ML20236G980
| ML20236G980 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/27/1987 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20236G984 | List: |
| References | |
| 1433, TAC-66415, NUDOCS 8711030373 | |
| Download: ML20236G980 (27) | |
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\ TOLEDO EDISON DONALO C. SHELTON ve.non-ruw -
1)ocket No. 50-346 ~
License No 'NPF-3 Serial No. 1433 October 27, 1987 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Subject:
License Amendment Application'to Support Enhanced Feed and' Bleed Modifications (TAC No. 66415)
Gentlemen:
The enclosed application for license amendment requests that the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1 Operating License, Appendix A,.
Technical Specifications be revised to. reflect the changes attached.. The .l proposed changes involve Section 3/4.3.2, Safety System Instrumentation, 1 Table 3.3-5, and Section 3/4.6.3, Containment Isolation Valves, Table 3.6-2. l Toledo Edison's letter of December 4, .1985 (Serial No. '1207): committed- to enhance feed and bleed capability for reactor decay heat removal under the~ l situation where all feedwater to the steam generators is lost. . The-initial concept of primary system depressurization'to. enable' initiation of High Pressure Injection (HPI), as was proposed in this letter, was: subsequently. 4 determined to be undesirable from a potential voiding standpoint. Toledo. "!
Edison's letter of June 25, 1987.(Serial No. 1382) provided a discussion.of the undesirability of the initial concept and proposed an-alternate-conceptual method for-reactor core cooling in the event of-the loss'of all feedwater. This proposal involves utilizing the high discharge headt ,
makeup pumps to provide redundant trains of feed to the reactor core'from !
the Borated Water Storage Tank (BWST).. A modification implementing this ,
concept is being pursued for the fifth refueling outage.-(presently scheduled '
to commence in February 1988) . -
I pt f 8711030373 871027 ( i l- PDR ADOCM 05000346 P PDR -
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THE TOLEDO EDISON COMPANY EDISON PLAZA ' 300 MADISON AVENUE TOLEDO. OHIO 43652 ? ,j t ,
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d Docket No. 50-346 License No. NPF-3
. Serial No. 1433 Page 2 To prevent interruption of the feed and bleed process and potential damage to the makeup pumps due to deadheading the pumps, both of which may' occur on actual or spurious Safety Features Actuation System (SFAS) actuation, it is proposed that the SFAS signal to valve MU-33.(the makeup line containment isolation valve) be deleted. Furthermore, it is proposed that no SFAS signal be provided to the proposed redundant train's corresponding valve, MU-6421, and that these valves provide containment isolation via remote manual means. The lack of an SFAS signal to automatically close these valves is consistent with General Design Criterion (GDC) 55, which !
permits alternative containment isolation provisions for specific classes of lines if they can be demonstrated acceptable on'some other defined basis. NUREG-0800, Standard Review Plan (SRP), defines the acceptable alternatives; specifically,Section II of SRP 6.2.4 details system acceptance !
criteria with' allowable exceptions for' engineered safety feature or engineered safety festure-related systems and systems needed for safe ,
shutdown of the plant. In addition, GDC 55 states _that automatic isolation valves shall "take the position that provides greater safety" upon loss of l actuating power. The makeup valves MU-33 and MU-6421 will be routinely (
open during normal operation in order to provide a makeup flowpath to the i Reactor Coolant System. If power is lost to the valves, it is proposed they remain open to afford greater safety in order to permit the passage {
of water for core cooling upon a loss of all feedwater and, thereby, J prevent possible core damage. It should be noted that if these valves i were to close on a SFAS signal, an interruption in core cooling would l occur and the possibility of their failure to reopen emerges. Therefore, l to preclude these adverse possibilities, the SFAS signal.is proposed to be l deleted. i The design of the two affected penetrations has incorporated appropriate features to minimize the probability and consequences of a postulated j rupture, as required by GDC 55. The makeup lines components, from MU-33 ;
and MU-6421 downstream to the HPI lines, are constructed with "Q" compo-l nents, are seismic Category I, and are protected against hazards such as pipe whip and jet impingement. In addition, the makeup system upstream of hU-33 and MU-6421 is seismic Category I. Isolation of the makeup line to the HPI line can be accomplished by a check valve (MU-169) down stream of MU-33 and is further protected by the makeup pump discharge check valve upstream of MU-33. Therefore, no adverse consequences related to containment integrity or HPI operability are created. (It should be noted that this arrangement of check valves will also exist for the new redundant line j
containing MU-6421).
l Based on the above, changes are proposed to the technical specifications as follows:
- 1. Technical Specification 3/4.3.2, Table 3.3-5: Deletion of references to the Makeup System as containing valves which receive an SFAS signal. (Affected Sections of the table are 1.1, 2.1 and 5.c)
l Dockat No. 50-346 License No. NPF-3 . ,
- Serial No. 1433 Page 3 l
- 2. Technical Specification 3/4.6.3, Table 3.6-2: Deletion of Penetration 19 ;
containment isolation valve MU-33 from Section A of the. table; i include MU-6422 (the redesignated MU-33 tag number) and MU-6421 in Section C of the table for Penetrations'19 and 50, respectively.
These changes will provide for a reliable enhancement of feed and bleed capability at the DBNPS. These changes must be issued prior to startup- j from the fifth refueling outage to support the commitment to have- ]
enhanced functional feed and bleed capability at that time. l l
Very trul yours, 1
RMC:dem
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cc: DB-1 NRC Resident Inspector ,
A. B. Davis, Regional Administrator (2 copies) !
A. W. DeAgazio, NRC/NRR Davis-Besse Project Manager State of Ohio I
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?Dockst NS.J'0-346: 5 '
, lLicense No.LNPF-3' w Serial '.No . . ,1433 ' ' '
- Enclosure Page 1 o- ,, ,
' APPLICATION FOR-AMENDMENT 'i I
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(. . FACILITY;0PERATING> LICENSE NO.:NPF-3
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. DAVIS-BESSE NUCLEAR POWER$TATION.
UNIT NO.-1: 4 j .i Attached aret requested changes to the'. Davis -Be'sse Nuclear Power Station,;
Unit No.'l Facility Operating' License;No."NPF-3. .Also included are the.
Safety Evaluation and'Significant: Hazards: Consideration.' '
. The proposed change-(submitted ~under cover.:1'etter Serial;No.:1433)" concern:.
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Section 3/4.3.2, Safety System Instrumentation,; Table'3.3-5 P. ,,
Section 3/4.6.3, Containment Isolation Valves,= Table 3.6-2 c-
. 3 j By S D. C. Shelton, Vice PresideistThuclear
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Sworn and subscribed before me this 27th day:of October,.1987. +
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Notarf Public, State of. Ohio i My commission expires.- I u .
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Dockat.No.iS0-346i '
' License No. NPF-3~ 'T
- ' Serial No'. 1433'
' Enclosure Page-2 The following information is provided to.' support issuance of the' requested changes to the Davis-Besse Nuclear Power' Station,, Unit.'No. 1~.
- Operating License No. NPF-3, Appendix;A, Technica1' Specifications Tables 3.3-5 and 3.6-2.
, A. Time require'd'to Implement: -This change will_be implemente'd by' the licensee-by the end of the.fifth. refueling outage. ~However, . .
expeditious issuance is required to support the. modification 1for:the~-
fifth refueling-outage.
B. ' Reason for Change (FCR No.- 87-0131):' ~ Revise the~ Technic'al Specifications to delete the Safety Features Actuation System signal _
to makeup valve MU-33, and add'a sister valve to a redundant train to support the enhanced fced and bleed, modifications. -
i C. Safety Evaluation: See attached' Safety l Evaluation (httachment No. 1).
D. Summary Significant Hazards Consideration: See. attached Summary Significant Hazards Consideration (Attachment No t 2) ...
E. Significant Hazards Consideration: See attached Significant Hazards Consideration (Attachment No. 3).
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Docket No. 50-346 License No. NPF-3
- Serial No. 1433 Attachment 1 Page 1 '
SAFETY EVALUATION Description This Safety Evaluation supports a change to the Davis-Besse Technical Specifications, Appendix A of the Operating License, to delete the License -
Safety Features Actuation System (SFAS) closure signal from MU 33, the makeup (MU) line containment isolation valve. The change is necessary because the makeup system serves a function similar to that of engineered safeguards systems for a specific off-design basis upset, i.e., a complete loss of feedwater. Deletion of the SFAS closure signal remains consistent with General Design Criterion (GDC) 55, which identifies isolation requirements on lines that are part of the reactor coolant pressure i boundary and also penetrate the containment building. GDC 55 permits alternative containment isolation provisions for specific classes of lines ,
if they can be demonstrated acceptable on some other defined basis. The I Standard Review Plan (NUREG-0800) defines acceptable alternatives.
Makeup System containment isolation valve MU 33 presently receives an automatic closure signal on SFAS. incident level 3, which is actuated on l either high containment pressure or low-low Reactor Coolant System (RCS) pressure. However, to compensate for an off-design basis event involving l a total loss of normal and emergency feedwater systems, Davis-Besse i relies upon the feed and bleed cooling method, in which the MU system provides the only source of post-trip decay heat removal. Flow will exit the Reactor Coolant System (RCS) via the PORV and/or the pressurizer code safety valves which results in subsequent containment pressurization.
The Davis-Besse feed and bleed modification will implement numerous modifications to the MU system and PORV in order to imp.ove this cooling capability. (See the attached figure for the proposed modification diagram).
Calculations demonstrate that the 18.4 psia high containment pressure setpoint which initiates SFAS incident levels 2 and 3 will be exceeded l early in a feed and bleed scenario and containment p; essure may also approach the SFAS level 4 setpoint if only one containment air cooler is operating (under single failure considerations). Peak containment pressure has been calculated at approximately 35 psia, compared to a minimum allowable SFAS bistable field setpoint of 36.825 psia. In order to direct all available makeup flow to the RCS, the feed and bleed modification includes provisions to isolate the makeup pump minimum recirculation line. To prevent interrr.ption of the feed and bleed process and potential damage to the mr.keup pumps by deadheading them, it is necessary for MU 33, and its elater valve in a parallel MU injection line being added under the feed and bleed modification, to remain open throughout the transient.
Docket No. 50-346 License No. NPF-3
- Serial No. 1433 Attachment 1 Page 2 This can be best accomplished by removing the SFAS automatic closure feature from MU 33 and by not providing such a closure signal on the new containment isolatien valve in the parallel train (MU 6421). This will ensure the greatest reliability possible for the enhanced feed and bleed system while still maintaining the required containment isolation features.
The specific changes involve deletion of MU 33 from Technical Specification Table 3.3-5 to reflect elimination of its SFAS actuation capability.
Table 3.6-2 must be modified to include the new remote manual operated containment isolation valve, MU 6421. An additional Technical Specification change addressed is to redesignated MU 33 as MU 6422 in Table 3.6-2. This is incorporated to improve correspondence between the two MU trains from a '
human factors standpoint.
Systems Affected Containment Isolation Makeup System Safety Features Actuation System References j i
- 1. Technical Specifications 3/4.3.2 and 3/4.6.3
- 2. USAR Sections 3.6.2.7, 6.2.4, 7.3 and 9.3.4
- 4. NUREG 0800, Standard Review Plan (Section 6.2.4)
- 5. NUREG 0737, Clarification of TMI Action Plan Requirements (Set;11on II.E.4.2)
- 6. Drr ? egula.-r y Guide 1.141, Containment Isolation Provisions for Fluid Systems ,
- 7. ANS/ ANSI 56.2-1984, Containment Isolation Provision for Fluid Systems !
After a LOCA
- 8. ANS/ ANSI 51.1-1983, Nuclear Safety Criteria for the Design of Nuclear Power Plants j
- 9. FCR 86-0432, Enhanced Feed and Bleed Capability j
- 10. P& ids M-031 and M-033 l
- 11. Toledo Edison Check Valve Reliability Program Report.
Safety Functions and Functions of Systems The containment vessel isolation system is designed according to the requirements of General Design Criteria (GDC) 54, 55, 56 and 57 of 10CFR50, Appendix A. Isolation valves are provided in lines penetrating )
the containment vessel to ensure no uncontrolled release of radioactive material can occur during a design basis accident. Penetrations not serving accident-consequences-limiting systems will either be closed by automatic isolation signals (SFAS) during such an event or be maintained l
Docket No. 50-346 License.No. NPF-3 Serial No. 1433 Attachment 1 Page 3 locked closed. The particular penetrations involved (No. 19 for the existing line and No. 50 for the new one) are classified as Type IV penetrations according to Section 6.2.4 of the. Davis-Besse USAR since they' provide a.flowpath for an engineered safety features (ESF) system, namely High Pressure Injection (HPI). Type IV penetrations do not have closure requirements since their design function is to open during a LOCA to permit passage of fluid used for reactor core cooling or' containment depressurization.
i The normal functions of the Makeup System are to maintain-Reactor Coolant System (RCS) inventory during plant. operation, to provide Reactor Coolant Pump (RCP) seal injection and to continuously purify the reactor coolant. Breaches in the RCS pressure boundary which~are within the j capacity of the MU System will be accommodated by that system and do not l result in actuation of engineered safety features. In the event that all j feedvater is unavailable, core cooling is accomplished by feed and biced. l This entails injection of water from the Borated Water Storage Tank (BWST) j into the.RCS using the MU System and removing steam / water via the PORV l and/or pressurizer code safety valves after passing through the reactor-Core.
I The design purpose of the Safety Features Actuation System is to protect '
the reactor core during a loss of coolant accident (LOCA) and to mitigate the consequences of a LOCA. This is accomplished by automatically j isolating the containment vessel to prevent release of radioactive fission products and by initiating operation of emergency core cooling and containment depressurization equipment. The system consists of four identical and redundant sensing and logic channels and two redundant ,
actuation channels. Parameters monitored are RCS pressure, containment '
pressure, containment radiation and BWST level. SFAS output logic consists of five separate levels, each actuating or providing permissive signals for specific systems and also establishing varying degrees of-containment isolation.
Effects on Safety General Design Criterion 54 requires that piping systems penetrating reactor containment "be provided with leak detection,' isolation, and.
containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems." General' Design Criterion 55 specifies containment-isolation requirements for lines which are also part of the RCS pressure boundary. It states:
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-Docket No.E50-346
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-License No. NPF-3 , ,
Serial No. 1433. ,1 Attachment 1 Y H
Page 4 Each line that is part.of the reactor coolant pressure boundary and' ;j
.that penetrates primary reactor containment shall be provided with ,
containment isolation valves as follows, unless.it can be
' demonstrated that the' containment isolation provisions for a i specific class of lines, such as instrument lines, are. acceptable cnt ,
some other defined. basis: <
(1)- One locked closed: isolation valve inside and one locked closed, j isolation valve outside' containment; or j l
(2) One automatic isolation valve'inside and one locked' closed:
isolation valve outside containment; or ~!
(3) One locked closed isolation valve inside and'one automatic isolation. valve outside containment. A simple _ check valve may. ,
not be used as~the. automatic isolation valve outside' n. l containment; lor' (4) One automatic isolation valve inside'and one automatics isolation valve outside containment. A simple check valve.may.
not be used as the automatic isolation. valve outside. _!
containment.
Isolation valves outside containment shall be-located asiclose to containment as practical and'upon loss of_ actuating power, automatic' isolation valves shall be designed to take the position that pro" ides greater safety. ~l Other appropriate requirements to minimize ~the probability?or consequences of an accidental rupture of these lines or of u lines connected to them shall be provided as'necessary to assure adequate safety. Determination of the appropriateness j of.these requirements,_such as higher quality in design, ~
fabrication, and testing, additional provisions-for inservice inepection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include-consideration of the population density,'use_ characteristics, and physical characteristics of the' site environs.
The requirements for containment isolation systems presented in the Standard Review Plan (SRP) elaborate on'the GDCs._ Specific criteria necessary-to meet relevant GDC requirements and guidelines for acceptable alternate containment' isolation-provisions forfcertain' classes?of lines L
are presented in the SRP.Section II of SRP 6.2.4 enumerates system acceptance criteria along with allowable exceptions for engineered safety feature or engineered safety features-related systems and systems needed: 1 for safe shutdown of the' plant. Remote manual isolation valves are-i
Dockat No. 50-346 License No. NPF-3
. Serial No. 1433 Attachment 1 Page 5 permitted in these situations provided possible leakage outside containment can be detected. This position is repeated in NUREG 0737 (Ref._5). The nature of the makeup system during feed and bleed cooling is similar to that of an ESF system since it provides the only means of maintaining adequate core water inventory to preclude offsite-radiation releases in excess of 10CFR100 limits.
The General Design Criterion states that automatic isolation valves shall take the position providing greater safety upon loss of actuating power. Both MU 33 and the new valve, MU 6421, will be maintained normally open to provide a makeup flowpath during power operation. Since these j valves will be motor operated, they will fail "as-is" upon loss of power. '
This is the position affording greater safety as it will permit the passage of water for core cooling upon a loss of all feedwater, thereby preventing possible core damage.
In the design of the two affected penetrations, along with the inside and outside isolation valves, appropriate features to minimize the probability. q and consequences of accidental rupture have been considered as required 1 by GDC 55. These components are "Q," seismic Category I, and protected 1 against hazards such as pipe whip and jet impingement. Further, the interfacing MU System piping upstream of MU 33 is inherently precluded' from creating any adverse consequences with regard to containment integrity and HPI operability, as explained in detail in a subsequent paragraph.
According to ANS/ ANSI 51.1 (Ref. 8) the performance criterio . for' primary containment is to confine radioactive material leakage during any operational activity or event. These limits are prescribed by 10CFR50, Appendix I for normal operation and by 10CFR100 for transients or accidents.
The system design should satisfy the requirements of ANS/ ANSI 56.2 (Ref. 7). The latter standard delineates specific isolation system design requirements for satisfying these criteria.
ANS/ ANSI 56.2 expands upon the GDC 55 requirements regarding containment isolation and defines a class of lines forming an exception to the provisions identified. It states that "the objective of containment isolation shall be to allow the normal or emergency passage of the following through the containment boundary while preserving the integrity of the containment boundary: (1) Engineered safety feature system fluids, or (2) Fluids of systems which are not required to function following a loss-of-coolant accident but, if available, can be used to accompJish a function similar to an engineered safety feature system." Examples given of such systems include PWR fluid systems required for RCP operation. It is noted that the concept of " functions similar to engineered safety features" is applicable to MU 33. This standard further defines '
" safety-related function" as those plant features necessary to acsure the capability to prevent or mitigate the consequences of accidents that could result in offsite exposures comparable to the guidelines of e
Dockot No. 50-346 License No. NPF-3 Serial No. 1433 Attachment 1 Page 6 10CFR100. Although the feed and bleed cooling system does not serve in a l safety related capacity according to traditional licensing criteria, .
i.e., direct mitigation of a design basis accident or transient, it I provides the only means of core heat removal during a complete loss of feedwater event. Therefore, the system serves a " safety-related" function according to the above criteria. Isolation of the system by closure of MU 33 results in operation outside the analyzed conditions for feed and bleed and could result in core uncovery leading to potential radiological release for this off-design basis event. As such, removal of the SFAS automatic isolation feature from MU 33 is appropriate.
Section 3.8 of ANS/ ANSI 56.2 discusses the requirements for the system providing the signal which initiates containment isolation (i.e.,'SFAS).
For those systems which employ phased isolation, as is the case at Davis-Besse, the standard requires automatic isolation of all systems ,
except those serving ESF or ESF-related functions. It further states J that only those ESF-related systems which can be justified to remain operational can be excluded from automatic isolation during the initial phase. This standard permits the isolation function associated with the preceding two categories of valves to be accomplished remote manually instead of automatically. Remote manual isolation is acceptable for ESF or ESF-related systems to permit them to perform their accident prevention or mitigation function provided certain conditions are satisfied. These conditions include provision of features allowinE the control room operator to detect possible failure of the fluid lines inside and/or outside containment to preclude degradation of containment isolation or the operation of ESF, and maintaining the capability to isolate the affected line remotely. An analysis of the consequences of a leak or line break, as required by Section 3.8.7 of ANS/ ANSI 56.2, will determine how quickly a non-isolated system must be isolated in the event of a leak and whether automatic or manual isolation is appropriate.
The configuration of the existing and proposed interfaces between the MU and HPI Systems at Davis-Besse inherently precludes the possibility of a MU line leak or rupture affecting either containment integrity or HPI.
Check Valve MU 169, installed downstream of MU 33 and prior to the point of connection to HPI, is within the "Q" boundary of the HPI system and is intended to prevent diversion of emergency core cooling fluid into the MU System when the MU pumps are not operating. Any MU system leakage or rupture occurring upstream of MU 169 would not, therefore, impair the performance of the HPI system. Further, the makeup system pipiug'is installed Seismic Category I, so line breaks need not be considered-concurrent with an accident situation, and tre pump discharge piping pressure rating is equivalent to that of HPI. Additional protection of the low pressure MU suction piping is afforded by a relief valve with an 88 psig setpoint and 450 gpm flow capacity and by the pumps' discharge check valves MU 196 and MU 197, A special test conducted on these valves in 1984 confirmed their ability to prevent reverse flow. A check valve reliability program established in 1986 recommended performing a detailed
a r A Docket No. 50-346 License.Ho. NPF-3
. Serial No. 1433 Attachment 1 Page 7 internal inspection of either HP 57, MU 196,,or MU 197 (these are similar-valves and see similar service) during the next refueling in order to- '
ascertain possible wear. . Such inspection will.be conducted in the upcoming-refueling outage'and is part of the Davis-Besse check valve' reliability and inspection program.
Although credit cannot be taken-for MU 169 serving a containment .
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isolation' function, since GDC 55 prohibits consideration of check' valves-as automatic isolation valves outside containment, containment integrity will still be assured. For the!8Atuation of HPI train.2 operating during a LOCA, Penetration 19 will remain. pressurized substantially above.any pressures possible in the containment atmosphere, effectively eliminating 1 the possibility of fission product leakage via this flowpath. . Should HPI' pump 2 fail to operate during a design basis event,.resulting in the penetration. remaining depressurized, containment integrityLis assured by-the redundant containment isolation valve inside, HP 57. An additional failure of this check valve is beyond the single failure criteria-and need.
not be postulated. -Check Valve HP 59, in series with HP 57, underwent-internal inspection and modification to its disc anti-rotation stops in 1984 to reduce its chances of failure, thereby providing increased confidence of proper seating. Additional recommended evaluations based on the check valve reliability program are described above. The HPI function for the condition of one HPI pump inoperable will be provided by the 3 redundant train. The makeup system is in continuous operation during normal plant operation. As such, any leakage would be observed and ]s appropriate corrective measures implemented to ensure a leak-tight system is maintained. Technical Specification 6.8.4(a) requires establishment of a program to minimize ~1eakage outside containment'from several systems, including makeup. The program consists of preventive maintenance and/or visual inspection requirements along with periodic integrated system leakage testing. .These existing requirements and operational monitoring are considered sufficient to preclude system leakages during an event requiring containment isolation. Further, the MU system is not intended to circulate fluid from the containment emergency. sump; pump suction is restricted to either the MU tank or'BWST. This will be. insured by appropriate procedure modifications.
Since the low-low RCS pressure signal is combined with the containment high pressure signal in the SFAS logic, this change also removes the-closure signal for these valves on low-low RCS pressure. Closure of the MU system containment isolation valves on low-low RCS pressure during'a design basis LOCA is not a necessary function. The MU pumps will.
normally continue to operate, thereby increasing' the flowrate of Larated
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water into.the. reactor vessel and enhancing core cooling. Further, s discussed above, the containment isolation functions remain satisfiei even if the penetration is depressurized due to the MU and HPI pumps not operating. !
s:
e Docket No. 50-346' License No. NPF-3 Serial'No. 1433 Attachment 1 Page 8 The failure evaluation presented above satisfies the GDC 55 requirement to minimize the probability and consequences of an accidental rupture of ,
such lines to ensure adequate safety and meets the intent of the 1 requirement in ANS/ ANSI 56.2 to analyze the consequences of a leak or break in an unisolated line. Further, since all: protection features j which assure containment integrity and preclude degradation of HPI are inherently provided, no time requirement for isolation of MU 33 is- )
warranted. Therefore, this isolation capability can be provided as -q remote manual from the control room rather than-automatic. Reverse flow i testing of MU 169 is presently part of the periodic valve test program.to {
ensure proper functioning of the check valve. This valve will continue ]
to be tested, along with the corresponding.one-in the new line being installed. Similarly, the Type C leakage test requirement will be maintained j for MU 33 and established for MU 6421 to ensure the valves'will remain j capable of isolation if closed remote manually. 1 1
The preceding discussion applies equally to the new makeup line being j installed. This line will utilize Penetration 50 and will have a "Q" check '!
valve comparable to MU 169 between MU and HPI. I Unreviewed Safety Question Evaluation The proposed deletion of the automatic SFAS closure feature from'MU 33 -l and provision of only remote manual closure capability on MU 33 and on the ;
new line (MU 6421) will not increase the probability of an accident previously i evaluated in the USAR. The containment penetrations involved will' )
normally remain pressurized by HPI during a design basis event thereby j precluding leakage of fission products from containment.. If a single l failure affecting operability of one HPI train occurred, the containment isolation valve inside containment would ensure proper isolation.
Implementation of this modification will actually reduce the probability of core uncovery occurring as a result of an off-design basis complete loss of feedwater event due to improved availability of the makeup ;
flowpath. (10CFR50.59 (a)(2)(1)) . l I
The proposed action would not increase the consequences of an accident previously evaluated in the USAR because the containment is.olation -l function of the penetrations involved will still be achieved. Likewise, the flow of HPI into the reactor coolant system will be unaffected, thereby remaining within the bounds of the existing LOCA analysis.. Any event initiating SFAS isolation of the makeup line will also start HPI, ensuring the affected penetrations will be pressurized well above containment pressure to preclude leakage of fission products from ,
containment. The consequences of a complete loss of feedwater event will be enhanced by this change. (10CFR50.59 (a) (2) (1)) .
Implementation of the proposed change will not increase the probability of a malfunction of equipment important to safety since only the automatic closure feature of the two valves is affected. Changing the j
Dockst No. 50-346 License No. NPF-3 Serial No. 1433 Attachment 1 Page 9 mode of operation to remote manual provides adequate valve closure capability should isolation be required. This modification actually decreases the probability of makeup pump failure during feed and bleed.
Since the MU pump minimum recirculation line will be isolated during this mode of operation to prevent diversion of water away from the reactor, closure of the containment isolation valves would then result in ;
deadheading the MU pumps, with subsequent pump damage possible.
(10CFR50.59 (a) (2) (1)) .
l The proposed change will not increase the consequences of a malfunction )
of equipment important to safety since no.new failure mode is introduced.
Failure of a containment isolation valve to operate on demand is equivalent to the single active failure presently assumed in the safety analysis.
This change will significantly decrease the adverse consequences of valve l malfunction during feed and bleed operation. Closure of these valves upon I reaching a pressure of 18.4 psia in containment, as is the present situation, f creates a requirement for the valves to be reopened in order to re-establish l flow and prevent core uncovery. This provides additional opportunity for active component failure, which is undesirable. (10CFR50.59 (a) (2) (1)) . ,
Implementation of this change will not create the possibility for an accident of a different type than any previously evaluated in the USAR, Containment integrity will be assured as explained previously and leakage I or line breaks in the MU system will not adversely affect the HPI system l due to the presence of the check valve at the interfaces. Check valves are considered passive components so their failure need not be postulated. However, makeup line check valve MU 169 is presently included in the station in-service test program, and the corresponding valve in the new line will be added, to ensure their capability to prevent reverse flow is maintained at an acceptable level. (10CFR50.59 (a) (2) (11)) .
The proposed change will not create the possibility for a malfunction of equipment of a different type than any previously evaluated in the USAR.
The automatic closure feature of the valves is being deleted as this is no longer a necessary function. However, remote manual isolation capability will be retained to ensure valve closure should it be required. Failure to actuate on demand will continue to be the only failure mode as is the present situation. (10CFR50. 59 (a) (2) (ii)) .
The proposed modification will not reduce the margin of safety as defined in the basis for any Technical Specification. Implementation of this .
change will limit radioactive material leakage as assumed in the existing ?
safety analysis to main?ain site boundary radiation doses within the limits of 10CFR100 for any design basis event. Since the penetrations )'
involved serve accident consequences mitigation functions, they can be justified to remain open. The margin of safety for the off-design basis total loss of feedwater event will be increased due to improved reliability of the flowpath. (10CFR50.59 (a) (2) (iii)) .
l
Docket.No. 50-346
' License No. NPF-3
, Serial No. 1433 Attachment 1 Page 10 Conclusion i
As demonstrated by the preceding discussions, General Design Criterion 55 i is satisfied with the proposed modification to the MU system containment '
j isolation valves. Elimination of the auto-closure feature is' consistent !
with the Standard Review Plan and industry standards due to the. ' j engineered safety features-related function which the Makeup System- ;
serves. l 1
Pursuant to the above evaluation, removal of the SFAS closure from MU 33 j and provision of only remote manual closure on its sister. valve in'the )
new train will not result in an Unreviewed Safety Question. ]
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I Dockat No. 50-346 i License No. NPF-3 i Serial'No. 1433 l Attachment 2
SUMMARY
SIGNIFICANT HAZARDS CONSIDERATION 1
l Description of Amendment Request: This amendment request proposes to delete the Safety Features Actuation System (SFAS). signal from makeup system valve MU-33 to support enhanced feed and bleed capability during a 1 postulated loss of all feedwater event. The valve will' retain remote l manual isolation capability. This change would be reflected in a i revision to TS 3/4.3.2, Safety System Instrumentation, Table 3.3-5 and TS 3/4.6.3, Containment Isolation Valvec, Table 3.6-2. i i
The change would also add a new valve, MU-6421, to Table 3.6-2. This valve performs the same function as MU-33 on a redundant train to be -)
installed as part of the feed and bleed modification, and would change the !
designation of MU-33 to MU-6422.
Basis for the Proposed No Significant Hazards Consideration: Toledo Edison committed to provide enhanced feed and bleed capab'lity at the l Davis-Besse Nuclear Power Station Unit 1 as part of its response to the
- June 9, 1985 Loss of Feedwater Event. This capability will provide assurance of reactor decay heat removal during a postulated loss of all feedwater. The change involves the modification of the existing makeup line and installation of a redundant makeup line for single failure l
considerations. The makeup lines connect or will be connected to the High Pressure Injection System line to provide a direct flow path to the reactor. Removal of the existing SFAS signal to MU-33, and not providing an SFAS signal to the new valve, ensures feed and bleed will not be interrupted and precludes the possibility of the failure of the valve to reopen if it were allowed to close on SFAS. The modification and redefinition of the system function make the makeup system an engineered safety feature - related system for providing core cooling during an off-design basis loss of feedwater event. Remote manual isolation of the makeup system and other safety system design interface features prevent the failure of the makeup system from affecting any other safety system.
The design and function of the system are consistent with Standard Review Plan 6.2.4,Section II for engineered safety features or engineered safety features - related system and systems needed for safe shutdown of-the plant, and satisfies the requirement of General Design Criterion 55 for containment isolation. l It has been concluded that the change will not result in a significant l increase in the probability or consequences of a previously evaluated j accident nor create the possibility of a new or different kind of an I accident. A significant reduction in the margin of safety is not involved.
l Therefore, the proposed amendment is determined not to involve a significant l
1 hazards consideration.
1 a
Docket'No. 50-346 !
License No. NPF-3 Serial No. 1433 i Attachment 3 i
- Page 1 !
4 l
SIGNIFICANT HAZARDS CONSIDERATION Description This Significant Hazards Consideration supports a change to the Davis-Besse Technical Specifications, Appendix A of the Operating License, to delete the Safety Features Actuation System .(SFAS) closure signal from )
MU 33, the makeup (MU) line containment isolation valve. The change is l warranted because the makeup system serves a function similar to that of l engineered safeguards systems for a specific off-design basis upset,
- i.e., a complete loss of feedwater. Deletion of the SFAS closure signal
)
l remains consistent with General Design Criterion (GDC) 55, which identifies l isolation requirements on lines that are part of the reactor coolant pressure boundary and also penetrate the containment building. GDC-55 permits alternative containment isolation provisions for specific classes of lines if they can be demonstrated acceptable on some other i defined basis. The Standard Review Plan (FUREG-0800) defines acceptable alternatives.
Makeup System containment isolation valve MU 33 presently receives an automatic closure signal on SFAS incident level 3, which is actuated on either high containment pressure or low-low Reactor Coolant System (RCS) pressure. However, to compensate for an off-design basis event involving a total loss of normal and emergency feedwater systems, Davis-Besse relies upon the feed and bleed cooling method, in which the MU system provides the only source of post-trip decay heat removal. Flow will exit the Reactor Coolant System (RCS) via the PORV and/or the pressurizer code safety valves which results in subsequent containment pressurization.
The Davis-Besse feed and bleed modification will implement numerous modifications to the MU system and PORV in order to improve this cooling capability. (See the attached figure for the proposed modification diagram).
Calculations demonstrate that the 18.4 psia high containment pressure setpoint which initiates SFAS incident levels 2 and 3 will be exceeded early in a feed and bleed scenario and containment pressure may also approach the SFAS level 4 setpoint if only one containment air cooler is operating (under single failure considerations). Peak containment pressure has been calculated at approximately 36 psia, compared to a minimum allowable SFAS bistable field setpoint of 36.825 psia. In order to direct all available makeup flow to the RCS, the feed and bleed modification includes provisions to isolate the makeup pump minimum recirculation line. To prevent interruption of the feed and bleed process and potential damage to the makeup pumps by deadheading them, it is necessary for MU 33, and its sister valve in a parallel MU injection line being added to remain open throughout the transient.
Dockst No. 50-346 i License No. NPF-3 Serial No. 1433 Attachment 3 Page 2 This can be best accomplished by removing.the SFAS automatic closure feature from MU 33 and by not providing such a closure signal on the new containment isolation valve in the parallel train (MU 6421). This will ensure the greatest reliability possible for the enhanced feed and bleed' system while still maintaining the required containment isolation features.
The specific changes involve deletion of MU 33 from all locations in Technical Specification Table 3.3-5 to reflect elimination of its SFAS actuation capability. Table 3.6-2 must be modified to include the new remote manual operated containment isolation valve, FW 6421. An additional Technical Specification change addressed is to redesignated MU 33 as MU 6422 in Table 3.6-2. This is incorporated in the design to improve correspondence between the two MU trains from a human factors.
standpoint.
Systems Affected Containment Isolation Makeup System Safety Features Actuation System References
- 1. Technical Specifications 3/4.3.2 and 3/4.6.3 l
- 2. USAR Sections 3.6.2.7, 6.2.4, 7.3 and 9.3.4
- 4. NUREG 0800, Standard Review Plan (Section 6.2.4)
- 5. NUREG 0737, Clarification of TMI Action Plan Requirements i (Section II.E.4.2) l 6. Draft Regulatory Guide 1.141, Containment Isolation Provisions for Fluid Systems
- 7. ANS/ ANSI 56.2-1984, Containment Isolation Provision for Fluid Systems After a LOCA
- 8. ANS/ ANSI 51.1-1983, Nuclear Safety Criteria for the Design of l Nuclear Power Plants l 9. FCR 86-0412, Enhanced Feed and Bleed Capability
- 10. P& ids M-031 and M-033 l
- 11. Toledo Edison Check Valve Reliability Program Report.
1 Safety Functions and Functions of Systems The containment vessel isolation system is designed.according to the requirements of General Design Criteria (GDC) 54, 55, 56 and 57 of 10CFR50, Appendix A. Isolation valves are provided in lines penetrating the containment vessel to ensure no uncontrolled release of radioactive material can occur during a design basis accident. Penetrations not serving accident-consequences-limiting systems will either be closed by automatic isolation signals (SFAS) during such an event or be
4
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Docket No. 50-346 l
~
License No. NPF-3 {
Serial No. 1433 i Attachment 3 Page 3 4
maintained locked closed. The particular penetraticas involved (No. 19 for the existing line and No. 50 for the new one) are classified as Type IV penetrations according to Section 6.2.4 of the Davis-Besse USAR since they provide a flowpath for an engineered safety features (ESF) system, namely High Pressure Injection (HPI). Type IV penetrations do not have closure requirements since their design function is to open during a LOCA to permit passage of fluid used for reactor core cooling or containment depressurization.
The normal functions of the Makeup System are to maintain Reactor Coolant System (RCS) inventory during plant operation, to provide Reactor Coolant Pump (RCP) ceal injection and to continuously purify the reactor ;
coolant. Breaches in the RCS pressure' boundary which are within the
- I capacity of the MU System will be accommodated by that system and do not l result in actuation of engineered' safety features. In the event that all ,
feedwater is unavailable, core cooling is accomplished by feed and bleed. j This entails injection of water from the Borated Water Storage Tank (BWST) j into the RCS ueing the MU System and removing steam / water via the PORV i l
and/or pressurizer code safety valves after passing through the reactor l I
core. l The design purpore of the Safety Features Actuation System is to protect l
the reactor core during a loss of coolant accident (LOCA) and to mitigate j the consequences of a LOCA. This is accomplished by automatically I isolating the containment vessel to prevent release of radioactive fission l products and by initiating operation of emergency core cooling and )
containment depressurization equipment. The system consists of four identical and redundant sensing and logic channels and two redundant i actuation channels. Parameters monitored are RCS pressure, containment 1 pressure, containment radiation and BWST level. SFAS output logic i
consists of five separate levels, each actuating or providing permissive i
signals for specific systems and also establishing varying degrees of ;
containment isolation. 1 I
Effects on Safety l
General Design Criterion 54 requires that piping systems penetrating reactor containment "be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems." General Design Criterion 55 specifies containment isolation requirements for lines which are also part of the RCS pressure boundary. It states: ,
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5 Dock t No. 50-346 License No. NPF-3 Serial No. 1433 j
' Attachment 3 i Page 4 l
1 Each line that is part of the reactor coolant pressure boundary and -i that penetrates primary reactor containment shall be.provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on j some other defined basis:
-(1)- One locked closed isolation valve inside and one locked closed !
isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed ,
, isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside' containment; or (4) One automatic isolation valve inside and one automatic !
isolation valve outside containment.- A simple check valve may not be used as the automatic isolation valves outside containment.
Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the posit 1on that provides greater safety.
Other appropriate requirements to minimize the probability or l consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication,"and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.
The requirements for containment isolation systems presented in the Standard Review' Plan (SRP) elaborate on the CDCs. Specific. criteria necessary to meet relevant CDC requirements and guidelines for acceptable alternate containment isolation provisions for certain classes of' lines are presented in the SRP.Section II of SRP 6.2.4 enumarates system acceptance criteria along with allowable exceptions for engineered safety feature or engineered safety features-related systems and systems needed for safe shutdown of the plant. Remote manual isolation valves are
Docket No. 50-346 License No. NPF-3 Serial No. 1433 Attachment 3 Page 5 permitted in these situations provided possible leakage outside containment can be detected. This position is repeated in NUREG 0737 (Ref. 5). The nature of the makeup system during feed and bleed cooling is similar to that of an ESF system since it provides the only means of.
maintaining adequate core water inventory to preclude offsice radiation releases in excess of 10CFR100 limits.
The General Design Criterion states that automatic isolation valves shall take the posi' ion providing greater safety upon loss of actuating power. Both MU 33 and the new valve, MU 6421, will be maintained normally open to provide a makeup flowpath during power operation. Since these valves will be motor operated, they will fail "as-is" upon loss of power.
This is the position affording greater safety as it will permit the passage of water for core cooling upon a loss of all feedwater, thersby preventing possible core damage.
In the design of the two affected penetrations, along with the inside and outside isolation valves, appropriate features to minimize the probability and consequences of accidental rupture have been considered as required by GDC 55. These components are "Q," seismic Category I, and protected l against hazards such as pipe whip and jet impingement. Further, the !
interfacing MU System piping upstream of MU 33 is inherently precluded from creating any adverse consequences with regard to containment integrity and HPI operability, as explained in detail in a subsequent paragraph.
According to ANS/ ANSI 51.1 (Ref. 8) the performance criterion for primary l containment is to confine radioactive material leakage during any operational activity or event. These limits are prescribed by 10CFR50, Appendix I for normal operation and by 10CFR100 for transients or accidents. The system design should satisfy the requirements of ANS/ ANSI 56.2 (Ref. 7).
The latter standard delineates specific isolation system design requirements for satisfying these criteria.
, ANS/ ANSI 56.2 expands upon the GDC 55 requirements regarding containment l
isolation and defines a class of lines forming an exception to the provisions identified. It states that "the objective of containment isolation shall be to allow the normal or emergency passage of the following through the containment boundary while preserving the integrity l of the containment boundary: (1) Engineered safety feature system fluids,
- or (2) Fluids of systems which are not required to function following a loss-of-coolant accident but, if available, can be used to accomplish a 1 function similar to an engineered safety feature system." Examples gC/tn of such systems include PWR fluid systems required for RCP operation It is noted that the concept of " functions similar to engineered safety features" is applicable to MU 33. This standard further defines
! " safety-related function" as those plant features necessary to assure the l capability to prevent or mitigate the consequences of accidents that could result in offsite exposures comparable to the guidelines of
Docket No. 50-346 License No. NPF-3
~
Serial No. 1433 Attachment 3 Page 6 10CFR100. Although the feed and bleed cooling system does not serve in a safety related capacity according to traditional licensing criteria, i.e., direct mitigation of a design basis accident or transient, it )
provides the only means of core heat removal during a complete loss of feedwater event. Therefore, the system serves a " safety-related" function according to the above criteria. Isolation of the system by closure of MU 33 results in operation outside the analyzed conditions for feed and bleed and could result in core uncovery leading to potential radiological-release for this off-design basis event. As such, removal of the SFAS automatic isolation feature from MU 33 is appropriate.
Section 3.8 of ANS/ ANSI 56.2 discusses the requirements for the system providing the signal which initiates containment isolation (i.e., SFAS).
For those systems which employ phased isolation, as is the case at Davis-Besse, the standard requires automatic isolation of all systems except those serving ESF or ESF-related functions. It further states that only those ESF-related systems which can be justified to remain operational can be excluded from automatic isolation during the initial phase. This standard permits the isolation function associated with the preceding two categories of valves to be accomplished remote manually instead of automatically. Remote manual isolation is acceptable for ESF j or ESF-related systems to permit them to perform their accident j prevention or mitigation function provided certain conditions are j satisfied. These conditions include provision of features allowing the control room operator to detect possible failure of the fluid lines inside and/or outside containment to preclude degradation of containment isolation or the operation of ESF, and maintaining the capability to isolate the affected line remotely. An analysis of the consequences of a leak or line break, as required by Section 3.8.7 of ANS/ ANSI 56.2, will determine how quickly a non-isolated system'must be isolated in the event of a leak and whether automatic or manual isolation is appropriate.
The configuration of the existing and proposed interfaces between the MU and HPI Systems at Davis-Besse inherently precludes the possibility of a MU line leak or rupture affecting either containment integrity or HPI.
Check Valve MU 169, installed downstream of MU 33 and prior to the point of connection to HPI, is within the "Q" boundary of the HPI system and is intended to prevent diversion of emergency core cooling fluid-into the 3 MU System when the MU pumps are not operating. Any MU system leakage or !
rupture occurring upstream of MU 169 would not, therefore, impair the performance of the HPI system. Further, the makeup system piping is installed Seismic Category I, so line breaks need not be considered concurrent with an accident situation, and the pump discharge piping pressure rating is equivalent to that of HPI. Additional protection of l the low pressure MU suction piping is afforded by a relief valve with an 88 psig setpoint and 450 gpm flow capacity and by the pumps' discharge check valves MU 196 and MU 197. A special test conducted on these valves-in 1984 confirmed their ability to prevent reverse flow. A check valve reliability program established in 1986 recommended performing a detailed-1
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Dockst No.50-346, '; .dM1 , g;jlhf Li~ense c No.'NPF-3 ,
Serial No. 1433 3[
S.
1 3m
. Attachment 3, "
m Page 7" s' c-N. ...
p QW' internal; inspection'of either HP 57, MU'196, or MU 197 (these~are sids ar valves.and see similar service) during the.next' refueling in order.to.
ascertain possible. wear. Such inspect. ion will'be conducted in the upcoming g refueling outage and is part of- the hyta-Besse check valve reliability N
'and inspection program.= "', __
~dithough credit.cannot be taken for MU 169 serving a Containment '
isolation function,'since GDC;55, prohibits consideration of check valves %y '
as automatic. isolation' valves outside containmen3, containment integrity. ,,
will still be assured. For the situation of HPI_ train-2 operating during '
N-a LOCA, Penetration 19 vill reniain pressurized subptautially above any : (> g pressures possible in the containment atmosphere, effectively eliminating ..e iJ the possibility of fission product Icakage via this flowpath. Should HPI pump 2 fail to operate during a' design basis event, resulting in?the penetration remaining depressQrized, containment. integrity is assured by the redundant containment isolation valve,inside? HP 57. -An additional failure of this check valve is beyond'the. single failure briteria.and need-not be postulated.. Check Valv'e HP 59, in series with HP 57; underwent.
internal inspection and modification to its disc anti-rotdtion stops in 1984 to reduce its chances.of-failure', thereby providing1 1ncreased confidence =
of proper seating. Additional: recommended evaluations based on'the check valve reliability program Mre described above. The HPI function for the condition of one HPI pump inoperable will be provided by.the redundant-train. The makeup system is in continuous operation during normal plant-operation. As such, any leakage would be observed and appropriate ..
corrective measures implemented to ensure a leak-tight system'is maintained.
l Technical Specification 6.8.4(a) requires establishment of.a prog, ram to minimize leakage outside containment from several systems, including-makeup. The program consists of preventive . maintenance and/or vthnal D-inspection requirements along with periodic integra.ted system lekkage testing. These existing requirements and operational monitoring 3re considered sufficlent to preclude' system leakages during.an event. requiring containment isolation. Further, the MU system is'not' intended to circulate fluid from the containment emergency sump; pump suction is restricted to either the MU tank or BWST. This will be insured by< appropriate procedure modifications.
Since the low-low RCS pressure signal is combined with the containment high pressure signal in the SFAS logic, this. change also' removes the . 4 closure signal for these valves on low-low RCS pressure. Closure of, thel h MU system containment isolation valves on low-low RCS pressure during'a W j s design basis LOCA is not a necessary function. . The.MU pumps will normally continue to operate, thereby increasing the_flowrate of borated-water into the reactor vessel and enhancing core cooling. ..Further, as discursed above, the containment isolation fvnetions remain satisfied even if the penetration is depressurized due to the MU and HPI pumps not ; , , %[
operating.
,of UUwy, -r 4 y
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L i Docket No. 50-346 i License No.'NPF '
Serial No.'1433 :]
Attachment 3 -
Page 8 The failure ~ evaluation presented ab'ove satisfies the GDC 55 requirement to-miaimize the probability and consequences of an accidental rupture'of?
such lines to ensure adequate safety and meets the' intent of,the.
requirement in ANS/ ANSI 56.2 to analyze the consequences of a leak or i break in.an unisolated line.- Further, since all protection features j which assure containment integrity and preclude degradation of HPI are 'i inherently provided, no time requirement for isolation of MU 33 is warranted. Therefore,1this. isolation capability can be provided as j^
remote manual from the control room rather than automatic. Reverse; flow testing of MU 169 is presently part of.the periodic valve test program to ensure proper functioning of the check valve. This valve will continue ..f to be tested, along with the corresponding one in the new line being ,
'r>
installed. Similarly, the Type C leakage test requirement will be 9 i maintained for MU 33 and established for MU 6421 to ensure the valves will- '{ '
remain capable of isolation if closed remote manually. 7 The preceding discussion applies equally to the new makeup line being installed. This line will utilize Penetration 50 and will' have a '.'Q" check valve comparable to MU 169 between.MU and HPI.
I Significant Hazards Consideration The proposed changes do not involve a significant hazards-consideration because the operation of the Davis-Besse Nuclear Power Station,' Unit 1 in -
accordance with these changes would: ,
- 1. Not involve a significant increase in the probability or- l consequences of an accident previously eveluated (10CFR50.92(c)(1)). i The probability or consequences of an accident previously evaluated
~
)
is not significantly increased because containment isolation and other engineered safety feature functions are not affected. The ability of existing Emergency Core Cooling. Systems (ECCS)Lto' function as required remain within the bounds of the existing LOCA analysic.
These changes will serve to reduce the probability of core uncovery resulting from a complete loss of feedwster and, therefore, the consequences of a complete loss of feedvater is also reduced.
- 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated (10CFR50.92(c)(2)). ;
The proposed Technical Specification changes will not create the possibility of a new or different kind of accident-from any'previously. '
evaluated because containment isolation, capability is maintained.-
MU System failures will not adversely affect HPI performance due to !
the presence of check valves et the system interfaces. Though the- l automatic closure feature of MU-33 and MU-6421Lwill be deleted, remote manual isolation capability is.still provided to ensure valve q closure, if required. Failure of the valve to close on demand l remains the only failure mode, which is consistent with the present-situation. O j
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License No. NPF-3 -[
Serial No. 1433 '9' Attachment 3 .
- M-Page-9 q 1- , ,
g Not involve a significant' reduction in the margin of safety
~
- 3. {
.- (10CFP$0.92(c) (3)) .
al d The propo' sed Technical Specificati6n"chay,e.s do not involve a 1 significant: reduction in the margin'of-safety because limitation of l ra'dosptive..nateritileakagethroughithepenetrations.isasassumdd,
-!. 'j in the hxisting safety analysis for any design basis event. The ' '
l margin of safety fot the off-design bapis complete.2oas of feedwater i y
event will be increared due to itarrovsd reliability of the / -
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Conclusio'n ;
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( i1 Therefor ~e, it is concluded that the propoJ ed T'/chnical Specification g changes do not involve a significant hazards ccarideration, t
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