ML18046B420

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Submits Addl Fee for Amends 62 & 59 to Licenses DPR-39 & DPR-48.Concern Expressed Over Criteria for Assessing Fees
ML18046B420
Person / Time
Site: Zion  File:ZionSolutions icon.png
Issue date: 01/14/1982
From: Lentine F
COMMONWEALTH EDISON CO.
To: Miller W
NRC OFFICE OF ADMINISTRATION (ADM)
References
NUDOCS 8201250352
Download: ML18046B420 (16)
v • d • e


Text

  • *. i e . UNITED STATES e

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 20, 1982 Docket No. 50-255 LS05-82-0l-053 Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 1945 WParnall Road Jackson, Michigan 49201

Dear Mr. Hoffman:

SUBJECT:

1. PALISADES DRAFT PROBABILITY RISK ASSESSMENT
2. CONSUMERS POWER CO. *DISPOSITION OF UNRESOLVED DIFFERENCES ON PALISADES Enclosed is a draft of a report entitled, 11 Summary of Risk Based Evaluation of Palisades SEP Issues" developed by the Sandia National Laboratories and Science Applications, Inc. This report is a probability risk assessment of some of the Systematic Evaluation Program (SEP) topics that have un-resolved differences. The report is under review by the NRC staff and we would appreciate your comments before this preliminary draft is revised.

The SEP topic review for Palisades has identified 28 topics with unresolved differences. These topics are listed in Enclosure 1 and a summary of the differences is in Enclosure 2. It is requested that you provide us with information on how CPCo plans to resolve these differences.

Your response to both these items is requested by February 2, 1982.

Sincerely,

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Dennis M. Crutchfield1/Chief Operating Reactors Branch No. 5 Division of Licensing Enclos*ures:

As stated cc w/enclosures:

See next page

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Mr. David P. Hoffman PALISADES Docket No. 50-255 I

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  • I M. I. Miller, Esquire U. s. Environmental Protection Isham, Lincoln &Beale Agency Suite 4200 Federal Activities Branch One First National Plaza Region V Office Chicago, Illinois 60670 ATTN: EIS COORDINATOR 230 South Dearborn Street Mr. Paul A. Perry, Secretary Chicago, Illinois 60604 Consumers Power Company 212 West Michigan Avenue Charles Bechhoefer, Esq., Cha*; rma*n Jackson, Michigan 49201 Atomic Safety and Licensing Board Panel
  • Judd L. Bacoft, Esquire U. S. Nuclear Regulatory Commission Consumers Power Company Washington, D. C. 20555 212 West Michigan Avenue Jackson, Michigan 49201 Dr. George C. Anderson Department of Oceanography Myron M. Cherry, Esquire University of Washington*

Suite 4501 Seattle, Washington 98195 One IBM Plaza Chicago, Illinois 60611 Dr. M. Stanley Livingston 1005 Calle Largo Ms. Mary P. Sinclair Santa Fe, New Mexico 87501 Great Lakes Energy Alliance 5711 Summerset Drive *

  • Resident Inspector Midland, Michigan 48640 c/o U. S. NRC Palisades Plant Kalamazoo Public Library Route 2, P. O. Box 155 315 South Rose Street Covert, Michigan 49043 Kalamazoo, Michigan* 49006 Township Supervisor Covert Township Route 1, Box 10 Van Buren County, Michigan 49043 Office of the Governor (2)

Room 1 - Capitol Building Lansing, Michigan 48913 William J. Scanlon, Esquire 2034 Pauline Boulevard Ann Arbor, Michigan 48103

~_______../ Palisades Plant ATTN: Mr. Robert Montross .*

Plant Manager Covert, Michigan 49043 i

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TOPICS WHICH DO NOT MEET CURRENT CRITERIA OR EQUIVALENT.

II-1 .A EXCLUSION AREA AUTHORITY AND CONTROL

. .I II-3.B FLOODING POTENTIAL AND PROTECTION REQUIREMENlS

  • i II-3.B.l . CAPABILITY OF OPERATING PLANT TO COPE WITH DESIGN.

~ BASIS FLOODING* CONDITIONS II-3 .C. SAFETY-RELATED WATER SUPPLY [ULTIMATE HEAT SINK (UHS.>>J Ill-1 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS ~SEISMIC AND. QUALITY) .

WIND AND TORNADO LOADINGS III-4.A TORNADO MISSILES III-5.A EFFECTS OF PIPE BREAK ON STRUCTURES,.SYSTEMS AND CammDNENTS INSIDE CONTAINMENT III-5.B PIPE BREAK OUTSIDE CONTAINMENT Ill-7 .A INSERVICE INSPECTION, INCLUDING PRESTRESSED CONCRETE !CONTAINMENT WITH EITHER GROUTED OR UNGROUTED TENDONS.

III-7.B

  • DESIGN CODES, DESIGN CRITERIA, AND LOADING COMBINATlmNS III-7 .C . DELAMINAHON OF PRESTRESSED CONCRETE CON~A~NMENT STRUi.tTURES III-8.A LOOSE PARTS MONITOR.ING AND CORE "BARREL\ VIB.*RATION PROmRAM ./

V-5 REACTOR COOLANT PRESSURE BOUNDARY (RCPB) LEAKAGE DE'Tii:TION V-10.B RHR RELIABILITY V-11.A. REQUIREMENTS *FOR ISOLATION OF HIGH AND LOW PRESSURE SWSTEMS .

VI-2.D MASS AND ENERGY RELEASE FOR POSSIBLE PIPE BREAK INSDmE CONTAINMENT Vl-3 CONTAINMENT PRESSURE AND HEAT REMOVAL CAPABILITY VI-4 CONTAINMENT ISOLATION SYSTEMS VI-6

  • CONTAINMENT LEAK TESTING VI-10.A TESTING OF REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES

. INCLL!DING QUALi FI CATIONS OF -ISOLATION DEVICES ._

VII-3 .~YSTEMS REQUIRED FOR SAFE SHUTDOWN

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I VIII-3.A STATION BATTERY CAPACITY TEST REQUIREMENTS IX-3 STATION SERVICE AND COOLING WATER SYSTEMS

  • IX-5 VENTILATION SYSTEMS XV-2 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT {PWR}

XV-12 SPECTRUM OF ROD EJECTION ACCIDENTS I

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  • TOPIC NO. TITLE J

U-1.A Exclusion Ar.ea Authority and Control Difference Sumnar,y Potential defects exist in the titles to some lands within the exclu.sion area boundary which rel ate to both surface ownership and mineral rights.

  • 10 CFR Part l00*.3(a) as implemented by SRP2.l.7 requires the licensee

.' to have the authority- to determine all activities. including exclusion or removal of personn~l and property from the area. . * . * ** .

TOPIC NO. TITLE II-3.B Flooding Potential and Protection Requirements II-3.B.1 Capability of Operating Plants to Cope With Design Basis Flooding Conditfons II-3.C . Safety-Related Water Supply [Ultimate Heat Sink (UHS)] . . . .

Difference Summary The unresolved deviation in the above inte"grated topics is that the calculated flood level du~ to a (wind driven wave) seiche. as r*equired by 10 CFR 50*(GDC 2) as implemented by Standard Review Plan 2.4.5, and Regulatory Guide 1.59 is 597.1 feet ms1. Flooding of-safety-related

  • equii:xnent would occur above 594.67 feet msl.

TOPIC NO. TITLE .,

III-1 Classification of Structures, Components*

And Syster.is (Seismic and Quality)

Difference Summary 10 CFR 50 {GDC 1) as implemented by Regulatory Guide 1.26 requires that structures systems and components important to safety'be designed, fabricated, erected and tested to quality standards conunensurate with the importance of the safety functions to be performed. Category .C joints of vessels,which would currently be classified by ASME Section III, 1977 as Class 2 or 3 but

  • built to ASME Section III; 1965 as Class C do not satisfy currentra,d1ography requirements.
  • TOPIC NO. TITLE III-2 Wind and Tornado Loadings Difference Summary

. The requirements .of lo CFR 50 (GDC. 2) .as implemented by *Regulatory Guide l_.1-17 and SRP 3~3 prescri~e structures, syst~s *. and comp~nents that should be designed to withstand the effects. of a tornado without loss of* capability to perform their safety func~ion.

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  • l The fo)Jowfog structt9s and components not to be able to withstand tornado wind loadings: -
l. Safety injection and refueling water tank
2. Supply and exhaust piping for emergency dies~l generators
3. Steel frame enclosure over the spent fuel pool TOPIC NO.
  • TITLE III-4.A Tornado Missiles Difference SuJTJnary 10 CFR 50 (GDC 2) as implemented by Regulatory Gufde 1.117 prescribes structures, systems, and components that should be designed to withstand the effects of a tornado, including tornado missiles, without loss of capability to perform* their safety function.

The following safety-related structures, systems. and components were I found to not be protected from tornado missiles:

1. Atmospher*ic relief stacks of steam relief walves Ii 2.

3.

Safety injection_ and refueJing water tank Compressed air system

4. Diesel generator cdr intake and exhaust piping TOPI'C NO. TITLE III-5.A Effects of Pipe Break on Structures. Systems and Components Inside Containment Difference Summary 10 CFR 50 (GDC 4), as implemented by Regulatory Guide l.46 and SRP 3.6.2*.

requires in part that sturctures, systems and components important to safety be appropriately protected against *dynamic effects, such as pipe*

whip and discharging fluids that may* result from equipment.failures. The effect*of pipe breaks inside containment was not a part of the original design *basis of the Palisades plant. Approximately 200 break locations exist for which the consequences of postulated pipe breaks have not been evaluated and which could adversly affect safety related systems or com--

ponents.

  • TOPIC NO. TITLE III-5.B Pipe Break Outside Containment Difference SulllTlary 10 CFR 50 (GDC 4), as implemented by SRP 3.6.l and 3.6.2. requires in part that structures, systems and_components important to safety be appropriately protected against dynamic effects, such as pipe whip and discharging fluids that may result from equipment failures.

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Reactor cQolant letdown piping break evaluation, the je:t expansion model >,

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and evaluation of the effects of cracks in seismic CategorY--1-modera-t~->

energy lines have not been adequately evaluated by the licensee.

TOPIC NO. TITLE IIl-7 .A Inservice Inspection, Including.Prestressed Concrete Containments With Either Grouted or Ungrouted Tendons Difference Summary

. ~ 10 CFR 50 (GDC 53) as implemented by Regul ato*ry Guide 1.35 in part

. ' requires the following:

1. Measured tendon force acceptance cr~teria with time
2. Tendon forces be reported rather than wire .forces.
3. ~onc~ete surr~unding the tendon end anchorages be visually
  • ins!;)ected during the integrated. leak rate tests. * -
  • The inseryice inspection program at Palisades does- not meet the items stated a bo v e .

TOPIC NO. TITLE III-7.B Design Codes, Design Criteria, and loading Combinations*

Difference Summary 10 CFR 50 (GDC l, 2 and 4) as implemented by SRP 3.8 requires. that structures systems and components be desigDed for the loading that will be imposed on them and that they conform to applicable codes and standards.

  • Code changes affecting specific types of structural *elements have been identified which may require facility modification.. in order to provide safety margins in structures which would be r~quired by_c~rrent versio_~s.

of the applicable codes and* standards. That is 22 specific areas of design 3

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code changes applicable to the Palisades plant design have been identified -

where the current code requires substantially

  • greater safety margins th~n the earlier version of the code.
  • TOPIC NO. TITLE III-7 .C .
  • Del ami-nation of Prestressed Concrete Containment Structures Difference Summary Standard Review Plan 3.8.l and the ASME Code, as part of the.imtplementa-tion of GDC 16, state that consideration should be given to radial forces for portions of prestressed containments with curvature. These forces were not considered in the initial design. There is no radial reinforcement in the dome concrete resulting in tensi_on forces being r~sisted by the concrete*and possible delamination. A post-construction method to consider these radial forces is by inspection.. There is no _

'inspection program currently being performed by the licensee to assure that delamination has not oc~urred.

TOPIC NO. TITLE

_III-8.A Loose Parts Monitoring and Core Barrel Vibration Program Difference Summary The requirements of 10 CFR 50 (GDC 13) as implemented by Regula'tOry Guide 1.133, Revision 1, and SRP Section 4.4 prescribe a loose parts monitoring program for the primary system of light-water-cooled reactors. Palisades does not have a loose parts monitoring program that meets the criteria of Regulatory Gui de 1 .133 ~

. Review Criteria .*

10 CFR 50 (GDC's 2 and 30) as implemented by SRP 5.2.5 and R. G. 1.45 requires the measurement of 1eakage from the reactor coolant pressure boundary {*RCPB) to the containment and interfacing systems and states design criteria for _the systems employed for such.

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    • '*j Fo~ systems emp1oyed for measurement of 1 ea kage from t.he Rtn to the containment, R. G. 1.45 states that: 1) system should be au airborne*

.*. particulate radioactivity *monitor that ;s SSE *qualified, 2):. a minimum

. . 'I of two others should be present which are OBE qualified, anal 3) all

., systems should have a sensitivity to detect 1 eakage of 1 gpa within

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1 hoµr. Those employed for measurement of intersysten leakage should include sensors for things such as radioactivity, flow, level, pressure,

. temperature, etc. and be OBE qualified. All the above systems should

1) have alarms and indicators in the main control room, 2) !be readily testable and calibrated during normal operation, and have t~eir avail-ability in the technical specifications.

Difference s*ummary l) The leakage detection systems incorporated for *measurement of 1 eakage

  • l from the r_eactor cool ant pressur~ boundary to the contai.inment are.

.,_(! not in conformance with Regulatory Guide l .4S criteria- Mith regard to* not being sufficient in 1} number of types of systems, 2) sensi-

, i tivity, 3) seismic resistance and 4) testability durin.gi normci.1

' .operation. * *

2) A section is 1ack*ing in* the Palisades Technical Speciffcatio*ns con-cerning. operability of the reactor coolant pressure boumdary to the co*nta i nment 1 ea kage detection system. * -
3) Information concerning the leakage detection systems fGr the detection of inter-system reactor coolant pressure boundary leakage and the CVCS*Makeup Flowrate is incomplete. Th~efore,, 'i!le carynot determine the extent to* which Regulatory Guide 1.45 is met.
4) The reactor cool ant inventory ba 1a nee is only capable o:f 1 gpm sensi-tivity', performed on a daily basis, not 1 gpm w/in 1 hr *. as would be required for reliance on this system. In addition *. 1l:he. seismic qualification and testability during nonrial operation requirements for detection systems are not met. Therefore, it is no.f: appropriate.

to rely upon this for leakage detection.

TOPIC NO*. TITLE V-1 O.B RHR Reliability V-11.A Requirements for Isolation of High Pressure/Low*

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Difference Surrmary

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(1) Overpressure relief capacity is required by 10 CFR* 50 *(GDC' s 19 and 43) as implemented by SRP 5.4.7. BTP ASB 5=1*>> and Regulatory Guide 1.139 for the Shutdown Cooling System (SCS) when in opera-tion. i.e.9 not isolated fro~ the reactor coolant system *. The Overpressurization Protection System (OPS) fulfills this function and is required by Technical Specifications when Primary Coolant

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. *;. *system (PCS} temperature is less than 250°F. However. the SCS .

can be placed in operation at temperatures uptQ 325°F; thus. there is the potential for overpressuri*zation of SCS when it is not isolated from the primary coolant system.

(2) 10 CFR 50 (GDC 19 and 34} as implemented by SRP 5.4.7. BTP RSB-5-1 and Regulatory Guide 1.139 require that the plant can be taken from normal operating conditions to cold shutdown using only safety-grade systems, assuming a single failure and utilizing either onsite or offsite power through the use of suitable procedures.

The Pali sades pl ant has safety-grade plant systems capable of safe

-shutdown under thes~ conditions; however, the plant operating procedures rely upon other non-safety grade systems and do not specify* how the cooldown would be. accomplished by_ the operator ;~

the event of failures in non-safety grade systems.

TOPIC NO. . TITLE

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V-11.A (El ectr.ical) Requirements fer I sol at ion of High and low .

Pressure Systems Difference Summary 1 O CFR 50 *(GDC *35) requires that the ECCS ha ye suitable interconnections.

and isolation. The High Pressure (HP} and Low Pressure (LP) Safety

. l Injection (SI) Systems share common headers with the* headers being separated from each other by a single check valve and a motor operated valve in series. The HPSI and LPSI pumps and the motor operated valves between the headers all start at the same time. In the event of a

. *! sinole check valve failure, some HPSI flow wi11 be diverted to the LPsi system and. the LPSI could be overpressurized and damaged.. The cor.se::;.:.:ences of this event ha_s not been analyzed by the licensee.

TO?! C r:D*. TITLE VI-2.D Mass and Energy Rel ease for Pos.tu1 ated Pipe*

Breaks Inside Containment

  • VI-3 Containment Pressure and Heat Removal Capability_

Difference Sunmary 10 CFR*SO*(GDC's 16, 3S, and SO) as implemented by SRP 6*.2.1.4 require that containment design conditi'ons are not exceeded during-an accident assu~ir.g a single failure. Failure of single MSIV results in a two steam generator blowdown, which exceeds containment design pressure by l

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(Operational Note: The MSIV check. valves at the Palisades plant have

  • failed to close on three occasions following a ~hutdown and cooldown.

. :I On Septeir~er 21, 1972, CV-0510 failed to operate because*1:he linkages were sticky on six of the solenoid valves. The stickiness was attributed

. to dirt,.and the solenoid valve linkages were cleaned amid relubricated. , .

  • The solenoid valve linkages have been covered with plastic covers to *
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mini*mize dirt pickup and a dry lubricant, recorrrnended bl' the manufacturer :

to limi~ long term dirt. pickup, is applied at every outage.

On May 19, 1973, CV-0501, and on August 12, 1973, CV-0510 failed to

.close because of the binding in the stuffing boxes. The inside and out-side diameter of the lantern rings and packing followers were machined, in accordance with factory recommendations, to increase 'the clearance *

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around the operating shaft. The valve packing is now inspected at each*

refueling outage and the MSIVs are exercised several timJes following each cold shutdown).. * *

  • TOPIC NO. TITLE

. VI-4 Containment Isolation Systems

  • Difference Summary
1. The isola~ion valving arrangements do not meet the requirements of 10 CFR SO (G~: 55 or 56) as implemented by SRP 6.Z.4 from the stand-point of valve location for penetrations l, 4, 4a, 10, 11, 25, 26, 30, 31, 33, 36, 37, 38, 39, 40, 40a, 41, 421 44, *45_- 46.; .47, 49, 52, 65, *67, 68, and 69.. * *
2. !sol aticn -.*al v2s differ from the explicit requirerrents of 10 CFR so*

{GDC 55, 56, ano 58) as implemented by SRP 6.2.4 frm *the standppint of valve ty;:ie by using one* check valve in series wt'th other type isolation valves located outside containment for penetrations 7, 8, 10~ 11, 1.4, 26, 30, 31, 37, 39, 41, 42, 45, 65, and fil.

3. Isolation barriers differ from the explicit requirements of 10 CFR SO .{GDC 55, 56 and 57) as implemented by SRP .6.2.4 from the stan~

point that pip~ caps or blind flanges are used as containment isola-tion bar~i~~3 as follows:

a. *Pen::r'-:-: or.s with pipes or test connections* capped outside contain-ment: 13, 17, 21, 25, 28, 29, 38, 39, 48,-and 73;
b. Penetrations with blind flanges inside containment: 18, 27, 29, *
  • and _73; and
c. Several lines associated with the* following penetrations which a.re -

equip;;2d with pipe caps: *the pe.rsonnel air lock (penetration 19);

emerg~ncy access air lock (penetration SO}; .and equipment hatch *

{penetration 51). .

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10 CFR 50 (GOC 55) as implemented by SRP 6.2.4 requires that.two automatic valves be provided for isolation. *Penetration 44 shows a manual isolation valve (3/4"-2084) in series with an air operated isolation valve that differs from the requirement from the stand-point of valve actuation.

5. Certain penetrations have been provided with remote ma,nual isola.;. *

' tion valves, which is acceptable. However. 10 CFR 50 (6DC 1 s 55, 56 and 57)*as implemented .by SRP 6.2.4 Item II.6.C require provisions should be made to allow the operator in the main control room to

    • J know when to isolate fluid systems equipped with remote manual "1 isolation valves.
6. 10 CFR (GDC 55 and 56) specify that automatic isolation valves should, .

upon .loss of actuating power, take the posit"ion that provides greater safety. The position of an isolation valve for normal and shutdown operating conditions, and pnst-accident condition, depends on the fluid system function. . !n the event of powi:ar failure to a: valve operator, the valve position should be consistent with the line function. In this regard, separate power supplies for isolation. valves in series

.may be required to assure the isolation of non-essential: lines.

1 .

~ - . TOPIC NO. TITLE  : I

.. i I

VI-6 Containment Leak Testing Difference Summary . '  ;  !

lO*CFR 59, Appendix J requires that tests be performed to assure that leakage through the primary reactor containment and systems and components penetrating.primary containment shall not exceed allowable leakage rate values as specified in the technical specifications or associated bases. . I I

The licensee has requested an exemption to the requirements of airlock leak testing if the airlock is opened between the six-month Type B test. This *  !

request for exemption has been denied and the .licensee has been requested to ' i modi.fy the technical specifications to require a full Type B test of the personnel airlock at a pressure of Pa.at least once every six months, with a verification of airlock door seal inte~rity within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each opening JI or the first of a series of openings during the interim between six month *. J tests, whenev~r containm~nt integrity is required. PJant modification will be required i~ order to perform "this testing.

TOPIC NO. TITLE Vl*lO.A Testing of Reactor Trip System and Engineered Safety Features Including Response Time Testing

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.-i 10 CFR 50.55a (h) through IEEE Std. 279-1971, Sections 3 (9) and.4.10 requires that response time testing be performed* on a periodic basis

-  ; at Palisades, only the control rods, the diesel generators start ~ime and load- sequencers, and some of the containment isolation valves are response time tested *.

TOPIC. rm. TITLE VII-1.A Isolati:n of Reactor Protection System fr:: r:Dn- -.-

Safety Systems, Jncl,uding Qualifications of Isolation Devices - ..

- i Difference Summary

  • 10 C~R 50.55a (h) through .JEC:E Std. 279-1971 requires that safety signals be isolated from non-safety signals and that no credible failU:Te at the output of an isolatfon device shall prevent the associated pro.t:ecti.on system channel from meeting the_ minimum perfonnanc~ requirements specified in the design bases.*
  • At Palisades, electrical sigr.a1s from RPS steam ge~erator pressure channel B *and reactor* coolant flow channel A are run* to the plant .computer without any isolation device.

TOPIC NO. TITLE VII-3 Systems Required for Safe Shutdown Difference Summary Electrical Items

1. 10 CFR 50 (GDC 17) requires* that two paths must be.avail.le from the safety busses to the offsi'te power system. One path m.1st be imniediately available anci the other path must be made available in a short period of tir::e. A"':. Palisades, it will take four "to six hours to establish tne aeiayed access path. Only two hours of battery capacity exist and the consequence of loss of all AC and DC have not been evaluated.
2. 10 CFR_50.55a (h) re~~~~e! that channels that provide signals for the same protective foncti on shall be independent and physically .-

separated to accomplish decoupling of the effects of unsafe environmental factors, electric transients, and physical accident consequences docurneii::: ~~ :~e design basis,- and to reduce_ the likelihood of interactfons t:twee.n channels during maintenance

  • qperations or in ths e*:e~: of channel malfunction. **

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At Palisades, the instrumentation for. high pressure scram signals

~oes not sa~isfy this r7quirement because the channels. are divided into t~o ~a1rs, ea~h pair shares a po~r source and cable routing,

' . the trip is not fail safe on loss of power and the logic is two out of four.

Because the Technical-Specifications permit operation with_only two operable channels (if one of the inoperable channels is tripped}, there are several scenarios in which the high pressure

_trips fail. The most simple of these is to fail the de source which is assumed to be cc...,.,vn to the remaining channels. Al-ternatively, assuming one of the four reactor protection system channels is bypassed, but not tripped (operating in a two out of three 1Qgic arrangement for reactor trip) and assuming the

  • -~ single failure of one race~*:?y or one de power source causes failure of .two high pressure trip signals, the reactor would not trip when required.

3 *. 10 CFR 50 (GDC.17) requires that both onsite and offsite power systems *shall provide power to systems and co111ponents important to safety. At Palisa9es the Boric Acid Injection System is not independent of* a loss of offsite *power because the Boric A.Cid Tank heaters (which ensure boron remains in solution) are not

  • powered from a safety related bus (i.e., cannot be powered from the diesel gsnerator). -

TOPIC NO. TITLE VII-3 -- Systems Re qui red for Safe *Shutdown Difference Summary 10 CFR 50 (GDC's 2 and 34) as.implemented throughSRP 5.4.7~ BTP RSB 5-1, and R. G. 1.139 in part require th~:~~; Seismic Category I water supply for auxiliary. feedwater have suffici~nt inventory to *permit operation at hot shutdown for four hours foiic.h.;..; ~J cooldown to conditions permitting shutdown cooling syste~ ~s:~) initiation. The inventory is based on* the *cooldown time assuming a sirigle active failure and either only onsite or only offsite power.

Sufficient safety-gr~de water is not maintained in a seismically qualified*

tank(s) to perform this function.

10 * -Enctcisiire 2- - -*--.

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  • ' . TOPIC .NO. TITLE

. *i VIII-3. A Station Battery Capacity Test Requirements

} Difference Summary

' ~ '

.10 CFR 50 (GDC 18) as imp.lemented by Regulatory Guide 1.. 129_requires

  • periodic testing for determining battery capacity and for demonstrating_ .
  • that the batteries will provide sufficient power* under accident conditions.

The Palisaqes program for testing the batteries does not-satisfy these requirements.

TOPIC NO. TITLE IX Station Service and Cooling Water Systems

  • Difference Summary l.O CFR*SO (GDC 44) requires that for onsite electric power system opera-tion {assuming offsite power is.not available) the ultimate heat sink
  • cooling water system safety function can be ciccompli shed 9 assuming a _

single failure.** With loss. of offsite power and the single failure of diesel 1-2,suffic.ient service water flow rr.ay not be . provided to prevent exceeding design temperatures in the comp.anent cool.ing water system.

The* capability exists to-throttle service wa~er flow to non-essential*

components. Procedures do not exist nor have the effects of temperatures in excess of design been evaluated.

TOPIC NO. TITLE IX-5 Ventilation Systems Difference Summary

l. The isolat*ion of the engineered safegt!:>rd equipment ventilated area remains questionable que to the presence of non-s~fety grade isolation dampers. * *.. - *
2. The "penetration and fan room" ventilation system performance is vulnerable to failure of either emergency d~ese1 generator*.. The failure of one d.iesel to start when r~::.:~:-:: :-esults in loss of either the supply or exhaust fan. The situation could possibly lead to service conditions exceeding the design* pa!"a:::eters of equipment housed in these areas.
3. The ventilation equipment for the "Auxiliary and Radwaste Areas9" "Turbine Building," "Intake Structure,: o.r.~ ***~*i-ewing Gallery, Switch:-

gear and Cable Spreading Areas," ser.vi c~ equipment deemed:. essential for.safety. However, these ventilation.systems* are neither safety grade, powered from emergency sou:ces ror singl_e failure proof.

11 i

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TOPIC NO.

  • TITLE

. ..:, ~

XV-2 Spectrum of Steam System Piping Failures Inside and Outside Containrr~nt I

Difference Summary

  • For analysis of a spectrum of steam line br~aks, 10 CFR 50 (GDC's 17, -

21 and 35} as implemented by SRP 15.1.5 require that the most severe single active component failure should be assur:ied and the effect of

  • loss of offsite power should be considered.
  • Other single failures in mitigating systems have not been analyzed to a sufficient extent so that it can be concluded that the effects of the worst sin;~a failure have been

.considered. These other single failures are:

l. Diesel ge~erator failure (with loss of.offsite power)
2. Failure of main feedwater isolation.

For some of the~e events, ~he licensee has recently submitted analyses*.

tOPIC NO. TITLE XV-12 Spectrum of Rod :*:Ejection Accidents Difference Summary 10 CFR Part 50 (GDC 28) as implemented by Regulatory Guide 1.77 and SRP 15.4.8 require that reactivity limits be established on*the reactivity control systef!l .. Our.aryalysis~ required to demonstrate the acceptability of the.

reac_t1 v1 t~ 11 m~ ts, ma inly _the rod ejection accident, was eva1uated for fuel- melt1~g (1.e., less than 200 cal/gm) but not fuel. cladding failures.

The a~aly~1s of the number of pins that would experience DNB for the limiting rod eJect1on event, and the effect on the dose calculations has not been .

performed.

12.

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