ML19248D439
| ML19248D439 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/03/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19248D437 | List: |
| References | |
| NUDOCS 7908160138 | |
| Download: ML19248D439 (14) | |
Text
f UNITED STATES b.
4 NUCLEAR REGULATORY COMMISSION 7 f G
WASHINGTON. D. C. 20555 5;
{
S-a
/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N05.38 AND 32 TO FACILITY LICENSE NOS. DPR-42 AND DPR-60 f40RTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 Introduction 1.0 Dy letter dated August 4, 1978, Northern States Power Company (fiSP) requested amendment of the Technical Specifications amended to Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant Unit Nos. I and 2 (PINGP).
The proposed amendments incorporate new definitions, limiting conditions for operation and surveillance requirements associated with the reactor cooling system overpressure protection system.
During our review of the proposed amendments we found that certain modifi-cations were necessary to meet our requirements.
These modifications were discussed with the NSP staff and they have agreed to the modifications.
By letter dated July 22,1977(I) NSP submitted a plant specific analysis to the Nuclear Regulatory Commission (NRC) in support of the proposed reactor vessel overpressure mitigating system for PINGP.
Ihis information ments other documentation submitted by NSP over the past 12 months. g e-The proposed lcw temperature overpressure protection and mitigating system includes sensors, actuating mechanisms, alarms and valves to prevent a reactor coolant system transient from exceeding the pressure-temperature limits included in the PINGP Technical Specifications as required by Arpendix G to 10 CFR Part 50 (Appenaix G).
In order to assure proper operation of the overpressure protection system, we requested and NSP has submitted proposed changes to the Technical Spec-ifications that are in accordance with the requirements presented in section 4.2 of this Safety Evaluation Report (SER).
Our review of all information submitted by NSP in support of the proposed overpressure mitigating system including the proposed Technical Specification, is completed and we have concluded that the system provides adequate protection from ove ?ressure transients and that the Technical Specifications proposed by NSF are in accordance with the requirements presented in section 4.2 of this SER.
(1)
Numbers in parenthesis refer to references which may be found on the last page of this SER.
7908160J U O
< n/
nn
/
OVU o,
2.0 Background
Over the last few years, incidents identified as pressure transients have scurred in pressurized water reactors.
The term " pressure transients,"
used in this report, refers to events dur% which the Appendix G temperature-pressure limits of the PIf1GP Tc
...ical Specifications, have been exceeded.
All of these incients occurred at relativ 41y low temperature (less than 200 F) where t'> reactor vessel material 1.cughness (resistance to brittle failure) was educed.
The " Technical Report on Reactor Vessel Pressure Transients", flUREG 0138 (7) summarizes the technical considerations relevant to this matter, discusses the safety concerns and existing safety margins of operating reactors, and describes the regulatory actions taken to reduce the likelihood of future events at operating reactors.
A brief discussion is precented here.
2.1 Vessel Characteristics _
Reactor vessels are constructed of high quality steel made to rigid specifi-cations, and fabricated and inspected in accordance with the time-proven rules of the ASME Boiler and Pressure Vessel Code.
Steels used are partic-ularly tough at reactor operating conditions.
However, if subjected to high pressures at lot temperatures, reactor vessel steels are less tough and could possibly fail in a brittle manner.
Therefore, power reactors have always operated with restrictions on the pressure allowed during the low temperature portions of startup and shutdown operations.
At operating temperatures, the pressure allowed by Appendix G limits is in excess of the setpoint of currently installed pressurizer code safety valves.
However, most operating pressurized water reactors (PWRs) did not have pressure relief devices to prevent a pressure transient during cold con-ditions from exceeding the lower Appendix G limits that correspond to the lower temperatures involved.
2.2 Regulatory Actions By letter dated August 13,197G,( } we requested that f1SP begin efforts to design and install plar.t systems to mitigate the consequences of pressure transients at low temperatures.
We also requested that operating procedures be examined and administrative changes be made to guard against initiating overpressure events.
It was our position that proper administrative controls are required to assure safe operation for the period of time prior to installation of the proposed overpressure mitigating hardware.
f1SP responded ( ' } with preliminary information describing interim measures to prevent these transients along with some discussions of proposed hardware.
The proposed hardware change was to install an optional low pressure actua-tion set point, n the pressurizer air operated relief valves.
tj / ';
- 0; NSP participatec member of a Westinghouse user's group which was formed to support the anab ;is effort required to verify the adequacy of the proposed system to prevent overpressure transients.
Using input data generated by the user's group, Westinghouse performed transient analyses which are used as the basis for plant specific analysis.
We requested additional information concerning the proposed procedural as provided the required responses (5,6) and the final hardware changes.
NSP plant speci fic analyses Il for PINGP.
2.3 Design Criteria Through this series of meetings and correspondence with PWR vendors and licensees, we developed a set of criteria for an acceptable overpressure mitigatinj system, which will prevent reactor vessel pressures in exce ;
of those allowed by Appendix G.
Specific criteria for systeu performa.
e are:
- 1) Operator Action:
ho credit can be taken for operator action for 10 miriutes af ter the operator is aware of a transient.
- 2) Singe, Failure:
The system must be designed to relieve the pressure transient given a single failure in addition to the failure that initi ated the pressure transient.
- 3) Testability:
The system must be testable on a periodic basis consis';ent wi th the sys tem's employment.
- 4) Seismic and ILLE 279 Criteria:
Ideally, the system should meet seismic Category I and It EE 279 criteria.
The basic objective is that the system should not be vulnerable to a common failure mode that would both initiate a pressure transient and disable the overpressure nitigating systen.
Events such as loss of instrumen+ air and loss of offsite power must be considered.
We also requested NSP to provide an alarm which monitors the position of the pressurizer relief valve isolation valves, in conjunction with the low pressure setpoint enabling switch, to assure that the overpressure mitigating system is properly aligned for shutdown conditions.
2.4 Design Basis Events The overpressure incidents that have occurred at PWRs to date have been the result of operator errors or equipment failures.
Two varieties of pressure transients can be identified:
a mass input type from charging pumps, safety injection pumps and safety injection accumulators; and a heat addition type which causes thermal expansion of the reactor coolant.
Sources of heat are steam generators (SG) and the reactor core.
6 l ')
b0 On Westinghouse plants, the most common cause of the overpressure transients of the mass input type has been initiated by isolation of the letdown path without turning off the charginy pump during low pressure operations. Other transients occur with lower frequency.
We selected those which result in the most rapid pressure increases for analysis.
The most limiting mass input transient we identified was inadvertent injection by the largest safety injection pump.
The most limiting thermal expansion transient we identified was the start of a reactor coolant pump wFen a temperature differential exists between the reactor coolant (cooler) secondary coolant (hotter) across the affected steam generator.
Based on the historical record of overpressure transients and the imposition of more effective administrative controls, we have concluded that the limiting events identified above form an acceptable basis for analyses of the proposed overpressure mitigating system.
3.0 System Description and Evaluation NSP adopted the " Reference Mitigating System" developed by Westinghouse and the user's group.
NSP proposed to modify the actuation circuitry of the existing air operated pressurizer relief valves to provide a low pressure set point during startup and shutdown conditions.
When the reactor vessel is at low temperatures, with the valves in low pressure set point mode a pressure transient will be terminated below the Appendix G limit by auto-matic opening of these relief valves at the lcw pressure set point.
A keylock switch is used to enable and disable the low pressure set point of each relief valve.
The overpressure mitigating low pressure set point is manually enabled at a RCS temperature of 275 F during plant cooldown and is manually disabled at the same temperature during plant heatup.
The system remains enabled whenever the temperature is below 275 F.
We find that the capacity of these air operated pressurizer relief valves is sufficient to prevent overpressure and that the proposed overpressure mitigating system is acceptable.
Discussion and evaluation of the details of the system proposed by NSP is presented below.
3.1 Air Supply The power operated relief valves (PORVs) are spring-loaded-closad.
Air is required to open the valves, which are supplied by a control air source. To assure ability of the valves to open upon loss of control air, a backup air supply is provided.
The backup air supply consists of a Seismic Category I passive air accumulator for each PORV.
Each accumulator contains 36 ft3 of air at a pressure of 80-100 psig and provides air for approximately 15 valve openings.
Check valves in the existing air lines prevent depressurizing the accumulators in the event of a ruptured air line.
50) 4 f9 Ot a
Pressure switches are insta' led to indicate loss of normal air supply and gauges are provided to veri fy operability of the backup supply.
We find the backup air supply to be acceptable.
3.2 Electrical Controls The proposed overall approach 'o e'!ininating overpressure events incorporates administrative, procedural and hardware controls with reliance upon the plant operator for the principal line of Jefense.
Preventive administra-tive/ procedural measures include:
(1) procedural precautions; (2) de-energi ing non-essential components during the cold shutdown mode of oper-ation; and (3) maintaining a non-water solid reactor coolant system condition whenever possible.
The basic design criteria which we apply in determining the adequacy of the electrical, instrumentation and control systems aspects of the low temperature overpressure protection system are:
1.
Operator Action - No credit can be taken for operator action until 10 minutes after the operator is aware, through an action alarm, that a pressure transient is in progress.
2.
Sinne Failure Criteria - The pressure protection system shall be designed to protect the reactor vessel given a single failure in addition to the failure thr.i. initiated the pressure transient.
3.
Testability - The system design shall include provisions for testing on a schedule consistent with the frequency that the system is relief upon for pressure protection.
4.
Seismic Design and IE_EE-279 Criteria - The pressure protection system shall satisfy both seismic Category I and IEEE-279 criteria.
In addition to complying with the above criteria, NSP has agreed to provide an interlock for the Enable / Disable switch with the position of the isolation valves associated with the POP.V.
This design feature will ensure a complete pathway from the pressurizer to a quench tank for pressure relief.
3.2.1 System Electrical Description The overpressure protection system has been designed to comi.ly with both seismic Category I and IEEE Std.279-1971 criteria with the ea:ception of the position alarm provided on the Enable / Disable switch.
Sin e this alarm function is similar to that provided by the Bypass and In,perable Status Panels, it likewise is not required to satisfy the seismic design requirement.
'Gb
^j;
+
The proposed overpressure protection system is designed with two independent channels and full testability.
Moreover, the installed pressure and temper-ature instrumentation at PINGP will provide a continuous and permanent record over the full range ot L;th pressure and temperature.
A keylock switch in each channel is used to manually enable or disable the low pressure set point of each PORV relief valve.
The overpressure protection system's low pressure setpoint is enabled at a reactor coolant system (RCS) temperature of 275 F during plant cooldown and is disabled at the same temperature during plant heatup.
The system remains enabled whenever the RCS temperature is below 275"F with the Enable / Disable switch awarm inter-locked with the position of the two PORV/ isolation valves.
We find these design features to be acceptable.
3.2.2 Electrical Equipment Modif' cation The following equipment will be added to each instrument loop for the overpressure protection system:
a.
One dual bistable with an alarm output and a valve actuat on output i
will be counted in spare locations in the RCS analog racks including the test jacks and test panels.
b.
Three control relays which provide instrument interface and appropriate train isolation will be mounted in sp;re locations in the relay room racks.
c.
One test jack for calibration signal input.
d.
One Enable / Disable switch per channel will be mounted on a panel in the control room.
e.
One test panel with a histable test switch and output indicating lights.
A spare point on an existing control room annunciator panel will be used for the alarm.
With the exception of the position alarm provided on the Enable / Disable switch, all electrical equipment added to each instrument loop will be protection grade and meet IEEE Std.279-1971 criteria.
All additional equip ont will be installed in accordance with the as-built design criteria for protection grade equipment.
We find these electrical modifications acceptable.
3.2.3 Al_a rm Sys tem _ Design In accordance with our position on the isolation alarm, NSP has added a design in the system which interlocks the position of the two PORV isolation
,nG V. l 'i valves with-the Enable / Disable switch alarm.
This feature ensures a complete pathway from the pressurizer to a quench tank for pressure relief.
The overpressure system annunciator logic design provides alarms in the control room if:
(1) the systen is enabled and a high reactor coolant system pressure exists; (2) the system is enabled and a PORV isolation motor operated valve (MOV) is closed; (3) the air supply to the PORV dcCumulators is not available; or (4) the system is disabled.
3.2.4 PORV Backup Air Supply Each PORV will have a backup air supply consisting of a seismic Category I passive air accumulator.
Check valves to prevent accumulator depressuri-zatior, in the event of a ruptured air line and pressure switches to indi-cate lcss of air and gauges will be installed to verify backup air supply opera bi li ty. We find this design acceptable.
3.2.5 Inadvertent Operation of Components Our position with regar] to the inadvertent operation of safety injection system (SIS) component, during cold shutdown operations requires the deenergizing of SIS ptnps and closure of safety injection
/SI) header /
discharge valves.
NSP has agreed to deenergize high pressure safety injection (HPSI) pumps and valves and other valves (except for operations allowed by the Technical Specifications) as denoted in the following list:
a.
b.
HPSI Injection Valve : 8306A d.
HPSI Injection Valve d 8806B; e.
Boron Discharge Valve s 8809C; f.
Accumulator Isolation Valve e 8800A; and g.
Accumulator Isolation Valve d 88008.
All of the valves w 11 be placed in the closed position prior to deenergizing them.
The SI pump and valve breakers are located in the Auxiliary Building and the breakers are controlled from the control room.
Position indication at the control switch for all valves which have their breakers in the OFF position will be lost. We find thi to be acceptable because position indication is available to control room operators from the safety injection monitor panels.
Required valve positions, valve breaker positions and pump motor breaker positions are also specified in the plant startup and shutdown procedures and require operator sign off prior to each step during startup or shutdown.
The procedures for starting (jogging) a reactor coolant pump will provide for the RCS AT across the affected steam generator to be less than 500F when the HCS is in a water,olid condition.
This limitation will not be applied when a steam bubble exists in the pressurizer.
n i
,,b/Y
'00 3.2.6 High Pressure Alarm Our position requires that a high pressure alarm for use during low RLS temperature be incorporated such that the operator's attention will be attracted to a transient in progress.
NSP has agreed to install a permanent computer high pressure alarm to satis fy this requirement. When the RCS temperature is ler.s than 2750F, the computer will scan RCS pressure once everv minute and, fo-a pressure above the setpoint, continue to alarm at this interval until the situation is corrected.
The temperature input is received from a cold leg reactor thermal detector (RTD) and the pressure input is received from a wide range pressure instrument.
The permanent alarm will be iocated on the control board in the control room. We find this design feature acceptable.
3.2.7 System Testability Our position with regard to testability is that the system be tested prior to any reliance upon it for overpressure protection.
NSP has agreed to perform a system functional test and a setpoint verification test prior to enabling the overpressure protection system during plant cooldown.
The system will be calibrated at refueling intervals.
The operability of the backup air supply system check valves will also be verified at refueling intervals.
We find this to be acceptable.
3.3 Testability The equipment design includes provision for testing.
NSP has stated that testing will be done on a schedule consistent with the frequency that the system is used for pressure protection.
A setpoint verification and functional test of the system will be performed prior to enabling the system during plant cooldown.
The system will be calibrated at refueling intervals and the operability of the check valves in the backup a ir supply system will also be verified at refueling intervals.
We find the type and frequency of system testing to be acceptable.
3.4 Appendix G The Appendix G curve submitted by NSP for purposes o f overpressure transient analysis is based on ten effective full power years of irradiation.
This is based or, the zero degree heatup curve since most pressure transients occur during isothermal metal conditions.
Further, margins of 45 psig and 100F are included for possible instrument errors.
The Appendix G limit of 1000F according to this curv ' i, 632 psig. We find that use of this curve is acceptable as a basis for evaluation of the overpressure mitigating system performance.
l')
bO!
3.5 Set Point Analysis The one loop version of the LOFTRAN code (9) was used to perform the mass input analyses. The four loop version was used for the heat input analysis.
Both versions require some input modeling and initialization changes.
LOFTRAN is currently under review by the NRC staff and is judged to be an acceptable code for treating problems of this type.
The results of this analysis are provided in terms of PORV setpoint over-shoot.
The predicted maximum transient pressure is simply the sum of the overshoot magnitude and the setpoint magnitude.
The POR'l setpoint is established so that even with the setpoint overshoot the resultant pressure is still below that allowed by Appendix G limits discussed in 3.4 above.
NSP used the following PINGP characteristics in determining the pressure reached for the design basis pressure trancients:
SI Pump Flowrate @ 500 psig 83.4 lb/sec 3
RCS Volume 6,528 ft PORV Opening Time 3 sec 2
SG Heat Trans fer area 51,500 ft Relief Valve setpoint 500 psig The Unit 1 Appendix G curve for ten years irradiation was presented with the anal ysis.
This curve is conservative with respect to Unit 2.
Westingnouse identified certair assumptions and input parameters as conservative with respect to the analysis.
Some of these are listed here.
- 1) One PORV was assumed to fail.
- 2) A three-second opening time was assumed for the PORV.
- 3) The RCS pressure boundary was assumed to be rigid and water solid with respect to expansion.
- 4) Conserva+ive neat transfer coefficients were assumed for the steani generator.
3.5.1 Mass Input Case The inadvertent start up of a high precsure saTety injection purrp with the plant in a cold shutdown condition,as selected as the limiting mass input case.
zno sf; j00
10 -
Westinghouse provided NSP with a series of curves based on the LOFTRAN analysis of a generic plant design which indicates PORV setpoint overshoot for this transient system volume, relief valve opening time and relief valve setpoint.
These sensitivity analyses were then applied to the PINGP parameters to obtain a conservative ectinate of the PORV setpoint overshoot.
We find this method of analysis to be acceptable.
Using the Westinghouse methodology, the PINGP PORV setpoint overshoot was determined to be 126 psi.
With a relief valve setpoint of 500 psig, a final pressure of 625 psig would be reached for the worst case mass input transient.
Since the Ui,it 1 ten year Appendix G limit ppplicable to Unit 2 also) at temperatures above 1000F is above 625 psig, we have concluded that the system performance is acceptable with a 500 psig low pressure relief valve setpoint.
3.5.2 Hea t Input _ Ca s_e.
The limiting heat input case selected as inadverten' startuo of a reactor coolant pump with a primary to secor.uary temperature d. iferent W across the steam generator of 500F, and w th the plant in a water solid condition.
For the heat input case, Westinghocse provided the licensee with a series of curves based on the LOFTRAN analysis of a generic plant design to determine the PORV setpoint ovenn% as a function of RCS volume, steam generator overall heat trans fer capability and initial RCS temperature.
For this transient, the reference relief valve selected was assumed to have a total opening of three sc:onds from the instant the signal to open is received until the vait. reached the full open position.
The calculated peak pressure for the heat input transient for a fixed AT of 500F was dependent upon the initial RCS temperature as shown here:
RCS Temperature Maximum Pressure 100 531 140 561 180 596 250 651 In all these cases, for the given RCS temperature, the Appendix G limits were not exceeded.
We find that the analyses of the limiting mass input and heat input cases show a maximum pressure transient below that allowed by Appendix G limits and are therefore acceptable.
,jl 3.6 Impl ementa t io n_ Sc hedul e NSP has made all electrical modifications required for the long term overpressure mitigating systen on both units and the backup air suppl, has been installed for both units.
4.0 Administrative Control s To supplement the hardware modifications and to limit the magnitude of postulated pressure transients to within the bounds of the analyses provided by NSP, a defense in depth approach has bcen adopted using procedural and administrative controls.
Those specific conditions required to assure that the plant is operated within the bounds of the analysis are included in the proposed Technical Specifications.
4.1 Procedures A number of provisions for prevention of pressure transients are contained in the PINGP Operating procedures.
These include precautions in the RCP operating procedures regarding temperature gradients in the RCS and temperature difference between the RCS and secondary side inventory in the steam generator.
Secondary side steam generator temperatures are monitored locally using surface-mounted therroccuples on the outside of the steam generators.
RCS temperatures are required by resistance temperature dectectors (RTD) in the loops.
The procedures for startup (and jogging) an RCD require the temperature difference between reactor coolant and the steam generator secondary side inventory be less than 500F hefore starting a pump associated with that steam generator when the RCS is water solid.
This limitation is not applicable when there is a ste r-bubble in the pressurizer.
Also, shutdown procedures have been revised to include provisions for maintaining a steam bubble in the pressurizer during most cold shutdown conditions.
In addition when the RCS is completely depressurized, procedures require that the pressurizer be vented to the pressurizer relief tank.
Although the RCS will be in a water solid condition for some tir:e whenever the plant is cooled down for refueling or for RCS maintenance the period of water solid operations has been reduced to the maximum extent possible, compatible with availability considerations (the total period of water solid operatien for each cooldown/heatup cycle has been reduced to 0
approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />).
Both HPSI pumps are de-energized below 200 F to preven t inadvertent starts (except for operations allowed by the Technical S peci f ic a tio ns ).
Above 200*F, the maximum allowable pressure by Appendix G is 1134 psig.
The relief capacity of one pressurizcr PORV is sufficient to relieve the combined flowrate from both HPSI pumps.
Thus, it is acceptable to have both HP5I pumps on line above 200 F.
4, 1
We find that the procedural and administrative controls described are acceptable, however, we have also required that certain procedural and administrative controls be included in the proposed Technical Specifications.
NSP has submitted Technical Specifications in accordance with the guidance in the following section.
4.2 Technical Speci fications To assure operation of the overpressure mitigating system, the licensee has submitted, for stiff review, Technical Specifications to be incorporated into the licensees for Prairie Island Units 1 and 2.
These specifications are consistent with the intent of the statements listed below.
NSP has assured that the Technical Specifications proposed are compatible with other NSP requirements.
1.
Both PORVs must be operable whenever the RCS temperature is less than the minimum pressurization temperature, except one PORV may be inoperable for seven days.
If these conditions are not met, the primary system must be depressurized and vented to the atmosphere or to the pressurizer relief tank within eight hours.
2.
Operability of the overpressure mitigating system requires that the low pressure set point will be selected, the upstream 1sulation valves open and the backup air supply charged.
3.
No more than one HPSI rump may be energized at RCS temperatures below 2000F (except as pennitted by Technical Specification 3.1.G.5).
4.
A reactor coolant pump may be started only if there is a bubble in the pressurizer or the SG/RCS temperature difference is less than 500F.
S.
The overpressure mitigating system must be testad on a periodic basis consisten+ with the need for its use.
5.0 Summary The administrative controls and hardware changes proposed by NSP provide protection of PINGP from pressure transients at low temperatures by reducing the probability of initiation of a transient and by limiting the pressure of such a transient to below the limits set by Appendix G.
We find that the overpressure mitigating system meets the criteria established by the NRC and is acceptable as a long term solution to the problem of overpressure transients. Any future revisions of Appendix G limits for PINGP must be considered with respect to this evaluation.
At that time, the overpressure mitigating system setpoint should be adjusted accordingly and corresponding adjustments in the license should be made.
Ci g-The overall low temperature overpressure protection system design in the area of electrical, instrumentation and control (EI&C) is in accordance with those design criteria originally prescri';ed by the staff and later amended in subsequent staff discussions with NSP.
We find the EI&C aspects of the proposed design to be acceptable on the basis that:
(1) the proposed system complies with IEEE Std. 279-1971 criteria and is designed as a seismic Category I system; (2) the system is redundant and satisfies the single failure criterion; (3) the design is such that the system requires no operator action for ten minutes after receipt of an overpressure action alarm; (4) the system is testable on a periodic basis.
Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint c f environmental impact and, pursuant to 10 CFR Section Sl.5(d)(4), that an environmental impact statement or negative declarati.on and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have conciuJed, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and sa fety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be ininical to the common defense and security or to the health and safety of the public.
Dated:
August 3,1979 iL bl REFERENCES 1.
Northern States Power Company letter \\Mayer) to NRC (Davis) " Reactor Vessel Overpressurization" dated July 22, 1977.
2.
flSP letter (Mayer) to NRC (Ziemann) " Reactor Vessel Overpressurizatnn" dated September 3,1976.
3.
NSP letter (Mayer) to NRC (Ziemann) "R' actor Vessel Overpressurization" dated Oc*ober 13, 1975.
a.
" Pressure Mitigating System Transient Analysis Results" prepared by Westinghouse for the Westinghouse Occurrences Group on Reactor Coolan:
System Overpressurization, dated July,1977.
5.
NSP letter (Mayer) to NRC (Ziemann) " Reactor Vessel Overpressurization" dated February 25, 1977 6.
NSP letter (Mayer) to NRC (Ziemann) " Reactor Vessel Overpressurization" dated April 4,1977.
7.
" Staff Discussion of Fifteen Technical Issues listed in Attachment G November 3,1976 Memorandum from Director, NRR to NRR Sta ff."
NUREG-0138, November,1976.
8.
NRC letter (Ziemann) to NSP (Mayer) " Prairie Island Nuclear Generating Plant Units Nos. I and 2" dated August 13, 1976.
9.
"Loftran Code Description", WCAP 7907, October 1972.
(lj
_'} ' b