ML20212P469
| ML20212P469 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/10/1987 |
| From: | Julie Ward SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20212P471 | List: |
| References | |
| JEW-87-359, TAC-64896, NUDOCS 8703160120 | |
| Download: ML20212P469 (12) | |
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$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT C P. O. Box 15830, Sacramento CA 95852-1830.(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA JEW 87-359 i-March 10, 1987 Director of Nuclear Reactor Regulation Attention:
Frank J. Miraglia, Jr.
Division of PWR Licensing-B U S Nuclear Regulatory Commission Washington D C 20555 Docket 50-312 Rancho Seco Nuclear Generating Station Unit #1 PIANT OPERATION WITH EITHER OR BOTH DHS TRAINS NOT OPERABLE
Dear Mr. Miraglia:
)
At the request of the Rancho Seco Project Manager, Syd Miner, attached for your information is a copy of the 50.59 Safety Analysis Log No. 889 Rev. 2 written for Operating Procedure B.2A, Plant Operations with Either or Both DHS Trains Not Operable and Casualty Procedure C.12, Loss of Decay Heat Removal System. The copies of Procedures B.2A Rev. 2 and C.12 Rev. 5 are the final issues with the Plant Review Committee (PRC) comments incorporated and for which the attached Safety Analysis is applicable. Also attached are a District Memo TS87-047 and B&W's calculation #32-1167923-00 referenced in the safety analysis.
If you have any questions, contact Bill Kemper, Nuclear Operations Manager, of my staff at extension 4353.
Since y,
ar Deputy General Manager, Nuclear Attachment cc: Syd Miner, NRC - Bethesda A. D'Angelo, NRC - Rancho Seco h00(
$N~AEi!R ggg;
'h DISTRICT HEADQUARTERS O 6201 S Street, Sacramento CA 95817-1899
)
E Log. N 3. - 850 Rev. 2' SAFETY REVIEW OF PROPOSED PROCEDURE CHANGE
- 1. DESCRIPTIC'J:
See Attached B.2A Rev. 2 & C.12 Rev. 5 IN
-PROCEDURE dN, DESIGN BASIS REPORT REQUIRED:
Yes C ' No %
NUMBER OF PAGES /w/arrachmentsl:
- 2. PRC RECOMMENDATION
- e. Change in Procedures as Described in SAR
.Yes @ No Q
- c. Unreviewed Safety Question
- Yes ) No @
h Technical Specification Addition
- Yes U No E
- d. Technical Specification Change
- Yes l
! No 2El
"^
DISPOSITION OF PRC
- a. Concurs with SA Evaluation and Findings....E
- g. 50.54(p) Security Review Required........ [
tL Recommends Proposal..................
- h. 50.54(q) Emergency Program Review Required a
- c. Send to MSRC for Concunence...........
L 50.54(al Quality Program Review Required... _!
- d. Procedure Change Rejected..............
t;u I) outad
- e. Retum to Cognizant Individual............
implementing Conditions
(' -- N
- f. MSRC Approval Prior to implementing......
1 FGu e.) # CM 4r #or d
C e i d I(4 s 't u J
Win 2.~ 27-97 inc cowRMAN ~
DATE
- 3. NUCLEAR OPERATIONS MANAGER REVIEW
- c. Unreviewed Safety Questiorf Yes No d IL Technical Specification Change Yes No 5
- c. Change in Procedures as Described in SAR Yes No DISPOSITION OF NUCLEAR OPERATIONS MANAGER
~~
s E No 0 impiemen6 Conditions
- a. Proposal May Proceed it Refer to MSRC.........'..............
I
- c. Procedure or Change Rejected............
W M
U v
NUCLEAR OPERATOf MANAGER
/ DATE 4
/
I. MSRC FINDINGS.
Yes C No O
- a. Unreviewed Safety Question tL Technical Specification Change Yes[
No
- c. Change in Procedures as Described in SAR Yes L' No I
.. DISPOSITION OF MSRC
- c. Recommends Proposal................. O
- d. Retum to Cognizant Individual............ l[
tL Send to NRC for Approval............... h
- e. Procedure or Change Rejected............ l_
- c. Proposal is Not to Proceed.............. J._
MSRC CHAIRMAN DATE
- 5. COMMISSION APPROVAL.OBTAINED: (if applicac/e/
MSrcC CHAfRMAN DATE
( 6. RETEST COMPLETE AND ACCEPTABLE: /// applicable /
- 7. OVERALL REVIEW: Procedure Change Complete. II/applicablel NUCLEAM TECHN' CAL SUQPORT SUPER.TENDENT CATE MAP.AGER RESPONSISLE CEoARTr.*ENT CATE
- 8. DOCUMENTATION COMPLETE:
- 9. ADDED TO MONTHLY REPORT Report Date OLoALITY MANAGER C.;'!
sMuo4964 2/04
m
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SAFETY ANdLYSIS RANCHO SECO NUCLEAR GEfJERATING STATI'ON Log No. 889 Rev. 2
! PRocEo.;81 ? UM5ER Af D TAE B.2A Rev. 2-P]ar.: C:crs: ion Ir. Cc1d Shutdown Uith Both DHS Trains Not Operable C.12 Rev. 5-Loss Of Decay Heat Esmoval Svsten
==
Description:==
(A description of the prepcsed procedure or procedure change.)
See Atteched R;ason for the Proposed Procedure or Procedure Change:
(A statement as to why the action is being requested.)
See Attached Evaluation and Basis for Safety Findings:
(The evaluation will address itself to specific sections of the SAR or Technical Specifications as applicable. Any effect on nuclear safety will be described.)
=
See Attached Safety Findings:
Yes No l
@ - O The proposed change is a change to the facility as described in the SAR.
O W The proposed change does involve an Unreviewed Safety Question (if an Unreviewed Safety Question is involved, check the appropriate reason).
O Probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased.
O Possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created.
O The margin of safety as defined in the basis for any Technical Specification is reduced.
O G The proposed change does involve a change in the Technical Specifications.
O E The proposed change does involve an addition to the Technical Specifications.
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UCENslNG ENGINEER oATE REVIEW ENGINEER oATE UCE ' ING MANAGER oATE
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FR::I:".I SAFETT A' ALYSIS LOG N0. 889 RE. 2
- 3.
- A, FIAN OPII.CICN IN CC13 SHUTDOWN WITH B;TH OES !?_CNS NOI ~?ILA.BLE C._* 2, LOSS OF LICAT HEAT RI"~,TAL SYSTEM PAGI 1 0F 7 i
s
- DIS:E PTION:
i Ekheir ot Operating Procedure 3.2A Rev. 2, Plant Operation in Cold Shutdow: WithABoth
'E DHS Trains Not Cperable, dated 2/24/87, provides major operator guidance for preparing the plant and caintaining cold shutdown, using the OTS3s for RCS hear removal, while neither DHS Train is operable.
Casualty Procedure C.12, Revision 5, Loss of Decay Heat Removal System, dated
/b
-2/24/87 is revised to provide additional operator guidance upon the loss of operability of both decay heat loops, with the RCS filled and verted.
Some of this revision is not generically applicable and will be removed at a later date.
3.2A Eev. 2 and C.12 Rev. 5 are attached.
lihi REASON FOR CHANGE:
Pur:ose The preparation of 3.2A is necessary to provide a normal operatirg procedure for the planned evolution for removing both decay heat loops fro: service for maintenance and repairs.
The revision to C.12 is an enhancement which provides more detailed guidance for maintaining decay heat removal using the OTSGs in the event that both DHR loops are inoperable with the RCS filled and vented and RCS temperature j{g exceeds cold shutdown.
EVALUATION AND BASIS FOR SAFETT FINDINGS:
Systems, Subsystems, and Components / Safety Functions Affected i
i I
Both procedure actions detail the possible use of the steam generators for decay heat removal. The use of steam generators for decay heat removal with RCS reaperatures below 2800F is not addressed in the USAR. Therefore, these procedure actions represent a change to the plant's procedures as described in the USAR.
Residual heat removal capability is necessary "to transfer fissio: product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reacror coolant pressure boundary are not exceeded" (USAR Section 1.5.30, criterion 34-Residual Heat Re= oval). Two separate and distinct cethods are provided for residual heat removal: 1) the steam generators, and 2) the Decay Heat Re: oval System (USAR Section 9.5).
USAR Section 1.5.30 cost fully discusses the application of the steam generators for decay heat removal,
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l PROCEDURE.SAFLTl ANALYSIS LOG NO. 889 LE7. ;
B.2A, PIANT OPEP.ATION IN COLD SHUTDOWN WITH BOTH DHS-TF.AINS NOT OPERA 3LE C.12, LOSS OF. DECAY HEAT REMSVAL SYSTEM PAGE 2 0F 7 l
Systems, Subsystems, and Cocoonents/ Safety Functions Affecred (Cont.)
including such considerations as feedwater supply, secondary steam venting, and RCS circulation.
The C.12 revision is consistent with this descripti:n of the procedure for employing the' steam generators for decay heat removal.
The method detailed in B.2A for applying the steam generator's heat removal i
capabilities at RCS temperatures below 2800F has not been described in the
~
USAR.-
_;s 1
- According to the USAR Section 1.5.30, the steam generators are used for decay heat removal "until the reactor coolant system is cooled to where the decay heat removal system becomes operational". The USAR references this distinct j
transition temperature as 2800F (USAR Sections 4.2.3.4-RCS Cooldown, 4.2.4.6-RCS Cooldown, 9.5.2.1.1-DHR System,10.2.2.1-Feedwater Supply).
Section 4.2.3.4 is representative:
" System cooldown to 2800F is accomplished by use of the steam generators and by bypassing steam to the condenser with the turbice bypass system.
The decay heat removal system provides the heat recoval for system cooldown below 2800F".
In actuality, while the decay heat removal system is the preferred method of
.cooldown below 2800F, the steam generators can also be employed in this temperature range.
Further, Technical Specification 3.1.1.5, Decay Heat Removal, requires the steam generators to be OPERABLE for decay heat removal below 2800F, dependent on the operability status of the DER Systen trains.
I-The " Bases" of Technical Specification 3.1.1.5 state, "When TAV is below 2800F, a single reactor coolant loop or DHR loop provides sufficient heat removal capacity for removing decay heat; but single failure considerations require at least two loops to be OPERABLE".
The USAR requires updating to address the capability of using the steam
generators for decay heat removal below 2800F, including clarification that the DHR system is the preferred system in this range.
Consideration should y also be given to enhancing USAR Section 4.3.8.4, RCS Minimum Operational Components, to more closely reflect Technical Specification 3.1.1, RCS Operational Components LCO.
Technical Specification 3.1.1.5 was added by License Amend:ect No. 71, approved July 3,1985. This Specification was required by NRC Letter (D. G.
[
Eisenhut to All Operating PWRs, June 11, 1980) because "a nu:ber of events i
have occurred at operating PWR facilities where decay heat re: oval capability has been seriously degraded due to inadequate administrative controls during shutdown modes of operation" (NRC SER on Amendment No. 71). The SER further j
states, "The revised Technical Specifications which are consistent with the NRC Standard Technical Specifications provide an improve:ent over the existint ones since -redundant decay heat removal will now be provided during hot
1 PROCED3E SAFETY ANALYSIS LOG NO. 889 REY. 2
~B.2A, PLANT OPERATION IN COLD SHUTDOWN WITH BOTH DHS TRAINS NOT OPERABLE C.12, LOSS OF DECAY HEAT RFS.0 VAL SYSTEM PAGE 3 0F 4 Systems, Su'bsystems, and Components / Safety Functions Affected (Cont.)
standby, hot shutdown, and cold shutdown....We therefore conclude that the proposed Technical Specifications provide an improvement over existing Technical Specifications with respect to redundant means of decay heat renoval capacity and are acceptable".
Pursuant to Technical Specification 3.1.1.5, in order for one reactor coolant loop to be OPERABLE for decay heat removal, the reactor coolant loop and associated steam generator and at least one associated reactor coolant pump must be OPERABLE. The definition of OPERABLE (Technical Specification 1.3) includes that the system must be " capable of performing its intended function within the required range" and that the " required auxiliaries are capable of performing their intended functions".
To take credit for the capability of steam generator decay heat removal, a number of systems are required, including emergency feedwater, main steam, seal injection and makeup, and emergency power (see attached Compliance interpretations and guidance).
OP B.2A provides an operating procedure for the planned evolution of filling 3
and venting the RCS, lining up the necessary equipment to provide OPERABLE decay heat removal capacity via the steam generators, direction for maintaining the plant in cold shutdown, and contingency guidance for the case that cold shutdown can not be maintained. OP B.2A maintains compliance with Technical Specification 3.1.1.5.
,7 4
i Effects on Safety Functions and Analysis of Effects The steam generators are normally used for decay heat removal 3bove 2800F (USAR Sections 4.2.3.4, 4.2.4.6, 10.2.2.1).
The use of steam generators for g
decay heat removal below RCS temperatures of 2800F is clearly acceptable per the requirement of T.S. 3.1.1.5.
At cooler RCS temperatures, the RCS stored energy is less. The RCS is closer to its thermally static state. With forced RCS flow, the steam generators can maintain RCS temperatures below 2200F, steaming through the ADVs (note C.12 Step 2.1.7.12).
With the Condenser and l
TBVs available, it may be possible to maintain RCS temperatures below 2000F.
l With low levels of decay heat, it is also possible to maintain RC temperatur s e
below 2000F by supplementing heat loss to ambient by filling and draining the steam generators. Enclosure 5.2 of Procedure B.2A Rev. 2 details this supplemental cooling method. This method drains the OTSG using the normal method to the condenser via the OTSG drain booster pumps.
OTSG fill will be from the CST by gravity or using the AFW pumps.
b B&W has analyzed this method of supplemental decay heat recoval using one OTSG in B&W 32-1167923-00. Based on B&W's analysis, approximately 20 gpm fill and drain rate will maintain the RCS temperature less than 1500F.
Fill water from 6>s the CST will meet OTSG makeup water chemistry requirement.
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PROCEDURE SAFEri ANALYSIS LOG NO. 889 REV. 2 I
B.2A, PLET CPEMTIT* IN CT 3 S:CDOWN WITH BOTH DHS ! MINS NOT CPEMBLE-C.12, LOSS OF DICAY HEAT REM 37AL SYSTEM.
PAGE 4 0F 7 Effects on Safety Functions and Analysis of Effects (Cont.)
This drain & fill of the OTS3's vill provide substantial margin for acconplishing scheduled caintenance activities and a buffer for any unexpected contingencies.
However, the primary method for decay heat removal, steam venting from the OTSG's, will be available throughout the decay heat outage and this supplenental fill & drain procedure will not be relied on for long term decay heat recoval.
5
_~Although the requirement of T.S. 3.1.1.5 will be satisfied during the DHS outage, ty) to RCS temperature of 2800, it was felt appropriate to revise Casualty Procedure C.12 to provide more detailed guidance for removing decay
. heat using operable Reactor coolant loops should RCS tecperature exceed 1900F. Casualty Procedure C.12 provides cultiple contingencies for steam venting paths from the OTSG's. RCS tenperature, sub cooling margin,' and circulation are specifically monitored. The incore thermocouples provide indication of the hottest RCS temperatures. With the DHR System out of service, natural circulation should be established in the RCS without operator action.
C.12 provides for establishing and verifying conditions required for natural circulation.
The plant conditions upon entering Casualty Procedure jg C.12 are bounded by the Complete Loss of All Unit AC Power Accident, USAR Section 14.1.2.8.
The associated analysis takes-credit for generator steaming and natural RCS circulation.
The USAR Section 14.1.2.8 analysis bounds the use of the C.12 revision.
(
For the present plant conditions, with the plant shutdown since December 26, 1985, greater than 46 weeks, the decay heat being generated in the core is i
very small.
"A conservative analysis (Calc. #Z-RCS-N0002) has been performed j{'
that estimates the heat up rates of the RCS at 3.90/hr. This analysis ia extremely conservative in that it limits the available heat sinks to the liquid inventory in the reactor vessel in the partially drained condition and l
the vessel itself. This would be a conservative assumption for a worst case in the initial stages before natural circulation is established" (from SAR Log No. 882, Revision 1).
In addition, B&W has performed an evaluation for Rancho Seco (Report #32-1167923-00, attached) which shows that natural circulation
. will safely core decay heat at low decay heat generation rates (196 kw). With the OTSG 1evels at 85% on the operate range and only accounting for convective and radiative heat losses, heatup from 950F to 1900F based on the B&W report, will take approximately 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />. STP 1011, Determination of Decay Heat j$5 Load, will be performed to more accurately quantify expected heat up rates in the present plant condition.
The procedure recognizes that at low decay heat loads, natural circulation will be intermittent and may take hours to days to develop.
Section 4.4 of Procedure B.2A and Section 2.1.7.9 of Procedure C.12 l
note this " slug-flow" natural circulation phenomenon which may be expected during the DHR outage.
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PROCED'JRE SAFETY ANALYSIS LOS NO. 889 REV. 2 3.2A, PLANT OPI?>.C ON IN COLD SH C D M VI H BOTH DHS TRAINS NOT OPEPJ.BLE C.12, LOSS OF DICAY HEAT REMOVAL SYSTEM PAGE 5 0F 7 Effects on Safety Functions and Analysis cf Effects (Cont.)
Considering the length of the scheduled =aictenance outage of bcth DER trains, and the "best-guess" RCS heat up rate preiictions, w'ith no decay heat removal in operation, it is not expected that the RCS temperature will exceed 2000F.
B.2A Step 4.3.1 requires a calculation of the time allowed to have the DER Pumps out of service and maintain Incore ~hercocouple temperature less than 1900F.
This calculation will be based on the results of STP.1011.
With the intent to assure RCS temperatures remain below 2000F, Enclosure 5.2 of B.2A details a method for supplemental decay hsat removal by filling and draining
/kg the OTSG's beginning at 1500F. There is reasonable assurance this contingency cooling will be fully adequate to caintain the RCS maximum temperature below 2000F. However, should the RCS te:perature increase to 1900F, Casualty Procedure C.12 will be implemented which provided for steaming the steam generators either to the condenser or to the atmosphere.
/kg Steps-3.7 and 3.10 ensures that no RCS boron concentration reduction evolutions occur without at least one RCP or DERP operating. Reactor shutdown '
cargin is maintained per OP B.6 (note B.2A Step 3.9).
USAR Section 3.2.2.1.5.D " Moderator Temperature Coefficient" discusses a reduction in boron concentration caused by reactor coolant heatup.
Clearly this is a reference to boron atoms /Cm3 i
change due to the effects of density changes caused by heatup of the water and does not refer to chemical boron concentration changes in ppa caused by the addition of water with a boron concentration less than l
that of the RCS. The requirement in Technical Specification, not to reduce boron concentration unless there is forced core flow applies to a chemical reduction in concentration rather than to an atos/Cm5 reduction due to temperature increase.
Increasing RCS temperature with no forced core flow is concluded to be in conformance to Tech Spec 3.1.1.1.B and USAR 3.2.2.1.5.D.
l/hg Throughout Procedure B.2A, precautions and administrative controls are used to protect against inadvertent boron dilution of the RCS from occurring and to maintain a shutdown margin of at least 1% delta k/k.
During the Decay Heat removal System Outage, administrative controls will be put on RCS makeup.
i Makeup will be from the BWST or the RC Drain Tank. Both sources will provide makeup with boron concentration slightly higher than the RCS.
A third method of makeup using the boric acid blender will be available but i
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will be strictly controlled to eliminate a possible source of dilution water, m
This mode of makeup is the third in the descending order of makeup sources during the decay heat outage. These provisions in Procedure B.2A provide adequate assurance against inadvertent boron dilution during the decay heat outage and are consistent with the require:ents of T.S. 3.1.1.1.B.
The boren concentrations for the current fuel burnup to maintal: a shutdown margin of 1% delta k/k is 1300 ppe assuming an RCS temperature of 680F.
The
'RCS boron concentration during the DHS outage will be above 1800 ppe. This
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PROCEDURE SAFETY /JALYSIS LOG NO. 889 REV. 2 B.2A, PLANT OPEFaTION IN CCID SHl.CDOWN WITH BOTH DHS I?aINS NOT C?IEA3LE C.12, LOSS OF DECAY HEAT F.I_ G AL SYSTEM PAGE 6 0F 7 Effects on Safety Functie:s and Analysis of Effects (Cont.)
will ensure that for all anticipated RCS temperature changes, adequate shutdown margin wil1~ be =aintained.
The boron concentration will be verified at lease once per day.
Stratification or changes of boron concentrations,in the RCS caused by low flow or no flow conditions was -investigated and documented in attachment 1 for
/
the DHR outage. Attachment 1 concludes that for the DHR outage (boron concentrations between 80-2S00F) ;the boron will remain in the solution and be evenly dispursed in the RCS by physical properties and by mixing caused by natural circulation. No techanises were identified for developing pockets of low boron water in the RCS curing the DHR outage.
As previously addressed, both RCS loops will remain operable throughout the DER outage in conformance with T.S. 3.1.1.5.
No Licensing Condition has been identified that prohibits leaving Cold Shutdown (see attached December 26, 1985, Confirmatory Action Letter, Martin to Rodriguez, which requires that 1
" prior to return to power, you will provide the NRC a briefing of your assessment of the root cause and your justification as to why the Rancho Seco facility is ready to resume power operations".)
No change in Technical Specifications Section 3 operability requirements upon exceeding Cold Shutdown have been identified. Technical Specification 3.1.1.5 specifies the decay heat removal requirements up to 2800F. All applicable Surveillance (Chapter 4) requirements to meet 3.1.1.5 operability requirements have been fulfilled. As required by Tech Spec 3.2.2, Low Temperature Overpressure Protection (LTOP) will be provided throughout the DHR outage.
The. applicability of T.S. 3.2.2 is addressed in Enclosure 5.1 of Procedure I
B.2A.
No other sources of overpressure have been identified that would affect the LIOP or the bases of T.S. 3.2.'?.
Containment Integrity (Technical Specification 3.6) and CR/TSC Essential HVAC Operability (Technical Specification 3.13) requirements are not applicable with the reactor t
suberitical by more than 1 % delta k/k.
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Neither of the subject procedure actions requires a change or addition to l
Technical Specifications.
I An RCS LOCA is not a required assumption. High energy line breaks are not considered for lines which do not exceed 2000F and 275 psig (USAR Section 4.1.2.3).
In the unlikely event that RCS inventory is lost, or that multiple failures of the steam generator decay heat removal capability occur, alternate i
methods of removing decay heat are available (see Casualty Procedure C.12 and USAR Section 9.5).
Acciden: analysis in USAR Chapter 14 are not affected since decay heat removal via ECS natural circulation and OTSG steaming is
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included in the Chapter 14 analysis.
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PROCEDURE. SAFETY A'!ALYS 5.
LOG NO. 889 REY. 2
' S.2A, PIANT OPERATION IN C~:.' SI~' D U *C BOTH DHS I?i1NS NOT ~?IRAl*_I C.12, LOSS OF DECAY HEA~ ?I:"VE SYS!T.-:
PAGE 6A 0F 7 Proposed Am' ndment No. *A5 r_b:.::ed en September 22, 1986 requested a waiver e
en surveillance of the Rea :::7essel In:ernal Vent Valves (RVVV) until Cycle
- 8 refueling.
However, it is nc: er;ec:ei that the approval of Proposed
. Amendnent No.145 will.be in ti:e fc: the decay heat outage. Therefore, the cperability of the RVVV is disc:ss ed b el:w with regard to the Decay Heat outage.
The RVVVs are large swing check va.1ves m:unted vertically between the inlet and outlet sides of the core su;pc t shield. The core support shield directs cold leg (inlet) flow-downvard into the annular space ~ just inside the vessel and contains core outlet flo - 1: the cen::a1 portion, directing it upward to the hot leg nozzles.. The vent valve a.ssenblies are installed so they can swing outward into the old ~ eg water spacer * 'should pressure on the outlet side of the core exceed ficv in the central portion, directing it upward to the hot leg nozzles. The vent Tal.ve assenblies are installed so they can swing outward into the cold ~ eg wa.te: space should pressure on the outlet side of the core exceed in*ec p:sssure. During normal operation and most plant transient conditions, these valves are held closed by both gravity and j
the higher pressure on the ccre inlet side.
The safety function of the reac:or vessel internals vent valves is to relieve the pressure that could be gene ated b r Toiding in the core following a loss entry of Energency Core Cocling Systen (ICCS) injection water of coolant accident. Steam for:ed in the core that would otherwise block the o N
through the valves allowing the core to remain covered. In the unlikely of a rupture of the reactor inlet pi;ing, the vent valves promote expulsion of steam produced in the core dire:tly to the break, thus enhancing the effectiveness of ECCS injection. The valves serve a safety related function,
(
and are subject to an inspectio: and m.ancal actuation to fulfill the following purposes:
- 1) Ensure valve operability l
- 2) Ensure that valves are not stuck ope: during normal operation
- 3) Demonstrate that the valves are fully open at the forces equivalent to the differential pressure assumed in the safety analysis.
During the decay heat outage, the plant vill remain in cold shutdown. Under these conditions and for RCS temperatures up to 2800F, the internal vent valves serve no safety function since a loss of coolant accident of a magnitude which would require the functioning of the RVVV is not a credible accident. Long term core coeli:g via DEE or RCS natural circulation is in
- effect at all times. The inter _al var valves are held in their required closed position primarily 17 gravity a.s indicated above for these modes of operation. Therefore, for F. S :e:peratu:e below 2800F, the reactor vessel internal vent valves are ::: required to be operable as pa:it of the reactor coolant loop for decay hea: ren: val.
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PROCEDURE SAFETY ANALYSIS
'. : : N:. 85:- R.E7. 2 B.2A, PIANT OPERATION IN COLD SHUTDOWN-k'ITH BOTH DRS TRAINS NOT OPERABLE C.12, LOSS OF DECAT HEAT REMOVAL SYSTEM
- 17E 5B 07 7
- Should the RVVVs become technically not operable because the r.:reti" lance interval for* inspection has lapsed, there is no effect en the :perabi2ty of reactor coolant loops used to satisfy Tech Spec 3.1.1.5 fcr :ht decay hat d
outage.
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PEOCED'JRE SAFETY A%'ALYSIS LOG NO. 889 REV. 2 B.2A, PIANT OPERATION IN COLD SHUTDOWN WITH BOTH DHS TRAINS NOT OPERABLE C.12, LOSS OF DECAY HEAT RIMOVAL SYSTEM PAGE 7 0F 7 Effects on Safety Functions and Analysis of Effects (Cont.)
These procedure actions will not result in system operation in any way that may increase the probability or consequences of any accident previously analyzed or that may result in an accident of a different type. The margin of safety as defined in the basis of Technical Specifications (particularly 3.1.1) is maintained. Therefore, these procedure changes do not involve an unreviewed safety question.
Summary These procedure actions do require a change to the USAR to address the use of steam generators for decay heat removal below RCS temperatures of 2800F. A change or addition to Technical Specifications is not required.
The probability of occurrence or the consequences of an accident or malfunction of equipment imp ~ortant to safety previously evaluated is not increased because the plant will remain in cold shutdown with required redundant methods of decay heat removal operable per T.S. 3.1.1.5.
The possibility for an accident of malfunction of a different type is not created because there are no changes to plant configuration or modes of operation not previously analyzed.
The margin of safety as defined in the basis for any Tech Spec is not reduced because the requirements of applicable Tech Specs is maintained throughout the DER outage. Therefore, the proposed procedure changes do' not involve an Unreviewed Safety Question.
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