ML18026A248

Part 21 Rept Re Substantial Safety Hazard in Design of Facility for Loss of Normal Spent Fuel Pool Cooling

ML18026A248
Person / Time
Site:	Susquehanna
Issue date:	11/27/1992
From:	Lochbaum D, Prevatte D AFFILIATION NOT ASSIGNED
To:	Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML17157C139	List: ML17157C138 ML17157C140 ML17157C141 ML17157C142 ML18026A248
References
REF-PT21-92 NUDOCS 9301050135
Download: ML18026A248 (90)

Text

Mr. Thomas T. Martin Regional Administrator, Region I

United States Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 November 27, 1992

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION DOCKET NO.

50-387 LICENSE NO.

HPF-14 10CFR21 REPORT OF SUBSTANTIAL SAFETY HAZARD

Dear Mr. Hartin:

Pursuant to the requirements of 10CFR21, Reporting of Defects and Noncompliance, this letter is submitted to report a "substantial safety hazard" that exists in the design of the Susquehanna Steam Electric Station (SSES) located near Berwick, Pennsylvania.

This report is being made by Hr. David A.

Lochbaum

who, through July of this year, worked as a contract engineer in Pennsylvania Power Light Company's (the licensee)

Nuclear Plant Engineering'ection, and Hr.

Donald C. Prevatte who is currently, and until the end of this year, working as a contract engineer in PPSL's Nuclear Plant Engineering Section.

The substantial safety hazard is as follows:

The SSES design for a

loss of normal spent fuel pool cooling fails to meet numerous regulatory requirements.'s a

. result, for a

design basis

accident, there is the, potential for meltdown o'f irradiated fuel outside primary containment and the failure of all safety-related systems in the reactor building.

For an operating

plant, 10CFR50.72 requires licensees to report in one hour any'instance of the plant (a) being in an unanalyzed condition that significantly compromises plant safety, (b) in a

condition that is outside the design basis of the plant, or (c) in a condition not covered by the plant's operating and emergency procedures.

It also requires that reports shall be made within four hours of any condition that alone could have prevented the fulfillment of the safety function of structures or systems needed to (a) shut down the reactor and maintain safe shutdown.

(b) remove residual

heat, (c), control radioactive

release, or (d).mitigate the accident.

All of these conditions exist at SSES for the design basis accident (DBA) loss-of-coolant accident (LOCA) or LOCA with a loss-of-offsite-power (LOOP) as a result of the heatup. of the spent fuel pool which mechanistically follows these accidents.

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GLUJi'

'930i0501$ 5 92ii27

PDR ADOCK 05000387 S

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On October 9,

1992, after seven months of attempts to convince PP8L's management to address these concerns as

required, the signatories to this letter declared to PP8L management our intent to report these concerns to the HRC ourselves unless they were properly handled by PP8L.

In response to our declaration and the

'actions it initiated, Pennsylvania Power 8

,Light Company submitted Licensee Event Report (LER) 92-016-00 to the Huclear Regulatory Commission on Hovember 17, 1992.

Although PP8L's report acknowledged that concerns had been raised, it dismissed them as having minimal

. safety significance.

The LER is incomplete, inaccurate, unbalanced and misleading in its presentation of our

concerns, the pertinent technical and licensing information, and its conclusions.

The purpose of this letter is to inform the HRC that we still consider these concerns to be a

"substantial safety hazard" which should have been reported by PP8L under 10CFR50.72.

The focus of our concerns is the inability to remove decay heat from the spent fuel pools for the various design events which mechanistically incapacitate the normal fuel pool cooling system and the resultant effects from loss of normal cooling on the safety-related systems and components in the reactor buildings.

The heart of the PPSL position stated in LER 92-016-00 is a

legalistic ar gument that the licensing basis of SSES does not require the loss of normal fuel pool cooling to be considered concurrently with other 'design basis events such as LOCA or LOCA/LOOP.

Me agree that loss of normal spent fuel pool cooling is not required to be postulated concurrently, but when it follows mechanistically as a result of the design basis events as it does at SSES, it must be considered.

PP8L cites in the LER as support for its position FSAR Section 9.1.3 and Appendix 9A which it contends contain the only design basis requirements for the fuel pool cooling failure which must be considered

- basically, failure due to a seismic event.

Me contend that there're other conditions within the SSES licensing basis as described throughout the FSAR which will mechanistically cause fai 1 ur e of the non-safety related norwal fuel pool cooling

system, such as hydrodynamic loads associated with a

LOCA, environmental conditions associated with a

LOCA, LOOP, failure of the non-safety related service water

systea, and

randoa, single failures.

In 1988, PP8.L introduced another failure mode when it implemented procedures to manually de-energize non-lE loads in the reactor building following a LOCA without a

LOOP.

Me contend that as with all other systems described in the

FSAR, the design and operation of the fuel pool cooling system cannot be taken out of context of its mechanistic relationships with the other

systems, events, and licensing bases without review and approval by the HRC.

The current design and operation of SSES for a

loss of normal spent fuel pool cooling, even for failure Page 2

due to the seismic event indicated by PP8L to be within the licensing

basis, clearly do not meet the design or licensing basis requirements if the effects on safety related structures, systems'nd components in the reactor buildings are considered.

The design basis accident for SSES is the LOCA with a concurrent LOOP.

For this event, it must be assumed that the normal fuel pool cooling system will fail as described above.

Therefore, the removal of decay heat from thespent fuel pools must be accomplished by the design basis method (and only safety related method available) described in Section 9.1.3 of

,the FSAR; allowing the fuel pool to boil and providing makeup from the safety related emergency service water (ESW) system.

However, at the present time there are no design provisions, analyses or procedures which adequately define

how, within the applicable regulatory requirements, this function will be accomplished.

The effects of the boiling spent fuel pools on safety related equipment in'he reactor buildings are also unanalyzed.

These deficiencies exist even for the loss of spent fuel pool cooling event described in FSAR Appendix 9A.

When these concerns are addressed within the context of the regulatory requirements, it appears that the necessary steps to provide makeup water from the ESW system following a design basis LOCA cannot be performed due to very high radiation in the areas where valves

. must be manually operated.

Additionally, the current Eg temperature limits of virtually all of the safety related equipment in the reactor building will probably be exceeded by a large margin due to the heat and moisture put into the reactor building atmosphere by the boiling spent fuel pools.

To appreciate the significance of these

concerns, the magnitude of the potential effects for the design basis accidents must be considered:

1)

The currently calculated radiation levels at some of the ESM valves which must be manipulated are in the thousands of R/hour,'not including the associated airborne dose which may be in the hundreds of R/hr.

2)

The boiling fuel pools will add approximately 20 million BTU/hr of latent heat to the reactor building atmosphere which is not currently accounted for.in the calculations..'he total heat load in the reactor building that is currently accounted for is only 5.2 million BTU/hr, and even at this heat load there are a

number of areas where the accident teIIperatures slightly exceed the Eg temperatures.

3) At the design makeup rate from ESW to the fuel pools, 5.2 million gallons of water are introduced into the reactor buildings (both the acci'dent and non-accident units will be affected) either through evaporation/condensation or Page 3

spillover of the pools.

Hone of the possible detrimental effects of this water on the safety related structures, systems and components in the reactor buildings have been analyzed.

In fact, PP8L's own recent engineering evaluation for these concerns determined that the standby gas treatment system would isolate due to high inlet temperature.

In addition to these most significant safety

concerns, there are also related concerns of lesser safety significance which nonetheless constitute "substantial safety hazards".

These include the following:

4)

Fuel pool instrumentation for monitoring the cooling of the fuel pool post-LOCA (a

safety related function) is not environmentally qualified, and the readouts are located in an area which is not accessible to the operators post-LOCA.

5)

The design heat loads and the calculated times to boil for the spent fuel pools have not been updated to reflect changes that have been made in the fuel design, fuel cycle

length, and refueling procedures.

Following our October 9,

1992 declaration of intent to report to the NRC on these

concerns, there ensued large scale efforts within the PP5L Nuclear Department to analyze the concerns and define the actions needed to be taken.

This activity produced an engineering

report, NE-92-002 (attached).

This report described extensive modifications and procedure changes required for SSES to cope with a loss of normal fuel pool cooling event.

Although the repor t addressed many of our concerns, it did not, adequately address all of

them, and some of the proposed solutions are either technically inadequate and/or they do not meet regulatory requirements.

In general, the proposed solutions are not acceptable. for the following reasons:

1) They place heavy reliance on non-safety related equipment and functions.

2) They place heavy reliance on plant modifications which have not yet been implemented or even designed.

3) They place heavy reliance on procedure changes which have not yet been made.

4) They place heavy reliance on analyses which have not yet been performed.

5) They place heavy reliance on operator and EOF personnel training which has not yet been developed or performed.

Page 4

6) They place heavy reliance on operator actions following a

LOCA when there is already heavy dependence on operator actions and monitoring, and these additional actions must be performed under extremely adverse environmental conditions.

7) They rely on assessments of operator accessibility to the reactor building which in turn are based on assumptions of core damage which are unreviewed by the HRC and are substantially less than the assumptions required by HUREG-0737 and the SSES licensing basis reflected in Chapter 18 of the FSAR.

Additionally, the accessibility position taken in the report with re'spect to airborne radioactivity contributions is inconsistent with the requirements in HUREG-0737, 10CFR50 Appendix J, actual SSES Appendix J test

results, and the design of other plant systems (e.g.

secondary containment and the standby gas treatment system).

For NRC mandated DBA conditions, as stated in FSAR Chapter 18, the reactor building is inaccessible for days following a

LOCA.

8) They rely on probability arguments which may be acceptable in an Individual Plant Evaluation and in a justification for interim operation, but which are not acceptable substitutes for compliance with regulatory requirements, unless they are reviewed and approved by the NRC.

These have not been.

9) In some

areas, the report's conclusions are inconsistent with the facts presented.

For example, the report concluded that Zone III venting is acceptable, whereas the supporting documentation indicates that the 10CFR100 and 10CFR50 Appendix A Criterion 19 allowables for offsite and control room doses respectively are exceeded.

It should also be considered that the conclusions in this report represent PPE L's vision of systems, equipment and procedures in the future, not as they exist today.

Although essentially none of the technical inforaation from the report is contained in their LER, this information along with their legalistic arguments discussed

earlier, has provided the underlying bases for their determinations of operability and reportability.

But the law requires determinations of operability and r eportability to be.

based on the plant conditions as they exist at the time of discovery as discussed in considerable detail in NURE6-1022.

In addition to these technical

concerns, we also must point out the conditions adverse to quality that PP&L's handling of this case (and other recent safety concerns) demonstrates in violation of 10CFR50 Appendix B.

Since our concerns were first discovered and reported'n Harch of this year, there has been a programmatic failure by PP8L to properly evaluate these concerns.

PP8L repeatedly attempted to improperly dismiss these concerns m-indef ini te1y def et their eval uati on by methods i net ucfi~g Page 5

classifying them as design basis document issues (in clear violation of the guidelines expressed in NUREG-1397), selectively applying regulatory requirements to permit favorable conclusions, claiming that the NRC had already reviewed and approved the design deficiencies based on the FSAR/SER text, and even claiming that an informal, undocumented agreement had been made with the NRC at the time of initial licensing of the plant.

Our experience and our knowledge of the difficulties encountered by other eng'ineers with nuclear safety concerns for SSES indicates that PP5L's program for handling nuclear safety issues is itself cause for concern.

While PP8L cites data to support their contention that their discrepancy management system is effective, most of their data points represent relatively minor discrepancies which are easy to resolve.

However, for large problems with extensive or uncertain resolution such as in this case, the system lacks the ability to assure proper evaluation and subsequent implementation.

In these

cases, PPIL's treatment violates their own administrative procedures controlling discrepancy management.

In the nuclear power

industry, organizations such as PPEL and individuals such as ourselves have legal and ethical responsibilities..

PP&L has not fulfilled its responsibilities in this case and has forced us to fulfill our s by submitting this letter.

The more detailed technical descriptions for these concerns and the history of their treatment are contained in numerous letters,

memos, and documents.

A listing of pertinent documents is contained in Attachment 1 to this letter, with copies of these documents provided as the remaining attachments to this letter.

We expect that.you may require additional information from us regarding this matter.

Me will make every effort to support your requirements.in a

timely manner.

We can be reached at the a'ddr esses and telephone numbers listed below.

Me would also greatly appreciate being kept informed of your actions regarding this matter.

Thank you for your consideration.

Sincerely,

~g David Lochbaum Donald C. Prevatte C

80 Tutt 1 e Road Watchung,'NJ 07060 (908) 754-3577 7924 Woodsbluff Run Fogelsville, PA 18051 (215) 398-9277 Page 6

Distribution List Mr. Thomas T. Martin (with all attachments)

Mr.

G.

S.

Barber Senior Resident Inspector US Nuclear Regulatory Commission P.O.

Box 35

Berwick, PA 18603-0035 (with all attachments)

US Nuclear Regulatory Commission Attention:

Document Control Clerk Mail Station Pl-137 Mashington, DC 20555 (with all attac'hments)

Director, Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Mashington, DC 20555 (with Attachment-1)

The Honorable Ivan Selin Chairman US Nuclear Regulatory Commission Mashington, DC 20555 (with Attachment 1)

The Honorable Kenneth C. Rogers Commissioner US Nuclear Regulatory Commission Mashington, DC 20555 (with Attachment 1)

The Honorable James R. Curtiss Commissioner US Nuclear Regulatory Commission Mashington, DC 20555 (with Attachment 1)

The Honorable Forrest Commissioner US Nuclear Regulatory Mashington, DC 20555 J.

Remi ck Commission (with Attachaent 1)

The Honor able Gai 1 De Commissioner US Nuclear Regulatory Mashington, DC 20555 Planque Comwission (with Attachment 1)

Page 7

Attachment 1

List of Attachments 7

s 10 List of Attachments PP8L Memo from Dave Lochbaum and Don Prevatte to Mark Mjaatvedt, "Susquehanna Steam Electric Station Spent Fuel Pool Boiling Issues",

March 19, 1992 (ET-0149)

PP&L Engineering Discrepancy

Report, "Loss of Spent Fuel Pool Cooling Event Design Discrepancies",

Originated April 16, 1992 and Dispositioned October 6,

1992 (EDR G20020)

PP8L Operability Statement, "EDR 8G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies",

April 23, 1992 PP8L Memo from Dave Lochbaum and Don Prevatte to Joe Zola "Supplemental Information for EDR G20020 on Boiling Spent Fuel Pool",

June 22, 1992 (ET-0471)

PP8L Draft Screening Morksheet prepared by Art White, "EDR No ~

G20020", July 1, 1992

Handout, "EDR G20020 References",

July 15, 1992 White Paper prepared by David A.

Lochbaum and Donald C

Prevatte, "Safety Consequences of a Boiling Spent Fuel Pool at the Susquehanna Steam Electric Station", July 27, 1992 PP&L Memo from 6.

D. Hiller to G.

T.

Jones, "Fuel Pool Cooling Deficiencies",

August 18, 1992 (ET-0586)

PP&L Memo from D.

C.

Prevatte to G.

T.

Jones, "Fuel Pool Cooling Deficiencies",

August 20, 1992 (ET-0587)

PP&L Memo fr om A.

Dyszel to T.

C.

Dal piaz, "U2 RI05 Fuel Pool Decay Heat Evaluation",

August 21, 1992 (PLI-72230) 12 13 15 PP8L Memo from J.

M ~

Kenny to 6.

T ~

Jones and C ~

A ~

"EDR on Fuel Pool Cooling", August 25, 1992 PP&L Memo from Geor ge T.

Jones to Glenn D. Hiller Pool Cooling EDR '

620020, 600005",

August 27 (PLI-72267)

PP8L Memo. from 6lenn D.. Hiller to George T.

Jones Pool Cooling EDRs

G20020, 600005",

August 31 (PLI-72297)

PP8L Memo fr om Kevin M.

Br inckman to Geor ge T

"Review of Fuel Pool Cooling",

September 1

(PLI-72288)

Myers Fuel 1992 Fuel 1992 Jones 1992

'.. 9301050135

Attachment 1

List of Attachments (continued)

~N 16 17 18 19 20 22 23 24 25 26 27 28 PP&L Memo from J.

R.

Miltenberger to G.

T.

Jones, "Spent Fuel Pool Cooling", September 9,

1992 (PLI-72367)

PP8L Letter from James E.

Agnew to David A.

Lochbaum, "EDR

G20020, Spent Fuel Pool Design Discrepancies",

October 7,

1992 (ET-0785)

PP8L Memo from G.

D. Hiller to G. 0. Hiller, "Assignment of EDR", October 7,

1992 (ET-0?80)

Letter from David A.

Lochbaum and Donald C.

Prevatte to George T.

Jones, "Reportability of Boiling Spent Fuel Pool Concerns",

October 9,

1992 PP&L Memo from D.

A.

Lochbaum and D.

C.

Prevatte to George T.

Jones, "EDR System Concerns",

October 13, 1992 (PLI-72365)

PP&l Memo from George T. Jones to G.

D. Hiller, "Spent Fuel Pool Issue",

.October 14, 1992 (PLI-72640)

PP&L Memo from George T.

Jones to G.

D. Miller, J.

S.

Stefanko and M.

M. Simpson, "Spent Fuel Pool Cooling Issue",

October 14, 1992

( PL I-? 2641 )

PP&L Memo from George T.

Jones to All Nuclear Engineering Managers and Supervisors, "Engineering Discrepancy (EDR)

Program",

October 14, 1992 Letter from David A.

Lochbaum and Donald C.

Prevatte to George T. Jones, "Disagreement with Screening, Reportability and Operability Evaluations for EDR G20020",

October 14, 1992 Nemo fr om Charles A.

Hyers to George T.

Jones, "Fuel Pool Cooling Issues Repor tabi 1 ity / Operability", October 20, 1992 I

PP8L Memo from Glenn D.

Miller to George T.

Jones, "Evaluation of EDR G20020 Spent Fuel Cooling Issue",

October 21,

'1992 (PLI-72711)

PP&L Memo from David A.

Lochbaum and Donald C. Prevatte to George T.

Jones, "Evaluation of EDR 620020 Reportability/Operability",,

October 26, 1992 (PLI-72739)

PPSL Memo from David A.

Lochbaum and Donald C. Prevatte to George T.

Jones, "Response to Evaluation of EDR G20020",

October 28.

1992 (PLI-72751)

Attachaent 1

List of Attachments (continued)

N 29-

'PP&L Memo from Glenn D.

Miller to George T.

Jones, "Evaluation of EDR G20020

- Spent

.Fuel Pool Cooling Issue",

October 29, 1992 (PLI-72763) 30 PP&L Engineering

Report, "Loss of Fuel Pool Cooling Event Evaluation for EDR PG20020",

October 29, 1992 (KE-92-002 Rev.

0) 31 32 PP&L Memo from Glenn D.

Hiller to George T. Jones, "Revised Evaluation of EDR G20020 Spent Fuel Pool Cooling Issue",

October 29, 1992 (PLI-72764)

PP&L Memo from David A.

Lochbaum and Donald C. Prevatte to George T.

Jones, "Position on EDR G20020 and Planned Actions",

November 2,

1992 (PLI-72783) 33 PP&L Hemo from David G. Kostelnik and Mark R. Hjaatvedt to George T.

Jones, "Comments on PLI-72783 Regarding EDR G20020",

November 11, 1992 (PLI-72857) 34-PP&L Letter 'rom H.

G.

Stanley to the U.S.

Nuclear Regulator y Commi ssion, "Licensee Event Report 92-016-00",

November 17, 1992 (PLAS-546)

PP&L Safety Eval uation Summar y, "Procedure EO-IP-'055'",

1988 (SER No'.88-127)

Attachment 2

PP8L Memo from Dave Lochbaum and Don Prevatte to Mark Mjaatvedt, "Susquehanna Steam Electric Station Spent Fuel Pool Boiling Issues",

March 19, 1992 (ET-0149)

Note:

This memo documents the discovery of the problems with the loss of normal spent fuel pool cooling event and the reporting of these problems to a supervisor in the PP8L Nuclear Plant Engineering Section.

Approximately four weeks later, the authors of this memo were directed to initiate an Engineering Discrepancy Report on the concerns.

PP8L's decision to generate an EDR on these concerns may have been driven by schedule interests

- the authors, as preparer and technical reviewer of reactor building heat load calculations'to support the PP8L Power Uprate Project, would not sign off on the calculations until these concerns were addressed.

Upon generation of the EDR, the authors signed off the calculations conditionally with a

note that the-results might be affected by the disposition of the EDR.

PP8L needed these calculations issued in order to submit their engineering report on power upr ate to the NRC in June 1992.

0

~

~

~

MEMORANDUM TO:

PROM:

Xae)e-"Mgaatvedt Dave Lochhaum Don Prevatte DATE: March 19, 1992 JOB FILE

SUBJECT:

Paver Uprato NUMBER: ET-01A9 P88-1 COPZES: Distribution REPLY:

No SUSQUEHANNA STEAM ELECTRZC STATZON SPENT PUEL POOL BOZLZNG ISSUES Potential problems resulting from a boiling spent fuel pool have been uncovered during the preparation of the engineering evaluation of the fuel pool cooling system and the reactor building heat load calculations at power uprate conditions.

These

problems, their brief history and recommendations to resolve these issues are presented in the attachment to this letter.

The consequences of these problems willbe made incrementally worse by power uprate and therefore must be addressed before implementation of power uprate.

More importantly, however, these problems affect the current operation of SSES and must be evaluated and resolved as soon as possible.

If additional information is needed, please contact Dave Lochbaum at ETN 220-7768 or Don Prevatte at ETN 220-7781.

David A. Lochbaum Donald C. Prevatte DISTRIBUTION:

J A Bartos M B Detamore J

M derley SRMS Corres File A6-3 (w/a)

A6-2 (w/a)

A6-3 (w/a)

A6-2 (w/a)

Page 2 of 10 ACtachment

Boiling Spent Fuel Pool Issues REACTOR BUILDING HEAT LOADS Problems Reactor building design heat loads do not account for the boiling spent fuel pool event.

History:

The calcs of reactor building pre-uprate and uprate heat loads for Zone I, II and III under normal and accident conditions (calcs M-RAF-052,

-053,

-054) assume the spent fuel pool temperature remains at 125'F for all cases.

This assumption relies upon use of the service water system to remove heat from the fuel pool heat exchangers post-LOCA and the fuel pool cooling assist mode of RHR to remove heat from the fuel pool post-LOOP.

Neither of these operating modes is safety related and therefore may not be available.

The design provision for the loss of fuel pool cooling event is to permit the fuel pool to boil and use ESW to maintain the level in the pool above the top of the fuel.

ESW provides redundant seismic Category I makeup lines to each of the two spent fuel pools.

If the spent fuel pool is permitted to boil, the heat loads in the reactor building, particularly in Zone III, increase significantly.

These higher heat loads have not been considered in reactor building analyses to date.

The equipment qualification of safety-related equipment in the reactor building may therefore be adversely affected if the heat loads from a boiling spent fuel pool are considered.

Recommendation:

An EDR was prepared on this condition.

The options available to resolve this problem include:

1) Analyzing the reactor building heat loads for the boiling spent fuel pool case and update associated analyses for equipment qualification.

2) Providing design capability to maintain spent fuel pool temperature

< 1254F using safety-related equipment such that existing reactor building heat load analyses are adequate.

Page 3 of 10 Attachment

Boiling Spent Fuel Pool Issues IZ.

PUEL POOL TIME-TO-BOIL AND RADIOLOGICAL RELEASE ANALYSES First Problem:

The analytical 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil for the spent fuel pool is nonconservative for the maximum normal heat load in the spent fuel pool.

History:

Bechtel calc 200-0048 Rev. 1, "Boiling Spent Fuel Pool" dated May 7, 1982, determined time-to-boil using the equation:

Time-to-Boil =

(m

• C

• OT) / Q, where m =

mass of water in fuel pool, lb C =

specific heat, BTU/lb-'F OT = difference between final pool temperature (212'F) and initial pool temperature (125 F) I F

Q =

fuel pool decay heat load, BTU/hr This calc used a decay heat load of 9.79x106 BTU/hr for Unit 1 and 7.92x106 BTU/hr for Unit 2 to determine times-to-boil of 25.15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 31.087 hours0.00101 days <br />0.0242 hours <br />1.438492e-4 weeks <br />3.31035e-5 months <br /> respectively.

FSAR 9.1.3.1 establishes the maximum normal heat load as that heat load resulting from 2840 assemblies discharged to the fuel pool by a routine refueling schedule.

FSAR Tables 9.1-2a and 9.1-2b report the maximum normal heat load for Units 1 and 2 as 12.6x10 BTU/hr.

These values were determined in Bechtel calc 153-9 Rev.

1, "Fuel Pool Decay Energy and Temperature".

The spent fuel pool decay heat values used in the boiling spent fuel pool calc and for the FSAR discussion were based upon assumptions for cycle operating lengths, fuel exposures, and reactor power level.

SSES has subsequently operated differently than had been assumed such that the decay heat loads in the filled spent fuel pools may exceed 9.79x106 BTU/hr, resulting in a shorter than analyzed time-to-boil.

Calc NFE-B-NA-053 Rev.

0, "Decay Heat from a Full Spent Fuel Pool (ASB9-2 Method)",

determined decay heat from a filled spent fuel pool using actual fuel operating history through 1991 and assumptions which bound operation after power uprate.

This recent calc reported a

maximum normal heat load of

Page 4 of 10 Mtachment

Boiling Spent Puel Pool Issues

=17x106 BTU/hr.

The methodology use in this calculation is conservative and may over predict actual decay heat loads by

=204.

Calc M-FPC-009 was drafted to determine the spent fuel pool time-to-boil and recpxired ESW makeup rate for power uprate.

Preliminary results from this calc indicate the fuel pool boils 19.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after loss of fuel pool cooling for the design heat removal capacity of the fuel pool cooling system (13.2xl06 BTU/hr).

This calc determined a time-to-boil of

=15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the '17x10 BTU/hr heat load calculated for the power uprate case.

Recommendation:

The basis for the time-to-boil analysis should not be the maximum normal heat

load, since this value is subject to assumptions of reactor operation which are extremely difficult to predict.'wo options are proposed:

1) The time-to-boil analysis for the loss of normal spent fuel pool cooling case should use the design capacity of the fuel pool cooling system since this value bounds any normal heat load stored in the fuel pool. If the normal heat load in the fuel pool exceeded 13.2x10 BTU/hr, then modifications to the fuel pool cooling system would be necessary to enable the system to maintain pool temperature less than 1254F.

2)

The time-to-boil analysis for the loss of normal spent fuel pool cooling case should use a range of spent fuel pool decay heat. loads up to at least the design capacity of the fuel pool cooling system.

This method bounds any maximum normal heat load for the fuel pool while limiting overly conservative times in the years while the fuel pool is partially filled.

Basically, this method provides time-to-boil as a function of decay heat load in the spent fuel pool.

This relation can be used for more realistic time-to-boil for current conditions if actual decay heat load in the spent fuel pool is known.

Page 5 of 10 Attachment

Boiling Spent Fuel Paol Issues Second Problem:

The analytical 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil for the spent fuel pool does not account for the emergency heat load in the spent fuel pool.

Histary:

FSAR 9.1.3.1 establishes the emergency heat load for the spent fuel pool as that heat load following a full core offload which completely fills the fuel pool.

The FSAR specifies the emergency heat load to be 32.6x10 BTU/hr.

Bechtel calc 200-0048 Rev.

1, "Boiling Spent Fuel Pool" dated May 7,

1982, determined time-to-boil for the maximum normal heat load case only.

The actual decay heat load in the spent fuel pool exceeds the maximum normal heat load during every refueling outage at SSES in which the core is fully offloaded.

SSES currently imposes administrative controls during refueling outages when the core is fully offloaded into the spent fuel pool to reduce the potential for loss of fuel pool cooling.

Decay heat is removed from the spent fuel pool during these periods by RHR shutdown cooling (when the fuel pool to reactor cavity gates are removed) and by cross-tieing the operating unit's fuel pool cooling system to the outage unit's fuel pool.

However, a

seismic event in this configuration could cause loss of fuel pool cooling at a time when the time-to-boil is significantly less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

Calc M-FPC-009 was drafted to determine the spent fuel pool time-to-boil and required ESW makeup rate for power uprate.

Preliminary results from this calc indicate the fuel pool boils 7.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after loss of fuel pool cooling for a decay heat load of 36.2x106 BTU/hr, which is the currently analyzed emergency heat load.

Recommendation:

The time-to-boil analysis should be expanded to include decay heat loads up to at least the design capacity of the RHR fuel pool cooling assist mode.

Operating procedures, off-normal procedures and SSES outage management policies should be reviewed and revised as necessary to ensure that appropriate controls are implemented when the fuel pool decay heat load exceeds the capacity of the fuel pool cooling system and proper responses are taken in event fuel pool cooling is lost.

Page 6 of 10 Attachment

Bailing Spent Puel Paal Issues Thir4 Problem:

The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates.

History:

Bechtel calc 200-0048 Rev.

1, "Boiling Spent Fuel Pool" dated May 7,

1982, determined the evaporation rate from a boiling spent fuel pool using the equation:

Evap Rate

=

Q / (h hf),

where Q =

fuel pool decay heat load, BTU/hr h

= enthalpy of vapor at boiling, BTU/lb hf = enthalpy of water at boiling, BTU/lb This calc used a

decay heat load of 9.79x106 BTU/hr to determine evaporation rate.

As reported

above, the maximum normal heat load specified 'in FSAR Table 9.1-2a is 12.6x106 BTU/hr and the emergency heat load specified in FSAR 9.1.3.1 is 32.6x106 BTU/hr.

When the decay heat load in the spent fuel pool exceeds 9.79x106 BTU/hr, the evaporation rate from the boiling pool will exceed the rate assumed in the radiological release analysis.

SSES currently applies the 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil determined by calc 200-0048 as the criterion in deciding when to permit common RHR work during an outage.

Therefore, when decay heat loads are less than 9.79x10 BTU/hr, the time-to-boil is longer than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> and the radiological release in event of loss of fuel pool cooling is bounded by the results from calc 200-0048.

The 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> criterion for common RHR work prevents this work from beginning prior to =Day 18-21 each outage.

Since core offloading typically starts on Day 5 and is completed by Day 10 or 11, this means that for at least 7 days, the decay heat load in the spent fuel pool is significantly higher than the heat load used to derive the evaporation rate used in the radiological release analysis.

The consequences from a loss of fuel pool cooling may be offset by a longer time-to-boil if the fuel pool to reactor cavity gate is removed and the fuel pools are cross-tied, but credit for the additional water inventory available cannot be taken without administrative controls and a time-to-boil analysis for this configuration.

Page 7 of 10 Attachment

Boi.ling Spent Fuel Paal Issues Recommendation:

The radiological release analysis should use appropriate evaporation rates for the decay heat loads used in the associated time-to-boil analysis.

The method and results from these analyses should be clearly stated and conveyed to SSES to ensure that adequate administrative controls are implemented during normal operation and in refueling to ensure the radiological release analysis results bound actual plant conditions for all operating configurations.

Fourth Problem:

The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms.

History:

Bechtel calc 200-0048 Rev.

1, "Boiling Spent Fuel Pool" dated May 7, 1982, determined the radiological release consequences from a boiling spent fuel pool.

This calc assumed 12 month operating cycles and 184 bundle equilibrium reload sizes to determine the activity terms for failed fuel in the fuel pool.

SSES currently has 18 month operating cycles with =230 bundle reloads which will increase to

=254 bundles after power uprate.

Since calc 200-0048 implies that most of the activity results from the most recent discharge

batch, the effect of increasing the discharge size from 184 bundles assumed in the calc to 230 and 254 bundles would appear to be nonconservative with respect to the radiological release analysis.

Recommendati.an:

The basis for the radiological release analysis should not be the projected operating conditions, since these conditions are subject to assumptions which are extremely difficult to predict.

The radiological release analysis should assume conditions which will bound future actual operating conditions.

For example, a reload batch size of 320 bundles was assumed in calc NFE-B-NA-053, "Decay Heat from a Full Spent Fuel Pool (ASB9-2 Method)", because this size represents the maximum reload batch size possible under core design criteria.

Page 8 of 10 Attachment

Boiling Spent Puel Pool Issues ZZZ.

ESN MAKEUP TO THE SPENT FUEL POOL Rirst Problem:

The impact of the ESW makeup water to the spent fuel on equipment in the reactor building has not been evaluated.

History:

The ESW system and the ultimate heat sink are designed to provide adequate makeup to the spent fuel pool for 30 days following loss of normal spent fuel pool cooling.

Based on the original ESW makeup flow of 60 gpm to each fuel pool, the spray pond inventory allocates 5 million gallons of water for this purpose.

However, the consequences of this quantity of water on equipment in the reactor building has not been evaluated.

EDR G00005 was written in 1990 to address discrepancies between the spent fuel pool discussion in FSAR Chapter 9 and actual SSES operation.

This EDR also questioned the ESW makeup flow to the spent fuel pool since the 60 gpm flow rate had not been demonstrated to be achievable.

Calc M-FPC-009, "Spent Fuel Pool Boiling Analysis",

was drafted to determine the ESW makeup flow required for the design heat removal capacity of the fuel pool cooling system (13.2x10 BTU/hr) and for the heat removal capacity of the RHR fuel pooi cooling mode (32.6x10 BTU/hr).

These ESW makeup flows were determined to be 31.8 gpm and 67.5 gpm respectively.

The interim disposition to EDR G00005 pointed out that the maximum ESW makeup flow case occurs when all of the reactor core is offloaded to the spent fuel pool, so the higher ESW makeup flow rate could be obtained by the reduced ESW system flow required when there is no fuel in the reactor.

When the ESW makeup flow to the fuel pool exactly matches the boil-off rate from the fuel pool, that quantity of water vapor must also either exit the building via the standby gas treatment

system, bring the building to 1004 humidity or condense somewhere within the reactor building.

When the ESW makeup flow to the fuel pool exceeds the boil-off rate, there will also be overflow once the skimmer surge tank fills and level control is lost.

Eventually in either case, the quantity of water added via the ESW system ends up going thxough the standby gas treatment system or as water in the reactor building.

The consequences of up to 2.5 million gallons of water in each reactor building

Page 9 of 10 Mtachment

Boiling Spent Fuel Pool Issues could include flooding of the ECCS pumps rooms, inoperability of the ECCS pump room coolers, inoperability of safety related equipment due to higher than analyzed humidity and degradation of the standby gas treatment system due to moisture loading.

Recommendation:

A comprehensive evaluation for the boiling spent fuel pool event needs to be performed which accounts for the water present in the reactor building due to boil-off and overflow from the spent fuel pool.

This evaluation must address the effects of this water on the operability of systems and components in the reactor building.

Second Problem:

The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be permitted under post-LOCA conditions.

History:

Off-normal operating procedure ON-135-001, "Loss of Fuel Pool Cooling/Coolant Inventory", requires the operator to manually open the valves in ESW makeup line to fuel pool if all other means of adding water to the fuel pool are lost.

The procedure calls for the valves to be left open until the desired water level is obtained.

Since these valves are in the reactor building, it may be impossible for them to be manually operated as directed under all conditions including post-LOCA.

In addition, even if the throttle valve is initially adjusted so that the ESW makeup flow to the fuel pool exactly matches the boil-off rate, the subsequent exponential decline in fuel pool decay heat load would require the throttle valve to be'periodically adjusted to reduce the ESW makeup flow unless the fuel pool is permitted to overflow.

Recommendation:

The required operation of the ESW makeup flow valves should be evaluated from the perspective of accessibility and usage over the entire 30 day period of the boiling spent fuel pool event to ensure that all necessary valve manipulations can be made.

Page 10 of 10 Attachment

Boiling Spent Fuel Pool Issues Third Problem:

The instrumentation available to the operator post-LOCA may not provide adequate indication of spent fuel pool temperature-and level to allow proper response to a loss of fuel pool cooling event.

History:

Off-normal operating procedure ON-135-001, "Loss of Fuel Pool Cooling/Coolant Inventory", requires the operator to manually open the valves in ESW makeup line to fuel pool if all other means of adding water to the fuel pool are lost.

The operator enters this procedure upon annunciation of low level in the spent fuel pool or high temperature in the fuel pool cooling system.

Each spent fuel pool has temperature indication (TE-15333) and level indication (LT-15332).

Each skimmer surge tank has level indication (LT-15312).

The skimmer surge tank piping to the fuel pool heat exchangers has temperature indication (TE-15313) and each fuel pool heat exchanger outlet piping has temperature indication (TE-15316A,B,C).

The level and temperature instruments providing these alarms may not be qualified for all conditions, such as post-LOCA, in which they would be required to function.

In addition, these instruments may not be powered from class 1E sources such that they would be available post-LOOP when the fuel pool heat exchangers would be without service water.

Recommendation:

The spent fuel pool temperature and level instrumentation, as a minimum, should be verified to be or made to be qualified for all reactor building environmental conditions and required accident conditions.

Attacheent 3

PP8L Engineering Discrepancy

Report, "Loss of Spent Fuel Pool Cooling Event Design Discrepancies",

Originated April 16, 1992 and Dispositioned October 6,

1992 (EDR G20020)

PPgL SUSQUEHANNA STEAM ELECTRIC STATION ENGINEERING DISCREPANCY REPORT FILE: R42-15

1. EpR No. (y QOOat 0

2. REV. No. 0

3. PAGE1 of~

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18 DISCREPANCYREVIEtWCOMPLETE:

SUPERVISING ENGINEER, ENGR PROJECTS Full Signature Date FORM EPM-QA-122A, Rev. 3 Page 1 of 1

~pootjnrenM'onofbeses tequlied.

Use EDR Condncetion Sheet.

PPfiL SUSQUEHRL28iE STEAK ELECTRZC STlLTZOS EHGZHEERZHG DZSCREP2LSCX REPORT PZLE R42 1.5 (cont)nuation sheet) 1.

EDR No.(>>~

2.

REV No.~

3.

Page 2-of 6

9. Potential Engineering Discrepancy (continued) the fuel pool cooling system used for normal operation and the RHR fuel pool cooling assist mode used for abnormal heat loads are not designed to satisfy seismic category I and single failure criteria.

The following discrepancies for the loss of. spent. fuel pool cooling event were discovered during the system evaluations for power uprate:

A.

B.

Reactor building design heat loads do not account for the boiling spent fuel pool event.

The current calculations for reactor building Zone I, II and III heat loads assume a spent fuel pool temperature of 1254F for all cases.

The reactor building heat load analyses and attendant temperature analyses upon which equipment qualification environmental parameters are based do not account for the additional heat load from boiling spent fuel pool(s).

'The additional heat load could be as high as 26.4x106 BTU/hr compared to the current maximum reactor building heat load of 5.5X106 BTU/hr (Unit 1 LOCA case).

Therefore, the design environmental conditions of safety related equipment in the reactor building may be exceeded if the heat load from the boiling spent fuel pool(s) is considered (essed(

beleo).

The impact of the ESW makeup water to the spent fuel pool on equipment in the reactor building has not been evaluated.

The ESW system and the ultimate heat sink are designed to provide adequate makeup to the spent: fuel pool for 30 days following loss of normal spent fuel pool cooling.

Based on the original design ESW makeup flow of 60 gpm to each fuel pool, 5

million gallons of the spray pond inventory is allocated for this purpose.

The water added to the spent fuel pool via the ESW system boils off and exits through the standby gas treatment system or condenses in the reactor building, or the water overflows the pool.

The consequences of up to 2.5 million gallons of water in each reactor building could include flooding of ECCS pump rooms, inoperability of ECCS pump room coolers, emergency switchgear and load center room coolers and/or other safety related equipment due to higher than analyzed temperature and humidity conditions, and degradation of the standby gas treatment system due to moisture loading.

The standby gas treatment system is designed for 1004 relative humidity conditions in the reactor building, but a system design calculation ORM EPM-QA-122B, Rev.

2 age 1 of 1

SUSQU1BGiHHK STEAM ELECTRXC STATZOH EHGZHEERXHG DZSCREPAHCY REPORT PUB: R42-15

<contlamtton sheet) 1.

EDR No.6~o 2.

REV No. C>

3. Page 3 of 6

9. Potential Engineering Discrepancy (continued)

(M-SGT-015) which determined that water buildup in the ductwork before the inlet HEPA filter would not degrade system performance does not consider the potential collapse/failure of the ductwork from the weight of this water.

C.

D.

The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be possible.

The off-normal operating procedure (ON-135-001) requires the operator to manually open the valves in the ESW makeup line to the fuel pool if all other means of adding water to the fuel pool are lost.

The procedure calls for the valves to be left open until the desired water level is obtained.

Since these valves are in the reactor building, it may be impossible for them to be manually operated as directed under all conditions including post-LOCA without unacceptable risk to the operator from the high radiation levels in the building and potentially high temperature and humidity conditions.

The maximum gamma dose rates reported for EQ purposes for Unit 1 reactor building elevations 749'-1" to 818'-1" ranged between 140 and 360 R/hr (C-1815 Sh 7-10).

In

addition, even if the throttle valve is initially adjusted so that the ESW makeup flow to the fuel pool exactly matches the boil-off rate, the subsequent exponential reduction in fuel pool decay heat load would require the throttle valve to be periodically adjusted to lower the ESW makeup flow unless the fuel pool is permitted to overflow.

The instrumentation available to the operator post-LOCA does not provide adequate indication of spent fuel pool temperature and level to allow proper response to a loss of fuel pool cooling event.

The off-normal operating procedure ON-135-001, "Loss of Fuel Pool Cooling/Coolant Inventory", requires the operator to manually open the valves in ESW makeup line to the fuel pool if all other means of adding water are lost.

The operator enters this procedure upon annunciation of low level in the spent fuel pool or high temperature in the fuel pool cooling system.

Each spent fuel pool has temperature indication (TE-15333) and level indication (LT-15332).

Each fuel pool skimmer surge tank has level indication (LT-15312).

The skimmer surge tank piping to the fuel pool heat exchangers has temperature indication FORM EPM-QA-122B, Rev.

2 Page 1 of 1

PPfcL SUSQUEHAHHK STEAK ELECTRIC STATIOH EHGZHEERZHG DZSCREP2LHCY REPORT FILE: R42-I.5

<cene)nuatien sheec) 1.

EDR No. s>~~

2 ~

REV No. O 3.

Page 4 of 6

9. Potential Engineering Discrepancy (continued)

(TE-15313) and each fuel pool heat exchanger has temperature indication (TE-15316A,B,C) in its outlet piping.

The level and temperature instruments providing these alarms may not be qualified for the temperature and humidity conditions, such as post-LOCA, in which they would be required to function.

In addition, these instruments are not powered from class 1E sources such that they would he available post-LOOP when the fuel pool heat exchangers would be without service water.

E. The analytical 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil for the spent fuel pool is nonconservative for the maximum normal heat load in the spent fuel pool.

The original design calculation (200-0048) used a decay heat load of 9.79x10 BTU/hr for Unit 1 and 7.92x10 BTU/hr for Unit 2 to determine times-to-boil of 25.15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 31.087 hours0.00101 days <br />0.0242 hours <br />1.438492e-4 weeks <br />3.31035e-5 months <br /> respectively.

FSAR 9.1.3.1 establishes the maximum normal heat load as that heat load resulting from 2840 assemblies discharged to the fuel pool by a routine refueling schedule.

FSAR Tables 9.1-2a and 9.1-2b report the maximum normal heat load for Units 1 and 2 as 12.6x10 BTU/hr.

The spent fuel pool decay heat values used in the boiling spent fuel pool calculation and for the FSAR discussion were based upon assumptions for cycle operating lengths, fuel exposures, and reactor power level.

SSES has subsequently operated differently than had been assumed such that the decay heat loads in the filled spent fuel pools may exceed 9.79x106 BTU/hr, resulting in a shorter than analyzed time-to-boil.

A recent calculation prepared for power uprate (NFE-B-NA-053) determined decay heat from a filled spent fuel pool using actual fuel operating history through 1991 and assumptions which bound operation after power uprate.

This calculation reported a maximum normal heat load of =17x106 BTU/hr.

Another recent calculation (M-FPC-009) determined the spent fuel pool time-to-boil and required ESW makeup rate for power uprate.

Preliminary results from this calculation indicate the fuel pool boils 19.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> FORM EPM-QA-122B, Rev.

2 Page 1 oi'

PPfcL SUSQUNGLH2GL STEER ELECTRZC ST2LTIOH EHGZHEERZHG DZSCREPAECY REPORT PILE: R42-1.5

<conetnuac<on sceeen) 1 EDR No.bd~>

2.

REV No.O 3.

Page 5 of 6

9. Potential Engineering Discrepancy (continued) after loss of fuel pool cooling for the design heat removal capacity of the fuel pool cooling system (13.2x106 BTU/hr).

This calc determined a time-to-boil of =15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the =17x106 BTU/hr heat load calculated. for the power uprate case.

F.

The analytical 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil for the spent fuel pool does not'ccount for the emergency heat load in the spent fuel pool.

FSAR 9.1.3.1 establishes the emergency heat load for the spent fuel pool as that heat load following a full core offload which completely fills the fuel pool.

The FSAR specifies the emergency heat load to be 32.6x106 BTU/hr.

The original design calculation (200-0048) determined time-to-boil for the maximum normal heat load case only.

The actual decay heat load in the spent fuel pool exceeds the maximum normal heat load during every refueling outage at SSES in which the core is fully offloaded.

I SSES currently imposes administrative controls during refueling outages when the core is fully offloaded into the spent fuel pool to reduce the potential for loss of fuel pool cooling.

Decay heat is removed from the spent fuel pool during these periods by RHR shutdown cooling (when the fuel pool to reactor cavity gates are removed) and by cross-tieing the operating unit's fuel pool cooling system to the outage unit's fuel pool.

However, a seismic event in this configuration could cause loss of fuel pool cooling at a time when the time-to-boil is significantly less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, which is not reflected in the off-normal operating procedure (ON-135-001).

Another recent calculation (M-FPC-009) determined the spent fuel pool time-to-boil and required ESW makeup rate for power uprate.

Preliminary results from this calculation indicate the fuel pool boils 7.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after loss of fuel pool cooling for a decay heat load of 36.2x106 BTU/hr, which is the currently analyzed emergency heat load.

FORM EPM QA 122Bg Revs 2

Page 1 of 1

PPfcL SUSQU&DQDGL STEAK ELECTRZC STATZOH EHGZHEERXHG DZSCREPASCT REPORT PILE: R42-15

<conerwac<on eeee) 1.

EDR No. 6-gc03G 2.

REV No.O 3.

Page 6 of 6

9. Potential Engineering Discrepancy (continued)

The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates.

The original design calculation (200-0048) used a

decay heat load of 9.79x10 BTU/hr to determine evaporation rate.

As reported

above, the maximum normal heat load specified in FSAR Table 9.1-2a is 12.6x106 BTU/hr and the emergency heat load specified in FSAR 9.1.3.1 is 32.6x10 BTU/hr.

When the decay heat load in the spent fuel pool exceeds 9.79x106 BTU/hr, the evaporation rate from the boiling pool will exceed the rate assumed in the radiological release analysis.

H. The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms.

The original design calculation (200-0048) assumed 12 month,operating cycles and 184 bundle equilibrium reload sizes to determine the activity terms for failed fuel in the fuel pool.

SSES currently has 18 month operating cycles with =230 bundle reloads which will increase to ~254 bundles after power uprate.

Since the calculation implied that most of the activity results from the most recent discharge

batch, the effect of increasing the discharge size from 184 bundles assumed in the calc to 230 and 254 bundles would appear to be nonconservative with respect to the radiological release analysis.

The analysis for maximum time prior to makeup to a boiling spent fuel pool is based upon nonconservative assumptions.

The original design calculation (175-

14) determined the time using evaporation of the entire fuel pool water inventory.

The maximum time should be based upon a minimum fuel pool water level which is sufficiently above the top of the fuel to provide the shielding required to allow corrective operator actions.

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Page 1 of 1

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies A.

Reactor Building Design Heat Loads Do Not Account for the Boiling 8pent Fuel Pool Event RE{}UIREMENT:

10 CFR 50.49 requires that electrical equipment must be qualified to the temperature "for the most severe design basis accidents."

10 CFR 50 Appendix A General Design Criterion 4 states that "structures,

systems, and components impo'rtant to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance,

testing, and postulated accidents, includi ng loss-of-'coolant accidents. "

CONCERN:

Secondary containment design analyses are required to account for all heat loads in the reactor building including from the boiling spent fuel pool.

The existing design reactor building heat load calcs consider sensible heat from the b'oiling pool, but neglect latent heat.

These calcs indicate little margin to design temperatures in many rooms for a maximum heat load in the reactor building of approximately 5.5x10'TU/hr.

The total design heat load from the spent fuel pools is 26.4x10 BTU/hr, which would add at least approximately 20.9x10 BTU/hr to the existing maximum heat load.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the boiling spent fuel pool event has not been fully considered in reactor building heat load calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to room temperatures in the reactor building exceeding design EQ values.

June 22, 1992 Page 2

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies B.

The Impact of the EBH Makeup Hater to the Spent Fuel Pool on Equipment in the Reactor Building Has Not Been Evaluated REQUIREMENTB'0 CFR 50 Appendix A General Design Criterion 4 states that "structures,

systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance,

testing, and postulated 'ccidents, including loss-of-coolant accidents. "

Standard Review Plan (NUREG-0800) 3.4.1 states that the review of "plant flood protection includes all structures, systems and components (SSC) whose failure could prevent safe shutdown of the plant or result on uncontrolled release of significant radioactivity..." and that this review "also includes consideration of flooding from internal sources."

CONCERNS FSAR 9.1.3.3 states the design ESW makeup function "is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days following the loss of the FPCCS capacity."

The ultimate heat sink and ESW are designed to provide 1.5 million gallons of water to each fuel pool over the 30 day period.

In the LOCA-LOOP condition, the reactor building HVAC system in Zone I; II and IIIisolation mode recirculates refueling floor air throughout all three zones.

The water added to the fuel pools ends up in the reactor building following boil-off and.

overflow.

,The effects of this water on the structures, systems and components in the reactor buildings have not been included in design analyses.

The potenti'al for common mode failures of multiple ECCS and safety-related systems such as the standby gas treatment system exists.

June 22, 1992 Page 3

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the boiling spent fuel pool event has not been fully considered in EQ and flooding effects calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to common mode equipment failures due to water/humidity.

June 22, 1992 Page 4

e

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies C.

The Manual Valve Manipulations Required to Provide ESR Makeup Plow to a Boiling Spent Fuel Pool May Not Be Possible REQUIREMENTS:

10 CFR 20.1 requires licensees to "make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to unrestricted areas, as low as is reasonably achievable."

10 CFR 50 Appendix A General Design Criterion 19 requires suitable design features to limit control room radiation exposure to 5 rem.

GDC 19 also requires design features for equipment outside the control room to permit operation in accordance with suitable procedures.

10 CFR 50.47(b)(11) states that licensees assure that "means for controlling radi ologi cal exposures, in an emergency, are established for emergency workers.

The means for controlling radiologi cal exposures shall include exposure guidelines consistent with EPA Emergency 8'orker and Lifesaving Activity Protective Action Guides."

NDI-6.4.3 specifies that the whole body dose for life saving actions "shall not exceed 75 rem" and the whole body dose for entry into a hazardous area to protect facilities or equipment "shall not exceed 25 rem."

CONCERN:

The ESW system is required to provide makeup to the pools following loss of fuel pool cooling.

Either a seismic event

.or loss of offsite power can lead to loss of fuel pool cooling.

Both conditions are assumed to occur concurrent with a

LOCA in the DBA for containment analyses.

However, the post-LOCA design EQ dose rates in the reactor building areas where the manual valves are located are 140-360 R/hr and will prevent these valves from being accessed without excessive radiation exposure to the operator.

In addition, the reactor building temperature, humidity and emergency lighting conditions would not be conducive to the location and manipulation of manual valves which are used infrequently.

10 CFR 20's ALARA provision requires plant design to minimize radiation exposure.

Application of the emergency dose guidelines to this manual valve operation is contrary to the intent of 10 CFR 20.1 and 10 CFR 50 App A GDC 19.

June 22, 1992 Page 5

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the boiling spent fuel pool event analysis depends on makeup from ESW to prevent uncovering irradiated fuel and subsequent.

fuel damage from overheating, and

4) the potential consequences from the boiling spent fuel pool event willsignificantly increase if adequate makeup cannot be established, or

5) personnel will receive unnecessary radiation exposures which exceed 10 CFR 20.1/GDC 19 requirements and probably exceed 10 CFR 50.47 guidelines in order to align the makeup path.

June 22, 1992 Page 6

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies D.

The Znstrumentation Available to the Operator Post-LOCA Does Not Provide Adequate Zndication of Spent Fuel Pool Temperature and Level to Allow Proper Response to a Loss of Fuel Pool Cooling Event REQUZREMENTS~

10 CFR 50 Appendix A General Design Criterion 63 states that "appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (l) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions. "

Regulatory Guide 1.97 defines accident-monitoring instrumentation to include "those vari abl es to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events. "

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design

'ncludes "the instrumentation provided for initiating appropriate safety actions."

Standard Review Plan (NUREG-0800) 7.1 states that "information systems important to safety include those systems which provide information for manual initiation and control of safety

systems, to indicate that plant safety functions are being accomplished, and to provide information from which appropriate actions can be taken to mitigate the consequences of anticipated operational occurrences and accidents."

CONCERNS The ESW system is required to provide makeup to the pools following loss of fuel pool cooling.

A loss of offsite power can result in loss of fuel pool cooling.

The loss of offsite power, will also disable the fuel pool temperature and level.

instruments monitored by the operator and used to initiate the safety action of providing ESW makeup to the boiling spent fuel pool.

June 22, 1992 Page 7

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the boiling spent fuel pool event analysis depends on makeup from ESW to prevent uncovering irradiated fuel and subsequent fuel damage from overheating, and

4) the potential consequences from the boiling spent fuel pool event willsignificantly increase if adequate makeup cannot be established and lack of monitoring could prevent the required safety action from being initiated properly.

0 June 22, 1992 Fage 8

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies E.

The Analytical 25 Hour Time-to-Boil for the Spent Fuel Pool is Nonconservative for the Maximum Normal Heat Load in the Spent Fuel Pool REQUZREMENTS:

FSAR Appendix 9A states "conservative results showed that t'e pools would not boil until at least 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after the loss of cooling."

FSAR Table 9A-2 states the total decay heat loads in the Unit 1 and Unit 2 fuel pools assumed in the loss of spent fuel pool cooling analysis are "9.79" and "7.92" BTU/hr x 106 CONCERNS The maximum normal heat load in the spent fuel pool is presently higher than 9.79x10~ BTU/hr and will also increase as a result of power uprate.

The fuel pool will boil in less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> for any fuel pool heat load greater than 9.79x10'TU/hr.

(See Figure 1 from Calc M-FPC-009 attached).

This concern does not affect the present operation of SSES because the existing decay heat loads in the fuel pools are less than 9.79x10~ BTU/hr.

NOTE!

The original determination of maximum normal heat load relied on assumed reactor operating parameters such as fuel type, fuel discharge average

exposure, and operating cycle length.

These parameters have changed since the original calculation and will probably continue to change as fuel design and fuel management evolves.

An approach to bound all such variables would consider the maximum normal heat load in the spent fuel pool to be equal to the design capacity of the fuel pool cooling system (13.2x106 BTU/hr).

This approach would bound all heat loads capable of being handled by the fuel pool cooling system without depending upon predictions of fuel and core designs.

June 22, 1992 Page 9

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies F.

The Analytical 25 Hour Time-to<<Boil for the Spent Fuel Pool Does Not Account for the Emergency Heat Load in the Spent Fuel Pool REQUIREMENTS'tandard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design are reviewed to determine that "a seismic Category, I makeup system and an appropriate backup method to add coolant to the spent fuel pool are provided" and that "engineering judgement...

used to determine that the makeup capacities and the time required to make associated hookups are consistent with heatup times or expected leakage."

SSES Safety Evaluation Report (NUREG-0776) 9.1.3 states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss of spent fuel pool cooling accidents."

FSAR 9.1.3.1 states that "during an emergency heat load (EHL) condi tion, one RHR pump and heat exchanger are available for fuel pool cooling."

CONCERN.

The emergency heat load condition requires an RHR loop to remove decay heat from the spent fuel pool.

A single failure of the valve in the RHR line to the fuel pool, even without a concurrent seismic event or loss of offsite

power, could initiate a loss of fuel pool cooling in which the time-to-boil would be significantly, less than the 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> assumed in the radiological release analysis and in plant operating procedures.

This potential exists presently during every refueling outage when the full core is offloaded to the spent fuel pool.

This concern affects the present operation of SSES during

refueling outages because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools will boil in as little as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following loss of fuel pool cooling with the existing decay heat loads in the pools during refueling, and

3) the spent fuel pool boiling analysis assumes a minimum time to boil of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

June 22, 1992 page 10

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies G.

The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconservative Evaporation Rates REQUIREMENT:

FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

CONCERN The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the evaporation rate from the pools based upon maximum normal heat loads of 9.79 and 7.92 x 10'TU/hr.

As discussed in Item (E) above, the present maximum normal heat load exceeds 9.79x10~

BTU/hr and will increase after power uprate.

Therefore, the actual rate at which water evaporates from the boiling spent fuel pool is higher than analyzed which introduces nonconservatism into the offsite dose calculation.

This concern does not affect the present operation of SSES (except during refueling outages as noted in Item F above) because the'xisting decay heat loads in the fuel pools are less than 9.79x10'TU/hr.

June 22, 1992 Page 11

EDR 620020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies H.

The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconservative Activity Terms RE{}UIREMENT:

FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

CONCERN+

The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the source terms in the spent fuel pools based upon assumptions for fuel design and cycle operation.

SSES has been operated with different fuel types and longer cycles than assumed in the analysis which introduces nonconservatism into the offsite dose calculation.

In addition, the conclusions reported in FSAR Appendix 9A regarding the thyroid doses from FSAR Table 9A-1 are not valid for all cases.

FSAR Table 9A-1 only addresses offsite doses from activity released from the two boiling spent fuel pools.

Since the boiling spent fuel pools can occur as a result of the LOCA-LOOP with SSE DBA, these thyroid doses should be added to the doses resulting from the LOCA.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the potential consequences from the boiling spent fuel pool event may significantly increase due to higher source term activity associated with 9x9 fuel, larger discharge batch sizes, and higher bundle exposures, and

4) the offsite dose resulting from the boiling spent fuel pool is not considered in the total offsite dose resulting from the DBA LOCA-LOOP.

June 22, 1992 Page 12

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies I.

The Analysis for Maximum Time Prior to Makeup to a Boiling Spent Fuel Pool is Based Upon Nonconservative Assumptions REQUIREMENT:

Calc 175-14 determined the maximum time available before makeup to a boiling spent fuel pool is required.

CONCERN The time determined by this design calculation is based upon how long it would take to completely evaporate the entire spent fuel pool water inventory.

Allowing the entire spent fuel pool to evaporate prior to makeup would have severe and unanalyzed consequences:

a) reactor building radiation doses would significantly

increase, b) offsite radiological doses would significantly increase due to skyshine, and c) fuel integrity of the irradiated fuel would be challenged as it was uncovered.

This concern does not appear to affect the present operation of SSES because no document or procedure is known to use the results of this calc.

However, an exhaustive search was not performed.

June 22, 1992 Page 13

Figure 1 Time to Boil vs. FPC Heat Load FPCCS Design Capacity g

~pE '

O 3

E Analytical Limit 10 15 20 25 30 Fuel Pool Cooling Heat Load, MBTU/hr 35 40

125 F Pool Temp 110 F Pool Temp

Attachaent 6

PP&L Draft Screening Morksheet prepared by Art White, "EDR No.

G20020", July 1,

1992,

~

Note:

The first nine text pages of this dr,aft evaluation of

~ EDR G20020 prepared by an engineer within the PP&L Engineering Discrepancy Management Group were taken almost verbatim 'rom the authors'emo dated June 22, 1992 (Attachment 5).

The final three pages of 'analysis'or EDR G20020 provide ample evidence of PP&L's reliance upon probability arguments, use of realistic instead of design conditions, and oversimplification of issues while assessing the safety significance of concerns.

SCREENING WORKSHEET UNIT 1

& 2 EDR No.G20020 SUSQUEHANNA STEAM ELECTRIC STATION PENNSYLVANIA POWER

& LIGHT COMPANY PREPARED BY DATE REVIEWED BY DATE

DISC ANC LOCAT ON

SUBJECT:

Loss of Spent Fuel Cooling Event Design Discrepancies DESCR PTION OF CONDI ON The regulatory requirements for cooling the spent fuel pool are based upon:

10 CFR Appendix A Design Criterion 61 which states that the fuel storage system shall be designed

" to prevent significant reduction in fuel storage coolant inventory under accident conditions" and Standard Review Plan (NUREG-800) 9.1.3 for the spent fuel pool cooling and cleanup system which states that the "safety function to be performed by the system in all cases remains the same; that is, the spent assemblies must be cooled and must remain covered with water during all storage conditions.

The SSES design utilizes non-seismic, non-Class IE powered fuel pool cooling and cleanup systems for cooling the fuel pools.

In the event of a

loss of spent fuel pool

cooling, the design provision at SSES is to allow the fuel pools to boil with adequate makeup provided to maintain the water level in the pools above the fuel.

The SSES design requirements are based upon:

FSAR Appendix 9A which states that it is assumed "a seismic event causes the loss of cooling to both spent fuel pools" and that "if cooling is not restored before the pool boils, then makeup water from the Category I Emergency Service Water System can be added to the pool to keep the fuel covered at all times," and FSAR 6.2.1.1.1(a) states that "The LOCA scenario used for containment functional design includes the worst single failure (which leads to maximum coincident containment pressure and temperature),

postulated to occur simultaneously with loss of offsite power and a safe shutdown earthquake (SSE)."

Since an analyzed design basis accident (DBA) at SSES is a

LOCA with a concurrent LOOP and

SSE, and either a seismic event or a loss of offsite power will result in a loss of spent fuel pool cooling, the consequences of this DBA include boiling spent fuel pools.

The SSES design was (NUREG-0776) 9.1.3 which states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss os spent fuel pool cooling accidents."

The following design discrepancies for the loss of spent fuel pool event:

A. Reactor Building Design Heat Loads Do Not Account for the Boiling Spent Fuel Pool Event Requirement:

10 CFR 50.49 requires that electrical equipment must be qualified to the temperature "for the most severe design basis accidents."

10 CFR 50 Appendix A General Design Criterion 4 states that "structures,

systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance,

testing, and postulated accidents, including loss-of-coolant accidents."

Concern:

Secondary containment design analyses are required to account for all heat loads in the reactor building including from the heat load calcs consider sensible heat from the boiling pool, but neglect latent heat.

These calcs indicate little margin to design temperatures in many rooms for a maximum heat load in the reactor building of approximately 5.5E6 BTU/hr.

The total design heat load from the spent fuel pools is 26.4E6 BTU/hr,. which would add at least approximately 20.9E6 BTU/hr to the existing maximum heat load.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event, 2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads, 3) the boiling spent fuel pool event has not been fully considered in reactor building heat load calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to room temperatures in the reactor building exceeding design EQ values.

B.

The Impact of the ESW Makeup Water to the Spent Fuel Pool on Equipment in the Reactor Building has not been evaluated Requirements:

10 CFR 50 Appendix A General Design Criterion 4

states that "structures,

systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance,

testing, and postulated accidents, including loss-of-coolant accidents."

Standard Review Plan (NUREG-0800) 3.4.1 states that the

review of "plant flood protection includes all structures, systems and components (SSC) whose failure could prevent safe shutdown of the plant or result in uncontrolled release of significant radioactivity...."

and that this review "also includes consideration of flooding from internal sources."

Concern:

FSAR 9.1.3.3 states the design ESW makeup function "is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days following the loss of the FPCCS capacity.>>

The ultimate heat sink and ESW are designed to provide 1.5 million gallons of water to each fuel pool over the 30 day period.

In the LOCA-LOOP condition, the reactor building HVAC system in Zone I, II and III isolation mode recirculates refueling floor air throughout all three zones.

The water added to the fuel pools ends up in the reactor building following boil-off and overflow.

The effects of this water on the structures, systems and components in the reactor buildings have not been included in design analyses.

The potential for common mode failures of multiple ECCS and safety-related systems such as the standby gas treatment system exists.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a, current design bases

event, 2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads, 3) the boiling spent fuel pool event has not been fully considered in EQ and flooding effects calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to -'common mode equipment failures due to water/humidity.

C.

The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be possible.

Requirements:

10 CFR 20.1 requires licensees to "make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to unrestricted areas, as low as is reasonably achievable.<<

10CFR 50 Appendix A General Design Criterion 19 requires suitable design features to limit control room radiation exposure to 5 rem.

GDC 19 also requires design features for equipment outside the control room to permit operation in accordance with suitable procedures.

10 CFR 50.47 (b)

(11) states that licensees assure that

"means for controlling radiological exposures, in an emergency, are established for emergency workers.

The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving ActivityProtective Action Guides.

~'DI-6.4.3 specifies that the whole body dose for life saving actions "shall not exceed 75 rem" and the whole body dose for entry into a hazardous area to protect facilities or equipment "shall not exceed 25 rem."

Concern:

The ESW system is required to provide makeup to the pools following loss of fuel pool cooling.

Either a seismic event or loss of offsite power can lead to loss of fuel pool cooling.

Both conditions are assumed to occur concurrent with a LOCA in the DBA for containment analyses.

However, the post-LOCA design EQ dose rates in the reactor building areas where the manual valves are located are 140-360R/hr and will prevent these valves from being accessed without excessive radiation exposure to the operator.

In addition, the reactor building temperature, humidity and emergency lighting conditions would not be conducive to the location and manipulation of manual valves which are used infrequently.

10CFR 20's ALARA provision requires plant design to minimize radiation exposure.

Application of the emergency dose guidelines to this manual valve operation is contrary to the intent of 10 CFR 20.1 and 10 CFR 50 App A GDC 19.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event, 2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads, 3) the boiling spent fuel pool event analysis depends on makeup from ESW to prevent uncovering irradiated fuel and subsequent fuel damage from overheating, and

4) the potential consequences from the boiling spent fuel pool event willsignificantly increase if adequate makeup cannot be established, or 5) personnel will receive unnecessary radiation exposures which exceed 10 CFR 20.1/GDC 19 requirements and probably exceed 10 CFR 50.47 guidelines in order to align the makeup path.

D. The instrumentation available to the Operator Post-LOCA does not provide adequate indication of spent fuel pool temperature and level to allow proper response to a loss of fuel pool cooling event

Requirements:

10 CFR Appendix A General Design Criterion 63 states that "appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions."

Regulatory Guide 1.97 defines accident-monitoring instrumentation to include "those variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is pmvided and that are required for safety systems to accomplish their safety function for design basis accident events."

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design includes "the instrumentation provided for initiating appropriate safety actions."

Standard Review Plan (NUREG-0800) 7.1 states that "information systems important to safety include those systems which provide information for actual initiation and control of safety

systems, to indicate that plant safety functions are being accomplished, and to provide information from which appropriate actions can be taken to mitigate the consequences of anticipated operational occurrences and accidents."

Concern:

The ESW system is required to provide makeup to the pools following loss of fuel pool cooling.

A loss of offsite power can result in loss of fuel pool cooling.

The loss of offsite power will also disable the fuel pool temperature and level instruments monitored by the operator and used to initiate the safety action of providing ESW makeup to the boiling spent fuel pool.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event, 2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads, 3) the boiling spent fuel pool event analysis depends on makeup from ESW to prevent uncovering irradiated fuel and subsequent fuel damage from overheating, and

4) the potential consequences from the boiling spent, fuel pool event willsignificantly increase if adequate makeup cannot be established and lack of monitoring could prevent the required

safety action from being initiated properly.

E. The analytical 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil for the spent fuel pool is nonconservative for the maximum normal heat load in the spent fuel pool+

Requirements:

FSAR Appendix 9A states "conservative results showed that the pools would not boil until at least 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after the loss of cooling."

FSAR'able 9A-2 states the total decay heat loads in the Unit 1 and Unit 2 fuel pools assumed in the loss of spent fuel pool cooling analysis are " 9.79E6 BTU/hr and 7.92E6 BTU/hr.

Concern:

The maximum normal heat load in the spent fuel pool is presently higher than 9.79E6 BTU/hr and will also increase as a

result of power uprate.

The fuel pool will boil in less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> for any fuel pool heat load greater than 9.79E6 BTU/hr.

(See Figure 1 from, Calc M-FPC-009 attached).

This concern does not affect the present operation of SSES because the existing decay heat loads in the fuel pools are less than 9 '9E6 BTU/br'ote:

The original determination of maximum normal heat load relied on assumed reactor operating parameters such as fuel type, fuel discharge average

exposure, and operating cycle length.

These parameters have changed since the original calculation and will probably continue to change as fuel design and fuel management evolves.

An approach to bound all.such variables would consider the maximum normal heat load in the spent fuel pool to be equal to the design capacity of the fuel pool cooling system (13.2E6 BTU/hr).

This approach would bound all heat loads capable of being handled by the fuel pool cooling system without depending upon predictions of fuel and core designs.

F. The analytical 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time-to-boil for the spent fuel pool does not account for the emergency heat load in the spent fuel pool.

Requirements:

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design are reviewed to determine that " a seismic Category I makeup system and an appropriate backup method to add coolant to the spent fuel pool are provided" and that "engineering judgement.

.used to determine that the makeup capacities and the time required to make associated hookups are consistent with heatup times or expected leakage."

SSES Safety Evaluation Report (NUREG-0776) 9.1.3 states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss of spent.

fuel pool cooling

accidents."

FSAR 9.1.3.1 states that "during an emergency heat load (EHL) condition, one RHR pump and heat exchanger are available for fuel pool cooling."

Concern:

The emergency heat load condition requires an RHR loop to remove decay heat from the spent fuel pool.

A single failure of the valve in the RHR line to the fuel pool, even without a

concurrent seismic event or loss of offsite power, could initiate a loss of fuel pool cooling in which the time-to-boil would be significantly less than the 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> assumed in the radiological release analysis and in plant operating procedures.

This potential exists presently during every refueling outage when the full core is offloaded to the spent fuel pool.

This concern affects the present operation of SSES during refueling outages because:

1) the boiling spent fuel pool is a current design bases

event,

2) the fuel pools willboil in as little as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following loss of fuel pool cooling with the existing decay heat loads in the pools during refueling, and

3) the spent fuel pool boiling analysis assumes a minimum time to boil of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

G. The Radiological Release analysis for a boiling spent fuel pool uses nonconservative evaporation rates Requirement:

FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

Concern:

The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the evaporation rate from the pools based upon maximum normal heat loads of 9.79E6 BTU/hr and will increase after power uprate.

Therefore, the actual rate at which water evaporates from the boiling spent fuel pool is higher than analyzed which introduces nonconservatism into the offsite dose calculation.

This concern does not affect the present operation of SSES (except during refueling outages as noted in Item F above) because the existing decay heat loads in the fuel pools are less than 9.79E6 BTU/hr.

H. The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms.

Requirement:

FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

Concern:

The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the source terms in the spent fuel pools based upon assumptions for fuel design and cycle operation.

SSES has been operated with different fuel types and longer cycles than assumed in the analysis

,which introduces nonconservatism into the offsite dose calculation.

In addition, the conclusions reported in FSAR Appendix 9A regarding the thyroid doses from FSAR Table 9A-1 are not valid for all cases.

FSAR Table 9A-1 only addresses offsite doses from activity released from the two boiling spent fuel pools.

Since the boiling spent fuel pools can occur as a result of the LOCA-LOOP with SSE

DBA, these thyroid doses should be added to the doses resulting from the LOCA.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases

event, 2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the potential consequences from the boiling spent fuel pool event may significantly increase due to higher source term activity associated with 9X9 fuel, larger discharge batch

sizes, and higher bundle exposures, and

4) the offsite dose resulting from the boiling spent fuel pool is not considered in the total offsite dose resulting from the DBA LOCA-LOOP.

I. The analysis for maximum'ime prior to makeup to a boiling spent fuel pool is based upon nonconservative assumptions Requirement:

Calc 175-14 determined the maximum time available before makeup to a boiling spent fuel pool is required.

Concern:

The time determined by this design calculation is based upon how long it would take to completely evaporate the entire spent fuel pool water inventory.

Allowing the entire spent fuel pool to evaporate prior to makeup would have severe and unanalyzed consequences:

, 1) reactor building radiation doses would significantly

increase,

2) offsite radiological doses would significantly increase due to skyshine, and

3) fuel integrity of the irradiated fuel would be challenged as it was uncovered.

This conoern does not appear to affect the present operation of SSES because no document or procedure is known to use the results of this calo.

However, an exhaustive search was not performed.

ANALYSIS I.

DOES E

ENGINEERING DISCREPANCY APPEAR TO CREATE CALCULATED ACCIDENT SE UENCE FRE UENCY?

HIGH BASIS:

No, the postulated concern is based on postulating a

'BA

LOCA, a

LOOP and an SSE all simultaneously.

The probability of such an event approaches zero, it is so vanishingly small.

II.

0 THE E G NEER G DISCREPANC APPEAR 0

A T

D INED G

"DEFENSE-I -DE TH" AGA NST AN ACC DENT S

U NCE E

U PME T OR PROCEDURE RE ED?

BASIS: No, the postulated concern takes no credit for manual action.

From a realistic point-of-view, there is no basis to assume that fuel damage will occur to the extent that manual actions can be taken to line up ESW and RHR in the spent fuel area.

III. DO ENG EER NG DIS EPANC APP TO ADV SE I

SYS EM OR CO PONE EXPLICITLY LISTED N

T E TECHNICAL SP C

CATIONS?

YES NO X TECHNICAL SPECIFICATION SECTION(S)

BASIS: This discrepancy has no basis in fact, and takes no credit for expected operator action.

10

ZV.

DOES THE DZSCR PANCY APPEAR TO COMPROMISE THE CAPABI ITY OF A SYS EM OR COMPONENT TO PERFORM ITS SAFETY RELATED FUNCT ON AS DESC IBED IN THE SAFETY ANALYSIS REPORT?

YES NO X

SAR SECTION(S)

BASIS:

This function as EDR's basis philosophies concern has no effect on any safety related described in the Safety Analysis Report.

The appears to be an invalid application of design to (realistic),post accident manual actions.

V.

DOES THE DISCREPANCY APPEAR 0 ADVERSELY IMPAC Y APPLICABLE CE SING COMMITMENTS?

YES NO X

REFERENCE BASIS:

'The discrepancy does not appear to adversely impact any applicable licensing commitments.

In fact, the SER specifically addresses the Spent Fuel Cooling

Function, and its ESW makeup and RHR cooling function.

VZ.

SAFETY SIGNIFICANCE ASSESSMENT Address lant-s ecific features which affect the safet si 'ca ce o t e concern.

Provide a rea 'stic assessme t of the actua sa et conse uences and im l'catio s o

the

~conce SAFETY SIGNIFICANCE

SUMMARY

NONE MINIMAL MODERATE CONSIDERABLE BASIS: There is no safety significance to this EDR since it has no basis in fact if one does not accept the premise that no action can or will be taken by operations personnel to stop a boiling pool from boiling, or to inhibit it from boiling in the first place.

Basically, it is a misapplication of plant design parameters, such as postulated fuel melt, to post accident operator actions.

11

• This is an i itial assess e t.

The screening function is to be considered a continuous process.

A re-evaluation of the screening status (not necessarily

formal, except when

,determined to be "significant")

should take place by referencing this procedure at each stage of EDR processing (e.g.

EDR implementation) to determine if the issue is now a "safety concern" and is subject to Reportability and/or Operability determinations.

12

AttachIIent 7

Handout, "EDR G20020 References",

July 15, 1992 Note:

This handout was prepared by the authors and distributed during a meeting on July 15, 1992 to discuss EDR G20020.

The handout summarizes the documents researched by the authors while preparing the EDR and subsequently defending its merits.

eI EDR G20020 References Design Bases and Related Issues FSAR 6.2.1.1. 1 states that the "LOCA scenario used for containment functional design includes the vorst single failure (vhich leads to maximum coincident containment pressure and temperature),

postulated to occur simultaneously vith loss of offsite pover and a safe shutdown earthquake (SSE)."

FSAR Appendix 9A states that "it vas assumed that a seismic event causes the loss of cooling to both spent fuel pools."

FSAR Appendix 9A states that "ifcooling is not restored before the pool boils, then makeup water from the Category I Emergency Service Water System can be added to the pool. to keep the fuel covered at all times."

I SSES Safety Evaluation Report 9.1.3 states that "makeup from the seismic Category I emergency service water systems would keep the fuel covered during loss of spent fuel pool cooling accidents."

Becht'el Spec M-192 for the High Density Spent Fuel Storage Racks (June 1977) states that the "seller shall perform analysis to determine the makeup flow rate required to maintain the pool water level under conditions of maximum heat load, none of the cooling systems available and pool water boiling."

Letter PLI-7457 from A. M. Male to R. J. Shovlin (July 1979) states that "the spent fuel pool cooling system is designed to maintain temperature at or below 1254F, The system is further backed up by the Seismic Category I Appendix B qualified emergency systems which have sufficient capacity to, handle this load.

If all of these redundant systems are somehov unavailable, it will still take more than one day before boiling begins.

This is more than sufficient time for onsi te personnel to provide from many alternate vater sources enough make up water to keep the pool from boiling."

Technical Report NPE-84-002 (December 1983) states that "SSES is designed to accept and mitigate a loss of coolant accident (LOCA) concurrent with a complete loss of offsite power (LOOP)" and "the assumption vas made, in the design of SSES, that the LOCA and LOOP would occur simul taneously, and the simultaneous occurrence of LOCA and LOOP becomes the design basis event."

July 15, 1992 Page 1

EDR G20020 References EWR MIS 86-0637 determined that the RHR fuel pool cooling assist mode lines could be deleted from the ISI program since the "present design uses the ESP makeup line as the ultimate heat removal source" with this source being "sufficient to cover the maxi mum boiloff of a full core offload."

EWR MIS 85-0740 stated that "the RHR, assist mode to fuel pool cooling is a non-safety function and therefore may be deleted from the ISI program boundaries" and this mode is "non-safety and adequate cooling is still available form boiling and ESW makeup."

NSAG 4-90 (September 1990) reported that in the RHR system design "the fuel pool cooling assist and the shutdown cooling modes share a

common suction line.

Therefore, the system can not operate in both modes at the same time."

EPRI Report NP-2301 (March 1982) reported that 27% of loss of offsite power events at nuclear plants had been caused by weather related problems.

This report also stated that in 54 of all the loss of offsite power events at nuclear

plants, the duration exceeded 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

NSAC Report 182 (March 1992) reported 21 loss of offsite power events lasting longer than one hour at. nuclear plants between 1980 and 1991, with the longest event lasting 18:58.

Telecon from Michael Rose (PP&L) to Mort Renslo (Bechtel) of November 9,

1981 states that "according to Bech tel 's Civil and Structural Design Criteria for the Susquehanna Steam Electri c Station... This criteria states Fuel Pool Structure shall be designed for water boiling during accident condition."

July 15, 1992 Page 2

EDR G20020 References A.

Reactor Building Design Heat Loads Do Not Account for the Boiling Spent Fuel Pool Event FSAR 6.2.2.1(d) states that the safety design bases for the containment removal system is that the system "shall maintain operation during those environmental conditions imposed by the LOCA. "

EWR 830658 (March 1983) noted "the initial boiling rate corresponds to =3000 cfm of 100% water vapor at one atm.

Is the equipment which vill be exposed to this atmosphere qualified for it?"

SEA-ME-099 (December 1987) analyzed reactor building temperatures for LOCA, LOCA/LOOP and LOCA/false LOCA cases assuming spent fuel temperatures remained at

1254F, but listed as a nonconservatism that fuel pool heatup in the LOCA/LOOP case would result in higher heat loads from the RHR systems, fuel pool walls and fuel pool surface.

July 15, 1992 Page 3

EDR G20020 References B.

The Impact of the ESW Makeup Water to the Spent Fuel Pool on Equipment in the Reactor Building Has Not Been Evaluated FSAR 6.3.1.1.3 states that separation barriers for ECCS "shall be constructed between the functional groups as required to assure that environmental disturbances such as fire, pipe rupture, falling objects, etc., affecting one functional groups will not affect the remaining groups.

In addition, separation barriers shall be provided as required to assure that such disturbances do not affect both RCIC and HPCI."

FSAR 9.1.3.3 states that "the design makeup rate from each ZSW loop is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days folloving the loss of FPCCS capacity."

EWR 830658 (March 1983) noted "condensation may be expected from this evaporation which will run dovn to lover levels of the R.B.

Will this cause loss of essential equipment, parti cularly electrical?

Has an evaluation been performed?"

Minutes from Bechtel meeting on HVAC systems (February 1980) states that original requirement for SGTS was "to handle fumes from a

boiling fuel pool," but that SGTS will not be able to handle this mixture since the room will become too hot.

"This requirement vill be deleted from the FSAR."

July 15, 1992 Page 4

EDR G20020 References C.

The Manual Valve Manipulations Required to Provide ESP Makeup Flow to a Boiling Spent Fuel Pool May Not Be Possible FSAR 9.1.3.2 states that "the manual supply valves in these emergency makeup lines are accessible apart from the refueling floor."

FSAR 18.1.20 (NUREG-0737 Item II.B.2) states that "each licensee shall provide for adequate access to vital areas and protection of safety equi pment by design

changes, increased permanent or temporary shi el ding, or postacci dent procedural controls.

The design review shall determine which types of corrective actions are needed for vital areas throughout the facility."

FSAR 18.1.20.3.3.4.1 defines vital areas as those "which will or may require occupancy to permit an operator to ai d in the mitigation of or recovery from an accident."

FSAR

18. 1. 20. 3. 2. 1 states that "a review was made to determine which systems, could be required to operate and/or be expected to contain highly radioactive materials following a

postulated accident where substantial core damage has occurred."

FSAR 18.1.20.3.2.5 states "exposures for areas not continuously occupied (frequent and infrequent occupancy) must be determined case by case, that is, multiply the task duration by the area dose rate at the time of exposure. "

FSAR 18.1.20.3.3.3 states that "GDC 19 is also used to govern desi gn bases for the maximum permi ssi ble dosage to personnel performing any task required post-accident These requirements translate roughly into the obj ecti ves to be met in the post-accident review as given below.

Radiation Exposure Guidelines Occupancy Dose Rate Obj ectives Continuous 15 mR/hr Frequent 200 mR/hr Infrequent 500 mR/hr Accessway 5 R/hr Dose Objective 5 Rem for duration 5 Rem for all activities 5 Rem per activity Included in above doses" FSAR 18.1.20.3.4.3 states that the review results "show that the reactor building will be generally inaccessible for several days after the accident due to contained radiation sources."

FSAR Figure 18.1-4 shows Room I-105 where ESW valves 153500/153501 are located to be in Rad Zone VIIIwith dose rates over 5000 R/hr.

July 15, 1992 Page 5

EDR G20020 References FSAR Figure 18.1-6 shows Room I-514 where ESW valves 153090A&B and 153091A&B are located to be in Rad Zone V with dose rates between 5 and 50 R/hr.

NDI-6.4.3 establishes the whole body dose for life saving to be 75

Rem, with a

dose limit of 25 Rem for less urgent measures to protect equipment.

July 15, 1992 Page 6

EDR G20020 References D.

The Instrumentation Available to the Operator Post-LOCA Does Not Provide Adequate Indication of Spent Fuel Pool Temperature and Level to Allow Proper Response to a Loss of Fuel Pool Cooling Event Page 7

EDR G20020 References E.

The Analytical 25 Hour Time-to-Boil for the Spent Fuel Pool is Nonconservative for the Maximum Normal Heat Load in the Spent Fuel Pool FSAR 9. 1. 3. 1 states that "the pool will begin to boil 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after loss of cooling."

FSAR Appendix 9A states that "the conservative results shoved that the pools would not boil until at least 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after the loss of cooling. "

FSAR Table 9A-2 reports the total decay heat load to be 9.79x106 BTU/hr in the Unit 1 SFP and 7.92x10'TU/hr in the Unit 2 SFP for the boiling spent fuel pool analysis.

Bechtel Calc 200-0048 (July 1977) determined a 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time to boil for the Unit 1 SFP and a

31 hour3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br /> time to boil for the Unit 2 SFP based upon 12 month operating cycles and 184 bundle reload sizes.

PP&L Calc NFE-B-NA-053 (February 1992) determined a spent fuel pool maximum normal heat load of =14.6x10~ BTU/hr and emergency heat load of =30x10~ BTU/hr using actual SSES operating history through 1991 and projected operation until the pool is filled.

NSAG Report 4-90 states that "Appendix 9A of the FSAR states that at least 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> vould be required to boil the spent fuel pool under worst case loading."

July 15, 1992 Page 8

EDR 620020 References F.

The Analytical 25 Hour Time-to-Boil for the Spent Fuel Pool Does Not Account for the Emergency Heat Load in the Spent Fuel Pool July 15, 1992 Page 9

EDR G20020 References G.

The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconservative Evaporation Rates July 15, 1992 Page 10

S

EDR G20020 References H.

The Radiological Release Analysis for a Boiling Spent Fuel Pool-Uses Nonconservative Activity Terms July 15, 1992 Page 11

EDR G20020 References The Analysis for Maximum Time Prior to Makeup to a Boiling Spent Fuel Pool is Based Upon Nonconservative Assumptions NSAG Report 13-'84 (December 1984) reported that water level in the spent fuel pool dropping to within five inches of the top of the irradiated fuel "would cause radiation levels on the 818'levation of the reactor building in excess of 200,000 rem/hour."

July 15, 1992 Page 12

Figure 1

Fuel Pool Cooling Unit 1 Spent Fuel Pool Unit 2 Spent Fuel Pool Unit 1 FPCCS, if

~

AC power available

~ piping intact

~ Ul SW available

~ decay heat

< 13.2x10 No Yes Yes Unit 2 FPCCS, if

~

AC power available

~ piping intact

~

U2 SW available

~ decay heat

< 13.2x10 No Unit 2 FPCCS, if

~

AC power available

~ piping intact

~

U2 SW available

~ pools crosstied (or U2 SFP not handled)

~ decay heat

< 13.2x10 No U

Yes T

Yes Unit 1 FPCCS, if

~

AC power available

~ piping intact

~ Ul SW available

~ pools crosstied (or Ul SFP not handled)

~ decay heat

< 13.2xl0 No Unit 1 RHR FPC Assist, if

~ Ul RHR SDC not required

~ manual valves aligned

~ Ul RHRSW/UHS available

~ Ul RHR Uncontaminated Yes 0

0 Yes Unit 2 RHR FPC Assist, if

~

U2 RHR SDC not required

~ manual valves aligned

~

U2 RHRSW/UHS available

~

U2 RHR Uncontaminated No No Unit 2 RHR FPC Assist, if

~

U2 RHR SDC.not required

~ manual valves aligned

~

U2 RHRSW/UHS available

~

U2 RHR Uncontaminated Yes Yes Unit 1 RHR FPC Assist, if

~ Ul RHR SDC not required

~ manual valves aligned

~ Ul RHRSW/UHS available

~ Ul RHR Uncontaminated No No Loss of Ul SFP Cooling Loss of U2 SFP Cooling Figure 2

Figure 2

June 25, 1992

Figure 2

Fuel Pool Heatup Figure 1

Figure 1

Loss of Ul SFP Cooling Loss of U2 SFP Cooling Time to Boil function of

~ initial pool temperature

~ pool water inventory

~ decay heat Time to Boil function of

~ initial pool temperature

~ pool water inventory

~ decay heat Unit 1 Spent Fuel Pool Boiling

~

RB Heat Load Impact

~

EQ Impact

~

SGTS Impact

~

RB-HVAC Impact

~

RB Flooding Unit 2 Spent Fuel Pool Boiling

~

RB Heat Load Impact

~

EQ Impact

~

SGTS Impact

~

RB-HVAC Impact

~

RB Flooding Figure 3

Figure 3

June 25, 1992

Figure 3

Fuel Pool Boiling

& Makeup Figure 2

Figure 2

Unit 1 Spent Fuel Pool Boiling Unit 2 Spent Fuel Pool Boiling Cond Xfer Water, if

~

AC power available

~ piping intact

~ manual valves aligned No RWST to Cask Storage Pit, if

~

AC power available

~ piping intact

~ manual, valves aligned No Ul ESW System, if

~ manual valves aligned No Yes Yes Yes A

D E

9 U

T M

A K

E U

P Yes Yes Yes Cond Xfer Water, if

~

AC power available

~ piping intact

~ manual valves aligned No RWST to Cask Storage Pit, if

~

AC power available

~ piping intact

~ manual valves aligned No U2 ESW System, if

~ manual valves aligned No U2 ESW System, if

~ manual valves aligned No Yes Yes Ul ESW System, if

~ manual valves aligned No Loss of Ul SFP'ater Level

~

RB Dose Impact

~ Offsite Dose Impact

~ Fuel Damage Impact Loss of U2 SFP Water Level

~

RB Dose Impact

~ Offsite Dose Impact

~ Fuel Damage Impact June 25, 1992

IOO 90 Loss of Spent Fuel Pool Cooling FPCCS Capacity 80 70 Dual FPCCS Capacity 60 50 O

~~

40 30 Analytical Limit 20 10 0

5 10 15 20 25 30 35 40'5 50 55 Total Decay Heat Load, MBTU/ht.

Single Fuel Pool Crosstied Pools One Unit Refueling ""--" Crosstied Refueling

Attachaent 8

white Paper prepared by David A. Lochbaum and Donald C.

Prevatte, "Safety Consequences of a Boiling Spent Fuel Pool at the Susquehanna Steam Electric Station",

July 27, 1992 Note:

This paper was handed to the PP&L Manager of Nuclear Plant Engineering in a

meeting requested by the authors.

This paper was prepared when the authors became convinced that the PP8L Engineering Discrepancy Management Group and the PP8L Supervisor, Engineering Projects were unable to properly evaluate EDR 620020.

The timing of this paper was dictated by the end of Mr. Lochbaum's contract at PPKL.

~t

/t

'I