ML20028E870
| ML20028E870 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/25/1983 |
| From: | Baxter T SACRAMENTO MUNICIPAL UTILITY DISTRICT, SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Buck J, Kohl C, Rosenthal A NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| Shared Package | |
| ML20028E871 | List: |
| References | |
| TAC-44673, TAC-48907, NUDOCS 8301280251 | |
| Download: ML20028E870 (2) | |
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00CKETED U$RC SHAW, PITTMAN, PoTTs & TROWBRIDGE'
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January 25, 1983 822-1090 Alan S. Rosenthal, Esquire Dr. John H. Buck Chairman Atomic Safety and Licensing Appeal Atomic Safety and Licensing Appeal Board Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Washington, D.C.
20555 Christine N. Kohl, Esquire Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 In the Matter of Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station)
Docket No. 50-312 Administrative Judges Rosenthal, Buck and Kohl:
Please find enclosed copies of two letters from licensee Sacramento Municipal Utility District to the NRC Staff, dated January 7 and 14, 1983, which transmit information relevant to the matters under review by this board.
Respectfully submitted, 8301280251 830125 PDR ADOCK 05000312 Thomas A. Baxter G
PDR Counsel for Licensee t
TAB:jah Enclosures cc:
Service List attached
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DOCKETED UNITED STATES OF AMERICA UShRC NUCLEAR REGULATORY COMMISSION
'83 sy 27 p{ 99 BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
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SACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docket No. 50-312
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(Rancho Seco Nuclear Generating
)
Station)
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SERVICE LIST Alan S. Rosenthal, Esquire David S. Kaplan, Esquire chairman Secretary and General Counsel Atcmic Safety and Licensing Appeal Sacramento Municipal Utility District Board P.O. Box 15830 U.S. Nuclear Regulatory Conmission Sacramento, California 95813 Washington, D.C.
20555 Roy P. Iessy, Esquire Dr. John H. Buck Office of the Executive legal Director Atanic Safety and Licensing Appeal U.S. Nuclear Regulatory Ccmnission Board Washington, D.C.
zusd5 U.S. Nuclear Regulatory Comnission Washington, D.C.
20555 Christopher Ellison, Esquire California Energy Otmnission Christine N. Kohl, Esquire 1111 Howe Avenue Atanic Safety and Licensing Appeal Sacramento, California 95825 Board U.S. Nuclear Regulatory Comnission Lawrence Coe Lanpher, Esquire Washington, D.C.
20555 Hill, Christopher and Phillips, P.C.
1900 M Street, N.W.
Elizabeth S. Bowers, Esquire Washington, D.C.
20036 Chairman Atcznic Esfety and Licensing Board Docketing and Service Section U.S. Nuclear Regulatory Ccmnission Office of the Secretary Washington, D.C.
20555 U.S. Nuclear Regulatory Ocmnission Washington, D.C.
20555 Dr. Richard F. Cole Atanic Safety and Licensing Board Panel U.S. Nuclear Regulatory Comnission Washington, D.C.
20555 Mr. Frederick J. Shan Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Ccmnission Washington, D.C.
20555
.SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S street, Box 15830, Sacramento, Califomia 95813; (916) 452-3211 i
January 7,1983 l
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DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4 US NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT NO 1 l
HPI N0ZZLE INSPECTIONS 1
c Your letter of January 5,1983, requests the bases for the reasons cited in our December 14, 1982 filing with the Atomic Safety and Licensing Appeal Board, regarding inspection frequency for the unmodified HPI nozzles at Rancho Seco Unit No. 1.
As stated in that filing, we do not feel there is a safety reason which warrants the once-per-cycle inspection proposed by the Appeal Board.
1 A Task Force was formed by the B&W Owners ~to study the loose thermal sleeve and nozzle cracking problems.
This task force had a meeting with your staff on December 16, 1982, to present the status of their effUrt.
The Task Force feels, and we agree, that there is no reason to expect a thermal sleeve to become loose which has remained tightly in position to date. We do feel, however, as does the Task Force, that an augmented, inservice inspection i
program is justified and'have agreed with their preliminary conclusion of a once-in-every-five-cyc?e radiographic insptction. We feel this is a l
conservative schedule since, as stated in our filing, we feel that a loose or missing sleeve does not present a safety concern.
The only through-wall cracking tu occur as a result of a loose sleeve was at the Crystal River Three plant which had a unique configuration with a valve body welded directly to the nozzle safe-end.
The Task Force hcs also made a preliminary recom-mendation for a once-per-cycle examination for the next five cycles before i
l changing to the once-every-five-cycle frequency. We feel there are no unique j
operating parameters during the next five cycles to justify waiting until.
I that time to change to the once-every-five-cycle frequency. We, therefore, feel the gradual change in frequency as discussed in our filing to be justi-fied at this time. As also stated in our filing, these nozzles are not used for system makeup (a routine, continuous function), and have experienced over six years of plant orgration to date.
I want to emphasize that the Task Force recomendation is ' preliminary and would not become an Owners' Group position until finalized and reveiwed,.by N lIUS /}
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JOHN F STOLZ age 2 January 7, 1983 each utility's management so that ALARA and cost considerations can be included in the decision., Since these considerations are not included in the Task Force effort, we feel that' their preliminary findings are conserv-ative; however, we agree with their once-every-five-cycle recommendation.
This is why we commented on the Appeal Board's tentative conclusion of a once-per-cycle inspection frequency.
ohn J. Mattimoe General Manager l
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( )SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830, Sacramento, C.S*ornia 95813; (916) 452-3211 Janua'ry 14, 1983 DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4 U S NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT N0 1 UPGRADED AUXILIARY FEEDWATER SYSTEM-NUREG-0737 ITEM II.E.1.1 Your letter of December 8,1982, requested additional information for your
- se in evaluating our upgraded auxiliary feedwater system. The information requested is attached to this letter with item numbers corresponding to those in Enclosure 1 to your letter.
If we can provide any additional infor-mation to assist in your review, please advise.
1Lb A
John J. Mattimoe General Manager Enclosure
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED 12/08/82 1.
Provide the seismic and quality group classification of all the components and piping for the AFWS design shown in Figure 3.1-1 t
of the proposed upgrade design submittal.
(Part I.B.1.C)
RESPONSE
3 In letters dated July 17, 1981 and July 6,1982, the District provided a discussion of the seismic qualification of the existing auxiliary feedwater system.
Since the upgraded system is considered to be a safety grade system, all components being added are both Seismic Cate-gory I and Quality Class I.
Copies of the above letters are attached for your reference.
2.
Verify that all AFWS essential components are located in seismic Category I structures or are provided with protection against failure of nonseismic Category I structures.
(Part I.B.2.a) 1
RESPONSE
All essential components of the auxiliary feedwater system are located either outside or in Seismic Category I Structures and are protected against the failure of any Nonseismic Category I Structures.
3.
Identify all AFWS components which are not protected from tornadoes, floods, external missiles and internally generated missiles.
(Part I.B.2.b)
RESPONSE
Tornado and tornado generated missiles are not a part of the design basis for the Rancho Seco plant.
All safety related components are however protected against a 101 mile per hour wind or missiles this wind could generate.
In addition, all auxiliary feedwater system components, as are all safety related components, are protected from floods and internally-generated missiles.
4.
Verify that all essential AFWS components are protected against the effects of pipe break in high and moderate energy lines. These include i
the effects of pipe wnip, jet impingement and internal flooding.
(Part I.B.2.c)
RESPONSE
Since the present auxiliary feedwater system is mechanically designed to Class I standards, it has been included in analyses performed for f
high-energy and moderate-energy line breaks. This study is not yet completed for the upgraded system including controls; however, this study will be completed before components will be installed to insure they are protected against these phenomena.
The effects of pipe whip, jet impingement and internal flooding are a part of'the District's high-energy and moderate-energy line break studies.
l (2) 5.
Document the provisions made for supplying safety-grade control air and verify that the air accumulators have sufficient capacity to supply control air for two hours following a loss of offsite power or loss of all AC power.
(Parts I.B.2.d, I.B.2.f)
RESPONSE
The Rancho Seco Control Air System is not safety grade, however for the valves in question, a backup nitrogen air supply is being installed.
This backup system will consist of safety grade nitrogen bottles and will have sufficient capacity to supply control air for two hours.
6.
Verify that AFW pump damage will not occur before water flow is initiated if the primary water source is not available on demand.
(Part II.A.4)
RESPONSE
Since the primary water source for the auxiliary feedwater system is Class I, and the system is entirely passive in nature, it is inconceivable that it would not be availabe on demand. The system is designed so water is present at the pumps themselves at all times. Condensate storage tank level is alarmed in the control room to provide the operator time to realign the source of water should that become necessary.
7.
Verify that, for the short-term, procedure A.51 requires an operator to be stationed at the AFWS flow control valves following the loss of all AC power, and that adequate lighting and communications with the control room are available to assure AFW operation for two hours independent of all AC power.
(Part II. A.5)
RESPONSE
In letters dated November 24, 1982 and December 15, 1982, the District has provided its position regarding procedures for a loss of all AC power.
We feel that specific procedures to handle the hypothetical case of a loss of all offsite and onsite AC power is unnecessary and undesirable.
We have committed, however, to conduct operator training in the use of -
existing procedures to cover this scenario.
In addition, lighting and the communications are available at the control valves to allow the specified operator actions.
Copies of the above letters are attached for your refer-ence.
8.
Verify that the condensate storage tank (CST) is protected against tornadoes and tornado missiles, or alternatively commit to provide another method of pump protection in the event of catastrophic damage to the CST.
(Part II.B.4)
RESPONSE
The Rancho Seco Design Basis does not include tornadoes or tornado generated missiles.
Consequently, the condensate storage tank is not protected against these; however, it is protected against a 101 mile per hour wind and the missiles that this wind could generate, since this
(3) is the design basis for safety related equipment.
Therefore, another method of protection is not necessary, nor would it be consistent with the Rancho Seco Design Criteria.
9.
Propose methods to increase AFWS reliability into the high range, using time to steam generator dryout as the upper time limit for necessary manual or automatic actions.
(Part II.C)
RESPONSE
Prior to responding directly to the above request, it should be noted that Section II.C of the Staff's " Status Report - Rancho Seco - Auxiliary Feedwater System" attached to the subject December 8,1982 letter implies that the reliability study submitted by the District only provided values of Rancho Seco AFW unavailability "considering that 20 minutes would be available for operator action."
In fact, system unavailabilities were calculated for two cases, operator intervention within twenty minutes and fully automatic initiation.
The " twenty minute" case does allow for operator action to correct system failures and corresponds to a mission success criterion of prevention of core damage. The District continues to consider that this is the appropriate acceptance criterion for the Rancho Seco AFW.
The " fully automatic" case calculations, however, can be correlated with a mission success criterion of prevention of steam generator dryout.
That is, for a plant such as Rancho Seco which has an anticipstory reactor trip on loss of main feedwater, automatic actuation of the AFW will enable auxiliary feedwater to be delivered to an intact steam generator prior to d ryout.
Relative to automatic AFW actuation and the above request regarding methods to increase the reliability of the Rancho Seco AFW, several possible changes have been identified which could decrease failure probability per demand by a factor of two to three. These are:
e Install a third, diesel-motor-driven, pump.
e Install a third, diesel generator backed, electric motor driven pump.
The District is of the position, however, that the benefit of any of these changes is not significant relative to the cost of such a modification and its impact on overall plant safety. Major improvements in AFW reliability are currently being made - for example, providing for automatic loading of Pump P-319 on to a diesel generator backed bus - and the system has a high degree of reliability when the appropriate acceptance criterion of prevention of core damage is utilized.
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(4) 10.
Provide the results of flow testing to verify that 780 gpm can be delivered within 40 seconds to the steam generator at 1150 psi. Alter-natively, verify the adequacy of 760 gpm supplied to the steam generator within 50 seconds.
RESPONSE
As discussed in die Staff's " Status Report - Rancho Seco Auxiliary Feedwater System" attached to the subjcct December 8,1982 letter, the original design basis for the Rancho Seco Auxiliary Feedwater System was delivery of 780 gpm within 40 second's to a steam generator at 1150 psig. As part of the current AFW upgrade effort the design basis event (a loss of main feedwater) was initially reanalyzed for delivery of 760 gpm within 50 seconds.
This loss of main feedwater transient analysis was performed by B&W using the TRAP computer code.
Major assumptions and input parameters are identified in Table 10-1 attached.
The sequence of events for the transient is summarized in Table 10-2.
The results of the analysis are given in Table 10-3 and the following acceptance criteria were met:
e RCS Peak Pressure Less.Than 110% Design Pressure e DNBR Greater Than Minimum Allowable e Site Boundary Doses Less Than Allowable In addition, the following desirable conditions were confirmed:
e Prevention of Solid Pressurizer e No Loss of Subcooled Margin e No Quench Tank. Rupture Disk Blow Out Major conservatisms in the analysis were:
o One AFW Pump / Train Operable e Decay Heat is 1.2 Times ANS e No Consideration fo'r Steam Relief Via Turbine Bypass System o No Anticipatory Reactor Trip on Loss of Main Feedwater Subsequent to the above analysis, further evaluation has been performed for delivery of 760 gpm of auxiliary feedwater within 70 seconds. This effort determined that the criteria and conditions identified above continued to be met, and it is the District's current intent to use 760 gpm/70 sec as the design specification for the upgraded Rancho Seco AFW.
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TABLE 10-1 RANCHO SEC0 NUCLEAR GENERATING STATION, UNIT 1 LOSS OF MAIN FEEDWATER TRANSIENT ANALYSIS Major Assumptions and Input Parameters A.
Thermal Hydraulics Parameter Value 1.
Power level (102%), MWt 2827 2 '. RC Pump Heat, MWt 16 3.
Primary Flow rate 8
Total, lbm/hr 1.378 x 108 Core Heat Transfer, lbm/hr 1.295 x 106 4.
Secondary Flow rate, lbg/hr 6.12 X 10 5.
Feedwater Ter.1peratuge, F
470 6.
Steam Temperature, F 570 7.
SG Outlet Pressure, psia 925 8.
SG Inventory, lbm 39,600 9.
Initial Pressurizer Pressure, psia 2170 10.
Initial Pressurizer Inventory, ft3 800 AFWTemperature,g(minimum)
AFW Flow rate, gp 760 11.
F 120 12.
B.
Kinetics Parameter Value 1.
Doppler Coefficient, (ak/k)/ F
-1.57 x 10-3 2.
Moderator Coefficient, (ak/k)/ F 0
3.
Boron Worth, ppm /(%Ak/k) 117 4.
Shutdown Worth, (%Ak/k) 2.6 5.
Boron Concentration ppm 1154 C.
Trip and Setpoint Times Parameter Value 1.
Main Feedwater Ramp Down, (100-0% flow), sec 5
2.
AFW Actuation to Full Flow, sec Motor Driven, sec 7 (50 w/o onsite power)
Turbine Driven, sec 50 3.
High Pressure Trip Setpoint, psia 2400 4.
Trip Delay Time, sec 0.4 5.
Turbine Trip After Reactor Trip, sec 0.5
TABLE 10-1,(C:ntinued) 1-D.
Valve Data Parameter Valve 1.
PORVObm/hr) 0 5
0 - 2270 psia 1.0 x 10
> 2270 psia 2.
Pressurizer Safety Valves (lbm/hr) 0 - 2515 psia 0
5
> 2515 psia 6.0 x 10 3.
Main Steam Safety Valves (lbm/hr) 0 - 1065 psia 0
6-1065 - 1085 psia 1.7 x 106 1085 - 1105 psia 3.4 x 106 1105 - 1117.5 psia 5.1 x 106
> 1117.5 psia 6.8 x 10 1
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TABLE 10-2 i
RANCHO SEC0 NUCLEAR GENERATING STATION, UNIT 1 LOSS OF MAIN FEEDWATER TRANSIENT ANALYSIS Sequence of Events Event Time (Seconds)
Steady State Full Power (102%)
0.0 Loss Of Main Feedwater 2.0 Zero Flow of Main Feedwater 7.0 i
OTSG Low Level Signal to AFW 16.0 High Reactor Coolant P'ressure Trip Signal 16.2 Reactor Trip 16.6 i
Turbine Trip 16.7 Pressurizer Safety Valves Open s 18.0 Main Steam Safety Valves Open s 22.0 AFW Pump Initiates 43.0 h
m.
1 TABLE 10-3 RANCHO SEC0 NUCLEAR GENERATING STATION, UNIT 1 LOSS OF MAIN FEEDWATER TRANSIENT ANALYSIS-Results Parameter Value Time to Match AFW Heat Removal Capability to Decay s 250 sec Heat 3
Subcooled Margin at 250 sec 2.6 x 10 lbm a
flass Relieved Through a
Pressurizer Valves 57 F Mass Relieved to Atmosphere 4
(0 - 250 sec) 3.1 x 10 lbm 4
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) S~ ~Ul rinr-SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s street, ses 15830. Sacramento. Califomia 95813;(916) 452 3211 July 17, 1981 lu Ak lo-N-CI b)Cvi+oNPS/ W DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION DARRELL G EISENHUT DIRECTOR DIVISION OF LICENSING U S NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 DOCKET 50-312
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RANCHO SECO NUCLEAR GENERATING STATION UNIT NO 1 AUXILIARY FEEDWATER SYSTEM SEISMIC QUALIFICATION PleaserefertoourresponsedatedJuly13,OV-01<.
1981, to your February 10,.
1981 Generic Letter No. 81-14. Our July 13, 1981 response was not notarized. This letter repeats our earlier response and is properly notarized in accordance with the requirements of the Generic Letter.
The Sacramento Municipal Utility District has reviewed your letter of February 10, 1981 (Generic Letter No. 81-14), requesting information on the seismic qualification of the Rancho Seco Unit 1 auxiliary feed-water system.
The entire Auxiliary Feedwater (A W) System at Rancho Seco was designed and constructed to Seismic Category I requirements. The analytical techniques, testing, evaluation methods and acceptance criteria used in the design of the A W System are consistent with the methods used for other safety-grade systems at Rancho Seco Unit 1.
The AW System has been included Within the scope of seismic related Bulletins 79-02, 79-04, 79-14, 80-11, and IE Information Notice 80-21.
The analysis methodology for Seismic Class I systems such as the A W System uses the " response spectrum" approach. The response spectrum approach uses the natural period, mode, shape and dampening factors that are characteristic of the system and equipment, to determine the dynamic loads that would result from a seismic event.
General primary stresses, including those resulting from gravity loads, operating loads y 181. $
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DARRELL G EISENHUT Page 2 July 17, 1981 and operating temperatures, are combined with the seismic stresses resulting from the amplified earthquake response spectrum.
The combination of stress is required to be below allowable stress set forth in the appropriate codes and standards, such as ASME Section III. A more detailed description of the methodology is given in Appendix SB of the FSAR.
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. Mattimoe Assistant General Manager and Chief Engineer Sworn to and subscribed before me this 17th day of July, 1981.
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