ML20006B466
| ML20006B466 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/26/1990 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19310C663 | List: |
| References | |
| NUDOCS 9002020250 | |
| Download: ML20006B466 (10) | |
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PROPOSED CHANGE NO. 66 COOPER NUCLEAR STATION TECHNICAL SPECIFICATIONS 4
PROPOSED PAGE CHANGES
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e SAEETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 REACTOR COOLANT SYSTEM " rEGRITY c
Aeolicability:
Aeolicability:
Applies to limits on reactor Applies to trip settings of the coolant system pressure, instruments and devices which are
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provided to prevent the; reactor system safety limits from -being i
exceeded.
Obiective:
Obiective:
To establish a limit below which' the To define the level of the process-integrity of the reactor coolant variables at which automatic system is not threatened due to an protective action is initiated to overpressure condition.
prevent the pressure safety' limit from being exceeded.
Action Soecifications:
If a Safety Limit is exceeded, the reactor shall be in at least hot 1.
The limiting safety system settings shutdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, shall be as.specified below:
Protective Action /Limitine Safety Specifications!
System Settinc 1.
The reactor vessel domo pressure A.
Scram on Reactor ' Vessel high shall not exceed 1337 psig-at any pressure-time when irradiated fuel is present in the reactor vessel.
$1045 psig B.
Relief valve settings-t
-s1210 psig i
C; Safety valve settings-
~ $1277 psig 2.
Action 'shall be taken to decrease the reactor - vessel dome pressure j
l below 75 psig-or the shutdown l
cooling isolation valves shall be L-closed.
l 2.
The reactor vessel dome pressure shall not exceed 75 psig at-any time l
when operating the Residual Heat Removal pump in the shutdown cooling mode.
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2.2 BASES The eight relief valves and three safety valves are sized, and the maximum upper limit on the set pressures are established, in order to ensure that the peak vessel bottom pressure remains below 110% of the vessel design pressure, as required by Section III of the ASME code.
The relief valve settings satisfy the. code requirements that the lowest valve set point be at or below the vessel design pressure of 1250 psig.
The peak vessel dome pressure limit of 1337 - psig was established based upon the ASME limit of 1375 psig (110% of vessel design pressure) and the difference between the pressure at the bottom of the vessel head and that in the steam dome of the reactor vessel.
The relief valve nettings are also sufficiently above the normal operating pressure range to prevent unnecessary cycling _
caused by minor transients.
Reanalysis in Reference 7 for the limiting vessel overpressure event, _ 'the MSIV-Closure with flux scram event, was performed with the setpoints of al_1-' the relief valves assumed to be at an upper limit of 1210 psig and all the safety valves at 1277 psig.
The upper limit setpoint for the relief valves was established to ensure.that the peak vessel pressure will remain below 1375 psig.
The peak vessel bottom pressure resulting from the Reference 7 evaluation was 1322 psig,, which -
provides sufficient margin to ensure that the 1375 psig (110% of the vessel design j
pressure) limit will not be exceeded.
Results - of the overpressure protection 1
analysis for the current cycle are provided in the current reload licensing document.
The actual relief valve and safety valve setpoints will be grouped'as specified in Section 4.6.D.
A sensitivity study on peak vessel pressure to the' failure to open-of one of the lowest set point safety valves was performed for a typical high power density BWR (Reference 8). The study is applicable to the Cooper reactor and shows - _. l that the sensitivity of a high power density. plant to the failure of a safety valve is approximately 20 psi.
A plant specific analysis for the Cooper overpressure transient would show results equal to or less than this value.
The design pressure of the shutdown copling piping of the Residual-Heat Removal System is not exceeded with the reactor vessel steam dome less than 75 psig.
REFERENCES 1.
Topical Report, aummary of Results Obtained from a Typical' Startup and Power a
Test Program for a General Electric Boiling Water Reactor", General Electric Company, Atomic Power Equipment Department (APED-5698) 2.
Station Nuclear Safety Operational Analysis (Appendix G) 3.
Station Safety Analysis ( Ection XIV) 4.
Control and Instrumentation ('Section VII).
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5.
Summary Technical Report of-Reactor Vessel Overpressure' Protection (Question 4'20, Amendment 11 to SAR) 6.
" Supplemental Reload Licensing Submittal for Cooper Nuclear Station Unit 1,"
(applicable reload document).
7.
"SRV Setpoint Tolerance Analysis for Cooper Nuclear Station," General Electric Company, NEDC-31628P, October 1988.
8.
Letter from I. F. Stuart (GE) to v. Stello (NRC) dated December 23, 1975.
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LIMITING CONDITIO74S FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6.D Safety and Relief Valves 4.6.D.
Safety and Relief Valves 1.
During reactor power operating 1.
Approximately half of the safety conditions and prior to reactor valves and relief valves shall be startup from a Cold Condition, or-checked or-replaced with bench whenever reactor coolant pressure is checked-valves once per operating greater than atmospheric and cycle.- - All valves ' will be tested i
temperature greater than 212'F, all every two cycles, three safety valves and the safety modes of all relief valves shall be The safety valve function of safety.
operable, except as. specified in and relief valves shall actuate with o
3.6.D.2.
the following settings:
2.
a.
Relief Valves l
a.
From and after the date that the 1080 psig 33 psi (2 valves) safety valve function of one relief 1090 psig i 33 psi (3 valves) valve is made or found to be 1100 psig i 33 psi (3: valves) inoperable, continued reactor b.
Safety Valves operation is permissible only during-r
_l the succeeding thirty days unless.
1240 psig i 37 psi (3 valves) such valve function is sooner made operable.
2.
. At least. one of the relief valves shall be' disassembled and inspected.
'i b.
From and after the date that the each refueling outage.
safety valve function of two relief 1
valves is made or found to-be 3.
Deleted i
inoperable, continued reactor i
operation is permissible only during 4
Deleted-j the succeeding seven days unless such valve function is sooner made
.5.
Once per operating cycle,-with the
-i operable.
7, reactor pressure ;E 100 psig, each
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relief -valve shall-be. manually 3.
If Specification 3.6.D.1 is not met,
' opened until the main turbine bypass an orderly shutdown shall be valves have closed to compensate for-initiated and the reactor coolant relief. valve opening.
pressure shall be reduced to a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
6 4.
From and after the date that a.
Operability of. the relief valve-position indication on any one-position
. indicating pressure relief valve is made or found to be switches and the safety: ~ valve.
inoperable, continued reactor position indicating thermocouples operation is permissible only during shall be demonstrated once..per the succeeding thirty days unless operating cycle.
1 such valve position indication is sooner made operable, b.
An Instrument - Check of the safety-and relief valve position indicating -
devices shall be performed monthly.
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3.6.C & 4. 6.C BASES (cont'd.)
indicates _ that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized _ by gradual crack growth.
This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or i
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u cracks, associated with such leakage would grow rapidly. However, the establishment of allowable unidentified leakage greater than that given in 3.6.C on the basis of the data presently available would be premature because of uncertainties associated with the data.
For leakage of the order of 5 gpm, as specified in 3.6.C, the 1
experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for l
rapid propagation.
Leakage less than the magnitude 'specified can be detected
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l reasonably in a matter of a few hours utilizing the available leakage detection-schemes, and if' the origin cannot be determined in a reasonably short time the plant should be shutdown to allow further investigation and corrective action.
The total leakage rate consists of all leakage, identified'and unidentified, which flows to the drywell floor drain and equipment drain sumps, l'
The capacity of the drywell floor sump pumps is 50 gpm and the capacity 'of the drywell equipment sump pumps is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with margin.
1 Reactor coolant leakage is also sensed by the containment radiation monitoring l unit which senses gross beta, -gamma particulate and iodine as well as by oxygen and hydrogen analyzers.
Leakage can also be detected by area temperature detectors, humidity detectors and pressure instrumentation. Due to the many and varied ways of detecting primary leakage, a 30 day allowable repair time is justified.
D.
Safety and Relief Valves l
l The safety and relief valves are required e to be operable 'above - the pressure (113 psig) at which the core spray system.is not designed to deliver full flow. The pressure relief system for Cooper Nuclear Station has been sized to meet two design bases. First, the total safety / relief valve capacity has been established-to meet
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the overpressure protective criteria of the ASME code. ' Refer to' Specification 2.2 and the corresponding bases for the maximum opening upper limit pressure which has been established for Cooper Nuclear Station.
Second, the distribution of E this.
required capacity between safety valves and relief valves has been set to meet design basis IV.4.2.1 of subsection IV.4 which states that the nuclear system relief valves shall prevent opening of the safifty valves during normal plant isolations and load rej ections.
The details of the complete safety analysis, which includes demonstration of compliance with the ASME code requirements for vessel overprotection, is presented ~
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in Reference 1.
Results of the overpressure protection analysis for the current cycle are provided in the current reload licensing document.
Experience in relief and safety valve operation shows~that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
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.q 3.6.D & 4.6.D BASES (cont'd.)
The relief and safety valves are bench tested every second. operating cycle to ensure their set points are within the 13 percent tolerance._
To avoid increasing the probability of setpoint drift, all valves are to be reset to-within 1% of their nominal setpoint prior to being returned to service.
1 Additionally, once per operating cycle, each reliof valve is tested manually.
with reactor pressure above 100 psig to demonstrate its ability to pass steam.
The requirements established above apply when the nuclear system can be pressurized above ambient conditions. _These-requirements'are applicable at nuclear system pressures below normal operating precsures because abnormal l
operational _ transients could possibly start at these conditions such that 1
eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure', than starting at rated conditions. The valves need not be functional-when 'the. vessel head is removed, since' the-nuclear system cannot be pressurized.
The position indicating pressure switches for the relief valves and the thermocouples for the safety valves serve as a diagnostic aid to the operator in the event of a safety / relief _ valve failure.
If _ position indication is -
lost, alternate means are available to the operator to. determine if a safety
-l valve is leaking.
5 E.
Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism,. nozzle-assembly and/or riser, - would increase the cross-sectional flow ' area for blowdown following the design bases double-ended line break.
Therefore, if a, failure occurs, repairs must be made.
The detection technique is as follows.
. With ' the. two recirculation pumps balanced in speed to within 1 5%,, tha-flow rates in both recirculation loops will be verified by Control Room monitoring instruments. If _ the two flow rate values do not differ by more than 10%, riser and nozzle assembly integrity has been verified.
If they do differ by 10% or more, the core ' flow rate measured by the jet pump diffuser differential pressure system must be checked egainst the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 1
10% or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel'of fail,ed jet pump nozzle.(or riser)-
and the plant shut down for relairs.
If.the potential blowdown flow area is i
increased, the syste.m resistance to the_ recirculation pump is also reduced; hence, the affected drive.-pump will "run.out" to a substantially higher flow-rate (approximately 115% _ to 120% for a single: nozzle failure)..
If the two loops are balanced in flow at the same. pump speed, 'the resistance characteristics cannot have changed.
Any imbalance between drive loop flow-
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rates would be indicated by the plant process instrumentation.
In addition, l
the-affected jet pump would provide a leakage path past the core thus reducing; the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but. the net effect would be 1
a slight decrease (3% to 6%)lin the total core flow measured. This decrease, j
together with the loop flow-increase, would result in a lack of correlation
.j between measured and derived core flow rate.
Finally, the affected jet pump
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diffuser differential pressure signal would be reduced because the backflow 4
would be less than the normal forward flow.
l A nozzle-riser system failure could also_ generate the coincident failure of a M
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BASFS (cont'd)
When a snubber is found inoperable, an engineering evaluation is performed, in addition ta the determination of the snubber mode of failure, in order to determine if any safety related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of fcilura has imparted a significant effect or degradation on the supported component or system.
In caser where the cause of failure has been identified, additional snubbers, having a high probability for the same type of failure or that are being used in the same cpplication that caused the failure, shall be tested.
This requirement increases the probability of locating inoperable snubbers without testing 100% of the snubbers.
Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance programs.
To further increase the assurance of snubber reliability, functional tests should be psrformed once each refueling cycle. Ten percent of each type of snubber represents an edequate sample for such tests. Observed failures on these samples should require testing of additional units. Snubbers in high radiation areas or those especially difficult to rsmove need not be selected for functional' tests provided operability was previously varified.
The service life of a snubber is evaluated via manufacturer input and consideration of the snubber service conditions.
The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.
REFERENCES 1.
"SRV S** point Toleranco Analysis for dooper Nuclear Station," General Electric Company, NEDC 31628P, October 1908.
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q GENERAL ELECTRIC COMPANY AFFIDAVIT-I, David J. Robare, being duly sworn, depose and state as follows:
1 1.
I am Manager, Licensing Services, General Electric Company, and have been delegated the -
function of reviewing the information described in paragraph 2 which is sought to be withheld and have been authorized to apply for its withholding.
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f 2.
The information sought to be withheld is contained in the GE proprietary report NEDC 31628P (Class III), "SRV Setpoint Tolerance Analysis for Cooper Nuclear Station",'
7 October 1988. This report provides the basis for revised plant Technical Specification changes to SRVs and SVs consistent with current ASME Code provisions. The safety limit setting' tolerance in Section 2.2 and the surve,illance requirements of Section 4.6.D of the Tech Specs :
could be revised. In addition, it provides recommended setpoint tolerances to be applied prior -
to returningvalves to service.
"A trade secret may consist of any formula, pattern, device or compilation of information :
which is used in one's business and which gives him an opportunity to obtain an advantage over competitors who do not know or use it.- A substantial element of secrecy must exist so' that, except by the use of improper means, there would be difficulty in acquiring i
information...Some factors to be considered in determining whether given information is one's -
trade secret are (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3) the extent 3
of measures taken by him to guard the secrecy of the information;(4) the value of the-1 information to him and to his competitors; (5)'r difficulty with which the information co him developing the information; (6) the ease o properly acquired or duplicated by others".
3.
Some examples of categories of information which fit into the definition of Proprietary q
Information are:
i
- a. Information that discloses a process, method or apparatus where prevention of its use.by.-
General Electric's competitors without license from General Electric' constitutes a competitive economic advantage o_ver other companies;-
- b. Information consisting of supporting data and analyses, including test data, relative to a process, method or apparatus, the application of which provide a competitive economic q
advantage, e.g., by optimization or improved marketability; 1
- c. Information which if used by a competitor, would reduce his expenditures of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality or licensing of a similar product; i
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GENERAL ELECTRIC COMPANY' AFFIDAVIT i
- d. Information which reveals cost or price information, production capacities, budget levels or commercial strategies of General Electric, its customers or suppliers; j
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- e. Information which reveals aspects of past, present or future General-Electric
.' customer funded develop _ ment plans and programs of potential commercial value to r
j General Electric; p
- f. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection;
- g. Information which General Electric must treat as proprietary according to agreements-i with other parties.
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4.
Initial approval of proprietary treatment of a' document is typically made by the Subsection-Manager of the origmating component, the person who is most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such l.
documents within the Company is limited on a "need to know" basis and such documents are j
L clearly identified as proprietary. -
J
' 5.. The procedure for approval of external release of such a document typically requires revie'v by the Subsection Manager, Project Manager, Principal Scientist or other equivalent authority, by the Subsection Manager of the cognizant Marketing function (or delegate) and by the Legal-4 Operation for technical content, competitively effect and determination of the accuracy of the
. proprietary designation in accordance with the standards' enumerated abo've.. Disclosures -
9 outside General Electric are generally limited to regulatory bodies, customers and potential.
customers and their agents, suppliers and licensees then only with appropriate protection by applicable regulatory provisions or proprietary agreements. ' '
6.
The document mentioned in paragra ah 2 above has been evaluated in accordance with the above criteria and procedures and has >een found to contain information which is proprietary and which is customarily held in confidence by General Electric.
7.
The information to the best of my knowledge and belief has consistently.been held in-confidence by the General Electric Company, no public disclosure has been made, and it is not -
available in public sources. All dtTclosures to third parties have been made pursuant to regulatory provisions of proprietary agreements which provide for maintenance of the-information in confidence. '
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' 8.
Public disclosure of the information sought to be withheld is likely to cause substantial harm to -
the competitive position of the General Electric Company and deprive or reduce the availability of profit making opportunities because it would provide other parties, including competitors, with valuable information.
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i GENERAL ELECTR',0 COMPAh7 STATE OF CALIFORNIA
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COUNTY OF SANTA CLARA
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David J. Robare, being duly sworn, deposes and says' That he has read the foregoing affidavit and the matters stated therein are truly and correct to i
the best of his knowledge, information, and belief.
i Executed at San Jose, California, this 6 day of 0CT0064 19 f9.
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David J. Robare General Electric Company i
Subscribed and sworn before me this 4 day of 848/eo 19 g -
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OFFICIAL SEAL f'
MARY L XENDALL
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W comm. Es. Mar. as, ises Notary Public, State of California i
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