ML20028E324
| ML20028E324 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/14/1983 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.1, TASK-TM TAC-44673, TAC-48907, NUDOCS 8301210223 | |
| Download: ML20028E324 (9) | |
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SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830, Sacrarnento, California 95813; (916) 452-3211 January 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 use 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.
h John J. Mattimoe General Manager Enclosure Ad'f4 8301210223 830114 PDR ADOCK 05000312 P
PDR AN ELECTRIC SYSTEM SERVING MORE THAN 600,000 IN THE NfARI 0F Calif 0PNIA
i RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED 12/08/82 i
I 1.
Provide the seismic and quality group classification of all the components and piping for the AFWS design shown in Figure 3.1-1 of the proposed upgrade design submittal.
(Part I.B.1.C) 4
RESPONSE
In letters dated July 17, 1981 and July 6,1982, the District provided a discussion of the seismic qualifict. tion 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)
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
J Tornado and tornado generated missiles are not a part of the design basis for the Rancho Seco plant. All safety related components are houever 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 the effects of pipe whip, 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 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.
(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 l
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 followir.g 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.
i We have committed, however, to conduct operator training in the use of existing procedures to cover this rcenario.
In addition, lighting and the communications are available at the control valves to allow the specified l
operator actions.
Copies of the above letters are attached for your refer-ence.
l 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 j
l 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
4 1
(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) l
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 anticipatory 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 ryou t.
Relative to automatic AFW actuation and the above request regarding methods to increase the reliability of the Rancho Seco AFW, several possible changes i
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.
Install a third, diesel generator backed, electric motor driven pump.
e I
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 i
high degree of reliability when the appropriate acceptance criterion of l
prevention of core damage is utilized.
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- 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 gom supplied to the steam generator within 50 seconds.
RESPONSE
As discusse.1 in the Staff's " Status Report - Rancho Seco Auxiliary Feedwater System" attached to the subject December 8,1982 letter, the original design basis for the Rancho Seco Auxiliary Feedwater System was delivery of 780 gpm within 40 seconds to a steam generator at 1150 psig. As part uf 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:
o Prevention of Solid Pressurizer e No Loss of Subcooled Margin e No Quench Tank Rupture Disk Blow Out Major conservatisms in the analysis were:
e One AFW Pump / Train Operable e Decay Heat is 1.2 Times ANS i
e No Consideration for Steam Relief Via Turbine Bypass System l
e 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 l
continued to be met, and it is the District's current intent to use l
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 10 4
6 4.
Secondary Flow rate, lbg/hr 6.12 X 10 Feedwater Temperatu F
470 5.
Steam Temperature, ge, F
570 6.
7.
SG Outlet Pressure, psia 925 8.
SG Inventory, lbm 39,600 9.
Initial Pressurizer Pressure, psia 2170 10.
Initial Pressurizer Inventory, ft3 800
- 11. AFW Flow rate, gpg (minimum) 760
- 12. AFW Temperature, F
120 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.
flain Feedwater Ramp Down, (100-0% flow), sec 5
2.
AFW Actuation to Full Flow, sec tiotor 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 (Continued)
D.
Valve Data Parameter Valve 1.
PORV (lbm/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 TABLE 10-2 RANCHO SECO NUCLEAR-GENERATING STATION, UNIT 1 LOSS OF MAIN FEEDWATER TRANSIENT ANALYSIS Sequence of Events Event Time (Seconds) i Steady State Full Power (102%)
0.0 I
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 Pressure Trip Signal 16.2 Reactor Trip 16.6 Turbine Trip 16.7
+ 18.0 Pressurizer Safety Valves Open Main Steam Safety Valves Open s 22.0-AFW Pump Initiates 43.0 i
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TABLE 10-3 RANCHO SEC0 NUCLEAR GENERATING STATION, UNIT 1 LOSS OF MAIN FEEDWATER TRANSIENT ANALYSIS Resul ts 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 flass Relieved Through Pressurizer Valves 57 F Mass Relieved to Atmosphere 4
(0 - 250 sec) 3.1 x 10 lbm l
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- -h.ED SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street. Box 15830. Sacramento California 95813;(916) 452 3211 July 17, 1981 M4 & /o-N-f/
WCu>foNPs/ W DIRECTOR OF NUCLEAR REACTOR REGULATION A*1TENTION DARRELL G EISENHUT DIRECTOR DIVISION OF LICENSING U S NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NO 1 AUXILIARY FEEDWATER SYSTEM SEISMIC QUALIFICATION 0%
O t<-
Please refer to our response dated July 13, 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 (AW) System at Rancho Seco was designed and constructed to Seismic Category 1 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 A W 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 1 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 p LCt. 5 1 ~F2._
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I DARRELL G EISENIIUT Page 2 July 17, 1981 and operating temperatures, are combined with the seismic stresses resulting from the amplified carthquake 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 5B of the FSAR.
,/ Jt417 John Mattimoe Assistant General Manager and Chief Engineer Sworn to and subscribed before me this 17th day of July, 1981.
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July 6, 1982 DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4 US NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20555 DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NO 1 SEISMIC QUALIFICATION OF THE AUXILIARY FEEDWATER SYSTEM - GENERIC LETTER 81-14 Your letter of May 7,1982 requested additional information concerning seismic qualifications of the Auxiliary Feedwater System at Rancho Seco Unit No.1. The attached report describes these seismic qualifiactions for the items questioned.
We trust that the information provided will be sufficient to allow you to complete your review.
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. ohn' J. Mattimoe jg Assistant General Manager i
and Chief Engineer EN Attachment bc: J. J. Mattimoe D. G. Raasch (2)
R. A. Dietcrich R. J. Rodriguez R. W. Colombo L. G. Schwieger (2) 4th. Floor Files 3rd. Floor Files Gordon Deppe (Bechtel - RS, MS-208)
Paul Goodman (Bechtel - Norwalk)
J. V. McColligan Tom Baxter D. Holt (B&W Lynchburg)
Kris Valvekar (Bechtel - Norwalk)
J. J. Field L. J. Marino (Bechtel - RS, MS-208)
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DESCRIPTION The attached system diagram shows the Auxiliary Feedwater System (AR1S) and its seismic boundaries of Rancho Seco nuclear Power Project. The AR1 pumps take suction from th e co nd en sa te storage tank which is a seismic Category I component. One pump is driven by a Class lE electric motor, whil e the other is equipped with a dual drive, operable by an independent Class lE electric motor or a steam turbine. Each pump motor is powered by independent circuit trains and Class 1E power supplies.
The steam supply to the turbine is provided by connections from the main steam lines originating from both Once Through Steam Generator s (OTSG 's).
Motor operated and manual valves are arranged such that steam from either OTSG can drive the arf pump turbine. Motor and air operated valves are arranged such that each pump can provide auxiliary feedwater flow to either of the two OTSG's. The valves and their motor operators are all qualified to Seismic Category I.
IE BULLETINS AR1S piping was originally designed to Seismic Category I criteria described in the project FSAR. NRC IE Eulletins 79-02 and 79-14 required that the District walkdown, verify and re-evaluate all Seismic Category I piping and supports.
Bulletin 79-02 was concerned with concrete expansion anchors and the District tested some of the pipe supports constructed with base plates and anchor bolts. The results of this anchor bolt verification program were presented to NRC in a letter to R.
H.
Engelken, dated June 26, 1980.
NRC IE Bulletin 79-14 was concerned'with seismic qualification of as buil t piping configuration.
The District inspected and evaluated all Seismic Category I spiping, including the AR15 piping.
The results of the field verification' program were transmitted to NRC in a letter to R.
H.
Engelken dated August 1, 1980.
IE Bulletin 79-07 required that the District submit a review of the computer codes for inter and intra-modal summa tion used for j
piping seismic stress analysis. This res ponse to the IE Bulletin was transmitted to the NRC in a letter to Engelken dated April 24, 1979. The District's reponse to IE Bulletin 79-04 showed that re-evaluation 'of analytical weights for Velan swing check valves was required for only two valves. These two valves 9ere not in the l
AR1S. This r esponse was tr an smit ted to the NRC in a letter to l
Engelken dated April 26, 1979.
IE Bulletin 80-21 did not concern the District since no Valve parts cast by Malcom Foundry are in use or planned for use at Rancho Seco. This fact was documented in a letter to Engelken dated November 21, 1980.
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The District res ponse to IE Bulletin 80-11 is currently pending a dditional information and calcula tions from the District's architect / engineer. However, as noted in a previous reply (ref:
letter to Engelken, dated January 19, 1981) Rancho Seco does not have any masonry walls supporting Class I equipment. Furthermore, there are not any masonry walls supporting AR1S components.
PIPING AFWS piping including steam supply to the turbine, and Seismic Category I branch piping were analyzed as such up to seismic boundaries indicated on the diagram. Branch lines were included in the seismic analysis beyond the point of seismic /nonseismic interface to three or thogonal restraints, structural anchor or underground embedment. All seismic analyses are performed with the response spectrum method which utilizes the na tu ral period, mode shapes and appropriate damping factors for the Ani piping system.
For a more detailed description of techniques used in the analysis of Seismic Category I piping, please refer to the FSAR Appendices SA and 58. Presently, the District is modifying sections of the auxiliary feedwater line inside containment near the steam generators. This modification involves providing six inlet nozzles to the OTSG instead of one. The piping will be reanalyzed. Because of added flexibility of the inlet header, rigid supports and snubbers will be added to qualify the piping seismically.
EOUIPMENT Seismic stress calculations for the Ani pumps were prepared by B&W, Canada.
The pump motors were supplied and accepted for seismic application by Hitachi America, Ltd. The steam turbine used on the dual drive pump was supplied by Terry Turbine, Co. and the seismic certification was furnished via vendor calculations.
The condensate storage tank was supplied by Conseco and certified to Seismic I
by vendor calculations.
The vendor certified calculations associated with all of the aforementioned equipment were reviewed by the District's architegt/ engineer, Bechtel Power Corporation, prior to acceptance.
VALVES (Refer to attached piping diagram)
All valves and in-line equipment were ptoperly included in the lumped mass seismic model of the piping system. The stresses in the piping and acceleration on the valves were kept within allowable limits to guarantee the operability of valves for seismic condition.
The air operated j
control valves, (FV20526 & FV20527) however, were not qualified to l
any seismic requirement except the stresses in the body of the valve were kept below the allowable stress limits of the piping code.
There are bypasses provided around these air operated 2
1 1
control valves.
The bypass line valves are full
- size, motor operated, seismically qualified engineered safety feature system valves.
(SFV 20577 & SFV 20578) Motor operated valves (HV 31826
& HV 31827) located on the crossover piping which in ter co nne cts the "A"
train and "B"
train :ere supplied by Anchor Valve Company and underwent seismic testing by an independent testing laboratory, Henry Vogt Company.
The motor operated valve (FV 30801) controlling the steam inlet to the AFW pump turbine was supplied by the turbine vendor and designed to Seismic Category I.
Motor operated valves (HV20569 & HV 20596) are located on the crossover steam line inter conne cting the OTSG's and supply steam for the pump turbine. These valves were built by Velan Engineering and designed to Seismic Category I.
ELECTRICAL Electrical circuits have been designed to Seismic Category I for the following auxiliary feedwater system compunents:
AFW Pump P-318 AFW Pump P-319 Control Valve FV 20527 Control Valve FV 20528 Motor Operated Valve HV 31826 Motor Operated Valve HV 31827 Motor Operated Valve FV 30801 Motor Operated Valve HV 2056 9 Motor Operated Valve HV 20596 Motor Operated Valve SFV 20577 Motor Operated Valve SFV 20578 STRUCTURES Auxiliary feedwater piping and equipment are located either outdoors or inside the reactor bu ilding. The foundations of all l
the outdoor equipment have been designed to withstand the design l
basis earthquake. For a description of the seismic qualification l
of the reactor building and inside st r u ctu r es, refer to the FSAR l
Appendix SB.
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of 9 SACRAMENTO MUN!CIPAt. UTILITY DISTRICT O 6201 SMUD 3-EP (916) 452-3211 November 24, 1982 DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4 U S NUCLEAR REGULATORY COMMISSION WASHINGTON O C 20555 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT N0 1 EMERGENCY PROCEDURES AND TRAINING FOR STATION BLACK 0UT EVENTS DA C/4-In our June 1,1981 letter, we responded to your February 25, 1981 Generic letter 81-04 regarding the need for emergency procedures and training for station blackout events. Our position was that a specific procedure to handle the hypothetical case of loss of all offsite and onsite AC was unnecessary and undesirable. We further stated that we felt the District training in the areas pf. emergency systems, power distribution systems and heat removal by natural circulation using the steamdriven auxiliary feedwater pump was adequate to allow our operators to determine a proper. course of action. Consequently, the District pro-posed to take no additional action relative to the issue.
Recently, Tom Cogburn of the District Licensing staff was contacted by your Mr. Sydney Miner regarding resolution of this issue.
Mr. Miner explained that a commitment to prepare a written procedure was unnecessary; that a comitment to conduct operator training in the use of existing pro-cedures and tying these together to cover operator response to the loss of all AC scenario would be adequate to resolve the issue. The District is willing to comit to perform the necessary training to cover the loss of all AC event. We feel that much or all of the netessary material is presently covered in our operator training program. The District commits to review our current training program to determine additional material is needed and to add training material where needed. Our approach will be
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Director of Nuclear Reactor Regulation Attention John F Stolz Chief November 24., 1982 to cover the training for this event from a symptom oriented perspective to insure that our operators can recognize the appropriate symptoms (e.g. loss of RC flow, no HPI, no pressurizer heaters, no condenser vacuum, etc.) and understand the necessary actions and procedures to safely respond.
John J. Mattimoe General Manager JJM/TC/vb
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J. J. Mattimoe D. G. Raasch R. J. Rodriguez R. A. Dieterich J. V. Mc Colligan R. W. Colombo L. G. Schwieger (2)
Supervisors Tom Baxter Paul Goodman (Bechtel-Norwalk)
Harvey Canter (NRC - Ranch)
G. Deppe (Bechtel - Ranch)
Dave Holt (B&W - Lynchb'urg) 4th Floor Files M
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SACRAMENTO MUNICIPAL UTILITY oisTRICT [] 6201 s street. Box 15830. sacramento California 95813; (916) 452-1711
(,4(w December 15, 1982 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 STATI0'l UNIT NO 1 STATION BLACK 0UT TRAINING (GL 81-04)
OK-In our letter of November 24, 1982, we committed to provide training of our operators on response to a loss of all AC power as necessary to complement our existing operator training program. Since this event is not presently covered as a specific topic in either the requalification or licensing training program per se, we will formally cover it in the next licensed operator requalification cycle and in subsequent operator licensing training programs.
The next requalification cycle begins af ter the annual exams which are scheduled to be completed by January,1983 and continues throughout 1983.
L, N ohn J. Mattimoe bc:
J. J. Mattimoe D. G. Raasch General Manager R. J. Rodriguez R. A. Dieterich p
$g THC/ch J. V. McColligan R. W. Colombo L. G. Schwieger (2)
Supervisors Tom Baxter Paul Goodman G. Deppe Harvey Canter Dave Holt (B&W) 4th Floor Files 3rd Floor Files 44 V IU^h u
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