ML20205T233

Submits Revised Responses to Util Re Role of Operators in Mitigating High Energy Line Breaks.Analysis Complete & Certain Mods Incorporated Into Flow Path.Revised Figure 1 Re Flow Paths Encl

ML20205T233
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/31/1987
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
P-87131, TAC-61894, NUDOCS 8704070245
Download: ML20205T233 (6)

Text

hPublic Service- =lh, .

2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 March 31, 1987 Fort St. Vrain Unit No. 1 P-87131 l

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk l Washington, D.C. 2055 Attention: Mr. H. N. Berkow, Director

Standardization and Special l Projects Directorate Docket No. 50-267

SUBJECT:

Role of Operators in Mitigating l

High Energy Line Breaks at Fort St. Vrain

REFERENCES:

1) PSC Letter Warembourg to Berkow, dated June 26, 1986 l

(P-86438)

2) PSC Letter Walker to Berkow, i dated March 14, 1986 '

(P-86208)

Dear Mr. Berkow:

The role of operators in mitigating high energy line breaks (HELB) at l Fort St. Vrain (FSV) has been previously discussed in Referance 1.

As stated in that letter, the operator actions were based on preliminary results of the analysis to verify the cooldown flow paths and procedures. The analysis is now complete and certain l modifications have beer, incorporated into the flow paths. The l purpose of this letter is to revise the responses of Reference 1 to incorporate these modifications and the resulting changes in operator actions.

l 8704070245 870331 PDR ADOCK 05000267 \

V PDR b()ON

. 1 P-87131 Page 2 "

March 31, 1987 f The attached Figure 1 replaces the figure from Reference 1. It shows the flow paths to line up secondary coolant flows to the steam generators and primary coolant motive power flow. The emergency condensate header is the primary flow path and the actions required to line it up are outlined here. The emergency feedwater line is a backup and is not described in detail in this letter.

All required supporting syitems (i.e. bearing water, instrument air, etc.) are qualified to operate and do not require manual line-up.

However, various single failures can be postulated which would require local operator action. These actions are not described in detail in this letter but involve manual valve overrides or bypass valves.

1. NRC Question No. 1 List and briefly describe local manual actions that would be

, taken by the operators following a HELB accident or equivalent l primary coolant leak.

PSC Response The Equipment Qualification Program flow path for the resumption of forced cooling following a HELB involves the use of firewater to drive one circulator and provide cooling water to the economizer, evaporator, superheater (EES) section of one steam generator (refer to attached Figure 1). The following local manual actions will be required by the operators following a HELB:

• Alignment of the firewater system to the emergency condensate system is required to provide a firewater flow path to the steam generator and circulator pelton wheel.

Two manual valves (HV-4518, HV-4519) must be opened to align the flow pat!q And two manual valves (HV-4520, V-45201) must be closed to ise a portions of the firewater system not required for safe 3r.utdown to preclude firewater from being directed to any sprinklers activated by the elevated ambient temperatures. These four valves are located outside the harsh environment.

The discharge valve (HV-22821 for Loop 1 or HV-22822 for Loop 2) in the flow path for firewater cooling must be manually opened. The valves are located in the Reactor Building and rely on the manual override handwheels.

. P-87131 Pag: 3 March 31, 1987

• Local manual action would "oe required to line up secondary cooling for certain pipe break locations in combination with specific single active failures. The valves are environmentally qualified and therefore should be operable from the control room. However, if one did fail, the following actions would be performed.

Break Location Single Active Failure Manual Action Feedwater or Main HV-2237 Note 1 or Steam Loop 1 Piping FV-2205 Note 2 Feedwater or Main HV-2; J Note 1 Steam Loop 2 or Piping FV-2206 Note 2 i Note 1 Local operator manual action would consist of l turning a handwheel to manually override the valve.

Note 2 Local operator manual action would consist of I opening two small valves to admit hydraulic fluid to the valve actuator from a local accumulator, l

l

• The operator must verify the steam generator outlet pressure via a local indicator (PI-22129-2 for Loop 1 or PI-22130-2 for Loop 2). The operator must then monitor outlet l

temperature from a local temperature gauge (TI-22823 for i Loop 1 and TI-22B24 for Loop 2). The temperature gauges are located in the Reactor Building and the pressure gauges are located in the Turbine Building.

Following a primary coolant leak (i.e., Design Basis Accident No. 2, Rapid Depressurization/ Blowdown), primary coolant circulation and secondary coolant heat removal could be interrupted by actuation of SLRDIS. Although there are actions which the operator would perform from the Control Room to ensure safe shutdown of the plant, there are no immediate or necessary actions outside the Control Room which are required of him.

2. NRC Question No. 2

Provide an evaluation of the ability of the operators to perform these actions in potentially high temperature environments. This l evaluation should address the same factors covered in approving i FSV operation at 35 percent power. (Please see our letter dated l February 7, 1986, Enclosure 5, Confirmatory Action 12.)

l

~

. P-87131

Pags 4 March 31, 1987 PSC Response Opening of the discharge valve, reading the gauges and overriding the single failure rely on access in the possible harsh environment. The approximate time that the above actions would be required would be 1/2 hour following break detection or interruption of forced circulation. The composite EQ Program temperature profile shows that temperatures would be 140 degrees Fahrenheit or less (as seen in Reference 2). Access with or without cool suits would be possible.

All other manual actions identified in our response No. I do not require access to the harsh environment, including the access routes from the Control Room to the action locations.

3. NRC Question No. 3 If future modifications to the plant are contemplated that effect this issue, please describe these modifications and your proposed l schedule for their completion.

i PSC Response j During the outage at FSV, several system modifications have been implemented to reduce the operator access requirements in areas of high temperature environment following a HELB, and to facilitate the initiation of safe shutdown cooling. The modifications listed in Reference 1 have been completed. In addition, a modification to install new 6" vent line discharge paths (HV-22819, HV-22820, HV-22821 and HV-22822) has been completed.

Future modifications may be considered for justifying power levels beyond 82%. However, these modifications have not been defined at this time. Likewise, as we gain experience with the present system, we may find it advantageous to make more modifications to reduce operator manual actions, but such modifications are not presently in our plans.

+

Please note that this letter supersedes Reference 11 in Attachment 4 (Safety Analysis Report) to PSC Letter, Williams to Berkow dated January 15, 1987,

Subject:

" Proposed Technical Specification Change Eliminating Reliance on the Reheater Section of the Steam Generators for Safe Shutdown Cooling," (P-87002).

4

. ,. ,. . - - _ _ . , , . . - , - - , - . ,,- -n - - . -,

~

P-87131 Page 5 March 31, 1987 If you have any questions on this subject please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, S V Vw D. W. Warembou &-

rg, Manager Nuclear Engineering Division DWW/KD:pa Attachment i

l 1

Safe Shutdown Flow Path Using the EES Section ,

Figure 1 Fire Water Pumps V-45201 W

Turbine Bldg.

P- 01 Ring Header HV-4520 d

Y HV-4518 'I HV-5252

&a J L P-45015 3 r HV-4519 V-52989 j g To -

To Circulators "t, & 2 V-4525 Deaerator~

C-2101, C-2102 7' r' r%

a Emer. Water Booster Pumps Emer. Condensate ' leader 7 Flash

.. 'a Tank T-5201 P-2 "

r ' V-45223 To Atm.

L F Emer. Feedwater Header P-2 1 P HV-22821 To Circulators J L C-2103, C-2104 h'

HV-2201 I

k HV-2237 k

HV-

' I 2203 FV-2205 Steam Generator EES Section HV-22819 1 P d '

HV-2223 X ) ) X X Various CL.Y CL.I B-2201 I I' I' I bl.II Drain Normal Sogg i, -

N LJ kJ L Bypass g Lines to Condenser From CL II 1 r1 r1 r valves /

1 FHV-2238 1 rHV-2204 BFP's ' '

HV-2202 FV-2206 HV-2224 CL.I X V CL.II  :  : CL.I CL.I -m y CL.II F HV-22820 B-2202 To Turbine t

Note: This is a simplified diagram only showing the main HV-22822 lines and valves. Valves are shown in the nonnal '

operating position. " '- To Atm.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -