GO2-16-104, LAR to Change EAL Scheme

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LAR to Change EAL Scheme
ML16214A374
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 07/28/2016
From:
Energy Northwest
To:
Office of Nuclear Reactor Regulation
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ML16211A338 List:
References
GO2-16-104
Download: ML16214A374 (40)
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4. Other (Site-Specific) Indtcat1ons This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the Fuel Clad barrier, including indications from containment air monitors or any other {site-specific) instrumentation.
5. Emergency Director Judgement This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to 110nitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgement that the barrier may be considered lost or potentially lost.

(See also IC SGl, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.) RCS BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6) The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

1. RCS Leak Rate The "Loss* EAL is based on design basis accident analyses which show that even if HSIV closure occurs within design limits, dose consequences offsite from a "puff* release would be in excess of 10 millirem. Thus, this EAL is included for consistency with the Alert emergency ch*ssification. The potential loss of RCS based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems.

Core uncovery is not a significant concern for a 50 gpm leak, however, break propagation leading to significantly larger loss of inventory is possible. Many BWRs may. be unable to measure an RCS leak of this size because the leak would likely increase drywell pressure above the drywell isolation set point. The system normally used to 1110nitor leakage is typically isolated as part of the drywell isolation and is therefore unavailable. If primary system leak rate information ts unavailable, other indicator~ of RCS leakage should be used. Potential loss of RCS based on primary system leakage outside the drywell is determined from site-specific alarms in the areas of the main steam line tunnel, main turbine generator, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.

2. Drywell Pressure The (site-specific) drywe11 pressure 1s based 'on the drywell high pressure alarm set point and indicates a LOCA.

A higher value may be used if supporting documentation is provided which indicates the chosen value is less than the pressure which would be reached for a SO gpm Reactor Coolant System Leak. There is no "Potential Loss" EAL corresponding to this item. 5-21

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4.

Does the main steam line break with isolation belong in the Fission Product

    • .. ~..

Barrier Chart? If the isolation occurred., there is no longer a barrier threat and should not be added to other "losses" for entry to a SAE or GE? Agreed, this condition should be removed from the FPB chart but must still be classified under system failures due to the probable offeite dose release from the puff release. S. Whv isn't scconda.rv containment failure considered a barrier loss? Secondary containment is not considered a true fusion product barrier. Its loss is addressed in Technical Specifications and failure to meet the action statement would result in an Unusual Event.

6.

If Drywell prcssmc exceeds the isolation setpoint due to a loss of drywell coolin2 (or inertin~ error), does this reQuire an Alert declaration? No - Jn general the FPB chart is built around a LOCA or slow developing accidents. The lcey qualifier is "accidents." Meeting a FPB chart trigger by itself should be handled under the applicable Technical Specification LCO. However, if there is an accident in progress, then an emergency declaration is wa"anted Individual plants may clarify the EAL conditions to ensure that non-accident conditions do not require emergency classification. C*. Page 17

WNP-2 6.2.1.1.3 Design Evaluation 6.2.1.1.3.l Summary Evaluation The key design parameters and the maximum calculated accident parameters for the pressure suppression containment are as follows: Design Cale. Accident Parameter Parameter Parameter

a.

Drywell pressure 45 psig 34.7 psig

b.

Drywell tempera-340°F 328°F ture

c.

Suppression cham-45 psig 27.6 psig ber pressure

d.

Suppression cham-275°F 220°F ber temperature The foregoing design and maximum calculated accident para-meters are not determined from a single accident event but from an envelope of accident conditions. As a result, there is no single design basis accident (DBA) for this containment system. A maximum drywell near the end of a accident (LOCA)

  • occurs for either steam line.

Both and suppression chamber pressure occurs blowdown phase of a loss-of-coolant-Approximately the same peak pressure the break of a recirculation line or a main accidents are evaluated. The most severe drywell temperature condition (peak tempera-ture and duration) occurs for a small primary system rupture above the reactor water level that results in the blowdown of reactor steam to the drywell (small steam break)

  • In order to demonstrate that breaks smaller than the rupture of the largest primary system pipe will not exceed the con-tainment design parameters, the containment system responses to an intermediate size liquid break and a small size steam break are evaluated.

The results show that the containment design conditions are not exceeded for these smaller break sizes. 6.2-5

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WNP-2 AMENDMENT NO. 47 November 1992 WNP-2, along with some other BWRs, proposed the use of the HPCS diesel as Alternate AC (Reference 1.5-27) as this third diesel is in excess of the number required to meet the minimum redundancy requirements (i.e., single failure for safe shutdown), (Reference 1.5-23, Appendix I, Q/A 105). While HPCS could provide decay heat removal during a SBO, it could not provide power to an emergency bus. Therefore, it was proposed that coping analyses would also be performed to address instrumentation, battery capacities, air supply and loss of HVAC. This approach has been found to be acceptable to the NRC although they disagreed that it should be classified as Alternate AC (Reference 1.5-24). 1.5.2.2 Evaluation to NUMARC 87-00 Chapters three through seven of NUMARC 87-00 contain guidance which, if implemented, would serve to satisfy the five NUMARC SBO initiatives and obtain compliance with the Station Blackout Rule, with the exception of Quality Assurance and Technical Specification considerations. The following sections address the content of these chapters. 1.5.2.2.1 Required Coping Duration Using the data listed in Table 1.5-1 and the methodology provided in Chapter 3 of NUMARC 87-00 the coping period for WNP-2 was determined to be four hours. 1.5.2.2.2 Station Blackout Response Procedures NUMARC 87-00 requires procedure changes to deal with:

1.

SBO Response. Changes in this category are to include; requiring priority efforts to restore ac power with site specific instructions on how this is to be accomplished, providing for core cooling and containment isolation, dealing with the loss of HVAC (e.g., the opening of control room cabinet doors), and the need to strip battery loads if necessary to conserve de power.

2.

AC Power Restoration. These changes are to provide guidance to operations and the load dispatcher concerning the course of actions to be taken to restore ac power.

3.

Severe Weather Guidelines. These changes are to provide guidance for operators to determine the proper course of action due to the onset of severe weather; in particular hurricanes.

1. 5-13
4.

As written, EAL l.b. ofSU6 would not allow the plant staff to notify offsite authorities of the declaration of the Unusual Event. It would be better if the EAL dealt with a degradation of offsite communications rather than a total loss. This would permit notification of the state and local authorities that a problem exists. The availability of one method of ordinary ojfsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when atraordinary means (relaying of information from radio transmissions, individuals being sent to offeite locations, etc.) are being utilized to make commu~cations possible. Licensees may choose to adopt more stringent thresholds for entry into the Unusual Event if desired

s.

For SU7, DC bus voltage may not be an indication available in the control room. Can alternative methods be used to determine the vulnerability of DC loads? If control room DC bus voltage indication or low voltage alarm is not available, local indication should be used The only alternative would be to observe DC-supplied loads for the effects of low voltage. This alternative, however, would defeat the anticipatory nature of the EAL as described in the basis and would therefore not be acceptable. 6a. Consider deleting SA3/EAL l.a. The statements pertaining to temperature can stand alone to correctly classify the event. 6b. If alternate cooling methods not required by Technical Specifications can be used to keep the plant in cold shutdown; is the declaration of an Alert necessary perSA3? Separate statements are included to recognize additional plant capability to maintain cooling of the reactor. A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary unplanned acursion above 200 "F when the heat removal function is available. Both statements are therefore necessary. fj) Page27

A/t.IJHA~~~SP- ~C77 SYSTEft MALEUNCJION GENERAL DtER&ENCY S&2 Failure of th* Reactor Protection Syst.. to Ca11pl1t1 an Autoaat1c Ser.. and Manual Scr111 was NOT Successful and There 1s Indication of an Extr... Challenge to the Ability to Cool the Core. OPERATING MOOE APPLICABILITY: Power Operation EXAMPLE Ef'ER&ENCY ACTION LEVEL:

1.

(Site-specific) indications exist that automatic and manual scram were not successful.

2.

Either of the following: BASIS

a. (Site-specific) indications exists that the core cooling is extremely challenged.

OR.

b. (Site-specific} indication exists that heat removal is extremely challenged.

Automatic and manual scram are not considered successful if action away from the reactor control console is required to scram the reactor. Under the conditions of this IC and its associated EALs, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as emergency boration in PWRs, or standby liquid control in BWRs, the continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence. For PWRs, the extreme challenge to the ability to cool the core is intended to mean that the core exit temperatures are at or approaching 12oo*F or that the reactor vessel water level is below the top of active fuel. For plants using CSFSTs, this EAL equates to a Core Cooling RED condition. For BWRs, the extreme challenge to the ability to cool the core is intended to mean that the reactor vessel water level is below 2/3 coverage of active fuel. Another consideration is the inability to initially remove heat during the early stages of this sequence. For PWRs, if emergency feedwater flow is insufficient to remove the amount of heat required by design from at least one steam generator, an extreme challenge should be considered to exist. For plants using CSFSTs, this EAL equates to a Heat Sink RED condition. For BWRs, (site-specific) considerations include inability to remove heat via the main 5-80

condenser, or via the suppression pool or torus (e.g., due to high pool water temperature. In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design {typically 3 to S~ power) a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time. 5-81

A '

7.

Must all control rods be inserted for a scram to be considered successful? If not, what is the threshold for detennining a partial scram to be adequate per SS21 For PWRs, the scram is considered unsuccessful when enough control rods have not inserted to cause the reactor power to fall below that percent power associated with the ability of the safety systems to remove heat and continue to decrease. This is typically between 3% and 6% rated thermal power. The wording of the EAL should include the concept that reactor power is below [X]% and decreasing under existing plant collditions. Subsequent actions necessary for the reactor to be prepared/or a cool down (normal boration to a xenon-free cold slrutdown boron concentration} and depressurization are not to be considered Jn SA2, the discriminator is whether the automatic reactor protection system.functioned If it did not or if reactor power was not decreased to below X% and decreasing due to insufficient rod insertion, an Alert is declared Note this assumes that the manual reactor trip function immediately available to the operators was effectively used to complete a successfi1l scram (power< [X}% and decreasing). Failure of the manual reactor trip function from the control room is the discriminator for the Site Area Emergency per SS2. Jn this situation, wording should be chosen to include the concept of "power not less than [X)% and decreasing and manual reactor trip function has not been effective. " If emergency boration is needed to bring the reactor to below X"A and decreasing, the scram was unsuccessful. For BWRs, a scram is considered unsuccessful if it does not result in achieving a state in which the reactor will remain shut down ullder all conditions without boron injection. For SA2, ifmarrual actions result in the reactor being shut down under all conditions without boron injection, an Alert is declared. Escalation to a Site Area Emergency (SS2) is not required If sufficient control rods are not inserted to reduce reactor power to below the APRM downscale setpoints, an immediate Site Area Emergency (SS2) is declared If the APRM downscale setpoint is achieved, but suppression pool temperature is greater than Boron Injection Temperature, a precursor exists for a threat to containment and thus a Site Area Emergency is warranted Page 28

DIC 1805.l Instrument Setpoint Change Request ISCR I 3 '-// WEA-RIS-14 Setpoint Change Prepared By/Date ~* 7 Reviewed By/Date /J ~"' JL ~Engineer Reviewed By/Date __ & ______ £p ____ -=-______ ..._I _fl-_~ )_-<t_f..__ Control System Engineer Reviewed By /Date Approved By/Date ~ System Engi eer Implementing Work Order ____ 'WR * ) 'f t!JOS~lt:/ I I j,~/°l 7 t I

ISCR WEA-RIS-14 Setpoint Change 1.0 FUNCTIONAL DESCRIPTION The purpose of WEA-RIS-14 is to monitor the release of radioactivity from the Radwaste building ventilation exhaust duct to the environment. This instrument provides input to the plant process computer, recorder PRM-RR-3 and annunciators. These alarms are as follows: a failure alarm on ANN 4.851.Sl-6.5 (low counts), a Hi alarm on ANN 4.602.A5-6.4, and a Hi Hi local alarm. Changing the setpoints will require a revision of the IMDS as well as the channel calibration procedure PPM16.3. l and the channel functional procedure 16.3.3. Operations should review this change as the Hi alarm point triggers operations to enter the emergency operating procedures. 2.0 SETPOINT AND ALLOWABLE VALUES WEA-RIS-14 has two setpoint values which will be changed, a HI alarm setpoint which will be changed to 200cps and a HI HI alarm setpoint which will be changed to 8000cps. Both alarms have tolerances determined by the procedure being performed, channel functional or channel calibration. The channel calibration, PPM 16.3.1, is the most restrictive the HI HI and the HI alarm point tolerances are +/-3 %. The channel functional, PPM 16.3.3 places the admin setpoint tolerances at +20% -17%. Current setpoints for WEA-RIS-14 are; HI HI 12000 CPM HI 6000 CPM New setpoints will be; (determined by the performance of PPM 12.11.5 HI HI 8000 CPM HI 200 CPM 3.0 DESIGN INPUTS AND BASIS REQUIREMENTS ODCM; 3.2.2 3.3 Table 6.1.2.1-1.5

4.0 REFERENCES

Page 2of3

ISCR WEA-RIS-14 Setpoint Change FSAR section 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS PPM 12.11.5 PPM 16.3.1 PPM 16.3.3 5.0 AFFECTED DOCUMENTS IMDS for WEA-RIS-14 PPM's 16.3.1, 16.3.3 Calculations: Page 3 of 3

USE CURRENT REVISION A ;u;;~~ p~~~TEM i WNP-2 PLANT PROCEDURES MANUAL PROCEDURE NUMBER APPROVED BY DATE

  • 12.11.5 GOS - Revision 2 10/01197 VOLUME NAME CHEMISTRY PROCEDURES SECTION SPECIAL CALCULATIONS TITLE RADIATION MONITOR SETPOINT CALCULATIONS PROCEDURE NUMBER REVISION PAGE 12.11.5 2

1 of 21

1.0 PURPOSE To provide instructions for determining the setpoint(s) for radiation monitors. 2.0 PREREQUISITES None

3.0 REFERENCES

3.1 Offsite Dose Calculation Manual 3.2 Setpoint Calculation E/1-02-92-1050 3.3 Setpoint Calculation E/I-02-92-1051 3.4 Setpoint Calculation E/I-02-94-1309 4.0 PRECAUTIONS AND LIMITATIONS None PROCEDURE NUMBER RE'vlSION PAGE 12.11.5 2 2 of 21

5.0 PROCEDURE 5.1 Due to the repetitious nature of these calculations, use of a spreadsheet is recommended. If one is used, the data may be transferred to the corresponding tables in this procedure, or a hard copy of the spreadsheet may be printed and attached. If a hardcopy is attached, mark the corresponding tables NIA, and reference the appropriate hardcopy.

5. 2 For liquid effluent monitor setpoint calculations, go to Step 5. 19.

5.3 Based on the following, determine the appropriate gas mixture to be used for this radiation monitor: 5.3.1 For the Reactor Building: PROCEDURE NUMBER 12.11.5

a.

The values from ODCM Table 3-15 are recommended, but a different mixture may be selected based on operational conditions or expected effluent activity.

b. The ODCM Table 3-15 values are for a design base mixture at 30 minutes decay, and produce a setpoint which is appropriate for operation of the offgas system in the charcoal bypass mode. The setpoint resulting from the use of these values is lower than the value obtained by using a gas mixture appropriate for off gas system operation with the charcoal beds in service.
c. If a gas mixture other than ODCM Table 3-15 is used, it may be necessary to limit the operation of the offgas system to require the use of the charcoal beds.
d. If a mixture other than ODCM Table 3-15 is used, document this fact, and the reasoning for its use, and attach to this procedure.
e.

The setpoint for the Reactor Building effluent is on the Low Range Monitor, PRM-RE-lA.

f.

X/Qj for the Reactor Building release pathway is 1. 6E-06 sec/m3, from ODCM Table 3-3.

g. Forty percent (40%) of the site boundary dose rate limit is allocated to the Reactor building, Fj is 0.4.
h. The maximum exhaust flow rate (Rj) is 4.48E+07 cc/sec (95,000 cfm).

REVISION PAGE 2 3 of 21

5.3.2 For the Turbine Building:

a.

The values from FSAR Table 11.3-7 are recommended, but a different mixture may be selected based on operational conditions or expected effluent activity.

b. If a mixture other than FSAR Table 11.3-7 is used, document this fact, and the reasoning for its use, and attach to this procedure.
c.

The setpoint for the Turbine Building effluent is on the Low Range Monitor, TEA-RIS-13.

d. X/Qj for the Turbine Building release pathway is 1.4E-05 sec/m3, from ODCM Table 3-3.
e.

Forty percent (40%) of the site boundary dose rate limit is allocated to the Turbine building. Fj is 0.4.

f.

The maximum exhaust flow rate (Rj) is 1. 79E+08 cc/sec (380,000 cfm). 5.3.3 for the Radwaste Building:

a.

The values from FSAR Table 11.3-7 are recommended, but a different mixture may be selected based on operational conditions or expected effluent activity.

b. If a mixture other than FSAR Table 11. 3-7 is used, document this fact, and the reasoning for its use, and attach to this procedure.
c.

The setpoint for the Radwaste Building effluent is on the Low Range Monitor, WEA-RIS-14.

d. X/Qj for the Radwaste Building release pathway is 1.4E-05 sec/m3, from ODCM Table 3-3.
e.

Twenty percent (20 % ) of the site boundary dose rate limit is allocated to the Radwaste building. Fj is 0.2.

f.

The maximum exhaust flow rate (Rj) is 3.91E+07 cc/sec (83,000 cfm). PROCEDURE NU/viBER REVISION PAGE 12.11.5 2 4 of 21

5.4 Determine the fraction for each nuclide of total activity released as follows: Fraction = Af.. 1t.. = _12. iJ M Tj where: 5.4. l 5.4.2 5.4.3 Nuclide Kr83m Kr85m Kr87 Kr88 Kr89 Xel31m Xe133m Xe133 Xel35m Xe135 Xel37 Xe138 PROCEDURE NU:MBER 12.11.5 Measured or estimated concentration of nuclide i in release path j (µCi/cc) Measured or estimated total concentration of all gases in release path j (µCi/cc) Enter the selected nuclide concentration or activity for each nuclide in the Mij column. Mark nuclides not present in the mixture NIA. Enter the selected total concentration or activity for each nuclide in the MTj column. Calculate the fraction for each applicable nuclide and record in the 7rij column. Table 5.4a Mij MTj 11"jj D c"'> 0 C> D c_) 0 c, c.> ~ C'J C) 0 C> 0 0 0 0 0 C> J.oc c: I ["..-0 P: . ) I. I CJ2.1=.,-r D 0 D '-/:)Or:_ I 'i--:;vE / 91fs£1 0 b 0 0 0 D REVISION PAGE 2 5 of 21

5.5 Determine the maximum release rate based bn whole body dose as follows: Fj500 Max release rate = QTj = --m---------- (µCi/sec) where: _x_ L (Ki) (1tij) Qj i=l Fj = Fraction of total dose allocated to this release pathway (Dimensionless) 500 = Whole body dose rate limit X/Qi = Maximum normalized diffusion coefficient for this pathway at and beyond the site boundary. (sec/m3) Ki = Total whole body dose factor due to gamma emission from nuclide i, as listed in the ODCM or Reg Guide 1.109, revision 1 (mrem/year per µCilm3) 1T'ij = Is the fraction of the total activity for nuclide i (Dimensionless) m = Total number of nuclides in the gaseous effluent j = Release pathway PROCEDURE Nlll\\IBER REVISION PAGE 12.11.5 2 6 of 21

5.5.1 Enter the 1t"ij values from Table 5.4a. Mark nuclides not present in the mixture NI A. 5.5.2 Multiply the 1t"ij value by its corresponding Ki, and record the result in the Ki'll"ij column. 5.5.3 Sum the values in the (11"ij)(Ku column. Table 5.5a Maximum Release Rate Nuclide 1t"ij Ki (1rij)(K) Kr83m 0 0.0756 C) Kr85m c:.> 1170 CJ Kr87 0 5920 {';; Kr88 C) 14700 0 Kr89 C) 16600 n Xel31m cJ 91.5 CJ Xel33m C) 251 CJ Xel33 /.ct:1C-/ 294 5 3)-i.? I Xel35m 0 3120 C> Xel35 c:;:/ i'-ti::-1 1810 /'15(£3 Xe137 c) 1420 0 Xe138 C) 8830 {) L 'Jl"ij*Ki= /.sJi53

5. 5. 4 Determine QTj by entering the appropriate values for the applicable effluent pathway in the spaces provided below and completing the indicated calculation:

(µCi/ sec) (c :- > soo = 't?utn (/*Ji=--) ) ( / 'i} i;: 1 ) (µCi/ sec) PROCEDURE NUMBER REVISION PAGE 12.11.5 2 7 of 21

5.6 Determine the maximum release rate based on skin dose as follows: Fj3000 Max release rate = Orj = --m--~------ (µCi/sec) i; ~ (Li + 1. lM) ( 1t ij) where: Fj = Fraction of total dose allocated to this release pathway (Dimel16ionless) 3000 = Skin dose limit X/Qi = Maximum normalized diffusion *coefficient for this pathway at and beyond the site boundary (sec/m3) Li = Skin dose factor due to beta emission from nuclide i, as listed in the ODCM or Reg Guide 1. 109, revision 1 (mrem/year per µCi/m3) Mi = Air dose factor due to gamma emission from nuclide i, as listed in the ODCM or Reg Guide 1.109, revision 1 (mrem/year per µCi/m3) 7rij = Is the fraction of the total activity for nuclide i (Dimensionless) m = Total number of nuclides in the gaseous effluent j = Release pathway PROCEDURE NUMBER REVISION PAGE 12.11.5 2 8 of 21

I 5.6.1 Enter the 7ru values from Table 5.4a. Mark nuclides not present in the mixture NI A. 5.6.2 Multiply the 7rij value by its corresponding Li + 1. lMi, and record the result in the (~+ l.1MJ1ru column. 5.6.3 Sum the values in the (7rij)(Li+ l. lMJ column. Table 5.6a Maximum Release Rate Nuclide 7rjj Li+l.lMi (7ru)(Li+ l. lMJ Kr83m 0 0 C> Kr85m c) 2.81E+03 c) Kr87 0 l.65E+04 0 Kr88 c> l.91E+04 c:) Kr89 (J 2.91E+04 c,J Xe13lm c""> 6.48E+02 c~ Xe133m c) l.35E+03 0 Xe133 J.')?2--B-I 6.94E+02 I.) r,, r:. 2-- Xe135m (? 4.41E+03r -~ '( ~-.,-, '(-'( } Xe135 y f~ ~ - I 2.30E+03 1 y .J 2..~IZ. 5 J.. ),, J.. .}_ ). Xe137 () l.39E+04 i) - Xe138 c' l.43E+04 C) L(7rij)(Li+ l. lMi) = 53s:i;;:; 5.6.4 Determine QTj by entering the appropriate values for the applicable effluent pathway in the spaces provided below, and completing the indicated calculation: (µCi/sec) = I :US"! (µCi/ sec) PROCEDURE NUMBER REVISION PAGE 12.11.5 2 9 of 21

5. 7 Compare the whole body and skin Orj values calculated above and select the lower of the two values. Record the selected value below:
5. 8 Determine the maximum allowable concentration as follows:

QT. Max concentration = c. = _J (µCi/ cc) TJ R. J where: Total allowed concentration of all noble gases (µCi/cc) Maximum acceptable release rate for all noble gases (µCi/sec) ~ = Effluent release rate (cc/sec) _ QTj _ ( i/.r;&r J ) _

c. - -

TJ Rj ( 3. ff E 7 ) 119?--f (µCi/ cc) 5.9 Determine the maximum allowable concentration for each nuclide as follows: Max concentrationi = cij =re ij CTj (µCi/ cc) where: 11"ij = As defined in Section 5.4 As defined above PROCEDURE NUMBER REVISION PAGE 12.11.5 2 10 of 21

5.9.1 Enter the 7rij values from Table 5.4a. Mark nuclides not present NIA. 5.9.2 Enter the Crj value from Step 5.9. 5.9.3 Multiply the applicable 7rij values by the CTj value, and record the result in the cij column. Table 5.9a Maximum Nuclide Concentration Nuclide 7rij CTj cij Kr83m 0 f.1ciF-'I (_, Kr85m c' CJ Kr87 rl ,,.-) Kr88 c' CJ Kr89 c-> CJ Xel3lm C' cJ Xe133m c u Xe133 l y) £ -/ ).1&>E-5' Xe135m 0 0 Xe135 i tK C-/ 9 7y e-5 Xe137 0 c_, Xe138 c <II"'" () PROCEDURE NUMBER REVISION PAGE 12.11.5 2 11 of 21

5.10 For gamma radiation detectors, (REA), determine the efficiency for the applicable nuclides as followS: 5.10.1 5.10.2 Nuclide Kr83m Kr85m Kr87 Kr88 Kr89 Xe13lm Xe133m Xe133 Xe135m Xe135 Xe137 Xe138 Enter the Xe133 efficiency in column for Exe-l33* Multiply the Relative Abundance for nuclides present in the mixture times the Exe-!33* and record the result in the t;j column. Mark nuclides not present in the mixture NI A. Table 5. lOa Relative Efficiency Gamma Relative Exe-133 Eij Abundance Abundance /'71'- 0.09 0.245 /(/;~, / \\ 0.74 2.016 0.837 2.281 \\ 1.372 3.738 \\ 1.6 4.360 \\ 0.0196 0.053 \\ 0.14 0.381 \\ 0.367 1.000 \\ 0.81 2.207 \\ 0.939 2.559 \\ 0.33 0.899 \\ 1.222 3.330 ~ 5.11 For beta radiation detectors, (TEA, WEA), determine the efficiency for the applicable nuclides as follows: 5.11.l Enter the Xe133 efficiency in column for Exe-l33* PROCEDURE NUMBER REVISION PAGE 12.11.5 2 12 of 21

5.11.2 Multiply the Relative Efficiency for nuclides present in the mixture times the Exe-m* and record the result in the Eu column. Mark nuclides not present in the mixture NIA. NOTE: Some nuclides may be present in the mixture th~t do not decay via beta emission. These nuclides are listed in the table belo~, but have a relative efficiency of zero. Table 5.1 la Relative Efficiency Nuclide Relative Efficiency1 l) i 'I' "( '< 'I' '< 'C\\ Kr83m 0.00233, t/Jo If 7 ~j Kr87 4.19 Kr88 2.05 Kr89 2 56 i--~~--+--------~+-----+------------~-+--~--f Xe131m 0.456 Xe133m 0.674 Xe133 LOO Xel35m ~9~233.... Xe135 ~ 1.93 i Xel37 '2.5i'l...l A} /fir-Xel 38 2.40 ~* (1) Setpoint Calculation E/I-02-94-1309 PROCEDURE NUMBER REVISION PAGE 12.11.5 2 13 of 21

5.12 Determine the count rate above background as follows: m Count Rate above Background = c. R j = L cij Eij ( cpm) i=l where: C.~ = Count rate above background (cpm) = = 5.12.1 5.12.2 5.12.3 Maximum acceptable concentration of noble gas i (µCi/cc) Effluent monitor detection efficiency for nuclide i (cpmlµCilcc) Enter the applicable Cij values from Table 5. 9a. Mark nuclides not present in the mixture NI A. Enter the applicable Eij values from Table 5. IOa. Mark nuclides not present in the mixture NIA. Multiply the applicable values in the Cij column times the values in the Eij column, and record in the C. Ry column. Mark nuclides not present in the mixture NIA. 5.13 Determine the C.~ value by summing the values in the C.Rij column. This is the setpoint count rate above background. PROCEDURE NUMBER REVISION PAGE 12.11.5 2 14 of 21

I Nuclide Kr83m Kr85m r87 Kr88 Kr89 Xel31m Xel33m Xe133 Xel35m Xe135 Xe137 Xe138 Table 5.12a Counts above Backgn;mnd ,At} A/. /r1 / I J ? '* r::: 7 ('J-*...Jl._J r.,,.,- I ~

  • 7 5.14 Based on available instrument data, select an operating background. Record the value below:

Background = _If-'-. _c_> __ 5.15 Calculate the instrument setpoint as follows: Setpoint = Background + C. Rj = ( Lt D ) + ( h*ZF-*?, ) = 't lt** > 5.16 Based on instrument range, accuracy, and the ability to accurately read instrument values, select a setpoint that is less than or equal to the setpoint value determined in Step 5.15. -~ ---? c;* - Selected Setpoint = __ Q'_* _,_0 5.17 If the selected setpoint is less than 80% of the calculated value, document the rational for this selection, and attach to this procedure. PROCEDURE NUMBER REVISION PAGE 12.11.5 2 15 of 21

5.18 Assign the selected setpoint to a specific alarm as follows: 5.18.1 PROCEDURE NUMBER 12.ll,5 For TEA-RIS-13 and WEA-RIS-14:

a.

Assign the selected setpoint to the High-High alarm.

b.

Determine the nominal, upper level and lower level vhlues for the High-High alarm as follows: NOTE: Since the High-High alarm setpoint is based on a regulatory limit, the upper level value must be LE than the selected setpoint. Upper level (UL) = Selected Value = '?? n-' Nominal value = (Selected Value)

  • 0.91 = J.:;co Lower Level (LL) = (Selected Value)
  • O. 82 =

7 2 I 6

c.

Determine the normal background value of the monitor based on instrument operating history. Background = ---~-0_, -------

d.

Multiply the background value by 4.66. High Alarm = (Background)(4.66) = ( 7' c )(4.66) = I J-C:

e.

Round the High Alarm !ill to the nearest interger value. High Alarm = 2 ° 0

f.

Assign this value to the High alarm setpoint, or if is desirable to assign a different value, document the justification for the different High alarm setpoint and attach to this procedure. g, Determine the nominal, upper level and lower level values for the High as follows: NOTE: The High alarm is not based on a regulatory limit, therefore it is acceptable to have values that are greater, or less than the selected

value, Upper level (UL) = (Selected Value)
  • 1.2 = ) '/{,

Nominal value = Selected Value = 1 C' {; Lower Level (LL) = (Selected Value)

  • 0.80 = /to REVISION PAGE 2

16 of 21

5.18.2 For PRM-RE-lB (REA):

a.

There is only one alarm on this system. Assign the selected value from Step 5.16 to the alarm setpoint..

b. Determine the nominal, upper level and lower level values as follows:

NOTE: Since this alarm setpoint is based on a regulatory limit, the upper level value must be LE than the selected setpoint. ,4/t Upper level (UL) = Selected Value = rtj/4 Nominal value = (Selected Value)

  • 0. 91 = ____ _

Lower Level (LL) = (Selected Value)

  • 0.82 = ---+---
5. 19 For liquid effluent monitor setpoint determinations, perform t 5.19. l 5.19.2 5.19.3 5.19.4 PROCEDURE NUMBER 12.11.5 From the instrument master data or instrument calibration records, determine the Cs137 efficienc or the monitor. Record value below:

Efficiency = ----+---- cpm/ µCi/cc or cps/ µCi/cc From instrument perfo ance data, determine the instrument background. Record below: Background = -----'.------ cpm or cps Calculate the maximum coun rate above background as follows: Cmax = = (2. OE-05) (-----1-) = ----- Calculate the maximum setpoint a follows: NOTE: An additional factor of 0. is included to reduce the maximum setpoint to 80 percent of 1 MPC. his is in accordance with ODCM Section 2. 9. Smax = (Cmax + Ba round)(O. 8) Smax =.._ _____ + ____ _,(0.8) Smax = ----'----- REVISION PAGE 2 17 of 21

5.19.5 5.19.6 PROCEDURE NUJ\\IBER 12.11.5 Based on instrument operating patameters and accuracy, select a monitor setpoint that is less than or equal ito the maximum value Smax determined above. If this value is less than 80% of the Smax value, attach a justificatio to this procedure. Record the selected value below: At Setpoint = -+:7._,1 ..... P~ 1 -=---- NOTE: The adjustment the service water radiation monitor setpoints is due the Reg Guide 1. 97, po accident, requirement to remain operable during and after an accident.

  • results in a substantial reduction in these setpoints that is not required for ot er effluent monitors.

For SW-RIS-604 and SW-RIS-605, adj t the setpoint for potential instrument error as follows:

a.

Convert the calculated setpoint to a volt ge value as follows: SetpointVoltage = lO[log(Setpoi t+l)) 7 Setpoint Voltage = _,,__ ___ _ Table EPN SW-RIS-604 SW-RIS-605 (1) Voltages determined by calculations E/1-02-91-1050 and E/1-02-91-1051.

b. Subtract the Correction Voltage, from Table 5.19.6a abov Corrected Voltage = Setpoint Voltage -

Corrected Voltage = --~-- - Correction Voltage Corrected Voltage = =<..:=---- REVISION PAGE 2 18 of 21

PROCEDURE NUl.IBER 12.11.5

c.

Convert the corrected voltage back into cps as follows: ( [ (Corrected Voltage) *(..2.)J) Corrected cps = 10 10 -1 , AC Corrected cps = 711 1

d. Select a value less. th~ equal to the corrected cps. '

Selected value = ___ __.;:::,_

e.

Determine the nominal, upper lev and lower level values as follows: NOTE: Since this alarm setpoint is ba d on a regulatory limit, the upper level value must be LE than the se cted setpoint. Upper level (UL) =Selected Value = --~- Nominal value = (Selected Value)

  • 0.91 = ___

Lower Level (LL) = (Selected Value)

  • 0.82 =

£

f.

Assign the above determined values to the High-High alarm.

g. Select a value for the High setpoint at a point above the normal background variation, and below the High-High alarm setpoint. This value should be set reasonably close to the normal operating value, but far enough above it to minimize spurious alarms. The selected value should be assigned to the Lower level point. The Nominal and Upper Level values should be 111 % and 122 % of the LL, respectively.

REVISION PAGE 2 19 of 21

5.19.7 PROCEDURE NU:MBER 12.11.5 For TSW-RIS-5:

a.

Determine the nominal, upper level and lower level values as follows: NOTE: Since this alarm setpoint is based on a regulatory limit, the upper level value must be LE than the selected setpoint. AL Upper level (UL) = Selected Value = 't: Nominal value = (Selected Value)

  • 0.91 = ----'"----

Lower Level (LL) = (Selected Value)

  • 0.82 = ----...,1-----
b. Assign the above determined values to the High-Hi alarm.
c.

Determine the normal background val instrument operating history. f the monitor based on Background = --------+-----

d.

Multiply the background value by 4. High Alarm = (Background)(4.66) = ( (4.66) = ___ High Alarm = --1----

f.

Assign this value t the High alarm setpoint, or if is desirable to assign a different value, do ent the justification for the different High alarm setpoint and attac

g.

Determine the nominal, upper level as follows: lower level values for the High NOTE: The High alarm is not based on a regufa~ory limit, therefore it is acceptable to have values that are greater, or less than the selected value. Upper level (UL) = (Selected Value)

  • 1.2 = _....__ __

/ /.// Nominal value = Selected Value = _ __.. __ _ Lower Level (LL) =(Selected Value)* 0.80 ~,,,,,_. ___ _ REv1SION PAGE 2 20 of 21

6.0 DOCUMENTATION Maintain the completed procedure in the permanent plant file in accordance with the appropriate record procedure(s). 7.0 ATTACHMENTS None PROCEDURE NUMBER REVISION PAGE 12.11.5 2 21 of 21

*--*-~---,.*-*-------~ ---*-'""""'" *-* --*-**-----"--

Wt:A-R1s-ll/ s~;::>onvTS PPYn I Z.. I/. S"" GA l<:..J.JLA r1 tJrv~ J-1, !-11 N!J/Yll N/TL uppEe.. l()~ g_ooE3 ¥, 7'1 e.$ /;2.1'1 -110% -. g % N!)IY'l !Nit L tlfPlf/e LPW£Je 2.0t>/F2 2.4C>Z. l,,.6()1:FZ +/- zo~~ JM})S C/il 181'<..19 /l()N 561" Pt1/N'/5. = /VtJ#Jl/l/A L +/- 3 % M ~ Nt9/?JIN-4 l -= 8..POG3 ~ - ~s:. = L.J3. 3 3 H ~ t./f¥JfTl'Z..

. ¥. oo tf3 -1
  • l> 3
  • 8: ()()ES = 3. 2 l.f l:F 3 '7 l"'h..

ff. 2 '{ tF 5 ~ - ~ / 37.. ~ 3 i-{ ?!:: wwez ""<i..tJO!i.5-.o3-8',0oG3 = ~ 7bE s $--'~

7. 76 3 ~

~ ~ s : I ZCJ. 3 3 l1.:e; I( 1 N1Jm11VAL = 2-oo ~ * :;s * :J-3 3./-/. ::- l!pptR2 = 200+. ()3. *2co = 206 y,....__ J/H./J.S 1-/1 /-/,

c.

W'- .LI 206 '°M_" ~ ,S. ': 3. t/3 1 {, -i: Lllwce = 200-.03-200 = 19'-f ~ }'VI. /CJ'(5;;,. - bo.s = 3.23 H-2:- A.bm 1 N $ 7 Pt;l7V7S -::: NIQ7 /NA -120 % - 17 % N~m1NA L -= ~tX>E3 ~ - ;,.s = 133. 3 3 H <<tP?Z.== ~ooe:~ -f, 2(?.oof"3) ~ 9.6t;c~ :Y~ a ~ 7.6DEi3 m.

  • 60.S

~ /tx:::J H~ L~wez. = S.OOG3 -,, I ((8-0DF.fj ::: ~,.bC( E 3 7~ 6.bt/63~ - 6os =- J/D,67 H~

J-1, NJn?rlVAt..-::: zoo 7'/"Y\\.

== 3... 33 H-:: llppEIZ.: 20D+~z(Zoo) : 240 Y-~ 24D~- ~.s = ft f-l:e~ LIJWfF/2 :: 2 ()0 -: J 7(2o:=>) :: I 66 7"' r'Y\\. yY\\. J J /6b~

  • 60s

-== 2~ 7? rte:

EMERGY WNP-2 DIC: 1801.l NORTHWEST INSTRUMENTMASTERDATASHEET Page 1 of3 System Engineer EPN: WEA-RIS-14 SERVICE: Radwaste Bldg. Ventilation Monitor Loop: WEA-RIS-14 Rev. #28 Date: 9116f 99 Work Order: Input From : WEA-RE-14 ~ Cal Procedure: CC: 16.2.1RC:16.2.2 CFT: 16.2.3 Output To: PRM-RR-3, Alarms INS1RUMENT CALIBRATION P ARAME1ERS Process Desired Output and Loop Check Data Input Tolerance CPS Meter CPM PMR-RR-3 CHAN. 3 & TDAS x408 +/- 10% (.9 to I.I) CPM +/- 10% (.9 to 1.1) .167 10 1 +/-.2 1.67 100 10 +/- 2 167 lOK .IK +/- 20 16.7K 106 IK +/- 200 167K 10' IOOK +/- 20K Bi-Stable Parameters - Limits Switch ID Seto int/ Action at Reset Tolerance Tolerance Setpoint Kl See Addenda Close Record See Addenda K2 See Addenda Close Record See Addenda K3 See Addenda Open Record See Addenda Notes: Power Supply: PP-7 A-Z Ckt. 20 VIA BD-RAD-24 Fuse 2TB 110-4 Setpoint reference in ODCM Section 3.6 ISCR #106, 128, 173, 318, 394, 515, 706, 717, 973 wea-ris-14.r28.doc Plant Computer (A0834) mv+/- 6.5 mv 32 53.3 96 138.7 160 Logic Affected High-High Alarm High Alarm Failure Alarm

EMERGY WNP-2 DIC: 1801.1 *. NORTH WEST . INSTRUMENT MASTER DATA SHEET Page 2 of 3 EPN: WEA-RIS-14 SER VICE: Radwaste Bldg. Ventilation Manito Loop: WEA-RIS-14 Rev. #28 Date: 9/16/99 Addenda: I. ODCM Table 6. l.2. l. l-1.5al Radioactive Gaseous Effluent Monitoring Instrumentation, Radwaste Building Ventilation Exhaust Noble Gas Activity monitor, Low Range

2. Failure alarm (switch K3) activates on any one of the following:

a) High Voltage !NOP b) Loss of Instrument Power c) Low Detector Counts. NOTE: Due to low background levels, this portion of the failure alarm has been INIIlBITED until a live zero is provided. "FAIL TRIP ADJUST" potentiometer R69 should be adjusted to provide a voltage= -10 to -15 VDC at TP-10 d) Contact change of state detected by PRM-RR-3 which in tum causes ann. Drop 4.851.Sl-6.5

3.

Annunciator, 4.602.A5-6.4, (RW Bldg. Vent Rad High), actuated by this switch (K2) is used by Operations personnel to signify required entry into Emergency Operating Procedures. If chan to alarm setpoints are made, an evaluation must be made by operations to determine is changes t Emergency Operating Procedures are required.

4.

Settings controlled by POC approval of PPM 16.3.2 wea-ris-14. r28. doc

EIWERGY NORTHWEST Page 3 of 3 EPN: WEA-RIS-14 Loop: WEA-RIS-14 TOL IUGH +/-5V WNP-2 DIC: 1801.l INSTRUMENT MASTER DATA SHEET SER VICE: Radwaste Bldg. Ventilation Monitor Rev. #28 Date: 9/16/99 OPERATING PARAMETERS OPERATING PARAMETERS (VOLTS DC) PR.EV COMPLETION date: 04-20..95 UL: 1030 NEW VALUES PER PPM 16.2.2 COMP, DATE:---- INITIAL: ___ UL: VOLTAGE DVM Only 1025 VDC U..: LL: 1020 HV INOP +/- 15V (Inputs t<J IO) I>VM Only 975 UL:990 LL: 960 UL: __ ___,__VDC LOWER LEVEL DISC +/- 5 "lo UPPER LEVEL DISC +/-5"/o SWITCH NO. cc TOL IUGH VOLTAGE +/-3% HVINOP (Inputs to KJ) +/-3% UPPER LEVEL DISC SWITCH NO. Kl m-m Alarm K2 Ill Alarm K3 DWNSCL wea-ris-l 4.r28.doc ..0.6 -12.6 UL:..0.63 LL:..0.57 UL: *13.2 .LL:-12.0 CALIBRATION SETPOINT PARAMETERS METER OPERATING PARAMETERS (CPM) (VOLTS DC) info only UL: 137.33 8000 (I) 133.33 U..: 129.33 UL: 3.43 200 (I) 3.33 U..: 3.23 Requires live (D) See Note 2 zero CFf/ADMIN SETPOINT PARAMETERS LL: UL: _____ VDC ~---- LL: UL: ___ -~ ........ -vnc LL: NEW VALUES PER PPM 16.2.2 COMP.DATE: INITIAL: UL: VDC U..: UL: VDC U..: 0 SeeNote3 CFr METER TRIP SETPOINT TOL (CPM) (HZ) info only +20% UL: 160.00 -17% 8000 (I) 133.33 U..: 110.67 +20% UL: 4.00 -17"/o 200 (1)3.33 LL: 2.77 Requires live zero (D) See Note 2}}

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