ML18092B225
| ML18092B225 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/31/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| LCR-85-07, LCR-85-7, LCR.85-7, NLR-N86094, NUDOCS 8608050104 | |
| Download: ML18092B225 (9) | |
Text
Public Service Electric and Gas Company Corbin A. McNeill, Jr.
Vice President -
Public Service Electric and Gas Company P.O. Box236, Han cocks Bridge, NJ 08038 609 339-4800 Nuclear July 31, 1986 NLR-N86094 REF:
LCR 85-07
- u. S. Nuclear Regulatory Commission Off ice of Nuclear Reactor Regulation Division of Licensing Washington, DC 20555 Attention:
Mr. Steven A. Varga PWR Project Directorate #3 Division of PWR Licensing A SUPPLEMENTAL INFORMATION TO REQUEST FOR AMENDMENT FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
Dear Mr. Varga:
Attached are our responses to ten questions asked by staff reviewers regarding our request for amendment LCR 85-07, submitted on October 25, 1985, related to Boron Injection Tank deletion.
Should you have any further questions, please do not hesitate to contact us.
Attachment 8608050104 860731 PDR ADOCK 05000272 P
PDR Sincerely,
Mr. Steven A. Varga C
Mr. Donald C. Fischer Licensing Project Manager Mr. Thomas J. Kenny Senior Resident Inspector Mr. Samuel J. Collins, Chief Projects Branch No. 2 DPRP Region I 2
7/31/86
REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGES REGARDING SALEM 1 & 2 BORON INJECTION TANKS (BITS)
REACTOR SYSTEMS BRANCH QUESTIONS (1)
For the large steam line break accident, please provide the minimum DNBR calculated assuming both present condition (BIT boron concentration of 20,000 ppm) and proposed condition (BIT boron concentration 0 ppm).
In accordance with the regulations of 10CFRS0.92 the staff requires the information to determine whether a significant reduction in margin of safety is involved.
Answer:
The minimum DNBR varies with each core load.
The Reload Safety Evaluation verifies only that the number remains above 1.30.
Reload Topical Report WCAP 9272 discusses how this verification is done.
PSE&G's position is that verifying that the DNBR remains above 1.30 during a large steam line break (a condition IV event) assures the change is not a significant safety hazard since 1.30 provides a high assurance of no fuel damage.
In addition, PSE&G considers that a one-to-one type comparison of the DNBR's is not entirely appropriate since the DNBR is calculated based on a revised model for determining the core heat flux and RCS temperature and pressure following the steam line break.
Also, the deletion of concentrated boric acid in the BIT results in improvements to safety which are not quantified by the DNBR (e.g., prevention of crystallization in the safety injection lines).
NRC has acknowledged these concerns in Generic Letter 85-16, Elimination of Boric Acid (in the BIT).
(2)
Page 3 of your submittal states:
"The heat tracing will be deleted and the BIT will be maintained at a boron concentration* between 4 weight % and 0 weight %".
Therefore assured means would still be required to maintain the fluid inside the BITS and associated SI piping at a minimum of 65°F (the minimum solution temperature of 4%
State how this will be accomplished.
- (NOTE:
We assume you mean boric acid concentration.)
1
Answer:
The statement in page 3 of our submittal is misleading.
Once the BIT to BAT recirculation lines are cut and capped, there is no credible way that the BIT concentration will increase above the concentration of the RWST, 2200 ppm or less.
Since the minimum solution temperature of 2200 ppm boric acid is below the freezing point of water, there will be no requirement to maintain either heat tracing or room heating.
All the piping in question is located inside safety related buildings with HVAC.
Salem will not eliminate or lower the setting of any of the existing heat tracing associated with the BIT until the BIT has been flushed to 2200 ppm, or less, and the connections to the BAT's and recircul~tion system are being cut and capped (or similar positive method of assuring no leakage between the BAT's and the BIT).
- Your assumption is correct.
We meant boric acid concentration.
(3)
The Salem FSAR apparently does not contain detailed piping and instrumentation drawings of the SI piping from the charging pumps via the BITs to containment penetration.
We also can not find the BIT location in the building drawings (i.e. Figure 3.6-26, 3.6-27 and 6.3-1 do not provide sufficiently detailed information).
Therefore, please provide detailed piping and instrumentation schematics and layout drawings for the above flow path.
(the latter could simply consist of a marked up building drawing.)
Also state whether the connections to the BAT and other concentrated boric acid sources will be blanked off.
Answer:
The following drawings have been transmitted to Mr. Don Fischer, NRC Project Manager for Salem:
0 0
0 Drawings 205234A8761 - Rev. 21, P&ID for Safety Injection Sheets 1-4, Unit 1 Drawing 205228A8761 -
Rev. 28, P&ID for CVCS Sheets 1-3, Unit 1 Drawing 204805A8752 - Rev. 3, General Arrangement Aux.
Bldg. 84', Reactor Containment 78' & 81' 2
0 0
0 Drawing 207452A8813 - Rev. 22, CVCS Piping Aux. Bldg, Elevation 84', Unit 1 Drawing 207458A8813 -
Rev. 16, CVCS Piping Reactor Containment, Unit 1 Drawing 207463A8844 - REv. 19, RHR & SI Piping Plan Elevation 78' & 81', Sheet 1 of 2, Unit 1 Unit 2 drawings were not provided.
For all practical purposes they are identical to Unit 1.
Refer to our answer to question 2, above, as to our intentions on capping off the BAT to BIT connections.
(4)
Provide assurance that periodic sampling of the BIT and connected SI piping will be conducted to determine boron concentration.
Also discuss whether the Salem Plant procedures will include requirements for periodic flushing of* the BIT and connected SI lines.
Answer:
Since the concentration in the BIT will be kept at or below 2200 ppm, the same as for any portion of the Safety Injection, RHR, Containment Spray, or other portions of the C/SI Systems, no requirements are being imposed to periodically sample or flush the BIT.
(5)
FSAR Figure 10.3.1 does not show check valve in the main steam lines.
Are the main steam stop valves designed to provide positive flow isolation for both forward and reverse flow?
Answer:
The Salem MSIV's are Hopkinson-type parallel slide, double-disk gate valves.
These valves are designed to provide flow shutoff in either direction.
No check valves are required to prevent backflow.
(6)
State what investigations have been made regarding the effect of the proposed decrease in the BIT boron concentration on the amount of superheat resulting from a steam line break, including the effect on equipment that can be affected by the superheated steam which may subsequently react in such a way as to increase the severity of the accident.
3
Answer:
The effect of superheat from a MSLB inside containment is discussed in WCAP 8822 supplement 2.
This report was transmitted to the NRC by Westinghouse (NS-NRC-85-3071 dated 10/7/85, E. P. Rake, Jr. to Harold Denton) and is applicable to Salem which has a large dry containment.
The conclusion of this report is that superheat will not have a significant impact on the Temperature-Pressure Profile inside the containment.
The Mainsteam Lines(MSL) exit the containment through the inboard and outboard penetration areas.
The MSL's then exit these penetration areas and run through an open area before entering the turbine building.
Refer to FSAR figure 3.6-11 and -15.
Outside containment the only safety related equipment which would be exposed to a MSLB environment is equipment in the inboard and outboard penetration areas required:
0 0
to initiate and perform MSL isolation, and to provide a SG pressure input to the SI and Reactor Trip logic.
MSIV closure as well as the SI initiation and reactor trip signals will occur long before SG tubes are uncovered on a MSLB (approximately 1.5 minutes into a MSLB); therefore, superheat will not have any effect on the initial closure of the MSIV or initiation of the SI and reactor trip signals.
In order to address the issue of superheat described in Information Notice 84-90 "Main Steam Line Break Effect on Environmental Qualification of Equipment,"
PSE&G has contracted Westinghouse to perform a subcompartment Pressure-Temperature analysis of the penetration areas.
Long term qualification of safety related equipment will be addressed as required based on the results of this analysis.
Since this is the only outstanding issue, PSE&G feels that approval of BIT elimination should not be predicated on the results of this analysis.
4
-~
DIVISION OF PWR LICENSING -
4 ENGINEERING BRANCH QUESTIONS (1)
- a.
State the value of re-evaporization used in the MSLB analysis.
Answer:
For large breaks 100 percent re-evaporization of the condensate is used.
No condensate re-evaporization is assumed for small steamline break analysis, a calculated fractional re-evaporization due to convective heat flux is used.
(1)
- b.
Provide justification for values larger than 8 percent.
Answer:
The Westinghouse containment model utilizes the analytical approaches~described in the following
References:
- 1.
Bordelon, F. M. and Murphy, E. T., Containment Pressure Analysis Code (COCO), "WCAP-8327, June 1974 (Proprietary) and WCAP-8326, June 1974 (Non-Proprietary).
- 2.
Letter to Mr. D. B. Vassallo, Chief, Light Water Reactor Projects Branch 6, USNRC, From Mr. C.
Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, dated March 17, 1976 (NS-CE-992).
- 3.
Letter to Mr. J. F. Stolz, Chief, Light Water Reactor Projects Branch 6, USNRC, From Mr. C.
Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, dated August 27, 1976 (NS-CE-1883).
- 4.
Hsieh, T. et al, "Environ.mental Qualification Instrument Transmitter Temperature Transient Analysis, "WCAP-8936, February 1977 (Proprietary) and WCAP-8937, February 1977 (Non-Proprietary).
5
(2)
Discuss briefly why a different number of blowdown cases and break sizes were evaluated in the revised containment temperature and pressure analysis for the proposed reduction in Boron concentration.
Answer:
The spectrum of blowdown cases and break sizes evaluated in the revised analysis represents current Westinghouse methodology.
The additional cases run in the revised analysis are due to the inclusion of entrainment in the analysis.
Specifically, the additional cases are the smallest break which results in moisture entrainment and the largest break resulting in a dry steam blowdown.
The additional cases bound the occurance of entrainment.
6
I PLANT SYSTEMS BRANCH QUESTIONS (1)
With respect to the analysis presented in reference 1, supporting removal of the Boron Injection Tank (BIT), the mass and energy releases following a main steam line break were calculated using the Westinghouse computer code LOFTRAN.
The version of the LOFTRAN code used does not account for heat transfer to the steam during steam generator tube bundle uncovery.
This heat transfer effect will lead to superheating of the steam being produced in the steam generator, and the exiting steam temperature may be substantially above the qualification temperature of equipment.
The phenomenon of steam superheating and its effect on containment response has not been addressed in the main steam line break analyses that have been submitted.
Provide a blowdown analysis which includes steam superheating effects.
Discuss and justify the assumptions and conservations made relative to containment response analysis.
(2)
The analysis.. pte~ented in reference 1 did not address the effect of BIT removal on safety related equipment outside containment.
Provide an analysis of the environmental conditions outside containment which includes the steam superheating effect discussed in the preceeding question.
Discuss and justify the adequacy of the qualification of all safety related equipment outside containment that may be exposed to the superheated steam blowdown.
References:
- 1.
Letter from Corbin McNeill, Jr. (Public Service Electric and Gas Company) to Steven Varga (NRC), dated October 25, 1985.
Answer to (1) and (2):
Refer to our reply to Reactor Systems Branch question No.
- 6.
7