ML20058K659
| ML20058K659 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 12/10/1993 |
| From: | ILLINOIS POWER CO. |
| To: | |
| Shared Package | |
| ML20058K656 | List: |
| References | |
| NUDOCS 9312150350 | |
| Download: ML20058K659 (9) | |
Text
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Attachment 1
- to U-602192 INSTRUMENTATION Page 2 of 3 ,
, BASES ,
3/4.3.2 CONTAINMENT AND REAC10R 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 4
i 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 7% accuracy and calibration capability.
AMihm 3 /4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ,
D 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 system at the same time. l Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the d'ifference 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.
g Setpoint and Allowable Value also contain additional margin for instrument 8 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 ;
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instrumentation.
-l 3/4.3.4 RECIRCUtATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scr:m (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 i to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDO-10349, dated March 1971 and NED0-24222, 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-EOC-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 _;
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l CLINTON - UNIT 1 B 3/4 3-3 revised by letter dated 5/21/93 w
9312150350 931209 J r 1 )
DR ADOCK 0500
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. Attachment I to U-602192 Page 3 of 3
- Differential and ambient temperature instrumentation is provided for cenain equipment rooms or areas !
to effect automatic isolation of the affected systems in response to a 25-gpm equivalent steam leak. The l ACTIONS specified in Technical Specification 3.3.2 for inoperable differential teniperature :
instmmentation permit an extension ofthe allowed outage time for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when a sufficient ;
number of ambient temperature channels remain OPERABLE in the affected area to maintain the i capability for automatic isolation in response to a steam leak. ;
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2 Attachment 2 to U-602192 Page 1 of 4 Prooosed Revision to Bases Changes Submitted for Technical Specification 3.6.5 By letter U-601726 dated August 31,1990 lilinois Power (IP) submitted proposed changes to Clinton Power Station (CPS) Technical Specification 3.6.5, "Drywell Post-LOCA Vacuum Relief Valves.".
Subsequent to the submittal, and based on discussions with NRC staff personnel involved in the review -
and approval of the proposed Technical Specification changes, it was determined that IP should revise the Bases changes proposed in the original submittal. (Note: The Technical Specification changes requested per letter U-601726 were recently approved in Amendment No. 84 to the CPS Operating License. IP's originally proposed Bases changes were not included in the amendment. In the NRC's Safety Evaluation issued with the amendment, it is acknowledged that "the licensee has withdrawn [the Bases] portion of their submittal and has committed to submit a new Bases section to fully describe the post-LOCA function of[the drywell post-LOCA vacuum relief) valves.") The bases changes requested herein incorporate the needed revision to the originally proposed bases changes, and therefore, these
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proposed bases changes supersede the presiously proposed changes. The proposed bases changes involve no changes to the Technical Specifications themselves.
The changes proposed for CPS Technical Specification 3.6.5 primarily consisted of proposed revisions to each of the three Action Statements of this Technical Specification. [ Technical Specification 3.6.5 addresses the eight drywell post-loss of coolant accident (post-LOCA) vacuum relief valves employed in the CPS design in which two valves are utilized in series for each of four drywell penetrations.] In particular, as one of the requested changes, IP proposed to revise Action "a" to allow one or both normally closed drywell post-LOCA vacuum relief valves in a single penetration to be inoperable for openh. for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The LCO itself, it should be noted, would remain unchanged as it requires all eight drywell post-LOCA vacuum relief valves to be OPERABLE and closed.
In the technical discussion within IP's letter for the above proposed change, it was noted that no credit is taken in the safety analyses for operation of the drywell post-LOCA vacuum relief valves for limiting the negative pressure of the drywell following a LOCA. Analyses presented in CPS USAR Section
- 6.2.1.1.4.1 demonstrate that the drywell negative differential pressure design limit of 17 pounds per square inch differential (psid) is not exceeded during a design basis accident assuming that the drywell post-LOCA vacuum relief valves do not open. Notwithstanding, it was noted in IP's submittal that the valves are required to open to support operation of the drywell-containment atmosphere mixing system.
In light of the above, IP proposed corresponding changes to the Bases for Technical Specification 3/4.6.5, Specifically, IP proposed revising the Bases to read as follows:
Although no credit is taken for their operation following a DBA LOCA, drywell vacuum relief valves are provided on the drywell to pass suflicient quantities of gas from the containment to the drywell to prevent an excess negative pressure from developing in the drywell. Opening of drywell vacuum relief valves is also required to support operation of the drywell-containment.
atmosphere mixing system. Two vacuum relief valves in series are provided in each of four drywell penetrations. The drywell vacuum relief penetrations must be closed in order to prevent steam bypass of the suppression pool in the event of a LOCA.
After IP's submittal of proposed changes to Technical Specification 3.6.5 and associated Bases, and after discussions with NRC staff personnel, it was noted IP's submittal did not address the fact that
c l Attachment 2 :
to U-602192 Page 2 of 4 ,
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credit is also taken for opening of three of the four drywell vacuum relief penetrations to mitigate the efTects of suppression pool swell following a design basis LOCA (reference CPS USAR Section A3.8.3.1 of Attachment A3.8 which discusses the effects of dynamic phenomena on the containment' i
and drywell). That is, although credit is not taken for the drywell post-LOCA vacuum relief function relative to negative pressure effects on the drywell, some of the valves are expected to open for pressure >
equalization to mitigate pool swell and associated drag load effects on equipment, components or #
structures in the drywell. It was agreed that this has no impact on the proposed changes to the j Technical Specifications (as those changes remain technically justified), but IP's proposed changes to the {
associated Bases should be revised. Since IP had proposed adding the words "although no credit is ;
taken for their operation following a DBA LOCA" (referring to the drywell vacuum relief valves' post- l LOCA drywell vacuum relief function), this could be seen as conflicting with the credit taken for these i valves opening (for three of the four penetrations) in the suppression pool swell loading analysis. IP now proposes, therefore, to remove these words from the proposed changes to the Bases. The complete and corrected proposed changes to the Bases for Technical Specification 3/4.6.5 are reflected l on the following pages (pages 3 and 4 of this attachment). These changes supersede the Bases changes proposed in IP's original submittal. 1
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Attachment 2
-CONTAINMENT SYSTEMS to U-602192 :
, BASES Page ) of 4 i
3/4.6.5 DRYWEtt POST-LOCA VACUUM RELIEF VAtVES 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 drywelig following a lage break L.0tA. l
> INSERT ATTACHED '
}f4.6.6 SECONDARY CONTAINMENT F 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 l 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. :
Establishing and maintaining a vacuum in the secondary containment with the standby gas treatment system once per 18 months, along with the surveillance i 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 values are not verified in the surveillances since no credit for !
dilution 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 i 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. The !
specified heater dissipation is based on a bus voltage of 460 volts. Heater
.- test results shall be adjusted to account for actual bus voltage. !
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-
- centration within the containment below its flamable limit during post-LOCA CLINTON - UNIT 1 B 3/4 6-8 letter dated 6/29/93
Attachment 2 !
to U-602192 l Page 4 of 4 In addition, this function controls rapid weir wall overflow (following a large-break LOCA) to minimize drag and impact loadings on essential equipment and systems in the drywell above the weir wall. .
OPERABILITY for opening of the drywell vacuum relief valves is also required to support operation i of the drywell-containment atmosphere mixing system.
The drywell post-LOCA vacuum relief valve penetrations are required to be closed in order to prevent steam bypass of the suppression pool in the event of a LOCA.
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Attachment 3 .
to U-602192 Page 1 of 3 Proposed Change to Basis for Technical 2 S ecifications 3/4.5.1 and 3/4.5.2 .
The third proposed changed to the Te;hnical Specification Bases is for Technical ;pecification 3/4.5.1 and 3/4.5.2, " Emergency Core Cooling Systems-Operating and Shutdown." This change is requested to correct the value specified for a particular differential pressure (AP) at which the high pressure core spray system (HPCS) will deliver a particular flow. This change is the result of recently completed ;
analyses to suppon relaxation of the tolerance required for the safety-mode lift setpoint for the main steam safety / relief valves (SRVS).
Illinois Power (IP) recently completed an efTort to relax the required tolerance for the SRV safety-mode l lifl setpoint from *1% to *3% in support of the current refueling outage (and future outages) at (CPS).
This effort was supported in part by the NRC staffs review and approval of a generic licensing topical ;
repon prepared and submitted by the Boiling Water Reactor Owners' Group (BWROG) (reference ;
NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing i Topical Report," submitted to the NRC by BWROG letter dated July 9,1990). IP's approach to the SRV setpoint relaxation was first discussed with the NRC staffin a meeting conducted on May 5,1993. .
The key point of the meeting was IP's interest in relaxing the tolerance relative to the ASME XI Insenice Testing acceptance criterion for determining whether additional valves must be tested (based :
on the test results obtained for an initial sample of removed valves). IP also noted that it would ,
complete a plant-specific analyses and safety evaluation to support the tolerance relaxation and that the !
associated Technical Specification changes would be incorporated through implementation of the Improved BWR-6 Standard Technical Specifications. It was agreed that a Relief Request for the CPS Inservice Testing program would be submitted based on the completed analyses and safety evaluation.
The relief request, No. 2036, was submitted by IP in July 1993 and was recently approved by the NRC by letter dated September 13,1993. ?
As noted previously, a plant-specific analysis for increasing the SRV setpoint tolerance was recently completed for CPS. This analysis concluded that relaxation of the SRV safety-mode setpoint tolerance l from *1% to
- 3% is acceptable. The original
- 1% SRV tolerance did not represent a limiting setpoint l
condition required to ensure plant safety. Evaluations contained in the overall analysis include vessel . l overpressure protection, abnormal operational occurrences, Emergency Core Cooling System / Loss-of- l Coolant Accident performance, high pressure makeup system performance, containment response and j integrity, and Anticipated Transient Without Scram (ATWS) performance. !
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-With respect to high pressure makeup system performance, the plant-specific analysis she ed that l increasing the SRV safety-mode setpoint tolerance to 3% increases the High-end pressure at which the !
high-pressure makeup systems are required to operate by approximately 23 psi. The analysis l nevertheless demonstrated the current HPCS system and the Reactor Core Isolation Cooling (RCIC) system each have sufficient margin to deliver the required system flow rate at the new maximum i operating pressure. However, this 23 psi increase in the maximum discharge head for the HPCS system j necessitates a change to the Bases for Technical Specifications 3/4.5.1 and 3/4.5.2 due to the following sentence:
"The capacity - -,: [HPCS] system is selected to provide the required core cooling. The HPCS pump is 6esigned to deliver greater than or equal to 467/1400/5010 gpm at differential pressure of 1177/1147/200 psid."
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- Attachment 3 i to U-602192 Page 2 of 3 Within this sentence "1177" should be changed to "1200" to reflect the 23 psi increase in the maximum !
vessel operating pressure due to the SRV setpoint tolerance increase. No comparable sentence appears in the Technical Specification Bases for the RCIC system; therefore, no such change is required there, (Also, the Technical Specifications themselves are unaffected since the SRV analysis affects only the ,
maximum operating pressure, not the pressures and flows identified in the pump-test surveillance requirements.) The proposed change to the Bases for Technical Specifications 3/4 5.1 and 3/4.5.2 is ,
reflected on the associated marked-up page from the CPS Technical Specification Bases, provided as. -
page 3 of 3 of this attachment..
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?' Attachment 3 of U-602192 4 Page 3 of 3 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.
The HPCS pump is designed to deli ' reater than or equal to 467/1400/5010 gpm at differential pressures of 117 /1147/200 psid. Initially, water from the reactor core isolation cooling CIC) 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 ater.
- 1200 With the HPCS system inoperable, adequate core cooling is assured by the i 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 sa will automatically provide makeup at reactor operating pressurefety s on aanalysis, reactor low water level condition. The HPCS out-of-service period of 14 days, as i specified in the corresponding ACTION statement, is based on the demonstrated e OPERABILITY of redundant and diversified low pressure core cooling systems. ,
The surveillance requirements provide adequate assurance that the HPCS 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.
Upon failure of the HPCS system to function properly after a small break loss- .
of-coolant accident, the automatic depressurization system (ADS) automatically l 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 l time to limit fuel cladding temperature to less than 2200*F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure '
i exceeds 100 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.
ADS automatically controls seven selected safety-relief 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.
The surveillance requirements for the ADS include a requirement to manually open each ADS valve. This requirement includes an exception to the provisions of Specification 4.0.4. This exception allows reactor steam conditions to be established which are adequate to open the ADS valves without resulting in unnecessary wear on the valves and to ensure that proper reactor pressure control can be maintained while opening and reclosing the valves. Reactor steam conditions which are considered adequate to perform the test thus include the establishment of sufficient reactor pressure as well as sufficient CLINTON - UNIT 1. B 3/4 5-2 Amendment No. 81 I 1
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