ML20063D055

Motion for Leave to File Contentions 21-26 Re Turbine Missiles,New Mark III Containment Concerns,Seismic Evaluation of BWR Core Thermal Hydraulics & in-core Thermocouples.W/Certificate of Svc

ML20063D055
Person / Time
Site:	Perry
Issue date:	08/18/1982
From:	Hiatt S OHIO CITIZENS FOR RESPONSIBLE ENERGY
To:	Atomic Safety and Licensing Board Panel
References
NUDOCS 8208270463
Download: ML20063D055 (21)

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August 18, 1982 / i UNITED STATES OF AMERICA OfC"77D '

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NUCLEAR REGULATORY COMMISSION ,

Before the Atomic Safety and Licensine Board I) fig , ,

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In the Matter of ) c;c,,,

CLEVELAND ELECTRIC ILLUMINATING ) Docket Nos. 50- 40' .

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50-441 ' 'CC COMPANY, Et A1. ) .

) (Operating Lice'n'se)

(Perry Nuclear Power. Plant, )

Units 1 and 2) )

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OHIO CITIZENS FOR RESPONSIBLE ENERGY MOTION FOR LEAVE TO FILE ITS CONTENTIONS 21 THROUGH 26 Ohio Citizens for Responsible Energy ("0CRE")-hereby moves the Licensing Board to grant OCRE leave to supplement further its Petition to Intervene by filing its Contentions 21 through 26 in the above-captioned proceeding. OCRE will first provide general explanations of each contention and then will address the filing requirements of 10 CFR 2.714._

Contention 21 Turbine Missiles OCRE contends tnat the placement and orientation of the PNPP turbine-generators are unacceptable because low trajectory turbine missiles could strike safety-related targets, thereby endangering the safe operation of the facility. This concern was identified as an open item in Section 3.5.1.3 of the Perry SER, NUREG-0887. The ACRS has also expressed dissatisfaction with the progress being made on the resolution of this issue (ACRS Report on the Perry Nuclear Power Plant, Unit 1, dated July 13, 1982).

The Applicants' F5AR, in its treatment of this issue, refers to a report prepared by their A/E, Gilbert Associates 8208270463 820818 - -

PDR ADOCK 05000440 G PDR 35C3

V' Inc., entitled "An Analysis of Low Trajectory Turbine Missile 2," GAI Report Hazards, Perry Nuclear Power Plant, Units 1 and No. 1848, October 1976. This report indeed indicates that. the following structures are within the. low trajectory missile strike z'o n e : control room; cable spreading.. room; HVAC equipment room; intermediate building; auxiliary building; electrical penetration area; reactor buildings of both Units 1 and 2. The estimated damage to these structures resulting from turbine missile impact includes rendering the control room inoperable, the collapse of buildings on safety-related electrical cables mad equipment, and penetration of the containment. .

Obviously these consequences are unacceptable. This sit-uation must be corrected before Perry can be allowed to operate.

Contention 22 New Mark III Containment Concerns Recently . J.M. Humphrey, a. former employee of General _

Electric, identified a number of concerns pertaining to the Mr. Humphrey, Mark III containment, such as is employed at PNPP.

who was for 3 years GE's Lead Systems Engineer for Containment, and'who was involved in the STRIDE program, approached Mississippi Power & Light, applicant for Grand Gulf, with a list of 22 major issues,~ some of which have been further divided into sub-issues (66 total). Although some of these issues did not apply to Grand Gulf, most are still unresolved, and remain so for Perry as well (see July 14, 1982 letter to D. Davidson, CEI from A. Schwencer, NRC requesting additional infonmation on these concerns). 'OCRE therefore adopts as sub-parts to this contention the 66 concerns identified by Mr. Humphrey (and listed in

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Attachment 1).

Contention 23 Seismic Evaluation of BWR Core Thermal-Hydraulics OCRE contends that the Applicants' seismic analysis (and the NRC Staff's review of same in the SER) is deficient because this analysis totally neglects the response of the core thermal-hydraulic design to a seismic event. Because the'BVH1 uses a' two-phase moderator / coolant, it is inherently sus'ceptible to power excursion transients resulting from events affecting void' distribution.

An earthquake could cause sloshing of the water in the reactor vessel,'thus resulting in Toid collapse and/or redistribution. See_ Dr. Richard E. Webb, The Accident Hazards of Nuclear Power Plants (University of Mass., 1976) at 28.

The seismic analy'ses performed for PNPP deal only with the response of components and structures and ignore the core thermal-hydraulic response. This analysis must Ee7erformed before Perry is allowed to operate. In light of the recommendation

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of the ACRS that studies be conducted to evaluate margins avail-able following an earthquake of greater severity than the safe shutdown earthquake (SSE), OCRE suggests that this analysis be based on an earthquake of greater severity than the SSE.

Contention 24 In-Core Thermocouples .

Applicants should conform to the requirements of Regulatory Guide 1.97, Revision 2, and TMI Action " Plan item II.F.2 by In-core thermocouples installing in-core thermocouples at Perry.

provide an indication of inadequate core cooling (ICC) and are a redundant and diverse means by which to detect reactor coolant

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level. As discussed in Section 4.4.7 of the SER, the Staff, which previously required in-core thermocouples in BWRs, has now agreed with General Electric and the BWR Owners Group that the issue should be broadened from the specific requirement' for in-core thermocouples to that of monitoring inadequate' core ,

cooling. OCRE contends that in-core thermocouples should be

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used at Perry.

GE and the BWR Owners Group argued against thle use of in-core thermocouples, citing excessive cos'ts and claiming that the thermocouples offer no advantage in monitoring ICC or reactor water le'el.v The la'tter claim is based upon heat transfer calculations which indicate an excessive time constant in thermocouple response; the Owners Group believes that this J

could provide ambiguous information to plant operators (see

~" Thermal Analysis of In-Core Thermocouples in BWRs," prepared However, an analysis performed by S. Levy, Inc., November 1981).

by Battelle Laboratories' indicates that the time lag might only be 1-1} minutes (letter from C.L. Wheeler, Battelle, to W.V.

Johnston, NRC, dated April 6,1981). GE does admit that thermo-couples could be useful in one situation: loss of coolant in-ventory with no water makeup systems a'vailable (General Electric Evaluation of the Need for BWR Core Thermocouples, November 1981).

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OCRE contends that,these analyses ignore another condition in which in-core thermocouples can provide vital information:

a fuel bundle blockage accident. GE, in its evaluation of this accident (Appendix B of the report mentioned above), makes

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several key assumptions as to the course of this accident so as to support its conclusion that thermocouples are of no value.

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I That these assumptions are arbitrary and unproven is discussed in The Accident Hazards of Nuclear Power Plants, by Dr. Richard E. Webb, at 59-61. OCRE suggests that signals from thermo-couples located near the fuel bundle experiencing blockage would hiert operators that this situation was occurring so that they could scram the reactor. This action would limit the severity of the everheating of the fuel bundle and avoid the possibility of propagating core damage. Rolying on other~

measured variables, such as fission product activity and hy-drogen concentration, to indicate this condition, as. CHI suggests, requires that fuel damage must have already occurred before any corrective action can be taken.. .OCRE maintains that this is unacceptable, as this type of accident can lead to a cascading

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core meltdown. _. ,

! Contention 25 Steam Erosion i

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' OCRi contends that Applicants are not prepared to provent, discover, assess, and mitigate the effects of steam erosion on I

components of PNPP which will be subjected to steam flow.

! Steam erosion has been idantified as the cause of recent failures

' of valves and piping (MSIVs and turbine exhaust lines: see NRC Information Notices 82-22 and 82-23). The NRC Staff has identified Applicants' lack of an inservice testing program for pumps and valves and leak testing of valves as an open item in Section 3.9.6 in the SER.

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Contention 26 Control Room Fire Suppression Applicants are proposing a carbon dioxide fire suppression system for use in the control room. The Staff has identified this as an open item in Section 9.5.1.6.2 of.the SER for the following reasons: (1) C0g has not been tested and approved as a suppression agent for use'in the GE Power Generation

' Control Complex (PGCC) design implemented at Perry; (2) 002 may leak from the underfloor into the control room, possibly causing injury to operators or forcing evacuation of the control room. The Staff instead advocates using Halon 1301.

OCRE contends that all advantages and disadvantages of each suppressent should be thoroughly evaluated before choosing

• a particular system, especially in regard to toxicity. For instance, NFPA 12, Article 121 lists reduced visibility and possible oxygen deficiency as hazards resulting from the dis-charge of large amounts of CO 2. However, according to NFPA 12, Articles A-1200 to A-1202, the hazards resulting from the use of Halon 1301 are twofold: those due to the natural agent, bromotrifluoromethane (CBrF 3), and those due to its decomposition products.

The effects of CBrF3 itself include dizziness, impaired coordination, and reduced mental acuity. It is recommended that personnel do not remain in an area where Halon 1301 concentrations exceed 7% and that they remain no more than a few minutes in an area with Halon concentrations less than 7%.

Persons can be quickly incapacitated by higher levels (10-15%).

Halon decomposition products include halogen acids, free i=' -

4 halogens, and carbonyl halides. These substances are both hazardous to personnel and. corrosive to equipment. In addition, free halogens can poison charcoal filters in the control room HVAC system. These decomposition products cannot be avoided in the presence of flame, since the mechanism by.which Halon

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inhibits combustion involves its decomposition.

Furthermore, halons are kn'own to cause degradation of

'the stratospheric ozone layer.

OCRE believes that the Staff has neglected many toxico-logical and environmental ' factors in its evaluation of this issue. .

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Filing Requirements under 10 CFR 2.714 OCRE has met the' requirements for late filing listed in 10 CFR 2.714(a)(1). All of the contentions filed herein are based upon the Perry SER, NUREG-0887. The SER constituted OCRE's first notice of the concerns identified in Contentions 21, 22, and 26. Contention 23 is based on the deficiency of the Staff's analysis in the SER. Similarly, Contention 24 was filed at this time because prior to the issuance of the SER, OCRE assumed that in-core thermocouples would be required at Perry. The Staff required them at Grand Gulf (Grand Gulf l

SER, NUREG-0831 at 22-22). Contention 25, in addition to

! referring to the Staff's finding as stated in the SER, is based i

l upon two recently issued NRC Information Notices. Thus there exists good cause for this late filing.

OCRE has only this' forum in which to protect its interests; in addition, no other parties to this proceeding have raised l

these issues. That OCRE's participation will aid in the development of a sound record has been affirmed by the Licensing Board (Memorandum and Order of July 12, 1982, LBP-82-53, at 5).

Wh'#, the admission of these contentions might cause delay, this should not be of concern to any party, since Applicants

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recently requested that 'the completion dates for Units 1 and 2 of Perry be extended to 1985 and 1991, respectively (see Attachment 2). These factors thus favor the! admission of these contentions into this proceeding.

Respectfully submitted, Susan L. Hiatt OCRE Representative

1. 8275 Munson Rd.

Mentor,-OH -44060 (216) .55-3158 2

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, f ATTACEHENT 1.

, HUMPHREY CONTAINMENT CONCERNS

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1. Effects of' Local Encroachments on Pool Swell Loads 1.1 Presence of local encroachments such as the TIP platform, the drywell personnel airlock and the, equipment and floo'r drain sumps may increase j the pool swell velocity by as much as 20 per cent. J 1.2 Local en'croach=ent.s in the pool may cause the bubble breakthrough height

. to be higher than expected. -

1.3 Additional submerged structure loads may be applied to submerged structures near local' encroachments.

1.4 Piping impact loads may be' revised as,"a result of the higher pool swell j

velocity.

1.5 Impact loads on the HCU floor may be imparted and the HCU modules may f ail which could prevent successful scram if the bubble breakthrough height is raised appreciably by local encroachmento. ,-

1.6 Local encroachments or the s. team tunnel may cause the pocl swell and

, froth to move horizontally,and apply lateral loads to the gratings around the HCU floor. \

. 1.7 GE suggests that at least 1500 square feet of open area should be maintained in the ECU floor. In order to avoid excessive pressure

, differentials, at least 1500 ft.2 of opening should be maintained at each containment elevation. *

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2. Safety Relief Valve Discharge Line Sleeves 2.1 T'ne knnular regions between the safety relief valve lines and the drywell wall penetration sleeves may produce condensation oscillation (c.o.)

• - frequencies near the drywell and containment wall structural resonnnee frequencies.

'. 2 .'2 The'-potential condensation oscillation and chugging loads produced through the annular area between the SRVDL and sleeve may apply unaccounted for loads to the SRVDL. Since the SRVDL is unsupport'ed from the quencher to the inside of the drywell vall, this may result"in failure of'the line.

2.3 The potential condensation oscillation and chugging loads produced through the annular area between the SRVDL and sleeve may apply unaccounted for loads to the penetration sleeve.

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The loads may also be at or near the natural frequency of the sleeve.

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3. ECCS Relief Valve Discharge Lines Bel'ou the Suppression Pool Level 3.1 The design of the STRIDE plant did not consider vent clearihg, J condensation oscillation and chugging loads which might be' produced by the actuation of these relief valves. -

3.2 The STRlDE design provided only nine inches of submergence above the RER relief valve discharge lines at low suppression pool levels.

3.3. Discharge fro = the RHR relief valves may produce bubble . discharge or

.other submerged structure loads on equipment in the suppression pool.

3.4 The RHR heat exchanger relief valve discharge lines are provided with ~

vacuum breakers to prevent negative pressure in the' lines when disch'arging steam is condensed in the pool. If the valves experience repeated actuation, the vacuum breaker sizing may not be adequate to prevent drawing slugs of water back through the discharge piping. These slugsjof water may apply impact loads *ro the relief valve or ,be discharged back into the pool at thelnext relief valve actuation and

. apply impact loads to submerged structuyes.

3.5 The RHR relief valves'must be capable of correctly functioning fo}]owing ,

an upper pool dump which may increane the suppression poni level as much as five fact creating higher back pressurou on the relief vn3ves.

3.6 If the RHR heat exchanger relief valves discharge steam to the upper

.s levels of the suppression pool following a design basis accident, they will significantly aggravate suppression jool temperature stratification.

'3.7 The concerns related to the RER heat exchanger relief valve, discharge ~

lines should also be addressed for all other relief lines that exhaust into pool. (p. 132 of 5/27 /82. transcript)

4. Suppression Poo'1 Temperature Stratification 4.l' The present containment response analyses for drywell break accidents

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assume that the ECCS systems transfer a significant quantity of water f rom the suppression pwl to the lower regions of the drywell through the break. This results in n pool in the dryvcil which is essential]y tuolated from the supprer:nf on pon) nt a t,cmpernture of npproximatcly 135'F. ,The containment response analysis asuumen that the drywel) pool is thoroughly mixed with the suppression pool. If the inventory in the dryuell. is assnmed to be isolated and the remainder of the heat is discharged to the suppression' pool, an increase in bulk pool temperature of 10*F may occur.

4.2 The existence of the drywell pool is predicated upon continuous operation of the ECCS. The current ecergency procedure guidelines require the operators to throttle ECCS operation to maintain vessel level below level .

8. Consequently, the dryvell pool cay never be formed.

4.3 All Mark III analyses presently assume a perfectly mixed uniform suppression pool. These analyses assume that the te=perature of the suction to the RER heat exchangers is the same as the bulk pool temperature. In actuality, the temperature in the lower part of the pool' l

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where the suction is located will be as much as 7b*F cooler than the bulk pool' temperature. Thus, the heat transf er through the RRR heat exchanger

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will be less than expected.

4.4 The long term analysis of containment pressure / temperature response

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assumes that the wetvell airspace is in thermal equilibritm_ with the

' suppression pool water at all times. The calculared bulk pool

- temperature is used to determine the airspace temperature. If pool thermal stratification were considered, the surface temperature, which is in direct contact-with the airspace, would be higher. Therefore the

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nirspace temperature (and pressure) would he higher.

4.5 A number of factors may aggravate cuppress.lon pool thermal stratification. The. chugging produced through the first row of horizontal vent s will not produce any mixing from the nuppicnnfon pon1 layerc below' the vent rou. An upper pon) dtuup inny cont i Sul c. t o The large volume additional suppression pool. temperature stratification.

of water from the upper pool further 9tboerges RRR heat exchanger effluent discharFe which will decreas,e qxing of the hotter, upper

regions of the pool. Finally, operation of the containment spray

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. eliminates the heat exchanger ef fluent discharge jet which contributes to mixing.

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- 4.6 The initial suppression. pool temperature is assu=ed.to be 95*F while the

' naxiun.t= expected service water temperature is 90*F for all GGNS. accident analyses as noted in FSAR table 6.2-50. If the service water te=perature a.* .is consistently higher than expected, as occurred at Kuosheng, the RER system may be' required to operate ne'arly continuously in order to maintain suppression pool temperature at or below the maximum permissible

.l- value. .

4 I 4.7 All analyses completed for fthe Mark 111 are generic in nature and do not I consider. plant specificinteradtions' oft}neRRRsuppressionpoolsuction and discharge.

spray mode will decrease ,

4.8 Opera' tion of the RHR system in the containment the heat' transfer coefficient.through the RER heat exchangers due ,to l

decreased system flow. The ESAR analysis assumes a constant heat

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- transfer rate from the suptression i pool even with operation of the

'* containment spray.

4.9 The.effect on the long term containment response and the operability of the spray system due to cycling the containment sprays,on and off to maximize pool cooling needs to be addressed. Also provide and justify the criteria used by the operator for switching from the. containment i-spray. mode 'to pool cooling mode, and back again. (pp. 147-148 of 5/27/S2 transcript) 4.10 Justify that the current arrangement of the discharge(pp. and. suction points 150-155 of l

of the pool cooling system maximizes pool mixing.

' 5/27/82 transcript) ,

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5. Dryvell to Containment Bypass Leakage

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5.1 The v'o rst case of dryuell to cont:sinment bypass -leakage has*be'en established as a small break accident. An intermediate break accident will actually produce the most significant dryuell to containment leakage prior to initiation of containment sprays. _

5.2 Under Technical Specification limits, bypass leakage' corresponding to A/ 6 = 0.1 ft.2 constitute acceptable operating conditions.

S= aller-than-1BA-sized breaks.can maintain break flow into the dryuell

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for long time periods', however, because the RPV uould be deoressurized over a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period. Given, f or exa=ple , an SBA with 'A/8 = 0.1, projected time period for containeent~ pressure'to reach 15 psig is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. In the latter 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of'the depressurization the containment would presumably experience ever-increasing overpressurization.

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- 5.3 . Leakage from the dryuell to containment vill inerense the temperature and e

pressure in the containment. The operators will have to uce the

• ' containments spray in order to ma'intain containment temperature and.

pressure control.. Given the decreased effectiveneno of the RllR cystem in

- accomplishing this objective..in 'the containment opray mode, the bypar.s leakage may increase the. cyclical duty of the containment sprays.

o j 5.4 Dire.ct leakage from the dryuell to the containment may dissipate hydrogen

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outside the region where the hydrogen recombiners take suction. The anticipated leakage exceeds the capacity of the dryuell purge

_. a compressors. This could lead to pocketing '

of hydrogen which exceeds the

- concentration limit of 4% by, volume.

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5.5 Equipment may be exposed to local conditions which exceed the f.4 environmental qualification envelope as a result of dir'ect dryuell to containment bypass leakage.

t Y 5.6 The test pressure of 3 psfg specified for the perf. odic operational

. .dryuell leakage rate tests does not reflect additionalThis presourization pressure alsoin the dryuell which will' result from upper pool dump.

does not reflect additional dryuell pressurization resulting from I~

throttling of the ECCS to' maintain vessel. level which is required by the

. curreht EPGs.

, 5.7 Af ter upper pool du=p, the level of the pool will be 6 feet higher, and dryuell-to-containment dif ferential pressure will be great.er than 3 psi.

i The dryvell H 2purge mpress r head is nominally 6 psid. The concern is that af t er an upper pool dump, the purge compressor head may not be j

sufficient to depress the weir annulus enough to clear the upper vents.

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In such a case, H mixing u uld not be achieved.

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reaching the 5.8 The possibility of high temperatures in the drywell without 2 psig high pressure scram level because of bypass. leakage through the dryuell vall should be addressed. (pp. 168-174 of 5/27/82 transcript)

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6. RHR Perminsive on Containment Spray 6.1 General Electric had recommended that the dryucll purge compressors'and the hydrogen recombiners be act'ivated if the reactor vessel water ' level drops to within one foot of the top of active fuel. This requirdment was not incorporated in the emergency procedure guidelinss. --

. 6.2 'Ger.eral Electric has recom= ended that an interlock be provided to require

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containment spray prior to starting the. recombiners because of the large

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quantities of heat input to the containment. Incorrect implementation of this. interlock could result in inability to operate the recombiners

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vithout containment spray.

6.3 The recombiners may produce " hot spots" near the reco=biner exhausts

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which might exceed the environmental qualification envelope or the containment design temperature.

6.4 For th'e containment air monitoring sys7em furnished b General Electric, the analyzers are not capable of measurigg hydrogen concentration at volumetric steam concentrations above 60 . Effective measurement is

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precluded by condensation of steam in the equipment.

6.5 Discuss the possibility.of local temperatures due to recombiner operation being hip,her than the temperature qualification profiles for equipment in 3'

the region around.and above the recombiners.. State'what instruction ~s', if any, are available to the operator to actuate contain=ent sprays to keep

,this temperature below Aesign values. (pp.,183-185 of 5/27 /82.

transcript) s

7. Containment Pressure Response - .

7.1 The containment is assumed to be in thermal equilibrium with a perfectly mixed, uniforn temperature suppression po~ol. As,noted under topic.4, the surface temperature of the pool vill be higher than the bulk pool temperature. This may produce higher than expected containment temperatures and -r pressures. ,, .

l 7.2 The computer code used by General Electric to calculate environmental

. qualification parameters considers heat transfer from the suppression pool surface to the containment atmosphere. This is not in accordance

. , with the existing licensing basis for Mark III environ = ental j

qualification. Additionally, the bulk suppression pool temperature was used in the analysis instead of the suppression pool surface temperature.

7.3 The analysis assumes that the containment airspace is in thermal equilibrium with the suppression pool. In the chort term thin is non-conservative for Mark 111 du'e to adiabatic compression ef fects and finite time required for heat and mass to be transf erred between the pool and containment volumes.

8-. Containment Air Mass Effects ,

l 8.1 This issue is based on consideration that some Tech Specs allow operation at parameter-values that differ from the values used in assumptions for FSAR tran'sient analyses. Normally analyses'are done assuming a nominal I . .

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i ennt mininent prompuro equal to ambient (0 pcIE) a tempenafuta near maxJmum upo n a t t ny, ( 9 0 " 11 ) anil elo not 1imit the ils ywe i I piannuia equet io ihe containment pressure. The Tech Specs operation under conditions. cuch an a positive containment pressure (1.5 psig) . temperatures less than to maxicum (60 or 70*F) and drywell pressure can be negative with respect the containment (-0.5 psid). All of these differences would result in

. transient response,different than the FSAR descriptions.*'

8.2 The draf t GGNS technical specifications permit operation.of the plant with containment pressure ranging between 0 and_-2 p'sig. Initiation of -

, containment spray at a pressure of -2 psig may reduce the containment pressure by an additional 2 psig which could lead to buckling _and

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failures in the containment liner plate.

.8 . 3 If the containment is maintained at -2 psig, the top row of vents could admit blowdown to the suppression pool during an SBA without a LOCA signal being developed.

B.4 Describe all'of the possible methods both before and after an accident of eresting a c,ondition of low air mass i'negde the containment. Discuss the

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_~ effects on'the containment design external pressure of actuating the containment sprays. (pp. 190-195 of 5/27 /82 transcript)

9. Final Dryvell Air Mass 9.1 11e current FSAR analysis is based upon continuous injection of

~,_ f relatively cool ECCS vater into the drywell through a broken ~ pipe following a design basis accident. The EPG's direct the operator to throttle ECCS operation to maintain reactor vessel level at about

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level 8. Thus, instead of releasing relatively cool'ECCS vater, the break will be releasing saturated. stea=.which cight produce' higher containment pressuritations than currently anticipated. Therefore, the drywell air which would have been drawn back into the drywell will remain in the containment and higher pressures will result in both the l

containment and the drywell.

9.2 The continuous steaming produced by throttling the ECCS flow will cause increased direct lenknge from the drywe~11 to the containment. This could

.rcsult in 1ncreased containment pressuren.

1 9.3 it appears that some confucion exists as to whether SBA's and stuck open SRV accidents are treated as transients or design basis accidents.

Clarify how they are treated and indicate whether the initial conditions were set at nominal or licensing values. (pp. 202-205 of 5/27/82 transcript).

10. Drywell Flooding Caused by Upper pool Dump _

10.1 The suppression pool may overflow fro = the weir wall when the upper pool is du= ped into the suppression pool. Alternately, negative pressure between the drywell and the containment which occurs as a result of nor=al operation or. sudden containcent pressuritation could produce l

similar overflow. Any cold wate'r spilling into the dryvell and striking

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hot equipment may produce thermal f ailures.

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10:2 Describe the interface requirement (A-42) that, specifies that no flooding of the drywell shall occur. Describe your intended methods to follow this interface or justify ignoring this requirement. (pp. 209-226' of 5/27/82 transcript)

11. Operational Control of Drywell to Containment Differentiai-Pressures

- Mark III load definitions'are based upon the levels in the suppression pool' and the' drywell weir annulus being. the same. The CGNS technical

• npecifi cations permit elevation differences between these pools. This

,' may effect load definition for vent ci ca ri ng..

12. _

Suppression Pool Makeup LOCA Seal in .

The upper pool dumps into the supprescion pool nutomatJen3 3y fullowing a LOCA signal with a thirty minute delay timer. If the signal which starts the timer disappears on the solid state logic plants, the timer resets to zero p'reventing upper pool-dump. ,

13. Ninety Second Spray Delay

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The '.'B" loop of the containment sprays includes a 90 second timer to prevent simultaneous initiation of the redundant containment sprays.

Because 6f instrument drift in the sensing instrumentation and the timers, GE estimates that there is a 1 in 8 chance that the sprays will actuate simultaneously. Simultaneous actuation could produce negative *

. pressure transients in the containment and aggravate

• temperature stratification in the suppression pool. x

14. RHR Backflow Throukh Containment Spray _

A f ailure in the check valve in. the LPC1 line to the reactor vessel could result in, direct leakage from the pressur'e~ vessel to the containment

- atm'osphere. This leakage might occur as the LPCI motor operated isolation valve is closing and the motor operated isolation valve in the This could produce unanticipate,d conta'inment spray.line is opening. , , ,

in. creases in the' containment sp, ray.

15. Secondary Containment Vacuum Breaker Plenum Resnonse The . STRIDE plants had vacuum breakers between the containment and the secondary containment. With suf ficiently high flows through the vacuum breakers to containment, vacuum could be created in the secondary

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16. Effect of Suppression Pool Level on Temperature Mensurement

.Some of the suppression pool temperaturc sensorn nre located (by. GE recommendation) 3" to 12" below the pool surface to pr' ovide early warning of.high pool temperature. Bovever, i,f the suppression pool is drawn down ~

below the level of the temperature s'ensors, the operator could be misled by erroneous readings and required safety action could be delayed.

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17. Emergency Procedure Guidelines The EPGs contain a curve which specifies limitations on suppressidn' pool level and reactor pressure vessel pressure. The curve presently do'es not adequately account for upper pool dump. At present, the operator would be required to initiate automatic depressurization when the only' action required is the opening of one additional SRV.

Ef f ects of Insulation Debris 18.

~ 18.1 Tailures of reflective insulation in the drywell'may lead to blockage of the gratings above the weir annulus. This may increase the pressure required in the drywell to clear the first row of drywell vents and.

perturb the existing load' definitions.

18.2 Insulation debris may be transported through the vents 'in the drywell wall into the suppression pool. This debris could then cause blockage of the suction strainers. - --

19. Submergence' Effects on Chugging Loads.' l 19.1 The chugging loads were originally defined on the basis of 7'.5 feet of submergence over the drywell to suppression pool vents. Following an upper pool dump, the cubmergence will actually be 12 feet which may effect chugging loads.

19.2 The effect of local encroachments on chugging loads needs to be addressed. (pp. 251-252 of 5/27/82 transcript)

20. Loads on Structures Piping and Equipment in the Drywell During Reflood

• During the latter stages of a LOCA, ECCS overflow from the primary system,,can cause drywell depressurization and vent backflow. The GESSAR defines' vent backflow vertical impingement and drag loads, to be applied to drywell: structures, piping, and equipment, but no horitontal loading is specified. ,

21. Containeent Makeup Air For Eackup Purge This Regulatory Guide 1.7 requires a backup purge H2 re val capabil,ity.

backup purge for Mark 111 is via the drywell purge line which discharges to the shield annulus which in turn is exhausted through the standby gas J

l treatment system (SGTS). The containment air is blevn into the drywell via the'sdrywell purge compressor to provide a positive purge. The compressors draw from the containment, however, without hydrogen lean air makeup to the containment, no reduction in containment hydrogen concentration occurs. It is necessary to assure that the shield annulus volume contains a hydrogen lean mixture of air to be admitted to the containment via contain=ent vacuum breakers.

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22. Miscellaneous Emergency Procedure Guideline Concerns The EPGs currently in ' existence' have been prepared with the' intent 6f coping with degraded core accid'ents. They cay contain requirements conflicting with design basis accident conditions. Someone needs to carefully review the EPG's to assure that they do not conflict with the expected course of the design basis accident.

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THE CLEVELAND ELECTRIC ILLUMIN ATING COMPANY P.o. Box 5000 a CLEVELAND. oHlo 44101 m TELEPHONE 1216) 622-9800 m ILLUMINATING BLOG. e 55 PUBLIC soVARE Serving The Best Location in the Nation Dalwyn R. Davidson tsCE PRES 0ENT .

SYSTEM ENGINEERING AN0 CONSTRUCTION July 21,1982 l'

Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U. 5. Nuclear Regulatory Commission Washington, D. C. 20555 Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Extension of Construction Permit Numbers CPPR-148 and CPPR-149

Dear Mr. Denton:

Enclosed herewith is an application for amendment of construction permit numbers CPPR-148 and CPPR-149 to extend construction completion dates.

We interpret this to be a Class II amendment per 10 CFR Part 170, and enclose a

, check for $1,200.00.

,, Very truly yours, Daiwy R. Davidson Vice. President l System Engineering and Construction j DRD:mb cc: Jay Silberg, Esq.

John Stefano l Max Gildner a

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• i h00 $

w(ebug.f1,200 ffdd: S l

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2B20722002r920721 PDR ADOCK 05000440 ^

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

)

THE CLEVELAND ELECTRIC. ) Docket Nos. 50-440 ILLUMINATING COMPANY, et al. ) 50-441

)

(Perry Nuclear Power Plant, ) --

Units I and 2) . ) __

APPLICATION FOR AMENDMENT OF CONSTRUCTION PERMITS NOS. CPPR-148 AND CPPR-149 TO EXTEND CONSTRUCTION COMPLETION DATES The Cleveland Electric illuminating Company, Duquesne Light Company,

^

Ohio Edison Company, Pennsylvania Power Company, and The Toledo Edison Company (hereinafter "Permittees") are the co-holders of Construction Permits Nos. CPPR-148 and CPPR-149 authorizing construction of the Perry Nuclear Power Plant, Units 1 and 2. 3 ,,

\

Construction Permit No. CPPR-148 Eurr~ently specifies December 31, 1982 as the latest date. for completion of construction of Unit 1. Construction Permit No. CPPR-149 currently specifies. June 30, 1984 as the latest date for completion of construction of Unit 2. Permittees' currently scheduled dates of commercial operation are May,1984, for Unit I and May,1988, for . Unit 2. In order to provide for further time contingencies as explained below, and pursuant to 10 CFR 550.55(b),

Permittees respectfully request that the Nuclear Regulatory Commission amend Construction Permit No. CPPR-148 to specify November 30, 1985 as the latest date for completion of construction of Unit I and Construction Permit No. CPPR-149 .

to specify November 30,1991 as the latest date for completion of construction of Unit 2.

The extensions of time for the construction completion dates are needed because of the following: ,

1. Since construction of Perry began, projections of the growth rate in the demand for electricity have been significantly reduced t

as a result of the slowdown in industrial growth, increased C207270036 820721 ' -

( ..

PDR ADOCK 05000440

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availability of natural gas, and conservation efforts by customers. i This reduced growth rate has delayed the need for the capacity to be supplied by the Perry units.

2. Numerous changes and additional requirements for plant design and analysis have been incorporated, including those required by the Commission as a result of the Three Mile Island accident and during the course of the NRC's regulatory review.

These substantial changes have dictated successive extensions of our project schedule to reflect the time required for completion of additional procurement and construction activities.

3. Increasing financing-requirements caused by changes. in plant d; sign, increased plant construction costs and the sustained high rates of inflation during the past several years, have increased the difficulties in obtaining capital funds.

• THE.. CLEVELAND ELECTRIC ILLUMINATING COMPANY By Daiwyn If/ Davidson Vice Preident System Engineering and Construction and subscribed before me, ay of w ,1982 O .h / /U ablic ission expires on:

A /?,/9$b AROUNE M. WILDE ry Put,lic. State of Ohio assion Expres April 17,1985 ccrded in 12ke County)

. .j

,. a, 1,If CERTIFICATE OF SERVICE C{E}Ed This is to certify that copies of the foregoing OHIO .g C- I 0 23 f 1 CITIZENS FOR RESPONSIBLE ENERGY MOTION FOR LEAVE TO FILE ITS CONTENTIONS 21 TEROUGH 26 were served by deposit in the.

U.S. Mail, first class, postage prepaid, this 18th day of cr.ett og 3ggpt7y August, 1982 to those on the service list below. CX ECc. ?i..sERvg

. n. n 2

Susan L. Hiatt

' SERVICE LIS'T Peter B. Bloch, Chairman Daniel D.08159.

Wilt Esq.

Atomic Safety and Licensing Board P.O. Box U.S. Nuclear Regulatory Comm'n Cleveland, OH .44108 Washington, D.C. 20555 Dr. Jerry R..Kline Atomic Safety and Licensing Board U.S. Nuclear Regulatory Comm'n Washington, D. C. 20555 Frederick J. Shon Atomic Safety and Licensing Board U.S. Nuclear. Regulatory Comm'n ,

Washington, D.C. 20555 Docketing and S.ervice Section -

Office of the~ Secretary' U.S. Nuclear Regulatory Comm'n Washington, D.C. 20555 l

t Stephen H. Lewis, Esq.

- Office of the Executive l Legal Director U.S. Nuclear Regulatory Comm'n Washington, D.C. 20555 Jay Silberg, Esq.

1800 M Street, N.W.

Washington, D.C. 20036 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear' Regulatory Commission Was hington, D.C. 20555 L