ML18101A187
| ML18101A187 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/19/1994 |
| From: | Hagan J Public Service Electric & Gas Co, of New Jersey |
| To: | NRC/IRM |
| Shared Package | |
| ML18101A188 | List: |
| References | |
| LCR-94-18, NLR-N94137, NUDOCS 9408260210 | |
| Download: ML18101A187 (21) | |
Text
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Public Service Electric and Gas Company Joseph J. Hagan Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President
- Nuclear Operations AUG 191994 NLR-N94137 LCR 94-18 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR AMENDMENT STEAM GENERATOR LOW AND LOW-LOW LEVEL SETPOINTS SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 In accordance with the requirements of 10CFR50.90, Public Service Electric and Gas Company (PSE&G) hereby transmits a request for amendment of Facility Operating Licenses DPR-70 and DPR-75 for Salem Unit Nos. 1 and 2.
Pursuant to the requirements of 10CFR50.91(b) (1), a copy of this request for amendment has been sent to the State of New Jersey.
This request would reduce the minimum setpoints and allowable values for the steam Generator Level-- Low-Low and Low reactor protection system signals.
The changes would increase operating margin and reduce the potential for unnecessary reactor trips based on improved steam generator level channel accuracy.
This request for amendment satisfies a corrective action identified in Licensee Event Report (LER) 311/94-008-00, dated July 27, 1994, which reported a reactor trip on Steam Generator Level-- Low-Low during plant startup.
Changes to the Technical Specification bases are also included to modify and expand the description of the relationship between setpoints, allowable values and the plant safety analyses, based on the improved Standard Technical Specifications of NUREG-1431. includes the description and justification for the proposed changes, including PSE&G's Determination of No.
Significant Hazards Consideration. contains the Technical Specification pages revised with pen and ink changes.
PSE&G requests an amendment to be implemented at each Salem unit no later than upon restart from the first outage of sufficient duration following issuance of the amendment.
These
~,
9408260210 940819 PDR ADOCK 05000272 P
PDR f(CD\\ I j
Document Control Desk NLR-N94137 AUG 1 9 1994 implementation provisions would allow adjustment of the setpoints, procedure changes and testing.
Approval is requested to allow implementation at Unit 2 during its eighth refueling outage, which is scheduled to begin in October, 1994.
Affidavit Attachments (2)
Sincerely, C
Mr. T. T. Martin, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J. C. Stone, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. C. Marschall (S09)
USNRC Senior Resident Inspector Mr. K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
REF:
NLR-N94137 STATE OF NEW JERSEY SS.
COUNTY OF SALEM J. J. Hagan, being duly sworn according to law deposes and says:
I am Vice President - Nuclear Operations of Public_ Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Salem Generating Station, Unit Nos. 1 and 2, are true to the best of my knowledge, information and belief.
Subscribed and Sworn79 before me t!li)
/q~~ d~ay of~~d, 1994
.*~
/, pJ; f
~
Q/M/11 Notary_PublicOfew Jersey My Commission expires on KIMBERLY JO BROWN NOTARY PUBLIC OF NEW JERSEY My Commission Exoires Aoril 21. 1998
NLR-N94137 ATTACHMENT 1 I.
DESCRIPTION OF THE PROPOSED CHANGES Revise Salem Generating Station (SGS) Unit Nos. 1 and 2 Technical Specification (TS) Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, and Table 3.3-4, Engineered Safety Feature Actuation System Instrumentation Trip Setpoints, as follows:
- 1)
Change the Stearn Generator Water Level--
Low-Low setpoint from ~16% Narrow Range Span (NRS) to ~9.0% NRS, and the allowable value from ~14.8% NRS to ~8.0% NRS.
This signal is used for reactor trip (Table 2.2-1) and Auxiliary Feedwater (AFW) system actuation (Table 3.3-4).
- 2)
Change the Low Stearn Generator Water Level setpoint from
~25% NRS to ~10.0% NRS and the allowable value from ~24% NRS to ~9.0% NRS.
This change would affect the reactor trip functional unit for Stearn/Feedwater Flow Mismatch and Low Stearn Generator Water Level (Table 2.2-1).
In addition to the proposed Technical Specification changes, revisions to the Technical Specification Bases B2.2.1, Reactor Trip System Instrumentation Setpoints and B3/4.3.1 and 3/4.3.2, Protective and Engineered Safety Features (ESF) Instrumentation are included to clarify the general relationship between setpoints, allowable values and analytical limits used in the safety analyses.
A change to the bases for the Stearn/Feedwater Flow Mismatch and Low Stearn Generator Water Level trip is also included (B2.2.1), to delete specific setpoint values.
II.
REASON FOR THE PROPOSED CHANGES The proposed changes would increase operating margin relative to steam generator level.
This would help preclude unnecessary reactor trips and AFW system actuations during plant evolutions involving steam generator water level changes (e.g., plant startup), while continuing to ensure the analytical limits in the safety analyses remain valid.
License Change Request (LCR 92-14), originally submitted via PSE&G letter dated 2/5/93 (NLR-N93001), requested deletion of the reactor trip on Stearn/Feedwater Flow Mismatch coincident with Low Stearn Generator Level.
NRC has reviewed the LCR, as documented in its draft Safety Evaluation Report (SER) letter dated December 2, 1993.
Issuance of the NRC amendment is on hold pending PSE&G's replacement of the existing feedwater control system with the Westinghouse advanced digital feedwater control system.
The reduction in the steam generator low level setpoint 1 of 5
and allowable value proposed herein would increase operating margin associated with the trip function until removal of the trip function is implemented.
The Bases changes were developed using the improved Standard Technical Specification bases for Trip Setpoints and Allowable Values (NUREG-1431, B3.3.1 and B3.3.2).
These changes would incorporate a description of the relationship between setpoints, allowable values and analytical limits, which are determined by PSE&G's setpoint methodology consistent with standard industry practice.
III.
JUSTIFICATION FOR-THE PROPOSED CHANGES The proposed changes are based on reduced channel uncertainties that have been calculated by PSE&G using a setpoint methodology consistent with Instrument Society of America standard ISA-867. 04, which is endorsed by NRC Regulatory Guide 1.105, Rev. 2.
The reduction in channel uncertainty is primarily the result of replacing the Rosemount 1153 series level transmitters with Rosemount 1154HH transmitters, which have improved environmental specifications.
The total accident channel uncertainty previously prepared utilizing the Rosemount 1153 transmitters resulted in a 15.3% NRS error.
This total accident uncertainty as calculated by PSE&G utilizing the Rosemount 1154HH transmitters has been reduced to 7.407% NRS.
The reduction of total accident uncertainties included a more specific analysis of process measurement uncertainties resulting in an improved overall value.
Based on the reduced uncertainties, the proposed setpoints and allowable values would continue to ensure the trip settings assumed in the plant safety analyses remain valid.
- 1)
The Steam Generator Water Level--
Low-Low signal initiates a reactor trip and actuation of the Auxiliary Feedwater (AFW) system.
This signal is used as a primary protection signal for postulated design basis events including loss of normal feedwater, loss of offsite power and feedwater line break.
The safety analyses assume reactor trip and AFW actuation occurs at 0.0% NRS (i.e., analytical limit).
The total calculated channel uncertainty for the low-low level channel is +7.407%, -3.458%.
Because the low-low level signal protects against conditions involving decreasing steam generator level, the positive uncertainty value (+7.407%)
is subtracted from the setpoint to determine whether there is adequate margin relative to the analytical limit. The proposed setpoint of ~9.0% and allowable value of ~8.0%
would ensure the analytical limit of 0.0% NRS is met with excess margin.
- 2)
The Low Steam Generator Water Level signal coincident with the Steam Flow/Feed Flow Mismatch signal initiates a reactor trip.
This signal is not credited in any safety analyses, but increases the overall reliability of the reactor protection system.
The low-low level signal uses three channels per steam generator, one of which is used by the 2 of 5
level control system.
The low level coincident trip is designed to ensure compliance with IEEE-279-1971, Section 4.7.3 relative to control and protection system interactions, such that a control system failure and random single failure of a protection channel would not result in loss of the protection function.
Because it is not credited in the UFSAR Chapter 15 safety analyses, there is no analytical limit associated with the low level signal.
The uncertainties calculated for the low level signal are identical to that of the low-low level signal (+7.407%, -3.458%).
The proposed setpoint of ~10.0%
NRS and allowable value of ~9.0% NRS would continue to provide backup protection to the low-low level trip signal with excess margin to the analytical -limit used for the low-low level trip.
The reduction in setpoint would increase the margin available for steam generator level recovery when a flow mismatch condition exists.
The bases changes for the low level coincident trip would delete the specific setpoints for the flow mismatch and low steam generator level signals.
Specific setpoint values are maintained in the Technical Specifications, and need not be duplicated in the bases.
The overall intent of the trip function is not changed.
As stated in the current bases, the trip would continue to be initiated before the steam generators are dry, reducing demands on the AFW system and minimizing thermal transients.
Note that the AFW system is actuated by the Low-Low Level signal (not the low level coincident signal).
The analytical limit for the low-low setpoint (0.0% NRS) is above the top of the steam: generator tubes.
The accompanying bases changes relative to setpoints and allowable values are based on the improved Westinghouse Standard Technical Specifications (NUREG-1431 B3.3.1).
Significant differences between the NUREG-1431 bases and the changes included herein are as follows:
o NUREG-1431 refers to "RTS/ESFAS Setpoint Methodology Study" where the Salem bases refer to ISA-S67.04-1982.
This is because the Salem setpoint methodology is consistent with the ISA standard, which is widely used in the industry and endorsed by NRC in Regulatory Guide 1.105, Rev. 2.
PSE&G does not use a document called "RTS/ESFAS Setpoint Methodology Study," and believes it is more appropriate to refer to an industry standard than a plant-specific study in the Technical Specification bases.
o NUREG-1431 refers to the CHANNEL OPERATIONAL TEST (COT) as the test which is capable of detecting those measurement uncertainties comprising the difference between the Trip Setpoint and Allowable Value.
The Salem bases refer to the CHANNEL FUNCTIONAL TEST, which is equivalent to the COT in NUREG-1431.
CHANNEL FUNCTIONAL TEST is a defined term in 3 of 5
the Salem Technical Specifications, whereas COT is not.
o NUR~G-1431 includes a paragraph relative to the ability to test channels on-line "to verify that the signal or setpoint accuracy is within the specified allowance requirements of
[UFSAR Chapter 6]... "
This paragraph is not included in the Salem bases.
Salem UFSAR Chapter 6 does not specify channel "allowance requirements."
The Salem Technical Specifications and bases already define test requirements in sufficient detail such that the paragraph in NUREG-1431 is not considered necessary.
The bases changes are not unique to the steam generator level setpoints and allowable values proposed herein, but are included because they are useful in describing the method in which setpoints and allowable values are established, consistent with an approved industry standard, to ensure that the reactor protection system channels protect the limits of the safety analyses.
IV.
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The proposed changes for Salem Unit Nos. 1 and 2:
(1) do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The Steam Generator Water Level--
Low-Low signal and the Low Steam Generator Level coincident with Steam Flow/Feed Flow Mismatch signal are designed to mitigate design basis transients involving significant reductions of steam generator inventory (e.g., Loss of Normal Feedwater, Turbine Trip, Loss of Offsite Power, Feedwater Line Break).
The setpoints and allowable values for these protection signals are prescribed by Technical Specifications such that performance of the signals is consistent with the plant safety analyses, considering the effects of channel uncertainties.
The proposed reductions to the setpoints and allowable values for the low-low and low steam generator level signals would not affect the probability of any transient that the protection signals are designed to miti~ate.
The changes would reduce the probability of unnecessary reactor trips and Auxiliary Feedwater (AFW) system actuations by providing greater operating margin for plant evolutions involving steam generator level changes (e.g., plant startup).
Therefore, the proposed changes do not involve any increase in probability of an accident previously evaluated.
The changes to the Steam Generator Water Level--
Low-Low signal would not result in any increase in consequences of a previously analyzed accident because the proposed setpoint and allowable value would continue to ensure the safety analysis assumptions remain valid.
As described in the accompanying changes to the Technical Specification Bases, the channel uncertainty calculations performed to establish the relationships between the setpoints, allowable values and safety analyses are consistent 4 of 5
with NRC Regulatory Guide 1.105, Revision 2.
Low Steam Generator Level coincident with Steam Flow/Feed Flow Mismatch signal is not credited in the UFSAR Chapter 15 safety analyses.
The proposed changes to the low steam generator level setpoint and allowable value would continue to provide reliable backup to the low-low level trip signal, consistent with IEEE-279-1971.
Therefore, the proposed changes would not involve an increase in consequences of any previously analyzed accident.
(2) do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes would continue to ensure the appropriate reactor protection system functions (reactor trip* and AFW initiation) are initiated in the event that steam generator water level decreases to the value used in the plant safety analyses.
The proposed changes would not involve any changes in protection system logic or function, and do not involve any plant configurations that could adversely affect the initiation or progression of any accident sequence.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3) do not involve a significant reduction in a margin of safety.
The proposed setpoints and allowable values would continue to ensure that the assumptions in the safety analyses remain valid, with appropriate consideration of protection system channel uncertainties.
Therefore, the proposed changes do not involve a reduction in margin of safety.
Therefore, PSE&G has concluded that the changes proposed herein do not involve a*significant Hazards Consideration.
5 of 5
NLR-N94137 ATTACHMENT 2
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TABLE 2.2-1 <Continued)
REACTOR TRIP SJSTEM INSTRUMENTATION TRIP' SETPO!MTS FUNCJI ONAL UN IT
- 13.
Ste** Gener*tor W*ter Level--Low*Lo1t
~
l!_ll!__SETPOINT
~~of n*rrow apen***ch at***
....---t ALLOWABLE YALUES r*n11e I nat ruaentl q.CJ%~ of n*rrow range I nat ru*ent gen*r*tor ap*n*e*ch at*** gener*tor
- 14.
St***IF*edw*t*r flo1t Ml***tch *nd LOii St***
Gener*tor U*t*r Level
- 15.
Undervolt*11*-l**ctor Coolent Pu*p*
- 16.
Underfrequency*leector Coolant Pu*p*
- 17.
Turbine Trip A. Low Trip Syate*
Preaaure I. Turbin* Stop V*lve Cloaure
- 11.
Sefety Injection Input fro* SSPS
- 19.
leector Coolent Pu*p lreeker Poaltlon Trip
~ 401 of full ot*** flow et RATED THERMAL POWEi coincident with ate**
generetor 11eter level ~
of n*rro1t ren11* lnatru*ent apen--eech
- t*** generetor
/().0
~ 2900 volt****ch bu*
~ 56.5 Hz - e*ch bu*
~ 45 pat11
~ 151 off full open Mot Appllc*bl*
Not Appllcebl*
- '.~~~~~~
~ 42.51 of full ate** flow *t RATED THERMAL POWEi coincident with ate**
gener*tor 1t*t*r level >....a.44 of
- '""""j";;t°'~**--****
at*** genor*tor a 7.0
~ 2150 volta-eech bu*
~ 56.4 Hz - eech bu*
~45 patg
~ 151 off full open lot Appllceble Not Appltcebl*
LIMITING SAFETY SYSTEM SETTINGS BASES rel1ab1lity of the Reactor Protection System. This tri is redundant to the Steam Generator Water Level Law-Law trip. The Steam/Feeciitater Flow Mismatch portion of thii tri~ is act vated w en t e steam flow exceeds the feeciitater flow by ~ 1.42 x 10 lbs/hour. The Steam Generator LC1111 Water level portion of the trip is activated when the water level dro s below 25 ercent as indicated b the narrC1111 ran e instrument. These rip values include sufficient al1C1111ance in excess of normal operating va ues to preclude spurious trips but will initiate a reactor trip before the steam gene ators are dry. Therefore, the required capacity and starting time requireme ts of the auxiliary feeciitater pumps are reduced and the resulting thermal tra sient on the Reactor Coolant System and steam generators is minimized.
Undervoltage and Underfrequency - Reactor Coolant Pump Susses The Undervoltage and Underfrequency Reactor Coolant Pu~ bus trips provide reactor core protection against Ile as a result of loss of voltage or underfrequency to more than one reactor coolant pu~. The specified set points assure a reactor trip signal is generated before the low flow trip set point is reached.
Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical pC1111er transients. For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pu~ bus circuit breakers shall not exceed 0.9 seconds.
For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is
- reached shall not exceed 0.3 seconds.
Turbine Trip A Turbine* Trtp causes-. a direct. reactor*trip when* operat1,ng, above. P-9. Each.
of the* turbine-trips provf de. turbine. protection and reduce the. severtty of the ensuing transient.
No credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.
SA 1.£M - UNIT 1 B 2-7 Amendment No. 85
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.e 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS Manual Reactor Trip The Manual Reacto1! Trip is a reriundant channel 1.o tli..: auto111atic protective i nst.run1e1.ta tfor. chann1:* ls and µrovi d.::s mariu~.1 reta,~tor trip capability.
Power Range, Neutron F1ux The Power Range, Neutron F1ux channc*1 high setpoint provides reactor core protection agaiost reactivity exc.ursions which are too rapid to be protected by temperature and pressure protective circuitry. The low set point provides redundant protection in the power range for a power excursion beginning from low power.
The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range chann,:!ls indicate a power level of above approximately 9 percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RATED THERMAL POWER).
Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid f1ux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.
SALEM - UNIT 1 B 2-3
LIMITING SAFETY c 9..... EM sE"""TT......,..I,._NG... s;;._ _________________ _
BASES through the pressurizer safety valves. No credit was taken fc7 operation of this trip in the accident analyses;* however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
- .oss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.
Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop flow.
Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature AT trip set point is adjusted to the value specified for all loops in operation. With the Overtemperature AT trip set point adjusted to the value specified for 3 loop operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational transients and anticipated transients with 3 loops in operation.
Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.
Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall SALEM - UNIT 1 B 2-6 Amendment No. 96
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TABLI J.l-4 (f.ontinued)
INOINllRID SAFETY fll\\TURI ACTUATIQH SYSTIM INSTRUHINTATIQll TRIP SITPQINTS FUNCTIONAL UNIT
- 5.
TURBINE TRIP AIU> FllDWATIR ISOLATION
. 6.
A.
Ste.. Generator Water Level --
Hi9h-Hi9h SAFIOUAIU>I BQUIPMINT CONTROL SYSTBN (SICI
- 7.
UNDIRVOLTAOB, VITAL BUS
- a.
Lo** of Voltage
- b.
Su*tained Degraded Voltage
- 8.
AUXILIARY FBIDWATIR
- a.
- b.
- c.
- d.
- e.
- f.
g
- Aut011Atic Actuation Logic Manual Initiation St*.. Generator Water Level--
Low-Low Und*r*olta9* - RCP s.1.
Trip of Main reedwater Pump*
Station Blackout
_ TRIP llTPQINT
$ 67\\ of narrow range in*trument *pan each
- tea11 generator Not Applicable
~ 70\\ of bu* voltage ALLOWULI VALUES
$ 68\\ of narrow range instrument span each
- team generator Not Applicable.
~ 65\\ of bue voltage
~ 91.6\\ of bu* voltage for
$ ll 11econd*
~ 91\\ of bue voltage for
$ 15 *econde Not Applicable Applicable
~
of narrow range in*trument *pan each
- team generator
~ 70\\ RCP bu* voltage Hot Applicable Not Applicable
~**-*~~~
~
~
of narrow range in*trument epan each steam generator
~ 65\\ RCP buo voltage See 1 above (All s.1. set~int*)
Not Applicable Not Applicable See 6 and 7 above (SEC and Undervoltage, Vital Bue,
e e
3/4.J INSTBtlMENIATION BASES 3/4.3.1 and 3/4.3.2 PRQTECIIYE AND ENGINiEBED SAfETX FEATtJBES CESFl INSTIUJMENTATION The OPERABILITY of the protective and ESF in*trumentation *y*t91Da and interlock* *n*ure that 1) th* a**ociated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceed* it* **tpoint, 2) th* *pecifi*d coincidence logic and *ufficiant redundancy i* maintained to permit a channel to b* out of **rvic* for testing or maintenance con*i*t*nt with maintaining an appropriate level of reliability of the Reactor Protection and Engin.. red Safety Feature* in*trumentation and, 3)
- ufficient *y*tem functio~apa.bility i* available from diver** parameter*.
Th* OPERABILITY of the** my*tem* i* required to provide the overall reliability, redundance and div*r*ity ***umed available in the facility de*ign for th* protection and mitigation of accident and tran*i*nt condition*. The integrated operation of each of th*** *y*tem* i* con*i*t*nt with the
- umption* u*ed in th* accident analy****
~
~
E Th* *urveillance requirement* *pecif ied for th*** *y*tem9 *n*ure that th*
overall *y*tem functional capability i* maintained comparable to th* original de*ign. *tandard*. Th* periodic *urveillance t**t* performed at th* minimum frequenci** are *ufficient to demon*trate thi* capability.
Specified
- urveillance interval* and *urveillance and maintenance outage time* have been determined in accordance with WCAP-10271, *sval~ation of Surveillance Prequencie* and OUt of Service Time* for the Reactor Protection In*trwnentation Sy*tem,* and Supplement* to that report.
Surveillance interval* and out of **rvic* time* were detez:mined ba*ed on maintaining an appropriate level of reliability of the Reactor Protection Sy*t911l and Bngin.. red Safety Peatur** in*trumentation.
Th* mea*ur...nt of rempon** time at the 9P90if ied frequenci** provide*
aa*urance that th* protective and SSP action function a**ociated with each channel i* camplet9d within the time limit a*eum.d in the accident analy****
Ro cr*dit waa tak9n in the analy..
- for tho** channel* with reapon** time*
indicated ** not applicable.
MllPOD.. t1- -y be dmonatrated by any..ri** of aequential, overlappi119 or total channel teat -.au~nta provided that *ucb teat*
.d..onatrau the total channel reapon** time u defined. senaor r**pon-time verification..
y be d.mon*trated by eitlmr 1) in place, onmit* or offait* t**t me~*
or 2) utiliaincJ repla~nt.enaora with certified reapon**
time*.
3/4,3.3 MQllITQBilq IllSTIUJJllllTATIQI 3/f,3,3,1 MPIATIQI MQII'l'OBIIG IISTBJDllllTATIQI Th* OPSltABILI'rY of the radiation monitoriDCJ channel* *n*ur** that
- 1) the radiation level* are continually -**ured in th* area* **rved SAI&ll -
URIT 1 a 3/f 3-1 Amendment No.142
-1
TABLE 2.2-1 <Continued>
REACTOR TRIP SYSTEM llSTRUMEITATIOI TRIP SETPOllTS FUNCTIONAL UNIT
~
J.RLP SETPOllT
- 13.
Ste** Generator W*ter ~~~
of narrow r*nge ln*tru*ent Level--Low*Low
- p*n-each *t*a* generator ALLOWAILE VALUES
~
of narrow range lnatru**nt apan-each *t*** generator
- 14.
St***/feedwater Flow Mla*atch and Low St***
Generator Water Level
~ 401 of full at*** flow at RATED THERMAL POWEi coincident with atea*
~ 42.51 of full atea* flow at RATED THERMAL POWEi coincident with atea*
- 15.
Undervoltage-leactor Coolant Pu*p*
- 16.
Underfrequency-leactor Coolant Pu*p*
- 17.
Turbine Trip A. Low Trip Syate*
generator water level ~
of narrow range lnatru*ent apan--each at*** generator JO.O'fl
~ 2900 volta-each bua
~ 56.5 Hz - each bua
~ 45 palg gener*tor water level ~
of n*rrow range lnatru*ent apan--each atea* generator
~ 2150 volt*-**ch bua
~ 56.4 Hz - each bua
~45 palg
~
Preaaure m
- 1. Turbl~* Stop V*lve 0......
Cloaure
- 11.
S*fety Injection Input fro* SSPS
- 19.
Reactor Cool*nt Pu*p Br**k*r Poaltlon Trip
~ 151 off full open
~ 151 off full open Not Applicable lot Applicable Not Applicable lot Applicable I
REPL..~~E bVllrl
'J:(')G /;~I j 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2. 1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Power Range, Neutron Flux The Power Range, Neutron Flux channel high satpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pre$sure protective circuitry. The low set point provides redundant protection in the power range for a power excursion beginning from low power.* The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicatP-a power level of above -approxiutely 9 percent of RATED THERMAL POWE:t) and is auto-matically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approxiutely 9 percent of RATED THERMAL POWER).
Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events fro11 any power level. Specifically, this trip complements the Power Range Neutron Flux Hig~
and Low trips to ensure that the criteria are met for rod ejection ~rom partial power.
SALEM - UNIT 2 B 2-3
LI~ITING SAFETY SYSTEM SETTINGS BASES Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.
Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop flow.
Above 367. (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational"transients and anticipated transients when 3 loops are in operation and the Overtemperature delta T trip set point is adjusted to the value specified for all loops in operation. With the Overtemperature delta T trip set point adjusted to the value specified for 3 loop operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational transients and anticipated transients with 3 loops in operation.
Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.
Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This tri is redundant to the Steam Generator Water Level Low-Low trip. The Steam Feedwater Flow Mismatch portion of this trip is activated when.the6steam flow exceeds the feedwater flow by greater than or equal to 1.42 x 10 lbs/hour. The Steam Generator Low Water level portion of the trip is activated when the water level drop* below 24 ercent as indicated b the narrow ran e instrument. These rip values include sufficient allowance in excess of normal operating values t preclude spurious trips but will initiate a reactor trip before the steam*gen rator* are dry. Therefore, the required capacity and starting time requiremen
- of the auxiliary f eedwater pumps are reduced and the resulting the 1 transient on the Reactor Coolant System and steam generators is minimized.
SALEM - UNIT 2 B 2-6 Amendment No. 72
H ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT
- 7.
UNDERVOLTAGI, VITAL BUS
- a.
Lo** of Voltage
- b.
Suetained Degraded Voltage
~ 70\\ of bus voltage
~ 91.6\\ of bus voltage for S
1 J seconds
- 8.
AUXILIARY FBIDWATIR
- a.
- b.
c *
- d.
Automatic Actuation Logic Not Applicable Manual Initiation Steam Generator Water Low-Low Undervoltage -
RCP Not Applicable Level--~ of narrow range in*trument *pan each eteam generator
~ 70\\ RCP bus voltage
~
- e.
S.I.
See 1 above (all s.I. 11etpoint11t
- f.
'l'rip of Main Feedwater P\\.lnp Not Applicable ALLQWABLE VALUES
~ 65\\ of bus voltage
~ 91\\ o: bus voltage for S 15 seconds Not Applicable Not Applicable of narrow range ln*trument span each
- team generator
~ 65\\ RCP bus voltage Not Applicable Station Blackout See 6 and 7 above (SEC and Undervoltage, Vital Bus)
- g.
- 9.
SEMIAUTOMATIC TRANSFER TO RECIRCULATION RWST Low Level
- b.
Automatic Actuation Logic 15.25 ft. above Inetrument tape Not Applicable 15.25 +/- 1 ft. above instrument taps Not Applicable
I TAserf / I 3/f,J IHSTBUMIHTATIOlf BAS SS
- --*m** *********************************--**********************************
3/4.J.1 *ns! 3/4.3.2 PRQTlctIVI AHp IN9IlfllRIP SAlljty PIA'l'tlBIS <ISP>
IHS'l'RtJKl1ft'A'11QH Th* OPIRABILITY of the protective and ISP in*trumentation *y*t.... and interlock* *n*ur* that 1) the a**ociated ISP action and/or reactor trip will be i'Ltiated when th* parameter monitored by each channel or cambination thereof exceed* it* **tpoint, 2) th* *pecified coincidence loqic and
- uff icient redundancy i* maintained to permit a channel to be out of **rvic*
for te*ting or maintenance con*i*t*nt with maintaining an appropriate level of reliability of the Reactor Protection and Sngin.. red Safety Feature*
in*trwnentation and, 3) *ufficient *y*tem functio~capability i* available frcm diver** parameter*.
~
Th* OPIRABILITY of th*** *y*t... i* required to provide th* overall reliability, redundance and diver*ity a**~ available in th* facility d**iqn for the protection and mitigation of accident and tran*ient condition*. Th*
integrated operation of each of th*** *y*t... i* con*i*tent with the a**umption* u*ed in the accident analy****
7 Th* *urveillanc* requir... nt8 *pecified for th*** *y*t... *n*ur* that the overall *y*tem functional capability i* maintained comparable to the original d**iqn *tandard*. Th* periodic *urv*illance t**t* perfonled at th* minimull frequenci** are *ufficient to demcn*trate thi* capability. Specified
- urveillance interval* and *urveillance and aaintenance outage time* have been determined in accordance with tfCaP-10271,... aluation of surveillance Prequenci** and Out of ser.ice T1-* for the -ctor Protection In*trumentation Sy*t..,* and Supple118Dt* to that report. Surveillance interval* and out of HrYice time* weft dfterllined baited On m&intaininCJ an appropriate level of reliability of the bactor Protection Sy*t.. and Sn9in.. red Safety Feature* inmtrumentation.
The mea8Ur-nt of n*ponH t1-at the *pecif ied frequenci** provi..
- a**urance that the protective and 1sr action function a*110Ciated with each channel i* ccmpleted within the t.W. Uait u9\\199d in tbe accident analy****
Ko credit va* tak8n in the aaalyH* for tbo.. cbanD81* with re9P0n** time*
indicated u not applicable.
ttm. MJ' be dmonmtrated by any.ari** of HqUential, Oflerl.....
- total chamutl temt *a*~t* proYid8d that ncb t**t*
d~
~
tftal channel re*ponH ttm. u deU.llM. -.Or r**pon** time ver1f1aatla. MJ' be dwmftrated by eitUr 1) in place, ouit* or off*it* te*t
-~*
or 2) ut1liain; replac...ut HIUIOr* vi~ c9Rified re9P0ue 3/f,3,3 llQII'fOBlllQ IKSTBtJMllTATIQI 3/f,3,3.1 RAPIATIQI llQII'J'QBllQ IIS'l'IUWiftTA'fIQI Th* OPSRABILIT!' of the radiation monitorinCJ channel* en*ur** that 1) th*
radiation level* are continually mea*ured in the ar*** Hrvacl by the individu&l chann*l* and 2) th* alara or automatic action i* initiated when th*
radiation level trip **tpoint 1* uceed-4.
SALD -
tJllIT 2
- 3/4 3-1
.Amendment No. 121
Insert 1 The Trip Setpoints are the nominal values at which the bistables are set.
Any bistable is considered to be properly adjusted when the "as-left" value is within the band for CHANNEL CALIBRATION accuracy (i.e., +/-rack calibration+ comparator setting accuracy).
The Trip Setpoints used in the bistables are based on the analytical limits stated in the UFSAR.
The selection of these Trip Setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account.
To allow for calibration tolerances, instrumentation uncertainties, instrument drift, and severe environment errors for those Reactor Protection System (RPS) channels that must function in harsh environments as defined by 10 CFR 50.49, the Trip Setpoints and Allowable Values specified in the Technical Specification Limiting Conditions for Operation (LCO's) are conservatively adjusted with respect to the analytical limits.
The methodology used to calculate the Trip Setpoints is consistent with Instrument Society of America standard ISA-S67.04-1982, which is endorsed via NRC Regulatory Guide 1.105, Rev. 2.
The actual nominal Trip Setpoint entered into the bistable is more conservative than that specified by the Allowable Value to account for changes in random measurement errors detectable by a CHANNEL FUNCTIONAL TEST.
One example of such a change in measurement error is drift during the surveillance interval.
If the measured setpoint does not exceed the Allowable Value, the bistable is considered OPERABLE.
Setpoints in accordance with the Allowable Value ensure that the safety analyses which demonstrate that safety limits are not violated remain valid (provided the unit is operated within the LCO's at the onset of any design basis event and the equipment functions as designed).
The Trip Setpoints and Allowable Values listed in the LCO's incorporate all of the known uncertainties applicable for each channel.
The magnitudes of these uncertainties are factored into the determination of each Trip Setpoint.
All field sensors and signal processing equipment for these channels are assumed to operate within the allowances of these uncertainty magnitudes.
___J