ML17157C142

Engineering Assessment of Fuel Pool Cooling Piping EDR-G20020.

ML17157C142
Person / Time
Site:	Susquehanna
Issue date:	10/21/1992
From:	PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML17157C139	List: ML17157C138 ML17157C140 ML17157C141 ML17157C142 ML18026A248
References
NUDOCS 9301050151
Download: ML17157C142 (96)

Text

ENGINEERING ASSESSMENT OF FUEL POOL COOLING PIPING EDR-G20020 Prepared by: c~P Date: /<ll~~

Reviewed by: Date: )cO Za 9Z Approved by: Date: i~la/ l~

Rev. 1 10/20/92 9301050i5i 92ii27 05000387 "PDR ADOCN, S PDR

ENGZNEERZN A ESSMENT OF F EL P OL OLZN PZPZNG - EDR G20020

~ Backgr ound A preliminary engineering assessment has been made of the SSES Fuel Pool Cooling piping to judge its ability to withstand earthquake and hydrodynamic loadings. Structural integrity and system operability were considered. This assessment is based upon a review of approximately 25 FPC fabrication isometrics and associated pipe support drawings. Also reviewed were approximately 15 other fabrication isometrics and the associated pipe support drawings for the service water and condensate systems.

The conclusions reached as to the operability of the FPC and interfacing piping have been based, to a large degree, on engineering judgement. This assessment was performed in order to provide a preliminary assessment of EDR 620020.

~ Current Design All of the Fuel Pool Cooling piping which comprises the primary system flow path was reviewed. This includes 6", 8", and 10" lines extending from the skimmer surge tank, through the heat exchangers and pumps and back to the spent fuel pool; reference P&ID H-153. Almost all of the subject piping is ASHE Class 3, however, a review of the FPC Design Report, DR-114, shows that only a small portion of the system has been dynamically analyzed, i.e.

a section of pipe coming from the surge tank and the pipe directly adjacent to the fuel pool. The majority of the FPC piping has not been dynamically analyzed, i.e. seismic and hydrodynamic loads not analyzed.

The Unit I isometrics reviewed contained approximately 1000 linear feet of FPC large bore piping, typically of 6", 8", or 10" diameter. There are

'also a number of small bore branch lines connected to the main runs which were not individually reviewed. The Unit II FPC piping is similar in quantity, size and layout.

The FPC non-seismic piping is supported using spring can supports, rigid rod supports and at some limited locations, rigid struts. The design of these supports included deadweight and thermal expansion loads only. No evidence of the consideration of dynamic loads for "two-over-one" concerns has been found. Also noted in the review was the lack of many analytical anchor points which typically isolate branch connections and thereby result in smaller, more manageable analytical boundaries. There is at least one branch line which cross-connects with Unit II piping that is not anchored or analytically isolated.

The service water piping supplying the fuel pool heat exchangers was also reviewed, reference P&ID H-110. This piping is supported using primarily spring can hangers and -rigid rod hangers. The pipe ranges in size from 4" to 14" diameter. In addition, the condensate lines supplying fuel pool Page 2 of 5

~ Hydrodynamic Loadings Fuel Pool Coolin It is the judgement of the reviewer that there is a low risk to system operability if the FPC piping experiences hydrodynamic loadings in its current design configuration. This judgement is based on the following.

Hydrodynamic events due to a LOCA are most impacting for those piping systems located inside containment, attached to containment or located at lower elevations in the reactor building. The FPC piping is located in the reactor building at relatively high plant elevations (above 764') where the effect of hydrodynamic events is much less. A review was made of the FPC piping which was analyzed for both seismic and hydrodynamic loads at these elevations. It was found that the load contribution due to hydrodynamic events resulted in approximately a 25X maximum increase over pipe support deadweight loads. It is likely that the FPC supports have enough reserve capacity to accommodate a 25X increase in loads. The reduced effect due to hydrodynamic loads on supports implies a reduced effect on pipe stress and equipment.

Interfacin S stems It is the judgement of the reviewer that there is a moderate risk to system operability for interfacing systems such as service water and condensate.

This judgement is based on the following.

These systems are located at lower elevations of the reactor building where the effect of hydrodynamic loads are higher. A review was made of similar piping located low in the reactor building which was dynamically analyzed to determine the load contribution due to hydrodynamic loads. The load increases were only slightly larger than found at upper reactor building elevations. It is judged that the supports have sufficient design margin to accommodate the increases due to hydrodynamic loads. As with the support loads, it is anticipated that pipe stresses and equipment loadings would be slightly larger at these lower elevations but not of a magnitude which would affect system operability.

These interfacing systems are judged to be slightly more vulnerable to hydrodynamic loads due to the increased magnitude as described above, the quantity of pipe and the uncertainties associated with field supported piping, i.e. no hanger details or support configurations available.

~ Conclusions Based upon the preliminary review performed on the Fuel Pool Cooling Piping and interfacing systems as described above, it is concluded that there is a high risk to system operability when considering the potential effects of seismic loadings (OBE/SSE). It is judged that pipe support modifications would be required to ensure system qualification.

Page 4 of 5

LOSS OF FUEL POOL COOLING EVENT EVALUATION NE-092-002 Rev. 0 ATTACHMENT 2 RADIOLOGICAL EVALUATION IN SUPPORT OP EDR4 G20020 (EP-548)

~ ~

MEMORANDUM TO: G. D. Miller, A6-3 DATE: October 21, 1992 FROM: D. A. Matchick, A6-2 COPXES: C. J. Kalter, A9-3 K. E. Shank, A9-3 JOB: NUMBER: EP - 54 8 FXLE: REPLY: No

SUBJECT:

Radiological Evaluation in Support of EDR G20020 As per you request,, radiological engineering evaluations of in-containment and off-site doses have been performed in support of the resolution of EDR G20020. The attachments to this memorandum summarize the results of those evaluations. Documentation packages containing the computer code runs and detailed results of these analyses will be transferred to you under separate cover.

MZLMEM.WPF

Radiological Evaluation I. Airborne Doses The TACT5 computer code was run to provide realistic estimates of radioactivity concentrations that might be expected on the Refuelling Floor (818') and a representative room (20'x20'x20'). The TACT5 code is validated and verified for design work under EPM-QA-104. The Chapter 15 DBA"LOCA model as obtained from PP&L calculation FX-C-DAM-014 was used as the base model, with a realistic estimate of 1% cladding failure as the source term. The 1% fuel cladding failure parameter was provided by Nuclear Fuels. A TACTS run was conducted for a 100%

cladding failure, and the activity concentrations were linearly scaled back to 1% cladding failure. Activity concentrations were then input into the MICROSHIELD 3.0 computer program to compute gamma exposure dose rates in the subject areas.

The MICROSHIELD 3.0 computer program is validated and verified for design use under EPM-QA-104.

From an earlier analysis, SE<<B-NA-074, it. is known that gross airborne activity concentrations tend to rise to a maximum at, around 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> post-accident, and hold constant until about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident in the reactor building. This is because the activity flow out of the drywell into the reactor building competes with radioactive decay processes to hold the gross activity concentration rather constant over this period. As such, the total activity concentrations of iodines, noble gases, and cesium at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> were used in the MICROSHIELD analyses for airborne dose rates.

For the Refueling Floor, the exposure dose rate at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> post,-accident for the above scenario with 1% cladding failure was found to be 1.411 R/hr. This dose considers gammma only. Protective clothing to shield the beta component will be required.

For a representative 20'x20'x20'oom in the reactor building, the exposure dose rate for the above scenario with 1% cladding failure was found to be 46.8 mR/hr.

This dose considers gamma only. Protective clothing to shield the beta component will be required.

The TACT5 code was subsequently used to determine doses on the refueling floor from a design basis (conservative) source term. This source term consists of 100t of core noble gases and 25% of core iodines being released into primary containment instantaneously at the start of the accident. Gamma exposure doses on the refueling floor from the design basis containment leakage rate of 1.0 4 per day and a more realistic .25% per day were calculated for various times post accident. The gamma exposure doses at the center of the refueling floor as a function of time are:

Time Post-accident Dose 1% da Dose .25% da 4 days 404.0 R/hr 158.5 R/hr 30 days 27.1 R/hr 12.4 R/hr 45 days 6.5 R/hr 3.2 R/hr 60 days 1.6 R/hr .9 R/hr II. Contained Sources It is postulated that a valve in the RHR Pump room will be required to be manually manipulated. The approximate dose rate at the valve location at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> post accident was determined. In this case, the source is radioactivity in suppression pool water from 1% failed fuel cladding which is contained in RHR piping . The analysis consisted of using the MICROSHIELD 3.0 computer program to estimate the exposure dose rate from two sections of piping, a long section below the area the operator would be working in, and a shorter section of pipe where the valve would have to be manually opened.

2. Additional DBA LOCA Doses Due To Purge Of Reactor Building TABLE ZONE III-Drywell Leakage = 0.25% / day ADDITIONAL DBA LOCA DOSE DUE TO ZONE III PURGE Dose AT LPZ Control Room Doses (REM) (REM)

Time At Which Purge THYROID WHOLE BODY THYROID WHOLEBODY BETA SKIN Occurs 2 hr 70.8 0. 178 12. 6 0.056 1.08 8 hr 195. 8 0. 202 34.9 0.054 1.28 24 hr 276.4 0.112 49.3 0.020 0. 826 4 day 192.5 0.034 34.3 0.006 0. 305 30 da 17.6 0.0022 3.14 0.0002 0.024 45 da 4. 53 0.0005 0.807 ne 0.016 60 day 1. 16 0.0001 0.208 neg. 0.014

LOSS OP PUEL POOL COOLINQ EVENT EVALUATION NE-092-002 Rev 0 ATTACHMENT 3 ZONE ZZZ VENTING EVALUATION L

LOSS OP FUEL POOL COOLZNG EVENT EVALUATION For the purposes of this evaluation, this process will be defined as Zone ZZZ Venting as following:

Isolating the connections between Zone ZZZ and the recirculation plenum and providing Zone ZZZ ventilation by running filtered exhaust only or unfiltered exhaust and supply if radiation levels permit.

This section examines the various aspects of how to provide Zone III venting and ensure that it is a viable option that will not compromise other plant systems. The areas evaluated are:

Zone III Venting Fuel Pool Cooling Instrument Air Supply 480 VAC Circuit Breaker Changeout For Zone III Venting SGTS Operability Assessment of Emergency Ventilation Options for Zone III Venting during Fuel Pool Boil Scenarios EQ Ecpxipment Located in HVAC Zone III

2.2 To 1V212B HD17564A HD-17564 B HD-17566B 2.3 To 1V213B HD-17502B HD-17502 B PDD-1750 1B 3.0 Jumper out LOCA contact in Isolation Stop Logic for 1V212B and 1V217B (E-192 sh1) 4.0 Start Zone III HVAC. If HVAC does not start due to Loss of Off-site Power provide Class 1E power to 1V212B,1V213B and 1V217B 5.0 Provide Class 1E power to Zone III HVAC Fans 5.1 1V217B 5.1.1 Install and connect, cable from 1B227 to 1B261-024 (1-3/C 42) 5.1.2 Change wiring in 1B227 and 1B261-24 cubicles to remove 0/L and contactor contacts from power circuit 5.1.2 Block open all circuit breakers on 1B261 except 1B261-024 5.1.3 Close the 1B227 supply circuit breaker and 1B261-024 circuit breaker 5.2 1V212B and 1V213B 5.2.1 Install and connect cable from 1B220-034 to 1B280-023 ( 3-1/C 4 500Kcmil) 5.2.2 Block open all circuit breakers on 1B280 except for the 1V212B and 1V213B circuit breakers 5.2.3 Close the 1B220 and 1B280-024 supply circuit breakers 6.0 Start Zone III HVAC Train B

FUEL POOL COOLING INSTRUMENT AIR SUPPLY Instrument Air is required to hold open the following Dampers for Zone III HVAC 'B'ans:

1V212B 1V213B 1V217B HD-17564A HD-17502A HD-17514A HD-17564B HD-17502B HD-17514B HD-17566B PDD-17501B HD-17511B HD-17513B HD-17518B PDD-17512B Instrument Air also supplies the instruments that actuate FSL-17512B which is used in the control of 1V217B.

The Instrument Air Compressors and Controls are power from Diesel Generator Backed supplies. In the event of Loss of Off-site Power, the Instrument Air Compressors and Control have electrical power.

The Above Zone supplies.

III Dampers are powered from Diesel Generator backed

~SC Loss o 0 f-Site Power with o Seismic e The above listed Zone III Dampers for the 'B'ans can be opened by using normal procedures to open the Dampers.

LOCA LOOP with no Seis ic Event The control logics for the following Dampers have LOCA actuation contacts in the circuits which close the Dampers and prevent their opening for a LOCA. These contacts MUST be defeated so the Dampers

.can be opened by normal procedures.

HD-17564A HD-17502A HD-17514A HD-17564A HD-17502 B HD-17514B The other listed Dampers open when their respective Fans start.

The Instrument Air System is considered lost for a seismic event since the tubing in the system is not designed with seismic supports For this event the above listed Dampers must be mechanically locked open 'in preparation for starting the Zone 1V217B Fans.

III 1V212B, 1V213B and

LO88 OP FUEL POOL COOLING EVENT EVALUATION SGTS OPERABILITY In a postulated seismic event that presumably causes the loss of the fuel pool cooling system, Zone III will be isolated (normal Zone ZZI HVAC shutdown, Zones I and II isolated from recirc plenum) and the SGTS will be used to maintain Zone III pressure (overpressure control). No radiation source terms will exist in Zone IZI. The only possible source term that might exi~t would be from leaking fuel in the fuel pool.

Currently five (5) leaking fuel rods are maintained in the pool. Thzs is within that assumed in Appendix 9A. Thus the analysis contained in FSAR Appendix 9A is bounding (this will need to be confirmed each cycle). Note that the Appendix 9A offsite dose analysis takes no credit for operation of SGTS.

Even without SGTS, acceptably low.

it still concludes that offsite doses are In the scenario evaluated here, assuming no pool cooling is established, it is possible that the SGTS system will. be operating with incoming Zone *III air conditions created by the high temperature pools. This could result in elevated Zone XII air temperatures and high humidity.

Carried to extremes, these conditions could result in condensation collecting in the duct work and SGTS filter train. It is expected that it would take a significant amount of time for this to occur. This is in part due to the large time to boil that exists during the current operating cycle (See Figures 1 and 2). It is deemed probable that pool cooling would be reestablished prior. to this condensation becoming threatening to the SGTS train components or ductwork.

If during this scenario SGTS is rendered inoperable (as identified above this is not expected to occur or, at worst, it is not expected to occur for an extended period of time) pressure control of Zone IZX can be achieved by Zone III venting.

Note that calculation 175-017 revision 3 determined that with 10500 CFM at 180 F, 1004 relative humidity, the heaters (OE101Aa and OE101B) needed to be rated for 90KW. In this scenario the flow rate through SGTS will be limited to the air inleakage to zone ZZZ. This flow rate (on the order of 2000 CFM) will be significantly lower than the 10500 CFM evaluated in the calculation. (10500 CFM is the maximum system flow rate). Thus for the scenario evaluated here, the heat required of the air stream to drop the relative humidity to acceptable levels will be less than at the conditions evaluated in the calculation. The calculation thus bounds the conditions expected in this scenario. The installation of the 90 KW heaters is verified by review of drawing E242 sht 3 revision 15.

s

~Ventin cirrose The purpose assessment

~durin Fuel pool

~ 'e LOSS OP FUEL POOL COOLINQ EVENT EVALUATION

~o'cenarios Ic-of this evaluation ie to provide a qualitative of three proposed options for ventilating Zone III during fuel pool boiling scenarios. All three options require that Zone III be isolated from the recirculatzon plenum and involve the following scenarios:

I) Pool boiling with a vent path and no fans in operation; II) Pool boiling with the normal HVAC supply, exhaust and filtered exhaust fans in service; III) Pool Boiling with only the filtered exhaust fan running.

Assum tions

1) Zone III is completely isolated. This includes complete isolation from the secondary containment common recirc plenum and the Standby Gas Treatment System. The required isolation lineups are achieved by manually closing various dampers and/or the installation of blank flanges.

2) Electrical power is available to the Zone III fan motors and their associated control schemes.

3) For the purposes of damper operation and flow control, ita) istheassumed that either:

instrument air system is functional, or b) balancing damper's can be adjusted manually.

4) Additional assumptions are stated where applicable.

References

1) PP&L Calculation M-FPC-009 Rev. 0

2) Combustion Engineering Steam Tables

3) Bechtel Calculation 175-19 Rev; 2

4) Bechtel Calculation 175-17 Rev. 3

5) Bechtel Specification 8856-A-7

6) 1985 ASHRAE Fundamentals Handbook

7) PP&L P&IDs M-175 Rev.25 and M-2175 Rev. 15

8) PP&L Specification M-1070 Rev. 1

9) FSAR Table 6.2-17

10) ASHRAE Psychrometric Chart 43 High Temperature

11) PP&L Drawing FF108570 Sht. 6401 Rev. 4

12) PP&L Drawing FF108880 Sht. 1301 Rev. 1 Conclusions The following conclusions are based on a qualitative analysis which assumes the worst, case: rate of pool boiling as determined in Reference 1. Hence these conclusions also apply to scenarios which involve a more realistic rate of pool boiling.

LOSS OP PUEL POOL COOLING EVENT EVALUATION In the event that Zone III venting is required after an accident, actions can be taken to physically isolate the Zone III air volume from the rest of the reactor building air spaces. Zone III can also be isolated from the common recirculation fan plenum and the Standby Gas Treatment System.

In this configuration, it is possible to start a filtered exhaust fan to take advantage of the filtering capability of a filtered exhaust train. Also, if available, it zs possible to re-establish ventilation to this volume (Zone III) with those systems which normally serve this function.

Although the function of the normal supply and exhaust systems are not related to safety, there is a high probability that portions of these systems and their associated components will be available to aid in post accident ventilation cooling, including seismic scenarios.

The following rationale supports this argument:

1) The performance of these major plant components is closely monitored as part of, the predictive maintenance program.

Since failure of a major HVAC component could jeopardize plant reliability, thxs equipment is maintained to high standards. While these components are not related to safety, they maintain the secondary containment pressure within the Tech Spec operating envelope.

2) Although portions of the associated duct work are not seismically qualified, significant portions of the "Non-Q" duct work are qua).ified as "Class A" which is supported to withstand the safe shutdown, earthquake. The remaining "Class B" 'duct work is not seismically designed. However, as with many other non-seismic plant components, design specifications require that it be capable of withstanding modest seismic accelerations of .05g (horizontal) and .03g (vertical) without exceeding code allowable stresses.

3) There are four complete sub-systems, comprised of a total of twelve fans, which provide the normal Zone III HVAC function. Considering this equipment complement, is likely that at least one set of components will be it available to establish supply and exhaust ventilation flow.

E i ment Evaluatio The following discussions are intended to provide a general qualitative assessment of the post accident availabilxty and operation of the normal Zone III HVAC system components.

Option I is not addressed in these evaluations since equipment operation is not required to support this mode.

LOSS OP PUEL POOL COOLING EVENT EVALUATION The effects of exhaust fan operation at the higher temperatures would be at worst, accelerated component wear.

The capacity of the Zone CFM. Per the attached III filtered exhaust fans is 6500 calculations, it is seen that this capacity is less than the worst case volumetric pool boiling rate. Although the calculation implies that the remaining vapor would exit through the 2.2 ft'pening, it is more realistic to expect that the direction of flow would actually be canto the Zone III volume for the following reasons:

1) In this "volumetric balance", no credit is taken for condensation which would occur on the wall and ceiling surfaces. These surfaces will in fact, provide cooling since they would ultimately transport heat to the environment.

2) More realistic estimates of the pool boiling rate are within the volumetric capacity of the filtered exhaust fans.

Fan Motors & Powe Su l E me t Per the discussions above, fan flow and dP would be essentially the same as during normal operation. The fan motors are located in the >>no-zone>> spaces on elevations 799', which are not connected to Zone III and hence, would not be subjected to the high temperature and humidity 779'nd levels induced by fuel pool boilznq. If the normal supply and exhaust fans are running (Option II), these >>no-zone>>

areas will be supplied with- ventilation cooling, as during normal operation. If only a filtered exhaust fan is in operation, the ambient operating temperature may be slightly elevated, but nonetheless acceptable.

Ductwork Dam ers As previously stated, portions of the Zone III normal supply and exhaust ductwork are not seismically qualified. However, significant portions of the >>Non-Q>> duct work are qualified as "Class A>> which is supported to withstand the safe shutdown earthquake, and the remaining "Class B>> duct work is also designed to withstand small seismic loadings.

therefore reasonable to assume that. significant portions of It is duct required to support operation of the Zone be intact, or only require minor repair.

III fans will The design, fabrication, and installation of all of the associated ductwork is governed via PP&L Specification M-1070 which provides assurance of high quality workmanship and design configuration. Further; all ductwork and balancing dampers are qualified for a continuous operating temperature of 150'F at a relative humidity of 1004.

In the event that automatic flow control cannot be re-established, the system's ductwork contains multiple flow

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EQ Equipment Located in HVAC Zone III A list of EQ equipment located on elevation 779 and above was generated from SEIS. This list was reviewed by Kevin Browing/Mark Mjaatvedt to determine which equipment was required to operate during the period when high humidity conditions from fuel pool boiling might occur. Items identified are as follows:

Plant ID ~Corn T'~e Old Binder New Binder FT-07557 Tavis PC8 EQDF-29 EQAR-016 ll II II II II II PDT-07554A1,A2,A3,B1,B2,83 HDM-07545A,B NH-90 EQDF-31B EQAR-074 PDDM-07554A,B NH-90 EQDF-31B EQAR-074 PDSL-07544A,B Dwyer dp EQPL-JOI EQAR-087 OV201A,B WHSE Motor EQDF-26 EQAR-057 The binder on each of these items was reviewed to determine if items were qualified/tested for 100X humidity conditions during normal and/or accident conditions.

Tavis PCB transmitters These transmitters were not subjected to 100X humidity in either normal or accident tests. These devices are potted which may allow an evaluation to be performed that eliminates humidity as a qualification concern. The old binder SCEW sheets state that these transmitters are qualified for 90X humidity.

NH-90 Actuators Actuators were exposed to steam and spray during first 12 hours of accident testing. The electical compartment was not sealed and the environment was allowed entry. Low humidity was maintained during thermal aging. Anticipate that 100X humidity during normal and at least the early portion of the accident can be justified. The old binder SCEW sheets state that these actuators are qualified for 100X humidity.

Dwyer Pressure Switches The test report states that switches are qualified for 100X humidity during both normal and accident conditions. A cursory review of the test indicates that 100X humidity conditions were applied only during the accident test. Anticipate that 100X humidity during normal and accident conditions can be justified. The old binder SCEW sheets state that these switches are qualified for 100X humidity.

'Westinghouse Fan Motors Aged motorettes were exposed to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in a condensing atmosphere while de-energized. While still in the chamber and visibly wet the motorettes were tested to 600 VAC for 10 minutes both phase to phase and phase to ground. Anticipate that 100X humidity during normal and accident conditions can be justified. The old binder SCEW sheets state that these motors are qualified for 100X humidity.

e LOBB OF FUEL POOL COOLING EVENT EVALUATION NE-092-002 Rev. 0 ATTACHMENT 4 TIME TO MAKE-UP FOR FUEL POOL

LOSS OP PUEL POOL COOLING EVENT EVALUATION ASSUMPTIONS:

1. Fuel pool level is at the normal level of 817'1". (J653 sht 62)

2. Evaporation rate to boiling will be based on prior 170'F 170 F pool temperature. is the approximate average of the pool temperature between the initial assumed temperature of 110'F and the final or boiling temperature of 210'F.

3. The minimum acceptable level is the bottom of the fuel pool weir which is at 816' 3/4" ( M29-9). Before pool level reaches this level, makeup needs to be provided.

Note that the minimum acceptable level per technical specifications is:

779'4" Bottom of fuel pool (C1932 sht 5) 14'8" Height of fuel bundles (XN-NF-304,998)

+

+ 22'ech Spec margin (3/4.9.9) 816'hus use of 816' 3/4" provides considerable margin.

4. No fuel pool level indication is available. If available, then actual conditions should be used versus it is this estimate.

Total volume and thus mass that will be evaporated prior to reaching the minimum acceptable pool level:

Fuel pool surface area = 1350 FT2 Depth 817r1r 816r- 11r- 9167 FT Volume 1237.4 FT3 Using (conservatively) a water density at 110 F = 61.8 (Crane), the mass loss will be:

1237.4 FT3 (61.8 ibm/FT3) =.76477.5 Lbm For the current operating cycle 11/92 to 9/93:

Once boiling occurs; Q = heat load H = Latent heat of Vaporization Mass loss = Q/H H at 210'F = 1031 BTU/LBM

LOSS OF FUEL POOL COOLXNG EVENT EVALUATZON NE-092-002 Rev. 0 ATTACHMENT 5 REVZBED EVALUATZON OP EDR G20020 SPENT FUEL POOL COOLXNG ZSSUE (PLZ-727 6i )

October 29, 1992 George T. Jones A6-2 SUSQUEHANNA STEAM ELECTRIC STATION REVISED EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-72764 FI E A45-1A The attached revised evaluation of EDR G20020 is provided for your information. The revision accounts for a change to the engineering report reference, corrects a typographical error on page 10, expands the event tree to three pages and provides several graphs used as part of the PORC presentation on Monday October 26, 1992. The conclusions remain unchanged.'lenn D. Hiller CC: G. J. Kuczynski SSES J. E. Agnew . - A6-3 C. A. Myers - A2-4 H. R. Mjaatvedt - A6-3 M. W. Simpson A1-2 D. F. Roth SSES H. G. Stanley - SSES J. H. Kenny - A2-4 J. S. Stefanko A9-3 F. G. Butler A6-3 J. R. Miltenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records - AS-2 D. A. Lochbaum Enercon

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October 21, 1992 George T. Jones A6-2 SUS(UEHANNA STEAN ELECTRIC STATION EVALUATION OF EOR 620020 - SPENT FUEL POOL COOLING ISSUE

<<7 1 F The attached evaluation of EDR G20020 is provided in response to your memo PLI-72640. This evaluation was prepared by myself and a team of engineers working on the action plan to resolve the subject EDR.

ln the course of reviewing this issue in detail, we have concluded that seven of the nine identified discrepancies are not valid deficiencies. This is explained in detail in the evaluation. The two remaining discrepancies are valid deficiencies but are not considered safety significant nor reportable.

I believe .that the technical basis on several of these issues has been clarified considerably in the course of the past week. Therefore, I suggest providing this evaluation to Nuclear Regulatory Affairs for reconsideration of the reportability aspects.

We are continuing with the remaining action items as requested. A detailed engineering design report and justification for interim operation will provide more detail than contained in the attached evaluation. We are scheduled to meet with PORC on monday October 26, 1992 at 2:30 pm to review this issue. We are working with Systems Engineering, Operations and NRA-Compliance with respect to potential compensatory measures.

Glenn D. Hiller CC'. J. Kuczynski SSES A2-4 J. E. Agnew -

A6-3 C. A. Ayers H. R. Hjaatvedt A6-3 N. W. Simpson Al-2 D. F. Roth - SSES H. G. Stanley - SSES J. H. Kenny - A2-4 J. S. Stefanko - A9-3 F. G. Butler - A6-3 J. R. Hiltenberger A6-1 D. C. Prevatte - A6-3 Nuclear Records A6-2 D. A. Lochbaum - Enercon

0 October 21, 1992 Page 1 Rev. 1 Evaluation of EOR 620020 This document contains an evaluation of the discrepancies documented in EDR G20020, "Loss of Spent Fuel Pool Cooling Event Design Discrepancies."

Conclusions of the author with respect to reportability of these concerns and operability impact on SSES are also provided.

8 The design bases for the Fuel Pool Cooling System are found in FSAR section 9.1.

The portions of the design basis relevant to EDR G20020 are as follows:

1. Maintain the fuel pool water temperature below 125F under "normal maximum heat loads" defined as 12.6 HBtu/hr (equivalent to a typical fuel cycle discharge schedule which fills the fuel pool, last quarter core offload at 6.7 days after shutdown).

Maintain fuel pool water temperature at or .below 125F during the "emergency heat load" condition of 32.6 MBtu/hr (equivalent to a full core offload 10.5 days after a shutdown following a typical fuel cycle discharge schedule which fills the fuel pool) utilizing the RHR system (with or without normal fuel pool cooling) for fuel pool cooling. This mode of operation applies "during periods of higher than MNHL generation in the fuel pool, eg., storing of a full core of irradiated fuel shortly after shutdown". The RHR system is used under these conditions to assist the FPCCS in dissipating the decay heat. Thus, any heat load in excess of 12.6 HBtu/hr is considered to be within the design basis for the RHR FPC assist mode of operation.

3. Redundant Seismic Category I ESW connections to each pool are provided to allow for makeup of evaporative losses in the event of failure of the FPC system. The conditions are bounded by a fuel pool time-to-boil analysis based on the same typical fuel cycle discharge schedule as in basis Pl except the time after shutdown is 10.5 days instead of 6.7 days resulting in a heat load of 9.S MBtu/hr. (This explains the difference between the two different heat loads, ie., 12.6 MBtu/hr for basis 81 and 9.8 HBtu/ht for basis f3. This is not a discrepancy.) The ESW makeup line is sized on the basis of this calculation (Reference FSAR section 3.13).

4. The cause of the Loss of Spent Fuel Pool Cooling event is stated to be a seismic event.

5. All piping and equipment shared with or connecting to the RHR intertie loop are Seismic Category I and can be isolated from any piping associated with the non-Seismic Category 'I fuel pool cooling system.

valuation of Di cre ancies Noted in R G200 0 EDR G20020 describes nine discrepancies relating to the loss of spent fuel pool cooling event. This discussion will sumnarize each issue. The reader is

October 21, 1992 page 3 Rev. 1 t m: Anal tical Time-to-Boil "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."

As stated in basis tl the maximum normal heat load is 12.6 MBtu/hr. As stated in basis 4'2 the time-to-boil analysis is based on a heat load of 9.8 MBtu/hr.

These two design bases are in fact consistent and are based on the same "typical fuel discharge schedule" and refueling outage scenario. The difference in the heat load is due solely to the time after shutdown assumed for purposes of establishing the design basis.

Focusing on the time-to-boil analysis, a time after shutdown value of 10.5 days is used. This is the time at which it is assumed that refueling is completed and the reactor cavity to fuel pool gates are reinstalled. Prior to that point the additional water stored in the reactor cavity is also available as a heat sink and the RHR system is available for fuel pool cooling. For times greater than 10.5 days the appropriate heat load will be even lower than the analyzed value of 9.8 MBtu/hr . For the SSES Unit 2 5RIO the time from reactor shutdown to fuel pool gates installed was 38 days. The decay heat in the Unit 2 pool at that time is calculated to be 5.65 MBtu/hr (Reference NE-092-002). The corresponding time-to-boil is 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />.

The EDR goes on to discuss other calculations which result in different heat loads using various assumptions. Calculation NFE-B-NA-053 was performed by Nuclear Fuels:to account for actual fuel discharge history and future offloads accounting for power. uprate conditions. The fuel pool heat load versus time curves obviously will increase subsequent to power upr ate, however, these curves do not apply to the existing design. As long as the calculated decay heat is less than 9.8 MBtu/hr at the point where the fuel pool gates are reinstalled the original design basis time-to-boil. calculation is still valid.

Calculation M-FPC-009 determined time-to-boil conditions post power uprate. This calculation shows that the time-to-boil for the design basis heat load of 9.8 MBtu/hr is slightly greater than 25 hours.

In conclusion, the design basis for the time-to-boil condition is established by the 9.8 MBtu/hr value used in the original calculations. This design basis is met by planning the outage so that the fuel pool is not isolated from the reactor cavity or the RHR system prior to a point in time where the actual heat load is 9.8 MBtu/hr or less.

This discrepancy is not a valid deficiency, is therefore not reportable and has no impact on the operability of the plant.

October 21, )992 Page 6 Rev. 1 tern G: Radiolo ical Release Calculation fo Hoilin S ent uel Pool "The radiological release analysis for a boiling spent fuel pool uses nonconser vative evaporation rates."

This discrepancy is directly related to the heat load assumed for the time-to-boil analysis. The evaporation rate used in the dose calculation is based on a heat load of 9.8 MBtu/hr which is the design basis heat load for the time-to-boil calculation. Heat loads in excess of 9.8 MBtu/hr obviously result in higher evapo'ration rates. Since the discussion under Item E above establishes that 9.8 MBtu/hr is the correct original design basis and still bounds current operation there is no discrepancy in the offsite dose calculation. It uses an evaporation rate consistent with the design basis heat load.

This discrepancy is not a valid deficiency, not reportable and has no impact on plant operability.

October 21, 1992 page 7 Rev. 1 te I: Anal sis for Ha Time Prior to Hakeu "The analysis for maximum time prior to makeup to a boiling spent fuel pool is based upon nonconservative assumptions. Tta 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."

The purpose of the referenced calculation was to determine refueling floor atmosphere conditions under various operating modes. The evaporation rates and assumptions used in the cited portion of the calculation were used solely to determine if condensation could be expected under fuel pool boiling conditions.

The conclusion of the calculation regarding time to boil the pool dry is not relevant to any operator action. Operator actions are based on maintaining normal pool level and temperature conditions. In any case, the cited nonconservatism would have a minor effect on the calculated 19 days to boil the pool dry, a result which is not used elsewhere in the design.

This discrepancy is not a valid deficiency, is not reportable and has no impact on plant operability.

October 21, 1992 Page 9 Rev. 1 Item D: Instrumentation "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 instrumentation available to the operator is not required to be qualified since the design basis loss of spent fuel pool cooling is not coincident wit the DBA LOCA conditions. This instrumentation is powered from an uninterruptible power supply and its'ssociated 1E AC source.

The minimum water level required per Tech Specs is below the weir elevation.

Since ESW makeup is provided to the pool the operators will know that when they see a rise in skinmer surge tank level the fuel pool level is at least as high as the weir. This provides a confirmation of adequate pool level without requiring access to the refueling floor.

Furthermore, on the basis of the discussion in item C above, access to the refueling f1oor is possible under all considered conditions. Therefore, it is possible to verify adequate fuel pool level visually from the refueling floor which is accessible from several locations.

While the available instrumentation is adequate for operator actions and meets the regulatory requirements of Reg Guide 1.13, improvements to the instrumentation have been recommended in the past and should be implemented.

This would enhance plant safety.

In conclusion, the existing instrumentation is adequate for performance of required operator actions for the current design basis and for scenarios not included in the current design basis.

This discrepancy is not a valid deficiency, it is not reportable and has no impact on plant operability,

October 21, 1992 Page 11 Rev. 1 (severe weather related), and .000665year (extremely severe weather related).

The probability of recovery from the LOOP within specified times was also calculated as follows:

/ m

~Time hr P Recover within T hrs 12.0 97.96X 24.0 99.53X 60.0 99.923K 75.0 99.953K Thus, it can be reasonably concluded that offsite power will be available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the initiating event.

The remaining factor in addressing restoration of fuel pool cooling is access to the reactor building. This issue is discussed under Item C above. Except for degraded core conditions access to the reactor building is feasible after the first twelve hours of the initiating event (Reference EP-548).

Notwithstanding the above basis, we have also considered the scenario where offsite power is not available and access to the lower reactor building elevations is restricted (based on FSAR chapter 18 contained source terms).

Under these conditions (representative of a degraded core event) access to the refueling floor remains available. Provision for fuel pool cooling is made through use of the plant fire protection system. Venting of Zone III via the filtered exhaust system is also possible for this scenario. While access to level and temperature instruments would be questionable it is possible to verify adequate pool level visually from the refueling floor which is accessible at several locations.

A Loss of Fue1 Pool Cooling Event Tree is attached to this evaluation to help guide the reader through- these various postulated scenarios.

In conclusion, for the design basis loss of fuel pool cooling the plant as currently designed and analyzed is acceptable. For other scenarios not specifically included in the design basis we have reasonable assurance that the effects of a loss of fuel pool cooling can be mitigated without adverse consequences on the plant.

This discrepancy is a valid deficiency. It is not a safety significant issue because we have established reasonable assurance that the effects of a loss of fuel pool cooling can be mitigated without adverse consequences on the plant and public health and safety and is therefore not reportable. The evaluation above also shows that this concern does not impact plant operability.

In consideration of this concern, additional analyses are warranted to further quantify the effects of evaporation and boiling conditions on the Zone III atmosphere and the potential transport of moist air to other locations in the reactor building for conditions outside of the current design basis.

October 21, 1992 Page 13 Rev. 1 n ineerin Re ort on oss of S ent Fuel Pool Coolin A detailed report, NE-092-002, is being prepared to document this evaluation in IM further detail. This report contains technical, input from several engineering groups and will provide a comprehensive set of references on this subject. The report will be completed by October 28, 1992.

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LOSS OP PUEL POOL COOLING EVENT EVALUATION NE-092-002 Rev. 0 ATTACHMENT 6 EXPECTED NUMBER OP PUEL PAILURES DURING A DBA LOCA tPLI-72696)

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October 20, 1992 G. D. Hiller A6-3 SUSQUEHANNA STEAM ELECTRIC STATION EXPECTED NUHBER OF FUEL FAILURES DURING THE DBA LOCA CCN 741087 FILE A7-8 P -7 696 Attachment I provides an evaluation of the expected number of fuel rod failures during the design basis LOCA based on the power uprate SAFER/GESTR analysis results. This information is provided in support of the assessments in response to EDR G20020.

Senior Project Engineer-Nuclear Nuclear Fuels Engineering AJR/el mel397a. a jr Attachment cc: J. M. Kulick A9-3 C. R. Lehmann A9-3 J. S. Stefanko A9-3 D. A. Hatchick A6-1 H. R. Hjaatvedt A6-3 NR File A6-2

ATTACHMENT I FORNAKC U G A OC During a LOCA there are two fuel rod failure mechanisms that have the potential to cause damage to the fuel rods. First, cladding rupture can occur following fuel rod ballooning when the, cladding temperature rises and the fuel rod internal pressure exceeds the external coolant pressure. Current LOCA analysis methodologies include empirical models to calculate the onset of ballooning and rupture. Second, if the fuel rod cladding reaches oxidation of high the temperatures, the cladding becomes embrittled by steam zircaloy cladding and can fragment upon introduction of the emergency core cooling water. This thermal shock phenomenon is precluded for events which remain below the 2200'F and 17K peak local oxidation criteria in 10CFR50.46 (Ref. NUREG-1230). The best estimate and licensing basis LOCA analysis results (Ref. NEDC-32071P) for Susquehanna using the GE SAFER/GESTR methodology were reviewed to determine the extent of fuel=damage that is expected during the design basis LOCA.

The SAFER/GESTR methodology (Ref. NEDC-23785P, Vol II) for calculating fuel rod ballooning. and rupture consists of a cladding hoop stress versus temperature curve below which cladding perforation is assumed not to occur.

This methodology was used for the Susquehanna power uprate project and is conservative for current conditions in that the analysis assumed higher LHGR and MAPLHGR values than the current plant operating limits. No ballooning was predicted for the best estimate DBA LOCA case and no fuel rod failures are predicted for the licensing basis DBA LOCA case (Ref. Telephone call with D. Pappone (GE) on 10/19/92). Therefore, fuel failure following ballooning is not expected to occur for the design basis LOCA.

Figure 1 shows the thermal shock failure curve and associated experimental data (Ref. NUREG-1230, Section 6. 14). For Susquehanna, the peak licensing basis cladding temperature is 1510'F (1094'K) and peak licensing basis local oxidation is less than 0.25K. These results are well below the failure boundary in Figure 1. Therefore, fuel failure as a result of oxidation embrittlement and subsequent thermal shock is not expected to occur for the design basis LOCA.

In summary, the potential fuel failure mechanisms and SAFER/GESTR LOCA analysis results have been reviewed. No fuel rod failures are expected to occur for the design basis LOCA.

ef rence

1. NUREG-1230, "Compendium of ECCS Research For Realistic LOCA Analysis,"

December 1988.

2. NEDC-32071P, "Susquehanna Steam Electric Station Units 1 and 2, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Hay 1992.

LOSS OF FUEL POOL COOLING EVENT EVALUATION NE 092 002 Rev 0 ATTACHMENT 7 USE OF UHS RITE RHR IN FUEL POOL COOLING ASSIST MODE tET 0870)

.MEMORANDUM TO: D. Kostelnik A6-3 DATE: October 27, 1992 FROM: J. Cajigas A6-3 COPIES: M. Mjaatvedt A6-3 Nuc. Records File A6-2 Eng. Tech. File A6-3 JOB: SSES NUMBER: ET-0870 FILE: P88-6 REPLY: No

SUBJECT:

USE OF UHS WITH RHR IN FUEL POOL COOLING ASSIST MODE The RHR system can be aligned in the fuel pool cooling assist mode in the event of a loss of fuel pool cooling condition. Following a DBA, use of the RHR system in fuel pool cooling mode requires two ultimate heat sink (UHS) spray divisions to be available. Loop "A" would support the "A" RHR system loops, in fuel pool cooling mode, for both fuel pools while loop "B" supports suppression pool/shutdown cooling at each unit: Since the design basis single failure of the UKS safety analysis is failure of one loop of sprays, this alignment is beyond the design basis of the spray pond.

The expected maximum normal fuel pool cooling heat load on the UHS is 13 million BTU/HR per pool or 26 million BTU/HR per RHRSW loop (or per UHS spray division).

UHS thermal analyses performed to date seem to indicate that this RHR fuel pool cooling heat load and DBA shutdown heat loads can be dissipated by two spray pond loops without exceeding the system design temperature. This scenario, however, has never been quantified by an analysis. Further, the current RHRSW operating procedure will need revision to provide sufficient guidance on how to obtain the optimum spray efficiencies required for this alignment.

Per the above discussion, it can be concluded that the UHS design should support the RHR fuel pool cooling mode provided this alignment is justified with UHS thermal analyses and the RHRSW operating procedures are revised to provide proper spray arrays alignment instructions.

uhsfp.wp5

LOSS OP PUEL POOL COOLZNG EVENT EVALUATZON NE-092-002 Rev. 0 ATTACHMENT 8 DRAZNAGE OP CONDENSATZON PROM THE 818 ~ ELEVATZON DURZNG PUEL POOL BOZLZNG (ET-0871)

Page 1 of 2 M E M 0 R A N D U M TO: D.G. Kostelik DATE: 10/28/92 FROM: K.G. Browning JOBe EDR G20020 NUMBER: ET-0871 COPIES: ET Memo File FILETS REPLY: None Required SUBJECT Drainage of Condensation from the 818'levation During Fuel Pool Boiling This memo is written in follow-up to our discussion regarding the possibility of condensate flowing down the reactor building stairwells during fuel pool boiling scenarios.

Problem During accident scenarios where the fuel pools are allowed to boil, condensation will occur on the wall and ceiling surfaces of the refueling floor elevation. Since the stairway access doors to the 818'levation (stairwells 101, 102, 201,

& 202) are not water tight, it is possible for this condensation to eventually drain down the stairwells to the basement of the reactor building.

Investi ation To address this concern, a review of the 818'loor Plan Drawings (C-243 thru C-255), along with the Area Drainage Drawings (P-25-7 thru P-34-7) was performed. The results are summarized below.

Resul ts The high point elevation for the refueling floor is 818'1".

In certain places, such as the steam dryer/separator pit and washdown areas, the high point is several inches higher since these areas are curbed. However, in general, area perimeters are defined by the high point elevation.

From this high point perimeters, the floor is sloped 2" to drains which are located at an elevation of 817'11". There are over 40 drains on the 818'looring which are generally located at the centers of these perimeters. Each of these drains has a diameter of 4" which provides for ample collection pathways for the fuel pool condensate. Since there is such a large cumulative collection area, flow of the condensate would be essentially unrestricted.

There are several hatchways on the 818'levati'on, including the train & truck bay hatches which are potential flow paths.

However, flow thxough these plugs would be minimal, if any, EPM-101B, Rev. 1

LOSS OP FUEL POOL COOLZNG EVENT EVALUATZON NE-092 002 Rev. 0 ATTACHMENT 9 SGTS PZRE DAMPERS (ET-0750)

ENGINEERING TECHNOLOGY MEMORANDUM TO: M. R. Mjaatvedt A6-3 DATE: October 20, 1992 PROM: D. J. Kohn A6-3 COPZES: G. D. Miller A6-3 J. Z. Agnew A6-3 J. E. Schleicher A1-2 S. E. Davis SSES JOB: NUMBER ET-0750 PZLE: A20-1, EDR G20020 REPLY No SUBJECT EDR G20020 Thi s confirms our discussion concerning the SGTS fire dampers between the Control Structure and the Unit 1 Reactor Building (P-line wall) .

Specification 323c indicates that all fusible links that operate fire dampers are rated at 160-165 F. The highest temperature should be kept at least 30~F below the fusi ble link rating or 130 F. If this temperature is too low, then the fusible links could be replaced with links rated at a higher temperature. Links rated at 212 F and.286 F are readily available. These provide for a maximum operating temperature of about 190~F od 250~F. Both would be acceptable since the fusible links would operate and close the fire dampers before allowing 325 F air to pass through the fire wall. Basically the acceptance criteria for a fire wall is to keep the cold side of the fire wall below 325~F.

The following dampers would require modification:

FPD-3-27-8-1sc FPD-3-27-8-3sc FPD-3-29-8-1 FPD-3-29-8-2 FPD-3-29-8-3 FPD-3-29-8-4 If I can provide additional information, please contact me.

c:fwp51)data)G20020

Attachaent 31 PP8L Memo from Glenn D. Miller to George T. Jones, "Revised Eval-uation of EDR G20020 - Spent Fuel Pool Cooling Issue", October 29, 1992 (PLI-72764)

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October 29, 1992 George T. Jones A6-2 SUSQUEHANNA STEN ELECTRIC STATION REVISED EVALUATION OF EDR G20020 SPENT FUEL POOL COOLING ISSUE PLI-72764 FILE A45-1A The attached revised evaluation of EDR G20020 is provided for your information. The revision accounts for a change to the engineering report reference, corrects a typographical error on page 10, expands the event tree to three pages and provides several graphs used as part of the PORC presentation on Monday October 26, 1992. The conclusions remain unchanged.

Glenn D. Miller cc: G. J. Kuczynski SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W. Simpson - A1-2 D. F. Roth SSES H. G. Stanley SSES J. M. Kenny - A2-4 J. S. Stefanko A9-3 F. G. Butler - A6-3 J. R. Miltenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records A6-,2 D. A. Lochbaum - Ener con ~

~ ~ ~

~~v I5E'8 w RQ +gg c+ ~

October 21, 1992 George T. Jones A6-2 SUS(UEHANNA STEN ELECTRIC STATION EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PL I-72711 FILE 45-lA The attached evaluation of EDR G20020 is provided in response to your memo PLI-72640. This evaluation was prepared by myself and a team of engineers working on the action plan to resolve the subject EDR.

In the course of reviewing this issue in detail, we have concluded that seven of the nine identified discrepancies are not valid deficiencies. This is explained in detail in the evaluation. The two remaining discrepancies are valid deficiencies but are not considered safety significant nor reportable.

I believe that the technical basis on several of these issues has been clarified considerably in the course of the past week. Therefore, I suggest providing this evaluation to Nuclear Regulatory Affairs for reconsideration of the reportability aspects.

We are continuing with the remaining action items as requested. A detailed engineering design report and justification for interim operation will provide more detail than contained in the attached evaluation. . We are scheduled to meet with PORC on Monday October 26, 1992 at 2:30 pm to review this issue. We are working with Systems Engineering, Operations and NRA-Compliance with respect to potential compensatory measures.

Glenn D. Miller CC: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt A6-3 M. W. Simpson - A1-2 D. F. Roth SSES H. G. Stanley - SSES J. M. Kenny A2-4 J. S. Stefanko A9-3 F. G. Butler - A6-3 J. R. Miltenberger - A6-1 D. C. Prevatte A6-3 Nuclear Records - A6-2 D. A. Lochbaum - Enercon

October 21, 1992 Page 1 Rev. 1 Evaluation of EDR G20020 This document contains an evaluation of the discrepancies documented in EDR G20020, "Loss of Spent Fuel Pool Cooling Event Design Discrepancies."

Conclusions of the author with respect to reportability of these concerns and operability impact on SSES are also provided.

Desi n Basis The design bases for the Fuel Pool Cooling System are found in FSAR section 9. 1.

The portions of the design basis relevant to fOR G20020 are as follows:

Maintain the fuel pool water temperature below 125F under "normal maximum heat loads" defined as 12.6 MBtu/hr (equivalent to a typical fuel cycle discharge schedule which fills the fuel pool, last quarter core offload at 6.7 days after shutdown).

2. Maintain fuel pool water temperature at or below 125F during the "emergency heat l.oad" condition of 32.6 HBtu/hr (equivalent to a full core offload 10.5 days after a shutdown following a typical fuel cycle discharge schedule which fills the fuel pool) utilizing the RHR system (with or without normal fuel pool cooling) for fuel pool cooling. This mode of operation applies "during periods of higher than HNHL generation in the fuel pool, eg., storing of a full core of irradiated fuel shortly after shutdown". The RHR system is used under these conditions to assist the FPCCS in dissipating the decay heat. Thus, any heat load in excess of 12.6 HBtu/hr is considered to be within the design basis for the RHR FPC assist mode of operation.

3. Redundant Seismic Category I ESW connections to each pool are provided to allow for makeup of evaporative losses in the event of failure of the FPC system. The conditions are bounded by a fuel pool time-to-boil analysis based on the same typical fuel cycle discharge schedule as in basis ¹1 except the time after shutdown is 10.5 days instead of 6.7 days resulting in a heat load of 9.8 HBtu/hr. (This explains the difference between the two different heat loads, ie., 12.6 HBtu/hr for basis ¹1 and 9.8 HBtu/hr for basis ¹3. This is not a discrepancy.) The ESW makeup line is sized on the basis of this calculation (Reference FSAR section 3.13).

4. The cause of the Loss of Spent Fuel Pool Cooling event is stated to be a seismic event.

5. All piping and equipment shared with or connecting to the RHR intertie loop are Seismic Category I and can be isolated from any piping associated with the non-Seismic Category I fuel pool cooling system.

Evaluation of Discre ancies Noted in EDR G20020 EDR G20020 describes nine discrepancies relating to the loss of spent fuel pool cooling event. This discussion will summarize each issue. The reader is

October 2), )992 Page 2 Rev. )

referred to the complete text of the EDR.

General Statement The introductory paragraph of the EDR states: "...the design provision for the loss of spent fuel pool cooling event is to permit the fuel pool to boil and maintain its water level above the fuel through makeup from the ESQ system. This design provision is necessary because 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."

As stated in design basis ¹5 above the RHR fuel pool cooling assist portion of the piping is designed to seismic category I requirements. No credit however is taken for this mode of operation in the fuel pool boiling analysis in the FSAR.

Credit is taken for this mode for emergency heat load situations as defined by basis ¹2.

Discussion of EDR Items A throu h I In order to discuss and evaluate each of the nine discrepancies listed in the EDR it will be more logical to review them in a different order. Items E & F both relate to the time-to-boil calculations and will be reviewed first followed by items G through I, which are related to the time-to-boil concern. Items C & 0 involve operator action considerations and will be discussed next. Finally items A & B relating to the evaporation effects will be discussed.

October 21, 1992 Page 3 Rev. 1 tern: Anal tical Time-to-Boil "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."

As stated in basis O'I the maximum normal heat load is 12.6 HBtu/hr. As stated in basis k2 the time-to-boil analysis is based on a heat load of 9.8 NBtu/hr.

These two design bases are in fact consistent and are based on the same "typical fuel discharge schedule" and refueling outage scenario. The difference in the heat load is due solely to the time after shutdown assumed for purposes of establishing the design basis.

Focusing on the time-to-boil analysis, a time after shutdown value of 10.5 days is used. This is the time at which it is assumed that refueling is completed and the reactor cavity to fuel pool gates are reinstalled. Prior to that point the additional water stored in the reactor cavity is also available as a heat sink and the RHR system is available for fuel pool cooling. For times greater than 10.5 days the appropriate heat load will be even lower than the analyzed value of 9.8 NBtu/hr. For the SSES Unit 2 5RIO the time from reactor shutdown to fuel pool gates installed was 38 days. The decay heat in the Unit 2 pool at that time is calculated to be 5.65 HBtu/hr (Reference NE-092-002). The corresponding time-to-boil is 45 hours.

The EDR goes on to discuss other calculations which result in different heat loads using various assumptions. Calculation NFE-B-NA-053 was performed by Nuclear Fuels to account for actual fuel discharge history and future offloads accounting for power uprate conditions. The fuel pool heat load versus time curves obviously will increase subsequent to power uprate, however, these curves do not apply to the existing design. As long as the calculated decay heat is less than 9.8 HBtu/hr at the point where the fuel pool gates are reinstalled the original design basis time-to-boil calculation is still valid.

Calculation H-FPC-009 determined time-to-boil conditions post power uprate. This calculation shows that the time-to-boil for the design basis heat load of 9.8 HBtu/hr is slightly greater than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

In conclusion, the design basis for the time-to-boil condition is established by the 9.8 HBtu/hr value used in the original calculations. This design basis is met by planning the outage so that the fuel pool is not isolated from the reactor cavity or the RHR system prior to a point in time where the actual heat load is 9.8 HBtu/hr or less.

This discrepancy is not a valid deficiency, is thet efore not reportable and has no impact on the operability of the plant.

October 21, 1992 Page 4 Rev. 1 Item F: Time-to-Boil for Emer enc Heat Load "The analytic'al 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."

As discussed above, the time-to-boil conditions apply to configurations where the spent fuel pools are isolated from the reactor cavity (ie., non-refueling configurations). As is correctly stated in the EDR, current practice is to fully offload the core during each refueling outage. Specific calculations are performed by Nuclear Fuels to determine the ability of the FPC system to remove the combined decay heat of the cross-tied refueling pools. Tests are also conducted to determine that the actual heat removal capability exceeds the actual fuel pool heat loads during the outage (Reference TP-235-011). Normally the reactor cavity is maintained flooded and cross-tied to the fuel pools; One loop of Core Spray is always operable in this configuration. One division of RHR is maintained in shutdown cooling mode except for a brief period required for the common RHR system outage window.

Design basis ¹2 states that heat loads in excess of the NNHL are considered to be emergency heat loads. The design of the RHR system to assist the FPC system during emergency heat load conditions assures that fuel decay heat is removed.

No time-to-boil calculation for this configuration is required since the RHR system will be in operation or available. At any rate, such a calculation should consider the effect of the additional water inventory available from the flooded reactor cavity, cask storage pit and dryer and separator storage pool which are all cross-connected during this time. Hakeup inventory is also available from Core Spray and the RHR system is normally in-service except for the common RHR system outage window.

In conclusion, no time-to-boil analysis is required for the emergency heat load desigri basis. Single failures of the RHR system are not required for this design basis for the emergency heat load (Reference SRP 9. 1.3).

This discrepancy is not a valid deficiency, not reportable and has no impact on plant operability.

October 21, 1992 Page 5 Rev. 1 Item G: Radiolo ical Release Calculation for Boilin S ent Fuel Pool "The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates."

This discrepancy is directly related to the heat load assumed for the time-to-boil analysis. The evaporation rate used in the dose calculation is based on a heat load of 9.8 HBtu/hr which is the design basis heat load for the time-to-boil calculation. keat loads in excess of 9.8 HBtu/hr obviously result in higher evaporation rates. Since the discussion under Item E above establishes that 9.8 NBtu/hr is the correct original design basis and still bounds current operation there is no discrepancy in the offsite dose calculation. It uses an evaporation rate consistent with the design basis heat load.

This discrepancy is not a valid deficiency, not reportable and has no impact on plant operability.

October 21, 1992 Page 6 Rev. 1 Item H: Nonconservative Activit Terms "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 approximately 230 bundle reloads which will increase to approximately 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 original radiological release analysis as referenced above is conservative for the following reasons:

(I) the activity levels used as a source term are based on IX failed fuel.

All of the failed fuel rods are assumed to be in the offloaded batch of 184 fuel assemblies. Therefore, increased batch sizes will not increa'se the amount of the source term used in this analysis.

(2) the activity levels used for the iodine source term are based on saturation level inventories for a core operating at 3440 NWt for one thousand days. Therefore, the fuel cycle length will not affect the source term.

In conclusion, the offsite dose calculation remains valid.

This discr epancy is not a valid deficiency, not reportable and has no impact on plant operability.

October 21, 1992 Page 7 Rev. 1 Item I: Anal sis for Hax Time Prior to Hakeu "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."

The purpose of the referenced calculation was to determine refueling floor atmosphere conditions under various operating modes. The evaporation rates and assumptions used in the cited portion of the calculation were used solely to determine if condensation could be expected under fuel pool boiling conditions.

The conclusion of the calculation regarding time to boil the pool dry is not relevant to any operator action. Operator actions are based on maintaining normal pool level and temperature conditions. In any case, the cited nonconservatism would have a minor effect on the calculated 19 days to boil the pool dry, a result which is not used elsewhere in the design.

This discrepancy is not a valid deficiency, is not reportable and has no impact on plant operability.

October 21, 1992 Page 8 Rev. 1 Item C: Manual ESW Valve Actions "The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be possible."

In-plant post-accident radiation levels are analyzed for SSES to the requirements specified in NUREG-0737. This document requires that post-accident radiation levels be determined for, purposes of vital area access by plant operators to perform short-term first priority actions. It specifies that radiation levels be determined on the basis of contained sources, and core damage source terms equivalent to those used for 10CFR100 calculations. These assumptions are clearly based on degraded core conditions which are beyond the design basis LOCA.

Airborne radioactivity sources from containment leakage are required to be analyzed For environmental qualification of equipment but not for personnel access.

A review of FSAR chapter 18 shows that access to the equipment necessary to provide makeup to the fuel pool from ESW is restricted for the approximately the first 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> following the design basis event (Figure 18.1-9). This analysis is based on a conservative source term equating to IOOX fuel damage resulting from core melt conditions as originally utilized for offsite dose calculations used to determine plant siting adequacy. These source terms were based on experiments involving heated irradiated uranium dioxide pellets.

An evaluation of actual fuel thermal response during design basis accidents results in no predicted fuel failures (Reference PLI-72696). Thus, the source term resulting from the DBA LOCA would only be equivalent to the radioactivity present in the reactor coolant as a result of normal operations (allowing for fuel defects as permitted by Technical Specifications). To bound the potential effects of a design basis accident, a realistic yet conservative analysis using an assumed IX fuel damage resulting from core degradation under LOCA conditions was performed (Reference EP-548) and concludes that access to equipment necessary to mitigate the effects of a loss of fuel pool cooling following a DBA LOCA is assured.

In conclusion, post-accident operator actions are viable for all potential scenarios under consideration, for both the current design basis and those outside of the current design basis.

This discrepancy is not a valid deficiency, is not reportable and has no impact on plant operability.

October 21, 1992 Page 9 Rev. 1 Item 0: Instrumentation "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 instrumentation available to the operator is not required to be qualified since the design basis loss of spent fuel pool cooling is not coincident with the DBA LOCA conditions. This instrumentation is powered from an uninterruptible power supply and its'ssociated IE AC source.

The minimum water level required per Tech Specs is below the weir elevation.

Since ESW makeup is provided to the pool the operators will know that when they see a rise in skimmer surge tank level the fuel pool level is at least as high as the weir. This provides a confirmation of adequate pool level without requiring access to the refueling floor.

Furthermore, on the basis of the discussion in item C above, access to the refueling floor is possible under all considered conditions. Therefore, it is possible to verify adequate fuel pool level visually from the refueling floor which is accessible from several locations.

While the available instrumentation is adequate for operator actions and meets the regulatory requirements of Reg Guide 1. 13, improvements to the instrumentation have been recommended in the past and should be implemented.

This would enhance plant safety.

In conclusion, the existing instrumentation is adequate for performance of required operator actions for the current design basis and for scenarios not included in the current design basis.

This discrepancy is not a valid deficiency, it is not reportable and has no impact on plant operability.

October 21, 1992 Page 10 Rev. I Item A: Reactor Buildin Desi n Heat Loads "Reactor building design heat loads do not account for the boiling spent fuel pool event."

The reactor building temperature analysis is performed for DBA LOCA conditions.

The presumption of item A is that the spent fuel pool will reach boiling conditions prior to restoration of fuel pool cooling subsequent to a LOCA. The existing temperature analysis does account for the sensible heat load from the fuel pool at 212F.

A loss of spent fuel pool cooling event can result from several conditions. The design basis condition is a seismic event as analyzed in the FSAR. The Fuel Pool Cooling system is not designed for seismic loads. In this case, the Fuel Pool Cooling system is assumed to be lost. An evaluation of the plant response shows that several methods are available to assure that the spent fuel remains cooled.

These include: (I) the RHR system can be used to cool the fuel pools with alternate shutdown cooling of the reactor using Core Spray and RHR for suppression pool cooling; or (2) allow the fuel pool to boil with makeup supplied by ESW with consideration of either SGTS operating on Zone III or providing a vent path from Zone III. offsite for SGTS.

If available, normal reactor building ventilation would be credit The dose analysis takes no used to provide cooling and venting of the Zone III atmosphere. Under any of these scenarios transport of moist air to other portions of the reactor building would not occur.

This scenario is the design basis for loss of fuel pool cooling.

Other scenarios not included in the design basis include LOCA and LOOP events, and combinations thereof. The time frame for consideration of operator actions is based on reasonable expectations for the time-to-boil condition. As stated previously, for the current operating practice, the fuel pool heat load prior to reactor restart is approximately 5.65 HBtu/hr. Time to boil under this condition is on the order of 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />. Note that this is the shortest possible time-to-boil for the current fuel cycle. With the pools cross-tied the time-to-boil is greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

For a LOCA scenario, the FPC system will be lost initially due to the Aux Load Shed provisions. Although the Fuel Pool Cooling system and other non-safety related systems are not specifically analyzed for the effects of hydrodynamic loads it is expected that they will be able to perform their normal functions following a broad spectrum of design basis events. Credit for these systems is not needed to meet the design basis, however, plant operators will utilize any equipment available to them during emergency situations. Therefore, in the course of evaluating the effects of a DBA LOCA on the fuel pool cooling system, we acknowledge the availability of normal plant systems in responding to the emergency.

Independent of the LOCA condition, offsite power is needed to restore normal cooling systems. The SSES Individual Plant Evaluation considered loss of offsite power (Reference IPE Appendix F). The IPE conservatively estimated the incidence of LOOP to be .04/year (plant related), .008/year (grid related), .00807/year

October 21, 1992 Page ll Rev. I (severe weather related), and .000665year (extremely severe weather related).

The probability of recovery from the LOOP within specified times was also calculated as follows:

/

T~ime hr P Recover within T hrs 12.0 97.96X 24.0 99.53X 60.0 99.923X 75.0 99.953X Thus, it can be reasonably concluded that offsite power will be available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the initiating event.

The remaining factor in addressing restoration of fuel pool cooling is access to the reactor b'uilding. This issue is discussed under Item C above. Except for degraded core conditions access to the .reactor building is feasible after the first twelve hours of the initiating event (Reference EP-548).

Notwithstanding the above basis, we have also considered the scenario where offsite power is not available and access to the lower reactor building elevations is restricted (based on FSAR chapter 18 contained source terms).

Under these conditions (representative of a degraded core event) access to the refueling floor remains available. 'rovision for fuel pool cooling is made through use of the plant fire protection system. Venting of Zone III via the filtered exhaust system is also possible for this scenario. While access to level and temperature instruments would be questionable it is possible to verify adequate pool level visually from the refueling floor which is accessible at several locations.

A Loss of Fuel Pool Cooling Event Tree is attached to this evaluation to help guide the reader through these various postulated scenarios.

In conclusion, for the design basis loss of fuel pool cooling the plant as currently designed and analyzed is acceptable. For other scenarios not specifically included in the design basis we have reasonable assurance that the effects of a loss of fuel pool cooling can be mitigated without adverse consequences on the plant.

This discrepancy is a valid deficiency. It is not a safety significant issue because we have established reasonable assurance that the effects of a loss of fuel pool cooling can be mitigated without adverse consequences on the plant and public health and safety and is therefore not reportable. The evaluation above also shows that this concern does not impact plant operability.

In consideration of this concern, additional analyses are warranted to further quantify the effects of evaporation and boiling conditions on the Zone III atmosphere and the potential transport of moist air to other locations in the reactor building for conditions outside of the current design basis.

October 21, 1992 Page 12 Rev. I Item B: Im act of ESW Hakeu Water "The impact of the ESW makeup water to the spent fuel pool on equipment in the reactor building has not been evaluated."

The analysis under item A above applies to this issue as well. This evaluation shows that with the current plant design and for existing design basis conditions the effects of a loss of fuel pool cooling are 'acceptable.

This discrepancy is a valid deficiency. As with item A it is not a safety significant issue and is not reportable. The evaluation above also shows that this concern does not impact plant operability.

In consideration of this concern, additional analyses are warranted to further quantify the effects of evaporation and boiling conditions on the Zone III atmosphere and the potential transport of moist air to other locations in the reactor building for conditions outside of the current design basis.

October 21, 1992 Page 13 Rev. 1 En ineerin Re ort on Loss of S ent Fuel Pool Coolin A detailed report, NE-092-002, is being prepared to document this evaluation in further detail. This report contains technical input from several engineering

)~

groups and will provide a comprehensive set of references on this subject. The report will be completed by October 28, 1992.

Design Basis Decay Heat Last Batch Offload to Single Fuel Pool 14 13 Max Normal Heat Load 12 11 Time-to-Boil Gale 10 9

8 O

D U2 5RIO 10 20 25 30 35 40 Days After Shutdown

Actual Decay Heat U2 Cycle 5 1/3 Core, Pool Isolated 14 50 13 Time-to-Boil 45 12 11 40 10 35 8 Decay Heat O 30 CI 25 20 10 15 20 25 ~

30 35 40 45 Days After Shutdown

Figure A - Loss of Fuel Pool Cooling Due to Seismic Event Sefsntc Evert.

Loss of Fuel Pool Cooling.

Zone III Isoided with No SGTS nmng?

Use RHR FPC assist with Use RHR FPC assist with Attendee SDC If available, Alternate SDC lf avalable, otherwise:. otherwise:.

Allow to boil with ESW maketp Allow to boil with ESW make. 4 Zlllverted.

If SGTS sftldown Is reqrsrerL them.

Vert ZIIL

Figure B - Loss of Fuel Pool Cooling Due to LOCA lOCA Erst loss d fPC, Yes No Scarce T>>ss R stall Yss No Ntsls Pow>> Arel&>) ONske Pows Aratabhf No Asst>>t WAC & cestae Use fHIFPC as>>sta fkc ldkhrIAccccshtcf fhfkESWAccesshtsf FPC wlh ha >> V cyst>>as doe to bol wlh ESW t scalakts, cth>>whr ssaksy & Lllred ed.

Nest<<s FPC<<khn>>>>V Alow to bol <<lh fk o Uss flS FPC assht a Alowte bol wlh Eke Usa FHI FPC asVst a

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I clat>>as aralatk, Rdscthn sskeoy & LNI ~ low to bol wlh ESW Rotsctkm soakage & Lll slow to bol wlh ESW cth>><<Is c. rstld, aaksup & Lllrsssk raced tacking I Lhlrertsd.

Use lHIFPC asshta ato<< to bot wlh ESW cask<<y & LNI rslcd.

Figure C - Loss of Fuel Pool Cooling Due to Other Causes Other Everts.

Loss of Fuel Pool Coolintf.

Offslte Power Availablel Use RHR FPC assist with Restart RB HVAC. Alternate SOC If avalabte, otherwise:.

Restore FPC withnormal Allowto boll with ESW makeup slrstems. 5 tillverled.

AttachIIent 32 PP8L Memo from David A. Lochbaum and Donald C. Prevatte to George T. Jones, "Position on EDR G20020 and Planned Actions", November 2, 1992 (PLI-72283)

Note: This memo provided PP8L with an update as to the authors'ntentions with respect to the declaration made on October 9, 1992 (Attachment 19) on a report to the NRC on November 2, 1992. The authors deferred a report to the NRC pending PP8L's promised formal report of their own on this matter.

November 2, 1992 George T. Jones A6-2 SUSQUEHANNA STEAM ELECTRIC STATION POSITION ON EDR G20020 AND PLANNED ACTIONS PLI-72783 FILE A45-1A In a meeting with you on October 9, 1992, we submitted a letter again expressing our deep concern reqarding the boiling spent fuel pool safety issues we raised in EDR G20020. After months of unsuccessful attempts to have these issues properly addressed per PP&L procedures and federal regulations, we formally requested of you that the screening, reportability, and operability processes required by procedures be completed, the attendant documentation be provided to us, and that these concerns be formally communicated to the NRC in accordance with the Code of Federal Regulations by November 2, 1992. We stated that should these actions not be taken by that, date, or should they be technically inadequate or incomplete, it was our intent to report these concerns to the NRC, also as required by federal regulations.

Since that meeting, significant efforts have been mounted by

,PP&L to address these concerns, significant progress has been made, and we have been assured by you that a formal report will be made to the NRC.. The latest manifestations of these efforts are a revision to Mr. Miller's October 21 evaluation which, in essence, still maintains that our concerns are safety significance and not reportable, and a report, NE-of'o 092-002, Rev 0, Loss of Fuel Pool Cooling Event Evaluation for EDR P G20020, 10-29-92, which is substantially more balanced, more complete, and better documented than any previous assessments, but with which we still have major disagreement on a number of basic technical points.

Therefore, in recognition of this continuing progress and ycur assurance, we will not make our report to the NRC on November 2. However, our fundamental positions with respect to the major technical concerns and the obligation for prompt reporting have not changed. We believe that today, within the NRC mandated design basis requirements, most 1f not all of the safety-related systems and functions in the reactor buildings are "In an unanalyzed condition that significantly compromises plant safety, in a condition that is outside the design basis of the plant, and in a condition not covered by the plant's operating and emergency procedures".

We know that the above cited report is the primary product of the efforts of the last three weeks, and we expect that in generating your report to the NRC you will give it heavy I

consideration. We urge you to review considering the following comments.

it very carefully, Although the report provides a wealth of information which could be used as justifications for interim operation, we strongly disagree with its conclusion that "...it is possible in all cases temphasis added) to provide adequate pool cooling and protect safety related equipment in Zones I and ZX from the environment produced from a boiling pool", as well as many of the positions the report takes to support these conclusions. Although we agree it would be possible in some cases, it would not be possible for the design basis case of LOCA/LOOP and many lesser cases. Additionally, the report's conclusions are based on expectations for future analyses, future modifications, and future procedure changes, not the plant as it exists today. The evaluations of operability and reportability are required by law to be based on the plant today, not our expectations for the future.

We believe that the report's conclusions are incorrect for the following reasons:

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

2. They place heavy reliance on modifications to the plant 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 accomplished or even developed.

6. They place heavy reliance on operator actions during a LOCA when there is already heavy reliance on operator, actions and monitoring. These additional actions must be performed under extremely adverse environmental conditions in the reactor building.

7. The conclusions are based on assessments of operator accessibility to the. reactor building which in turn.

are based on assumptions of core damage which are unreviewed by the NRC and are substantially less than the assumptions required by NUREG-0737 and our licensing basis reflected in Chapter 18 of the FSAR.

Additionally, the accessibility position taken in the report with respect to airborne radiation is inconsistent with NUREG-0737; 10CFR50, Appendix J;

actual Appendix J test results for SSES; the design of other plant system systems (e.g. secondary containment and SGTS); and common sense. For NRC mandated DBA conditions,= the reactor building is inaccessible for days into the accident.

8. They rely on probability arguments, arguments which may be acceptable for an IPE or a JIO, but, which are not acceptable substitutes for NRC mandated design bases, unless they are reviewed and approved by the NRC. These have not been.

9. In numerous areas, the report's conclusions are not consistent with the facts presented, e.q., the report concludes that Zone III ventinq zs acceptable; the supporting documentation, by contrast, shows that the 10CFR100 and 10CFR50, Appendix A, Criterion 19 allowables for offsite and control room doses respectively are exceeded.

Our more detailed specific comments on the report are provided on the attached sheets.

We would like to express our appreciation to you and to all who have worked hard in addressing these concerns. We have come a long wayi we still have some distance to go; and like you, we are committed to going that distance. As always, we are at your service..

ave A. Loc um ona . reva e

cc: C. A. Myers A2-4 G. D. Mailer A6-3 R. R. Sgarra A2-4 J. M. Kenny A2-4 J. E. Agnew A6-3 D. F. McGann SSES S&A-4 G. J. Kuczynski SSES H. G. Stanley SSES J. R. Miltenberger A6-1 H. W. Keiser TW-16 R. G. Byram . A6-1 W. R. Corcoran 21 Broadleaf Circle Windsor, CT 06095 J. S. Kemper 115 Polecat Road Glen Mills, PA 19342 R. L. Doty A9-3 A. F. Zorfida SSES A. R. Sabol A2-5 W. R. Licht A6-1 J. S. Stef anko A9-3 F. G. Butler A6-3 J. A. Zola A6-3 M. R. Mjaatvedt A6-3 C. A. Boschetti SSES T. J. Sweeney SSES G. D. Gogates SSES M. J. Manski Enercon J. D. Richardson Enercon

COMMENTS OF DAVID A. LOCHBAUM AND DONALD C. PREVATTE ON NE-I 0 9 2 0 0 2 g REV 0 g LOSS OF FUEL POOL COOL NG EVENT EVALUAT ON I FOR EDR G20020, 10-29-92 Abstract, first sentence, EDR G20020 also expressed concerns with the existing FSAR discussion for the effect seismic event of loss of fuel pool cooling from (i.e.,boil-off on reactor building heat loads and effects and spillover).

2 Page 3, the assumption of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to boil is improper given that EDR G20020 challenges the basis forChapter that time

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and EDR G00005 questions the accuracy of FSAR 9 and FSAR Append>.x 9A.

1st sentence, even though the FSAR assumes a

3. Page 5, seismic event causes the loss of fuel pool cooling, loss it is only one of many mechanisms.

4. Page 5, 2nd sentence, although the FSAR postulates the loss during a refueling outage, it can occur at any time.

5. Page 5, second paragraph, the current heat load is not lower than 9.8 BTU/hr until well beyond 10.5 days because SSES performs full core offloads each refueling outage.

6. Section 3.2, 2nd paragraph, although the instruments are UPS powered, they are not 1E qualified, not seismically qualified , and not environmentally qualified the for LOCA or for the boiling spent fuel pool. Therefore, instruments would not necessarily remain operational.

7. Section 3.2, 3rd paragraph, the skimmer surge tank level is not a reliable indicator of fuel pool level. Page 6, is the procedure to drain current the skimmer surge tank described here a part of plant procedures and operator training?

8. Section 3.3, this section confirms our concern in EDR G20020. Although it refers to actions that can be taken, these are not part of the present plant procedures.

9. Section 3.4, second paragraph, although it NUREG-0737 does not require consideration of airborne is true that sources due to leakage of systems outside containment, this is aimed at the ECCS systems which operate post-LOCA It is and form an extension of the containment boundary.

not necessary to consider their leakage because they are normally filled and pressurized, and therefore they would not be expected to leak. Additionally, the leakage involved would be liquid leakage and would be a relatively small contribution to the total airborne dose.

However, NUREG-0737 does not exclude the leakage through containment isolation valves which are closed post-LOCA.

The leakage of these plus the Type B penetration leakage constitutes the leakage of concern which is consistent with the requirements of 10CFR50, Appendix J. It is a real and significant valve as our tests show, inandFSAR not be ignored. Additionally, PP&L response it may Chapter 18 to the NUREG requirement recognizes this leakage, it just doesn't account for it.

Page 8, first full paragraph, FSAR Chapter 18 does not that the reactor building will be inaccessible for several days .following the LOCA it plainly states this.

~im v~1 On page 8, second paragraph, from a "design basis perspective " the Appendix J leakage is required to be considered, not just contained sources. Appendix J leakage also doesn't include leakage from the water filled ECCS systems that operate post-LOCA.

Page 8, third paragraph, the model uses the designis basis it also leakage is not just from Type B penetrations; from Type C penetrations which is consistent with the NUREG 0737 requirements.

Page 8, fourth paragraph, access to the refueling floor is not possible post-accident for the design basis accident due to airborne radiation levels xn the hundreds of R/hour per PP&L analyses in the files.

Page 9, first paragraph, first sentence, the assumption of 1% fuel cladding damage is not conservative; this is the allowable fuel cladding damage for normal operation.

This assumption basically allows for no damage as a result of the accident. This is non-conservative.

Second, the RHR system operating in the fuel pool assist mode is unanalyzed in the first place, and is particularly unanalyzed while at the same time responding to a LOCA. Additionally, the effects of this mode of operation in a LOCA on the RHRSW system, the ESW system and the spray pond are unanalyzed.

Third, the exposure to the operator in manipulating the ESW valves for fuel pool makeup is unanalyzed.

Section 3.5, this section confirms our point, the RHR fuel pool cooling assist mode is unanalyzed for post-LOCA response Section 3.6, this paragraph is true. However, for design basis for the safety function of fuel pool cooling, this is not adequate.

Section 3. 7; there are several problems with this section.

e is through a manual First, the makeup flow to the pool not calibrated and for which throttling valve which is instructions there are not throttling in the procedures.

The procedures only require that the valve be opened.

Unless otherwise stated, that means fully opened. At the full open position, the valve would likely flow considerably more than 30 gpm; the number that has been used for previous analyses has been 60 gpm/pool. At this flow rate, over 30 days, this is 5.2 million gallons of water more than the sumps can hold.

a Second, for the design basis accident condition, the water cannot be purged out of the building for several reasons: The pumps are non-1E powered, non-seismically qualified, non-environmentally qualified, etc.; the same is true of the radwaste processing system; the radwaste systems are not designed to handle this volume; the radwaste systems are not designed to handle DBA LOCA contaminated water; and finally, this is an unreviewed safety question.

Third, this does not address all of the other flooding effects, e.g., possible structural failure of HVAC ductwork, blockage of HVAC ductwork, spillage from HVAC ductwork on safety-related components, etc.

13. Section 3.8 addresses recovery. It is predicated on access to the areas, significant operator actions in a potentially hostile environment for the DBA LOCA, use of non-safety related systems, modifications not yet made or even designed, and unanalyzed, nonexistent procedures.

This may be possible, but is clearly outside many of the regulatory and design basis requirements.

14. Section 3.8 takes credit for operating the filtered exhaust of the normal HVAC system. It is unlikely that this system would provide adequate filtering function for pool boiling conditions due to wetting of the charcoal.

This is an unanalyzed condition.

15. Page 13, the report does not consider the seismic event.

This is in conflict with 10CFR50, Appendix A, Criterion 2 and FSAR Section 6.2 which describes the worst case scenario for containment functional design as LOCA/LOOP with concurrent SSE. The report's analysis non-conservative and inconsistent in that all consequences of SSE (e.g., loss of fuel pool cooling) are not considered.

16. Page 14, Event fl, this description requires both loops of RHR to be available. It failure in one of the loops.

does not address single

17. Page 14, Event g1, LOCA unit, the last statement is not necessarily true. Current design basis calculations for

the reactor building temperature require the fuel pool cooling system to be out of service along with the other non-1E powered loads in the reactor building. The service water system is also non-safety related and cannot be taken credit for in the design basis.

18. Page 14, non-LOCA unit, RHR in the fuel pool cooling mode as well as RHRSW, ESW, and the spray pool are unanalyzed for the condition described . Use of the condensernon- as a heat sink is outside the design basis since safety related and requires non-1E power.

it is 19 Page 15, concern ¹1, the last statement is not necessarily true for design basis requirements, even without a LOOP.

20. Page 15, concern ¹2, ESW could not fulfillthe purpose of makeup to the pool, due to inaccessibility of the valves.

21 Page 15, concern ¹3, this concern states that for flows

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greater than 2000 gpm the skimmer surge tank runs dry.

Operating procedures require a flow rate of 5000-6000 gpm for the RHR fuel pool-cooling assist mode. It would appear that this is more than just a concern, it appear to be a major discrepancy in the stated plan.

would 22 Page 16, third paragraph, if actions are not taken immediately to isolate Zone III, for the DBA, then all zones become inaccessible, even the non-LOCA unit zone, due to airborne contamination. It is unreasonable to expect that the operator can take the necessary actions to isolate Zone III in time before the contamination is

'3.

spread to all zones with the current design.

Page 16, last paragraph, this paragraph discusses exposures being within legal limits for the case that does not even represent NUREG 0737 requirements.

However, of ALARA.

it does not appear not to consider the concept

24. Page 19, next to last paragraph, this states that if RHR is not available for fuel pool cooling, boiling No.'t with makeup is allowed if Zone III is vented. this point in the accident, Zone III will already'e highly contaminated from the accident since up until this time, the three zones were cross connected. Therefore, be impossible to,gain access, and even if it access were will possible, the release of the unfiltered airborne radiation from Zone III would likely exceed the 10CFR100 limits.

25. Event ¹3, page 20, fuel pool boiling may be permitted by current procedures, but it is an unanalyzed condition.

Although it may not rely on non-safety related systems for the non-boil scenario in this case as pointed out, it does involve an analyzed, untested mode of RHR, RHRSW, ESW and spray pond operation.

26. Event ¹4, page 21, this .description dismisses the LOCA/

LOOP for the outage situation as "sufficiently small so as'not to be cons>dered." This is a probability argument, appropriate for an IPE or JIO, but not appropriate for the design basis except with the approval of.,the NRC.

27. Section 5.1, page 22, we do not agree that "...

possible in all cases to provide adequate pool cooling it is and protect safety related equipment in Zones I and II from the environment produced from a boiling spent fuel pool."

28. Section 5.1, Design Basis, pools it is will boil for the designcreate concluded that the fuel basis case. For today' plant, this would appear to a reportable condition in light of short term action ¹1 on page 24 which shows the SGTS to fail if the pool boils.

29. Section 5.1, Desiqn Basis, last paragraph, this concedes that "... if significant airborne radiation is will be necessary to evaluate Zone III venting present against it mixing zones and usinq SGTS." If the airborne neither option is available. If venting is chosen, the is high, operator gets too much exposure and the 10CFR100 limits are exceeded. If mixing 1s chosen, the reactor building safety systems plus SGTS are not qualified and are likely to fail. For these design basis conditions, we do not meet the design basis requirements.

30. Page 22, for the "realistic" case, even design basis conditions exist, it states if these non-that plant procedures must be changed to have a workable system.

This would appear to meet the test, of reportabxlity (required, not voluntary) under 10CRF50.72, paragraph (b)(ii)(C),-by being "In a condition not [currently]

covered by the plant's operating and emergency procedures."

31. Figure 1, the time-to-boil is shown for initial fuel pool temperatures of 100 F and 110 F, yet the initial design temperature for the fuel pool is 125 F which would represent shorter times-to-boil.

32.. Figure 2, same as Comment 27 above.

33. Figure 9, for the LOOP or LOCA/LOOP case, what prevents condensate from collecting in the ductwork at the low point and causing collapse or blockage?

0'

34. The report does not address the factis that for isolation of Zone III for which great credit taken, the airborne concentration in Zone I (II) will be significantly SGTS higher due to less dilution. Therefore, for the current flowrate the offsite dose will be significantly higher than shown by the current analyses. The airborne exposure to the operator in Zone I (II) will also be significantly higher. These are also currently unanalyzed conditions.

35. Attachment 2, Airborne Dose, Section I, 14 cladding failure is a non-reviewed safety question with respect to the assumptions required by NUREG-0737.

36. Attachment 2,Section II, Contained Sources, this section does not calculate the doses for the NUREG-0737 required case release from the core of 100> noble gases, 504 halogen inventory, and 14 other constituents.

37 Attachment 2,Section III, Offsite Consequences of

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Reactor Building Purge, this section states that the doses shown in the tables must be added to the FSAR, Chapter 15 doses. For the design basis case in Table 1, even without adding these doses to the Chapter 15 doses they exceed the 10CFR100 and 10CFR50, Appendix A, Criterion 19 allowables for offsite and control room doses respectively. Therefore, they don't support the position taken by the report that Zone allowable.

III venting is

38. Attachment 3, Zone III Venting, the actions required here include blanking off supply plenums since just closing the dampers will not be adequate. These actions would require substantial manpower and time under adverse pre-conditions even if all the materials and tools were staged. This entire analysis is based on design features and procedures that are not in place and are unanalyzed.

39. Attachment 3, Instrument Air Supply, Seismic Event, this section only considers the effects of the seismic event on the system tubing. The rest of the system would also be susceptible. In addition, failures from other causes since it it is susceptible to is non-safety-related.

40. Attachment 3, SGTS Operability, this section is based on conjecture and unanalyzed scenarios, and contradicts itself in two places; first, it it even says the condensation will be drawn into the filter train (this would incapacitate the filters) and second, it ductwork would likely fail structurally, both of which says the would constitute failure of SGTS.

41. Attachment 3, Assessment of Emergency Ventilation options

for Zone III Venting during Fuel Pool Boil Scenarios, this entire section is based on designs and procedures that are currently unanalyzed and don't exist, e.g., none of the fan loads, if they are to be powered from the emergency diesel generators, have been analyzed with respect to the diesel generator available capacity.

42. Attachment 3, EQ Equipment Located in Zone III, this section does not show that all devices required to respect to humidity.

qualified are indeed qualified withradiation Additionally, the temperature and effects are not addressed at all.

43. Attachment I, Fuel Performance During a LOCA, this analysis is an important consideration for a JIO, but of it is not in accordance with the "minimum" requirements NUREG-0737.

44. Attachment 7, Use of UHS with RHR in Fuel Pool Coolinq Assist Mode, this attachment concludes that (1)work this zs an unanalyzed condition, and (2) it will only if for single failure is not considered, both unacceptable a design basis.

45. Attachment 8, Drainage of Condensation from the During Fuel Pool Boiling, while this analysis818'levation appears to be very good with respect to conditions on the refueling floor, regarding what it does not address the questions happens to the water that goes down the drains, e.g., does safety-related it equipment create a flooding hazard for in the building.

46. Attachment 9, SGTS Fire Dampers, per this section, the existing fire damper fusible links are unacceptable alone would and design modifications are required. This constitute a reportable item.

Attachxent 33 PP&L Memo from David G. Kostelnik and Mark R. Mjaatvedt to George T. Jones, "Comments on PLI-72783 Regarding EOR G20020", November 11, 1992 (PLI-72857)

November 11, 1992 George T. Jones A6-2 SUSQUEHANNA STEAM ELECTRIC STATION COMMENTS ON PLI-72783 REGARDING EDR G20020 CTN 740281/003 PILE s A45-iA PLI- 72857 This letter provides you with our assessment of the comments contained in the subject letter regarding report NE-092-002, Rev. 0, "Loss of Fuel Pool Cooling Event Evaluation for EDR G20020". In this letter, Messrs. Lochbaum and Prevatte identify forty-six (46) comments related to specific sections of the report. They also provide general comments in the cover letter regarding the intent of the report and its conclusions.

It is not our intent to provide a resolution to each comment listed in the subject letter. We believe that this would be counter productive and simply result in additional letters from Messrs. Lochbaum and Prevatte restating their position.

We do, however, feel compelled to provide some comment on their letter in order to assure that the report is reviewed by others in an objective fashion.

Upon reviewing the statements made in the cover letter,.

apparent to us that the thrust of this letter is to promote it is the views of Messrs. Lochbaum and Prevatte by casting doubt as to the soundness of the evaluations presented in NE-092-002. This is evidenced by the statements which indicate that.

the realistic assessments made in the report cannot be applied to desiqn basis, and that the conclusions drawn in the report are incorrect because of the reliance on operator actions and non-safety-related equipment.

The report does not. state that realistic assumptions will be applied to design basis, and further indicates that, if design basis assumptions are made, then plant operability may be questionable. The intent of the realistic evaluation of each event was to demonstrate that, under the type of realistic assumptions permitted by the NRC for operability evaluations, the plant is operable for a loss of Fuel Pool Cooling event. The report recognizes the need to evaluate the issue further so that appropriate changes can be made to the plant design and design basis.

After reviewing the comments attached to the cover letter we believe that they can be grouped into the following categories:

1. Twenty-six (26) comments make statements to the effect that assumptions it is improper to use in place of design basis realistic assumptions/requirements. These are predominantly with regard to the realistic discussions of the made report. We

,assumptions.

believe it is appropriate to make such Examples are comments number 10, 11, and 19.

2. Three (3) comments represent statements taken out of context. A prime example of this is comment number

" 26, which reads:

"Event 44, page 21, this description dismisses the LOCA/LOOP for the outage sxtuation as "sufficiently small so as not to be considered." This is a probability argument, appropriate for an IPE or JIO, but not appropriate for the design basis except with the approval of the NRC."

The actual sentence, which is being made from a "JIO" viewpoint, reads as follows:

"Furthermore, the probability of a LOCA/LOOP, concurrent with both loops of RHR out of service in an outage, is sufficiently small so as not to be considered."

3. Eleven (11) comments represent statements made by Messrs. Lochbaum and Prevatte to promote their views on a specific area of the report. Examples are comments 3 and 4.

4. Five (5) comments represent areas where Messrs.

Lochbaum and Prevatte apparently misunderstood the report. Examples are comments 15 and 21.

5. There is one (1) comment which is made that This is identifies a true inaccuracy in the report.

comment number 1 which appropriately points out that the abstract does not fully I.dentify all of the concerns'raised by the EDR. This we'll be corrected if and when a revzsion is made to NE-092-002.

As previously stated, we are not planning to respond to PLI-72783 on a comment by comment baszs. We do not believe that this will be productive given the nature of the comments. We simply wish to express our opinion regarding these comments of for your consideration when evaluating the report in light

the comments made by Messrs. Lochbaum and Prevatte. If you wish to discuss this letter or the report, please feel free to contact Dave Kostelnik at x7788 or Mark Mjaatvedt at x7795.

David G. Kostelnik Ma k . 'aatvedt

//-

cc: G. J. Kuczynski SSES C. A. Myers A2-4 M. W. Simpson Al-2 H. G. Staniey SSES J. S. Stefanko A9-3 J. R. Miltenberger A6-1 J. E. Agnew A6-3 G. D. Mailer A6-3 M. R. Mjatvedt A6-3 D. G. Kostelnik A6-3 M. H. Crowthers A6-3 D. F. Roth A6-3 J. M. Kenny A2-4 R. R. Sgarro A6-3 F. G. Butler A6-3 D. C. Prevatte A6-3 D. A. Lochbaum Enercon Nuclear Records A6-2

Attachment 34 PP8L Letter from H. G. Stanley to the U.S. Nuclear Regulatory Commission, "Licensee Event Report 92-016-00", November 17, 1992 (PLAS-546)

Note: PP8L's formal report to the NRC on the concerns expressed in EOR 620020.