ML18026A266
| ML18026A266 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/28/1994 |
| From: | James Shea Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9410050332 | |
| Download: ML18026A266 (22) | |
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)W UNITED STATES NUCLEAR REGULATORY. COMMISSION VIASHINGTON, D.C. 20555-0001 September 28, 1994 LICENSEE:
Pennsylvania Power and Light Company FACILITY:
SUBJECT:
Susquehanna Steam Electric Station, Units 1 and 2
MEETING BETWEEN MESSRS.
DONALD PREVATTE AND DAVID LOCHBAUM AND NRC STAFF CONCERNING SUSQUEHANNA STEAM ELECTRIC STATION SPENT FUEL POOL COOLING ISSUES On September 6,
- 1994, the NRC staff (the staff) met with Messrs.
Donald Prevatte and David Lochbaum.
The meeting was a part of a continuing dialogue between the staff and Messrs.
Lochbaum and Prevatte concerning the report filed pursuant to 10 CFR Part 21 on November 27, 1992, describing potential design deficiencies in various spent fuel pool cooling (SFPC) systems at the Susquehanna Steam Electric Station.
The purpose of the meeting was to exchange information on the staff's assessment of certain specific technical issues raised in the Part 21 report.
The staff opened the meeting by describing possible sequences of events flowing from the initiating events postulated in the Part 21 Report.
The staff used a draft flow chart, provided by the licensee, to describe the systems and the operator actions available to aid recovery from a variety of loss of SFPC events.
The licensee had provided the draft flow chart to the staff prior to the meeting in response to the staff's questions regarding overall strategy for coping with postulated loss of SFPC events.
The document used at the meeting was a draft that the licensee has indicated will eventually be finalized and incorporated into existing emergency procedures.
The draft flow charts are attached as Enclosure 1.
The staff stated that, for the purposes of the meeting, the staff's discussions would recognize the current plant configuration as reflected in recent licensee correspondence.
Initial discussion focussed on the non-safety related SFPC system.
The staff described the availability of power to the SFPC system and supporting service water system following various initiating events.
The capability and qualification of recently installed remote indicating spent fuel pool level and temperature instrumentation as well as the status of procedure upgrades reflecting use of the new instrumentation was discussed.
Actions to initiate makeup to the spent fuel pools was also discussed.
Messrs.
Lochbaum and Prevatte commented that actions to isolate the non-accident unit from the accident unit were important for assuring operator access for makeup activities in the non-accident unit and that these actions should be reflected in procedures or on the flow chart, The discussion on the SFPC system concluded with Messrs.
Lochbaum and Prevatte's comment that adequate net positive suction head (NPSH) for restarting the SFPC pumps at elevated pool temperatures had not been evaluated.
The staff agreed that an assessment of this issue would improve overall understanding of the SFPC system's capability.
9410050332 940928 PDR ADOCK 05000387
'P gIg PIL< CKMTEE CIIP>
The next phase of the meeting focussed on the capability of the residual heat removal (RHR) system operating in the fuel pool cooling (FPC) assist mode.
The staff described pre-operational test results, calculations provided by the
- licensee, and the staff's assessment of available net positive suction head, which formed the basis for the staff's conclusion that the RHR system could provide adequate cooling in the FPC assist mode when operated in accordance with existing procedures.
The staff also discussed the load on the emergency diesel generators that would be imposed by use of the RHR FPC assist mode following a loss of off-site power.
Hessrs.
Lochbaum and Prevatte indicated that the staff presentation appeared to resolve their concern on this specific issue.
Hessrs.
Lochbaum and Prevatte had several comments on the use of the RHR FPC mode under various scenarios.
They expressed concern that use of a potentially contaminated accident unit RHR train in the FPC assist mode was not specifically evaluated or cautioned in the flow chart.
Additionally, they expressed concern that failure to control the availability of the spent fuel pool cooling system and RHR system under various plant operational conditions might undermine the defense in-depth concept represented in the flow chart.
The staff noted these
- concerns, and the staff will consider including the concerns related to the availability of fuel pool cooling in a generic review of spent fuel storage pool issues.
The staff provided some information regarding the impact of spent fuel pool heat loads on evaluations of the performance of the ultimate heat sink.
The staff indicated that some additional review of this issue was necessary and would be a topic for future meetings.
The staff briefed Hessrs.
Lochbaum and Prevatte on the content of the staff's draft generic action plan on spent fuel pool facilities.
The staff indicated that four to six plant audits would be performed starting in early 1995.
The staff stated that a temporary inspection instruction for audits would likely be available in the January 1995 time frame.
The staff stated that a copy of the final generic action plan would be provided to Hessrs.
Lochbaum and Prevatte.
The final phase of the meeting centered on the schedule of future meetings and future staff activities.
The staff stated that a draft safety evaluation for the Susquehanna plant-specific issues was expected in early October 1994 and that the draft would be provided to Hessrs.
Lochbaum and Prevatte.
Hessrs.
Lochbaum and Prevatte stated that they were interested in additional meetings with the staff.
Hr. Prevatte provided several lists of topics for future meetings.
These lists are included as Enclosure 2.
Hr. Lochbaum had provided a list of technical issues (Enclosure
- 3) prior to the meeting.
Hr. Prevatte stated that he believed that a discussion and understanding of related licensing issues would be beneficial since it would provide a foundation for future technical discussions.
The staff agreed to pursue additional meetings in the early October 1994 time frame and to contact Hessrs.
Lochbaum and
Prevatte in the near future to establish the exact agenda for the next meeting.
Enclosure 4 is a list of those who attended the September 6,
1994 meeting.
Docket Nos, 50-387/388
Enclosures:
1, Draft Flow Charts "Loss Of Spent Fuel Pool Cooling Event Recovery Flowchart" 2.
Notes on Major Issues Requiring Resolution Provided By Hr. Prevatte 3.
List of Cross Referenced Issues Provided By Hr, Lochbaum 4.
List of Meeting Attendees cc w/encl:
See next page
/s/
Joseph W. Shea, Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation DISTRIBUTION w Enclosure 4:
WRussel 1 /FHiragl i'a GHol ahan RZimmerman HThadani AThadani CMcCracken SVarga GHubbard CHiller
- EJordan, D/AEOD DISTRIBUTION w all
Enclosures:
,Docket File '<
HVirgilio PUBLIC SJones PDI-2 Reading LPrividy EWenzinger, RGN-I RMatakas JShea CPoslusny ACRS(10)
EKelly, RI HO'Brien WDean OGC OFFICE P
NAME t
'e PDI 2
PH OSSA/DD HVi ilio HThadani DATE
/
94 OFFICIAL E
0 COPY
/
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ply II Prevatte in the near future to establish the exact agenda for the next meeting.
Enclosure 4 is a list of those who attended the September 6,
1994 meeting.
Docket Nos.
50-387/388
Enclosures:
1.
Draft Flow Charts "Loss Of Spent Fuel Pool Cooling Event Recovery Flowchart" 2.
Notes on Hajor Issues Requiring Resolution Provided By Hr. Prevatte 3.
List of Cross Referenced Issues Provided By Hr.
Lochbaum 4,
List of Heeting Attendees cc w/encl:
See next page J
se W. Shea, Project Hanager P oject Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
'4 Pennsylvania Power
& Light Company Susquehanna Steam Electric Station, Units I
& 2 CC:
Jay Silberg, Esq.
- Shaw, Pittman, Potts
& Trowbridge 2300 N Street N,W.
Washington, D.C.
20037 Bryan A. Snapp, Esq.
Assistant Corporate Counsel Pennsylvania Power
& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Hr. J.
H. Kenny Licensing Group Supervisor Pennsylvania Power
& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Hr. Scott Barber Senior Resident Inspector U.
S. Nuclear Regulatory Commission P,O.
Box 35 Berwick, Pennsylvania 18603-0035 Hr. William P. Dornsife, Director Bureau of Radiation Protection Pennsylvania Department of Environmental Resources Commonwealth of Pennsylvania P.
O.
Box 8469 Harrisburg, Pennsylvania
'17105-8469 Hr. Jesse C. Tilton, III Allegheny Elec.
Cooperative, Inc.
212 Locust Street P.O.
Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administrator, Region I
U,S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hr, Harold G. Stanley Superintendent of Plant Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 Nr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural, Route 1,
Box 1797 Berwick, Pennsylvania 18603 George T. Jones Manager-Engineering Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Hr. Robert G.
Byram Senior Vice President-Nuclear Pennsylvania Power
& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Nr. David A. Lochbaum 80 Tuttle Road
- Watchung, New Jersey 07060 Hr. Donald C. Prevatte 7924 Woodsbluff Run Fogelsville, Pennsylvania 18051
MOOR LICENSING ISSUES RE UIRING RESOLUTION 9 6 94 1.0 Generic Definition ofLicensing Basis.
1,1 NUREG-1412, "Foundation for the Adequacy of the Licensing Bases".
1.2 Numerous NRC internal documents.
2.0 Specific Licensing Bases for Susquehanna.
2.1
Subject:
Requirement for Providing Cooling to Fuel Pool Under Postaccident Conditions.
Licensing Requirements:
a 10CFR50, Appendix A, Criterion 61, "Fuel Storage and Handling and Radioactivity Control".
b.
UHS water inventory analysis reported in FSAR for LOCA includes make to two boiling spent fuel pools.
2.2
Subject:
Specific documentation that loss of normal fuel pool cooling concurrent with LOCA is a part of the licensing basis.
Licensing Documents:
a.
UHS analysis of spray pond inventory requirements for LOCA reported in FSAR includes makeup to two boiling spent fuel pools.
b.
FSAR Chapter 6 tables ofelectrical loads carried by the diesel generators post-LOCA do not include spent fuel pool cooling pumps or service water pumps.
c.
PP&L Itr. to NRC, 5/24/93, that spent fuel pool cooling system is automatically load shed post-LOCA.
d.
PP&L Itr. to NRC, 1/17/89, that procedures were revised to manually shed 1E loads in reactor building post-LOCA, which includes fuel pool cooling system.
2.3
Subject:
Design Requirements for Fuel Pool Level and Temperature Monitoring Instrumentation post-LOCA.
Licensing Requirements:
10CFR50, Appendix A, Criterion 61, "MonitoringFuel and Waste Storage".
b.
Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear power Plants to Assess Plant and Environs Conditions During and Following an Accident" ENCLOSURE 2
2.4
Subject:
Requirement for Fuel Pool Cooling As Well as Core Cooling and Containment SSC to be Capable ofWithstanding Accident Conditions.
Licensing Requirements:
10CFR60, Appendix A, Criterion 4, "Environmental and Missile Design Bases".
10CFR60.49, "Environment Qualification ofElectrical Equipment Important to Safety for Nuclear Power Plants" 2.6
Subject:
Design Requirements for Determiniiig Postaccident Operator Radiation Exposure Conditions Inside Reactor Building.
Licensing Requirements:
a Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for BoilingWater Reactors".
NUREG 0737, Section lI.B.2, "Design Revie~ ofPlant Shielding and Environmental Qualification ofEquipment for Space/Systems Which May Be Used in Postaccident Operations" and Section IILD.l.l,"Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling-Water Reactors".
C.
10CFR60, Appendix A, Criterion 19, "Control Room" as amended by NUREG 0737,Section II.B.2.
d.
Special Inspection Report Nos. 60-387/84-10, 60-388/84-11 dated 4/23/84 of Susquehanna Plant and followup letters from PP&L to NRC nos. PLA-2133 dated 3/21/84 and PLA-2219 dated 8/4/84 and from NRC to PP&L dated 9/19/94.
e.
10CFR50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors" 2.6
Subject:
Design Requirements Regarding the Failure of Safety-Related Equipment; are any failures of safety-related equipment allowed by design?
2.7
Subject:
Design Requirements Regarding Loss-of-Offsite-Power (LOOP).
Licensing Requirements:
o 10CFR50, Appendix A, Criterion 17, Electric Power Systems".
o Regulatory Guide 1.137, "Fuel-Oil Systems for Standby Diesel Generators".
3.0 Regulatory Basis for. AllowingSubstitution ofPRA for Compliance with Regulations.
~ ~
MAJOR TECHNICALISSUES RE UIRING RESOLUTION 9 B 94
1.0 Issue
Normal fuel pool cooling is lost post-LOCA and cannot be restored.
NRC MatrixNumbers:
??????
Concerns:
1.
Design automatically sheds fuel pool cooling system on LOCA signal.
2.
Procedure required manual deenergization of system aRer 24-hours.
3.
System not environmentally qualified; willfail due to LOCA environment.
4.
Required supporting system, normal service water system, not available.
6.
System not single failure proof. Single random failure can incapacitate.
6.
System not lE powered.
2.0 Issue
RHR system and other supporting systems, within their design bases, are not capable of providing adequate cooling to the spent fuel pool in the Fuel Pool Cooling Mode.
NRC MatrixNumbers:
3 Related Material Not Yet Reviewed by Lochbaum/Prevatte:
PP&L Calc. M-RHR-039, Rev 0, 6/13/93.
Concerns:
Operators do not have access to the reactor building to align valves to operate system in this made due to radiation exposure levels.
RHR system is not single failure proof in this mode.
3.
RHR pumps have insuHicient net positive suction head (NPSH) in this mode to meet flow requirements.
b.
C.
d.
Level in fuel pool is too low; can't be raised without entering building.
Pool temperature is too high.
PP&L calculation confirmed system willnot work in this mode.
. PP&L preoperational test confirmed system would not operate in this mode under significantly less'emanding conditions than the design conditions.
Capability of the ultimate heat sink spray pond with this additional heat load is unanalyzed and doubtful.
This mode bypasses the primary containment and puts accident water into the fuel pool, thereby worsening building access, offsite radiation exposure, and control room operator exposure.
~ ~
0
3.0 Issue
Fuel pool level and temperature monitoring instrumentation are not adequate for accident conditions.
NRC MatrixNumbers:
??????????
Concerns:
1.
Existing and proposed new instrumentation does not have suHicient range to cover accident conditions.
2.
Instrumentation does not provide control room readout.
3.
Instrumentation is not environmentally qualified for accident and fuel pool boil conditions.
4.
Instrumentation (not just mounting) is not seismically qualiTied.
6.
Instrumentation is not 1E powered and qualiTied.
4.0 Issue
Standby gas treatment system (SGTS) is not qualiTied for conditions created by a boiling spent fuel pool.
NRC MatrixNumbers:
??????????
Concerns:
1.
SGTS fan motors and other equipment in the fan rooms are not environmentally qualified for the temperature conditions associated with a boiling fuel pooL 2.
SGTS cannot handle the volume of condensate which willenter the Qlter units.
3.
The SGTS ductwork is not designed to handle condensing vapor from the boiling fuel pool with regard to blockage, structural integrity, and leakage onto other safety-related equipment.
4.
Critical SGTS control instrumentation is not environmentally qualified for the conditions associated with a boiling fuel pool.
6.
Filter train heaters willcut out on high temperature.
6.0 Issue
PP&L's current procedures require shutting down ofreactor building recirculation fans to prevent carryover ofrefueling floor environment to the rest of the reactor building, this creates numerous unanalyzed conditions.
NRC Matrix Numbers:
????????
Concerns:
2.
Temperatures in numerous areas in rest of reactor building willexceed EQ limits.
(Admitted by PP&L, but speciTic equipment has not been identified.)
Radiation levels in numerous areas in rest of reactor building willexceed EQ limits.
3.
Analysis of offsite and control room radiation exposures are based on recirculation system being in operation.
4.
Pressure due to boiling willcarryover refueling floor conditions to rest ofreactor building even without recirculation fans.
6.0 Issue
As-yet-not-fully-quantified safety-related equipment in the reactor building fails due to flooding effects of the boiling spent fuel pool.
NRC MatrixNumbers:
??????
Concerns:
2.
3, RHR pump??? fails. Not allowed by regulations.
Core spray pump??? fails. Not allowed by regulations.
Analysis that shows other pumps don't fail has not been reviewed and is suspect.
4, Safety-related coolers ability to function in high humidity conditions has not been evaluated.
When they fail due to not being designed for latent heat cooling conditions, the associated safety-related equipment fails. Includes all safety-related pumps plus emergency switchgear and load center rom coolers which power almost all safety-related equipment in the building.
No analysis has been performed of safety-related equipment threatened by water as it makes it way from the refueling floor to the basement.
KNOWN CHANGES MADETO.DATE BY PP8cL AS A RESULT OF 10CFR21 REPORT, Fire damper fusible links in reactor building HVAC/SGTS systems have been changes from 165'F to 285'F.
2.
3, Two units'uel pool have been crosstied for normal operation.
Design modifications have been made to the fuel pool instrumentation (extent unknown).
4, RHR and service, water valves required for fuel pool cooling have been placed in inservice testing program.
5.
Emergency procedures changed to turn offreactor building recirculation fans.
E UIPMENTFAILURES KNOWNCONCEDED AS OF 9 94 1.
SGTS fails due to reactor building HVAC/SGTS fire dampers closing on high temperature.
Resolution - Fusible links replaced.
2.
SGTS fails due to water accumulation in intake ductwork. Resolution - unknown.
3.
SGTS fails due to exceeding qualification temperature of fan motors.
Resolution - unknown.
4.
SGTS fails due to exceeding EQ conditions ofcontrol instrumentation.
Resolution - unknown.
5.
RHR pump??? fails due to flooding. Resolution - unknown.
6.
Core spray pumps A and C fail due to flooding. Resolution - unknown.
7.
"Some" unspecified safety-related equipment due to exceeding EQ qualification conditions when reactor building recirculation fans are secured.
Resolution - unknown.
8.
RHR pump C room cooler would faiL Resolution - unknown.
9.
Core spray pump C room cooler would faiL Resolution - unknown.
10.
RHR pump C would experience failure due to MCC EQ temperature being exceeded.
Resolution - unknown.
11.
Core spray pump C would experience failure due to MCC Eq temperature being exceeded.
Resolution - unknown,
AUG-38-1994 14: 16, FRY CMS,EP David A. Lochbaum
, Nuclear Engineer 9331S842182 P. 81 80 Tattle Road Watchung,I070M 80% 7$43$Tl August 30, IQQ4 2:16 pm FAX MESSAGE Joe S~
e FROM!
Dave Lochbrunl NUMBER OF PACKERS gnctndtng sheet):~
Thanks for sending along the matrix. I reviewed every letter we have submitted to the NRC (excluding those to Ol and IG) and developed a cross reference of open/unresolved/new concerns to matrix items. h copy of this cross reference is provided for your information. This cross reference is not intended to represent the complete list of open/unresolved/ncw items. It is intcadcd solely to help mc prcparc for next week's mccdng.
Thc codes in the 4>> columa were simply used to sort the table entries.
Note there arc three new'ssues on IMs listing.
There is slot of duplication on this listing - I didn't attempt to consolidate items.
Many of the items oa this listing involve questions concerning the usc ofnonMcty related equlpmcat.
The RCIC System is a nonA Category I system at most, ifnot all, BWRs.
No credit fs taken for RCIC System operation in most, ifaot sll, plant transient <<nd student <<n<<lyscs.
Yet dm RCIC Sys(uu ls still 'cwuttullal'u the Topical Specifications and FSAR.
Its 'importaa< to safety'unction is tecognhed and formaIized. ~
actual cbLssiflcation of the RCIC System is of minor conuquence because it is designed, maintained and tested to cnsurc a high degree of rcllabillty. %ben the RCIC System ts unavanable, the ptant Ie iu an LCO of ftnite duration.
These measures provide, rcasonablc assurance that the RCIC System willbe avaQable, evea though pl<<ut safety <<u<<lycee sbvw aceptable results without RCIC.
The MES FPCCS <<nd RHR FPC Assist modes <<rc not 'RCIC-]i've.'uring <<n vut<<gc when the cathe mtu is offloaded into the fuel pool and tb: cavity gates installed,~ loops of RHR~ the entire FPCCS can be Inoperablc without entering an Lco, we Fpccs ann RHR t'pU Assist moaes are not aaugnett, mammnea and tested to thc degree rcqubed of RCIC. And also unlike RCIC, ifthe FPCCS and RHR FPC Assist mode are unavilable, plant safety analyses do not show acceptable results, Scc you next week in Allentown.
El<CLOSURE 3
Czoss Reference
- Letters to NRC 8, NRC Xatr&
Source 93/10/Ol page 7
93/11/07 Item 10 94/06/17 Item 2.1 94/06/17 Iten 1.1 94/01/24 Item A-1 94/05/25 Item 2.1 94/06/17 Item 2.2 NEif 94/05/25 Item 4.1 94/02/21 Itez D-1 94j02/21 Itea D-3 94j02/21 Itea D-4 93j10/Ql Iree 93/10/01 Iree RHR =FC manual valves removec froa ISI jIST'eliance on non-safety related systems is uncertain Loss cf FPC off-normal grocecure does not address feel pools being cross-tied RHR "-FC procedure canrot be followed verbat a in LOCA or LOOP event due to prerequisite Reliance on non-safety relat& systems is uncertain RHR PPC Valves in ZSI/1ST Program Loss of FPC off-noxmal procedure provides inadequate monitoring guidan"e Admix controls on RHR E. FPCS in refueling are inadequate (cozpare with RCIC controls)
Skimaer surge tank level improperly aralyzed with adverse impact or hPSH Non-conservative RB heat loaB analysis (SEA-EE-550 - assuned LOCA/L3OP case)
M-PPC-015 not avai:able for review Shut=ing Bown RB HVAC recirc fans is improperly analyzed Hydrocynawic ef=ects of SFP boiling are unanalyzeB Is accident @nit's reactor building accessibl
?
Operator Boses =or ESM aanual actions Matrix Ztaas 20 2, 20 27, 54 21, 47, 54 2,
$ 0 20 4, 57, 37 41, 54, 55 3r 4i c7 30, 52 19, 24, 52 C
8, 19, 24, 46 C
36, 39 40, 45 34, 38, 42 D
August 30, 1994 Page 1 of 5
Cross Reference
- Letters to NRC a NRC matrix Source 93!10/01 Tree 94!01/24 Iten A-3 94!06/17 Ktea 4.3.b 94!03/07 Item 2 94!05/25 Iten 4.3 94!04/21 page 1
94!04/21 page 1
94/04/21 page 2
93/10/Ol 1'ree 93/10/Ol Tree 93/10/01 Tree 94/01/24 Iten A-2 93/08/13 page 6
94/03/21 Iten 9
94i03/21 Iten 2 94/03/21 Iten 1 94i03/21 Iten 3
Issue Is non-accident unit's reactcr buildinc accessible?
Use of measu=ed containment leakage rate is improper Non-conservative assessments of operator dose Use of measured containment leakage rate is improper Airborne radiation dose est be corsidered Pressure adjusted leakage rate 's improper Use.of measured containment leakace rate is improper Containme~t airborne leakage rate is iaproper Are accident un=t's FECCS/SW= components operable?
Mill non-accident unit's FPCB hancle combined SPP heat loads?
Is accidest unit's FPCCS/SNS pipirc intact?
Reliance on FP S conflicts with RS load shed Licensing basis for SPP boiling GDC 63 BFP monitorin Licensing basis f'r LCOF no FPCS Licensing basis for EHN/RKS - SFP makeup Licensing basis for DBA IOCA -
FPCS load shed matrix Items 36, 39 40, 45 35, 45, 56 D
31, "3, 36 37, "9,'1 43, 44 35, 45, 56 t
36, 39, 45 48 35, 45, 56 E
35, 45, 56 35, 45, 48 K
13, 15, 20 27 27, 29 2,
15 August 30, 1994 Page 2 of 5
Cross Reference
- Let"ers to MRC & MRC Matrix Source 94!03/21 Item 4 94!03/21 Iten 11 94/03/21 Ite~ 8 93/10/01 Tree 93/10/01 Tree 93/10/01 Free 93/10/01, Free 93/10/01 1'ree 93/08/13 Iten 3-B 94 /01/24 Iten D-6 94/04/21 page 2
94i05/25 Itea 3.3 93 /10/01 Tree 93/10/01 Tree 93/10/01 Tree Licensing basis for DPA'XOCA - eanual load shed cf non-lass 1H loads GDC 4
'important to safety'esign GDC 61
"...postulated accident conditions.'s non-Class 18 power available for non-accichnt unit?
Are switchgear room coolers 9esigned for latent heat from SFP boiling?
Is flcoding fma SFP boiling controlled?
Are
"=CCS pump room coolers designed for latent heat from SPP boiling?
Is non-Class 1E power available for accident unit?
LOCA/LOOP affects botl reactor buildirgs very high" probability of SFP boiling in LOOP event per PP&I PEA LOCA/LOOP affects both reactor buildings Inadecuate EQ with SFP boiling Capability to restore cooling after SPP boilirg is unanalyzed
& uncertain Is an RHR loop available on non-accident unit?
Is an RHR loop ava=lable on accident unit?
Is adequate NPSH available for RHR FPC on non-accident unit?
Matrix Items 2,
29 2,
20 12, 29 19, 24, S1 14,
'23 19, 24, 51 M
12, 29 27, 31 2, 4, 12 22'7 27, 31 21 3,
4, 3C August 30, 199~
Page 3 of 5
Cross Reference
- Letters to NRC 6 HRC Matrix Source 93/10/01 Tree 93/10/01 Iree 93/10/01 Iree 93/10/Ol Tree 93/12/06 Item 2-C 93/10/01 page 7
93i12/06 IteN 3 93/10/01 page 7
93/12/06 Item 2-D 93j10/01 page 7
93/12/06 Item 2-H 93/10/01 page 7
93j10/01 page 7
94/05/25 Item 2.2 93/10/01 T ee Issue Are the RHR FPC conponerts on non-accident unit operabl
'?
Will acci8ent unit's RHR FPC transport accident source terms to SFP":
Are the RHR FPC cocponerts on accident unit operable?
Xs adequate NPSH available for RHR FPC on accident unit?
NPSH for RHR FPC Assist mode is uncertain RHR RPC manual act'ons prohibited by ¹gh radiation levels in reactor building BDG capacity fo" RHR PPC Assist is unknown RHR FPC after DBA LOCA would transport source terms to SFF uranalyzed Doses for RHR FPC Resist manual actions are non-"onservative RHR FPC NPSH at required flow rate is uncertain Availabilityof RHR FPC Assist post-LOCA is uncertain RHR FPC is not single failure proof SFP temperature when RHR FPC is in=tiated is uncertain RHR FPC Mode: Non-eafety and no= tested I
&"TS de g
d f teape a
a s
from SFP boiling?
Matrix Eteaw 20 47 20 3, '4, 30 3, 4,
$0, 44 44 d2 21, 47 20 4,
30 3,
20, 22 1
24
'6 Auguat 30,199-'age 4 of 5
Cross Reference
- Let"ers to MRC f RR Matric Source 93!10/01 Tree 94!05/25 Etea 3.1 94!05/25 Iten 3.6 94!05/25 Iten 3.2 93!11/07 Iten 9 93/10/01. Free 93/10/01
@age 7 Are SGTS components EQ'ed for SPP boiling conditions 2 SOTS >aust remain operable during design basis events
- therefore, SFP cannot boil SGZS ability to maintain 0.25" vacuum vith SFP bciling unverified
& uncertain SGTS component temperature cy;ylification UHS capacity for DBL LOCA is uncertain Ie RHRSW/3HS capacity available?
UHS capacity for DBL LOCA is uncertain Ifatrix Xtems 16, 24 50 16, 19 10, 11 10, 11 10, 11 94/06/17 Iten 4.3.a UHS capacity for DBA LOCA is uncertain 10, ~11 August 30, 1994 Page 5 of 5
LIST OF ATTENDEES MEETING BETWEEN MESSRS.
LOCHBAUM AND PREVATTE AND THE NRC SEPTEMBER 6
1994 NAME 1.
J.
Shea 2.
D.
Lochbaum 3.
D. Prevatte
'4.
M. Thadnai 5.
M. Virgilio 6.
S.
Jones 7.
E. Kelly 8.
L. Prividy 9.
J.
Kenny 10.
H. Woodeshick 11.
D.
Ney 12.
S. Maingi 13.
M. Pflieger 14.
D. Pearson ORGANIZATION NRR/PDI-2 Self Self NRR/PD I-2 NRR/DSSA NRR/SPLB NRC/RGN-I NRC/RGN-I PP&L PPKL PA DER/BRP PA DER/BRP The Mornin Call, Allentown, PA Self ENCLOSURE 4
0
).
I') iA C
LOSS OF SPENT FUEL POOL COGl ING EVENT RECOVERY FLOWCHART LOCA EVENTS POTENTIAL IMPAC, Si SEISMIC EVENT POTENTIAL IMPACTS:
SYSTEMS INTEGRITY POWER SUPPLY INTEGRITY SUPPORT SYSTEM AVAILABILITY OPERATOR ACCESS ENVIRONMENT (TEMP/HUM/RAO)
SYSTEM ANO PQ/ER SUPPLY INTEGRITY LOOP EVENT IMPACTS:
POWER SUPPLY FQR SYSTEMS WHAT IS TIME TO BOIL' NO IS g
W?J-
'FPC OPERATING YES ANO PROVIDE COOLING7 OK STOP IS POWER AVAILABLE'? (SFPC 5
~ IS IN-PLANT POWER SUPPLY ENTACT7
-SEISMIC LOADS
REFERENCE:
-LOCA LOADS
REFERENCE:
~IS POWER SUPPLY TRIPPED2
REFERENCE:
~IS OFFSITE POWER AVAILABLE/RESTORABLE(TIME)'?
REFERENCE:
ARE SUPPORT SYSTEMS AVAILABLE'
~ IS A MAKE-UP SOURCE OTHER THAN ESW AVAILABLE7 IF NO, USE ESW.
REFERENCE:
~ IS TANK/POOL LEVEL INSTR.
AVAILABLE7
REFERENCE:
~ IS SERVICE WATER AND IT'S SUPPORT SYSTEMS AVAILABLE7
REFERENCE:
YES ASSESS ENVIRONMENTAL CONDITIONS ON EQUIPMENT
~CAN EQUIP.
HANDLE TEMPERATURES7
REFERENCE:
~ CAN EQUIP.
HANDLE HUMIDITY7
REFERENCE:
~CAN EQUIP.
HANDLE RADIATION7
REFERENCE:
YES YES YES CAN OPERATORS ACCESS
- SFPC, ESW, 5
"SUPPORT SYSTEMS"'?
NO SFPC 5
SERVICE WATER SYSTEM INTEGRITY
~ CAN PIPING WITHSTAND SEISMIC LOADING'?
REFERENCE:
~CAN PIPING WITHSTAND LOCA LOADING7
REFERENCE:
~ CAN EQUIPMENT WITHSTAND SEISMIC LOADING7
REFERENCE:
~ CAN EQUIPMENT WITHSTAND LOCA LOADING2
REFERENCE:
NO
~ ARE RAD LEVELS IN RX BLDG ACCEPTABLE'
REFERENCE:
~ ARE RAD LEVELS OUTSIDE RX BLDG ACCEPTABLE7
REFERENCE:
~ ARE TEMPERATURE LEVELS ACCEPTABLE'
REFERENCE:
NO YES NQ INITIATE MAKE-UP TO SFP (S)
UTILIZING,AT LEAST, ONE QF THE FOLLOWING SYSTEMS:
R.F.
WATER.
NQ CAN OTHER UNIT PROVIDE NORMAL SFPC PRIOR TQ BOILING' YES YES STOP ARE BOTH SFP S
CROSSTIED'?
NO YES INITIATE KE-UP TQ SFP(S)
UTILIZTNC.-iiT, LEAST,QNE OF THE FOLLOWING SYSTEMS:
WATER.
CAN CASK PIT GATE'S BE'REMQVED7
~IS POWER AVAILABLE/RESTORABLE7 REFERENCE.
ADO RAO LEVELS PERMIT CRANE ACCESS7
REFERENCE:
ADO TEMPERATURES PERMIT CRANE ACCESS7 REFERENCE.
~ IS CRANE STRUCTURALLY INTACT' REFERENCE.
NQ STOP QK YES RES; SF( i COOLING CAN COOLING BE RESTORED PRIOR TO SFP BOILING7 NO CAN RHR FPC MODE BE USED'
~ ARE BOTH LOOPS OF RHR "COOLING" FUNCTIQNAL7
REFERENCE:
SCAN OPERATORS ACCESS REQUIRED VALVES7
REFERENCE:
~ IS SFP
(
SKIMMER SURGE TANK LEVEL INDICATION FUNCTIONAL 6 ACCESSIBLE'
REFERENCE:
~ ARE BOTH LOOPS QF RHR "COOLING" FUNCTIONAL7 REFElcENCEi SCAN OPERATORS ACCESS REQUIRED VALVES7
REFERENCE:
~ IS SFP
& SKIMMER SURGE TANK LEVEL INDICATION FUNCTIONAL 5
ACCESSIBLE'
REFERENCE:
YES NO NO NO ARE BOTH SFP'S CROSSTIED7 YES CAN OTHER UNIT PROVIDE COOLING PRIOR TO SFP '
BOILING2 NQ IF QNE LOOP IS FUNCTIONAL, ACCESSIBLE,AND LEVEL MONITORING IS AVAILABLE.
CONSIDER SWAPPING BETWEEN SFP COOLING AND RX DECAY HEAT REiMQVAL,AND IMPLEMENT HVAC ACTIONS.
THEN YES CAN COOLING BE RESTORED PRIOR TO SFP BOILING' YES STOP OK YES CAN DOLING BE RESTORED PRIOR TO SFP BOILING'?
IF QNE LOOP IS FUNCTIONAL, ACCESSIBLE.
AND LEVEL MONITORING IS AVAILABLE.
CONSIDER SWAPPING BETWEEN SFP COOLING AND RX DECAY HEAT REMOVAL.
AND IMPLEMENT HVAC ACTIONS.
YES STOP STOP OTHERWISE NQ STOP USE OTHER UNIT '
RHR SYSTEM PROVIDE SFP COOLING YES CAN OTHER UNIT PROVIDE COOLING PRIOR TQ BOILING7 NO YES IMPLEMENT HVAC ACTIONS.
LOSS QF SPENT FLIEL POOL t:00[.jNG EVENT RECOVERY FLQWr HART S
S IMPACTSi SYSTEMS INTEGRITY POWER SUPPLY INTEGRITY IMPACTS'UPPORT SYSTEM AVAILABILITY OPERATOR ACCESc ENVIRONMENT (TEMP/HUM/RA0)
SYSTEM AND POWER SLIPPLY INTEGRITY LOOP E
IMPACTS; POWER SUPPLY FOR SYSTEMS IMPLEMENT ON-135/235-001 "LOSS QF SFP COOLING" AND CHECK THE FOLLOWING NQ
.(S S( PC OPERATING AND QQI ING OK STOP IS POWER AVAILABLE'? (SFPC f.
~ IS IN-PLANT POWER SUPPLY ENTACT2
-SEISMIC LOADS QNO C] YES, BASIS i
-LOCA LOADS 0 NO CI YES. BASIS i
~ IS POWER SUPPLY TRIPPED2 0 NQ Cl YES, BASIS' IS QFFSITE POWER AVAILABLE/RESTORABLE(TIME)2 0 NO 0YES.BASIS'RE SUPPORT SYSTEMS AVAILABLE2
~ IS A MAKE-UP SOURCE AVAILABLE2 C] NO 0 YES, BASIS
~ IS TANK/POOL LEVEL INSTR.
AVAILABLE2 0 NO 0 YES,BASIS
~ IS SERVICE WATER AND IT'S SUPPORT SYSTEMS AVAILABLE2 CI NQ 0 YES,BASISi YES ASSESS ENVIRONMENTAL CONDITIONS SCAN EQUIP.
HANDLE TEMPERATURES'?
CINQ 0 YES,BASIS
~ CAN EQUIP.
HANDLE HUMIDITY2 CI NQ 0 YES, BASIS
~ CAN EQUIP.
HANDLE RADIATION2 0 NO 0 YES, BASIS YES YES YES CAN OPERATORS ACCESS SFPC II "SUPPORT SYSTEMS"'?
NO SFPC I)
SERVICE WATER SYSTEM INTEGRITY
~ CAN PIPING WITHSTAND SEISMIC LOADING2 CINQ 0 YES,BASISi
~ CAN PIPING WITHSTAND LOCA LOADING2 ClNO CI YES,BASIS' CAN EQUIPMENT WITHSTAND SEISMIC LQADING2 2 NQ U YES,BASIS
~ CAN EQUIPMENT WITHSTAND LOCA LOADING2 0 NQ CJ YES, BASIS NQ NQ
~ ARE RAD LEVELS IN RX BLDG ACCEPTABLE2 0 NQ 0 YES,BASIS
~ ARE RAD LEVELS OUTSIDE RX BLDG ACCEPTABLE2 0 NQ 0 YES,BASIS
~ ARE TEMPERATURE LEVELS ACCEPTABLE2 KINO 0 YES,BASIS YES NO RESTORE SFPC COOLING NQ CAN OTHER L(NIT PROVIDE NORMAL SFPC PRIOR TO BOILING2 YES YES OK STOP ARE BOTH SFP'S CRQSSTIED2 NQ YES CAN CASK PIT GATES BE REMOVED'
~ IS POWER AVAILABLE/RESTORABLE'?
0 NO 0 YES, BASIS (
ADO RAD LEVELS PERMIT CRANE ACCESS2 ClNQ 0 YES,BASIS~
ADO TEMPERATURES PERMIT CRANE ACCESS2 CINO 0 YES,BASIS(.
~ IS CRANE STRUCTURALLY INTACT2 CI NO CI YES,BASIS:
CAN RHR FPC MODE BE USED2 STOP IMPLEMENT HVAC ACTIONS AND OPEN'GTS, DUCT l
DRAINS 'N EP-PS-102 YES NQ CAN COOLING BE RESTORED PRIOR TQ SFP BOILING2 NO ARE BOTH SFP'S CRQSSTIED'?
YES CAN RHR FPD MODE BE LISED'?
~ ARE BOTH LOOPS OF RHR "COOLING" FUNCTIONAL2 GNO CI YES,BASIS SCAN OPERATORS ACCESS REQUIRED VALVES2 ClNO ClYES,BASIS
~ IS SFP f
SKIMMER SURGE TANK LEVEL INDICATION FLINCTIONAL 6 ACCESSIBLE2 CINO 0 YES,BASIS
~ ARE BOTH LOOPS OF RHR "COOLING" FUNCTIQNAL2 0 NO 0 YES, BASIS SCAN OPERATORS ACCESS REQUIRED VALVES2 0 NQ 0 YES, BASIS
~ IS SFP L SKIMMER SURGE TANK LEVEL INDICATION FUNCTIONAL f ACCESSIBLE2 0 NQ 0 YES, BASIS YES NQ NO CAN OTHER LINIT PROVIDE COOLING PRIOR TQ SFP '
BOILING2 YES QK r
IF ONE LDOP IS FUNCTIONAL, CFSSIBLE.AND LEVEL MQNI1ORING IS AVAILABLE.
QnklcIOER cWAPPING BETWEEN SFP COOLING AND RX DECAY HEAT REMOVAL,AND IMPLEMENT HVAC/SGTS ACTIONS IN EP-PS-102 THEN STOP OTHERWISE YES CAN CQQI IN((, RF RESTORED PRIOR TO SFP BQILING2 NO CAN OTHER LINIT PROVIDE COOLING PRIOR TO BQILING2 Y
ST YES YES CAN CQCLING BE REBTCRED PRIOR TQ SFP BnILING2 INQ IF QNE LOOP IS FLINCTIONAL, ACCESSIBLE,AND LEVEL
. MONITORING IS AVAILABLE.
CONSIDER SWAPPING BETWEEN SFP COOLING AND RX DECAY HEAT REMQVA(.AND IMPLEMENT HVAC ACTIOI IN EP-PS-I02,THEN OTHERWISE QK STOP STOP CAN OTHER UNIT '
RHR SYSTEM PROVIDE SFP CQQLING2 YES NQ Il"IPLEMENT HVAC ACTIONS AND OPEN SGTS DUCT DRAINS IN EP-PS-102 SPLLNG OF QNE OR BOTH SFP'S MAY OCCUR CONFIRM CClQLING~
BE RESTORED PRIQ~E ONSET OF BOILING IN ONE QR BOTH SFP'S.
INITIATE MAKE-UP TO SFP(S)
IN ACCORDANCE WITH QN-135/235-001 UTILIZING,AT LEAST, ONE OF THE FOLLOWING SYSTEMS'SW, CONDENSATE,DEMIN WATER,RHRSW, FIRE PROTECTION, R.F.
WATER.
YES WII L COQLIING BE RESTORED FOLLOWING THE ONSET QF BOIL[NG2 INITIATE MAKE-UP TO SFP(S)
IN ACCORDANCE WITH QN-135/235-001 UTILIZING,AT LEAST, ONE OF THE FOLLOWING SYSTEMS:
ESW.
CONDENSATE,DEMIN WATER,RHRSW, FIRE PROTECTION, R.F.
WATER.
SHLITDQWN REACTOR RECIRCULATION FANS AND OPEN SGTS DUCT DRAINS IN ACCORDANCE WITH EP-PS 102.TAB-BASIS IF NO SOURCE TERM IS PRESENT, CONSIDER SHUTDOWN OF SGTS AND VENT ZONE III IN ACCORDANCE WITH EP-PS-(02, TAB OTHERWISE LEAVE SGTS IN SERVICE.
RESTART NORMAL RX BLDG HVAC IF POSSIBLE IN ACCORDANCE WITH EP-PS-I02.TAB SHUTDOWN REACTOR RECIRCULATION FANS AND OPEN SGTS DUCT, DRAINS IN ACCORDANCE.'WITH EP-PS-I02,TAB BASIS IF NQ SOURCE TERM IS PRESENT.
SHUTDOWN SGTS IN ACCORDANCE WITH EP-PS-I02,TAB OTHERWISE LEAVE SGTS IN SERVICE, IF NQ SOLIRCE TERM IS PRESENT RESTART NORMAL RB HVAC IN ACCORDANCE WITH EP-PS-I02,TAB RX BLDG TEMPS IN EXCESS OF EQ LIMITS MAY OCCUR.
THIS HAS BEEN EVALUATED AS OK.BASIS'EACTOR BUILDING EQ TEMPS FOR SOME ECCS EQUIPMENT~ BE EXCEEDEO,BUT EQUIPMERT CAN OPERATE AT HIGHER TEMP QR FAILURE WILL NQT IMPACT LONG TERM CORE COOLING BASIS WIRKNG IS NQT RESTORED WITHIN 48 HOLIRS QF THE ONSET QF BOILING THE "A" LOOP QF CORE SPRAY MAY BE LOST DUE TO FLOODING BASIS I SUFFICIENT ECCS WILL EXIST TO PROVIDE FQR REACTOR/CONTAINMENT DECAY HEAT REMOVAL BASIS.
TtitJ~NG IS NOT RESTORED WITHIN
(
I SFP)
(2 SFP'S)
DF THE~SET QF KITLING,SGTS WILL BE DISABLED DUE TO WATER IN QUOTING/PLENLIM UNLESS SGTS OLIOS DRAINS ARE OPENED BASIS BASIS i "A" LOOP OF CORE SPRAY WILL BE LOST AT 48-72 HOURS AFTER ONSET OF BOILING BASIS l SGTS WILL BE LOST
[
I SFP)
(2 SFP'S)
HOURS AFTER THE ONSET QF BOILING UNLESS SGTS DUCT DRAINS ARE OPENED BASIS i BASIS i PROCEED WITH RESTORATION OF COOLING TQ SFP(S)
OK STOP PROVIDE SFP MAKE-UP FOR DURATION QF EVENT Arn~
U~
j+
fA