ML20133F364

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Proposed Tech Specs Providing Alternate Method of Testing S/Rvs During Shutdown Conditions Rather than During Unit Startup as Is Currently Done
ML20133F364
Person / Time
Site: Hatch  Southern Nuclear icon.png
Issue date: 01/07/1997
From:
GEORGIA POWER CO.
To:
Shared Package
ML20133F295 List:
References
NUDOCS 9701140158
Download: ML20133F364 (35)
v • d • e


Text

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l Enclosure 3 ,

Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications Page Change Instructions Unit 1 P_agg Instruction 3.4-8 Replace 3.5-6 Replace 3.6-19 Replace Unit 2 P_ age Instruction 1

3.4-8 Replace  !

l 3.5-6 Replace 3.6-19 Replace l

s 9701140158 970107 PDR ADOCK 05000321 P PDR HL-5276 E3-1

1 S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the S/RVs are as follows: with the Inservice Number of Setpoint Testing Program S/RVs (osia) 4 1110 i 33.3 4 1120 i 33.6 3 1130 i 33.9 Following testing, lift settings shall be within 1%.

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HATCH UNIT 1 3.4-8 96-34-12/13/96

ECCS -- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 165 psig, 18 months the HPCI pump can develop a flow rate 2 4250 gpm against a system head corresponding to reactor system pressure.

SR 3.5.1.10 -------------------NOTE--------------------

Vessel injection / spray may be excluded.

Verify each ECCS injection / spray subsystem 18 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 -------------------NOTE--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

SR 3.5.1.12 Verify each ADS valve relief mode actuator 18 months strokes when manually actuated.

HATCH UNIT 1 3.5-6 96-34-12/13/96

LLS Valves 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 Verify each LLS valve relief mode 18 months ,

actuator strokes when manually actuated. I SR 3.6.1.6.2 ------------------NOTE-------------------

Valve actuation may be excluded.

Verify the LLS System actuates on an 18 months actual or simulated automatic initiation '

signal.

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HATCH UNIT 1 3.6-19 96-34-12/13/96

j S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

SR 3.4.3.1 Verify the safety function lift setpoints In accordance j of the S/RVs are as follows: with the  ;

Inservice Number of Setpoint Testing Program S/RVs (osia) 4 1120 33.6 4 1130 i 33.9 3 1140 i 34.2 l Following testing, lift settings shall be within

  • 1%. I i

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HATCH UNIT 2 3.4-8 96-34-12/13/96 1

I 4

ECCS -- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY I

t i 1 SR 3.5.1.9 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i after reactor steam pressure and flow are j adequate to perform the test.

Veri fy, with reactor pressure s 165 psig, 18 months the dPCI pump can develop a flow rate 2: 4250 gpm against a system head

corresponding to reactor pressure.

SR 3.5.1.10 -------------------NOTE--------------------

Vessel injection / spray may be excluded.

l Verify each ECCS injection / spray subsystem 18 months l actuates on an actual or simulated i automatic initiation signal.

SR 3.5.1.11 -------------------NOTE--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

SR 3.5.1.12 Verify each ADS valve relief mode actuator 18 months strokes when manually actuated.

i (continued) l HATCH UNIT 2 3.5-6 96-34-12/13/96 i 1

)

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. . l LLS Valves 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 Verify each LLS valve relief mode 18 months actuator strokes when manually actuated.

SR 3.6.1.6.2 ------------------NOTE-------------------

Valve actuation may be excluded.

i Verify the LLS System actuates on an 18 months actual or simulated automatic initiation signal.

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i HATCH UNIT 2 3.6-19 96-34-12/13/96

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the S/RVs are as follows: with the Inservice Number of Setpoint Testing Program S/RVs (osia) 4 1110 33.3 4 1120 33.6 3 1130

  • 33.9 Following testing, lift settings shall be within 1%.

. ,c.

SR 3. . -------------------NOTE--------- --------

ot required to be perfo ntil 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> **"~

a eactor steam ure and flow are adequate erf he test.

Ver each S/RV opens manually 18 o hs tuated.

L, HATCH UNIT 1 3.4-8 LO fycndmant Nn 197

ECCS - Operating

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3.5.1 SURVEILLANCE REQUIREMENTS (continued) s SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

---__---==----.__-----__------------_---_--_

Verify, with reactor pressure s 165 psig, 18 months the HPCI pump can develop a flow rate a: 4250 gpa against a system head corresponding to reactor system pressure.

SR 3.5.1.10 -------------------NOTE------- - --

Vessel injection / spray may be excluded.

Verify each ECCS injection / spray subsystem 18 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 -------------------N0TE ==== ==------------

Valve a '.uation may be excluded.

-_-----_-------_--- __=------------------_

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

i SR 3.5.1.12 ------------- --- ------------

N requir to ---NOTE perf M dtil 12 rs afte tor steam ssure an lo are ade to perfo

--___-------.---_. est.

Verify each ADS val e s when manually 18 months i actuated. .

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HATCH UNIT 1 3.5-6 ^ fa ninent 5 . 105 -

LLS:Velvd 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 -------- -------

No re ire - -----NOTE-o be perf rm 7 -----til un 12 rs after eacto stea dome pre s e and flo a adequat to erform t tes Verify each LLS valve s whe anually 18 months actuated. p e e p gig { ,

CtcfudCf6 bod'6)

"W SR 3.6.1.6.2 ------------------NOTE-------------------

Valve actuation may be excluded.

Verify the LLS System actuates on an 18 months actual or simulated automatic initiation signal.

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4 HATCH UNIT 1 3.6-19 " Am;ndmat h . 195 /

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS

, SURVEILLANCE FREQUENCY s

SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the S/RVs are as follows: with the Inservice

Number of Setpoint Testing Program S/RVs (osia)
4 1120
  • 33.6
4 1130 33.9 3 1140 34.2 Following testing, lift settings shall be within 1%.
3. .2 -------------------NOTEZL------------ ' ' "

{SR ----

Not required to be performed u 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ~

Q reactor steam pres and flow are adequal o perfo test.

Ver each S/RV opens anually 18 ths ctuated.

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1 HATCH UNIT 2 3.4-8  % nt-No A 38_ .

IJ ' " "; M e E E M tE4W C 2 " " ' ~ 7 ? '" M .U D E G .MFLi .i E . " -- ECCS -Operating -

3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure :s 165 psig, 18 months the HPCI pump can develop a flow rate 2 4250 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.10- -------------------NOTE-------------------- - - ~ -

v. Vessel injection / spray may. be; excluded.
g.q3 ,

Hyge -

Verify each ECCS injection / spray subsystem 18 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 -------------------NOTE--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

SR 3.5.1.12 X---------- -------NOTE -------- ---- ---

Nh requir o be pe med un 12 ho s aft ctor ea press df w e ad to perf the tes .

~________________________ _______ ________

Verify each ADS valvc erm nually 18 months actuated. TM rnd 04flecfoC @8 t (continued) [

i HATCH UNIT 2 3.5-6 b Ainendment iiu. 137 [

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3.6-l.6 .

SURVEILLANCE REQUIREMENTS w

SURVEILLANCE FREQUENCY 4

I SR 3.6.1.6.1 ------ ------- p0TE-- ' ---- --------

N req i d to perfor un  ;

12 o sa er ea or tea e pre ure an . low a adequ e to p form t st. s  !

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i Verify each LLS valve,4psEs Wen m ually 18 months actuated. gjig[mg CLtftLCd O f [d}3 Q .-

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! SR 3.6.1.6.2 ------------------NOTE-------------------

Valve-. actuation-may- be-excluded.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _' .":1~"F

__19 _ ._. ._. _.. _ _ _ _

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-; 0E?f,yQ j:p_,t+, y Verify the LLS System actuates on-an-~ 18 ' months 1LT$iEGSk actual or simulated automatic initiation signal.

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I HATCH UNIT 2 3.6-19 CAmeninnt Nov135Y

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Enclosure 4 Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications Revised Bases Pages l

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S/RVs B 3.4.3 i

BASES I i APPLICABLE design pressure (110% x 1250 psig - 1375 psig). Sensitivity i SAFETY ANALYSES analyses hr.ve demonstrated that 8 or 9 S/RVs operating in ,

(continued) the pressure relief mode will maintain the reactor vessel l below 1375 psig. This LC0 helps to ensure that the l 1

acceptance limit of 1375 psig is met during the Design Basis i 4

Event. I From an overpressure standpoint, the design basis events are j J bounded by ' ~ 9 MSIV closure with flux scram event described l 4

above. Rei .ence 2 discusses additional events that are i

expected to actuate the S/RVs. )

i i j S/RVs satisfy Criterion 3 of the NRC Policy Statement  !

(Ref. 3). l i

l LC0 The safety function of eleven S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. I and 2), although niargins to the ASME Vessel Overpressure Limit are substantial. The requirements of this LC0 are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the I lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1?50 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on thr,se setpoints, but also include the additional uncertainties of 3% of the nominal setpaint drift to provide an ad&d degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

HATCH UNIT 1 B 3.4-14 96-34-12/13/96

S/RVs B 3.4.3

~

BASES

ACTIONS B.1 and B.2 (continued)  !

and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an i orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the S/RVs will open at the pressures assumed in the safety analysis of Reference 1. l The demonstration of the S/RV safety lift settings must be  !

performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is 3% for OPERABILITY; however, the valves are reset to 1% during the Surveillance to allow for drift.

Performance of this SR in accordance with the Inservice Testing Program requires an 18 month Frequency. The 18 month Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

(continued)

HATCH UNIT 1 B 3.4-16 96-34-12/13/96

S/RVs B 3.4.3 l

BASES (continued)

REFERENCES 1. FSAR, Appendix M. i

2. FSAR, Section 14.3.
3. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

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HATCH UNIT I B 3.4-17 96-34-12/13/96 a -

ECCS - Operating  ;

8 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 (continued) .

REQUIREMENTS l outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. 1 Operating experience has shown that these components usually '

pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

This prevents an RPV pressure blowdown.

SR 3.5.1.12 The pneumatic actuator of each ADS valve is stroked to verify that the pilot disc rod lifts when the actuator strokes. Pilot rod lift is determined by measurement of rod i travel. The total amount of lift of the pilot rod from the valve closed position to the open position shall meet criteria established by the S/RV supplier. SRs 3.5.1.11 and 3.3.5.1.5 overlap this SR to provide testing of the S/RV relief mode function. Additional functional testing is performed by tests required by the ASME OH Code (Ref. 13).

The Frequency of 18 months is based on the S/RV tests i required by the ASME OM Code (Ref. 13). Operating l experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based j on the refueling cycle. Therefore, the Frequency was j concluded to be acceptable from a reliability standpoint.

(continued)

HATCH UNIT 1 B 3.5-15 96-34-12/13/96

. .. . _ . _ _ .__ _ . _ ._ ~ - _ _ _.____

l ECCS -- Operating 8 3.5.1 BASES (continued)

REFERENCES 1. FSAR, Section 6.4.3.

t 2. FSAR, Section 6.4.4.

l 3. FSAR, Section 6.4.1.

4. FSAR, Section 6.4.2.

l 5. FSAR, Section 14.4.3.

l 6. FSAR, Section 14.4.5.

7. 10 CFR 50, Appendix K.
8. FSAR, Section 6.5.

I 9. NEDC-31376P, "E.1. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,"

December 1986.

10. 10 CFR 50.46.

l

11. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

12. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
13. ASME, OM Code - 1995, " Code for Operation and Maintenance of Nuclear Power Plants," Appendix 1.

9ETk i

i j HATCH UNIT 1 B 3.5-16 96-34-12/13/96 i

i LLS Valves B 3.6.1.6 BASES ACTIONS 8.1 and B.2 (continued) achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.

SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS The pneumatic actuator er each LLS valve is stroked to verify that the pilot disc rod lifts when the actuator strokes. Pilot rod lift is determined by measurement of rod travel. The total amount of lift of the pilot rod from the valve closed position to the open position shall meet criteria established by the S/RV supplier. SRs 3.6.1.6.2 and 3.3.6.3.6 overlap this SR to provide testing of the S/RV relief mode function. Additional functional testing is performed by tests required by the ASME OM Code (Ref. 2).

Also, the Frequency of 18 months is based on the S/RV tests required by the OM Ccde. 1 SR 3.6.1.6.2 The LLS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide complete testing of the safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditiens that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

HATCH UNIT 1 8 3.6-36 96-34-12/13/96

. . . . . = - . - - . - . . _ ,

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LLS Valves B 3.6.1.6 l BASES SURVEILLANCE SR 3.6.1.6.2 (continued)

REQUIREMENTS This SR is modified by a Note that excludes valve actuation.

This prevents a reactor pressure vessel pressure blowdown.

REFERENCES 1. FSAR, Section 4.11.

2. ASME, OM Code - 1995, " Code for Operation and Maintenance of Nuclear Power Plants, Appendix I."
3. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

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HATCH UNIT 1 B 3.6-37 96-34-12/13/96

i- S/RVs j B 3.4.3 1

BASES APPLICABLE design pressure (110% x 1250 psig - 1375 psig). Sensitivity i SAFETY ANALYSES analyses have demonstrated that 8 or 9 S/RVs operating in (continued) the pressure relief mode will maintain the reactor vessel l below 1375 psig. This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis ,

Event. I From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses additional events that are i expected to actuate the S/RVs.

S/RVs satisfy Criterion 3 of the NRC Policy Statement

, (Ref. 3). l I

LC0 The safety function of eleven S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. I and 2), although margins to the ASME Vessel Overpressure Limit are substantial. The requirements of this LC0 are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME l Code specifications require the lowest safety valve setpoint  ;

to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of 3% of the nominal setpoint drift to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

HATCH UNIT 2 B 3.4-14 96-34-12/13/96

S/RVs B 3.4.3 BASES ,

ACTIONS B.1 and B.2 (continued) and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times  !

are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an I orderly manner and without challenging plant systems. 1 SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the S/RVs will open at the pressures assumed in the safety analysis of Reference 1.

The demonstration of the S/RV safety lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is t 3% for OPERABILITY; however, the valves are reset to x 1% during the Surveillance to allow for drift.

Performance of this SR in accordance with the Inservice Testing Program requires an 18 month Frequency. The 18 month Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

1 l

(continued)

HATCH UNIT 2 B 3.4-16 96-34-12/13/96

S/RVs B 3.4.3 BASES (continued)

REFERENCES 1. FSAR, Supplement 5A.

l l 2. FSAR, Section 15, i

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3. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

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i HATCH UNIT 2 8 3.4-17 96-34-12/13/96

ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 (continued)

REQUIREMENTS outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

This prevents an RPV pressure blowdown.

SR 3.5.1.12 The pneumatic actuator of each ADS valve is stroked to verify that the pilot disc rod lifts when the actuator strokes. Pilot rod lift is determined by measurement of rod travel. The total amount of lift of the pilot rod from the valve closed position to the open position shall meet criteria established by the S/RV supplier. SRs 3.5.1.11 and 3.3.5.1.5 overlap this SR to provide testing of the S/RV relief mode function. Additional functional testing is performed by tests required by the ASME OM Code (Ref. 16).

The Frequency of 18 months is based on the S/RV tests  !

required by the ASME OM Code (Ref.16). Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in Reference 14. A Note to the Surveillance states that the instrumentation portion of the response time may be assumed from established limits. The exclusion of the instrumentation from the response time surveillance is ,

supported by Reference 15, which concludes that instrumentation will continue to respond in the microsecond ,

to millisecond range prior to complete failure.

(continued)

HATCH UNIT 2 B 3.5-15 96-34-12/13/96

, _ . . . - = . . . a_ . , , . - _.,1. _ s .- .., ,x =. . . . . _ . > ~ . _

ECCS - Operating B 3.5.1 1

l BASES SURVEILLANCE SR 3.5.1.13 (continued)

REQUIREMENTS l The 18 month Frequency is based on the need to perform the  !

Surveillance under the conditions that apply during a plant .

outage and the potential for an unplanned transient if the l Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually 1 pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability 1 standpoint.

6 4

REFERENCES 1. FSAR, Section 6.3.2.2.3.

2. FSAR, Section 6.3.2.2.4.
3. FSAR, Section 6.3.2.2.1. )
4. FSAR, Section 6.3.2.2.2.
5. FSAR, Section 15.1.39. l
6. FSAR, Section 15.1.40.
7. FSAR, Section 15.1.33.
8. 10 CFR 50, Appendix K. l l
9. FSAR, Section 6.3.3.

1

10. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 l SAFER /GESTR-LOCA Loss-of-Coolant Analysis,"

December 1986.

11. 10 CFR 50.46.
12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

13. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
14. Technical Requirements Manual.

(continued)

HATCH UNIT 2 B 3.5-16 96-34-12/13/96 l

t ECCS - Operating B 3.5.1 BASES REFERENCES 15. NED0-32291, " System Analyses for Elimination of (continued) Selected Response Time Testing Requirements,"

January 1994.

16. ASME, OM Code - 1995, " Code for Operation and t

Maintenance of Nuclear Power Plants," Appendix I.

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HATCH UNIT 2 B 3.5-16a 96-34-12/13/96

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i LLS Valves

! B 3.6.1.6 i

BASES I

ACTIONS B.1 and B.2 (continued) i achieve this status, the plant must be brought to at least j MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating i experience, to reach the required plant conditions from full l power conditions in an orderly manner and without
j. challenOng plant systems.

! ' SURVEILLANCE SR 3.6.1.6.1

REQUIREMENTS
The pneumatic' actuator of each LLS valve is stroked to l verify that the polit disc rod lifts when the actuator strokes. Pilot rod lift is determined by measurement of rod travel. The total amount of lift of the pilot rod from the valve closed position to the open position shall meet criteria established by the S/RV supplier. SRs 3.6.1.6.2 and 3.3.6.3.6 overlap this SR to provide testing of the S/RV relief mode function. Additional functional testing is performed by tests required by the ASME OM Code (Ref. 2).

Also, the Frequency of 18 months is based on the S/RV tests required by the OM Code.

SR 3.6.1.6.2 The LLS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function.

operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide complete testing of the safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

HATCH UNIT 2 B 3.6-37 96-34-12/13/96

f LLS Valves l B 3.6.1.6 l

BASES SURVEILLANCE SR 3.6.1.6.2 (continued)  !

REQUIREMENTS This SR is modified by a Note that excludes valve actuation.

This prevents a reactor pressure vessel pressure blowdown.

l REFERENCES 1. FSAR, Section 5.5.17.

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2. ASME, OM Code - 1995, " Code for Operation and Maintenance of Nuclear Power Plants," Appendix 1.

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3. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

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HATCH UNIT 2 8 3.6-38 96-34-12/13/96

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l Attachment Edwin I. Ilatch Nuclear Plant Valve Relief Request RR-V-11 1

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VALVE RELIEF REQUEST RR-V-11 SYSTEM: Main Steam Safety Relief Valves (SRVs)

VALVES: 1B21-F013A, B, C, D, E, F, G, H, J, K, & L 2B21-F013A, B, C, D, E, F, G, H, K, L, & M CLASS: Class 1-Main Steam Pressure Relief Valves With Auxiliary Actuating Devices Test Requirement:

ASME OM Code-1995, Appendix I, paragraph I 3.4.l(d) requires that valves that have been maintained or refurbished in place, removed for maintenance and testing, or 'coth, and reinstalled shall be remotely actuated at reduced or normal system pressure to verify open and close capability of the valve before resumption of electric power generation.

BASIS FOR RELIEF:

Exercising the main disk of the SRV after reinstallation can only be performed during reactor startup when there is sufficient pressure to actuate the disk. Past history indicates that the main disks routinely do not re-seat properly after being exercised during reactor startup, resulting in steam leakage into the suppression pool. This leakage results in a decrease in plant performance and the potential for increased suppression pool temperatures which could force a plant shutdown to repair a leaking SRV. P ois t operating history indicates that the exercising performed during reactor startup is of no significant benefit in ensuring the proper operation of the individual SRV assemblies.

System Description

The Unit 1 and Unit 2 S/RVs are the Target Rock Two-Stage, Model 7567F design. The S/RVs are dual-function valves capable of being independently opened in either the safety or the relief mode of operation. A total of 11 S/RVs are installed on each unit.

In the safety mode of operation, each S/RV opens when system pressure exceeds the valve's preset setpoint pressure, which is controlled by precompression of the setpoint spring acting down on the pilot disc.

Venting the volume behind the pilot disc creates a differential pressure across the main piston, thereby providing a force to open  !

the main disc and relieve system overpressure. Hence, reactor vessel i steam is allowed to flow directly through the main disc to seat I opening and to the suppression pool via the discharge piping. All 11  !

S/RVs operate in the safety mode, which provides the safety function of overpressurization protection. The requirements for this mode are listed in Technical Specification 3.4.3. i l

In the relief mode of operation, each S/RV is opened by an electro-pneumatic actuator, which consists of a three-way solenoid valve, an l attachment manifold, and a pneumatic operator. When the solenoid  !

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RR-V-11 (cont.)

l l valve is energized, pneumatic pressure is routed into the operator to lift the pilot rod against the force of the compressed setpoint spring. This allows system pressure to lift the pilot disc, venting the volume behind the disc, and opening the valve as in the safety mode discussed above. This mode of operation is used for ADS, LLS, j and remote manual operation. Technical Specifications 3.5.1 and  :

3.6.1.6 provide requirements for the ADS and the LLS System. Manual operation is not safety-related and is not addressed by Technical j Specifications. In each unit, seven S/RVs are part of ADS, while 1 the remaining four constitute LLS. l Current Testing at Plant Hatch Testing of Plant Hatch S/RVs is performed to satisfy Technical Specifications Surveillance Requirements and the ASME OM Code (1995),

" Code for Operation and Maintenance of Nuclear Power Plants."

Certain tests are performed with the S/RVs installed (in-situ), while others are performed as " bench tests" after the valve is removed and transported to a maintenance and testing facility. Current requirements are as follows:

1. Plant Hatch Units 1 and 2 Technical Specifications SR 3.4.3.2 I requires that each S/RV be opened by manual actuation to l demonstrate that the S/RV safety mode is operable. This is accomplished by showing that mechanically the valve is functioning properly and no blockage exists in the discharge piping. SR 3.4.3.2 is performed on an 18-month Frequency during reactor startup from refueling at a reactor pressure of at least 920 psig.
2. SRs 3.5.1.12 and 3.6.1.6.1 provide similar S/RV manual actuation testing for the ADS and LLS Functions, respectively, to demonstrate operability of the S/RV relief mode.
3. Remote manual actuation is also required by the ASME OM Code, Appendix I, paragraph 3. 4.1 (d) , to verify open and close capability of the valve before resumption of electric power generation. This applies to valves that were either maintained or refurbished in place, or removed for maintenance and testing and reinstalled. The remote manual actuation is performed at reduced or normal system pressure.

Plant Hatch currently meets these testing requirements by opening l and closing each S/RV using control room switches. Verification i

of valve opening and closing is made by monitoring the valve discharge piping temperature and/or pressure.

4. Plant Hatch Units 1 and 2 Technical Specifications SRs 3.5.3.11 and 3.6.1.6.2 require that the ADS and LLS S/RVs be opened on an
actual or simulated automatic initiation signal to demonstrate that the solenoids operate when initiated by a signal. Actual valve actuation is excluded from these tests which are performed on an 18-month Frequency.

Plant Hatch currently meets these testing requirements by performing the test in conjunction with Logic System Functional 2

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RR-V-11 (cont.) ,

1 Tests for the initiating instrument logic, which are also required by Technical Specifications.

5. ASME OM Code (1995) I 3.3.1 (d) and (e) require that S/RV auxiliary components be tested in place as follows: solenoid valve and pneumatic actuator integrity is verified by performance of leak rate tests, and solenoid valve electrical function is verified. These tests are performed following maintenance on the valves and together demonstrate operability of the valve pneumatic actuation system.

Current Testing at Outside Facilities During each refueling outage, which occurs on an 18-month Frequency, all 11 S/RV pilot assemblies and approximately 25% of the main stages are removed and shipped to Wyle Laboratories for "as-received" i testing, which includes visual inspection, leakage testing, pilot '

disc-to-seat sticking testing, and set pressure testing, all of which are performed prior to any maintenance on the valve. The leakage test and set pressure test are performed at a steam pressure of approximately 1010 psig. Both tests meet the requirements of ASME OM Code (1995) I 3.3.1(a), (b), and (c).

y Following the "as-received" testing, the S/RVs are given a dimensional inspection followed by refurbishment if required. This work is performed by the valve supplier, Target Rock Corporation.

Following valve refurbishment, post maintenance testing is performed at a steam pressure of approximately 1010 psig. This includes initial valve leakage testing; safety mode valve actuation to satisfy requirements for set pressure, reseat pressure, and main seat stroke time; and final leakage testing. Seat leakage tests are performed at approximately 1045 psig. Upon successful test completion, the valve i receives written certification from the lab and is returned to Plant Hatch for reinstallation. To receive certification, the valve must have zero seat leakage and meet the acceptance criteria for set and reseat pressure. These tests help meet the requirements of ASME OM Code (1995) I 3.3.1 and Technical Specifications SR 3.4.3.1 (for lift setpoint pressure verification).

General Change Justifiestion Leaking S/RVs result in the following challenges to Plant Hatch components and operation:

1. Extreme leakage during operation may cause the valve to inadvertently actuate, likely resulting in an unplanned plant shutdown, with its attendant challenges to plant safety systems and components. This has occurred recently at a domestic BWR plant.
2. Leaking S/RVs create operational problems with the suppression pool. S/RV leakage increases both pool temperature and level, requiring more frequent use of the suppression pool cooling mode of the Besidual Heat Removal (RHR) system.

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RR-V-ll (cont.)

3. Plant efficiency is impacted because the transfer of heat to the suppression pool is a source of thermal heat loss from the power generation steam cycle, thereby reducing electrical generating capacity. S/RV leakage results in radiological challenges since steam is an additional source of radioactive nuclides that become a potential source for personnel contamination.

As described previously, each S/RV pilot assembly and approximately 25% of the main stage valves are bench tested at Wyle Laboratories during each refueling outage. The valves are refurbished as necessary to meet the acceptance criteria of zero leakage, and are certified in writing as being leak free. The valves are reinstalled in the plant and remotely opened at a system pressure of at least 920 psig. Following this surveillance test, Plant Hatch typically experiences several leaking valves from what was originally a leak-free population. For example, in order to prevent the possibility of an inadvertent actuation, Plant Hatch was shutdown in March of 1996 as a result of a leaking S/RV. During this forced outage, two S/RVs were replaced with leak-tight valves, which were manually actuated at or above 920 psig. One of the S/RVs immediately began leaking following startup.

Also, the requirement for manual actuation during plant startup creates the need to progress in startup to a point where the S/PVs are required to be operable, but are not proven operable. The design of the Plant Hatch S/RVs dictates that a manual actuation of the valves be performed only when sufficient steam pressure is present to

" cushion" the valve when it reseats. Otherwise, significant valve damage may result. Therefore, the valve actuation may only be performed after the plant has progressed in startup to the point where steam pressure is sufficient. However, in order to achieve this, the unit has to move into a mode of operation requiring operable S/RVs before the test demonstrating operability can be performed. This is an exception to normal Technical Specifications operability requirements for mode enanges, but is recognized as necessary by a Note of exception given in the Technical Specifica.tions.

Several aspects of S/RV design and operation contribute to valve leakage. As mentioned earlier, these include simmer margin, reseat margin, testing pressure, pilot valve disc and rod configuration, and system and valve cleanliness. The manual actuation of the S/RVs at pressure allows these contributors to impact the ability of the valve to re-close completely. As stated previously, efforts not requiring prior approval are being made in parallel with this submittal to minimize the effects of these contributors. Also, a Technical Specifications revision request is presently being reviewed by the NRC to increase t.he simmer margin. However, elimination of valve testing at pressure conditions is expected to have the most positive impact in reducing the S/RV leakage.

l Additionally, reducing challenges to the S/RVs is a recommendation of NUREG 0737, "TMI Action Plan Requirements" item II.K.3.(16). The

! recommendation is based on a stuck open S/RV being a possible cause i

of a Loss of Coolant Accident. This submittal is consistent with that recommendation.

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RR-V-11 (cont.)

Specific Change Justification As an alternate to the testing required by ASME OM Code-1995, Appendix I, paragraph I 3. 4.1 (d) , GPC proposes to actuate the S/RVs in the relief mode before steam is generated. The solenoid valve will be energized, the actuator will stroke, and the pilot rod lift will be measured. This in-situ test will verify that, given a signal to the solenoid, the pilot disc rod will lift. If steam were present, the pilot disc would open and initiate opening of the main stage.

Alternate testing is justified since the remaining segments of the S/RV mode of operation are proven by other tests. The ability of the pilot disc to open is shown in the safety mode actuation bench test.

The integrity of the solenoid and pneumatic system for the S/RV is verified by performance of leak rate tests post maintenance, and the electrical function of the solenoid is verified. Automatic valve actuation is proven operable by logic system functional tests which includt verification that the solenoid actuates from the automatic signal.

Each refueling outage all 11 pilot assemblies and three or four main disc assemblies are sent to Wyle Labs and tested with system pressure. As a .esult, even though actual valve movement is not performed after the S/RV is re-installed in the plant, all pilot assemblies are tested with system pressure once per cycle and all the main discs are tested with system pressure approximately once every three cycles. This fact, together with the current and proposed S/RV testing, completely demonstrates operability of the valves.

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