ML20027B570

Forwards Util 820903 Submittal Responding to NRC Request for Info Re Resolution of Certain Unresolved Safety Issues

ML20027B570
Person / Time
Site:	Dresden
Issue date:	09/20/1982
From:	Fitzgibbons R COMMONWEALTH EDISON CO., ISHAM, LINCOLN & BEALE
To:	Atomic Safety and Licensing Board Panel
References
ISSUANCES-SP, NUDOCS 8209210295
Download: ML20027B570 (34)

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ISHAM, LINCOLN & BEALE '00Cpv u

COUNSELORS AT LAW THREE FIRST NATIONAL PLAZA CHICAGO. ILUNOIS 60602 R TUNC 288 1120 C N T1 AE k W.

WASHINGTON. O C 20036 202 831 9730 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of )

COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-237-SP (Dresden Station, Units 2 ) 50-249-SP and 3) ) (Spent Fuel Paol Modification)

Dear Administrative Judges,

Please find enclosed Commonwealth Edison Company's

(" Commonwealth Edison") submittal responding to the NhC Staff's request for information regarding the resolution of certain Unresolved Safety Issues at the Dresden Nuclear Station, Unit 2. Although no exceptions have been filed, it is Commonwealth Edison's understanding that the Appeal Board is exercising its sua sponte authority to review the Licensing Board's final decision in this proceeding. Since the enclosed submittal is arguably relevant to Board Question No. 2, Commonwealth Edison is providing the submittal in accordance with the full disclosure requirements set forth in Duke Power Con.pany (William B. McQuire Nuclear Station, Units 1 and 2) , ALAB-143, 6 AEC 623 (1973). Commonwealth Edison does not believe the enclosed submittal draws into question the Licensing Board's resolution of Board Question No. 2. Nor should this letter indicate Commonwealth Edison's assent to the appropriateness of the Appeal Board's sua sponte review.

Respectfully submitted, Robert G. Fito 1bbons, Jr.

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r r SERVICE-LIST Alan S. Rosenthal, Chairman U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Appeal Board Washington, D.C. 20555 Thomas S. Moore U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Appeal Board Washington, D.C. 20555 Dr. Reginald L. Gotchy U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Appeal Board Washington, D.C. 20555 John F. Wolf, Esq.

3409 Shepherd Street Chevy Chase, Maryland 20015 Dr. Linda W. Little 5000 Hermitage Drive Raleigh, North Carolina 27612 Dr. Forrest J. Remick 305 E. Hamilton Avenue State College, PA 16801 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555' Docketing and Service U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Richard Goddard U.S. Nuclear Regulatory Commission i

Maryland National Bank Building 7735 Old Georgetown Road Bethesda, Maryland 21202 Philip L. Willman Assistant Attorney General Environmental Control Division '

188 West Randolph Street Suite 2315 Chicago, IL 60601 l

[ CommonweaHh Edison

[ C C ) one First National Plaza. Chicago. I!!inois

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'(v) Xddies'slieplyIo' Post' Office Bod 67 Chicago, liiinois 60690 September 3, 1982 Mr. Gu s C. Lainas Assistant Director for Safety Assessment Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Dresden Station Unit 2 Unresolved Sa fety Issue Status NRC Do cket No . 50-237 Re ference (a): Gus C. Lainas letter to L. O.

De1 George dated Jul y 6, 1982.

Dear Mr. Lainas:

In Reference (a), Commonwealth Edison was requested to provide information regarding the resolution status of the following Unresolved Safety Issues at Dresden Unit 2:

(1) Waterhammer - (A-1)

(2) BWP, Mark I Pressure Suppression Containments - (A-6, A-7, and A-39)

(3) ' Anticipated Transients Without Scram ( A-9)

(4) BWR Nozzle Cracking - ( A-10)

(5) Reactor Vessel Materials Toughness ( A-11)

(6) Systems Interaction in Nuclear Power Plants ( A-17)

(7) Environmental Qualification o f Safety Related Electrical Equipment (A-24)

(8) Residual Heat Removal Requirements ( A-31)

(9) Control o f Heavy Loads %Bar Spent Fuel ( A-36)

(10) Seismic Design Criteria (A-40)

(11) Pipe Cracks at Boiling Water Reactors ( A-42)

(12) Containment Emergency Sump Reliability ( A-43)

(13) Station Blackout (A-44) ,

(14) Shutdown Decay Heat Removal Requirements ( A-45)

(15) Seismic Qualifications o f Equipment in Operating Plants (A-46)

(16) Sa fety Implications o f Control Systems ( A-47) i (17) Hydrogen Control Measures and Ef fects o f Hydrogen Burns on Sa fety Equipment ( A-48)

G. C. Lainas September 3, 1982 Attachment 1 to this letter provides the' requested in fo rma tio n . In all cases we believe~ that continued operation is completely the NRC. justified for the same reasons identified generically by Where we have additional justification because of special plant features, procedures, reviews or modifications, we have de-scribed this in the attachment. We understand that this information will be used to support the full-term operating license conversion pending before the NRC.

To the best of my knowledge and belief the statements con-tained herein and in the attachment are true and correct. In some respects these statements are not based on my personal knowledge but upon information furnished by other Commonwealth Edison employees.

Such information has been reviewed in accordance with Company practice and I believe it to be reliable.

matter toPlease address any questions you may have concerning this this office.

One (1) signed original and thirty-nine (39) copies o f this transmittal are provided for your use.

Very truly yours, f l " !-

Thoma s J. Raus ch Nuclear Licensing Administrator 1m Region III Inspector - Dresden SUBSCRIBED and SWORN to be forp me this tECd of h e d ,b day 1982 ,

0 baa.@k Ngtary Public 4957N

Attachment I TASK A-1 Waterhammer

1. Description of Problem Waterhammer events are intense pressure pulses in fluid systems caused by any one of a number of mecnanisms and system conditions ,

sucn as rapid condensation of steam pockets, steam-driven slugs of water, pump startup with partially empty lines, and rapid valve motion. Since 1971 over 200 incidents involving waterhammer in ,

pressurized and boiling water reactors have been reported. The l waternammers (or steam hammers) have involved steam generator feedrings and piping, the residual heat removal systems, emergency core cooling systems, containment spray, service water, feedwater and steam lines.

Most of tne damage reported has been relatively minor, involving pipe nangers and restraints; nowever, several waterhammer incidents nave resulted in piping and valve damage. The most serious waterhammer events have occurred in the steam generator feedrings of pressurized water reactors. In no case has any waterhammer incident resulted in the release of radioactive material.

Under Generic Task A-1, the potential for waterhammer in various systems is being evaluated and appropriate requirements and systematic review procedures are being developed to ensure that waternammer is given appropriate consideration in all areas of licensing review. A tecnnical report, NUREG-0582, "Waternammer in Nuclear Power Plants" (July 1979), providing the results of an NRC staff review of waternammer events in nucicar power plants and stating current staff licensing positions, completes a major subtask of Generic Task A-1.

2. Justification of Continued Operation Although waternammer can occur in any lignt water reactor and as approximately 118 actual and probable events nave been reported in boiling water reactors as of September 1979, none nave caused major pipe failures in boiling water reactors sucn as Dresden Unit 2 and none nave resulted in tne offsite release of radioactivity.

Dresden Unit 2 has installed several systems to preclude waternammer from occurring in emergency core cooling system lines. -

Tne Low Pressure Coolant Injection (LPCI), High Pressure Coolant Injection (HPCI), and Core Spray (CS) systems all nave a jockey pump to pressurize their respective lines and prevent the chance for any voids from forming. Also, a fill line routing water from tne Diesel Generator Cooling Water System to the Containment Cooling Service Water (CCSW) subsystems of LPCI has been installed to prevent voids from forming in tnese lines. In aadition to tnese waternammer precautions, a drain line nas been installed on the lowest elevation of e HPCI turbine inlet main steam line to remove possiole cont o.esate and preclude waterhammer damage f rom occurring.

E' B 2-In the event that Task A-1 identifies potentially.significant waternammer scenarios which nave not explicitly been accounted for in the design and operation of Dresden, Unit 2 corrective measures will oe considered at that time. Tnis task has not i dentified tne need for measures beyond those already implemented.

Basea on tne foregoing, we conclude that Dresden Unit 2 can be i operated prior to ultimate resolution of this generic issue without undue risk to the health and safety of the public.

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~3-TASK A-6 Mark 1 Snort Term Program TASK A-7 Mark I Long Term Program

1. Description of Problem During tne conduct of a large scale testing program for an advanced design pressure-suppression containment system (Mark III) for BWRs, new suppression pool hydrodynamic loads associated with a postulated loss-of-coolant accident (LOCA) were identified wnich nad not been explicitly included in the original design of the Mark I containment systems. These additional loads result from dynamic effects of drywell air and steam being rapidly forced into the suppression pool (torus) during a postulated LOCA event. In addition, recent experience at operating plants has indicated that tne dynamic effects of safety-relief valve (SRV) discharges to the suppression pool could be substantial and should be reconsidered.

Tne results of tne Mark I containment short-term program (STP) have provided assurance that the Mark I containment system of each operating BWR facility would maintain its integrity and functional capability during a postulated LOCA. However, the STP evaluation was conducted using a "most probable load" approach which was aimed at the identification of load magnitudes and load combinations whicn were most likely to be encountered during the course of a postulated design basis LOCA. In addition, the STP structural acceptance criteria were selected to assure that, for the most probable loads induced by a postulated design basis LOCA, a safety factor to failure of at least two existed for tne weakest structural or mecnanical component in the containment system for eacn operating Mark I BWR facility.

Consequently, since tne design margin of safety f or the containment systems of operating Mark I facilities nas been reduced from the margin believed to be present at tne time these facilities were originally reviewed and licensed, the need existed (1) to establish design basis LOCA loads wnich are appropriate for the life of the facility, and (2) to restore the originally-intended design safety margins for tne containment systems.

2. Justification for Continued Operation The safety issue addressed by this Task Action Plan (TAP) is applicable to boiling water reactor (BWR) facilities with the Mark '

I containment system design. A total of 25 sucn facilities nave been built or are being built in tne United States; of these, 22 are cu.rently licensed for operation.

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For Mark I BWRs currently licensed for operation, the NRC has concluded tnat tnere is reasonable assurance tnat continued operation, pending completion of this task, does not constitute an undue risk to the health and safety of tne public for the following reasons:

As documented in NUREG-0408, " Mark I Containment Short-Term Program Safety Evaluation Report," December 1977, based upon our review of tne generic "Short-Term Program Final Report" and Addenda submitted by the. Mark I Owner's Group and tne plant-unique analysis reports submitted by each licensee of an operating Mark I BWR facility, we nave concluded that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety of the puolic, during an interim period of approximately 2 years, while a metnodical, comprehensive Long-Term Program (LTP) is conducted.

This conclusion has been made based on NRC determinations (1) that the magnitude and character of each of tne hydrodynamic loads resulting from a postulated LOCA have been adequately defined for use in the Snort Term Program (STP) structural assessment of tne Mark I containment system, (2) that, for tne "most probaole" loads induced by a postulated LOCA, a safety factor to failure of at least two exists for the weakest structural or mechanical component in tne containment system for eacn operating Mark I BWR facility, and (3) that, based on (1) and (2), eacn Mark I containment system would maintain its integrity and functional capability in the unlikely event of a design basis LOCA.

The NRC nas reviewed tne Mark I Owner's Program Action Plan for the LTP and nave found.that it is reasonably designed to provide resolution of tne issues raised during our review of tne STP and to satisfy tne LTP objectives. The NRC is continually monitoring tne progress of tne LTP to assure that these requirements are satisfied.

As was the case during the conduct of the STP, if information becomes available during the course of the LTP which indicates that the safety factor to failure of a component of the containment system of a Mark I BWR facility is less than two, immediate corrective action could be required. Such action could take the form of structural modification, installation of load mitigating devices, or otner appropriate measures.

3. Program to Resolve Issue As previously stated tne NRC has. reviewed tne " Mark 1 Owners Group Long-Term Program". Tney issued their assessment of the program as NUREG -0651 " Mark 1 Containment Long-Term Program Safety Evaluation Report, Resolution of Generic Technical Activity A-7", dated July, 1980. Commonwealth Edison Company is addressing the issues as presented in NUREG - 0661 on a schedule commensurate with order dates specified by the Commission in a letter from Dominic Vassalo to Mr. DelGeorge dated January 19, 1982 and as clarified in a letter from T.J. Rauscn to Mr. Denton dated April 6, 1982.

TASK A-39 SRV, Pool Dynamic Loads

1. Description of Problem BWR plants are equipped with relief-valves tnat discharge into the wetwell. Upon relief valve actuation, the initial air column within the SRV discharge line is accelerated by the high pressure steam flow and expands as it is released into the pool as a nign pressure air ouoble. The high rate of air and steam injection flow in tne pool followed by expansion and contraction of tne bubble as it rises to the pool surface produces pressure oscillations on the pool boundary. This effect is referred to as the air-clearing pnenomenon.

Experience at several BWR plants with pressure suppression containments nas shown that damage to certain wetwell internal structures can occur during safety / relief valve (SRV) blowdowns as a result of air-clearing and steam quencning vibration pnenonmena.

In addition to the boundary loads, e.g., containment structures, reactor pedestal, the air injection and subsequent bubble motion produces pressure waves and water movement within the pool that produce drag loads on components in the pool.

Following the air-clearing phase, pure steam is injected into the pool. Condensation oscillations occur during tnis time period.

However, tne amplitudes of these vibrations are relatively small at low pool temperatures. Continued blowdown into the pool will increase the pool temperature until a threshold temperature is reacned. At tnis point, steam condensation becomes unstable.

Vibrations and forces can increase by a factor of 10 or more if the SRV continues to blow down. This effect is referred to as the steam quenching vioration phenomenon. Current practice for the BWR operating plants is to restrict the allowable operating temperature envelope via Tecnnical Specifications sucn that the threshold temperature is not reached.

In response to the concern on relief valve loads, letters were sent in 1975 to all licensees of operating BWR plants requesting tnat I

tney report on tne potential magnitude of relief valve loads, and on the structural capability of the suppression chamber and internal structures to tolerate such loads. As a result of the generic concerns, owners groups were formed by botn Mark I and II -

utilities. Tnrough these groups, integrated generic analytical and experimental programs have been developed to address tne subject of l SRV loads.

Witn respect to Mark III containments, tne staff nas establisned acceptance criteria for quencher loads. These criterion were conservatively establisned based on the data base available to us.

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One of tnese criteria requires the applicants to assume that, for the events involving multiple valve actuations, the bubbles from eacn SRV discharge reacn their peak pressures simultaneously and tnen oscillate in phase. GE, nowever, believes that this assumption is unrealistically conservative and will impose unduly severe loadings on equipment and piping in the plant. In early 1978, GE proposed an equipment reevaluation program, which considers a statistical approacn to determine the effects of bubble phasing considerations. Tnis Containment Loads Report - Mark III Containment," Rev. 2. Since tnis approach is expected to be applied generically to all plants witn Mark III containments, we have included this review item in the task action plan.

Recently, GE issued a Part 21 notification related to consecutive actuation of multiple safety / relief valves and concomitant load increases for BWR Mark III water pressure- uppression containments. Tnis concern resulted from a recent study perf ormed by GE of tne primary system pressure response following an isolation event. Tne results showed tnat more than one safety / relief valve could be actuated consecutively, as a result of a reactor isolation event. Tnis SRV load combination nas not been considered in the design. Discussions with GE have also revealed tnat tnis concern is generic to all BWR containments and, tnerefore, is included in tne task action plan.

Results of tnis task will be incorporated into Tasks 1.c and 2.d for establisning SRV load cases and load combination. A report of our evaluation will also be issued f or this particular concern.

2. Justification for Continued Operation As discussed in Section 1, the safety issue addressed by this task is tne possible damage to wetwell internal structures and the pool boundary tnat could occur due to air-clearing and steam quencning pnenomena resulting from safety relief valve (SRV) discnarge into tne suppression pools of BWR plants. It is of concern to all BWR plants using the Mark I, Mark II, or Mark III pressure suppression type containments.

This task will provide the basis for establishing acceptance criteria for safety relief valve loads and for suppression pool temperature limits. In conjunction witn Task A-7 (Mark I Long-Tern ,

Program) and Task A-8 (Mark II Containment Pool Dynamic Loads), a complete evaluation will be provided of suppression pool dynamic loads for BWR containments.

For Dresden 2 the justification for continued operation and licensing is based on an evaluation of operating experiences and the plant capability to tolerate SRV loads in the snort term. SRV operating experience nas shown tnat in all but a few instances, SRV discnarges have performed satisfactorily witnout any evidence of

damage either due to the hydrodynamic loads or pool temperature effect. In tnese cases wnere localized damage occurred at other result in a loss of the containment plants, the damage did notor release of radioactivity, or undue. risk to the health function, and safety of the public. In those cases, repairs were made and With respect to aoditional margin was included in the structures.

tne plant capaoility, the NRC has concluded that the plants have tne capability to tolerate SRV loads becauseallthe However, loadswill plants areberelated to the structural fatigue life.

required to demonstrate tne capability to meet the SRV loads criteria and pool temperature limit which will be established by tnis task.

In summary, it nas been concluded that the Tnerefore, SRV feels CECO loadsthat are the related to plants the structural fatigue life.

witn Mark I containment can be allowed to continue operation until completion of the Mark I Long-Term Program.

3. Program to Resolve Issue .

I Long-Term Program (NUREG-0661)

The NRC assessment of the Mark addresses the SRV discharge loads and specifies Tnerefore, these requirements issues will befor torus temperature monitoring.

resolved on a schedule as reported in Section 3 of Tasks A-6 and A-7.

TASK A-9 Anticipated Transients Without Scram

1. Description of Problem Nuclear plants have safety and control systems to ?imit the consequences of temporary abnormal operating conditions or

" anticipated transients." Some deviations from normal operating conditions may be minor; otners, occurring less frequently, may impose significant demands on plant equipment. In some anticipated transients, rapidly snutting down the nuclear reaction (initiatirig a " scram"), and thus rapidly reducing the generation of heat in tne reactor core, is an important safety measure. If tnere were a potentially severe " anticipated transient" and the reactor snutdown system did not " scram" as desired then an " anticipated transient witnout scram," or ATWS would have occurred.

1 All boiling water reactors, including Dresden Unit 2 have Deen required to provide recirculation pump trip in tne event of a reactor trip and to provide additional operator training for recovery f rom anticipated transient witnout scram events.

2. Justification for Continued Operation A Recirculation Pump Trip (RPT) provision has been incorporated into the Dresden Unit 2 design. An Alternate Rod Injection (ARI) subsystem is currently being installed and is scheduled for completion oy April, 1983. A commitment has also been made to modify the scram discharge system to include two instrument volumes that will incorporate diverse and redundant instrumentation. Inis work is seneduled for completion by Decemoer, 1984. Emergency procedures and operator training to cope witn potential anticipated transient without scram events have been implemented. These procedures and training will be revised as the ARI subsystem and scram discnarge system modifications are completed. Operator training and action as described in the Cordell Reed letter to H.

R. Denton dated March 16, 1982, significantly improved tne capability of the facility to withstand a range of anticipated transient without scram events.

Tne anticipated transient witnout scram rulemaking is currently scneduled for completion by Fall 1982. Based on our review, we feel that there is reasonable assurance that Dresden Unit 2 can be operated prior to ulimate resolution of this generic issue without endangering tne health and safety of the public.

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TASK A-10 BWR Nozzle Cracking

1. Description of Problem A. BWR Feedwater Nozzle Cracking Of the 23 operating BWRs with feedwater nozzle /sparger systems (normally 4 nozzles /spargers per BWR, nominal nozzle diameter being (10"-12")), 21 have been inspected to date (1-26-79) resulting in tne discovery of blend radius or bore cracking in all but three vessels. Altnough most cracks have been in the range of 1/2" to 3/4" total depth (including cladding), one crack penetrated the cladding into the base metal for a total depth of approximately 1.50 inches. The initiation of cracking is due to nign cycle fatigue caused by fluctuations in water temperature within the vessel in the sparger-nozzle region during periods of low feedwater temperature wnen the flow may be unsteady and intermittent. Once initiated, the cracks are driven deeper by the larger pressure and thermal cycles associated witn startup and shutdown.

Fracture analyses indicate that the cracks found to date in the feedwater nozzles constitutes a potential safety problem because the observed rate of crack growth with time in service is sucn that the margin of safety against f racture will be reduced below aceptable values unles the cracks are detected and ground out every few years. Obviously, repair by grindout can be repeated only a few times be' ore ASME Code limits for nozzle reinforcement are exceeded. Powever, repair by welding buildup of tne grindout has not been demonstrated to be acceptable. In addition, the inspection and removal of cracks by grinding has caused enough radiation exposure to personnel to be deemed unacceptable as a long-term solution.

B. Control Rod Drive Hydraulic Return Line Nozzle Cracking (CRDRL Nozzle)

Each of the 22 applicable BWRs nas one CRDRL nozzle of 3"-4" diameter, which is normally located approximately 4 feet below the level of the feedwater nozzles (in the Oyster Creek and Nine Mile Point vessels, the CRDRL nozzle is located at tne same level as the feedwater nozzles). Tnermal fatique cracks l nave been found by dye penetrant (PT) inspection of the CRDRL nozzle and the area immediately beneatn the nozzle at 12 units -

inspected to date (1-25-79). These cracks resemble those found j in the BWR feedwater nozzles, and the cause of cracking appears j to be thermal fatigue. All but two of the operating domestic j BWRs have some sort of thermal sleeve (tnere are several designs) in the CRDRL nozzle, but because of tne limited number l of inspections of nozzles with sleeves, the efficiency of tne sleeves is not known.

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To date, the principal activity of licensees has oeen to reroute or temporarily valve out the CRDRL. Although both accomplisn tne intended purpose of shutting off cold water flow to the nozzle, General Electric Company (GE) nas furtner recommended that the CRD system be operated in an isolated mode. GE recommends against retention of tne present CRDRL, even valved out, because of tne potential for stress corrosion in tne stagnant line. GE also recommends against operation with a rerouted CRDRL open to the reactor vessel. The recommendation to isolate the rerouted line was made on tne basis that return to the vessel is unneccessary for proper CRD system operation and that CRD makeup capability to the vessel will be maintained even when the return line is eliminated entirely.

Tne staff still considers the matter of CRDRL isolation to be an unresolved issue because of questions regarding tne amount of CRD pump flow whicn will be available to the vessel, the possiDie effects of isolation upon various drive parameters, and recently reported potential long-term deleterious effects on certain components of tne CRD nydraulic system. GE nas Degun an evaluation of component performance of affected portions of the CRD nydraulic system and nas commenced investigation of possible system modifications. The staff must assess these proposals prior to completion of its review of tnis subject. In the interim, the staff will review control rod test information from each facility which has modified its present CR0 system by valving out or rerouting. Additionally, to increase assurance of safety for continued operation, that staff is recommending inspections of the CRDRL nozzle blend radius and bore at eacn BWR during its next seneduled refueling outage. As in tne case of feedwater nozzles, we are especially concerned, particularly in the case of older units, that a potential safety problem could raise from deep cracks wnich would necessitate weld repair.

2. Justification for Continued Operation The staff anticipates that this task will result in long-term solutions that will provide: (1) assurance that a conservative margin of safety against vessel failure due to nozzle cracks in '

maintained at operating facilities, (2) more stringent licensing requirements concerning selection of materials and design for nozzles, thermal sleeves, and spargers; (3) more stringent inservice inspection and repair criteria; (4) modification of pnysical systems and/or operating procedures to minimize the occurrence of crack initiation and propagation; and (5) reliable inservice inspection tecnniques f or detection of nozzle flaws f rom positions exterior to the reactor vessel.

With respect to feedwater nozzle cracking, specific long-term corrective measures will include system and operational changes to reduce tne feedwater to reactor water temperature differential during low power operation, an improved thermal sleeve-sparger design to reduce bypass flow which exposes the nozzle surface in fluctuating water temperatures, and removal of clad from the nozzle surface whicn is believed to provide a surface more resistant to fatigue cracking. Implementing some combination of these measures after plants are already under construction or are operating is feasible, e.g., several utilities with operating reactors have already implemented clad removal and the first new thermal sleeve-sparger design has been installed in an operating plant.

With respect to control rod drive return line nozzle cracking, specific long-term corrective measures will include system modifications that assure proper control rod drive system performance witn the return line isolated (if one is installed by design) or eliminated by design. Control rod drive return line isolation nas been implemented at several operating faciliities as an interim corrective measure. Studies are currently underway to determine the acceptability of long-term operation in this manner.

If tnese studies (which are scheduled for completion in early 19/9) demonstrate no degradation of affected components, no further action in tnis regard will be necessary for plants so modified.

During the time period required to develop tne long-term solutions under tnis task, interim measures have been taken. Specifically, as required by the NRC inservice inspection using liquid penetrant examinations are being performed in accordance with the procedures and acceptance criteria set forth in detail in NUREG-0312, " Interim Technical Report on BWR Feedwater and Control Rod Drive Return Line Nozzle Cracking," July 1977. Edison is also utilizing ultrasonic inspection techniques in an effort to develop effective techniques that will allow early detection of subsurface flaws. Enhancement of ultrasonic testin personnel exposures.gTne techniques willand scheduling substantially reduce is extent of inspection based upon conservative estimates of crack growth from fracture mecnanics analyses assuming undetected flaws. Scheduling is tnus dependent upon the reactor's record of past repair (grindouts, clad removal, etc.), operating nistory (number of startup/snutdown cycles since dye-penetrant inspection), and licensee actions to minimize crack initiation by procedural or mechanical change. '

The staff has been actively involved in reviewing and approving the results of nozzle inspections and remedial actions proposed by licensees to assure continued safe operation. To date the extent of nozzle cracking at operating plants has been limited to deptns which Can be removed by grinding without exceeding ASME Code limits for nozzle reinforcement.

E 1 In addition tne staff has suggested that measures be taken at affected operating plants and by applicants for plants in the operating license review stage prior to opeation, to minimize the occurrence growth.

of conditions. conductive to crack initiation and These measures include monitoring feedwater temperatures and flow, minimizing rapid changes in feedwater flow and temperature, minimizing tne duration of cold feedwater injection, avoiding inadvertent or unnecessary HPCI injection, avoiding tne unnecessary introduction of cold water f rom the reactor water cleanup system, and eliminateing flow through the control rod drive return line (after assuring proper system operation in an isolated mode). Altnough cracking of the pressure vessel nozzles is important to safety, NRC staff analyses indicate that cracking that nas penetrated tne vessel cladding will grow at a slow enough rate sucn that the cracking does not pose a critical safety concern today that warrants immediate action. Ratner, the staff believes tnat sufficient time is available, due to the conservative design of the reactor pressure vessel, to permit continued operation of tne affected facilities while studies on tnese events continued on senedule.

Based on the interim measure being taken at operating facilities and the design margins available in the reactor pressure vessel, we nave concluded tnat operation of sucn f acilities does not present an undue risk to the health and safety of the public.

3. Program to Resolve Issue NUREG-0619 "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" contains the NRC required actions to resolve the subject safety issue. The following are references wnich define the Commonwealtn Edison program schedule and action for addressing tne NUREG requirements.

1. Letter f rom R.F. Janecek to Mr. Eisenhut dated February 23, 1981.

2. Letter f rom T.J. Rausen to Mr. Eisenhut dated November 6, 2981.

3. Letter from T.J. Rausch to Mr. Eisenhut dated February 23, 1982.

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TASK A-ll Reactor Vessel Material Toughness

1. Description of Problem Resistance to brittle fracture is described quantitatively by material property generally denoted as " fracture tougnness."

Fracture tougnness has different values and characteristics depending upon the material being considered. For steels used in a nuclear reactor pressure vessel, three considerations are important.

First, fra:ture toughness increases witn increasing temperature; second, fracture toughness decreases with increasing load rates; and tnird, fracture tougnness decreases with neutron irradiation.

In recognition of these considerations, power reactors are operated within restrictions imposed by the Technical Specifications on the pressure during heatup and cooldown operations. Tnese restrictions assure that the reactor vessel will not be subjected to a combination of pressure and temperature that could cause brittle fracture of the vessel if there were significant flaws in the vessel material. The effect of neutron radiation on the fracture toughness of the vessel material over the life of tne plant is accounted for in Tecnnical Specification limitations.

Tne principal objective of Task A-ll is to develop safety criteria to allow a more precise assessment of safety margins during normal operation, transients and accident conditions in older reactor vessels with marginal f racture tougnness.

2. Justification for Continued Operation CECO's letter dated March 31, 1982, from T.J. Rausch to R.R. Denton transmits proposed Tecn. Spec. changes for Dresden Unit 2 regarding reactor vessel tougnness.

In the letter it is empnasized the the beginning inherent jet pump water gap results in low end of lite fluences and subsequently insignificant snift in tne transition temp. due to irradiation.

Furtprmoreit is estimated that tne fluence level of 2X10 n/cm (E7, 1 mev.) in the Reg. Guide 1.99 graph for implementation of Appendix G, pressure temp. requirements will not be reached until 10 Effective Full Power Years. In addition the first Appendix H capsule date will soon become available providing actual snifts in RTNDT.

Based on tne above the RPV has adequate toughness for achieved experienced under the current specification.

Tnerefore, based upon tne foregoing, we conclude that Dresden can be operated prior to resolution of this generic issue without undue risk to the healtn and safety of the public.

TASK A-17 Systems Interaction in Nuclear Power Plants

1. Description of Problems In November 1974, the Advisory Committee on Reactor Safeguards requested that the NRC staff give attention to the evaluation of safety systems from a multi-disciplinary point of view, in order to identify potentially undesirable interactions between plant systems.

The concern arises because the design and analysis of systems is frequently assigned to teams with functional engineering specialities

-- such as civil, electrical, mechanical, or nuclear. Tne question is wnether the work of these functional specialists is sufficiently integrated in their design and analysis activities to enable them to identify adverse interactions between and among systems. Such adverse events might occur, for example, because designers did not assure tnat redundancy and independence of safety systems were provided under all conditions of operation required, which might happen if tne functional teams were not adequately coordinated. Task A-17 was initiated to confirm that present review procedures and safety criteria provide an acceptable level of redundancy and independence for systems required for safety by evaluating the potential for undesirable interactions between and among systems.

2. Justification for Continued Operation Current CECO review procedures and safety criteria provide reasonable assurance that an acceptable level of redundancy and independence is provided for systems that are required for safety. Furthermore, approximately 40 years of reactor operating experience at Dresden and Quad Cities Stations have snown no adverse systems interaction exist.

Therefore, CECO concludes that tnere is reasonable assurance that Dresden Unit 2 can be operated prior to the final resolution of this generic issue witnout endangering the health and safety of the public.

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Task A-24 Environmental Qualification of Safety Related Electrical Equipment 1 Description of Problem The NRC tnrough I.E. Circular 78-08, Bulletin 79-01, Bulletin 79-018 and Supplements, and Commission Memorandum and Order (CHI-80-21) brought to tne attention of the nuclear industry a potential Equipment.problem - the environmental qualification of Class IE Tnese NRC actions require the review of Class IE Equipment subjected to a harsh environment (resulting from a LOCA or HELB) to determine if the equipment can fulfill its design function in such an accident environment.

2. Justification for Continued Operation CECO nas provided the NRC with four major submittals which address justification for continued operation. The original System Component Evaluation Work sheets, reference 1 below, and most recently tne SCEW sheets associated with TMI Section Plan Equipment, of equipment reference 2, contain such justification for those pieces lacking proper documentation. In response to the NRC requestoffor reviews CECO additional justification resulting from preliminary submittals, continued Dresden Unit 2 operation. reference 3 was submitted Finally, in responsesupporting to the NRC's SER and Frankline's TER CECO submitted additional material justifying continued operation of Dresden 2 until equipment testing and/or replacement programs are completed.

1.

"Dresden Station UniE 2 Response to NRC Order Concerning Environmental Qualification of Class IE Electrical Equipment,"

J.S. Aoel to D.G. Eisenhut, October 31, 1980, NRC Docket No.

50-237.

2.

"Dresden Station Units 2&3 and Quad Cities Station Units 1&2 Environmental Qualification of TMI Action Plan Equipment,"

E.D. Swartz to D.G. Eisenhut, August 2, 1982, NRC Docket Nos.

50-237/249 and 50-254/265.

3.

"Dresden Station Unit 2, Environmental Qualification of Safety Related Electrical Equipment," J.S. Abel to H.R. Denton, dated March 2, 1981, NRC Docket No. 50-237. .

4.

"Dresden Stations Units 2&3 and Quad Cities Stations Units l&2 l Environmental Qualification of Safety Related Class IE Electrical Equipment," T.J. Rausch to H.R. Denton, dated September 4, 1981, NRC Docket Nos. 50-237/249 and 50-254/265.

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3. Corrective Action For equipment addressing equipmentlacking proper qualification documentation a scnedule developed. testing and/or replacement has been Tnis senedule is commensurate with the NRC draft schedule for completion of equipment qualification (the end of the senedule refueling outage originating after March 31, 1982). Tne test program targeted is presently for completion byunder way at Wyle Laboratories and is Mid-1983.

presently installed equipment Owner's Group testing of for completion by the Fall of 1983.is about to commence and is targeted Equipment replacement and relocation efforts are underway and will result in equipment changeouts commencing in begining January,during 1983. theThe presently scheduled refueling outage existing surveillance and maintenance program is qualification requirements.being reviewed to insure compliance with o

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TASK A-31 Residual Heat Removal Requirements

1. Description of Problem Long term cooling of the reactor coolant system is required in order to perform inspection and repairs. For this reason, the ability to transfer an heat safety important from tne reactor to tne environment after shutdown is function. It is essential tnat a power plant nave tne capability long term basis.

to remain in the cold-shutdown condition on a

2. Justification for Continued Operation The shutdown cooling system is available after the reactor coolant system has been sufficiently cooled and depressurized. The design of the 0

system is based on cooling the reactor coolant system from 350 F to 1250F witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown. The system consists of tnree loops which tie into the suction line of the reactor recirculation pumps. Each loop contains one pump and one neat exchanger which is cooled by tne reactor building closed cooling water.

Operating experience has shown that, at only eight nours after normal (main condenser) shutdown, only one pump and one heat exchanger are necessary to cool down. The flow path continues into the low pressure coolant injection system and then into each of the reactor recirculation loops. Various temperature and pressure interlocxs must be met before startup of the system. There is diversity in tne AC, DC, and emergency diesel power supplies to assure system isolation and protection.

Tne low pressure coolant injection system can be used if the snutdown cooling system is inoperable. The low pressure coolant injection system is capable of i njecting cooling water f rom the suppression pool or the contaminated condensate storage tank. This water is cooled by passing through the heat exchangers which are cooled by the containment cooling service water. The other modes of tne low pressure coolant injection system are suppression pool cooling, containment spray and suppression pool spray. The suppression pool cooling and suppression pool spray will condense steam and maintain the water temperature during regular operation and accident conditions. The containment spray will cool the air space inside primary containment tnrough a spray ring header, ,

The low pressure coolant injection system and containment cooling service water system have redundant trains. The low pressure coolant injection pumps are powered from essential service buses, and all motor-operated valves are powered from essential service motor control centers and are also accessible for manual operation if needed.

Based on the aDove, we conclude that Dresden Unit 2 can be operated prior to ultimate resolution of this generic issue witnout undue risk to the healtn and safety of the public.

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TASK A-36 Control of Heavy Loads Near Spent Fuel

1. Description of Problem In nuclear plant operation, maintenance and refueling activities, neavy loads may be handled in several plant areas. If these loads were to drop, they could impact on stored spent fuel, fuel in the core, or equipment that may be required to acnieve safe shutdown or permit continued decay heat removal. If sufficient stored spent fuel or fuel in the core were damaged and if the fuel is highly radioactive due to its irradiation history, the potential releases of radioactive material could result in offsite doses that exceed 10CFR Part 100 limits. If the load damaged equipment associated witn redundant or dual safe shutdown paths, the capability to achieve safe snutdown may be defected.

In this task a heavy load is defined as a load whose weight is greater than tne combined weight of a single spent fuel assembly and its handling tool.

Task A-36 was established to systematically examine staff licensing criteria and the adequacy of measures in effect at operating plants, and to recommend necessary cnanges to assure the safe handling of heavy loads once a plant becomes operational.

Additionally, with the increased spent fuel storage capacities at many operating plants, largely in tne form of increased density of fuel storage within tne pool, the potential for a given load to damage a large number of fuel assemblies nas increased.

2. Justification for Continued Operation As a result of NUREG-0612, CECO. has made several suomittals on control of heavy loads at Dresden. On February 22, 1982, a telepnone conference call was neld between the NRC Staff, the Franklin Researen Center (FRC) and CECO to discuss FRC draf t Tecnnical Evaluation Reports (TER's) concerning " Phase I" control of neavy loads at our Dresden and Quad Cities Stations.

As a result of our review of the draft TER's, and to document the conference call discussions, CECO provided their response to each ,

concern and recommendation that was identified by tne FRC in their TER's. From the response, along with the initial submittals for Dresden and Quad Cities, it is our understanding that this will form tne basis for a final Technical Evaluation Report from FRC for each station and ultimately the NRC Staff Safety Evaluation of "Pnase I" of tnis issue.

Tnerefore CECO concluces that there is reasonable assurance that Dresden Unit 2 can be operated prior to the final resolution of this generic issue witnout endangering the nealth and safety of the public.

I TASK A-40 Seismic Design Criteria - Snort Term Program f 1. Description of Problem Tne seismic design process required by current NRC criteria includes the following sequence of events.

A. Define the magnitude or intensity of the earthquake which will produce tne maximum vibratory ground motion at the site (the saf e shutdown earthquake or SSE) .

B. Determine the free-field ground motion at the site that would result if the SSE occurred.

C. Determine tne motion of site structures by modifying the f ree-field motion to account for the interaction of the site structures with tne underlying foundation soil.

D. Determine tne motion of the plant equipment supported by the site structures.

E. Compare the seismic loaJs, in appropriate combination witn otner loads, on structures, systems, and components to safety, with the allowable loads.

Wnile this seismic design sequence includes many conservative factors, certain aspects of the sequence may not be conservative for all plant sites. At present it is believed tnat the overall sequence is adequately conservative. Tne objective of this program is to investigate selected areas of the seismic design sequence to determine tneir conservatism f or all types of sites, to investigate alternate approaches to parts of the design sequence, to quantify the overall conservatism of the design sequence, and to modify the NRC criteria in the Standard Review Plan if changes are found to be justified. In tnis manner this program will provide additional assurance that tne healtn and safety of the public is protected, and if possible, reduce costly design conservatisms by improving (1) current seismic design requirements, (2) NRR's capability to evaluate tne adequacy of seismic design of operating reactors and plants under construction, and (3) NRR's capability to quantitatively assess tne overall adequacy of seismic design for nuclear plants in general. '

2. Justification For Continued Operation The objective of tne aforementioned task is to investigate selected areas of seismic design to determine tne conservatism for Dresden, to investigate alternate approaches to parts of tne design sequence, to gauntify tne overall conservatism of the design sequences and to modify tne licensing criteria if cnanges are found to be justified.

Tne results of tne task will be applicable to Dresden Nuclear Power Station.

It is anticipated that the results of tnis task will provide confirmation tnat current requirements provide an overall conservative approach to seismic design. The general result that is anticipated from this task is the development of better insight into seismic design considerations that will permit establishment of a set of integrated requirements providing for more realistic and effectsve designs witnout a loss of overall margin.

, Tnree general types of results are expected from this task. The l first is the ability to select seismic design grnund motion inputs for Dresden that are more appropriate for the site and thus will result in a more consistent level of seismic design.

Second, it is expected that these investigations will demonstrate tnat tne current metnods of analysis are conservative in relation to otner methods that could be justified and to provide a quantitative idea of how conservative tney are. Third, it is expected that this effort will demonstrate that tne overall safety margins attained using current methods are considerable. In the interim, it is believed that continuation of tne current licensing requirements will assure an acceptable level of safety in plant seismic design.

If the results of tnis task action plan are not as anticipated and the current criteria prove not to be adequately conservative, these results will not affect the seismic design criteria at Dresden; Decause the original design criteria as specified in the FSAR are j genefallymoreconservativethanthecurrentdesigncriteria.

Based on the discussion above, it is concluded that while this task is being perf ormed, continued operation and plant licensinq can proceed witn reasonable assurance of protection to the nealth and safety of tne public.

I 3. Added Tecnnical Justification The Systematic Evaluation Program (SEP) is currently in progress at i

Dresden Station. Tne last SER written June 30, 1982, Docket No.

l 50-237, LS05-82-06-130, SEP SAFETY TOPIC No. III-6, TITLE Seismic I Design Considerations, identified three concerns as identified in tne

! following.

Tne staff concerns are 10 CFR 50 (GDC 2), as implemented by SRP '

l Sections 2.5, 3.7, 3.8, 3.9, and 3.10 and SEP review criteria (NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants"), requires that structures, systems and components important to safety shall be designed to withstand the effects of natural pnenomena such as eartnquakes. The following l differences were identified:

j 1. Piping Systems - Tne staff has identified deficiencies regarding i

the existing piping supports. Therefore, we are unable to conclude tnat safety-related piping systems are capable of witnstanding the postulated SSE loads.

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2. Mechanical Equipment - The staff lacked sufficient design information for 3 out of 9 of the mechanical equipment items sampled. Therefore, the staff cannot conclude that mech'anical equipment is adequately designed.

3. Programs undertaken by the SEP Owners Group are intended to provide a set of general analytical methodologies for tne seismic qualification of cable trays and for documentation of the functionability of other safety-related electrical equipment subjected to seismic loads; tnese programs have not been completed.

Item 1 is being satisified by I&E review of Bulletin 79-14. As a

. result, piping supports are being modified and added to safety-related piping systems to upgrade the systems to withstand postulated SSE loads.

The current schedule to complete all modification work i s December 31, 1983. For the interim, operability studies nave been completed on safety-related piping and nave demonstrated that Dresden can be safely snutdown during a postulated SSE event.

Item 2 Sargent and Lundy completed a special study on the subject mecnanical equipment, which demonstrated that the subject mechanical equipment was adequately designed.

Item 3 is still in progress URS/Jonn A. Blume and Associates are perf orming analyses to qualify cable trays and developing documentation to demonstrate the functionability of otner safety-related electrical equipment subject to seismic loads. This study is scheduled to be completed in October 1982.

Based on tne SEP conclusions generated thus far and results of IE Bulletin 79-14, continued operation and plant licensing can proceed witn reasonable assurance of protection to tne health and safety of tne public.

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TASK A-42 Pipe Cracks at Boiling Water Reactors

1. Description of Problem Pipe cracking has occurred in the heat affected zones of welds in primary system piping in boiling water reactors (BWRs) since the mid-1960's. These cracks have occurred mainly in Type 304 stainless steel, which is the type used in most operating BWRs. The major problem is recognized to be intergranular stress corrosion cracking ,

(IGSCC) of austenitic stainless steel components that have been made l I

susceptible to this failure by being " sensitized," either by post-weld heat treatment or by sensitization of a narrow heat affected zone near welds.

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" Safe ends" (short transition pieces between vessel nozzles and the j piping) that have been nignly sensitized by furnace heat treatment l while attacned to vessels during fabrication were very early (late l 1950's) found to be susceptible to IGSCC. Because of this, the Atomic Energy Commission took the position in 1969 that furnace- j sensitized safe ends snould not be used on new applications. Most of the furnace-sensitized safe ends in older plants have been removed or clad with a protective material, and there are only a few BWRs that still have furnace-sensitized safe ends in use. Most of these, however, are in smaller diameter lines. A Earlier reported cracks (prior to 1975) occurred primarily in 4-inch diameter recirculation loop-bypass lines and in 10-inch diame te r core l spray lines. More recently cracks were discovered in recirculation riser piping (12-inch to 14-inch) in foreign plants. Cracking is most often detected during Inservice Inspection using ultrasonic

, testing tecnniques. Some piping cracks have been discovered as a result of primary coolant leaks.

In response to these occurrences of BWR primary system cracking, the NRC has taken a number of measures. These actions included:

Issuance of Regulatory Guide 1.44 on " Control of tne Use of Sensitized Stainless Steel."

Issuance of Regulatory Guide 1.45 on " Reactor Coolant Boundary

Leak Detection Systems." ,

I Closely following the incidence of cracking in BWRs, including foreign experience.

Encouraging replacement of furnace-sensitized safe ends.

l Requiring augmented in-serivce inspection (additional more frequent ultrasonic examination) of " service sensitive" lines, i.e., tnose tnat have experienced cracking.

l l Requiring upgrading of leak detection systems.

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, 2. Justification for Continued Operation

. CECO's letter f rom L.0. DeiGeorge to D.G. Eisennut, dated 7/7/81, and its references addresses this problem at developed by NUREG-0313, Rev. 1. CECO feels, because of the leak-before-break criterion, that pipe cracking is an availability problem and is not a safety issue.

Therefore, based upon the foregoing, we conclude that Dresden can be operated prior to resolution of this generic issue without undue risk to the healtn and safety of tne public.

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TASK A-43 Containment Emergency Sump Reliability

1. Description of Problem Following a postulated loss-of-coclant accident, i.e., a break in the reactor coolant system piping, the water flowing from the break would be collected in the suppression pool. This water would be recirculated through the reactor system by the emergency core cooling pumps to maintain core cooling. This water may also be circulated tnrough the containment spray system to remove heat and to draw water from tne suppression pool could disable the emergency cooling and containment spray systems.

Tne concern addressed by this Task Action Plan for boiling water reactors is limited to the potential for degraded emergency core cooling system performance as a result of thermal insulation debris that may be blown into tne suppression pool during a loss-of-coolant accident and cause blockage of the pump suction lines. A second concern, potential vortex formation, is not considered a serious concern for Mark I Containment due to tne large depth of the pool (approximately 25 feet) and the low approacn velocities.

2. Justification for Continued Operation With regard to potential blockage of the intake lines, the likelihood of any insulation being drawn into a emergency core cooling system pump suction line is very small. The potential debris in the drywell could only be swept into the suppression pool via the downcomer piping. However, tne downcomer pipes (approximately two feet in diameter) are capped with jet deflectors and would prevent any large piecas from reaching the suppression pool. Any smaller pieces reacning tne pool would tend to settle on the bottom and would not be

! drawn into tne pump suction since it is located several feet above tne pool bottom. In addition, boiling water reactor designs employ strainers within the suction piping, and net positive suction nead calculations for the pump are based on an assumed 50 percent blockage.

1 Accordingly, we conclude that Dresden can be operated prior to ultimate resolution of this generic issue without endangering the health and safety of the public.

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TASK A-44 Station Blackout

1. Description of Problem Electrical power f or saf ety systems at nuclear power plants must be supplied by, at least, two redundant and independent divisions. The systems used to remove decay heat to cool the reactor core f ollowing i a reactor shutdown are included among the safety systems tnat must j meet tnese reouirements. Each electrical division for safety systems 1 includes an of f site alternating current power connection, a standby <

emergency diesel generator alternating current power supply, and l direct current sources. i l

Task A-44 involves a study of whether or not nuclear power plants i should De designed to accommodate a complete loss of all alternating i current power, i.e., a loss of both the offsite and the emergency diesel generator alternating current power supplies. This issue '

arose because of operating experience regarding the reliability of alternating current power supplies. A number of operating plants have experienced a total loss of offsite electrical power, and more occurrences are expected in the future. During each of these loss-of-off site power events, the onsite emergency alternating current power supplies were available to supply the power needed by vital safety.eoutpment. However, in some instances, one of the redundant emergency power supplies nas been unavailable. In addition, there have been numerous reports of emergency diesel-generators failing to start and run in operating plants during periodic surveillance tests.

2. Justification for Continued Operation A loss of all alternating current power was not a design basis event for the Dresden Unit 2 facility. However, tnere are a number of items that assure safe continued operation.

First, Dresden Unit 2 Auxiliary Power System is connected to a highly reliable power grid. Sources of offsite power are available from ootn the 345KV and 138KV systems througn various switching comoinations. To date, Dresden 2 nas never experienced a total loss of offsite power and the probability of such an event has been snown to be extremely low.

l Second, if by some remote chance, offsite alternating current power ,

is lost, three diesel-generators and their associated distributed systems will deliver emergency power to saf ety-related eouipment.

Historical records have proven that tnese diesels are highly reliable and operational experience has proven this f act. Maintenance and surveillance procedures are currently in effect wnich will assure that tnis hign reliability is maintained at all times.

Tntrd, if botn offsite and onsite alternating current power are lost, the isolation condenser and HPCI may be used to remove core decay heat without reliance on alternating current power. This will assure tnat adeouate cooling can De maintained during the brief time period until eitner off site or onsite power sources are restored.

r Finally, specific Blackout Training Procedures have been implemented to assure to;t adequate core cooling will be maintained and that power will be restored in a timely manner.

Based on the above, CECO has concluded that tnere is reasonable assurance that Dresden Unit 2 can be operated prior to the ultimate resolution of tnis generic issue witnout endangering the nealth and safety of the puolic.

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Task A-45 Snutdown Decay Heat Removal Requirements

1. Description of Proolem Following a reactor shutdown, the radioactive decay of fission products continues to produce heat (decay heat) whicn must be removed f rom t he p rimary sytem. Tne alternative means of core cooling will be evaluated to determine the adequacy of the decay heat removal systems at Dresden Unit 2.

2. Justification for Continued Operation Tne principal means for removing heat in a boiling water reactor while at high pressure is via the steam lines to the turbine condenser. The condensate is normally returned to the reactor vessel by tne feedwater system. However, the isolation condenser is provided for core cooling in the event that the reactor becomes isolated from the main condenser. Tne hign pressure coolant injection system can provide makeup and cooling. Botn the isolation condenser and the high pressure coolant injection system have redundant cooling water systems.

If tne isolation condenser and the high pressure coolant injection system are unavailaole, the reactor system pressure can be reduced by the automatic depressurization system so tnat cooling by the low pressure coolant injection system and core spray systems can be initiated. The neat rejected to the suppression pool is removed through the low pressure coolant injection system heat excnangers which is cooled by the containment cooling service water.

The normal mode of snutdown is to remove decay heat througn the shutdown cooling system. The function of this system is to cool and maintain primary water temperature at 1250F.

Dresden Unit 2 has a dedicated diesel generator which has the capability of providing power to the equipment for Engineered Safety Systems Division II. A swing diesel generator will provide power to Engineered Safety Systems Division I. Tne capability has been provided, as an additional safety feature, for the Unit 3 diesel generator to provide power to tne Unit 2 diesel generator's Dus l

tnrough a Dus tie joined by two normally open circuit breakers. ,

Based on the above, we conclude that Dresden Unit 2 can be operated prior to ultimate resolution of this generic issue witnout endangering tne healtn and safety of the public.

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.T AS.K A-4 6_ S e i s_mic - Qu_a _l i f i c a t i o n o f Eq u_4 pme_n t - i n _ Op e r a t i n g - P l a n t s

1. Description of Problem The design criteria and methods for the seismic qualification of i

mechanical and electrical equipment in nuclear power plants have t undergone significant change during the course of the commercial nuclear power program. Consequently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably. The seismic qualification of the equipment in operating plants must, therefore, be reassessed to ensure the ability to bring the plant to a safe i

' shutdown condition when subject to a seismic event. The objective of this unresolved Safety Issue is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants in lieu of attempting to backfit current design criteria for new plants. This guidance will concern equipment required to safely shutdown the plant, as well as equipment whose function is not required for safe shutdown, but whose failure could result in adverse conditions which might impair shutdown functions.

2. Justification for Continued Operation Although many operating plants were designed before the development of current licensing criteria, the design rules and procedures incorporated inherent conservatisms. These include: (1) the margins between allowable stresses and ultimate strength of engineering materials, (2) the methods used for combining loads, (3) the inherent ducitility of materials, and (4) the seismic resistance of nonstructural elements which are not normally considered in design calculations.

An expanding data base of observations at large industrial facilities l

that have experienced strong ground motfon suggests that these facilities have significant seismic resistance capabilities. From the data, it can be concluded that the inherent seismic resistance of engineered structures and equipment is usually much greater than is assumed in both past and current analysis and design procedures.

Even facilities designed with very nominal seismic considerations, have been able to withstand severe seismic environments without loss of safety function. When even the most modest attention is paid in design to providing lateral loaf Carrying paths, significant capability results. Nuclear p's e plants have been designed using '

more vigorous techniques; th;' eft 'e, it is reasonable to expect even higher inherent margins tA& .cr .mplied from the data base of t observations. Because of .e :. vrience gained in the reivew of the SEP facilities and the continued staff review of seismic issues, it is concluded that operating plants can continue to operate without endangering the health and safety of the public, pending resolution of this Unresolved Safety Issue.

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Dresden Unit 2 nas also been seismically reviewed by the NRC's SEP program. The report on tnis review was published as NUREG/CR-0891.

Tne review concluded in Section 2.4 that in general Dresden Unit 2 could adequately resist an earthquake with a SSE value of acceleration of 0.2 g. Although this conclusion is predicated on certain assumptions, the following items support or make conservative this conclusion:

1. The NRC nas determined tnrough tne SEP Program that the Dresden 2 ground acceleration is 0.13 g- this is 35% less than the NUREG document found acceptable. (Reference letter from D.C.

Crutenfield to SEP Owners, June 17, 1981.)

2. In response to SEP findings (I.E. Information Notice 80-21) electrical equipment anchorages have been reviewed and upgraded as required.

3. In response to I.E. Bulletin 79-14 the "as-built safety related piping systems" are being reanalyzed and resupported as required.

-3. Corrective Action In response to SEP topic III-6 and the NRC's Senior Seismic Reivew Teau's report on Dresden Unit 2, CECO has initiated several actions to address NRC concerns. Tney are:

1. A tnorougn review was conducted by CECO of electrical equipment anchorage. This included a field walkdown of the anchorage of sucn equipment , seismic analyses and modifications to the anchorages if they were determined to provide less than the desired margin of safety. -

2. Tne wooden bracing on the battery racks will be replaced by angle iron to improve the margin of safety during seismic events.

3. CECO is involved in a SEP Owner Group witn J.A. Blume &

Associates trays.

to investigate the seismic adequacy of rod hung cable

4. CECO is also involved in an owners group which is presently -

investigating the response of equipment which nas been through and earthquake.

5. Response to I.E. Bulletin 79-14 has resulted in substantial reanalysis and modifications to supports for "as-built" safety related piping systems. Tnis work is presently in progress.

TASK A-47 Safety Implications of Control Systems

1. Description of Problem This issue concerns the potential for transients or accidents being made more severe as a result of control system failures or malfunctions. These failures or malfunctions may occur independently or as a result of the accident or transient under consideration. One concern is the potential for a single failure such as a loss of power supply, short circuit, open circuit, or sensor f ailure to cause simultaneous malfunction of several control features. Sucn an occurrence would conceivably result in a transient more severe than those transients analyzed as anticipated operational occurrences. A second concern is for a postulated accident to cause control system failures whicn would make the accident more severe tnan analyzed.

Accidents could conceivably cause control system failures by creating a narsh environment in the area of the control equipment or by pnysically damaging the control equipment. Although it is generally believed tnat such control system failures would not lead to serious events or result in conditions that safety systems cannot safely handle, in-depth studies have not been rigorously performed to verify tnis belief. Tne purpose of this " Unresolved Safety Issue" is to define generic criteria that will be used f or plant-specific review.

2. Justification for Continued Operation Tne Dresden Unit 2 control and safety systems have been designed with tne goal of ensuring that control system f ailures (either single or multiple fatiures) will not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a safe snutdown condition following any " anticipated operational occurrence" or " accident". Inis nas i been accomplished by either providing independence between safety and l nonsafety systems or providing isolating devices between safety and l nonsafety systems. Tnese devices preclude the propagation of I nonsafety system equipment faults such that operation of tne safety equipment is not impaired, l

A systematic evaluation of the control system design, such as contemplated for this " Unresolved Safety Issue," nas not been performed to determine whether postulated accidents could cause significant control system failures wnich would make the accident -

consequences more severe than presently analyzed. However, operating l experience of more than 40 reactor years nas shown that control l system failures can be corrected by operator or automatic actions.

l To date, no control system failure has resulted in a significant l safety hazard.

Based on the above, CECO has concluded that there is reasonable assurance tnat Dresden Unit 2 can be operated prior to tne ultimate l

resolution of tnis generic issue witnout endangering the health and l

safety of tne public.

TASK A-48 Hydrogen Control Measures and Effects Of Hydrogen Burns on Safety Equipment

1. Description of Proolem Following a loss-of-coolant accident in a light water reactor plant, combustiole gases, principally hydrogen can accumulate inside tne primary reactor containment as a result of: (1) metal-water reaction involving tne fuel element cladding; (2) the radioactive decomposition of the water in tne reactor core and the containment sump; (3) the corrosion of certain construction materials by the spray solution; and (4) any synergistic cnemical, thermal and radiolytic effects of post-accident environmental conditions on ,

containment protective coating systems and electric cable insulation.

Because of the potential for significant generation as the result of an accident, 10 CFR Section 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," and Criterion 41 of the General Design Criteria, " Containment Atmosphere Cleanup,"

in Appendix A to 10 CFR Part 50, require that systems be provided to control hydrogen concentrations in the containment atmosphere following a postulated accident to ensure tnat containment integrity is maintained.

On December 21, 1981, the NRC puolished a final ruling in the Federal Register concerning interim hydrogen control f or BWR primary containments. Tnis new ruling, 10 CFR 50.44(c)(3)(ii), requires that BWR plants witn Mark I or Mark 11 containments relying on purge /repressurization as the primary means of combustible gas control be provided with eitner an internal hydrogen recombiner or be provided with the capability to install an external recombiner following an accident. This new ruling requires tnat modifications be completed by the end of the first scheduled refueling outage of sufficient duration beginning after July 15, 1982.

2. Justification of Continued Operation Tne primary containment at Dresden Unit 2 relies on nitrogen inerting as tne primary means of combustible gas control. In addition to the primary containment nitrogen inerting, an Air Containment Atmospneric Dilution (ACAD) system and a Containment Atmospheric Monitoring (CAM) system have also been installed to monitor and control hydrogen gas concentration in the primary containment during a postulated LOCA.

It should be noted that a commitment has been made to upgrade the existing CAM system. This modification work is scheduled for completion by December, 1983.

Having technically reviewed the most current provisions stated in 10CFR50.44 (c)(3)(ii), it is tne position of Commonwealth Edison that since tne primary containment at Dresden Unit 2 relies on nitrogen inerting rather purge /repressurization (ACAD) as a primary means of comoustible gas control, installation of hydrogen recombiners is not required. Technical justification for this position is based on the results of a recent analysis performed by the General Electric Company, " Generation and Mitigation of Combustible Gas Mixtures in

r 1 1

1 Inerted BWR Mark I Containments" (NED0-22155). Results of the analysis performed snow tnat following a postulated LOCA, peak oxygen concentrations found witnin a BWR Mark I containment would remain Delow the Combustible gas limits at"all times Without the need for containment venting or hydrogen recombiners. The results of the analysis also show that an ACAD system is not required. In the event of a postulated LOCA, ACAD would further pressurize the drywell, and without controlled venting, would not effectively control combustible gases.

With regard to the results of this analysis and in consideration of the commitment to inert the Dresden Unit 2 primary containment with nitrogen, we conclude that Dresden Unit 2 can continue to operate without undue risk to the health and safety of the public.

i 22490*

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