Text

Proposed Tech Specs Re Secondary Containment & SGTS to Reflect Revised Secondary Containment Drawdown & post- Accident Analyses Results
	ML17292A551
Person / Time
Site:	Columbia
Issue date:	10/15/1996
From:	; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	;
Shared Package
ML17292A550	List: ML17292A549; ML17292A551;
References
	NUDOCS 9610230248
	Download: ML17292A551 (109)
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Attachment 3A Page 1 of 8

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CONTAINMENT SYSTEMS 3/4.6.5

~ ~ SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1 SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and ".

ACTION:

Without SECONDARY CONTAINMENT INTEGRITY:

.a. In OPERATIONAL CONDITION 1, 2, or 3, restore SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In OPERATIONAL CONDITION ", suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.6.5.1 SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

77a.lu.f~d e ng ry co t ng nt is g ess tha o e al to .2 inc o vy u wa er/ a ge

b. Verifying at least once per 31 days that:

1. All secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the secondary containment, is closed.

3. All secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic isolation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers/valves secured in position.

C. At least once per 18 months:

l. Verifying that andby gas tre ment subsystem will draw 20 MiSRu+~ ~ down the secondary containment to greater than or equal to 0.2 inch of vacuum water auge in less than or equal to and ca~ m ul'r

2. Operating andby gas treatment subsystem for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and maintaining greater than or equal to 0.25 inch of vacuum water gauge in th'e secondary containment at a flow rate not exceeding 2240 cfm.

>rradlated fuel is being handled in the secondary containment and during

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When CORE ALTERATIONS and operations with a potential for draining the reactor

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vessel. ~

WASHINGTON NUCLEAR - UNIT 2 3/4,6" 38 Amendment No.

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'Pbi0230248 qbiOi5 PDR ADOCK 05000597 I PDR

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Attachment 3A Page 2 of 8 CONTAINMENT SYSTEMS STAND BY GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and ".

ACTION:

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-a. With one stan by gas treatment subsystem inoperable, restore the l"gu'Ir~

inoperable subsystem to OPERABLE status within 7 days, or:

1. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. In OPERATIONAL CONDITION ", suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provi-sions o S eci ication 3.0.3 are not applicable.

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b. With both st y gas treatment subsystems inoperable in OPERATIONAL CONDITION ", suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS or operations with a potential for draining the reactor'essel: The provisions of Specification 3.0.3.

are not applicable.

SURVEILLANCE RE UIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the

. subsystem operates for a least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERABLE...

"When irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

WASHINGTON NUCLEAR - UNIT 2 3/4 6-41 Amendment No. l

k II Attachment 3A Page 3 of 8 CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subs stem by:

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l. Verifying that the s s em satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0 05X and uses the test procedure guidance in Regulatory Posi-tions C.5. a, C.5.c, and C.5.d of Regulat Guide 1.52, Revision 2, March 1978, at a system flow a e of cfm + 10X.

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2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than Oe175X; and Foe

3. Verifying a subsystem flow rate of cfm + lOX during system operation when tested in accordance with ANSI N510-1980.

C. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, .

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than Oe175X.

d. At least once per 18 months by:

Verifying that the pressure drop across .the combined HEPA filters and charcoal adsorber banks is less than 8 inches water cfm + 10K. ~ r~uirad e while opera in the filter train at a flow rate of subsysf~

2. Verifying that f r starts and solation dampers open on each of the o lowing test signals.

a. Manual initiation from the control room, and

b. Simulated automatic initiation signal.

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3. Verifying that the filter cooling bypass dampers can be manually opened and the an can be manually started.

Verifying that the heaters dissipate 20.7 + 2.1 kW when tested in accordance with ANSI N510-1980.

WASHINGTON NUCLEAR - UNIT 2 3/4 6-42 Amendment Nn. gj I

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Attachment 3A CONTAINMENT SYSTEMS Page 4 of 8 SURVEILLANCE RE UIREMENTS Continued

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 while operating the system at a flow rate of cfm + 10".

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f. After each complete or partial replacement of a charcoal adsorber bank by verifying that. the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a halogenated hydrocarbon r igerant test gas while operating the system at a flow rate of cfm + 10K.

5000 WASHINGTON NUCLEAR - UNIT 2 3/4 6-43 Amendment No.

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Attachment 3A CONTAINMENT SYSTEMS Pages of 8 BASES DEPRESSURIZATION SYSTEMS (Continued)

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety/relief valve inadvertently opens or sticks open. As a minimum this action shall include: (I) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety/relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety/relief valve to assure mixing and uniformity of energy insertion to the pool.

3 4.6.3 PR MARY CONTAINHENT ISOLATION VALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures for those isolation valves designed to close automatically that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3 4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell and between the reactor building and suppression chamber. This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. There are nine pairs of valves to provide redundancy and capacity so that operation may continue indefinitely with no more than two pairs of vacuum breakers inoperable in the closed position.

3 4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

Add 2ttscv'l 2 MASHINGTON NUCLEAR UNIT 2 B 3/4 6-5 Revision No.

<<gl Attachment 3A CONTAINMENT SYSTEMS Page 6 of 8 BASES Add SECONDARY CONTAINMENT (Continued) Zff5Lvi 2 The OPERABILITY of the standby gas treatment systems. ensures that sufficient iodine removal capability wi11 be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses. Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months during each 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

3 4.6.6 PRIM RV CONTAINMENT ATMOSPHERE CONTRO The OPERABILITY of the systems required for the detection and control of oxygen/hydrogen gas ensures that these systems will be available to maintain the oxygen concentration within the primary containment below the lower oxygen limit for oxygen/hydrogen mixture during post-LOCA conditions. Either drywell and suppression chamber oxygen/hydrogen recombiner system is capable of controlling the expected oxygen generation associated with radiolytic decomposition of water. The oxygen/hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA," September 1976.

Following an accident the inerted primary containment oxygen level is controlled to not exceed 4.8% volume with the catalytic recombiner system. By FSAR Figure 6.2-26 the containment will reach 4.8% oxygen approximately 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months after the accident if either recombiner system is operating.

To provide assurance that recombiners are capable of achieving the required oxygen removal, the feed and effluent streams will be sampled during surveillance testing to establish that the effluent hydrogen concentration is less than 25 ppm by volume for a feed of at least 1% hydrogen by volume. This will confirm a minimum efficiency of 99.75% for the expected range of post-accident conditions. This efficiency will be adequate to maintain the post-accident oxygen level below 4.8% by volume.

The CAC system employs a platinum on alumina catalyst to recombine the oxygen and hydrogen flow from the containment. During accident conditions, the gas mixture is preheated to approximately 450 to 550'F prior to entering the catalyst. This preheat increases the effectiveness of the hydrogen/oxygen recombination because it limits the potential for bed poisoning. In the test configuration, the blower is used as the only source of gas stream heating and the catalyst preheaters are not energized. The blowers are capable of heating the gas stream by compression. Temperatures at the blower exit are limited for test purposes to approximately 300'F due to the blower gas exit temperature trip setpoint.

WASHINGTON NUCLEAR UNIT 2 B 3/4 6-6 Revision No. 0

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REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 3A Page 7 of 8

~Ine 1 Secondary containment is considered OPERABLE and standby gas treatment (SGT) filtration can be credited when a differential pressure of 0.25 inch of vacuum water gauge is established. The time to establish this differential pressure is a function of SGT flow rate and secondary containment in-leakage, in addition to outside meteorological conditions, initial secondary containment temperature and humidity, standby service water temperature, and a postulated single failure of an electrical division. With an in-leakage of 2240 cfm at static conditions (no wind or temperature effects), the worst case secondary containment drawdown time occurs at O', no wind, a failure of the Division 1 diesel generator, 77'F standby service water temperature, and a secondary containment temperature of 50'F and relative humidity of 0%.

The design basis accident dose analysis assumes a secondary containment drawdown time of 20 minutes. No SGT filtration is credited for 20 minutes in the analysis.

The SGT flow rate must be greater than or equal to 5000 cfm within 2 minutes of an initiation signal. This is verified as part of the 18 month surveillance that verifies the filter train starts and the isolation dampers open on a manual and simulated automatic initiation signal. The 2 minute time period includes the time for. the emergency diesel generator to start, the SGT start sequence to occur, and an automatic swap-over to the lag subsystem assuming a failure of the lead subsystem. No SGT flow is credited for 2 minutes in the secondary containment drawdown analysis.

The secondary containment in-leakage is primarily a function of seams in the secondary containment superstructure and leakage through doorways. The differential pressure generally increases linearly with flow through the seams and doorways. An equation was derived from correlated performance curves to provide a method for accurately determining secondary containment differential pressure versus SGT flow rate relative to the 2240 cfm in-leakage limit.

The equation is as follows (Flow in cfm and ~P in inches water gauge):

Flow = 1988.3*~P + 3485.85"V~P The constant for the square root term in the equation is increased such that a differential pressure of 0.25 inch vacuum water gauge would result in a flow rate of 2240 cfm. The constant for the square root term was also conservatively increased to account for any in-leakage due to holes or widening of the seams in secondary containment which would generally result in quadratic losses rather than linear losses. This equation can be used to generate a curve to accurately establish an acceptable region for secondary containment differential pressure versus SGT flow rate relative to the 2240 cfm in-leakage limit. The 2240 cfm in-leakage SURVEILLANCE REQUIREMENT can be satisfied by measuring the secondary containment differential pressure at any SGT system flow rate greater than 2240 cfm and verifying that the measured differential pressure when corrected for outside wind and temperature effects is at a greater vacuum (larger negative pressure) than that shown on the curve for the measured SGT flow rate. A representative curve is included in the appropriate implementing procedures.

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 3A Page 8 of 8 Secondary containment must be maintained at a vacuum during normal operation to ensure secondary containment effluent is monitored as well as following an accident for accident mitigation. In the event that the normal secondary containment ventilation system is secured, secondary containment could become pressurized such that the maximum accident design basis pressure of 0 inch water gauge on all surfaces of secondary containment is exceeded. Therefore, the SGT system must be started for secondary containment to be considered OPERABLE when the normal secondary containment ventilation system is secured.

~In ~r2 The SGT system is designed to filter the effluent gases from secondary containment through a 99% efficient filter train to mitigate the post accident effects of a severe accident within primary containment or other postulated accidents outside primary containment. A secondary function of the SGT system is to filter the purge exhaust from primary containment whenever radiation levels within the primary containment exceed acceptable levels for direct purge to the environment via the reactor building exhaust system. The SGT system consists of two filter trains. The active system components in each filter train that are required for post-accident operation are redundant and configured in two subsystems to allow for lead/lag operation. The lag subsystem within a train may be considered as a backup to the lead subsystem, with either subsystem being capable of supporting the associated train in the performance of its required functions, In the event that the lead subsystem in a train fails to establish proper air flow within a set time delay from an SGT system start signal, the lead subsystem will automatically shutdown and the lag subsystem will start.

Only the lead fan and heater (including the support equipment and controls), or the lag fan and heater (including the support equipment and controls) in each filter train are required for OPERABILITY of each SGT filter train. However, in order to meet the requirement that two independent SGT subsystems be OPERABLE, both lead subsystems or both lag subsystems are required to be OPERABLE (i.e., the two independent subsystems required by the LCO must be powered from separate buses).

Inser Each required SGT subsystem will be operated for at least one hour every 18 months to ensure the system performs as designed.

~ f ~ II " ~ I N f II

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS ATTACHMENT3B ITS and Bases Markup B

PCIVs Page 1 of 20 3.6.1.3 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FRE(UENCY SR 3.6.1.3.7 Verify each automatic PCIV actuates to 24 months the isolation position on an actual or simulated isolation signal.

SR 3.6. 1.3.8 Verify each EFCV actuates to the 24 months isolation position on an actual or simulated instrument line break signal.

SR 3.6. 1.3.9 Remove and test the explosive squib from 24 months on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS SR 3.6. 1.3. 10 Verify the combined leakage rate for all In accordance secondary c ainment bypass leakage with the

'a'8 scfh when pressurized to r

paths is < Primary

>P.

Containment Leakage Rate Testing Program SR 3.6. 1.3. 11 Verify leakage rate through each MSIV is In accordance

< 11.5 scfh when tested at > 25.0 psig. with the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.12 Verify combined leakage rate through In accordance hydrostatically tested lines that with the penetrate the primary containment is Primary within limits. Containment Leakage Rate Testing Program MNP-2 3.6-15 Revision B

Page 2 of 20 Secondary Containment 3.6.4.1 ACTIONS cont i nued CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment C.1 --------NOTE---------

inoperable during LCO 3.0.3 is.not movement of irradiated applicable.

fuel assemblies in the secondary containment, during CORE Suspend movement of Immedi ately ALTERATIONS, or during irradiated fuel OPDRVs. assemblies in the secondary containment.

AND C.2 Suspend CORE Immediately ALTERATIONS.

AND C.3 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY r SR 3. .4. 1.1 Ver fy

.25 inc co ar of co ac m ai en ter vac um gau e.

s SR 3.6.4. 1. Verify all secondary containment 31 days equipment hatches are closed and sealed.

(continued)

WNP-2 3.6-39 Revision B

Page 3 of 20 Secondary Containment 3.6.4.1 SURVEILLANCE REgUIREHENTS continued SURVEILLANCE FRE(UENCY SR 3.6.4.1. Verify each secondary containment access 31 days inner door or each secondary containment access outer door in each access opening is closed.

quired SR 3.6.4.1. Verify each standby gas treatment (SGT) 24 months on a subsystem will draw down the secondary STAGGERED TEST containment to > 0.25 inch of vacuum BASIS water gauge in <

20 equi nv'fc.s vcguivad SR 3.6.4.1. Verify each SGT subsystem can maintain 24 months on a

> 0.25 inch of vacuum water gauge in the STAGGERED TEST secondary containment for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at ao~ BASIS flow rate < 2240 cfm. i~- luakag~

WNP-2 3.6-40 Revision

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B Page 4 of 20 SGT System 3.6.4.3

3. 6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System inCk~Ck 4 LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, Dur ing CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME V+dlV t ~vivid A. One SG subsystem A. 1 Restore SGT subsystem 7 days inoperable. to OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met in MODE 1, 2, ol 3. 8.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Required Action and ------------NOTE-------------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during r uvW movement of irradiated C.1 Place OPE LE SGT Immedi atel y fuel assemblies in the subsystem in secondary containment, operation.

during CORE ALTERATIONS, or during OR OPDRVs.

(continued)

WNP-2 3.6-45 Revision

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B Page 5 of 20 SGT System 3.6.4.3 ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend movement of Immedi ately irradiated fuel assemblies in the secondary containment.

AND C.2.2 Suspend CORE Immedi ately ALTERATIONS.

AND C.2.3 Initiate action to Immediately suspend OPDRVs.

(guised D. Two SGT subsystems D. 1 Enter LCO 3.0.3. Immedi ately inoperable in MODE 1, 2, ol 3~

Yap i V"d.

NOTE---------

E. Two SGT subsystems E.1 inoperable during LCO 3.0.3 is not movement of irradiated applicable.

fuel assemblies in the secondary containment, during CORE Suspend movement of Immediately ALTERATIONS, or during irradiated fuel OPDRVs. assemblies in the secondary containment.

AND E.2 Suspend CORE Immedi ately ALTERATIONS.

AND E.3 Initiate action to Immedi ate 1 y suspend OPDRVs.

MNP-2 3.6-46 Revision

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Attachment 3B Page 6 of 20 SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY V+tfiYCd SR 3.6.4.3. 1 Operate each SGT subsystem for 31 days

> 10 continuous hours with heaters operating.

SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

v~uir W SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 months actual or simulated initiation signal.

goiv~d SR 3.6.4.3.4 Verify each SGT filter cooling 24 months recirculation, valve can be opened and the fan started.

MNP-2 3.6-47 Revision

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B PCIVs Page 7 of 20 8 3.6.1.3 BASES APPLICABLE The single failure criterion required to be imposed in the SAFETY ANALYSES conduct of unit safety analyses was considered in the (continued) original design of the primary containment purge valves.

Two valves in series on each purge line provide assurance that both the supply and exhaust lines could be isolated even if a single failure occurred.

PCIVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 3).

LCO PCIVs form a part of the primary containment boundary. The PCIV safety function is related to minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during a OBA.

The power operated, automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal. Mhile the reactor building-to-suppression chamber vacuum breakers isolate primary containment penetrations, they are excluded from this Specification. Controls on their isolation function are adequately addressed in LCO 3.6. 1.6, "Reactor Building-to-Suppression Chamber Vacuum Breakers." The valves covered by this LCO are listed with their associated stroke times in Reference 4.

The normally closed PCIVs are considered OPERABLE when manual valves are closed or open in accordance with appropriate administrative controls, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves and devices are those listed in iS Refe nce 4 S drostatsca y teste va ve ea age exempt from ype C testing limits and must meet specific Cha NSl llew eaka e rate re uirements an secondary containment bypass valves must meet additiona eakage rate requirements.

Other PCIV leakage rates are addressed by LCO 3.6. 1.1, "Primary Containment," as Type B or C testing.

This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents.

(continued)

MNP-2 B 3.6-16 Revision

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"E B Secondary Containment B 3.6.4.1 Page 8 of 20 BASES APPLICABLE Secondary containment satisfies Criterion 3 of the NRC SAFETY ANALYSES Policy Statement (Ref. 3).

(continued)

LCO An OPERABLE secondary containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be diluted and processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.

APPLICABILITY In MODES I, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 Qr 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES I, 2, and 3. This time period ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal .

continued MNP-2 B 3.6-78 Revision@9

lt B Secondary Containment Page 9 of 20 B 3.6.4.1 BASES (continued)

SURVEILLANCE SR . .4. .1 RE(UIREMENTS h s R nsu es ha th se ond ry ont inm nt foun ar is s f ci w'tl le k gh to pre ud ex ilt ati h u der e p cte nd tio s. Th 24 our Fre uen y t is R as de elo ed as d o op rat ng xpe ien e r at d se on ary con ai en va uum ari ti s rin t e a pli ab

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e H ES'nd th low pro abi ity of DB o cur in tw en urv il anc s.

Fur he or , te 2 ho r F equ ncy is ons'der d eq at in vi ~

w ot er ind'cat'ons avai ab e i th co tro ro m,

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i clo iog al ma, to lec th op cat r t a ah oem]l s co dar co tai men va uum con iti n.

SR 3.6.4.1 and SR 3.6.4.1 Verifying that secondary containment equipment hatches and each inner access door or each outer access door in each access opening are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. SR 3.6.4.1.2 also requires equipment hatches to be sealed. In this application, the term "sealed" has no connotation of leak tightness. Haintaining secondary containment OPERABILITY requires verifying all inner doors or all outer doors in the access opening are closed. However, each secondary containment access door is normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being performed on an access. The 31 day Frequency for these SRs has been shown to be adequate based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

SR 3.6.4. 1. and SR 3.6.4.1 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.

To ensure that all fission products are treated, SR 3.6.4.1.4 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the lowest postulated pressure continued MNP-2 B 3.6-80 Revision

I B Secondary Containment Page 10 of 20 B 3.6.4.1 BASES 4,

Add SURVEILLANCE SR 3.6.4. 1. and SR 3.6.4. 1 (continued) Zk1sc,r P~

RE(UIREHENTS external to the secondary containment boundary. This is confirmed by demonstrating that one SGT subsystem will draw down the secondar containment to > 0.25 inches of vacuum 20 81irtO ks water gauge in s s . This cannot be accomplished i the e t i ment boundar is not intact.

'iilib SR .6.4.1. emonstra es that eac SG subsys em can main a)n 0.25 inches of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate s 2240 cfm. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months test period allows d y i t i t d state conditions. e e ese two tests are used to 7-ensure secondary containmen bou ary in egri y. ince these SRs are secondary containment tests, they need not be performed with each SGT subsystem. The GT subs stems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, either GT subsystem will perform this test. Operating experience as shown these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. Voiv W REFERENCES 1. FSAR, Sections 15.6.5 and 15.F.6.

2. FSAR, Section 15.7.4.

3. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

WNP-2 B 3.6-81 Revision

3, l

I' B SGT System Page 11 of 20 B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident'(DBA) are filtered and adsorbed prior to exhausting to the environment. Fi/6 Vras'~s >

The SGT System consists o two fully redundant s each with its own o network, dampers, r f e and controls. pipe ~,y, We~f 2 Each filter train consists of (components listed in order of the direction of the air flow):

a. A moisture separator; %ad

b. Two electric heater banks (one r and one

/a9 c );

c. A prefilter bank;

d. A high efficiency particulate air (HEPA) filter bank;

e. Two charcoal adsorber banks;

f. second HEPA filter bank; and /a

'nd A

g. Two centrifugal fans (one one } each with inlet flow control vanes.

The sizing of the SGT System equipment and components is based on the results of an infiltration analysis, as wel as an exfiltration analysis. The internal pressure of the boundary region is maintained at a negative pressure o 0.25 inch water au e when th ystem is in operation, which represents the internal pressure required to ensure zero exfiltration of air from the building using the 95K meteorological data.

continued WNP-2 B 3.6-89 Revision B SGT System Page 12 of 20 B 3.6.4.3 BASES BACKGROUND The moisture separator is provided to remove entrained water (continued) in the air, while the electric heaters reduce the relative humidity of the airstream to less than 701. (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter is provided to remove fine particulate matter and protect the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter is provided to collect any carbon fines exhausted from the charcoal adsorber.

The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident Au, Ized Fa~ that could re uire o eration of the system. Following In CacA initiation, f subsystem starts. SGT S stem flow are controlle automatically by modulating inlet vanes Add installed on the SGT fans. . Msu.f 3 APPLICABLE The design basis for the SGT System is to mitigate the SAFETY ANALYSES consequences of a loss of coolant accident and fuel handling accidents (Refs. 3 and 4). For all events analyzed, the SGT System is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment.

The SGT System satisfies Criterion 3 of the NRC Policy Statement (Ref. 5).

LCO Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous ind~~wd~C releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subs stem Add in even o a s n ac ive ai ure. In ad ition, on y Z~savt 4- e primary elec ric heater ank and centrifugal fan are re uired for OPERABILITY of each SGT subs stem.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

continued WNP-2 B 3.6-90 Revision

\

a B SGT System Page 13 of 20 8 3.6.4.3 BASES APPLICABILITY In MODES 4 and 5, the probability and consequences of these (continued) events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System OPERABLE is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.l rucpi v cd With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABI E redundant SGT subsystem and the low probability of a DBA occurring during this period.

B.l and B.2 If the SGT subsystem cannot be resto} ed to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.l C.2.1 C.2.2 and C.2.3 During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRYs, when Required Action A. 1 cannot be completed within the re uired Com letion Time th PERABLE SGT subsystem vguircd s ou e immediately placed in operation. This Required continued WNP-2 B 3.6-91 Revision

A P

B SGT System Page 14 of 20 B 3.6.4.3 BASES ACTIONS C.l C.2.1 C.2.2 and C.2.3 (continued)

Action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation will occur, and that any other failure would be readily detected.

An alternative to Required Action C.I is to immediately suspend activities that represent a potential for releasing radioactive material to the secondary containment, thus placing the unit in a condition that minimizes risk. If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of'a vessel draindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable.

If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

D.l Y'd (Pl I dd If both SGT subsystems are inoperable in MODE I, 2, or 3, the SGT System may not be capable of supporting the required radioactive release control function. Therefore, actions are required to enter LCO 3.0.3 immediately.

E. 1 E.2 and E.3 veau/red When two GT subsystems ALTERATIONS and movement are inoperable, if applicable, of irradiated fuel assemblies in CORE the secondary containment must be immediately suspended.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if continued WNP-2 B 3.6-92 Revision

Ip v s ~

ygl F

4 jl B SGT System Page 15 of 20 B 3.6.4.3 BASES ACTIONS l 1 E 2 , d E ( t; d) applicable, actions must be immediately initiated to suspend

.OPDRVs to minimize the probability of a vessel dr aindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action E.l has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE I, 2, or 3, the fuel movement is independent of reactor oper ations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 I"Cgui rW REQUIREMENTS Operating (from the control room) each SGT subsystem for z 10 continuous hours ensures that both su sys ems are OPERABLE and that all associated controls are functioning proper ly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on (automatic heater cycling to maintain temperature) for z 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6). The YFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specified test frequencies and additional information are discussed in detail in the VFTP.

continued WNP-2 B 3.6-93 Revision B SGT System Page 16 of 20 B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 Vgoivad . Add REQUIREMENTS ZHSLr7 5 (continued) This SR requires verification that each SGT subsystem starts upon receipt of an actual or simulated initiation signal.

The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. Mhile this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.4.3.4 I-cgoived This SR requires verification that the 'GT cooling recirculation valve can be opene an the filter fan started. This ensures that the ventilation mo e o GT System operation is available. Mhile this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES .

l. 10 CFR 50, Appendix A, GDC 41.

2. FSAR, Section 6.5.1.2.

3. FSAR, Sections 15.6.5 and 15.F.6.

4. FSAR, Section 15.7.4.

5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

6. Regulatory Guide 1.52, Rev. 2.

WP-2 8 3.6-94 Revision

Programs and Manuals B 5.5 Page 17 of 20 5.5 Programs and Hanuals 5.5.7 Ventilation Filter Testin Pro ram VFTP (continued) after any structural maintenance on the system housing; and, following significant painting, fire, or chemical release in any ventilation zone communicating with the system while it is in operation.

Tests described in Specification 5.5.7.c shall be performed once per 24 months; after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation; after any structural maintenance on the system housing; and, following significant painting, fire, or chemical release in any ventilation zone communicating with the system while it is in operation.

Tests described in Specification 5.5.7.d and 5.5.7.e shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.

a. Oemonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.054 when tested in accordance with Regulatory Guide 1.52, Revision 2, Section C.S.c and ANSI N510-1989, Section 10 at the system flowrate specified below:

ESF Ventilation System Flowrate (cfm)

SGT System /

CREF System 900 to 1100 45oo Vo 5SQO

b. Oemonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, Section C.5.d and ANSI N510-1989, Section 11 at the system flowrate specified below:

ESF Ventilation System Fl owrate (cfm)

SGT System t 4 CREF System 900 to 110 45oo A 5'5'oo continued WNP,-2 5.0-12 Revision

lt"~~

's I

B Page 18 of 20 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Fil ter Testin Pro ram VFTP (continued)

C. Oemonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, Section C.6.b shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTH 03803-1986 (Method B for the SGT System and Method A for the CREF System) at a relative humidity greater than or equal to the value specified below:

ESF Ventilation'System Penetration (5) RH (X)

SGT System 0.175 70 CREF System 1.0 70

d. Oemonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:

ESF Ventilation System Oelta P Flowrate (inches wg) (cfm)

SGT System < 8 9 CREF System < 6 900 to 1100 oo *5500

e. Oemonstrate that the heaters for each of the ESF systems dissipate the nominal value specified below when tested in accordance with ANSI N510-1989, Section 14.5.1:

ESF Ventilation System Wattage (kW)

SGT System 18.6 to 22.8 CREF System 4.5 to 5.5 5.5.8 x losive Gas and Stora e Tank Radioactivit Honitorin Pro ram This program provides controls for potentially explosive gas mixtures contained in the Hain Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

continued WNP-2 5.0-13 Revision

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHMCAL SPECIFICATIONS Attachment 3B Page 19 of 20 Inner I The time to establish the required vacuum in secondary containment is a function of SGT flow rate and secondary containment in-leakage, in addition to outside meteorological conditions, initial secondary containment temperature and humidity, standby service water temperature, and a postulated single failure of an electrical division. With an in-leakage of 2240 cfm at static conditions (no wind or temperature effects), the worst case secondary containment drawdown time occurs at O', no wind, a failure of the Division 1 diesel generator, 77'F standby service water temperature, and a secondary containment temperature of 50'F and relative humidity of 0%. The design basis accident dose analysis assumes a secondary containment drawdown time of 20 minutes. No SGT filtration is credited for 20 minutes in the analysis.

The secondary containment in-leakage is primarily a function of seams in the secondary containment superstructure and leakage through doorways. The differential pressure generally increases linearly with flow through the seams and doorways. An equation was derived from correlated performance curves to provide a method for accurately determining secondary containment differential pressure versus SGT flow rate relative to the 2240 cfm in-leakage limit.

The equation is as follows (Flow in cfm and ~P in inches water gauge);

Flow = 1988.3*aP + 3485.85*VIP The constant for the square root term in the equation is increased such that a differential pressure of 0.25 inch vacuum water gauge would result in a flow rate of 2240 cfm..The constant for the square root term was also conservatively increased to account for any in-leakage due to holes or widening of the seams in secondary containment seams which would generally result in quadratic losses rather than linear losses. This equation can be used to generate a curve to accurately establish an acceptable region for secondary containment differential pressure versus SGT flow rate relative to the 2240 cfm in-leakage limit. SR 3.6.4.1.3 verifies secondary containment boundary integrity by measuring the secondary containment differential pressure at any SGT system flow rate greater than 2240 cfm and verifying that the measured differential pressure when corrected for outside wind and temperature effects is at a greater vacuum (larger negative pressure) than that shown on the curve for the measured SGT flow rate. The measured differential pressure must be greater than the curve in ~ 20 minutes. A representative curve is included in the appropriate implementing procedures.

I~nert 2 The active system components in each filter train that are required for post-accident operation are redundant and configured in two subsystems to allow for lead/lag operation. The lead subsystem of each train is powered from a separate emergency diesel divisional bus (Division 1 or 2) than the lag subsystem. The lag subsystem within a train may be considered as a backup to the lead subsystem, with either subsystem being capable of supporting the associated train in the performance of its required functions.

IU PA O'

'.,ik3'

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 3B Page 20 of 20 Insert 3 In the event that the lead subsystem in a train fails to establish proper air flow within a set time delay from an SGT system start, the lead subsystem will automatically shutdown and the lag subsystem will start.

Insert 4 In addition, only the lead fan and heater (including the support equipment and controls), or the lag fan and heater (including the support equipment and controls) in each filter train are required for OPERABILITY of each SGT filter train. However, in order to meet the requirement that two independent SGT subsystems be OPERABLE, both lead subsystems or both lag subsystems are required to be OPERABLE (i,e., the two independent subsystems required by the LCO must be powered from the separate buses).

Inser~

This Surveillance also ensures that the SGT subsystem fan achieves

~

a 5000 cfm within 2 minutes. ~

Insert 6 In addition, secondary containment must be maintained at a vacuum during normal operation to ensure secondary containment effluent is monitored as well as following an accident for accident mitigation. In the event that the normal secondary containment ventilation system is secured, secondary containment could become pressurized such that the maximum accident design basis pressure of 0 inch water gauge on all surfaces of secondary containment is exceeded. Therefore, the SGT system must be started for secondary containment to be considered OPERABLE when the normal secondary containment ventilation system is secured.

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS ATTACHME<NT4 JCO (Revision 5) and Closure Summary

Attachment 4 Page 1 of 23 ATI NF R NCR 292-171, (JCO Rev. 5)

NCR 292-171, (JCO Rev. 4)

NCR 292-171, (JCO Rev. 3)

NCR 292-171, (JCO Rev. 2)

PER 290-017, (JCO Rev. 1)

NCR 288-357, (JCO Rev. 0)

. 1.0 oductio d om nent Tdentifica on Niagara Mohawk Corporation filed an LER on NMP-2 with the NRC in mid-1987 following discovery that assumptions used to evaluate secondary containment differential pressure drawdown time following a postulated LOOP/LOCA were not conservative. Vpon review of WNP-2 calculations of drawdown time, it was found that the WNP-2 analysis was also nonconservative. Like NMP-2, an assumed failure of certain emergency power bdses cause delay in the ability to achieve the negative secondary containment differential pressure. Further, the original WNP-2 analysis does not consider adverse environmental conditions that increase secondary containment leakage.

2.0 T~ue NCR 288-357, (JCO Rev. 0 and Rev. 1) originally documented a problem associated with ability of the standby gas treatment system for WNP-2 to effectively perform its safety function under all postulated post-accident conditions including consideration of adverse environmental conditions. Revision 2 and 3 to the original Rev. 0 and 1 assessments were required to address new issues which have been identified dealing with instrument calibration accuracy and setpoint error. Revision 4 was made to expand the SW temperatures shown in Figure 4 down to 32'F. Revision 4 was done to eliminate any pmcedure compliance issues at SW temperatures below 40'F.

Revision 5 considers reduced room cooler performance due to fouling to 65%

efficiency. A review of open BCO/JCO items (reference lOM SS2-PE-93-260) noted that revision 4 of this JCO considered 100% room cooler efficiency while the BCOs for NCRs 292-0476/0941/0950 use 65% efficiency. To correct the discrepancy, the

'I'I 4 N

Attachment 4 Page 2 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 secondary containment drawdown profile model was run with the reduced efficiency.

In Figure 1 the new draw down profile shows that -.25" w.g. is reached in less than 10 minutes. The model was also found to include the running heat load from the REA and RQA fans during the event, although no credit was assumed in analysis for these fans to aid in secondary containment drawdown since they trip on the F, A, and Z signals. The F, A, and Z signals also start SGT. When the heat loads of the REA and ROA fans were removed from the model, the secondary containment drawdown time is still less than ten minutes with the SW pond at 77 'F. Since the SW pond

temperature restrictions are no longer necessary, Figure 4 has been removed from this JCO. Plant procedures will reflect this change after JCO approval by POC by removing PPM 2.4.5 rev 23 Attachment 6.4, 7.0.0 rev 27 Att. 9. 10, 73.1 rev 18 Att. 9.4, and 7.0.2 xev 16 Att. 9.7.

Revision 5 also deletes references to future actions which would preclude single failure of the SGT inlet air valves. A Basic Design Change has been implemented to place the inlet air valves in a normally open condition.

/

Specifically, standby gas treatment flow indicator/recorders (SGT-FR-2A1, 2A2, 2B1, 2B2) are calibrated to read fiow in SCFM when fiow is at 212'F, 2 psig and 0% relative humidity. During any event in which SGT is initiated, these conditions are never reached. Consequently, for the range of conditions under which SGT is designed to operate, the actual standard flow through the SGT trains will be 5%

lower to 10% higher than indicated on the flow recorder.

In addition to this calibration correction factor, the SGT fan flow controller limiter) loop has a conservatively calculated accuracy of ~ 10%. Removing conservatism could lead to an accuracy of between g5 and +7%.

4 The effect of these two corrections is that the surveillance test results for system low and building leakage could have been 20% higher than previously believed.

This new information poses two problems. First, the higher total system flow (up to 20% higher) could have resulted in a trip of the SGT fan motors on thermal overloads ifthe fans had been called upon to operate under degraded voltage conditions. And second, the building leakage could have been up to 20% higher than

Attachment 4 Page 3 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 assumed in the secondary containment drawdown analysis resulting in. potentially under predicting the time to reach a -.25" w.g. building pressure.

In light of the above considerations, additional constraints must be placed on the plant to maintain the JCO basis. The analysis for JCO Rev 0 and Rev 1 required the SGT fiow limiter to be set at 5600 CFM to achieve a building drawdown to -0.25" w.g. within ten minutes to meet 10CFR100 and GDC 19 dose limits.

With the new information available, a new flow limiter setpoint (based on setpoint calculation E/I-02-91-1066) has;been established which considers degraded voltage, charcoal bed residence time and adverse environmental conditions. A summary of the setpoint calculation is provided below:

Analytical limit = 5812 ICFM (indicated cubic feet per minute = ICFM)

Correction for setpoint accuracy (nominal setpoint) = 5378 ICFM (rounded to 5380 ICFM)

Correction for flow element ICFM to ACFM (nominal inlet conditions', 90 F, 0%

R.H., 0.2 psig) = 5857 ACFM Lower setpoint limit = 4944 ICFM correction for flow element ICFM to ACFM (90'F, 0% RH, 0.2 psig) = 5385 ACFM From the setpoint calculation, the correction factor for flow limiter accuracy considers the effect of calibration error, instrument drift and environmental effects.

Calibration error and drift make up approximately 50% of the total loop setpoint

'0'F inlet is based on a 15'F temperature rise across the SGT heater coils and a 75 F nominal secondary containment temperature at the beginning of the accident. This condition represents the expected minimum building temperature during normal full power operation.

0% RH and 14.89 psia represent the most limiting combination of effected SGT operating conditions at the SGT Qow element and their use results in the most conservative correction factor (i.e., lowest Qow) per Calculation NE-02-92-06 for converting ICFiifto ACIM

Attachment 4 Page 4 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 accuracy with the remaining 50% attributed to environmental effects and other uncertainties.

Taking these accuracies into account in establishing limits on building leakage and available SGT system flow results in the following two limiting combinations of leakage and system flow:

g~e /~2 GRe&

(ACFM) (ACFM) (ACFM)

SGT fan flow = 5850 5385 53&5 (rounded from 5857)

Leakage = 1625 1475 1625 The Case 1 combination assumes the controller error is causing the flow limiter to control flow higher than the nominal setpoint. Case 2 assumes the controller error is causing the flow limiter to control and indicate flow lower than the nominal setpoint.

In both these cases, the system low capacity and associated leakage are derived based on the assumption that typically the total loop calibration and drift portion of the setpoint error act in only one direction for a given calibration period. On the other hand, environmental portion is assumed to act in the direction which provides the worse condition for the drawdown analysis; i.e., the error acts to reduce total system flow capacity and increase total building leakage . The assumption on calibration and drift error is considered valid based on historical trends observed during instrument calibration.

As a final consideration, Case 3 (Figure 3) was run assuming all instrument errors act to provide the worse case combination of building leakage and available SGT flow; i.e., 5385 ACFM SGT flow with the building leakage test inaccuracy in the opposite direction of 1625 ACFM. Even for this case, the SGT system is able to achieve building drawdown with single train SGT operation maintaining a negative 0.25" w.g. pressure differential and limiting dose consequences to less than This assumption is taken since building leakage is veriied annually to be less than 1475 CFM (adjusted for flow element calibration and standard conditions) and the accuracy of the flow limiterfrndicator could be as much as 10% high or low during the test. Subsequent system initiation during an accident (at some later time) could result in a diBerent environmental error being introduced into the system.

Attachment 4 Page 5 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 10CFR100 limits. Case 3 is not considered'a credible combination of SGT flow and building leakage, but is included to demonstrate acceptable consequences even in this unlikely condition. Cases 1 and 2 represent the worse case expected range of flow and leakage for the SGT and secondary containment.

Figures 1 and 2 (Case 1 and 2 respectively) present the revised drawdown analysis results and include consideration for the new flow limiter setting and range of possible secondary containment leakages.

Com onent afet Function The SGT system and secondary containment act to minimize and control radiological releases from the plant. Unfiltered release of primary containment leakage, and other radioactive gasses and particulates resulting from accidents outside the primary containment, is prevented by maintaining the secondary containment pressure negative 0.25" w.g. with respect to atmospheric pressure, and by filtering the efQuent gasses from the secondary containment. Secondary containment pressure boundary integrity is assured by periodically testing in-leakage and SGT capability.

The SGT system is required to have the Qow and pressure head capacity to maintain secondary containment at a -0.25" w.g. differential pressure. High efficiency filters and charcoal bed filters process the SGT flow. This capability must be maintained assuming a post-accident single active failure and design allowance to accommodate expected building accident environment and external meteorology.

WNP-2 Technical Specification requirements is secondary containment in-leakage maintained below 2240 cfm at -0.25" w.g. differential. The SGT is also used under certain conditions to filter primary containment purge gasses.

4.0 ustification For ontinued eration The critical parameters related to accomplishing the safety function of the SGT system and secondary containment are summarized in Table 1 for the original licensing bases, the 288-357 JCO, Rev. 1 and 2 and the new JCO (Revision 5 to

'CR 292-171). As identified in the Rev. 1 to the 288-357 JCO, standby gas treatment, secondary containment, standby service water, and weather modeling show that during post-LOCA, under certain adverse weather conditions, differential

'I Attachment 4 Page 6 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 pressure of the secondary containment does not'always meet the original licensing commitments. Certain combinations of post-LOCA single active failures and winter weather conditions act to minimize SGT performance, and increase secondary containment leakage. Wind increases the demand on the SGT to hold the leeward side of the Reactor Building sufficiently negative while simultaneously increasing the differential pressure, and thus the in-leakage, on the windward side of the building.

Differential temperature between the inside and outside of the building creates a differential pressure gradient from the bottom to the top of the secondary containment due to the density difference of the air inside and outside the building.

As a result, the lower portion of the building must be held at a high differential pressure (up to -0.75" w.g.) to assure that a negative 0.25" w.g. differential exists at the building roof line. This overall greater differential pressure proportionally increases building in-leakage. The net result of wind and winter temperatures is the inability to hold the upper portion of the secondary containment at a -0.25" w.g.

differential in cold weather, and to lengthen the time required to reach -0.25" w.g.

differential in moderate weather.

Steady state secondary containment pressure is established when building in-leakage is equal to standby gas treatment system (SGT) flow and secondary containment temperature transients have stabilized. Analysis shows that the time required to reach the steady state condition is a function of the assumed meteorological conditions at the time of a postulated LOCA, the assumed type of single active failure coincident with the LOCA, and the standby service water (SW) temperature. The transient analysis clearly indicates that the limiting single active failure is the assumed loss of one SGT train. Based upon single train design basis SGT flow and maximum Technical Specification allowable secondary containment leakage, the upper-most surface areas of the reactor building cannot be maintained at a quarter-inch water gage negative pressure with respect to atmospheric pressure. High SW water temperature acts to extend the time required to reach a steady state condition, but does not effect the final steady state pressure. However, there is no added restriction to maximum Tech Spec SW pond temperature as a result of this JCO (Tech Spec maximum temperature remains at 77'P).

Both SGT trains are redundantly powered by both divisions of standby emergency power. Consequently, the SGT is not susceptible to most of the single active failures that have relatively high probability of occurrence (e.g.; failure of an emergency diesel generator to start.) System design reviews completed since the

Attachment 4 Page 7 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 issuance of the Revision 1 JCO indicate that SGT is vulnerable to only a few single failures; specifically,

1) Electrical separation - hot shorts disabling active function of SGT valves and fans simultaneous in both divisions (14 valves and 2 fans affected).

2) A failure of either SGT train due to a single failure opening of the associated fire protection water deluge supply valve.

3) Lack of control logic to detect and transfer SGT operation to the redundant division on failure of a heater coil (i.e., SGT humidity control).

r Thus, from a failure analysis perspective, the SGT has features that provide more reliable operation than minimum design requirements would normally dictate.

However, even in the event only one SGT train is available, the new flow limiter setting of 5380 ICFM (5850 ACFM to 5385 ACFM) coupled with maintaining SW at or below 77'F and low secondary containment leakage ensures secondary I containment drawdown within ten minutes and limit dose releases to within the 10CFR100 and 10CFR50 GDC 19 limits.

Testing conducted for SGT flow/differential pressure capability, and testing of secondary containment integrity show that the SGT is capable of performing beyond design basis requirements, and that the secondary containment is significantly more leak-tight than required by Technical Specifications. Actions have been taken since this issue was first identified to further tighten the secondary containment boundary against leakage. Building leakage tests conducted over the last five years have measured building leakage at less than the 1475 ACFM.

Since the issuance of the original NCR 288-337 JCO, there has been a significant amount of engineering analyses and several engineering evaluations completed on the SGT performance and design. The intent of these further studies was to establish a long-term solution to the SGT issue and address concerns raised on the original JCO bases. This further work has resulted in a better building drawdown model, a plan to make the SGT system single failure proof (eliminadon of the three single failures identified above) and more accurate dose projection computer model. These enhanced calculation models and the aspect that SGT has limited susceptibility to single failures, have been integrated with the original analysis provided in the Rev. 0 and Rev. 1 JCOs of NCR 288-357. More detailed discussions of each of the changes from the original Rev. 0 and 1 JCOs bases are provided below.

Attachment 4 Page 8 of 23 J~tiQcation for Continued Operation, Rev. 5 NCR 292-171 4.1 8 ildin D wd wnM de Di erenc The WNP-2 secondary containment JCOs submitted to NRC on 9/89, discussed in 1/16/90 NRC staff meeting, and summarized in LER 89-040-01, had a number of basic assumptions that have been modified during the development of a change to the SGT design basis as new knowledge was added to this issue. Specifically, 4.1.1 Leaka e A um n: At the NRC meeting on 1/16/90, the NRC staff expressed concern on the assumption of uniform leakage over all secondary containment surfaces in calculating the drawdown time to a achieve -1/4" w.g. at all secondary containment surfaces. Due to NRC concern, WNP-2.created a revised model which splits secondary containment leakage into two portions. The first leakage parameter is leakage at ~de elevation while the second leakage parameter is leakage at the sheet metal siding. Due to the relative small number of penetrations through the WNP-2 concrete secondary containment structure (441'levation to 606'10" Elevation) the portion of leakage from these areas is expected to be small. Due to the large number of siding seams (over 30,000 feet) in the upper WNP-2 secondary containment structure (Elevation 606'10" to Elevation 669'), the portion of leakage from the siding is expected to be the majority of WNP-2 Technical Specification allowable leakage.

A special series of tests (PPM 8.3. 177) were conducted during the WNP-2 R-5 outage to characterize leakage. The results of these tests indicate that the majority of the leakage was at the upper siding seams.

For this JCO a high leakage at grade during winter (the most limiting design condition) was used and is conservative with regard to building drawdown calculation. The grade leakage in the analysis was set at 40% of the total leakage (although actual test data does not support significant leakage at grade).

4.1.2 Envir nment Tem erature At the NRC meeting on 1/16/90, the environment temperature assumptions in the analysis were discussed. The JCO submitted on 9/89 used the lowest average monthly temperature recorded in the WNP-2 FSAR. At the 1/16/90 NRC meeting, WNP-2 presented the concept of using 95% joint frequency meteorology data of wind and temperature as a design basis for secondary containment design. Since offsite dose consequences are based on 95% joint frequency meteorology of wind and stability, per R.G. 1.145, it seemed justified to apply a similar approach to secondary containment design.

The NRC staff did not voice concern or criticism with that approach in the

Attachment 4 Page 9 of 23 j'ustification for Continued Operation, Rev. 5 NCR 292-171 1/16/90 meeting or in response to information submitted on the-docket, therefore WNP-2 has included the 95% meteorology of wind and temperature into the analysis assumptions.

4.2 ~il 'I D As discussed above, there are very few single failures which could result in the loss of a SGT train. Consequently, we believe for all credible accident conditions, both

- SGT trains willbe operable, providing more than adequate flow to achieve a rapid building drawdown. Nevertheless this JCO basis is single SGT train operation.

In addition, to impmve SGT performance capability, operating pmcednrea have been revised to provide guidance to detect and switch heaters upon detection of high SGT humidity. These actions will compensate for the lack of automatic control logic to detect and transfer SGT operation to the redundant division on failure of a heater coil. An alarm and indication of SGT humidity is available in the control room to provide necessary information for operator action.

I 4.3 Do e Pro ec' de The original WNP-2 dose calculation model has been enhanced and expanded to support evaluations associated with the development of a change to the SGT design basis.

4.3.1 Buildin Drawd wn Time: The original WNP-2 model had limited ability to evaluate the dose impact of various drawdown times to achieve -1/4" w.g.

(i.e., time interval when SGTs filtering efficiency was equal to zero). The new dose analysis model is a FORTRAN code,,PADD, (Post-Accident Design Dose Model) which allows the user to specify the loss of SGT filtering efficiency for any time interval during the initial radioactivity release assumed to occur for an accident.

The PADD computer code is also a more precise calculation of dose due to improvements made in the method for calculating total dose (i.e., exact numerical solution versus average dose over large time steps).

4.3.2 ff- ite D e Im act: The PADD computer code also calculates off-site dose consequences (whole body and thyroid) for the exclusion area, low population zone, and control room as a result of changes in secondary

I i

Attachment 4 Page 10 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 containment drawdown times. The PADD code was revised to include the off-site dose calculations.

4.4 Stand ervice Water m erature Limi ti n The normal Technical Specification limit on Standby Service Water pond temperature is 77'F. During development of revision 5, the model was found to assume that the REA and ROA fans continued to run during the event, however, these fans trip on the F, A, and, Z signals. The F, A and Z signals also start SGT.

When the heat loads of the REA and ROA fans were removed from the model, it was shown that the secondary containment drawdown time is still less than ten minutes with the SW pond at 77 'F. Since the SW pond temperature restrictions are no longer necessary, Figure 4 is being removed from this JCO. YAP-2 is removing the added restrictions placed on the Tech Spec SW pond temperature.

Figure 1 presents the drawdown analysis for a SW pond temperature of 77'F.

4.5 lant Ac 'on to Su o J Ba i l) Maintain SGT flow limiter setting of 5380 ICFM (5850 to 5385 ACFM) to account for inaccuracies in SGT flow element and SGT flow element limiter loop and achieve secondary containment drawdown within ten minutes (see Table 2 for dose projections).

2) Maintain operator action to protect against heater coil failure (i.e., switch to lag fan if lead heater coil is inoperable).

5.0 Qgnclusion Given the current state of secondary containment integrity, the SGT can provide adequate differential pressure control with an adequate margin applied for variations in secondary containment leak-tightness and SGT flow performance. The fiow limiter setting of SGT (5380 ICFM) will ensure, even assuming a single failure, that secondary containment drawdown to -0.25" w.g. can be achieved within ten minutes.

The secondary containment pressure differential will remain greater than -0.25" 10

Attachment 4 Page 11 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 under the 95% meteorological conditions. Calculations show that both off-site and on-site post-accident doses remain well below 10CFR100 limits, 10CFR50 GDC 19, and are not significantly different than the results documented in the%NP-2 JCO which the NRC has reviewed and approved (1/3/90). Results of the drawdown model building response and the new dose projections are provided in Figure 1 and Table 2, respectively.

'Originator/Date POC Meeting Number POC Chairman/Date

'The parameter was qualitatively considered in the safety evaluation to have no impact on the operability of SGT under all design basis accident conditions. However, this evaluation was not documented.

11

Attachment 4 Page 12 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 dditionai R vielle Engineering PSE LD Sharp lp5 ng Date

~ gg'~G (f.

GW Brastad Date Safety Analysis R Vosburgh, Licensing e$ 1 s 0/Iq/qq JE Bekhazi Date Operations KL Fantbaugh 12

Attachment 4 Page 13 of 23 Justification for Continued Operation, Rev. 5 NCR 292-171 Table Rev. 0, JCO &, (Rcv. 2)

~O' ID R . I JCO Rev 5 JCO Key Parameters R.G. 13 Release Assumptions Secondary W.G. -.25" W.G. -.25" W.G.

Containment (worso location) (worse location)

RB Prcsmro Drawdown Time 2 rh minutes 10 minutes 10 minutes Secondary. 2240 CFM 1475 CFM 1475 ACFM to 1625 ACFM Containment Leakage Location of Not Spccitied Uniform Across 40% - Bonom RR doors Bldg. Etev. 60% - Top - Siding Leakage SW Tcrnp 75'F 77'F (70'F pcr PPM 2.4Q Initial Bldg 100'F 75OF 75'F Tcnlpelanire Outside Temp Not Specified 12 F (lowest 20'F (95% windltcmp (NCR issue) monthly avg.) meteorology)

Wind Not Spccificd IOD MPH (avg. 95% Met. Data Conditions (NCR issue) monthly wind Jan)

SGI'ow 4457 ACFM 5600 ACFM 5385 ACFM to 5850 ACFM Hoor Limiter Not Specified Qualitatively 8.0% (Cele FJI42-91-1066)

Accuracy Considered Degraded 80% Qualitatively 87% (Based on degraded Voltage Considered voltage Cate FJI42-9041)

How Element Not Considered Not Considcrcd -5 to + 10% (Rcf Cele Cahb. Error NE42-9246 bldg. tcmperaturc)

Assumed SGT or SW Single Failure

'he parameter was qualitatively considered in the safety evaluation to have no impact on the opetability of SGT under all design basis accident conditions. However, this evaluation was not documented.

13

Justification for Continued Operation, Rev. 5 NCR 292-171 Table 2 Parametric Dose Results for NCR 288-357 (JCO Rev 2)

SGT Flow Time for Control Room Exclusion Area LPZ Dose (ACFM) -.25 W.G. Dose (REM) Dose (REM) (REM)

(minutes) Thyroid Whole Beta Whole Thyroid Whole Thyroid Bod Bod Bod JCO JJev 2 2 SOT Trains 10600 10 25.3 0.2 5.4 4.8 134.9 3.7 116.4 1 SGT Train 5850 10 25.3 0.2 5.4 4.7 131.9 3.7 115.2 I SGT Train 5385 10 25.3 0:2 5,4 4.7 131.3 3.7 115.0 1 SGT Train 5385 17 25.9 0.3 5.4 5.0 208.7 3.8 146.5 4457 5.8 0.1 2.7 1.3E-7 8.3E-7 6.3E-7 3.1E-S 6, 15.6 JC.'0 Rev 2 5600 10 19.3 104 l0QEP I I 10CFR50 Appx XI A 30 5 30 25 300 25 300 GDC 19 Limits JABOT ES Dose determined for entire accident Dose deterniined aAer first two hours of accident Preliminary Analysis 14

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Attachment 4 Page 16 of 23 STANDBY GAS TREATM~WT JCO CLOSEOUT ITEM PUNCHLIST J ORev FSAR Sections 6.5.1 and 9.4.2 and Technical Specification 3/4.6.5 do not clearly 'refiect Branch Technical Position CSB 6-3, Page 6.2.3.8, which states that "a positive pressure is defined as only pressure greater than -0.25" W.G., to account for wind loads and the uncertainty in the pressure measurements," and the original intent to assure a negative pressure in the Reactor Building.

The JCO addressed this concern by not crediting SGTsystemflltration until -0.25" w.g. was established in the reactor building. The current LOCA dose analysis (NE-02-88-27) does not credit SGT system ftltration for 20 minutes post-accident. This change to the design basis is reflected in the proposed amendment to the secondary containment and SGT system Technical Specijications.

2. TER 88-0362-0 requests a redesign of the SGT sensor hP instrument loop to replace the four sensors with one d,P sensor on the Reactor Building roof.

The four hP sensors incorporate an auctioneering circuit and measure pressure at the 572'levation.

There is no need to pursue the design change requested in the 1ZR. The current secondary containment drawdown analysis (NE-02-94-19) is based on the existing design and the results of the analysis are re//ected in the proposed amendment to the secondary containment and SGT system Technical

~

Specifications. ~

3. Actions required to support JCO, Rev. 0 bases.

~ Evaluate the results of a future capacity test of SGT (TP 8.3.108) to determine system capability.

SGT system flow capacity is demonstrated as part of the monthly SGT surveillance test (PPM 7.4.6.5.3.1). The results of these tests are evaluated for acceptability consistent with the surveillance test program.

~ Evaluate potential design changes to minimize impact, i.e., relocate dP sensors to the roof, evaluate need for setpoint change.

The hP sensors are located on the 572'levation of the reactor building. The evaluation of the need to relocate the d,P sensors concluded that no design change or setpoint changes were necessary. The current secondary containment drawdown analysis PVE-02-94-19) is based on the existing design and the results ofthe analysis are reflected in the proposed amendment to the secondary containment and SGT system Technical Spectftcations.

~ Consider revising licensing basis to specifically identify design basis parameters of temperature, humidity, and wind with respect to SGT for both surveillance testing and post-LOCA operation.

IA 0

<t r

0

Attachment 4 Page 17 of 23 A change to the FSAR has been prepared that addresses the impact of temperature, humidity, and wind on SGT system surveillance testing and post-LOCA operation. The changes to the design basis parameters are reflected in the proposed amendment to the secondary containment and SGT system Technical Spectftcations. The changes to the FSAR cannot be incorporated until the NRC approves the proposed amendment to the Technical Spectftcations.

J~ORev 1 Recent analysis (Calculation ME-02-89-09), based upon Standby Gas Treatment, Secondary Containment, Standby Service Water, and weather modeling shows that, during post-LOCA or adverse weather, differential pressure of the Secondary Containment may not always meet the, FSAR commitments to be maintained at -0.25" W.G. and be established within two minutes following a LOCA.

This analysis was used to defend the 10 minute drawdown time used in the JCO by incorporating meteorology and service water conditions to determine the time to reach -0.25" w.g. Calculation NE-02-94-19 supersedes ME-02-89-09 as the basis for the changes to the FSAR and the proposed amendment to the secondary containment and SGT system Technical Spectftcations.

2. Based upon single Train design basis SGT flow and maximum Technical Specification allowable Secondary Containment leakage, the upper-most surface areas of the Reactor Building cannot be maintained at -0,25" W.G. pressure with respect to atmospheric pressure during low temperature and high wind conditions, Calculation NE-02-94-19 determined that all surfaces of the reactor building can be maintained at-0.25" w.g. under 95% meteorological data considerations. This is based upon 5000 cfm SGT system gow (one fan) and 2240 cfm reactor building leakage at -0.25" w.g.

3. Reviews have uncovered a problem in the meteorology data presented in FSAR Amendment 36.

In changing from the Hanford Dispersion Model used for initial licensing basis to the PAVAN methodology consistent with Regulatory Guide 1.145, incorrect input resulted in erroneous X/Q values being submitted in the FSAR. If these X/Q values were used in revising the Chapter 15 events, the resultant offsite doses are nonconservative with respect to R.G. 1.145 guidance.

An FSAR change has been prepared which revises the Chapter 1$ tables consistent with the XIQ values established in Calculation 1VE-02-88-27based on Murphy-Campe methodology. The revised XIQ values are consistent with R.G. 1.145 guidance. The changes to the FSAR cannot be incorporated until the NRC approves the proposed amendment to the secondary containment and SGT system Technical Spectftcations.

4. The performance criteria stipulated in the SGT LCOs and surveillance requirements do not specify environmental conditions or specifically permit application of any compensation for wind, temperature differential, etc. Verbatum compliance with the existing Technical Specification requirements would not require testing under the most adverse weather conditions or require adjustment to compensate for them.

Attachment 4 Page 18 of 23 The current surveillance procedure (PPM 7.4.6.5.1.2) provides for engineering to evaluate the data to determine whether the SGT system capacity is sufhcient to meet the secondary containment drawdown requirements (10 minutes for the current ICO) using 95% meteorology data. Upon NRC approval of the proposed amendment to the secondary containment and SGT system Technical Spectftcations, the surveillance procedure will be revised to incorporate the changes to the surveillance requirements.

Since the changes to the surveillance requirements simplify SGT system capacity determination, it is presumed that the test da'ta will no longer require engineering review for acceptablility. Completion ofthe revised surveillance procedure willensure that the SGT system capability to drawdown secondary containment consistent with design basis assumptions even under adverse weather conditions.

5. A design review of the SGT system to determine the susceptibility of an SGT Train to single failure has not been performed.

Evaluations (EANF-90-0167) were performed by the Supply System Safety and Reliability Analysis Group to assess the reliability of the SGT system. The results of the evaluations conjirm that each train of the SGT system is single failure proof.

6. The current condition of the SGT system and Secondary Containment do not meet the FSAR description under all reasonable environmental conditions; however, the resultant doses are within the 10CFR100 and GDC 19 requirements in 10CFR50, Appendix A.

An FSAR change has been prepared which revises the appropriate SGT system and secondary

~

containment descriptions to reference the current design basis assumptions related to environmental

~

conditions. The changes to the FSAR cannot be incorporated until the NRC approves the proposed

~

amendment to the secondary containment and SGT system Technical Spectftcations.

7. Actions required to support JCO, Rev. 1 bases.

~ Resolve SGT operating mode and setpoint issues.

.This action resulted Pom PER 290-001 7 and was incorporated into the long term resolution as described in NCR 288-0357 and 292-0171. Final closure of this issue and the associated corrective actions occurred on December 31, 1992.

~ Review and propose modification recommendations for the plant operating procedures, This action was addressed by NCR 292-0171. Operating procedures 2.3.1 and 2.3.5, as well as annunciator response procedures 4.811.K2 and 4.827.Kl were changed to incorporate the NCR recommended actions.

~ Reportability review.

A reportability evaluation was completed on May 14, 1990 which determined that post-accident doses remained within the 10 CFR 100 guidelines. Therefore, it was concluded that the condition was not reportable since the plant remained within its design basis.

Attachment 4 Page 19 of 23

~ Communication with the NRC.

A meeting was held between the NRC stag and Supply System stag on January 16, 1990 to discuss the issues addressed in NCR 288-0357 and PER 290-0017.

J~C2 For the range of conditions under which the SGT flow indicator/recorders (SGT-FR-2A1, 2A2, 2B1, and 2B2) are designed to operated, the actual standard flow through the SGT Trains will be 5% lower to 10% higher than indicated on the indicator/recorders. In addition, the SGT fan controller (limiter) has a conservatively calculated accuracy of +/-10%. The effect of these two conditions is that the surveillance test results for SGT system flow and Reactor Building leakage could have been 20% higher than previously believed.

This condition was reported to the NRC via the ENS on March 1, 1992 and was followed up on March 31, 1992 by LER 92-008. This condition was reported as a condition alone that could have prevented the fuiftllment of the structures and systems that are needed to control the release of radioactive material. The actions to resolve the condition are described in NCR 292-0171. Final closure of this issue and the associated corrective actions occurred on October 20, 1992.

2.~ A higher total SGT system flow (up to 20% higher) could have resulted in a trip of the SGT fan motors on the thermal overloads

~

if the fans had been called upon to operate under degraded voltage conditions. ~

See No. 1 above.

3. The Reactor Building leakage could have been up to 20% higher than assumed in the Secondary Containment drawdown analysis which could have resulted in under predicting the time to reach a building pressure of -0.25" W.G..

See No. 1 above.

4. SGT system design reviews indicate that the system is vulnerable to the following single failures:

~ Electrical separation - hot shorts disabling the active functions of SGT valves and fans in both Divisions simultaneously (14 valves and 2 fans affected).

The current secondary containment drawdown analysis (Calculation NE-02-94-19) only credits one fan in one train during design basis accident conditions. This is rejfected in the proposed amendment to the secondary containment and SGT system Technical Specifications.

~ A failure of either SGT Train due to a single failure opening of the associated fire protection water deluge supply valve.

These valves are maintained normally closed. This conftguration satisftes the single failure criteria for the current design basis and is rejfected in the proposed amendment to the secondary containment and SGT system Technical Specifications.

Attachment 4 Page 20 of 23

~ A single failure of the SGT inlet air valve to open in either Train.

These valves are maintained normally open. This conftguration satisftes the single failure criteria for the current design basis and is re/ected in the proposed amendment to the secondary containment and SGT system Technical Speciftcations.

~ Lack of control logic to detect and transfer SGT operation to the redundant division upon failure of a heater coil (i.e., SGT humidity control).

Annunciator response procedures (PPMs 4.8II.K2 and 4.827.KI) were changed to include a high moisturelhumidity annunciator response action to energize the lag heater when it is determined that the lead heater has failed. In addition, the long term reactor building analysis (NE-02-94-71) indicates that the humidity will remain within the capability of the SGT system heaters.

5. Basic assumptions used in the Reactor Building drawdown model and presented to the NRC were changed without NRC review.

A 10 CFR 50.59 evaluation (Control No. 92-07) was performed for the changes to the basic assumptions used in the analysis. The analysis concluded that the changes did not involve an unreviewed safety question and NRC review and approval was not required.

~ To improve SGT performance capability, operating procedures will be revised to provide guidance to detect and switch heaters upon detection of high SGT humidity.

See No. 4 (fourth bullet) above.

7. The WNP-2 dose calculation model has been enhanced and expanded to support evaluations associated with the development of changes to the SGT design basis.

Calculation NE-02-88-27 provides the basis for the development of the dose model.

8. Actions required to support JCO, Rev. 2 bases.

~ Reduce the SGT flow limiter setting to 5380 ICFM (5850 to 5385 ACFM) to account for inaccuracies in SGT flow elements and loops and to achieve Secondary Containment drawdown within ten minutes.

Calculation E/I-02-91-1066 provided the basis for an instrument setpoint change (ISCR 1107) that was incorporated into PPMs 7.4.6.5.1.2, 7.4.6.5.3,2A & 2B, 7.4.6.5.3.5A & 5B, 7.4.6.5.3.6A & 6B, 7.4.6.5.3.1, 7.4.6.5.3.4, 2.3.1, 2.3.5', 4.811.K2, 4.827.KI, and 7.0.0.

~ Maintain SW temperature at or below 72'F to achieve Secondary Containment drawdown within ten minutes,

Attachment 4 Page 21 of 23 The appropriate procedures were changed to maintain SW at or below 72'F, but were subsequently changed back to the Technical Specification limit of 77'F. See Nos. 3 & 4 ofJCO, Rev. 5.

~ Add operator action to protect against heater coil failure (i.e., switch to lag fan if lead heater coil is inoperable).

See No. 4 fourth bullet) ofJCO, Rev. 2.

Rev The Technical Specification limit on the Standby Service Water spray pond is 77'F. However, in order to assure Secondary Containment drawdown time is ten minutes or less, the spray pond temperature must be maintained below 72'F when outside temperature is at or near the minimum temperature associated with the 95% meteorology envelope (i.e., 20'F).

The appropriate procedures were changed to maintain SW at or below 72'F, but were subsequently changed back to the Technical Spectftcation limit of 77'F. See Nos. 3 & 4 ofJCO, Rev. 5.

2. The Standby Service Water spray pond temperature restriction of 72'F must be maintained until SGT modifications are complete.

The appropriate procedures were changed to maintain SW at or below 72'F, but were subsequently changed back to the Technical Specijication limit of 77'F. See Nos. 3 & 4 ofJCO, Rev. 5.

3. Actions required to support JCO, Rev. 3 bases.

~ Reduce the SGT Qow limiter setting to 5380 ICFM (5850 to 5385 ACFM) to account for inaccuracies in SGT Qow elements and loops and to achieve Secondary Containment drawdown within ten minutes.

Same as No. 8 (erst bullet) ofJCO, Rev. 2.

~ Maintain SW pond temperature in accordance with Figure 4 to ensure that Secondary Containment drawdown to -0.25" W.G. can be achieved within ten minutes.

Same as No. 8 (second bullet) ofJCO, Rev. 2.

~ Add operator action to protect against heater coil failure (i.e., switch to lag fan if lead heater coil is inoperable).

Same as No. 8 (third bullet) ofJCO, Rev. 2.

0 Attachment 4 Page 22 of 23 KQ 4

1. Expand the Standby Service Water spray pond temperatures shown in Figure 5 to 32'F to eliminate potential procedure compliance problems when the spray pond temperature falls below 40'F (previous minimum temperature addressed) ~

The SW spray pond temperatures shown in Figure 4 of JCO, Rev. 3 were expanded to 32'F (Figure 5 of JCO, Rev. 4). However, the Figure was subsequently removed from procedures when SW temperature was changed back to the Technical Specification limit of 77'F. See Nos. 3 & 4 ofJCO, Rev. 5.

2. Actions required to support JCO, Rev. 4 bases.

~ Reduce the SGT flow limiter setting to 5380 ICFM (5850 to 5385 ACFM) to account for inaccuracies in SGT flow elements and loops and to achieve Secondary Containment drawdown within ten minutes.

Same as No. 8 (first bullet) of JCO, Rev. 2.

~ Maintain SW pond temperature in accordance with Figure 4 to ensure that Secondary Containment drawdown to -0.25" W.G. can be achieved within ten minutes.

Same as No. 8 (second bullet) ofJCO, Rev. 2.

~ Add operator action to protect against heater coil failure (i.e., switch to lag fan if lead heater coil is inoperable).

Same as No. 8 (third bullet) ofJCO, Rev. 2.

Rev

1. A review of open BCO/JCO items (reference IOM SS2-PE-93-260) noted that Revision 4 of the JCO considered 100% room cooler efficiency. The BCOs for NCRs 292-0476, 292-0941, and 292-0950 used 65% room cooler efficiency.

The current secondary containment drawdown calculation (NF=02-94-19) conservatively credits a 50%

room cooler egciency while maintaining the drawdown limits.

2. The Secondary Containment drawdown profile model was found to include the running heat load from the REA and ROA fans during a LOCA event, although no credit was assumed in the analysis for these fans to aid in drawdown since they trip on the F, A, and Z signals.

The addition of these heat loads was conservative since they resulted in an increase in the drawdown time. See Nos. 3 & 4 below

~ ~ ~

Attachment 4

. Page 23 When the heat loads of the REA and ROA fans were removed from the Secondary Containment drawdown model, the drawdown time is less that ten minutes with the Standby Service Water spray pond temperature at 77'F. Thus, spray pond temperature restrictions are no longer of 23 necessary and Figure 4 was removed from the JCO.

Current procedures re/lect the restoration of the SW temperature to 77'F.

4. Plant procedures will reflect the restoration of the Standby Service Water spray pond temperature limit to 77'F (original Technical Specification limit) after JCO approval by POC by removing Attachment 6.4 from PPM 2.4.5, Revision 23; Attachment 9.10 from PPM 7.0.0, Revision 27; Attachment 9.4 from PPM 7.0.1, Revision 18; and Attachment 9.7 from PPM 7.0.2, Revision 16.

See No. 3 above (indicated attachments contained Figure 4).

5. A Basic Design Change has been implemented to place the SGT system inlet air valves in a normally open condition. References to future actions related to the inlet air valves were removed from the JCO.

BDC 92-0301-OA has been implemented to lock open the SGT system inlet air valves.

~

6.~ Actions required to support JCO, Rev. 5 bases.

~ Reduce the SGT flow limiter setting to 5380 ICFM (5850 to 5385 ACFM) to account for inaccuracies in SGT flow elements and loops and to achieve Secondary Containment drawdown within ten minutes.

Same as No. 8 (ftrst bullet) ofJCO, Rev. 2.

~ Add operator action to protect against heater coil failure (i,e., switch to lag fan if lead heater coil is inoperable).

Same as No. 8 (third bullet) of JCO, Rev. 2.

REQUEST FOR AMENDMENTTO SECONDARY CONTAPSIENT AND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS ATTACHMENT5 GOTHIC Model Comparison Information

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 5 Page 1 of 7 GOTHIC Comparison The purpose of this section is to document the applicability of the GOTHIC code for use in the secondary containment drawdown analysis. GOTHIC was developed as a general purpose thermal-hydraulic analysis tool for design, licensing, safety, and operating analysis of nuclear power plant containment and other confinement buildings. GOTHIC was developed by Numerical Applications, Inc. (NAI) with support through EPRI and many sponsors via a tailored collaboration project. The result is a state-of-the-art best estimate compartment analysis code that was verified by independent review and validated against experimental results. GOTHIC was used for WNP-2 secondary containment analysis because of its flexibility, ease of use, and applicability. Justification for its use is based on work performed by both the Supply System and NAI in combination with the GOTHIC users group.

Qualification of GOTHIC was performed by comparison of the code output with experimental data for containment applications and solutions to analytic problems. The qualification basis was the use of GOTHIC as a best estimate containment analysis code. The qualification that GOTHIC provides representative results when model situations fall within the report'oncluded range of analyzed problems. The valid range of GOTHIC was considered in selecting it for modeling the secondary containment and High Energy Line Break (HELB) evaluations for WNP-2.

Westinghouse has independently developed and submitted the AP600 containment design analysis, using a version of GOTHIC, for staff review. The Westinghouse version of the GOTHIC code differs only in that the steam and non-condensable properties were extended to encompass the required design parameters of the AP600.

Verification and validation of the GOTHIC containment analysis package was performed under an approved QA program. The Supply System has independently validated the vendor QA program and verified compliance with 10 CFR 50, Appendix B, requirements. Therefore, the code has been approved for use at WNP-2 as a vendor verified code. The code is currently maintained by the vendor QA program. GOTHIC was installed at WNP-2 under the Supply System's QA program which included running a series of test problems to assure proper code installation and performance as designed by the vendor.

In addition to the qualification of GOTHIC based on comparison with analytic and experimental results, a comparison with other thermal-hydraulic codes has been performed. Several participants of the GOTHIC users group have made comparisons with other containment codes, including CONTEMPT and RELAP4, and presented their findings at Users Group meetings.

The findings verify a reasonable correlation in results. NAI has developed a white the numerical and modeling differences between GOTHIC and CONTEMPT-LT/028.

paper'escribing The Supply System has developed comparisons of GOTHIC with RELAP4 for various HELB analyses.

The Supply System has been using GOTHIC over the past six years for various analyses.

Confidence in the appropriate use of the code has been developed through attendance at regular

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 5 Page 2 of 7 users group meetings, seminars, and comparison with previous analyses. One of the codes used previously at the Supply System for design basis analysis was the RELAP4 computer code.

RELAP4 was used to determine secondary containment response to various HELBs. To assure that GOTHIC would be properly used as an analysis tool, the results were compared with previous analyses performed with RELAP4. The following is a discussion of one of these comparisons.

GOTHIC was used in support of the WNP-2 power uprate to revise HELB environmental profiles. Prior to performing the final power uprate calculations, a benchmark case was prepared between GOTHIC and RELAP4 to assure that the code was being properly applied.

A Reactor Water Clean Up (RWCU) line break was selected for the benchmark case. Figure 1 of this attachment presents a schematic of the model nodalization used for both the GOTHIC and RELAP4 runs. Both model inputs were essentially identical with only minor changes due to input requirement differences between the two codes. Some of the major differences between RELAP4 and GOTHIC are in the way heat transfer coefficients and condensation are evaluated.

Since RELAP4 was developed as a licensing basis code for use in evaluating reactor response to a LOCA, it employs fixed heat transfer coefficients to assure compliance with 10 CFR 50, Appendix K, "ECCS Evaluation Models." However, GOTHIC has a selection of heat transfer and condensation options available which permits versatility in the type of problem being analyzed.

For the benchmark comparison, the GOTHIC analysis selected heat transfer coefficients consistent with RELAP4 methodology. Figures 2 through 9 of this attachment provide the results of the comparison between GOTHIC and RELAP4 for a RWCU HELB. The comparisons are favorable for both temperature and pressure in the various compartments.

Based on this comparison, it was demonstrated that GOTHIC will produce results comparable to RELAP4 and supports the use of GOTHIC for applications involving compartment type analysis such as the secondary containment analysis presented in this submittal.

In summary, GOTHIC is a versatile best estimate containment analysis code to accurately evaluate containment response to any number of transient or steady state conditions, The code was installed by the Supply System consistent with current approved procedures. Appropriate use of the code has been controlled by the use of vendor training and validation against previously performed analyses. The code was used within the allowable ranges of correlations and thermodynamic data. The Supply System considers that the qualification and verification and validation performed adequately justifies the use of GOTHIC for the secondary containment drawdown analysis.

'Numerical Applications Inc., "GOTHIC Containment Analysis Package Qualification Report,"

NAI 8907-09 rev 2, (EPRI report RP3048-1).

'George, T. L., "A Comparison of the GOTHIC-4.1 and CONTEMPT-LT/028-A Computer Programs," Numerical Applications, Inc., December 1994.

548hbench-Benchmark w/RELAP4-HELB.R51 1 a Attachment 5 Mon Jul 31 10&4:43 1995 GOTHIC Version 4.1 (QA~ - December 1994 Page 3 of 7

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Attachment 5 548hbench-Benchmark w/RELAP4.HELB-R511a Mon Jul31 0991971995 Page 6 of 7 GOTHIC Version 4.1(QA)-b - December 1994 Vol 1 - Pressure ProQe PR1 DC7 0.1 10 100 1000 1e+04 Time tsec)

GOTHIC 4.1 QA 02/24/95 15:51:47 Figure 6. Volume 1 Pressure Profile Comparison (DC7 RELAP4 Data)

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REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS h

ATTACHMENT6 No Significant Hazards Consideration and Environmental Considerations Evaluations

REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 6 Page 1 of 4 NO SIGNIFICANT HAZARDS CONSIDERATION In accordance 10 CFR 50.91, a licensee must provide to the NRC its analysis of a proposed amendment to the operating license using the standards in 10 CFR 50.92 concerning the issue of no significant hazards consideration. The Supply System has evaluated the proposed amendment to the Technical Specifications and determined that the amendment does not involve a significant hazards consideration. This determination has been performed using the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of no significant hazards consideration standards:

erai n f WNP-2in cc rd ncewi h he r ed men men will n inv 1 e i nific n incr e in he ro ili r c n ences of an acciden revi usl ev luated Secondary containment and the Standby Gas Treatment (SGT) system are not initiators or precursors to any accident. The SGT system acts as part of secondary containment to minimize and control airborne radiological releases from the plant following a design basis accident.

Therefore, operation of WNP-2 in accordance with the proposed changes will not cause a significant increase in the probability of an accident previously evaluated.

The proposed amendment to the Technical Specifications impacts the capability to demonstrate that the secondary containment and SGT system designs will maintain radioactive releases within 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits. As a result, a new (current) design basis accident dose analysis was performed using the source term criteria outlined in Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," to evaluate the proposed changes. The new analysis provides a conservative representation of the timing and release of radioactivity during a design basis accident.

The proposed amendment also deletes the normal (nonsafety-related) secondary containment ventilation system surveillance requirement to verify every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within secondary containment is ~ 0.25 inch of vacuum water gauge. This surveillance requirement is not necessary as current Technical Specification Limiting Conditions for Operation (LCOs) as well as the WNP-2 Final Safety Analysis Report (FSAR) adequately address secondary containment integrity requirements and ensure secondary containment effluent is monitored.

Deletion of the surveillance requirement has no impact on the secondary containment drawdown analysis or the design basis dose analysis. Thus, the analyses assumptions and conclusions remain valid.

The secondary containment and SGT system designs must accommodate a post-accident single failure and remain operable. In addition, certain plant specific parameters, such as SGT capacity, secondary containment in-leakage, outside meteorological conditions, secondary containment heat loads, available cooling capacity, emergency diesel start time and loading sequence, and drawdown time for secondary containment must be considered in the design analyses and dose assessments. The current design in conjunction with an assumed secondary containment leakage of 2240 cfm and a drawdown time of 20 minutes provide assurance that the

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REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 6 Page 2 of 4 radiological doses for a design basis accident are maintained below the 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits.

The dose analysis supporting the proposed amendment to the Technical Specifications includes analytical changes to the SGT flow rate, secondary containment drawdown time, mixing, and bypass leakage, and establishes a 95% meteorological basis. These analytical changes, in combination, result in a calculated increase in the offsite thyroid dose values and a decrease in the whole body dose values. Although the calculated offsite thyroid dose values are higher than previously calculated, they remain within the 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits. In accordance with Standard Review Plan (NUREG-0800), Section 15.6.5, "Loss-of-Coolant Accidents Resulting From a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary," the radiological consequences of a design basis accident are considered acceptable ifthey are within the guidelines of 10 CFR 100. Since the offsite thyroid dose values remain within these acceptance criteria, and since there is no increase in the control room thyroid dose values or any of the whole body dose values, the changes are considered acceptable and operation of WNP-2 in accordance with the proposed amendment to the Technical Specifications will not cause a significant increase in the consequences of an accident previously evaluated.

erati n f WNP-2 in cc rdance wi h the ro sed mendmen will no cre e he ossibilit f

fa new ordifferen kind cciden fr m an cciden revioi 1 ev 1 i ed Secondary containment and the SGT system are not initiators or precursors to any accident. The SGT system acts as part of secondary containment to minimize and control airborne radiological releases from the plant following a design basis accident.

The dose analysis supporting the proposed amendment to the Technical Specifications includes analytical changes to the SGT flow rate, secondary containment drawdown time, mixing, and bypass leakage, and establish a 95% meteorology basis. These analytical changes do not alter any safety-related equipment or functions or create any new failure modes. The changes will improve the capability of secondary containment and the SGT system to mitigate the consequences of a design basis accident by ensuring that secondary containment pressure can be drawn down from 0 inches water gauge to at least 0.25 inch of vacuum water gauge during the most adverse environmental conditions. The proposed changes reflect consideration of SGT capacity, secondary containment in-leakage, outside meteorological conditions, secondary containment heat loads, available cooling capacity, emergency diesel start time and loading sequence, and drawdown time for the limiting secondary containment elevation. Required instrumentation have been evaluated to ensure proper operation under normal and accident environmental conditions, including but not limited to pressure, humidity, seismic, temperature, and radiation. The evaluation method is based on American National Standards Institute/Instrument Society of America (ANSI/ISA) Standard S67.04-1988, "Setpoints for Nuclear Safety-Related Instrumentation," and guidelines in ISA draft Recommended Practice RP67.04, "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation."

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REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 6 Page 3 of 4 The proposed amendment to the Technical. Specification does not change plant equipment or functions, but serves to clarify and credit existing design features. Fault tree and single failure analyses were performed to ensure that the SGT system design, including the equipment and components, credited in the licensing basis for the proposed amendment meet the single failure criteria for credible failure modes. The proposed amendment also deletes the normal (nonsafety-related) secondary containment ventilation system surveillance requirement to verify every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within secondary containment is ~ 0.25 inch of vacuum water gauge.

Deletion of this surveillance requirement does not invalidate existing analyses or change plant equipment or functions. Thus, no new failure modes are created.

Based on equipment failure and qualification analyses performed and the above conclusions, the proposed amendment to the Technical Specifications does not change any safety-related equipment or functions, or create any new failure modes. Therefore, operation of WNP-2 in accordance with the proposed amendment to the Technical Specifications will not create the possibility of a new or different kind of accident from any accident previously evaluated.

erati n f WNP-2in ccor an ewi h he r o ed amendmen will n involve i nifican r uci nin mrin f fe Consistent with the current Bases for the Technical Specifications and the WNP-2 FSAR, secondary containment and the SGT system act to minimize and control airborne radiological releases from the plant to within 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits following a design basis accident.

The proposed amendment to the Technical Specifications impacts the capability to demonstrate that the secondary containment and SGT system designs will maintain radioactive releases within 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits. As a result, a new (current) design basis accident dose analysis was performed using the source term criteria outlined in Regulatory Guide 1.3 to evaluate the proposed changes. The new analysis provides a conservative representation of the timing and release of radioactivity during a design basis accident.

The proposed amendment also deletes the normal (nonsafety-related) secondary containment ventilation system surveillance requirement to verify every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within secondary containment is ~ 0.25 inch of vacuum water gauge. This surveillance requirement is not necessary as current Technical Specification LCOs as well as the WNP-2 FSAR adequately address secondary containment integrity requirements and ensure secondary containment effluent is monitored. Deletion of the surveillance requirement has no impact on the secondary containment drawdown analysis or the design basis dose analysis. Thus, it follows that deletion of the surveillance requirement will not impact the offsite and control room dose safety margins established by these analyses.

The dose analysis includes analytical changes which increase SGT system flow rate and secondary containment drawdown time, credit mixing within secondary containment, increase bypass leakage, and establish a 95% meteorological basis. The combined effect of these

t REQUEST FOR AMENDMENTTO SECONDARY CONTAINMENTAND STANDBY GAS TREATMENT SYSTEM TECHNICAL SPECIFICATIONS Attachment 6 Page 4 of 4 analytical changes results in an increase in the calculated offsite thyroid dose values. The calculated control room thyroid dose values and all of the whole body dose values are shown to decrease. Although the new thyroid dose values are higher than previously calculated, they remain within the 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits. The calculated thyroid dose values at the plant exclusion area boundary (EAB) (1.2 miles) increased from 72 Rem to 114.2 Rem and the calculated thyroid dose at the low population zone (LPZ) (3 miles) increased from 251 Rem to 275.6 Rem.

The LPZ is defined as all land within a 3 mile radius of the plant site and 0 persons reside within this area. The nearest residence is 4.1 miles from the plant site. There are no schools or hospitals within 5 miles of the plant site and the nearest population center is at 12 miles.

Considering the low population density in the area immediately surrounding the plant site, the increase in thyroid dose will have a small impact on the health and safety of the public.

Since the offsite thyroid dose values remain within the 10 CFR 100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19 limits, and since there is a small impact on the health and safety of the public, the increase in the offsite thyroid dose values are considered acceptable and operation of WNP-2 in accordance with the proposed amendment to the Technical Specifications will not cause a significant reduction in the margin of safety.

ENVIRONMENTALCONSIDERATIONS Based on the above evaluation, this proposed amendment to the secondary containment and SGT system Technical Specifications does not result in a significant hazards consideration. In addition, the proposed amendment does not create a potential for a significant change in the types or a significant increase in the amount of any effluents that may be released offsite, nor does the amendment involve a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the amendment meets the eligibility criteria for a categorical exclusion as set forth in 10 CFR 51.22(c)(9). Therefore, in accordance with 10 CFR 51.22(b),

an environmental assessment of this amendment is not required.

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