ML17174A408

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Forwards Draft Evaluation of SEP Topic III-5.B, Pipe Break Outside Containment. Requests That Util Address Outstanding Times & Respond within 30 Days
ML17174A408
Person / Time
Site: Dresden Constellation icon.png
Issue date: 01/17/1980
From: Ziemann D
Office of Nuclear Reactor Regulation
To: Peoples D
COMMONWEALTH EDISON CO.
Shared Package
ML17174A409 List:
References
TASK-03-05.B, TASK-3-5.B, TASK-RR NUDOCS 8002060388
Download: ML17174A408 (21)
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-rffie Local PDR Docket No. 50-237 ORB. Reading

DEisenhut Mra O. Loui& Peoples; Director of Nucl~ar L1censing-Cornmonwealth Edi son Company.

p. o. Box 767 Chicago.* I111nots 80690

Dear Mr. Peoples:

RVollmer OELD.

OI&E (3)

DLZiemann 0

1 Connor.

..HSmi th NSIC.

TERA ACRS (16)

DCRutchfield (2)

RE:* Topic III-5.B - Pipe-Break.Outside Containment.*

JAN 1 7 1980 Enclosed 1s a copy of *QUr draft *evalu~tion of Sy$tematic E*val.uation Program Topic II I-5.B-.

You are requested to examtne the facts* upon which the staff has -bas'ed its evaluation and respond 'either by confinning that the facts. '

are cor,rect *. or by fdent1fy1ng any errors. lf 1n error. please supply corrected

  • information for the docket.

We encourage-you tQ supply for the. docket any o~her material related to these topics that mig!lt a.ff_ect th'e ~taff's evaluation.

  • Your> response within 30 days of the date you *receive this letter *is requested.
  • If no *response is received within, that. time. we w11l assume th_at you.have

-no comments or corrections.

Enclosure:

Topic Ill-5~J3 cc w/encl9sure: _

s~e nex~ 'PC)9e. --~

  • s1 ncere'ly;
  • Original Sign81 by:
  • ~ennis t. Zieillanri

. Dennis L. Ziemann, Chief Operating-Reactors Branch #2 Division of Operating Reactors

  • --~--

.* L. ~

8002060,3g-8

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DATE

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'.... D. Louis Peoples cc Isham, Lincoln & Beale Counselors at law One First National Plaza, 42nd Floor Chicago, Illinois 60603 Mr. B. B. Stephenson Plant Superintendent Dresden Nuclear Power Station Rural Route #1 Morris, Illinois 60450 Anthony Z. Roisman tiatur2l Resources Defense Council 917 15th Street, N. W.

~ashir.gton, D. c.

20005 U. S. Nuclear Regulatory Commission ATTN:

Jimmy L. Barker P. 0. Box 706

~orris, Illinois 60450 Susan N. Seku l er Assi s:ant Attorney General

  • Environr.!ental Control Division 188 W. Randolph Street Suite 2315.

Chicago, Illinois 60601

~orris Public Library 604 Lfberty Street

~orris, Illinois 60451 Cha irnan Board of Supervisors of Grundy County Grundy County Courthouse

~orris, Illinois 60450 January 17, 1980 Department of Public Health ATTN:

Chief, Division of

  • Nuclear Safety 535 West Jefferson Springfield, Illinois 62761 Director, Technical Assessment Division Office of Radiation Programs (AW-459)

U. S. Environmental Protection Agen.cy

EI.S COORDINATOR 230 Sout~ Dearborn Street Chicago, Illinois 60604 Neil Smith Commonwealth Edison Company P. 0. Box 767 Chicago, Illinois 60690

SEP REVIEW OF PIPE BREAK OUTSIDE CONTAINMENT TOPIC III-5. B FOR THE

e.

DRESDEN NUCLEAR POWER PLANT UNIT 2

--./

8002060 3~/

e.

INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.B, 11 Pipe Break Outside Containment" is to assure that pipe breaks would not cause the loss of needed functions of safety-related systems, structures and components and to assure that the plant can be safely.shut down in the event of such breaks.

The needed functions of safety-rela~ed systems_ are those functions required to mitigate the effects of ~he pipe break.

The current criteria for review of pipe breaks outside containment are contained in Standard Review Plan 3.6.1 and 3.6.2 inc1uding their attached Branch Techn~cal Position~.

BACKGROUND In December 1972, the staff sent letters (Reference 1) to a.11 power reactor licensees requesting an analysis of the effects of postulated failures of high energy lines outside of containment.

A summary of the criteria and requirements in this letter is set forth below:

a.

Protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protected equipment, should be provided from all effects re~ulting from ruptures in pipes carrying*high energy fluid, where the temperature and pressure conditions of the fluid exceed 200°F and 275 psig, respectively, up to an including a double-ended ruputre of such pipes.

Breaks should be assumed to occur in those locations specified in the 11pipe whip criteria".

The rupture effects to be considered include pipe whip, structural (including the effects of jet impingement), and environmental.

b.

In addition, protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated.single active failure of protected equipment, should be provided from the environmental and structural

effects (including the effects of jet impingement) resulting from a single open crack at the most adverse location in pipes carrying fluid routed in the vicinity of this equipment.

The size of the cracks should be assumed to be 1/2 the pipe diameter in length and 1/2 the wall thickness in width.

In response to our letter, Commonwealth Edision submitted Dresden Station Units 2 & 3 Special Repo~t No. 37 "Analysis of Effects of Pipe Break Outside the Primary Containment", dated January 23, 1974, and Supplement 1 and Revision 1 to this report dated March 22, 1974 and February 18, 1975, respectively (References 2, 3, 4).

These submittals were reviewed and accepted by the staff in our safety evaluation report of May 12, 1976 (Reference 5).

On November 17, 1977, the licensee submitted Appendix E to Special Report No. 37 which analyzed the effect of pipe breaks outside containment on electrical conduits and cable trays in Dresc;ien 2 & 3 and ident*ified areas requiring modifications (References 6).

The NRC staff reevaluation of the effects of pipe breaks outside containment under SEP Topic III-5.B includes the comparison of Dresden 2 with current criteria for pipe breaks outside containment and the evaluation of Appendix E to Special Report No. 37.

The staff used an "effects oriented 11 approach to determine the acceptability of plant response to pipe breaks, i.e. each structure, system, component, and power supply which must function to mitigate the effects of the pipe break and to safely,shutdown the plant was examined to determine its susceptibility to the effects o.f the postulated break.

Break effects considered were compartment pressurization, pipe whip, jet impingement, spray, flooding, and environmental conditions of _temperature, pressure, and humidity.

2

(The effects of potential missiles generated by fluid system ruptures and rotating machinery were also consi.dered and are evaluated under SEP Topic III-4.C, "Internally Generated Missiles 11.)

Data for this assessment was gathered during a visit to the Dresden 2 plant on August 27-30, 1979.

EVALUATION The results of the SEP reevaluation of pipe breaks outside containment of Dresden 2 are provided in Table 1.

The following paragraphs provide additional information used to evaluate certain pipe breaks listed in Table 1.

Current criteria for pipe breaks outside containment assume the loss of offsite power if the pipe break results in a reactor or main turbine trip.

The Dresden 2 facility has never experienced a total loss of offsite power resulting from a unit trip (reactor and turbine).

The Dresden Station has experienced only one total loss of offsite power and this was caused by a tornado on November 12, 1965.

Therefore, for the Dresden site, the assumed loss of offsite power resulting from a unit trip is extremely conservative.

The safe shutdown systems which were examined from*the standpoint of protection from pipe break effects are identified in the SEP Safe Shutdown Review fqr Dresden 2 (Reference 8). These systems are:

3

(a) Reactor Control and Protection System (b) Pressure Relief System (4 electromatic/1 electro-pneumatic relief valves)

(c)

Low Pressure Coolant Injection System (d) High Pressure Coolant Inje~tion System (e) Containment Cooling Service Water.

(f) Instrumentation of shutdown/cooldown (g)

Emergency power (AC and DC) and control power for the above systems and components.

Table 1 provides an assessment of circulating water (CW) system breaks in the crib house.

Reference 7 contains an evaluation of the effects of a CW line break in the condenser pit of the turbine building.

To prevent a break in the condenser pit from flooding ciitical equipment, the licensee has installed flood doors between the condenser pit and the condensate pump room.

Permanent flood level switches have been installed in the pit to warn the operator of a flooding condition and to trip the CW pumps if necessary.

The flooding rates used in Reference 7 are much greater than would be assumed using current MELB criteria since. the licensee assumed a complete severance of a CW line. Also, as part of the permanent modifications resulting from the analysis in Reference 7, the diesel generator cooling water pumps (#2, #3, and #2/3) in the crib house*

have been replaced with submersible pumps, and containment cooling service water pumps 28 and 2C have been surrotinded with a watertight enclosure which extends from the 495' elevation to the 511 1 elevation.

The maximum historical fJood level for the river at the crib house is 508 1

4

At present, the fire protection water system is normally pressurized through a cross-connect from the service water system (SWS).

A moderate energy leak from a fire protection pipe anywhere in the plant would not result in a noticeable decrease in fire protection system pressure, so the leak would continue until it was. discovered by the plant operator or until the amount of water leaked had caused drain sump. level alarms or equipment damage indications in the control room.

To remedy this situation, the licensee plans to modify the fire protection water system so that a leak would result in a noticeable decrease in fire protection system pressure and activate an alarm.

This modifiCation is currently scheduled to be made in January or February 1980.

This modification will substantially improve the operators ability to detect and stop fire system leaks before essential equipment is flooded.

The control rod drive (CRO) hydraulic system return line, which penetrates primary containment on the 545' elev. of the Reactor building, is currently valved out of service as a result of ongoing studies of CRD return line nozzle cracking.

Pipe breaks associated with the CRD hydraulic control units, on the 517' elev.

of the Reactor Building, could involve 1) the drive insert and withdraw lines which lead through containment penentration to the drives, 2) the CRD hydraulic drive and cooling water lines, 3) the CRD charging water line, and 4) the CRD return line.

A break in the drive withdraw line would cause its associated control rod to insert (scram).

A break in any of the other lines would cause the rod to remain in position but the rod could still be inserted by the 5

. ~

e.

operator or by a reactor protection system scram signal.

Loss of electrical power to the ~RD hydraulic control unit would also result in a rod insertion.

Therefore, potential pipe break damage to the CRD hydraulic control units would not prevent control rod trip (scram) by the operator or the reactor protection system.

The Containment Cooling Service Water (CCSW) system pumps, which are in the Turbine Building north side (495 1 elev.) above the condensate pump room, take suction from the crib house via buried pipe.

No means to isolate this pipe at the crib house end exists; so a MELB in the condensate pump room at the CCSW pump end of the pipe would result in flooding the condensate pump room to the level of the river at the crib house.

Postulated site flood levels in excess of 511 1 would result in the complete loss of the CCSW system.

CONCLUSIONS Based on the information submitted by the licensee and obtained during our site visit to Dresden 2, we have determined that the following review areas have not been addressed in previous staff safety evaluations and should be resolved with the SEP:

1.

Postulated pipe breaks outside of the primary containment between the containment penetration and the first containment isolation valve have not been evaluated for the main s:eam lines, isolation condenser steam and condensate lines, and reactor water cleanup inlet line.

Currently 6

e.

the staff applies the provisions of Branch Technical Position MES 3-1 (Reference 10) section.8.1.b.; to the review of the postulated break areas.

The licensee will be required to compare the design of the Dresden 2 systems with these current regulatory provisions.

2.

The effects of postulated pipe breaks jn certain systems could result in damage to the containment isolation valves or power supply and control cables to the containment isolation valves for those systems.

The combi-nation of the single active failure provision and damage to the containment isolation valve could result in an unisolable break flow path.

The systems of concern are the isolation condenser system (steam line only),

and reactor water cleanup system (inlet line).

The staff currently applies the provisions of Stanch Technical Position ASB 3-1 (Reference.

11), Section B.2.c., to the review of these break areas.

The licensee will be required to compare the design of the Dresden 2 system with these current regulatory provisions.

3.

The consequences of a potential main feed system line break in the main feed regulatory valve area (turbine build 538 1 elev.) have not been determined with regard to possible damage to the engineered safeguards system electrical cables in that area.

The licensee will be required to evaluate the consequences of a possible feed line break in this area and provide adequate protection as re.qui red.

(This area is a 1 so being reviewed under the NRC fire protection evaluation to determine the effects of a fire on the electrical cables in that area.)

7

e.
4.

A main feed system line break in the main feed pump area (turbine building 517 1 elev.) could damage the bus duct connecting diesel generator #2 to the 4KV switchgear in the reactor building.

Current criteria postulate a loss of offsite power (caused by a un~t trip on loss of feedwater) and a potential single active failure of the #2/3 diesel generator.

This scenario would result in loss of all AC power to the unit.

5.

A properly oriented moderate energy line break of the reactor building closed cooling water system could cause the flooding of the redundant 4KV switchgear 23-1 and 24-1 (on the 545 1 elev of the reactor building) to the level of the curb surrounding the switchgear - about 5 11

  • Items 1, 2, 3 and 5 above constitute potential devations from current criteria for postulated pipe breaks outside containment.

Additional information will be obtained from the licensee to determine if 1) the systems identifed in items 1 and 2 conform to the provisions of References 10 and 11, 2) if the postulated flooding identified in item 5 represents a hazard to. switchgear 23-1 and 24-1, and 3) if a feedwater pipe rupture, as described in item 3, could damage electrical cables for system which ~re required to function to mitigate the effects of the rupture or to safely shutdown the plant.

Although item 4 above results in the total los$ of AC power on the site, the likelihood of a unit trip resulting in the'loss of offsite power is extremely small based on the operational history of this plant.

Thus, AC power would be available to operate the safe shutdown systems which were identified in the evaluation section of this report.

Nevertheless, should a loss of offsite power be 8

e.

postulated, the isolation condenser system would be available to remove reactor decay heat and commence a plant cooldown.

In this case, makeup water to the isolation condenser would be provided by the fire protection system.

The twenty minutes of shell side water in the condenser provides adequate time to start a diesel driven fire pump and assure a makeup flow path from the fire system to the isolation condenser.

Decay heat removal using the isolation condenser can be maintained while repairs are made on the diesel generator which was assumed to fail (single active failure) or until offsite power is regained.

The staff is continuing its evaluation of the postulated pipe breaks identified in*items 1, 2, 3 and 5 above and will update the evaluation as conclusions are reached.

9

-~

TABLE 1.

EFFECTS OF PIPE BREAK OUTSIDE CONTAINMENT Zone Crib House Reactor Build.

(613')

Reactor Build.

(589')

Pipe Break SWS (MELB)*

CW (MELB}

Fire system (MELB)

Isolation Condenser Steam and Condensate (llELB)

Affected Mitigating System None None None None

  • See last page of fable 1 for list of abbreviations.

Affected Safe Shutdown System DGCW pumps.

OGWS pumps None None Adequacy of Protection Remarks.

Adequate.

SWS leak (822 gpm) envelopes all MELB's in upper crib house.

SWS function is not lost.

SWS may flood DGCW pumps in lower crib house (see following remarks}.

Adequate.

CW leak envelopes all MELB's in lower crib house.

DGCW pumps (#2, #3, and #2/3) would be flooded, but these are submersible pumps.

CW system function and offsite power would not be lost since SWS pumps in upper crib house would not be flooded.

A control room flood alarm is activated at 3' level in lower crib house.

Adequate.

Floor drains are adequate to remove this leakage (approx. 100 gpm).

Fire system MELB envelopes other MELB's in this zone, e.g.

A demin. water.

Adequate.

Pipe whip of isolation condenser steam or condensate lines may damage both divisions of ESS cable trays.

These cable trays contain control cable for the isola-tion condenser constant vent valves which are located on the 509' level (References 9).

These valves fail closed on loss of electrical signal.

I J.~...

Zone Reactor Build (589') (Cont.)

..... Reactor Build.

(570')

TABLE 1.

(Continued)

Pipe Break Affected Mitigating System fire system and None clean and contaminated Demin. water (MELD)

St.CS (llELB)

None SLCS (HELB)

Isolation condenser steam & condenser lines (HELD)

I 11*1 f'...

None Isolation conden-ser steam isola-tion valve Affected Safe Shutdown System None None None 480V Switch-gear Div. I.

Adequacy of Protection Remarks Adequate.

floor drains and stairs on north side of zone and drains, stairs, and open hatch (20' x 20') on south side of zone provide adequate drainage for MHD's.

Adequate.

A Sl.. CS MELD outside containment would result in the containment isolation check valve inside containment seating with reactor system pressure.

This would isolate the flow path from the reactor recirculation system to the pipe break.

Adequate.

See above remarks.

Potentially inadequate.

An isolation condenser (IC) steam or condensate line break may damage the IC steam line isolation valve

~

and a failure of the IC steam line

~

isolation valve inside containment would result in a LOCA outside containment.

Also, the IC steam or condensate line break may damage the Oiv. I 480V switchgear; however, this is acceptable because the Div. II 480V switchgear would be available.

N Zone Reactor Build.

(570') (cont.)

Reactor Build.

Reactor Build.

(545') (Cont.)

Reactor Build.

(545') *(Cont.)

Pipe Break Fire system (MELB)

RWCU (HELD)

CRD Supply line (HELB)

  • Isolation Condenser Vent (llELB)

RBCCW (MELB)

CRD return line (HELO)

RBCCW or SWS (MELB)

TABLE 1.

(Continued)

Affected M1t1gat1ng System None RWCU 1solat1on valves None None None None None Affected Safe Shutdown System None Div. II Cable Tray valves Div. ll.

Cable Tray Div. 11 Cable Tray 4KV switchgear (both division)

None Reactor protection (RPS) instrument racks 2202-5 or 2202-6.

Adequacy of Protection Remarks Adequate.

Stair, floor drains, and open hatches provide adequate drainage-to prevent equipment flooding.

Potentially inadequate.

Break may A

damage cable tray with control &

WI' power for RWCU valves 1201-2, or the valve itself.

This may pr~vent closure of the valve.

Adequate.

Analyzed in Reference 6.

Adequate.

Only one division of ESS cable trays in affected.

These

_trays were analyzed in Reference 6.

Potentially inadequate.

Spray from RBCCW leak could inpinge on either

  • 4KV switchgear 23-1 or 24-1 and flood both 23-1 and 24-1 to level of 5 11 causing loss of both ESS divisions.

Adequate.

Sufficient distance separates line from Div. II cable trays overhead.

Adequate.

Spray from MELB affects only one of two redundant racks.

,J.. **

Zone Reactor Build.

(517 1 )

Reactor Build.

(476 1 )

Pipe Break CRD hydraulic control units (llELB)

CRD*supply line (UELB)

    • CRD supply line (UELB)

CRO hydraulic control units (UELB)

Fire system (MELB) llPCI steam line (llELB)

TABLE 1.

(Continued)

Affected Mitigating System None None None None None llPCI steam line isolation valve Affected Safe Shutdown System Control rods (RPS)

CRD modules (RPS)

Diesel generator

  1. 2 ESS Div. I or II cab 1 e trays None Torus structure Adequacy of Protection Remarks Adequate.

Ruptures of high energy portions of CRD control units or damage to units resulting from pipe whip would result in either a tripped rod {scram) or loss of CRD

~

supply to the affected control rod.

~

In the latter case, the control rod could still be scrammed manually or automatically by the RPS.

Adequate.

See remarks above.

Adequate.

Pipe whip could damage the diesel generator supply llus to the 4KV switchgear.

No reactor trip, or LOP would occur. lhe ability to trip the reactor, if necessary, would still be retained.

Adequate.

Only one division of ESS cable is affected by a single break. A Analyzed in Reference 6.

~

Adequate.

Floor drains and hatches provide adequate drainage to prevent equipment flooding.

Adequate.

Analyzed in Reference 4 and 6.

HELB restraints have been added to this line.

.J. **.

-~

Zone Reactor Build.

Corner Rooms (476 1

)

llPCI pump room Turbine Build.

(469')

TABLE 1.

(Continued)

Pipe Break Affected Mitigating System Fire system or CCSW lines (MELB) llPC I steam lf ne (llELB)

None Condensate Booster pump discharge (HELB)

Fire systems (MELB)

None None None CCSW suction and None discharge lines (MELB)

Affected Safe Shutdown System LPC I, CS pumps -

llPCI CCSW pumps 2C, 20 (on 495' level) ccsw ccsw

. Adequacy of Protection Remarks Adequate.

Flooding from a MELB in a corner room would affect only the equipment in that room.

Redun-dant equipment is available in the other.corner room.

Adequate.

Other safe shutdown systems can perform the functions of the HPCI system.

HPCI steam line isolation valves would close on high HPCI steam flow or high HPCI steam flow or high temperature in the HPCI room.

Adequate.

Break could affect cable tray (Div. II) for only two of four

  • CCSW pumps.

Flooding from a conden-sate system break would not reach other CCSW pumps (Reference 7).

Adequate.

Flooding level alarms and condensate pit sump level alarms provide sufficient warning time to e allow isolation of leak.

Adequate.

The enveloping flood condition is a non-isolable suction line leak which would take several hours to fill the condensate pump room.

The water level would event-ually reach the l eve 1 of lhe river at the crib house.

A maxinttoo his-torical flood level in the river

,J.. *.

U1 TABLE 1.

(Continued)

Affected Mitigating Affected Safe Adequacy of Zon_e __ ~~~~--~-P_i~p_e_B_r_e_a_k~~~~------__,Sy~s~t_e_m ________ ~---S~h_u_td_o_w_n __

S~y_s~t_em __ ~--P~r_o_t_e_c_t_io_n __ R_e_m_a_r_ks __ ~~~~----~

Turbine Build.

(495 1

)

Turbine Build.

(517'}

Condensate Booster pump discharge (llELB}

CRD water pumps and discharge lines

.. (HELB)

Various MELB systems Main feed (HELB}

None None None Emergency AC power CCSW pumps CCSW pumps CCSW pumps Emergency AC power (508'} would still be below the vent opening at the top of the CCSW pump 28 & 2C watertight enclosure (511'}

Adequate.

A single break could

,a affect Div. I power cables to pumps

  • 2C & 28 or Oiv~ II cables to pun~s 2C & 20.

In either case the remaining pumps could provide the safe shutdown function.

Adequate.

See remarks above for CCSW pump power cables.

Adequate.

See remarks for MELB's on Turbine Build 469' level.

MELB's on 495' level would drain to 469' 1eve1.

Adequate.

A main feed line break in the turbine building could damage the diesel generator #2 bus duct.

Current pipe break criteria would assume loss of offsite power (cause by turbine trip} and single active failure leading to loss of emergency AC power.

Reliability of offsite power at Dresden provides assurance that emergency AC power would not be lost.

.J.. *.

Zone Turbine Build.

(538')

Pipe Break Main feed (HELB)

TABLE 1.

(Continued)

Affected Mitigating System

  • Div. I & II Cable trays Affected Safe Shutdown System Ofv. I & II Cable trays Adequacy of Protection Remarks Potentially inadequate.

A feed line HELD could damage redundant Div. I &

II electrical cables.

The effect of this damage needs to be evaluated.

I *-..

TABl.E 1.. (f.onti mmrt) lf st of Abln*ev I atlons CCSW - Containment Cooling Service Water (also referred to as Emergency Service Water)

CRD - Control Rod Drive CS - Core Spray sy~tem (part of the emergency core cooling systems)

CW - Circulating Water system demin. - demineralized DGCW - Diesel Generator Cooling Water system ESS - Engineered Safety System HELB - High Energy Line Break HPCI - High Pressure Coolant Injection system (part of the emergency core cooling system)

IC - Isolation Condenser LOCA - Loss of Coolant Accident LOP - Loss of Offsite Power LPCI - Low Pressure Coolant Injection system (part of the emergency core cooling system)

MELD - Moderate Energy Line Break RBCCW - Reactor Building Closed Cooling Water System RPS - Reactor Protection System RWCU - Reactor Water Clean-Up system SLCS - Standby Liquid Control System SWS - Service Water System J.J. *.

References

1.

NRC letter, A. Giambusso to CE Co., dated December 18, 1972.

2.

Analysis of Effects of Pipe Break Outside Containment, Dresden Station Unit 2 & 3 Special Report No. 37, dated January 23, 1974.

3.

Supplement 1 to Dresden Station Unit 2.& 3 Special Report No. 37, dated March 22, 1~74.

4.

Revision 1 to Dresden Station Unit 2 & 3 Special Report No. 37, dated February 18, 1975.

5.

NRC letter to CE Co., dated May 12, 1976, transmitting License Amendment No. 16 and 14 to License Nos. DPR-19 and 25.

6.... Appendix E to Dresden Station Unit 2 & 3 Special Report No. 37_. dated November 17, 1977.

7.

CE Co. letter J. Abel to D. Ziemann dated August 23, 1973 concerning Flooding of Critical Equipment at Dresden Units l, 2, and 3.

8.

SEP Review of Safe Shutdown Systems for the Dresden Unit 2 Nuclear Power Plant.

9.

CE Co letter M. Turbak to G. Lear dated June 5, 1978 transmitting Dresden Station Units 2 and 3 Fire Protection Safe Shutdown Analysis.

10.

Branch Technic~l Position MEa 3-1, appended to Standard Review Plan 3.6.2.

11.

Branch Technical Position ASB 3-1, appended to Standard Review Plan 3.6.l.

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