaller breaks following depressurization by the ADS.
The LPCS is a primary source of emergency core cooling aftercthe' reactor vessel q
a is depressurized and a source for.. flooding of~the core in_ case of accidental draining.
1 The surveillance requirements provide adequate assurance that the LPCS system will be OPERABLE when required.
Although all active components are testable-and full flow can be demonstrated by recirculation through~a test loop.during a
reactor operation, a complete functional test requires. reactor shutdown.. The pump disenarge piping is maintained: full to prevent water hammer damage to piping and to start cooling at the earliest moment.
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The low pressure coolant injection ~(LPCI) mode of the RHR system is provided to.
assure that the core is adequately cooled following a loss-of-coolant accident.
The LPCI system, together with the LPCS system,~ provide, adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.
l The surveillance requirements provide adequate assurance that the LPCI system j
will be OPERABLE when required..Although all-active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete. functional test requires. reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
i ECCS division 3 consists'of the high pressure core spray system. - The high pres-l sure core spray (HPCS) system is provided to assure that the reactor core-is.
adequately cooled to limit fuel clad. temperature in the event of a small break j
in the reactor coolant system and loss of ~ coolant which does not result'in rapid depressurization.of the reactor vessel. The HPCS system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory 4
until the vessel is depressurized.
The HPCS system operates over a range of 1177 psid, differential' pressure between reactor vessel' and HPCS suct:fon source, to O psid.
CLINTON - UNIT 1 B 3/4 5-1 1
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EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING AND SHUTDOWN (Continued)
The capacity of the system is selected to provide the required core cooling.
l The HPCS pump is designed to deliver greater than or equal to 467/1400/5010 gpm at differential pressures of 1177/1147/200 psid.
Initially, water from the reactor core isolation cooling (RCIC) tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the RCIC tank water.
With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.
In addition, the reactor core isolation cooling system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition.
The HPCS out-of-service period of 14 days, as specified in the corresponding ACTION statement, is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.
The surveillance requirements provide adequate assurance that the HPCE system will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage.
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Upon failure of the HPCS system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200*F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.
This pressure is substantially below that for which the low i
pressure core cooling systems can provide adequate core cooling for events requiring ADS.
ADS automatically controls seven selected safety-reli.ef valves although the safety analysis only takes credit for six valves.
It is therefore appropriate to permit one valve to be out-of' service for up to 14 days without materially reducing system reliability.
3/4.5.3 SUPPRESSION POOL The suppression pool is required to be OPERABLE as part of the ECCS to ensure j
that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA.
This limit en suppression pool minimum water i
volume ensures that sufficient water is available to permit recirculation l
cooling flow to the core. The OPERABILITY of the suppression pool in l
OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.3.1.
CLINTON - UNIT 1 B 3/4 5-2 revised by letter dated 5/21/93
l 3.4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radia-tion doses to within the limits of 10 CFR Part.100 during accident conditions.
L 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volne will not exceed the value assumed in the accident analyses at the peak accident pressure of 9.0 psig, Pa.
As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La during performance of the periodic tests to account for possible degra-i l
dation of the containment leakage barriers between leakage tests.
Operating tx;,erience with the main steam line isolation valves has indicated l
that degrauet'on has occasionally occurred in the leak tightness of the valves-therefore the. special requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J to 10 CFR 50 (with the exception of exemption (s) j granted for main steam isolation valve leak testing).
j 3 /4. 6.1. 3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the con-tainment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2.
The speci-fication makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor oper-ation.
Only one closed door in each air lock is required to maintain the inte-grity of the containment.
The surveillance testing for measuring leak rate for the containment air locks is consistent with the requirements of Appendix J to 10 CFR 50 with the excep-tion of exemption (s) granted for the containment air lock leak testing.
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CLINTON - UNIT 1 B 3/4 6-1
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CONTAINMENT SYSTEMS BASES 3 /4. 6.1. 4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main.
steam line isolation valves in the postulated LOCA situations would be a small l
fraction of the 10 CFR 100. guidelines, provided the main steam line system from the isolation valves up to and including the MSIV-LCS motor operated boundary valve remains intact. Operating experience has indicated that degradation has occasionally occurred in the leaktightness of the MSIV's.such that the specified leakage requirements have not always been maintained continuously. The requirement for the leakage' control system will reduce the untreated leakage from the MSIV's when isolation of the primary system and containment is required.
3/4.6.1.5 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that.the. structural integrity'of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that-the containment.will withstand the maximum pressure of 15 psig in the event of a steam line break accident. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3 /4. 6.'l. 6 CONTAINMENT INTERNAL-PRESSURE The limitations on containment to secondary containment differential pressure ensure that the containment peak calculated pressure of 9.0 psig does not exceed the design pressure of 15.0 psig during design basis steam line break conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 3.0 psid. The limit of -0.25 to +0.25 psid for initial containment to secondary containment pressure will limit the containment pressure to 9.0 psid which is less than the design pressure and is consistent with the safety analysis for containment design pressure.
3/4.6.1.7 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE t
The limitation on containment average air temperature ensures that the
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containment peak air temperature does not exceed the design temperature of 185'F during steam line break conditions and is consistent with the safety analysis.
I CLINTON - UNIT 1 B 3/4 6-2 Lrevised by letter dated 5/21/93-s i
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CONTAINMENT SYSTEMS BASES 3/4.6.1.8 CONTAINMENT BUILDING VENTILATION AND PURGE SYSTEMS The 36-inch containment purge supply and axhaust isolation valves have permanently installed blocking devices'to restrict their opening to 50' during plant OPERATIONAL CONDITIONS 1, 2, and 3, since these valves.have not been demonstrated capable of closing from the full open position during an accident.
Maintaining these valves blocked during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system.
To provide assurance that the 36-inch valves cannot be inadvertently fully opened, they are blocked in accordance with staff's recommendations accepted in SSER 5, paragraph 6.2.4.1.
The use of the containment purge lines is restricted ~to the 12-inch purge supply and exhaust isolation valves since, unlike the 36-inch valves, the 12-inch valves close during accident conditions and therefore the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during purging operations.
The design of the 12-inch purge supply and exhaust isolation valves meets the requirements of Branch Technical' Position CSB 6-4, " Containment Purging During Normal Plant Operations."
The use of the 12-inch containment purge exhaust and supply lines shall be in accordance with the "Clinton Power Station Report on Containment Purge Operational Data Gathering and Evaluation Program and Proposed Containment Purge Criteria" provided in Illinois Power (IP) letter U-601410, dated April 4, 1989.
Section 6 of the report provides the criteria for governing operation of the containment building ventilation (36-inch) and the continuous purge (12-inch) systems.
The criteria balance ALARA dose to the worker with protection of the health and safety of the public.
Since continuous operation of the 12-inch containment purge system has been shown to be required to support access to the containment to perform Technical Specification surveillances during normal conditions while in OPERATIONAL CONDITIONS 1, 2 and 3, continuous operation of the 12-inch system is allowed except while venting the drywell for pressure control or while operating the 36-inch system.
Continuous containment purge using the 36-inch containment building ventila-tion system is limited to only OPERATIONAL CONDITIONS 4 and 5.
Intermittent use of the 36-inch system during OPERATIONAL CONDITIONS 1, 2, and 3 is permitted only for the purpose of reducing airborne activity levels, or containment pressure, and atmosphere control (excluding temperature and humidity), and shall be limited to the time limits identified in Specification 3.6.1.8.
Leakage integrity tests with a maximum allowable leakage rate for 36-inch supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage failures develop.
The 0.60 La leaking limit should not be exceeded when the CLINTON - UNIT 1 B 3/4 6-3 revised by letter dated 5/21/93
CONTAINMENT SYSTEMS BASES 3/4.6.1.8 CONTAINMENT BUILDING VENTILATION AND PURGE SYSTEMS (Continued) leak rate determined by the leakage integrity tests of these valves is added to the previously determined total for all valves and penetrations subject to Type B and C tests.
The limitations placed on the use of the containment building ventilation system during OPERATIONAL CONDITIONS 1, 2, and 3 are appropriate to minimize the poten-tial release of radioactive gases to the environment following a LOCA during normal plant operation.
The operation of the 36-inch system is consistent with the requirements of Branch Technical Position CSB 6-4, " Containment Purging During Normal Plant Operations."
3/4.6.2 DRYWELL 3/4.6.2.1 DRYWELL INTEGRITY Drywell integrity ensures that the steam released for the full spectrum of dry-well pipe breaks is condensed inside the primary containment either by the sup-pression pool or by containment spray.
By utilizing the suppression pool as a heat sink, energy released to the containment is minimized and the severity of the transient is reduced.
3/4.6.2.2 DRYWELL BYPASS LEAKAGE The limitation on drywell byoass leakage rate ensures that the maximum leakage which could bypass the suppression pool during an accident would not result in the containment exceeding its design pressure of 15.0 psig.
The integrated drywell leakage value is limited to 10% of the design drywell leakage rate.
The limiting case accident is a very small reactor coolant system break which will not automatically result in a reactor depressurization.
The long term dif-ferential pressure created between the drywell and containment will result in a significant pressure buildup in the containment due to this bypass leakage.
3/4.6.2.3 DRYWELL AIR LOCKS The limitations on closure for the drywell air locks are required to meet the restrictions on DRYWELL INTEGRITY and the drywell leakage rate given in Speci-fications 3.6.2.1 and 3.6.2.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.
Only one closed door in each air lock is required to maintain the integrity of the drywell.
CLINTON - UNIT 1 B 3/4 6-4 Q
CONTAINMENT SYSTEMS BASES 3/4.6.2.4 DRYWELL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the drywell will be maintained comparable to the original design specification for the life of the unit. A visual inspection in conjunction with Type A leakage tests is suffi-cient to demonstrate this capability.
l 3/4.6.2.5 DRYWELL INTERNAL PRESSURE The limitations on drywell-to-containment differential pressure ensure that the drywell peak calculated pressure of 19.7 psig does not exceed the design pressure of 30.0 psig and that the containment peak pressure of 9.0 psig does l
not exceed the design pressure of 15.0 psig during LOCA conditions. The maxi-l mum external drywell pressure differential is limited to 0.2'psid, which is l
well below the pressure at which suppression pool water will be forced over l
the weir wall and into the drywell. The limit of I.0 psid for initial l
l positive drywell to containment pressure will limit the drywell pressure to 19.7 psid which is less than the design pressure and is consistent with the safety analysis to limit drywell internal pressure.
3/4.6.2.6 DRYWELL AVERAGE AIR TEMPERATURE l
The limitation on drywell average air temperature ensures that peak drywell temperature does not exceed the design temperature of 330*F during LOCA condi-tions and is consistent with the safety analysis.
3/4.6.2.7 DRYWELL VENT AND PURGE The drywell purge system must be normally maintained closed to eliminate a potential challenge to containment structural integrity due to a steam bypass of the suppression pool.
Intermittent venting of the drywell is allowed for pressure control during OPERATIONAL CONDITIONS I, 2, and 3, but the cumulative time of venting is limited to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per 365 days.
Venting of the drywell is prohibited when the 12-inch continuous containment purge system or the 36-inch l
containment building ventilation system supply or exhaust valves are open.
l This eliminates any resultant direct leakage path from the drywell to the i
environment.
In OPERATIONAL CONDITIONS I, 2 and 3, the drywell isolation valves (IVQ002, IVQ003) have permanently installed blocking devices so as not to open more than 50*.
This assures that the valve would be able to close against drywell pressure buildup resulting from a LOCA.
I Operation of the drywell vent and purge 24-inch supply and exhaust valves during plant OPERATIONAL CONDITIONS 4 and 5 is unrestricted, and the l
l cumulative time for vent and purge operation is unlimited.
CLINTON - UNIT I B 3/4 6-5 revised by letter dated 5/21/93 l
CONTAINMENT SYSTEMS BASES 3/4.6.3 DEpRESSURIZATION SYSTEMS The specifications of this section ensure that the drywell and containment pressure will not exceed the design pressure of 30 psig and 15 psig, respectively, during primary system blowdown from full operating pressure.
The suppression pool water volume must absorb the ' associated decay and structural sensible heat released during a reactor blowdown from 1040 psia'.
Using conservative parameter inputs, the maximum calculated containment pressure during and following a design basis accident is below the containment design pressure of 15 psig.
Similarly the 'drywell pressure remains below-the design pressure of 30 psig.
The maximum-and minimum water volumes.for the suppression pool are 150,230 cubic feet and 146,400 cubic feet, respectively.
i These values include the water volume of the containment pool, horizontal' vents, and weir annulus.
Testing in the Mark III Pressure Suppression Test Facility and analysis have assured that the suppression pool temperature will not rise above'185'F,for the full range of break sizes.
Should it be necessary to make the suppression pool inoperable, this shall only be done as specified in Specification 3.5.3.
Experimental data indicates that effective steam condensation without excessive load on the containment pool walls will occur with a quencher. device and pool temperature below 200*F during relief valve operation.
Specifica-tions have been placed on the envelope of reactor operating conditions to assure the bulk pool temperature does not rise above 185'F in compliance with the containment structural design criteria.
In addition to the limits on temperature of the suppression pool water, operating procedures define the action to be;taken in the event a safety-relief valve inadvertently opens or sticks open.
As-a minimum this action shall include:
(1) use of all available means to close the valve,,
(2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy inserpion to the pool.
The containment spray system consists'of two 100% capacity trains, each with two spray rings located at different ' elevations about the inside circumference of the containment.
RHR A pump supplies one train and RHR pump B supplies the other. RHR pump C cannot supply the spray system. Dispersion of the flow of water is effected by 251 nozzles in each train, enhancing the condensation of-l water vapor in the containment volume and preventing overpressurization.
Heat rejection is through the RHR heat exchangers. The turbulence caused by the spray system aids in mixing the containment air volume to maintain a homogeneous mixture for H control.
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CLINTON - UNIT 1 B 3/4 6-6 re' vised by letter dated 5/21/93 i
CONTAINMENT SYSTEMS BASES 3/4.6.3 DEPRESSURIZATION SYSTEMS (Continued)
The suppression pool cooling function is a mode of the RHR system and functions as part of the containment heat removal system.
The purpose of the system is to ensure containment integrity following a LOCA by preventing excessive containment pressures and temperatures.
The suppression pool cooling mode is designed to limit the long term bulk temperature of the pool to 185*F considering all of the post-LOCA energy additions.
The suppression pool cooling trains, being an integral part of the RHR system, are redundant, safety-related component systems that are initiated following the recovery of the reactor vessel water level by ECCS flows. from the RHR system. Heat rejection to the standby service water is accomplished in the RHR heat l
exchangers.
The suppression peo! make-up system provides water from the upper containment pool to the suppression pool by gravity flow through two 100% capacity dump lines following a LOCA. The quantity of water provided is sufficient to account for all conceivable post-accident entrapment volumes, ensuring the long term energy sink capabilities of the suppression pool and maintaining the l
water coverage over the uppermost drywell vents.
The minimum freeboard distance above the suppression pool high water level to the top of the weir l
wall is adequate to preclude flooding of the drywell in the event of an inadvertent dump.
During refueling, neither automatic nor manual action can open the make-up dump valves.
3/4.6.4 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the contain-I ment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressuriza-tion of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A to 10 CFR 50 "and the requirements of NUREG-0660 as i
l clarified by NUREG-0737 as described in the USAR, Appendix D, item II.E.4.2 l
(Containment Isolation Dependability)." OPERABILITY of the automatic containment isolation valves which isolate the reactor water cleanup system suction containment penetration is also required in OPEPATIONAL CONDITION 5 with any control rod withdrawn to ensure that, in the event initiation of the standby liquid control system becomes necessary, the sodium pentaborate solution is not removed from the reactor coolant system.
The requirements for OPERABILITY of these valves agree with those conditions for which the standby liquid control system is required to be OPERABLE per Specification 3/4.1.5.
The opening of locked or sealed closed containment isolation valves on an l
intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication l
with the control room, at the valve controls, (2) instructing this operator to
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close these valves in an accident situation, and (3) assuring that I
environmental conditions will not preclude access to close the valves and that j
this action will prevent the release of radioactivity outside the containment.
i Measurement of the closure time of automatic containment isolation valves is l
performed for the purpose of demonstrating PRIMARY CONTAINMENT INTEGRITY and system OPERABILITY (Specification 3/4.6.1).
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CLINTON - UNIT 1 B 3/4 6-7 revised by letter dated 5/21/93
l CONTAINMENT SYSTEMS t
BASES L;
3/4.6.4 CONTAINMENT ISOLATION VALVE.S (Continued) l The Maximum Isolation Times (MIT) for primary containment automatic isolation valves listed in this specification are either the analytical times used in--
the accident analysis; described in the USAR; or times derived by applying l
margins to the test data obtained by performing testing in accordance with the i
Inservice Testing program (IST) outlined in Section XI of the ASME Code.. For non-analytical automatic primary containment-isolation valves, the.MIT is derived as follows:
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Valves with full stroke times less than or equal to 10 seconds, l
MIT - Initial Base Line Time X 2 2)
Valves with full stroke time greater tha'n 10 seconds',
MIT = Initial Base Line Time X 1.5.
CLINTON - UNIT 1 B 3/4 6-7a revised by. letter dated 5/21/93
CONTAINMENT SYSTEMS BASES 3/4.6.5 DRYWELL POST-LOCA VACUUM RELIEF VALVES Drywell vacuum relief valves are provided on the drywell to pass sufficient quantities of gas from the containment to the drywell to prevent an excess negative pressure from developing in the drywell.
3/4.6.6 SECONDARY CONTAINMENT The secondary containment completely encloses the primary containment, except for the upper personnel hatch.
It consists of the fuel building, gas control boundary, and portions of the auxiliary building enclosed by the extension of the gas control boundary and the ECCS cubicles and areas as described in USAR l
Figure 6.2-132.
The standby gas treatment system (SGTS) is designed to achieve and maintain a negative 1/4" W.G. pressure within the secondary containment following a design basis accident. This design provides for the capture within the secondary containment of the radioactive releases from the primary containment, and their filtration before release to the atmosphere.
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Establishing and maintaining a vacuum in the secondary containment with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
The inleakage vahes are not verified in the surveillances since no credit for dilutien was taken in the dose calculation. As noted however, adequate drawdown is verified once per 18 months.
The acceptance criteria specified in Figure 4.6.6.1-1 for the drawdown test is based on a computer model, verified by actual performance of drawdown tests, in which the drawdown time determined for accident conditions is adjusted to account for performance of the test during normal plant conditions.
The acceptance criteria indicated per Figure 4.6.6.1-1 is based on conditions corresponding to power operation (with the turbine building ventilation system in operation) and wind speeds less than or equal to 10 mph.
The acceptance criteria for plant conditions other than those assumed will be adjusted as necessary to reflect the conditions which exist during performance of the surveillance test.
The OPERABILITY of the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage.
The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.
Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31-day period is sufficient l
to reduce the buildup of moisture on the absorbers and HEPA filters.
i 3/4.6.7 ATMOSPHERE CONTROL The OPERABILITY of the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydrogen con-I centration within the containment below its flammable limit during post-LOCA 1
CLINTON - UNIT 1 B 3/4 6-8 revised by letter dated 5/21/93
o 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 SHUTDOWN SERVICE WATER SYSTEM i
The OPERABILITY of the shutdown service water system ensures that sufficient cooling capacity is available for continued operation of safety-related equip-l ment during accident conditions. The redundant cooling capacity of these sys-l tems, assuming a single failure, is consistent with the assumptions used in i
the accident analyses within acceptable limits.
l The ultimate heat sink (VHS) specification ensure that sufficient cooling i
capacity is available for continued operation of safety-related equipment for l
at least 30 days to permit safe shutdown and cooldown of the reactor.
The l
j surveillance requirements ensure that quantities maintained are consistent l
with the assumption used in the accident analysis as described in the USAR and l
the guidance provided in Regulatory Guide 1.27, January 1976.
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3/4.7.2 CONTROL ROOM VENTILATION SYSTEM i
l The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all design basis accident conditions.
Continuous operation of the system with-the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31 day period is sufficient to reduce the buildup of-moisture on the adsorbers and HEPA filters.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation-exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix "A",10 CFR 50.. Surveillance testing provides assurance that system and component performances continue to be in accordance with performance specifications for Clinton Unit 1, including applicable parts of ANSI N509-1980.
3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the Emergency Core Cooling System equipment.
The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig.
This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring the RCIC system.
The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the l
primary (non-ECCS). source of emergency core cooling when the reactor is i
pressurized.
With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCS system and justifies the specified 14 day out-of-service period.
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CLINTON - UNIT 1 B 3/4 7-1 revised by letter dated 5/21/93 4
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PLANT SYSTEMS BASES 3/4.7.3-REACTOR CORE ISOLATION COOLING SYSTEM (Continued)
The surveillance requirements provide adequate assurance that RCIC will be j
OPERABLE when required.
Although all active components are testable and full I
flow can be demonstrated by recirculation during reactor operation a complete functional test requires reactor shutdown.
ThepumpdischargepipIngismain-tained full to prevent water hammer damage.
3/4.7.4 SNUBBERS l
All snubbers are required OPERABLE to ensure that the structural integrity of l
the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.
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Snubbers are classified and grouped by design and manufacturer but not-by size.
For example, mechanical snubbers utilizing the same design features of the 2-kip, 10-kip, and 100-kip capacity manufactured by Company "A" are of the same type.
The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.
A list of individual snubbers with detailed information of snubber location, size and system affected shall be available at the plant.
The accessibility of each snubber shall be determined and reviewed by the Facility Review Group and approved by the Manager, Clinton Power Station.
The-determination shall be based upon the accessibility of the snubber during plant operations (e.g.,
radiationlevel, temperature, atmosphere, location,etc.).
The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.
When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to deter-mine if any safety-related component or system has been adversely affected by.
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the inoperability of the snubber.
The engineering evaluation shall determine i
whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.
A representative semple of the installed snubbers will be functionally tested during plant shutdowns at 18 month intervals.
Observed failures of these sample snubbers will require functional testing of additional units. To provide further assurance of snubber reliability, snubbers are visually inspected at the frequencies recommended in NRC Generic Letter 90-09.
Hydraulic snubbers and mechanical snubbers may each be treated as a different.
entity for the above surveillance programs.
The service life of a snubber is evaluated via manufacturer input and informa-tion through consideration of the snubber service conditions and associated installation and maintenance records, i.e., newl placed, spring replaced, in high radiation area,y installed snubber, seal re-in high temperature area, CLINTON - UNIT 1 B 3/4 7-2 Amendment No.61 JAN 1519ga i
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3/4.7.4 SNUBBERS (Continued) etc.... The requirement to monitor the snubber service. life.is included.to j
ensure that the snubbers periodically undergo a performance evaluation in view i
of their age and operating conditions. These records will provide statistical-bases for future consideration of' snubber service life. The requirements for
the maintenance of records and_the snubber service life review are not-intended to affect plant operation.
l 3/4.7.5 SEALED SOURCE CONTAMINATION ~
j The limitations on removable contamination for-sources' requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c). limits for plutonium.
j This limitation will ensure that leakoge from byproduct, source, and special nuclear material sources will not exc'eed allowable intake. values.
Sealed sources are classified into three groups. according to their use, with surveil-j lance requirements commensurate with the probability of damage to a' source in i
that group. Those_ sources which are frequently handled are required to be tested more often than those'which are not.
Sealed sources which are continu-ously enclosed within a shielded mechanism, i.e., sealed sources within radia-tion monitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3/4.7.6 MAIN TURBINE BYPASS SYSTEM i
The main turbine bypass system is required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis in USAR Chapter 15.
3/4.7.7 L10VID STORAGE TANKS The tanks listed in this Specification include all those outdoor storage tanks j
that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
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Restricting the quantity of radioactive material contained in each of the specified tanks provides assurance that in the event'of an uncontrolled
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release of the contents from any of these tanks, the resulting. concentrations would be less than the limits of 10 CFR Part 20, Appendix B, _ Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
CLINTON - UNIT 1 B 3/4 7-3 revised by letter dated 5/21/93 a
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l BASES 3/4.7.8 MAIN CONDENSER OFFGAS HONITORING 3/4.7.8.1 0FFGAS - EXPLOSIVE GAS MIXTURE jet air ejectors to ensure adequate dilution of hydrogen offgas from attaining hydrogen levels in excess of the flammability limit),
explosive gas mixtures contained in the offgas holduthis specificat maintained below the flammability limit of hydrogen.p system is_ monitored and Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance w the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
3/4.7.8.2 0FFGAS - NOBLE GAS RADI0 ACTIVITY RATE Restricting the gross radioactivity rate of noble gases-from the main condenser provides reasonable assurance that the total body exposure to an i
i individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is. inadvertently cischarged directly to the environment without treatment.
This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.
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3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2. AND 3/4.8.3 AC SOURCES. DC SOURCES. AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the AC and DC power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant AC and DC power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.
The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation.
The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least Division I or II of the onsite AC and DC power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite AC source.
Division III supplies the high pressure core spray (HPCS) system only.
The AC and DC source allowable out-of-service times specified in the applicable ACTION statements are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974.
When diesel generator lA or IB is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator i
1A or 1B as a source of emergency power, are also OPERABLE.
This requirement is intended to provide assurance that a loss of offsite power event will not result in a complete loss of safety function of critical systems during the period diesel generator lA or IB is inoperable.
The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.
It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.
Analysis has shown that testing which includes a semi-annual f ast start of the diesel generators is sufficient to demonstrate the capability of the onsite A.C. power systems to mitigate the consequences of the design basis event for the plant (i.e., large LOCA coincident with a loss-of-offsite power) while minimizing the mechanical stress and wer on the diesel generator.
The OPERABILITY of the minimum specified AC and DC power sources and asso-l ciated distribution systems during shutdown and refueling ensures that (1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.
The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Re3ulatory Guide 1.9, CLINTON - UNIT 1 B 3/4 8-1 revised by letter dated 5/21/93
BASES 3/4.8.1. 3/4.8.2. and 3/4.8.3 AC SOURCES. DC SOURCES. AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
" Selection of Diesel Generator Set Capacity for Standby Power Supplies",
March 10, 1971, Regulatory Guide 1.108, " Periodic. Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants,"
Revision 1, August 1977 and Regulatory Guide 1.137 " Fuel-Oil Systems for Standby Diesel Generators," Revision 1, October.1979.
The surveillance requirements for demonstrating the 0PERABILITY of the unit batteries are in accordance with the recommendations of Regulatory Guide 1.129
" Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, Regulatory Guide. l.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants,"
February 1977, IEEE Std 450-1975, "IEEE Recommended Practice for Maintenance,.
Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations," and IEEE Std 308-1974, "IEEE. Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations" with-exceptions '
noted in the CPS-USAR.
i Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance Lf battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with-the rated capacity.
Table 4.8.2.1-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity (sg). The limits for the ' designated pilot cell's float voltage and specific gravity, greater than or equal to 2.13 volts and greater than or equal to 1.195 (sg), or a battery charger current that has stabilized at a low value are based upon the manufacturer's recommended values, and are-characteristic of.a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than or equal to 2.13 volts and greater than or equal to 1.190 (sg), and an average specific gravity of all the connected cells greater than or equal.to 1.200 (sg), are the values recommended by the manufacturer to ensure the OPERABILITY and capability of the battery.
Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8.2.i-1 is permitted for up to 7 days.
During this 7 day period:
(1) the allowable values for electrolyte level-ensures no physical damage to the plates with adequate electron transfer l
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capability; (2) the allowable value for the average specific gravity of all the cells is the manufacturer's recommended specific gravity, which ensures l
that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity is the value recommended by the manufacturer to ensure that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.10 volts, ensures the battery's capability to perform its design function.
l CLINTON - UNIT 1 B 3/4 8-2 revised by letter dated 5/21/93
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REFUELING OPERATIONS BASES 3/4,9.12 INCLINED FUEL TRANSFER SYSTEM f
The purpose of the inclined fuel transfer system specification is to control i
personnel access to those potentially high radiation areas immediately l
adjacent to the system and to assure safe operation of the system.
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