ML18078A299

Forwards Response to Remaining NRC Requests for Addl Info. Info Should Be Incorporated Into Amend to Application & FSAR

ML18078A299
Person / Time
Site:	Salem
Issue date:	10/23/1978
From:	Mittl R Public Service Enterprise Group
To:	Parr O Office of Nuclear Reactor Regulation
References
NUDOCS 7810250069
Download: ML18078A299 (90)

Text

• October 23, 1978 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management Gentlemen:

RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION NO. 2 UNIT SALEM NUCLEAR GENERATING STATION I

• i DOCKET NO. 50-311

• I I

Public Service Electric and Gas Company hereby submits 40 copies of its responses to the remainder of your requests for additiorial inform~tion dated March 27, April 18 1 May 2, May 8,

~ay:..;-18, September 7, September 21, and October 6, 1978. The information contained herein will be incorporated into the Sale~ F$AR in an amendment to our application.

Shbuld you *h~ve any questioris, please do not hesitate to con~

tact us*.

fj;(:@you s,

.~ ..*. R. Lo Mittl

• - General Manager Licensing and Environment Engineering and Construction L

QUESTION 4.31 Pumps ~nd Valves Operability Assurance Program - NSSS and BOP Provide a listing of vital appurtenances attached to safety related pumps a~d valves and their method of seismic qualifi-cation to demonstrate their functional capability under SSE event.

ANSWER Safety-related air-operated valves supplied by PSE&G were either seismically tested or analyzed. Those valves vibra-tion tested were shown to be operable during a seismic event through cycling of the valve. These tests demonstrated operability of the actuator and other vital appurtenances physically attached to the valve which are necessary for its proper operation. Those valves analyzed were demonstrated to have natural frequencies above at least 30 Hz and considered to be rigid components without amplification due to a seismic event. Motors associated with safety related pumps provided by PSE&G have been tested or analyzed for operability under pbstulated seismic conditions as s~ecified iri Table Q7.18-l.

"Summary of Seismic Qualification for Safety Related Equipment."

For pumps and valves in the Westinghouse Scope of Supply, vital appurtenances which are attached to an active pump and motor or an active valve operator were designed and installed using the applicable standards and engineering practice in use at the time of ordering equipment. The system containing the active equipment will undergo cold hydro testing and hot functional SNGS-FSAR Amendment 43 Units 1 & 2 Q4.31 P78 144 70

r4/

QUESTION 4.31

~ ANSWER - Continued testing prior to start up. After startup, the components will be inservice inspected and tested as required at regular speci-fied intervals to assure continued functional availability *

• SNGS-FSAR Units 1 & 2 Q4.31-2 Amendment 43 P78 144 71

Question 5.59 Provide a table listing the structural heat sink~ within the containment. Specify and justify the uncertainty, as percent, to be applied for the maximum and minimum containment pressure response analyses. The table should include the following information; identification of the heat sinks, surface area of each face, material of construction, material thickness, and id~ntif ication of the atmosphere to which each face of the heat sink is exposed.

Answer Uncertainty of the heat sinks is considered as follows:

Containment Liner - 5%

Other Miscellaneous Items - 7 1/2%

Miscellaneous Mechanical Items - 10%

These uncertainty factors are based on an evaluation of un-avoidable variations resulting from material thickness and fabrication tolerances applied by manufacturers and fabri-caters.

The attached table pr6vides a listing of the structural heat sinks.

SNGS-FSAR Q5.59 Amendment 43 Units l & 2 P78 139 48

TABLE Q5.59-l STRUCTURAL HEAT SINKS Material Sides Surface Description Type Thickness Exposed (ft 2)

Containment Steel A-36 3/8" 1 47,546 Cylinder Steel A-36 3/8" 1 14,954 (Covered by insulation)

Reinforced 54 II* 1. 47,546 Concrete Reinforced 54" 1 14,954 Concrete (Section under steel covered with insulation)

Insulation 2.5" 1 14,909 Paint 7.5 mils 1 47,546 Containment Steel A-36 1/2" 1 30,788 Dorne Reinforced 42" 1 30,788 Concrete Paint 7.5 mils 1 30,788 Containment Reinforced 42" 1 I2, 222 Floors Concrete Paint 18 mils 1 12,222 Reactor Reinforced 12" 1 2,514 Cavity Concrete Paint 18 mils 1 2,514 SNGS-FSAR TABLE Q5.59-l Amendment 43 Units 1 & 2 Sheet 1 P78 140 33

STRUCTURAL HEAT SINKS Material Sides Surf ace Description Type Thickness Exposed (ft 2)

Refueling Stainless 1/4 1 7,942 Canal Steel 304 Concrete 51° 1 7,290 Crane Wall Reinforced 36° 2 13,707 Concrete Paint 14 mils 1 26,414 Operating Deck Reinforced 2 4,960

@EL. 130'-0 Concrete Paint 18 mils 1 9,920 Shield Walls Reinforced 36° 2 2,956 Above El. 130'-0 Concrete Paint 14 mils 1 5,912 Mainsteam/F.W. Steel A-441 1/4" 2 554

• Stops

• 3/8" 2 813 9 1/2 1/2° 2 5,532 5/8° 2 639 3/4° 2 7/8° 2 461 l" 2 1,515 1 1/4° 2 1,006 Paint 7.5 mils 1 26,928 SNGS-FSAR TABLE Q5.59-2 Amendment 43 Units 1 & 2 Sheet 2 P78 140 34

STRUCTURAL HEAT SINKS Material Sides Surface Description Type Thickness Exposed (ft 2)

Polar Gantry Steel A-36 1/8" 2 76 Bridge Crane 3/8 II 1 4,953 3/8" 2 589 1/2" 2 41 1/2"' 1 156 5/8" 1 374 5/8" 2 360 3/4" 2 41 3/4" 1 4,911 l" 1 348 l" 2 100 1 1/8" 1 358 1 1/2" 2 31 1 5/8" 1 6. 3 '

2" 2 13.5 2 1/4" 2 26 2 3/4" 1 1,553 3" 1 263.7 3 1/4" 1 3 3 1/4 II 2 1.8 SNGS""."'FSAR TABLE QS.59-3 Amendment 43 Units 1 & 2 Sheet 3 P78 140 35

STRUCTURAL HEAT SINKS Material Sides Surface Description T:t:pe Thickness Exposed (ft 2)

Polar Gantry A36 Steel 3 3/4" 1 14.6 Bridge Crane 4" 2 1.5 5 1/2° 1 27.1 6" 2 1 8 n- 1 10 .,2 1 1/2° 2 72.2 3/4° 2 33 Paint 7.5 mils 1 15,751.9 Manipulator Steel A-36 3/4 11 2 75 Crane 1/4 n 1 83 1/4" 2 10 3/8" 1 187 3/8° 2 390 3/4" 1 146 l" 2 100 Paint 7.5 mils 1 1,366 Orbital Service Steel A-36 1/8" 2 248 Bridge & Flat Form 1/4 11 2 2,408 SNGS-FSAR TABLE Q5.59-4 Amendment 43 units 1 & 2 Sheet 4 P78 140 36

STRUCTURAL HEAT SINKS Material Sides Surface DescriEtion Tl:pe Thickness Exposed (ft 2)

Orbital Service Steel A-36 3/8" 2 2,533 Bridge & Flat Form 1/2" 2 5,375 5/8° 2 2,369 3/4°' 2 227 ln 2 13 1 1/2° 2 25 3 1/6" 2 5,150 Paint 7.5 mils 1 36,496 Annular El-100' Cone. 18" 2 435 Cone. 6" 2 1,600 Paint 18 mils 1 4,070 Annular El.130' Cone. 10 1/2° 2 2,000 Paint 18 mils 1 4,000 Annulus Steel 3/16" 2 375 A36 1/4" 2 ~-

705.5 3/8" 2 1,261.5 1/2" 2 2,150 5/8 II 2 283.5 SNGS-FSAR TABLE Q5. 59-5 Amendment 43 Units 1 & 2 Sheet 5 P78 140 37

STRUCTURAL HEAT SINKS Material Sides Surf ace Description Type Thickness Exposed (ft 2)

Annulus 2 104 Fl. Framing El. 130 1 3/8n 2 129 Paint 7.5 mils 1 10,017 Annulus Steel A-36 2 600 Fl. Framing El. 100-0 l/4n 2 447.5 2 138.4 1/2" 2 1,829.5 5/8n 2 150 3/4" 2 123.5 l" 2 110.5 Paint 7.5 mils 1 9,290 Annulus Steel A-36 3/16 11 2 300 Columns Stair RC-1 & 2 l/4n 2 2,758 Equip. Hatch Cover Storage 3/8n* 2 1,167.5 Rack l/2n 2 129 5/8" 2 2,360 7/8n 2 412.5 Paint 7.5 mils -1 14,254

• SNGS-FSAR Units 1 & 2 TABLE QS.59-6 Sheet 6 Amendment 43 P78 140 38

STRUCTURAL HEAT SINKS Material Sid-es Surface Description Type Thickness Exposed (ft 2)

Annulus Steel A-36 1/4" 2 442.5 Feedwater Limit Stop EL-107-6 3/8" 2 389 Stir RC-3 Equip. Support 1/2" 2 219.5 3/4" 2 183.5 5/8" 2 35.5 1" 2 105.0 Paint 7.5 mils 1 2,750 Pressure Steel Al36 3/16" 2 84 Supports 1/4" 2 38 3/8 11 2 7.5 1/2 11 2 371. 5 5/8" 2 288 3/4" 2 156.5 1" 2 32.5 1 1/2" 2 16.0 Paint 7.5 mils 1 1,988 Steam Gen. Steel A-36 3/16" 2 70 Supports 1/4 11 2 15.8 3/8" 2 54.3 SNGS-FSAR TABLE Q5.59-7 Amendment 43 Units 1 & 2 Sheet 7 P78 140 39

STRUCTURAL HEAT SINKS Material Sides Surface Description TyEe Thickness ExEosed (ft 2)

Steam Gen. 1/2" 2 95.75 Manway Cover Platforms 5/8" 2 337.5 3/4 8 2 180 7/8" 2 434 ln 2 18 ,.037 1 1/4° 2 1,124.7 1 3/8° 2 258.5 1 1/2 11 2 752 1 7/8° 2 282 2 II 2 714. 5 2 1/4 II 2 1,296 2 3/4 11 2 625 3 II* 2 142.5 4n* 2 195.5 Paint 7.5 mils 1 16,764 Reactor Coolant Steel A-36 3/16" 2 960 Pump Supports 1/4" 2 507 3/8 11 2 630

• SNGS-FSAR Units 1 & 2 TABLE Q5.59-8 Sheet 8 Amendment 43 P78 140 40

STRUCTURAL HEAT SINKS Material Sides Surface :e iption Type Thickness Exposed (ft 2)

Pump 5/8° 2 284 s Plate l" 2 336.5 l 1/4° 2 131.5 1 3/8° 2 1,013 1 1/2° 2 152.5 l 5/8" 2 18 1 7/8" 2 756 2 II 2 57.5 2 1/2" 2 51.5  ;* 6 3 II 2 253~5 i. 5 4 II 2 34 6 II 2 28.5 Paint 7.5 mils 1 10,427 ~. 5 Frames Steel A-36 3/16 11 2 308.5 30'-0 ded 1/4" 2 38 1.1 El. 130'-0, Head 1/4" 1 56 wn El. 130'-0,

.* 5 3/8° l 50 3/8 11 2 21 FSAR l & 2 TABLE Q5.59-9 Sheet 9 Amendment 43 P78 140 41 43

STRUCTURAL HEAT SINKS Material Sides surface Description T:lEe Thickness Exposed (ft 2)

Trenches at Steel A-304 1/4" 2 76.5 EL. 81'-0 Dwg. 208915A 1/4" 1 279 8823 Sump Pit 1/2" 2 9 Liner Pls *-

and Screens Reactor Cavity Steel 1/4 n 1 445 Reactor Support Cooling Duct Paint 7.5 mils 1 445 Refueling Canal Stainless 3/16" 1 1,072 Cone. Forms Steel Type 3 &4 Reactor Cavity Steel A-36 3/16" 2 257 Reactor Pit 1/4" 2 352.5 Platforms Paint 7.5 mils 1 1,219 eactor Cavity isc. Steel

.. Steel A-36 - 3/16" 2 68. 3 ncore Seal 1/4" 2 41 Table Area 1/4" 1 18 3/8R 1 13 3/8" 2 11 1/2" 2 13.5 Paint 7.5 mils 1 299

• SNGS-FSAR Units 1 & 2 TABLE Q5.59-ll Sheet 11 Amendment 43 P78 140 43

STRUCTURAL HEAT SINKS Material Sides Surface Description Type Thickness Exposed (ft 2)

Missile Shield Steel A-36 1/4° 2 122 Control Rod Drive Mechanism 3/8° 2 7.7 EL. 130'-0 2 22.5 1 434 1 3.3 1 289 Paint 7.5 mils 1 1,169 Main Steam and Steel A-36 1/4 11 2 34.5 Feedwater Limit Stops EL. 140'-9 1/2 1/2° 2 180 5/8 11 2 41 7 /8 II 2 16

• Paint 1 II 1 1/8° 7.5 mils 2

2 1

64 34.5 370 Feedwater Steel A-36 3/8 ° 2 67 Limit Stops El. 96'-0 1/2° 2 97.5 Annulus Area 5/8 11 2 71 3/4° 2 122.5 l" 2 80 SNGS-FSAR TABLE Q5.59-13 Amendment 43 Units 1 & 2 Sheet 13 P78 140 45

STRUCTURAL HEAT SINKS Material Sides Surface Description Type Thickness Exposed (ft 2) 1 1/8" 2 13 1 1/2" 1 9 1 7 /8" 1 119 l 7/8" 2 86.5 Paint 7.5 mils 1 1,203 Annulus Area Steel A-36 3/16" 2 10.4 Steel Supports Below EL 100'-0 1/4" 2 407.5 3/8" 2 4,929.0 1/2" 2 388.4 Paint 7.5 mils 1 11,471 P~essurizer Shield Steel A-36 3/16" 2 1,476.0

& Plate Above EL. 130'-0 1/4" 2 312.1 3/8" 2 397.0 3/8" 1 190.0 l" 1 420.0 f" 2 1,475.5 Paint 7.5 mils 1 7,931

• SNGS-FSAR Units 1 & 2 TABLE QS.59-14 Sheet 14 Amendment 43 P78 140 46

STRUCTURAL HEAT SINKS Material Sides Surface Description Type Thickness Exposed (ft 2)

Ventilation and Steel 0.0625" 1 48,600 Misc. Ducting Reactor Head Lifting Steel 0.86" 1 1,178 Gear Av

• Containment Spray Stainless 0.148" 1 3,547 Piping Steel Miscellaneous Stainless 1.125" 1 466 Piping Uninsulated Steel Accumulators Steel a) Heads 1.41 11 1 1,760 b) Shell 2.75" 1 1,164 c) Skirt .375" 2 2,099 Cable Tray and Steel 0.25 2 35,000 Hangers II Conduit Steel 0.104" Avg. 1 15,000

• Junction Boxes Steel 0.109" 1 1,500 Service Water Steel 0.365" 1 5,712 Piping with 1-1/2" Insulation .

Service Water Piping Steel 0.216" 1 840 with l" Insulation RHR and SI Piping Stainless 1.125" 1 200 with 3-1/2" Steel Insulation RHR and SI Piping Stainless 1.125" 1 534 with 1-1/2" Insulation

• SNGS-FSAR TABLE QS.59-16 Amendment 43 Units 1 & 2 Sheet 16 P78 140 51

STRUCTURAL HEAT SINKS Material Sides Surface

• Description Type Thickness Exposed (ft 2)

Miscellaneous Steel 0.145" 1 7,455 Small Bore Piping Bare Piping Controls Trays, Steel 0.202° 2 5,950 Panels and Tubing Avg.

Insert Steel Steel A-36 5/6° 1 2,928 Hanger Steel Steel A-36 1/4° 2 710 1/4° 1 106.8 3/8° 2 2,476 3/8° l 106.8 Pipe Restraints Steel -A-36 1/4 II 2 288

& Steel Hangers (Large Pipes) 3/8" l 108

• 3/8° 1/2" 1/2° 2

1 2

446.5 27.8 35.2 3/4" 1 1,154.8 3/4" 2 84.5 7/8" 2 137 l" 1 106 1 II 2 226 1 1/4 n 1 25

• SNGS-FSAR TABLE QS.59-17 Amendment 43 Units 1 & 2 Sheet 17 P78 14 0 52

STRUCTURAL HEAT SINKS Material Sides Surface DescriEtion T;i;ee Thickness ExEosed (ft 2)

Continued Pipe Steel A-36 l 1/4" 2 14 Restraints Steel

& Hangers (Large l 3/8° *2 122.75 Pipes) 1 1/2° 1 73.9 1 1/2° 2 92.5 1 3/4" 2 69.88 20 1 166 2" 2 14.1 30 1 11.6 4 & 4 1/2" 2 111 Springs & Steel 3/16" 1 703.9 Spring Box for Hangers

-s1ee1 * - -

RWS/mlb P78 140 33/47 51/53

• SNGS-FSAR Units 1 & 2 TABLE Q5. 59-18 Sheet 18 Amendment 43 P78 140 53

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I QUESTION 5.62

• Describe and justify the analytical model used to conservatively determine the maximum containment temperature and pressure for a spectrum of postulated main steam line breaks for various reactor power levels.

a.

Include the following in the discussion:

Provide single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and con-tainment pressure/temperature response. The single failure analysis should include, but not necessarily be limited to:

main steam and connected systems isolation; feedwater, auxiliary feedwater, and connected systems isolation; feed-water, condensate, and auxiliary feedwater pump trip and run-out control system; the loss of or availability of offsite power; diesel failure when loss of offsite power is evaluated; and partial loss of containment cooling systems.

b. Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.

c. Discuss and justify the heat transfer correlation(s) e.g.,

Tagami, Uchida) used to calculate the heat transfer from the containment atmosphere to the passive heat sinks, and

• d.

provide a plot of the heat transfer coefficient versus time for the most severe steam line break accident analyzed.

Specify and justify the temperature used in the calculation of condensing heat transfer to the passive heat sinks; i.e.,

specify whether the saturation temperature corresponding to the partial pressure of the vapor, or the atmosphere temperature which may be superheated was used.

e. Discuss and justify the analytical model including the thermodynamic equations used to account for the removal of the condensed mass from the containment atmosphere due to condensing heat transfer to the passive heat sinks;

f. Provide a table of the peak values of containment atmosphere temperature and pressure for the spectrum of break areas and power levels analyzed;

g. For the case which results in the maximum containment atmosphere temperature, graphically show the containment atmosphere temperature, the containment liner temperature, and the containment concrete temperature as a function of time;

• - SNGS-FSAR Units 1&2 Q-5.62 Amendment 43 P78 139 54

• h.

i.

For the case which results in the maximum containment atmosphere pressure, graphically show the containment pressure as a function of time; and For the cases which result in the maximum containment atmosphere pressure and temperature, provide the mass and energy release data in tabular form.

j. Specify and justify the design temperature of the con-tainment structure liner and concrete, the design temperature of the internal concrete structures, and the temperature used to qualify the safety related instrumen-tation located within the containment.

k. Discuss the design provisions in the main feedwater control and power supply systems which assure that a single active failure cannot prevent the termination of main feedwater flow to the affected steam generator. Specify the design criteria for the equipment relied on to limit the mass and energy release to the containment. Justify relying on equipment for accident mitigation that are not engineered safety feature grade; i.e., designed to Safety Class 2 and Seismic Category I criteria *

• ANSWER Steambreak analyses using three different models will be per-formed for Salem. The first will utilize the Westinghouse containment models developed for the 1971 Equipment Qualif i-cation Program. These models and their justification are de-tailed in References 1 to 5. Westinghouse Public Service and the NRC discussed these models at the October 4 meeting in Bethesda.

The mass and energy blowdown used in this analysis will dry steam releases covering a spectrum of power levels, break sizes and single failures.

SNGS-FSAR Amendment 43 Units 1&2 Q-5.62 P78 139 55

• -- ---- ---~-~-

• These calculations will determine will determine the limiting large and small break cases. If necessary, thermal analyses will be performed utilizing the Westinghouse equipment models.

These results will fall under a peak temperature transient as presented in Reference 5 (approximately 385°)~

The second analysis performed will show a *comparison between the NRC proposed containment model and the Westinghouse small break containment model. The case studied will be the limiting small break case identified by the previous analysis. The cal-culated containment transients should be similar to those cal-culated for the small break with the Westinghouse model.

The third analysis performed will illustrate the effect of more realistic mass and energy releases for the large MSLB. The mass and energy release model used will include liquid carry-over (entrainment) from the steam generators as described in WCAP 8822. Salem specific param.eters will be incorporated in this model. These results will illustrate the insensitivity of the maximum temperature to containment assumption for the large break.

a. A list of the single failures to be considered in the Salem analyses is provided in response 5.82. All blowdown analyses with MARVEL will be done assuming the reactor coolant pumps SNGS-FSAR Amendment 43 Units 1&2 Q-5.62 P78 139 56

running since this results in more severe mass/energy releases as described in Section 3.1.7 of WCAP-8822. How-ever, all containment analysis will assume loss of offsite power and will consider loss of a diesel to evaluate the failure of a continment safeguards train.

b. The times assumed in the analysis for initiation of the containment sprays and fan coolers following the appro-priate containment pressure setpoints are 59 seconds and 35 seconds, respectively. These times are based on the loss of offsite power with delays consistent with the Technical Specification limits. The delay in containment spray de-livery includes the time required to obtain full speed of the containment spray pumps and fill the appropriate headers and piping.

c. The containment analysis utilizing the Westinghouse COCO computer code uses the Tagami condensing heat transfer coefficients as described in WCAP-8327 (Reference 7).

Justification of the use of these heat transfer coefficients has been provided in References 1, 4, and 6. A plot of the heat transfer coefficient versus time will be provided for the most severe steam line break accident identified by the analysis.

SNGS-FSAR Amendment 43 Units 1 & 2 Q-5.62 P78 139 57

d. The saturation temperature corresponding to the partial pressure of the vapor in the containment is conservatively assumed for the temperature in the calculation of condens-ing heat transfer to the passive heat sinks. This tempera-ture is also conservatively assummed for the calculation of heat removal by the containment fan coolers.

e. As described in the introduction of this response, several analytical approaches wil be used for the calculation to account for the removal of condensate from the containment atmosphere. The thermodynamic equations used for the studies performed with the Westinghouse containment model are described in References 5 and 7. Justification of this model is provided in References 1, 4, S, 6, and 7. As noted, appropriate sensitivity studies will be provided with the plant analysis.

f. This information will be provided in the final response upon completion of the analysis.

g. This information will be provided in the final response upon completion of the analysis.

h. This information will be provided in the final response upon completion of the analysis.

SNGS-FSAR Amendment 43 Units 1 & 2 Q-5.62 P78 139 58

~---~---------*

• i.

j.

Tables of mass/energy releases will be provided for the worst pressure and temperature cases.

Upon completion of the containment analysis, an evaluation of the safety related instrumentation will be performed to show conformance with the requirements of IEEE-323-1971.

k. The only non-safety grade equipment in the main feed system which is relied upon to terminate the main feed flow to the steam generators are the main feedwater control valves.

These valves are not seismic category I. However, each valve receives dual, independent, safety grade trip-closed signals from the protection system following a steam line break. Also, the valves are air-operated fail-closed design. Since the assumed break is inside containment in a seismic category I pipe, it is not assumed to be initiated by a seismic event. Therefore, to assume a coincident seismic event with the hypothetical pipe rupture is not required, and thus a seismic classification for the main feed regulation valve is not necessary to insure closure following a steamline break inside containment.

Because of the conservative nature of the transient calcula-tions used for the 1971 Equipment Qualification Program, the results of the Salem temperature transient calculation will

• SNGS-FSAR Units l & 2 Q-5.62 Amendment 43 P78 139 59

-*-*-*~*--**- -- .

,I I

/

l fall under the peak transient calculated for the 1971 Equipment Qualification Program and presented in reference 5 (approximately 38S°F). The pressure transient will fall below the design limits for the Salem 2 containment.

5.0 REFERENCES

1. Letter to Mr. D. B. Vassallo, Chief, Light Water Reactors Project Branch 6, USNRC from Mr. c. Eicheldinger, Manager, Nuclear Safety Westinghouse Electric Corporation, Dated March 17, 1976 (NS-CE-992).

2. Letter to Mr. D. B. Vassallo, Chief, Light Water Reactors Project Branch 6, USNRC from Mr. c. Eicheldinger, Manager, Nuclear Safety Westinghouse Electric Corporation, Dated July 10, 1975 (NS-CE-692).

3. Letter to Mr. D. B. Vassallo, Chief, Light Water Reactors Project Branch 6, USNRC from Mr. c. Eicheldinger, Manager, Nuclear Safety Westinghouse Electric Corporation, Dated April 7, 1976 (NS-CE-1021).

4. Letter to Mr. J. F. Stolz, Chief Light Water Reactors Project 6, USNRC from Mr. C. Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, Dated August 27,.

1976 (NS-CE-1183).

5. T. Hsieh et al, WCAP-8936, "Environmental Qualification Instrument Transmitter Temperature Transient Analysis,"

February 1977.

6. Letter to John F. Stolz, Chief Light Water Reactor's Project, Branch 6, USNRC from c. Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, Dated June 14, 1977 (NS-CE-1453).

7. Bordelon, F. M., Murphy, E.T., WCAP-8327, Containment Pressure Analysis Code (COCO), July 1974.

SNGS-FSAR Amendment 43 Units l & 2 Q-5.62 P78 139 60

QUESTION 5.63

• Provide the following information regarding the environmental qualification of safety related equipment located inside the containment.

1. Provide a comprehensive list of equipment required to be operational in the event of a main steam line break (MSLB) accident. The list should include, but not necessarily be limited to, the following safety related equipment:

a. Electrical containment penetrations

b. Containment recirculation pump motor

c. Pressure transmitters

d. Containment isolation valves

e. Electrical power cables

f. Electrical instrumentation cable

g. Level transmitters Describe the qualification testing that was done, including the test environment, namely, the temperature, pressure, moisture content, and chemical spray, as a function of time.

2. If a thermal analysis of safety related equipment which may be exposed to the containment atmosphere. following a main steam line break accident is presented, it is our position that the analysis should be based on the following:

a. A condensing heat transfer coefficient based on the recommendations in Branch Technical Position CSB 6-1, Minimum Containment Pressure Model for PWR ECCS Per-formance Evaluation, should be used. To lessen the sensitivity of the calculated condensing heat transfer coefficient on the time selected to calculate the maximum Tagami condensing heat transfer coefficient, we recommend that the greater of four times either the Tagami or Uchida condensing coefficients be used.

b. A convective heat transfer coefficient should be used when the condensing heat flux is calculated to be less than the convective heat flux. During the blowdown period it is appropriate to use a conservatively evaluated forced convection heat transfer correlation.

For example:

SNGS-FSAR Amendment 43 Units 1&2 Q-5.63 P78 139 27

Nu = C {Re) n where Nu = Nusselt No.

Re = Reynolds No.

C,h = emperical constants dependent on geometry and Reynolds No.

Since Reynolds number is dependent on velocity, it is necessary to evaluate the forced flow currents which will be generated by the steam generator blowdown. The CVTR experiments provide limited data in this regard.

Convective currents of from 10 ft/sec to 30 ft/sec were measured locally. We recommend that the CVTR test_

results be extrapolated conservatively to obtain forced flow currents to determine the convective heat transfer coefficient during the blowdown period. After the blow-down has ceased or been reduced to a negligibly low value, a natural convection heat transfer correlation is acceptable.

3. For each component where thermal analysis is done in con-j unction with an environmental test at a temperature lower than the peak calculated temperature following a main steam line break accident, compare the test thermal response of the component with the accident thermal analysis of the component. Provide the basis by which the component ther-mal response was developed from the environmental quali-fication test program. For instance, graphically show the thermocouple data and discuss the thermocouple locations, method of attachment, and performance characteristics, or provide a detailed discussion of the analytical model used to evaluate the component th~rrnal response during the test.

This evaluation should be performed for the potential points of failure such as thin cross-sections and temperature sensitive parts where thermal stressing, temperature-related degradation, steam or chemical interaction at elevated temperatures, or other thermal effects could result in the failure of the component mechanically or electrically. If the component thermal response comparison result in the prediction of a more severe thermal transient for the accident condition than for the qualification test, pro-vide justification that the affected component will perform its intended function during a MSLB accid~nt, or provide protection for that component which would appropriately limit the thermal effects.

SNGS-FSAR Amendment 43 Units 1&2 Q-5.63 P78 139 28

ANSWER l - Refer to question 7.30 2,3 As noted in the response to Question 5.62, an evaluation will be performed based on a comparison of the containment equipment test conditions versus the containment accident environments. If a thermal analysis is necessary, Westing-house will use a thermal analysis model similar to the model presented in Reference 5 of the response to Question 5.62. Any differences between the Westinghouse thermal analysis model and the proposed NRC interim model {as de-tailed in this question) will be discussed and justified.

a. The condensing heat transfer coefficient used for the thermal analysis wil be the same as used in the approved Westinghouse model for ECCS analysis. This model is documented in Appendix A of WCAP 8339 and has been re-viewed and approved by the NRC Staff, as comparable to the model recommended in Branch Technical Position CSB 6-1.

b. A covective heat transfer coefficient comparable to that recommended by the NRC will be used. If neces-sary, sensitivity studies will be performed to justify any model differences *.

SNGS-FSAR Amendment 43 Units 1&2 Q-5.63 P78 139 29

• t

c. Justification for each safety related component's test conditions will be provided utilizing the NRC recommended procedure given in Question 5.63 Part 3 *

• SNGS-FSAR Units 1&2 Q-5.63 Amendment 43 P78 139 30

.. QUESTION 5.64

• /

The information regarding the subcompartment analyses is in-complete. Therefore, provide the following information:

a. Provide and justify the pipe break type, area, and location for each analysis. Specify whether the pipe break was postulated for the evaluation of the compartment structural design, component supports design or both.

b. For each compartment provide a table of blowdown mass flow rate and energy release rate as a function of time for the break which results in the maximum structural load, and for the break which was used for the component supports evalu-ation.

c. Describe the nodalization sensitivity study performed to determine the minimum number of volume nodes required to conservatively predict the maximum pressure load acting on the compartment structure. The nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variation circumferentially, axially and radially within the compartment. Describe and justify the nodalization sensitivity study performed for the major component supports evaluation, where transient forces and moments acting on the components are of concern.

d. Provide a schematic drawing showing the compartment nodali-zation for the determination of maximim structural loads, and for the component supports evaluation. Provide suf-ficiently detailed plan and section drawings for several views, including principal dimensions, showing the arrange-ment of the compartment structure, major components, piping, and other major obstructions and vent areas to permit verifi-cation of the subcompartment nodalization and vent locations.

e. Provide a tabulation of the nodal net-free volumes and interconnecting flow path areas. For each flow path provide an L/A (ft-1) ratio, where L is the average distance the fluid flows in that flow path and A is the effective cross sectional area. Provide and justify values of vent loss coefficients and/or friction factors used to calculate flow between nodal volumes. When a loss coefficient con-

~ists of more than one component, identify each component,

_*its value and the flow area at which the loss coefficient applies.

SNGS-FSAR Amendment 43 Units l&2 os. 64-l P78 139 30

ANSWER An analysis will be performed utilizing the TMD computer code with an 18 node containment model in order to evaluate the pressure differentials across containment structures, steam generators and the pressurizer to RCS breaks. The nodel boundaries, which will be used are defined as illustrated in Figures QS.64-1 and QS.64-2. The latest version of TMD will be used, along with DEHL & DECL mass and energy releases. These releases represent a significant conservatism since the largest possible break will be less than a double-end oven due to existing pipe restimate.

Analyses have been performed in plants with similar containment geometry and reactor coolant system operating conditions. For these cases, the maximum calculated pressure differentials across the major equipment have been approximately 6 psi *

.The steamline break will be addressed either by comparison with a plant which has more restrictive geometry or by a plant specific analysis.

SNGS-FSAR Amendment 43 Units 1&2 os. 64-2 P78 139 31

An evaluation of the interior concrete structure due to the differential pressure effects caused by a loss of coolant acci-dent is in progress. As a first step in this analysis, the differential pressure between the compartment walls and floor slabs of the interior concrete structure are calculated. This calculation has been completed. The results indicated a maximum delta-P across any wall or floor slab of 14 psi. This value does not include any design margin.

The thickness of the concrete walls and floor slabs are often established to minimize the ~xposure to radiation. Because of this, the walls and slabs are normally two to three feet thick.

Salem, however, is constructed of reinforced concrete walls showing a typical thickness of three feet and floor slabs that range from three to five feet thick.

The layout of different PWR interior concrete structures do vary considerably depending upon the number of steam generator loops and also which A/E provided the layout for a particular plant. Keeping this in mind, Table QS.64-1 provides values of differential pressure calculated between the steam generator compartment and adjacent compartments for interior concrete structures of several plants, either in operation or under SNGS-FSAR Amendment 43 Units l.&2 .OS.64-3 P78 139 38

construction. In addition, results from the Westinghouse Reference Plant has also been included. The differential pressures listed in the table includes design margins of 20 - 40%. As it can be seen, the minimum delta-P is 24.5 psi and the maximum is 41.2 psi. In all cases, these values are greater than the differential pressure calculated for Salem.

Even if a 40% margin is applied to Salem, resulting in a delta-P of 19 psi, the Salem delta-Pis still less than those plants listed. It should be noted that the 40% margin is normally included during the early stages of evaluation because of the uncertainties of the final design configuration. In the case of Salem, the final design and configuration has been established and a 40% margin is not appropriate.

Considering the reasonably low value of differential pressure calculated for Salem along with the unusually thick interior concrete walls and floor slabs, we do not anticipate any problem in qualifying Salem for the compartment differential pressures loading.

However, to substantiate that the structure will be adequate, a detailed evaluation of the structure is in progress. This evaluation uses finite element representation of the interior concrete structure. Figures QS.64-3 through QS.64-7 show typical sections of the finite element model. These sections SNGS-FSAR Amendment 43 Units 1&2 QS.64-4 P78 139 39

are components of a complete model. The analysis includes evaluation of the concrete capacity when subjected to a loss of coolant accident combined with other normal operating and abnormal loads.

Detailed calcultions on supports are in progress. Based on the original stress calculations for the component supports, it is expected that all streses will be acceptable.

SNGS-FSAR Amendment 43 Units 1&2 05.64-5 P78 139 55

.. TABLE QS.64-1

• STEAM GENERATOR LOOP COMPARTMENT DESIGN PRESSURE Psg*

(psid) REFERENCE

= Psg 24.S Wolf Creek (SNUPPS) PSAR Table 6.2-1, Rev. 9, 8/75 -

41.2 Virgil C. Summer FSAR Tables 6.2-lla, lSa 35.0 Josephy M. Farley FSAR Section 6.2.1.3.9 29.1 South Texas Project FSAR Table 6.2.1.1-2 26.6 Alvin W. Vogtle PSAR Figures 6.2-BE, F,G

25. 0 Kewaunee FSAR Section 5.9-2 30.0 ~ 3 Loop Reference Plant Table 2-3 19** Salem - Current Analysis

• All include margin 1.2 - 1.4

• • 1.4 margin - However, for evaluation of existing designs the 40% margin is not applicable *

• SNGS-FSAR Units 1&2 Table QS.64-l Amendment 4 3 P78 139 56

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Question 5.65 Section 50.34 of 10 CFR part 50 requires that an analysis and evaluation of ECCS cooling performance following postulated loss-of-coolant accidents be performed in accordance with the requirements of Section 50.46. Appendix K, 0 ECCS Evaluation Models," to 10 CFR Part 50 sets forth certain required and acceptable features of evaluation models. Appendix K states in part, that the containment pressure used for evaluating cooling shall not exceed a pressure calculated conservatively for this purpose. It further requires that the calculation include the effects of opeation of all installed pressure reducing systems and processes. Branch Technical Position CSB 6-1, "Minimum Containment Pressure Model for PWR ECCS Performance Evaluation,n provides additional guidance for the performance of a minimum containment pressure analysis and should be used when the analysis is performed. Therefore, state the minimum containment pressure that has been used in the analysis of the emergency core cooling system. Justify this value to be conservatively low by describing the con-servatism in the assumptions of initial containment condi-tions, modeling of the containment heat sinks, heat transfer coefficients to the heat sinks, heat sink surface area and any other parameter assumed in the analysis. Provide the con-tainment pressure, temperature and sump temperature response for the most conservative assumptions. Your November 2, 1977 submittal on this matter was incomplete.

Answer The containment backpressure used in the 10CFRS0.46 related ECCS analysis will be provided in conjunction with the response to Question 14.30. The analysis will be done for the limiting break as required by 10CFRS0.46.

The containment backpressure will be calculated using the methods and assumptions described in nwestinghouse Emergency Core Cooling System Evaluation Model - Summary "WCAP-8339, Appendix A. Containment data and conditions used to calculate ECCS backpressure will be conservative with respect to minimizing containment pressure *

• SNGS-FSAR Units 1&2 Q-5.65 Amendment 43 P78 139 31

It will be conservatively assummed that both spray pumps and all safeguards fan coolers are operating. It will be also con-servatively assumed that the spray pumps and operating at their runout flow rated. A conservatively high safeguards fan cooler heat removal rate will be calculated by using minimum expected service water temperature.

The structural heat sinks used in the analysis, including the materials of ~onstruction, material thickness, atmosphere to which each face of the heat sink is exposed, and the surface area of each face will be provided in the response to question 5.59. Concrete walls are modeled to prevent overestimation of the energy absorbed by the concrete.

The condensing heat transfer coefficient used for heat transfer the steel containment structures for the limiting break will be provided in the response to 14.30. A conservative model for determining condensing heat transfer will be used in which, at the end of blowdown, a maximum condensing heat transfer co-

• efficient five times higher than that calculated using the Tagami correlation will be assumed. Prior to the end of blow-down, a parabolic increase from the stagnant heat transfer co-efficient to the maximum value will be assumed. During the long-term stagnation phase of the accident, characterized by low coefficients equal to that obtained from the Tagami data will be assumed.

SNGS-FSAR Amendment 43 Units 1&2 Q-5.65 P78 139 32

The containment temperature response will be prevented with the answer to question 14.30.

ing break.

This response will be for the limit-Containment sump temperature will not be used in the analysis because the maximum peak clad temperature occurs prior to the initiation of the recirculation mode for the containment spray system.

The mass and energy releases which will be used in the contain-ment backpressure calculation for the limiting break will be given the response to question 14.30 *

• SNGS-FSAR Units 1&2 Q-5.65 Amendment 43 P78 139 33

QUESTION 5.82 Provide the results of analyses for a spectrum of main steam line breaks within the containment. You should vary break size and power level to identify the break sizes producing the highest containment temperature and pressure. Provide mass and energy release data for these two break sizes. Provide a com-plete description for the method used to calculate flow from the steam generator into the containment. Include justification for all. equations and assumptions. Provide a comparison of the method used in the above analyses with the method described in WCAP-8843 (MARVEL). The analyses should include the resul.ts of postula.ted single failures in the steam and feedwater system.

ANSWER At each of four power levels (0, 30, 70 and 102% of nominal),

on evaluation of a full double-ended pipe rupture and a limited size split pipe rupture will be made. The size of the split ruptures is established by the steam line break protection equipment, and is the largest rupture which will not result in a steam line isolation signal due to low steam line pressure.

Determination of this break size is.discussed in Section 2.4 of WCAP-8822. All blowdown releases will be calculated using the MARVEL code (WCAP-8843) and will use the specific plant design parameters for the Salem unit. Slowdown effluent will be assumed to be dry steam. The analysis will include considera-tion of the effects of failures of a containment safegua~ds train, a main feedwater line isolation valve, a main steam line isolation valve, and a failure in the auxiliary feedwater runout protection system.

SNGS-FSAR Amendment 43 Units 1&2 Q5. 8 2 P78 139 35

QUESTION 5.83 Discuss the method by which unisolated steam in the main steam line and turbine plant is added to the containment. Provide a table of unisolated steam mass as a function of power level*

for when the steam line isolation valve closes and when it is assumed not to close.

ANSWER C~nsideration of unisolated steam line mass will be considered in the analyses as described in Sections III.l.C.l and III.2.B of Appendix A of WCAP-8822. The steam line volumes with and without a valve failure will be provided with the analyses

• SNGS-FSAR Amendment 43 Units 1&2 QS. 83 P78 139 34

QUESTION 5.84 Discuss the method by which unisolated f~edwater in the main and auxiliary feedwater systems is added to the affected steam generator following a postulated main steam line break. Pro-vide the mass of unisolated feedwater with and without a fa il ure of a feedwater isolation valve.

ANSWER Feedwater in the unisolated portions of the feedlines will be considered via use of the feedline flashing model in the MARVEL code. Discussion of this model can be found in Section 2.23 of WCAP-8843. Volumes of the unisolated feed lines with and with-out a feed line isolation valve failure will be provided with the analyses.

SNGS-FSAR Amendment 43 Units 1&2 QS. 84 P78 13 9 33

QUESTION 5.85 Justify by analysis the values assumed for feedwater flow into the ruptured steam generator following a main steam line break.

Flashing of the fluid in the feedwater lines should be con-sidered as well as the affect of reduced discharge pressure on the flow rates through the feedwater and condensate pumps.

ANSWER Calculations of the* feedwater flowrate to the steam generator with the broken steam line will be performed using the pump characteristics of the Salem feed pumps and the hydraulic resistances of the Salem feed system piping. The forcing function for the increased flow will be the pressure in the steam generator calculated with the MARVEL code. Assumptions used in these calculations will be provided with the analyses

• SNGS-FSAR Amendment 43 Units 1&2 QS. 85 P78 13 9 32

QUESTION 5.110

• In request for additional information 5.100 we asked that you state if the fundamental frequencies of the key subsystems are controlled to be either greater than twice or less than one-half the dominant frequencies of their supporting system. Your response stated that the fundamental frequencies of the key subsystems were considered in relation to the dominant frequen-cies of their supporting systems. However, you did not state if the above criteria were used to accomplish the adequate design of the key subsystems or some other criteria that may be proven to be just as adequate. Provide a more detailed response to this concern.

ANSWER The fundamental frequencies of key subsystems were considered in relation to the dominant frequencies of their supporting systems. Elimination of resonance was one of the principles of design. Various methods for seismic qualification were employed for key subsystems. In most cases, the key subsystems were considered to be very flexible and were analyzed/tested as a decoupled system from the supporting system. Refer also to the res,ponse to Question 5. 38 which addresses the approach to avoid the predominant input frequencies of components to earthquake inputs. Refer also to the responses to Questions 4 .12 I 5. 3 5 I 5. 3 7 I 7. 18 and 7. 2 9 *

• SNGS-FSAR Units 1&2 Q5.110-l Amendment 43 P78 138 25

QUESTION 9.59 The design of your component cooling water system (CCWS) pro-vides a single supply and return line from the component cooling pumps to the reactor coolant pumps. Each of these lines con-tains two motor-operated valves for containment isolation.

Inadvertent failure or closure of any one of these motor-operated valves could terminate CCW flow to the reactor coolant pump coolers, which potentially may lead to fuel damage, due to a locked rotor or break of the primary coolant boundary from the loss of the mechanical seal. Therefore, it is our position that you design this portion of the compon~nt cooling water system so that the following criteria are met:

a. A single f~ilure in the component cooling water system shall not result in fuel damage or damage to the reactor coolant system pressure boundary caused by an extended loss of cooling to the reactor coolant pumps. Single failure includes operator, error, spurious actuation of motor-operated valves, and loss of component cooling water pump.

b. A moderate energy leakage crack or an accident that is initiated from a failure in the component cooling water system piping shall not result in excessive fuel damage or a br~ach of the reactor coolant system pressure boundary when an extended loss of cooling to the reactor coolant pumps occurs. A single active failure shall be considered when evaluating the consequences of the accident. Moderate leakage cracks.should be determined in accordance with the guidelines of Branch Technical Position APCSB 3-1, "Protec-tion Against Postulated Failures in Fluid Systems Outside Containment."

To meet the two criteria above, that portion of the component cooling water system which supplies cooling water to the* reactor coolant pumps can be designed to non-seismic Category I require-ments and Quality Group D if you demonstrate that the reactor coolant pumps are capable to operate with loss of cooling for longer than 30 minutes without loss of function and the need for operator protective action, and, safety grade instrumenta-tion to detect the loss of component cooling water to the reactor coolant pumps and to alarm the operator in the control room is provided. The entire instrumentation system, including audible and visual status indicators for loss of component cooling water should meet the requirements of IEEE Std. 279-1971. Alter-nately, if it cannot be demonstrated that the reactor coolant pumps will operate longer than 30 minutes without loss of func-tion or operator corrective action, then your design must meet the following requirements for the entire component cooling watar system:

• SNGS-FSAR Amendment 43 UNITS 1 & 2 Q9.59-l P78 64 46

a. Safety grade instrumentation consistent with the criteria for the protective system shall be provided to initiate automatic protection of the plant. For this case, the component cooling water supply to the seal and bearing of the pumps may be designed to non-seismic Category I requirements and Quality Group D; or

b. The component cooling water supply to the pumps shall be capable of withstanding a single active failure or a moderate energy line crack as defined in our Branch Tech-nical Position APCSB 3-1 and be designed to seismic Category I, Quality Group C, and ASME Section III, Class 3 require-ments.,

ANSWER The information presented below is provided in response to a meeting held with the NRC staff on August 31, 1978.

1. Description of Cooling Water Supply to Pump/Motor Component cooling water is provided to the reactor coolant pump thermal barrier heat exchanger, as well as to the

• upper and lower motor bearing oil coolers. In addition, seal injection flow is supplied to the pumps from the chemical and volume control system. These cooling supplies

• are discussed in the following paragraphs and are shown schematically in Figure Q9.59-l. Detailed flow diagrams of the chemical and volume control system and the component cooling water system are shown in FSAR Figures 9.2-1 and 9.5-1, respectively.

Seal Injection System Seal injection flow, at a slightly higher pressure and at a lower temperature than the reactor coolant system, enters

• SNGS-FSAR UNITS 1 & 2 Q9.59-2 Amendment 43 P78 64 47

QUESTION 9.59 (Continued)

• the pump through a pipe connection on the thermal barrier flange (see Figure Q9.59-2) and is directed to a point between the pump radial bearing and the thermal barrier heat exchanger. Here the flow splits with a portion flow-ing down through the thermal barrier labyrinth (where it acts as a buffer to prevent reactor coolant from entering tha radial bearing and seal section of the pump) and into the reactor coolant system. The remainder of the seal injection water flows up through the pump radial bearing and the shaft seals and is discharged via the seal leakoffs.

The pump shaft seal section consists of three seals in series, which are contained within a seal housing. The

• seal arrangement is shown in Figure Q9.59-3. The No. 1 seal, the primary seal of the pump, is a controlled-leakage, film-riding seal. Most of the seal injection flow (dir~ct-

• ed to the seals) is discharged through the No. 1 seal leak-off; **which is piped to *the volume control tank. Minor leakage passes through the No. 2 and No. 3 seals, which are rubbing-face type seals. The No. 2 and No. 3 seal leakoffs are directed to the reactor coolant drain tank. This arrangement minimizes leakage of water and vapor into the containment *

• SNGS-FSAR UNITS 1 & 2 Q9.59-3 Amendment 43 P78 64 48

QUESTION 9.59 (Containment)

• Thermal Barrier System The thermal barrier is a welded assembly consisting of a flanged cylindrical shell, a series of concentric stain-less steel cans, a heat exchanger coil assembly, and three flanged water connections~

Component cooling water enters the thermal barrier through a flanged connection on the thermal barrier flange (see Figure Q9.59-2). The cooling water flows through the inside of the coiled stainless steel tubing in the heat exchanger and exits through another flanged connection on the thermal barrier flange.

During normal operation, the thermal barrier limits the

• heat transfer from the reactor coolant to the pump

.internals~ If a loss of seal injection flow ~hould occur, the heat exchanger in the thermal barrier assembly cools the reactor coolant before it enters the radial bearing and the shaft seal area. Conversely, if a loss of com-ponent cooling water to the thermal barrier heat exchanger should occur, the seal injection flow is sufficient to prevent damage to the seals.

Upper Motor Bearing Assembly/Upper Motor Bearing Oil Cooler The upper bearing assembly contains an oil-cooled pivoted-pad radial guide bearing (upper guide bearing), as well as

• SNGS-FSAR UNITS 1 & 2 Q9.59-4 Amendment 43 P78 64 66

QUESTION 9.59 (Continued}

a double acting oil-cooled Kingsbury-type thrust bearing (see Figure Q9.59-4}. The thrust bearing shoes are positioned above and below a common runner to accommodate thrust in both directions. The shoes are mounted on equalizing pads, which distribute the thrust load equally to all the shoes.

The oil is circulated through and cooled by component cooling *water in an external oil-to-water shell and tube heat exchanger (oil cooler)

• Lower Guide Bearing/Lower Motor Bearing Oil Cooler The lower guide bearing is a pivoted-pad radial bearing, similar to the upper guide bearing *

• The entire lower. guide bearing assembly is located in the lower oil reservoir, which contains an ~ntegral oil-to-water coil type heat exchanger (see Figure Q9.59~4).

2. Previous RCP Testing As discussed in_(l) above, component cooling water is provided to the reactor coolant pump thermal barrier heat exchanger, as well as to the upper and lower motor bearing oil coolers. Should a loss of CCW to the RCP's occur, the chemical and volume control system continues to provide seal injection flow to the RCP's; the seal injection flow

• SNGS-FSAR UNITS 1 & 2 Q9.59-5 Amendment 43 P78 64 67

QUESTION 9.59 (Continued)

• • is sufficient to prevent damage to the seal with a loss of thermal barrier cooling. However, the loss of CCW to the motor bearing oil coolers will result in an increase in oil temperature and a corresponding rise in motor bearing metal temperature. It has been demonstrated by testing, discussed below, that the reactor coolant pumps will incur no damage as a result of a CCW flow interruption of ten minutes.

Two RCP motors have been tested with interrupted CCW flow; these tests were conducted at the Westinghouse Electro Mechanical Division. In both cases, the reactor coolant pumps were operated to achieve "hot" (2230 psia, 552°F) equilibrium conditions. After the bearing temperatures stabli~ed, the cooling water flow to the upper and lower motor bearing oil coolers was terminated and bearing (upper thrust, lower thrust, upper guide and lower guide) temperatures were monitored. A bearing metal temperature of 185°F was established as the maximum test temperature.

When that temperature was reached, the cooling water flow was restored.

In both tests, the upper thrust bearing exhibited the limiting temperatures. Figure Q9.59-5 shows the upper thrust bearing temperature versus time. In both cases, 185°F was reached.in approximately ten minutes.

SNGS-FSAR Amendment 43 UNITS 1 & 2 Q9.59-6 P78 135 61

QUESTION 9.59 (Continued)

• The maximum test temperature of 185°F is also the suggest-ed alarm setpoint temperature and the suggested trip tern-perature is 195°F. It should be noted that the melting point of the babbitt bearing metal exceeds 400°F.

The information presented above constitutes the basis of the RCP qualification for ten minute operation without CCW with no resultant damage *.

3. Operating Procedures PSE&G Operating Procedures for Salem Unit 1 and 2 have been revised to address the loss of component cooling water to the reactor coolant pumps in sufficient detail to cover the concerns expressed. Upon ....a low component cooli_!lg flow .;:ilarm to any reactor coolant pump, the operator will trip the reactor and reactor coolant pumps within five minutes if flow cannot be restored to the reactor coolant _pumps.

This action will be performed prior to the motor bearing reaching its design operating temperature.

4. The locked rotor and mutiple pump seizure analyses will be provided the week of October 23, 1978.

SNGS-FSAR Amendment 43 UNITS 1 & 2 Q9.59-7 P78 135 62

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QUESTION 12.23 We have reviewed Section 12.9 of the Salem FSAR against the minimum requirements of Appendix E to 10 CFR Part 50 and our current positions in Annex A of Regulatory Guide 1.101 -

March 1977 (the Guide) and find that the following additional information is required for our review.

QUESTION 12.23.1 Appendix E at IV.C requires a description of your means for determining the magnitude of a release of radioactive materials, and criteria for notification and participation of offsite agencies, and criteria for initiating protective actions offsite. Per Sections 4.1.4 and 4.1.5 of the Guide your plans should include action levels (specific instrument readings and alarms - see definitions in the Guide) compatible with the criteria required by Appendix E. Per 4.1.5 of the Guide, emergency action levels for declaring your general emergency class (requiring immediate initiation of protective actions offsite) should be specified in terms of information readily available in the control room. An acceptable planning basis for this class is the most serious design basis accident analyzed for siting purposes. Per 4.1.4 of the Guide, action levels for declaring a site emergency (requiring mobilization, but without planned or predetermined protective actions, off-site) should be specified either in terms of in-plant instrument readings, or in terms of instrument readings for monitored pathways in the environment. Such information is not included in your FSAR. Thus, please provide the following information:

a. With respect to 4.1.5 of the Guide:

(1) Provide the results of calculations which demonstrate your capability to measure dose rates in containment assuming the release of containment of (i) all primary water, the radio-nuclide concentration of which would be the maximum allowed by your technical specifications, (ii) 100% of the gap activity in the core, and (iii) 100% of the noble gas and 50% of the radioiodine activi-ties in the core, all for operation at 100% power.

Simplifying assumptions may be made, but should be summarized.

(2) Specify the emergency action levels (specific instru-ment readings in the control room, including radio-logical, temperature, pressure, and meteorological) at which you could certainly recommend the initiation of protective measures offsite. Show that these action levels are related to plume exposure protective action guides for emergency Class 4 in the Delaware and New Jersey plans

• SNGS-FSAR Amendment 43 UNITS 1 & 2 Ql2.23-l P78 64 68

b. Per 4.1.4 of the Guide, specify the action levels based on in-plant indicators at which you would certainly declare a site emergency (corresponding to emergency Class 3 in New Jersey and Delaware). These should be related to monitored variables in containment, including the spent fuel storage pool area.

c. Per 4.1.4 of the Guide, specify derived action levels based on water, soil, vegetation, and milk radiological monitoring offsite at which you would declare a site emergency. In particular, for the cases of radionuclides deposited on pasture land, relate the Federal Radiation Council PAG for milk (10 rem) to instrument readings that would be observed about three feet above the ground (in the absence of a plume),

e.g., beta/gamma dose rates or count rates in windows when using a portable spectrum analyzer.

d. On page 12.9-27 of your FSAR you state that the concentra-tions of radioactivity being released to the atmosphere can be determined. Describe your capabilities to measure release rates of radioiodines to the atmosphere and features of that capability which assure that the results would not be masked by noble gases and noble gas daughter products, over the spectrum of conservatively analyzed accidents con-sidered in your FSAR.

ANSWER The Radiation Monitoring System (RMS) for Units 1 & 2 provide normal containment atmosphere monitoring of noble gases, particulates and iodine. The system also includes high range 103 R/hr area monitor. This high range monitor is designed to withstand unfavorable environmental conditions (pressure, tern-perature and humidity). FSAR Figure 7.5-2 provides a plot of containment dose versus time for a design basis LOCA, utilizing Regulatory Guide 1.4 assumptions. The gap activity case would be* a factor of approximately 60 lower than the design basis LOCA case *

• SNGS-FSAR UNITS 1 & 2 Ql2.23-2 Amendment 4 3 P78 64 69

The monitoring channels necessary for identification of a Class IV

• accident are located in both No. 1 & 2 Unit Control Rooms, with the exception of the meteorological channels, which are located in the No. 1 Unit Control Room. Figure 12.9-14 lists the combined States of New Jersey and Delaware action levels as well as the associated Salem emergency procedures which would be used to identify Class I through IV emergencies associated with these protective action guides (PAG).

In the unlikely event that the containment dose rate exceeds the high range area monitors (103 R/hr), contingency calcula-tions which perdict the worst case dose are available (refer to Section 12.9.7.2). Table 12.9-2 provides Control Room instru-mentation available for assessment of the magnitude of an accident.

The actuation of high alarms on the RMS channels Rlla, Rl2A, and Rl2B (which sample the containment atmosphere), as well as the high range containment area monitor would indicate a reactor containment accident. Additionally, the vent radiation monitors provide for identification and evaluation of an accident in the spent fuel pool areas. The Emergency Procedures provide for identification of all classes of emergencies.

The State' Governments of New Jersey and Delaware have the ultimate responsibility for performance of the necessary off-site monitoring, evacuation and condemnation of drinking water SNGS-FSAR Amendment 43 UNITS 1 & 2 Ql2.23-3 P78 64 70

and food stuffs in the event such actions should become neces-

• sary. Offsite emergency evaluation teams would be dispatched to monitor offsite radiation levels in the event of an abnormal release and to supplement or assist the various State organi-zations. It is the responsibility of each State, however, to organize and direct their respectiv~ offsite operations following notification of a Salem Nuclear Generating Station emergency.

The States of New Jersey and Delaware would be notified of an abnormal release long before radionuclides in drinking water, soil, vegetation and milk would concentrate to the Emergency PAG levels or the does rates would approach the emergency PAG levels.

Repose Levels and Derived Action Levels applicable to State Protective Action provides are as follows:

1. EMERGENCY PAGl, which is:

(i) 30 rem dose commitment to the thyroid or (ii) 10 rem dose commitment to the bone marrow or whole body for an exposed individual in the population.

(iii) As an operational technique, 10 rem dose commit-ment to the thyroid or 3 rem dose commitment to the bone marrow or whole body averaged over a suitable sample.

1 These recommendations are contained in the Proposed Response Recommendations in Case of an Event Involvin the Radioactive Contamination of a io og1ca

p. 16 & 17.

SNGS-FSAR Amendment 43 UNITS 1 & 2 Ql2.23-4 P78 64 71

Response level for EMERGENCY PAG The response levels equivalent to the EMERGENCY PAG are presented for both infants and adults, in order to permit use of either level and thus, insure a flexible approach to taking action in cases where exposure of the most sensitive portion of the population (infants and pregnant women) can be prevented.

Peak Ac ti vi ty 131I 137cs 90sr 89sr Infant Adult Infant Adult Infant Adult Infant Adult Milk, or Water 0.08 1.3 1.8 3.6 0.05 0.18 1.2 40 uCi per liter Pasture, uCi per 0.6 9 6 12 2 8 60 2000 square meter mR/hr at 3 ft bove pasture* 0.008** 0.12 0.18 0.48 0.15 0.55 8.0 267.0 The states of New Jersey and Delaware have developed detailed PIPAGs which include field monitoring and condemnation procedures. Ex-cerpts from these plans are included in this response as Figures 12.9-3, 12.9-12 and 12.9-14.

In order to keep this discussion in proper perspective, the levels of radioiodine concentrations in milk requiring notification are 3.5 pG/1 as per current technical specifications. The sampling frequency required by these technical specifications is one or two weeks based on last analysis *

• SNGS-FSAR UNITS 1 & 2 Ql2.23-6 Amendment 43 P78 62 44

• Open window ion chamber survey meter such as Juno or Cutie Pie with background subtracted.

• • Due to the extremely low dose rates from radioiodine deposited on the ground (.008 mR/hr in a .005 to .015 mR/hr natural back-ground) which are indicated by protective action levels for infants and the required sensitive instrumentation necessary for the detection of these levels. Alternate means of measurement may be required *

• SNGS-FSAR UNITS 1 & 2 Ql2.23-6 Amendment 43 P78 62 75

QUESTION 12.23.ld On Page 12.9-7 of your FSAR you state that the concentration of radioactivity being released to the atmosphere can be deter-mined. Describe your capability to measure release rates of radionuclides to the atmosphere and features of that capability which assure that the result would not be masked by noble gas and noble gas daughter products, over the spectrum of conserva-tively analyzed accidents considered in your FSAR.

ANSWER Refer to FSAR Section 12.9.8.2 *

• SNGS-FSAR Units 1 & 2 Ql2.23-7 Amendment 43 P78 62 48

I

.... QUESTION 12.23 We have reviewed Section 12.9 of the Salem FSAR against the minimum requirements of Appendix E to 10 CFR Part 50 and our current positions in Annex A of Regulatory Guide 1.101 -

March 1977 (the Guide) and find that the following additional information is required for our review.

QUESTION 12.23.1 Appendix E at IV.C requires a description of your means for determining the magnitude of a release of radioactive materials, and criteria for notification and participation of offsite agencies, and criteria for initiating protective actions offsite. Per Sections 4.1.4 and 4.1.5 of the Guide your plans should include action levels (specific instrument readings and alarms - see definitions in the Guide) compatible with the criteria required by Appendix E. Per 4.1.5 of the Guide, emergency action levels for declaring your general emergency class (requiring immediate initiation of protective actions offsite) should be specified in terms of information readily available in the control room. An acceptable planning basis for this class is the most serious design basis accident analyzed for siting purposes. Per 4.1.4 of the Guide, action levels for declaring a site emergency (requiring mobilization, but without planned or predetermined protective actions, off-site) should be specified either in terms of in-plant instrument readings, or in terms of instrument readings for monitored pathways in the environment. Such information is not included in your FSAR. Thus, please provide the following information:

a. With respect to 4.1.5 of the Guide:

(1) Provide the results of calculations which demonstrate your capability to measure dose rates in containment assuming the release of containment of (i) all primary water, the radio-nuclide concentration of which would be the maximum allowed by your technical specifications, (ii) 100% of the gap activity in the core, and (iii) 100% of the noble gas and 50% of the radioiodine activi-ties in the core, all for operation at 100% power.

Simplifying assumptions may be made, but should be summarized.

(2) Specify the emergency action levels (specific instru-ment readings in the control room, including radio-logical, temperature, pressure, and meteorological) at which you could certainly recommend the initiation of protective measures offsite. Show that these action levels are related to plume exposure protective action guides for emergency Class 4 in the Delaware and New Jersey plans *

• SNGS-FSAR UNITS 1 & 2 Ql2.23-l Amendment 43 P78 64 68

____.,,,.____**-**--*-----=-- __ .,:_

\'

b. Per 4.1.4 of the Guide, specify the action levels based on in-plant indicators at which you would certainly declare a site emergency (corresponding to emergency Class 3 in New Jersey and Delaware). These should be related to monitored variables in containment, including the spent fuel storage pool area.

c. Per 4.1.4 of the Guide, specify derived action levels based on water, soil, vegetation, and milk radiological monitoring offsite at which you would declare a site emergency. In particular, for the cases of radionuclides deposited on pasture land, relate the Federal Radiation Council PAG for milk (10 rem) to instrument readings that would be observed about three feet above the ground (in the absence of a plume),

e.g., beta/gamma dose rates or count rates in windows when

• using a portable spectrum analyzer.

d. On page 12.9-27 of your FSAR you state that the concentra-tions of radioactivity being released to the atmosphere can be determined. Describe your capabilities to measure release rates of radioiodines to the atmosphere and features of that capability which assure that the results would not be masked by noble gases and noble gas daughter products, over the spectrum of conservatively analyzed accidents con-sidered in your FSAR.

ANSWER The Radiation Monitoring System (RMS) for Units l &, 2 provide normal containment atmosphere monitoring of noble gases, particulates and iodine. The system also includes high range 103 R/hr area monitor. This high range monitor is designed to withstand unfavorable environmental conditions (pressure, tern-perature and humidity). FSAR Figure 7.5-2 provides a plot of containment dose versus time for a design basis LOCA, utilizing Regulatory Guide 1.4 assumptions. The gap activity case would be*a factor of approximately 60 lower than the design basis LOCA case.

SNGS-FSAR Amendment 43 UNITS 1 & 2 012.23-2 P78 64 69

The monitoring channels necessary for identification of a Class IV accident are located in both No. 1 & 2 Unit Control Rooms, with the exception of the meteorological channels, which are located in the No. 1 Unit Control Room. Figure 12.9-14 lists the combined States of New Jersey and Delaware action levels as well as the associated Salem emergency procedures which would be used to identify Class I through IV emergencies associated with these protective action guides (PAG).

In the unlikely event that the containment dose rate exceeds the high range area monitors (103 R/hr), contingency calcula-tions which perdict the worst case dose are available (refer to Section 12.9.7.2). Table 12.9-2 provides Control Room instru-mentation available for assessment of the magnitude of an accident.

The actuation of high alarms on the RMS channels Rlla, Rl2A, and Rl2B (which sample the containment atmosphere), as well as the high range containment area monitor would indicate a reactor containment accident. Additionally, the vent radiation monitors provide for identification and evaluation of an accident in the spent fuel pool areas. The Emergency Procedures provide for identification of all classes of emergencies.

The State Governments of New Jersey and Delaware have the ultimate responsibility for performance of the necessary off-site monitoring, evacuation and condemnation of drinking water SNGS-FSAR Amendment 43

• UNITS 1 & 2 Ql2. 23-3 P78 64 70

~ and food stuffs in the event such actions should become neces-sary. Offsite emergency evaluation teams would be dispatched to monitor offsite radiation levels in the event of an abnormal release and to supplement or assist the various State organi-zations. It is the responsibility of each State, however, to organize and direct their respectiv' offsite operations following notification of a Salem Nuclear Generating Station emergency.

The States of New Jersey and Delaware would be notified of an abnormal release long before radionuclides in drinking water, soil, vegetation and milk would concentrate to the Emergency PAG levels or the does rates would approach the emergency PAG levels.

Repose Levels and Derived Action Levels applicable to State

• Protective Action provides are as follows:

1. EMERGENCY PAGl, which is:

(i) 30 rem dose commitment to the thyroid or (ii) 10 rem dose commitment to the bone marrow or whole body for an exposed individual in the population.

(iii) As an operational technique, 10 rem dose commit-ment to the thyroid or 3 rem dose commitment to the bone marrow or whole body averaged over a suitable sample.

1 These recommendations are contained in the Proposed Response Recommendations in Case of an Event Involvinl the Radioactive Contamination of Food and An1ma Feeds, Bureau of Rad1olog1cal Health, HEW, December 1976 Draft, p. 16 & 17.

SNGS-FSAR Amendment 43 UNITS 1 & 2 Ql2.23-4 P78 64 71

Response level for EMERGENCY PAG The response levels equivalent to the EMERGENCY PAG are presented for both infants and adults, in order to permit use of either level and thus, insure a flexible approach to taking action in cases where exposure of the most sensitive portion 6f the population (infants and pregnant women) can be prevented.

Peak Activity 131I

• 137cs 90sr 89sr Infant Adult Infant Adult Infant Adult Infant Adult Milk, or Water 0.08 l. 3 1.8 3.6 0.05 0.18 1.2 40 uCi per liter Pasture, uCi per 0.6 9 6 12 2 8 60 2000 square meter mR/hr at 3 ft above pasture* 0.008** 0.12 0.18 0.48 0.15 0.55 8.0 267.0 Th~ states of New Jersey and Delaware have developed detailed PIPAGs which include field monitoring and condemnation procedures. Ex-cerpts from these plans are included in this response as Figures 12.9-3, 12.9-12 and 12.9-14.

-In order to keep this discussion in proper perspective, the levels of radioiodine concentrations in milk requiring notification are 3.5 pG/1 as per current technical specifications. The sampling frequency required by these technical specifications is one or two weeks based on last analysis.

SNGS-FSAR Amendment 43 UNITS 1 & 2 Ql2.23-6 P78 62 44

• Open window ion chamber survey meter such as Juno or Cutie Pie

• with background subtracted.

• • Due to the extremely low dose rates from radioiodine deposited on the ground (.008 mR/hr in a .005 to .015 mR/hr natural back-ground) which are indicated by protective action levels for infants and the required sensitive instrumentation necessary for the detection of these levels. Alternate means of measurement may be required.

SNGS-FSAR Amendment 43 UNITS 1 & 2 Ql2.23-6 P78 62 75

t. I j

.I

.I QUESTION 12.23.ld On Page 12.9-7 of your FSAR you state that the concentration of radioactivity being released to the atmosphere can be deter-mined. Describe your capability to measure release rates of radionuclides to the atmosphere and features of that capability which assure that the result would not be masked by noble gas and noble gas daughter products, over the spectrum of conserva-tively analyzed accidents considered in your FSAR.

ANSWER Refer to FSAR Section 12.9.8.2

• SNGS-FSAR Amendment 43 Units l & 2 Ql2.23-7 P78 62 48

""' ur NJ \rl) ui< 11r.l, \C) Nl1'.:U.At< !N(;!IJr,NJ' Al.f;ll'l' CA'l'f:GOHH:S

~J (al or Di!l(c)

~~~~r Jl"n..:y

..:~,:1.:i:.i-.1 C')de

• Stat2*

STATE OF NJ OR DEL

~r: m:' !Wlr::wMml' County weal Citizens NOCLF'.AR U'l'I LITY CONS P.(XJF:OCES Utility/NRC Emergency Level or Classifica-tion Control of Facility Radioeftluents (d) (e)

SM'.S Ernergency Plttn Reference REFF:RE:r-K:ES TO NJ EMERGEl'l'.:

PW!?oJING DXlJ-IENJ'S PIPAG MFJllO (a)

PIPl-G MAN\JhL {bl GENERAL PROC CD-DC BRP PROC

• REFFR~ES m' r,-:r.

EJiofDl"."iF.~X:Y Plt..*~;ra.:~

DXLP~ti!!'.:J

,(c)

SP CCO:: BRP (NJ PF.RSONNF.L Fully in EP fI-1, 16, IV N.A. App. App. 1, 2, 3 , 8 SP Cro:: (Del) or Control 22 VII c Alf.RT p. 7 VIII Col!lllunications tests & in- none none none or EPI II-1, 4, forr:iation releases, drills, UJCAL 5, 6, 7, surveillance by utility e, 9 BRP (NJ) direct or Q)unty CD UNIT or Possibility of EP tr-2, 3, p. 5 IV N.A. App. A;lp. 1, 2, 3, 4, 8 coo: (Del) direct (assistance may be none none PWll' some release 9, 15, v c &

SP BRP CU) needed) (Activate (single (Possibility of 16, lB, Am. tl, VII c SP crr;c (Del) ap;:il icable county (building) some on-site 22 p. l EOC's) evacuations)

Active surveillance may be needed BPP (NJ) direct SITE Kno..,.. releases EP il-2, 3, 4, p. 5 IV App. App. App. l, 2, 3, 4, 5, ca:.c (Del) direct or County CD Local CD (if none or (Consequences 6, 7, e, p. 6 v A& A, C, 7, a, 14 SP BRP (NJ) local assistance STATION probably confined 13, 15, p. 7 VI B D, H, SP cro:: (Dell is needed to site. Possible lli, 17. Am. fl, VII I, J site evacuations) lB, 21, p. l Wa~er Release (Code 2) 11 Airborne Release (Code 3) or 2 Surveilla.,ce required oc 3 for a possible Code 4 Actions Invol- Kno..,.. off-site EP #1-2, 5, p. 5 IV App. App. App. l, 2, 3, 4, 5, ving Citizens consequences a, p. 6 v A & A, C, 7, B, 9, 10, 11' BRP (NJ) direct or Property (LOCA) 13, 14, p. 7 VI B D, H, 14 cro:: (Dell direct or County CD .Local CD (assistance (Prompt ini tia- 15, lSa, l\m, tl, VII I, J,

~

SP BR? (?olJ) may be needed tion of protec- 16, 17, p. l K 4 SP cox: (Del

• -o-.--.

immediately tive actions (g-h) 18, 21, in LPZ would be 22 Protective action required necessary)

(Code 4)

(6J9) 832-2000 (can also use dedicated red emergency phone beti.ieen SNGS Control Room and NJ State Police Communications Center)

(302i 673-1437 (can also use Wit/AS dedicated line beti.ieen SNGS Control Room and r::el State Police Communications Center and r::el CD-DC) i

!-~~Fr:t~t.~..t:ES:

I (<o) llJ - PSELG PIPhG Memo of Understanding, as amended I {:J) ~;J - M<lcoUal entitled *:-;cw Jersey PIPAG Manual: An Emergency Response Plan for Major Nuclear Facilities*, as amended.

' (c) Del ~**clear k:cident or Incident Control Plan, as amended II rd

( .1) S:JGS e:iergency Plan, Section 12. 9 of FSAR, as amended 3:~:;s ~ergency Procedures, as amended ij : t) llJ C:.unty Nuclear Emergency Plans, as amended j r 1) EPA - '.>lanual of Protective Action Glides and Protective Actions for Nuclear Incidents, Latest Amendment ij "1) HUI, FD'+. - The Radioactive Contamination of Food and Animal Feeds, Latest l\mendment I;" "'"" Table 012.23-1

QUESTION 14.27 In Section 14.5, no analysis has been provided for postulated moderate energy line breaks. Provide such an analysis using the criteria in the J. O'Leary letter of July 1973, or Branch Technical Position APCSB 3-1, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment."

ANSWER Moderate Energy Pipe Failure Evaluations Moderate energy fluid systems outside of containment as defined in Section I.A below, have been ev~luated for the consequences of through-wall. leakage cracks.* Components required for the safe shutdown of the reactor were evaluated and shall be provided, as necessaryr with measures to ensure operability.

I. Definitions A. Moderate Energy Lines (MEL).

Moderate energy.piping includes those systems where both of the following conditions are met:

a) The maximum operating temperature is 20QOF or less, and b) The maximum operating pressure is 275 psig or less.

B *. Hazard For purposes of this evaluation, postulated leakag~

shall be considered* for the effects of resulting flooding or liquid spray on components required for safe unit shutdown.

II. Postutated Break Location A. Moderate energy piping that is located in areas con-taining systems and components important to safety were postulated to develop a through-wall leakage crack at the most adverse location to determine protection needed to withstand the effects of the resulting liquid spray and flooding.

B. Piping systems that by plant arrangement and layout are isolated and physically separated from systems and components important to safety, were not considered for postulated leakage cracks.

SNGS-FSAR Ql4.27-l Amendment 43 UNITS 1&2 P78 140 25

c. Moderate energy p1p1ng that is located in the same area as high energy fluid systems considered for postulated breaks, was not considered for postulated leakage cracks.

III Postulated Crack Size Through-wall leakage cracks were postulated in piping runs and branc~es over 1-inch nominal size. Crack size was assumed to be 1/2 the pipe diameter in length by 1/2 the*

pi.pe wall. thickness in width.

IV Evaluation Procedure A. A review of the auxiliary building was made to deter-mine those compartments or areas with components re-quired for safe reactor shutdown.

B. Each of the compartments or areas to be considered for postulated cracks was then physically inspected for flood and spray damage potential.

c. Crack postulated flow rates of the largest MEL in the*

given space were estimated on the basis of the Bernoulli equation, Q=KA (2gh) 1/2 where K, the ori.fice co-efficient, was assumed to be 0.6. An accumulation rate or flood level was then estimated based* on a comparison between floor drainage capacity and the postulated leakage rate. If 1 iquid ~ccumulation posed a flood threat to components within a compartment, an evaluation was mada to determine the possibility of damaging each component and the acceptability of such damage. Where necessary, modification to existing design was deter-mined to correct the condition.

D. Fluid spray consequences were evaluated on the basis of line-of-sight inspection between components in the given area or space and the highest pressure or most unfavorably oriented MEL. If it was determined that liquid spray in a given compartment or space could interact with components within that space, evaluation was made to determine acceptability of such interaction and, if necessary, modification to existing design was determined to correct the condition.

SNGS-FSAR Ql4.27-2 Amendment 43 UNITS 1&2 P78 140 26

V Inspection Results A. Area 1 - RHR Pump Rooms, Elevation 45 Flooding in the RHR pump rooms can occur from a postu-lated crack in service water piping in the pipe alley on elevation 84 which in turn communicates with elevation 45 via a pipe chase. Flooding due to a ser-vice water pipe ruptQre is discussed in the response to FSAR question 9.33. Alarms in the control room resulting from high RHR pump room pump level will alert the: control room operator who in turn will terminate flow.by remote valve realignment. Flooding to the RHR pump rooms could also occur as a result of MEL fluid from breaks on upper elevations running down staircases and conceivably into both RHR pump rooms.

To prevent this from occurring, curbs shall be installed on elevation 55 immediately above the RHR pump rooms such that fluid flow from MEL failures on elevations above, will only flow to one RHR pump room not both rooms.

Water spray from component cooling lines in the RHR

• pump rooms could affect safety related equipment in those areas. A single postulated MEL failure however will only involve one of two redundant RHR pump trains because of the partition wall between the separate

.rooms.. Therefore, spray failures in this area do not jeopardize safe shutdown capability of tpe plant.

B. Area 2 - 4160V Bus Room - Elevation 64 The only piping located in this area is ~n auxiliary feed suction line and a fire hose station. The feed suction line is normally empty and is not considered for postulated cracks. The fire piping will be pro-vided with a shroud to prevent impingement spray on electrical components.

c. Area 3 - Electrical Penetration Area - Elevation 84 This area does not contain MEL piping. No modifications a re necessary *

• SNGS-FSAR UNITS 1&2 Ql4.27-3 Amendment 43 P78 140 27

• D. Areas 4, 5 and 6 - Rod Control Reactor Trip Breakers and Miscellaneous Vital Electrical Gear - Elevation 84 These areas do not contain MEL piping. Curbing at ac-cess points into these areas has been provided to prevent flooding from adjacent areas.

E. Area 7 - Safety Injection Pump Room - Elevation 84 This area contains MEL piping. Floor drainage capacity however is adequate to prevent flooding of the compart-ment. Water spray from service water or demineralized water* piping could a*ffect safety injection pump motors *.

The safety injection pump motors will be protected from overhead spray by means of a protective shroud~

F. Area 8 - Component Cooling Heating Exchanger Rooms -

Elevation 84 This area contains, service water and fire protection MEL piping. Floor drainage capacity in the area is adequate to prevent flooding. Water spray from ser-vice water pipe cracks could affect 22 and 23 component cooling pump motors and associated controls. Component c6oling pump 21, however, is isolated in another com-partment and would not be affected by this fault.

Therefore, spray failures in this area' do not affecf safe shutdown of the plant.

G~ Area 9 - Auxiliary Feed Pump Room - El~vati6n 84

. This area contains service water, fire protection, com-ponent cooling, demineralized water and refueling water storage tank piping. Floor drain capacity in the area as well as drain capacity in the corridor to which this area is open, is adequate to prevent local flood-ing~

Water spray from the MEL p1p1ng in the area can affect safety related motor control centers 2C West and 2A West as well as control panels 205, 206, 207 and 213.

In addition, water spray can affect the auxiliary feed pump motors. To prevent water spray damage to these vital components, motor control centers and control panels will be protected to withstand the effects of spray. Auxiliary feed pumps motors will be provided with a shroud. to prevent spray damage *

• SNGS-FSAR UNITS 1&2 Ql4.27-4 Amendment 43 P78 140 28

H. Area 10 - Emergency Diesel Fuel Oil Transfer Pump Room - Elevation 84 No MEL piping is located in this area. No modifications are required.

I. Area 11 - Safety Injeqtion Charging Pump Rooms -

Elevation 84 This area contains,. service water, component cooling and spent fuel cooling MEL piping. Drainage is ade-quate to prevent flooding in the area. Portions of the MEL piping will be provided with a baffle* to pre-vent sp.ray damage to the charging pump motors.

J. Area 12 - Containment Spray Pump Area - Elevation 84 This area contains refueling water storage MEL piping.

Drainage in the area is adequate to prevent flooding.

Shrouds over Containment Spray Pump Motors will be provided to prevent water spray damage from MEL piping.

K. Area 13 - Rod Control, Relay Room and Battery Racks -

Elevation 100 This area does not contain any MEL piping. No modifi-cations are required.

L. Area 14 - Electrical Penetration Area - Elevation 100 This area contains service water MEL piping to the chiller condensers. Drainage capacity is adequate to prevent flooding from postulated cracks. Water spray in this area does not present any safe shutdown hazards.

• No modifications are required.

M. Area 15 - Emetgency Diesel Rooms - Elevation 100 These rooms contain service water, demineralized water and fuel oil MEL piping. The rooms do not con-tain floor drainage, hence a postulated service water line failure could conceivably cause flooding in a single room. The individual emergency diesel engines however, are physically isolated from each other and hence local MEL failures in one room will not affect the other rooms. No modifications are required in this area *

• SNGS-FSAR UNITS 1&2 Ql4.27-5 Amendment 43 P78 140 29

ii* , ,.._ .. *W N. Area 16 - Control Room - Elevation 122 This area contains service water, chilled water and heating water piping. A pressure tight steel enclosure is provided to encase all the piping. A drainage path is provided to remove any liquid from the enclosure.

o. Area 17 - A.C. Heating and Vent Room - Elevation 122 This area contains chilled water and service water MEL piping. Drainage in the area is adequate to prevent flooding. Water spray does not present any safe shut-down hazards. No modifications are required in this area.

P. Area 18 - Service Water Intake Postulated faulting of piping in this area is discussed in the response to question 9.33. No modifications are required in this area.

IV Implementation While efforts are being made to implement the above modif i-cations prior to initial fuel loading, if this cannot be accomplished, completion will be as soon thereafter as practical *

• SNGS-FSAR UNITS 1&2 Ql4.27-6 Amendment 43 P78 140 30

QUESTION 14.28 You state in Section 14.S.l.6 that "In order to preclude unde-sirable effects due to steam flooding *** ", as a result of a steam line break backdraft-type dampers are used to prevent steam flow into adjacent vital areas either through supply or exhaust ducting. Provide the following information:

a. Provide a detailed description including drawings on how the backdraft-type dampers operate to prevent steam from entering vital areas.

b. Assuming a steam line break and a failure of any one of the backdraft-type dampers to close, show that the vital equipment and systems that were being protected can operate in the steam environment so that the plant can be brought to . a safe cold. shutdown condition *

~ ~ ': .

c. Assuming a steam line break such that the steam flow is insufficient to shut the backdraft-type dampers, show that the vital equipment and systems in all the protected areas can operate in the steam environment so that the plant can be brought to a safe cold shutdown.

ANSWER

• Backdraft dampers are of. the hinged parallel blade design with interconnecting linkage to enable blades to operate in unison.

Gasketing is provided along the blade edges to limit blade leakage to design limits. Figure Ql4.28-l illustrates typical design detail and control logic for operation. A differential pressure transmitter actuates a solenoid valve which operates the damper drive mechanism upon receipt of a trip signal. The dampers are designed to fail safe in the event of a loss of power or air.

The backdraft dampers are provided as an integral part of the equipment and hardware installed to protect against the un~

likely event of a postulated break in a high energy piping SNGS-FSAR Amendment 43 Units 1 & 2 Ql4.28-l P78 65 11

./

,_ .. ,)- - -:......;'

system. As such, the dampers are designed to Category I seismic criteria and are intended to provide an additional level of protection along with pipe encapsulation sleeves, pipe whip restraints and impingement baffles where the pre-f erred physical separation approach was not feasible in im-plementing the regulating criteria on a backfit basis.

Periodic surveillance testing and inspection will be per-formed to assure operability of the dampers.

Steam leakage*of sufficiently low magnitude to cause environ-mental conditions below the trip point of the differential pressure transmitters is within the capacity of the ventilation system such that environmental conditions necessary for the proper operation of vital equipment and systems in the pro-tected area are maintained *

• SNGS-FSAR Units 1 & 2 Ql4.28-2 Amendment 43 P78 65 12

• 7, .,-~

.....t ANALYSIS OF THE REACTOR COOLANT SYSTEM SUPPORTS FOR POSTULATED LOSS OF COOLANT ACCIDENTS FOR SALEM UNITS 1 AND 2

~he reactor coolant system was analyzed for postulated pipe ruptures at the RPV inlet nozzle, the RPV outlet nozzle, and the RCP outlet nozzle. These pipe rupture locations produce the most severe loadings on the reactor pressure vessel and have the most severe consequences upon structures required to assure plant safety. The analyses and calculations dis-cussed below include the effects of a plant modification which reduces the severity of the postulated event. This modification is the addition of pipe displacement restraints in each primary shield wall pipe annulus. These restraints limit the break opening area for pipe ruptures postulated at the reactor vessel safe end locations *.

Thebreak opening areas which were assumed for the RPV in-let nozzle and RPV outlet nozzle breaks were 100 square inches and 76 square inches, respectively. Based on calcu-lations of the actual break size for other Westinghouse NSSS's of similar design, the assumed values are significantly higher than the actual areas. Therefore, the analysis is conserva-tive from a break area consideration.

A dynamic analysis of the reactor pressure vessel (RPV) was performed using the DARIMOSTAS 3 code in which time-history loads associated with each postulated break were applied to a mathematical model of the RPV and internals. The applied loads included RPV internal. hydraulic forces, RCL mechanical loads, and reactor cayity pressurization forces. All input to the analysis was specifically applicable to Salem Units

1. and 2. The resul.ts of the dynamic analysis included time-history displa6ements of the RPV and loads on the RPV sup-ports~ The RPV displacements were used as input for analysis of other RCS components, and the support loads were used to evaluate the RPV supports.

The cavity pressures were calculated using the TMD 2 com-puter code with the unaugmented homogenous critical flow correlation and the isentropic compressible subsonic flow correlation (y factor). The 70 node model used for Salem has been shown to have an adequate number of nodes by com-parison to a generic noding study. All insulation was as-

. sumed in place and uncrushed during the entire transient ex-cept on the broken loop nozzle. This insulation was assumed to crush to zero thickness. Although the sand in the inspec-tion ports is designed to be expelled during the hypothetical

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accident, no credit was taken in the anaJ.ysis for venting thru the inspection ports. Reactor cavity loads were calculated for a 100 square inch guillotine break opening at the cold leg nozzle safe end. This break area has been verified to be the maximum possible opening area. Since the break area calculated for a reactor vessel outlet nozzle break is much smaller than for the inlet break, and since the flow path data are similar for the two breaks, the pressurization forces would be more severe for an inlet nozzle break than for an outlet nozzle break. The highest calculated pressure in the reactor vessel annulus was calculated* to be 495.l psig. Transformation of the pressures into corre*pondin~ forces and moments acting on the reactor vessel were made in order to assess the im-pact of the transient reactor cavity pressures on the reactor vessel supports. The peak horizontal force was calculated to be 4 3 3 6 kips

• The internal hydraulic loads were calculated using the Multi-flex 1 code. Depressurization waves propagate from the postu-lated break location into the reactor vessel through either a hot leg or a cold leg nozzle. After a postulated break at the RPV inlet nozzle or at the RCP outlet nozzle, the depres-surization path for waves entering the reactor vessel is through the nozzle which contains the broken pipe and into the downcomer annulus which is the region between the core barrel and reactor vessel~ The initial waves propagate up, around, and down the downcomer annulus, then up through the region circumferentially enclosed by the core *barrel; that is, the f~el region~ As a result, the region of the down-comer annulus .close- to the break depressurizes rapidly but, because of restric.ted flow areas and finite wave speed (ap-proximately 3500 f~et per second), the opposite side of the core barrel remains at a high pressure. This results in a net horizontal force on the core barrel and RPV. As the depressurization wave propagates around the downcomer annu-lus and up through the core, the barrel differential pressure reduces, and similarly, the resulting hydraulic forces drop.

In the case of a postulated RPV outlet rupture, the waves follow a dissimilar depressurization path, passing through the outlet nozzle and directly into the upper internals re-g ion, depressurizing the core, and entering the downcomer annulus from the bottom exit of the core barrel. Since the depressurization wave travels directly to the inside of the core barrel (so that the downcomer annulus is not directly involved), the internal differential pressures are not as

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large as for th~ RPV inlet nozzle break, and therefore, the horizontal force applied to the core barrel is less for the hot leg break than for a* cold leg RPV inlet nozzle break.

For breaks in either the hog leg or cold leg, the depressuri-zation waves would continue to propagate by reflection and translation through the reactor vessel and loops. The reac-tor coolant pump outlet nozzle and reactor pressure vessel inlet nozzle pipe rupture locations have similar vessel internal hydraulic loads, but due to the influence of reac-tor, cavity pressure loads, the vessel inlet nzzle break generates larger net f~rces applied to the reactor vessel.

The reaetor coolant loop mechanical loads are applied to the RPV nozzles by the reactor coolant loop piping. For guillo-tine pipe separations, the loop mechanical loads result from the release of normal operating forces present in the pipe prior to the separation as well as from transient hy-draulic forces in the reactor coolant system. The magnitudes of the loop release forces are determined by performing a reactor coolant loop analysis.for normal operating loads (pressure, thermal, and deadweight). The loads existing in the pipe at the postulated break locations are calculated and are "released" at the initiation of the LOCA transient by application of the loads to the broken piping ends.

These forces are applied with a ramp time of 1 millisecond due to the assumed instantaneous break opening time.

All loads-that.could be applied to the *reactor coolant system as a resuit of the postulated breaks were iricluded in the analysis~

The reacto*r coolant loop piping stresses were evaluated

~gainst the faulted condition stress limit. The loads in-cluded in the evaluation result from deadweight, pressure, DBE, LOCA loop hydraulic forces, and reactor vessel motion.

The dynamic structural response of the loop piping was calcu-lated for the simultaneous occurrence of the hydraulic forces acting at changes in direction and changes in area, which vessel motion applied at the vessel location. The complexity of the reactor coolant loop/supports system re-quired the use of the WESTDYN3 computer program to obtain the displacements, forces, and stresses in the piping and support members. The WESTDYN computer program is capable of performing an elastic analysis of redundant piping sys-tems subjected to thermal, static, and dynamic loads. In

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r the LOCA evaluation, the faulted condition stress limits must be met for the unbroken legs of any reactor coolant loop. To ensure the RCL supports system integrity, the stresses in the unbroken leg of a broken loop as well as the unbroken loop piping were evaluated. The results show the reactor coolant loop piping meet the faulted con-dition requirements of ASME Section III and is capable of withstanding *the consequences resulting from the afore-mentioned breaks.

The external loads imposed on the reactor* coolant system com-ponents were evaluated. All the nozzles for RCS components are capable of withstanding all of the accident. loads.

The review included the following locations:

(1) Steam generator *primary inlet and primary outlet nozzles (2) Reactor coolant pump inlet and outlet nozzles (3) Reactor pressure vessel inlet and outlet nozzles The loads induced in the components at these locations were compared to loads which used in previous analyses for Salem Units 1 and 2 or for other plants which-iiave identical

• equipment. The criteria used in these previous analyses were those presented in the FSAR for the faulted condition.

The stresses in a nozzle are calculated from six components of load applied to the nozzle. The comparison of the in-dividual comporients of load for the pipe break loads derived in the analyses to the loads of previous analyses indicated that the resultant stresses will be lower than those pre-viously calculated.

A dynamic analysis was performed to evaluate the response of the control rod drive mechanisms (CRDM's) to the reactor pressure vessel motion during LOCA. The time-history dis-placements from the RPV dynamic analysis were applied to a mathematical model of the CRDMs. The resulting loads and stresses were compared to allowable to verify the adequacy of the CRDM's. For postulated breaks less than 3 ft2, the rods are not assumed to drop until approximately 2 to 3 seconds following the occurrence of RCS conditions which generate a reactor trip signal. Since the peak transient forces and displacements during LOCA occur SNGS-FSAR Amendment 43 Units 1&2 P78 139 62

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within the first 0.5 seconds following the break, the motion of the transient will not affect the ability of the rods to drop when required.

The fuel assembly response resulting from the RPV inlet nozzle break and the RCP outlet nozzle breaks were analyzed using time-history numerical techniques. The RPV outlet nozzle break was not analyzed since previous analyses have indicated that the fuel assembly impact forces and deflections for this break size and location were substantially less than the break considered. The motion o.f the vessel and core plates for the RPV outlet break compared to the inlet break confirms the higher severity of the vessel inlet break. The vessel motion from the LOCA analyses induce lateral loads on the reactor core which are analyzed using a finite element model similar to the seismic model of WCAP-82364. The reactor core finite element model which is used to asses fuel assembly interaction during lateral excitation consists of 15 fuel assemblies arranged in a planar array with inter-assembly gaps. The time-history motion of the upper and lower core plates and the barrel at the upper core plate elevation, which were obtained from the RPV dynamic analysis, were simultaneously applied to the reactor core model. The fuel ~ssembly stresses and grid impact forces were evaluated, and indicated substantial margin compared to allowable values.

The reactor internals were evaluated for all three postulated breaks.* This evaluation combined the response of the in-ternals from the RPV dynamic analysis with the core barrel shell response obtained from a separate analysis* of the core barrel. The pressure differentials across the core barrel wall were applied to a mathematical model of the core barrelt and the beam bending stresses and shear stresses were obtained. To properly evaluate the total stress re-sults in the core barrel, the DARIMOSTAS results from the RPV dynamic analysis and the results from the core barrel shell analysis were combined on a time-history basis. The maximum stresses were substantially below the allowable values.

The LOCA loads on the steam generator and reactor coolant pump support structures have been evaluated, and preliminary results show the supports are adequate for the postulated events. Final results for the RPV support show that it is adequate. The cooling box and concrete below the RPV SNGS-FSAR Amendment 43 Units 1&2 P78 13 9 63