ML20205P139

Requests That Certain Category B & C Valves Be Exempted from Inservice Testing Requirements of ASME Section Xi.Fee Paid

ML20205P139
Person / Time
Site:	Three Mile Island
Issue date:	05/15/1986
From:	Standerfer F GENERAL PUBLIC UTILITIES CORP.
To:	Travers W Office of Nuclear Reactor Regulation
References
0425A, 425A, 4410-86-L-0075, 4410-86-L-75, NUDOCS 8605210179
Download: ML20205P139 (7)

Text

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GPU Nuclear Corporation Nuclear  :::,ome:reo s Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Nurnber:

(717) 948-8461 4410-86-L-0075 Document ID 0425A May 15,1986 TMI-2 Cleanup Project Directorate Attn: Dr. W. D. Travers Director US Nuclear Regulatory Commission c/o Three Mile Island Nuclear Station Middletown, PA 17057

Dear Dr. Travers:

Three Mile Island Nuclear Station, Unit 2 (TMI-2)

Operating License No. DPR-73 Docket No. 50-320 Relief From ISI Test Requirements for Category B and C Valves GPU Nuclear letter 4410-84-L-0058, dated April 9,1984, requested relief from the Inservice Inspection (ISI) Test Requirements of ASE Section XI for various Category B and C Valves. Table II of the referenced letter identified those valves which GPU Nuclear had proposed to include in an ISI testing program subject to NRC approval of our request. Since the submittal date of that request, GPU Nuclear has submitted three (3) Technical Specification Change Requests (TSCR), i.e., Number 46, 49 and 51, as part of our Technical Specification Simplification Program, which are relevant. The changes implemented by TSCR No. 46 and those proposed by TSCR Nos. 49 and 51, which are currently being reviewed by the NRC, significantly change the TMI-2 Technical Specifications. Based on these TSCRs, GPU Nuclear has re-evaluated the need to conduct ISI testing on any of those valves identified in the o reference. Accordingly, based on the attached evaluation, GPU Nuclear requests that these valves also be exempted from the ISI testing requirements j of ASE Section XI.

8605210179 DR 860515 4h q p ADOCK 05000320 PDR  ;

[do GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

c Dr. Travers May 15,1986 4410-86-L-0075 Subject to your approval of the above request, it follows that THI-2 should be exempt from all ISI testing requirements based on the current plant status.

Therefore, exemption from the Inservice Inspection Program Requirements of 10 CFR 50.55a and the provisions of IWV-3410 and IWV-3510, for Category B and C valves is requested.

Per the requirements of 10 CFR 170, an application fee of $150.00 is enclosed.

Sincerely,

'C pf p

. R. Standerfer ice President / Director, TMI-2 FRS/RDW/eml Attachment

Enclosure:

GPU Nuclear Corp. Check No. 00023374

c ATTACHMENT

. 4410-86-L-0075 I Nuclear Services Closed Cooling Water (NSCCW) System The NSCCW system was deleted from the TMI-2 Technical Specifications by NRC Amendment Of Order dated August 8, 1985. Thus, testing of the valves listed below is not required since the NSCCW system is no longer essential for plant safety.

Valve Function NS-Vll A, B and C NS-P-1 A, B, C Discharge NS-V217 A and B Inlet to Instrument Air Comp.

NS-V218 A and B Outlet to Instrument Air Comp.

II Decay Heat Closed Cooling Water (DHCCW) System The DHCCW system was deleted from the TMI-2 Technical Specifications by NRC Amendment Of Order dated August 8, 1985. Thus, testing of the below valves is not required since the DHCCW system is not essential for plant safety.

Valve Function DC-V6 A and B DC-P-1 A/B Outlet NR-V40 A and B Inlet to Decay Heat Service Coolers NR-V42 A and B Outlet to Decay Heat Service Coolers III Reactor Building Normal Cooling Water (RBNCW) System The RBNCW system has never geen a post-accident Technical Specifications required system; therefore,;is not essential for plant safety.

Additionally, these valves are containment isolation valves per TMI-2 Surveillance Procedure 4210-SUR-3244.01, " Containment Integrity Verification - Recovery Mode." Thus, the following valves should be re-classified as Category A valves which are exempt from ISI Testing per NRC letter dated April 27, 1981.

Valve Function RR-V25 A, B, C & E Reactor Building Cooling Coil Outlet to Evaporator Cooler IV Standby Pressure Control (SPC) System The SPC system was deleted from the TMI-2 Technical Specifications by NRC Amendment Of Order dated August 8, 1985. Currently, the SPC system can be utilized to recover from an RCS leak. However, the SPC prc/ ides a secondary means of recovery; the primary means for recovering from an RCS leak would be to provide make-up by gravity feed from the BWST or by activation of the Reactor Building Recirculation System or use of both systems. Therefore, since the SPC system is not required for RCS make-uo, it is not essential for plant safety. Thus, testing of the below valves is not required.

m ATTACHMENT

, 4410-86-L-0075 Valve Function SPC-V4 A and 8 SPC-P-1 A/B Discharge SPC-V6 SPC to RCS Check Valve SPC-V32 SPC to RCS Check Valve SPC-V40 SPC to RCS Check Valve SPC-V71 SPC-T-1 Discharge Control Valve V Nuclear Services River Water (NSRW) System The NSRW (NR) valves and associated air handling (AH) valves listed in Table II of GPU Nuclear letter 4410-84-L-0058 have been categorized in terms of the systems supported as follows:

o Emergency Diesel Generator Operation o Service Building River Water Operation o Control Building Ventilation System o Control Room HVAC A. Emergency Diesel Generator Operation TMI-2 Techncial Specification Change Request (TSCR) No. 51, which was submitted to the NRC via GPU Nuclear letter 4410-85-L-0135 dated July 31,-1985, proposed deletion of the emergency diesel generators from the TMI-2 Technical Specifications. This proposcl was based on a safety evaluation which demonstrates a high probability that recovery from a loss of off-site power can be accomplished within eight (8) hours during which time power would supplied by the station batteries. Therefore, the safety evaluation states that the diesel generators are not required to maintain safe plant conditions.

The NSRW system supplies cooling water to the emergency diesel generators which, in turn, supply power to the NSRW pumps in the event of a loss of off-site power. Therefore, TSCR No. 51 proposed deletion of the NSRW system based on the justification that, subsequent to the deletion of the emergency diesel generators, the NSRW system will not be a safety related system; i.e., it will no longer service any Technical Specification required systems with the exception of the Control Room HVAC which is discussed separately in Section VII. Thus, GPU Nuclear believes that the following valves do require testing pending NRC approval of TSCR No. 51: M Valve Function NR-V1 A, B, C, D NR-P-1 A, B, C, D Discharge Valves NR-V2 A, B, C, D NR-P-1 A, B, C, D Discharge Valves NR-V33 A River Water to Emergency Diesel Generator Cooling NR-V34 8 River Water to Emergency Diesel Generator Cooling #g NR-V39 A and B Diesel Generator Cool 0utlet NR-Vll6 A and B River Water Pump House Fan Coil Inlet NR-Vil7 A and B River Water Pump House Fan Coil Outlet

n ATTACHMENT

, 4410-86-L-0075 NR-V186 A, B, C, D Lube Water to NR-P-1 A, B, C, D Check Valves NR-V234 A, B, C, D NR Pump Lube Water Supply AH-EP-5356 A Controls NR-Vll6 A AH-EP-5356 B Controls Damper D5356 AH-EP-5358 A Controls NR-Vll6 B AH-EP-5358 B Controls Damper D5358 B. Service Building River Water (SBRW) System The only NSRW valve associated with the SBRW system is NR-V241 which is the discharge valve for the SBRW Booster Pumps NR-P-3 A/B. The SBRW system provides cooling water for the Service Building HVAC system which is not a Technical Specification required system. Thus, testing of NR-V241 currently is not required; the SBRW system is not essential for plant safety.

C. Control Building Ventilation System This system provides ventilation for the Cable, Battery Switchgear, and Mechanical Equipment Rooms in the Control Building. The purpose of this system is to provide cooling water to the associated equipment in order to avoid degradation in severe summer conditions. The safety evaluation for TSCR No. 51 states that no significant equipment degradation will occur during the eight (8) hours conservatively assumed necessary to restore off-site power.

Thus, justification for not requiring ISI testing of the following valves is consistent with the rationale stated in Section V, Subparagraph A.

Valve Function NR-V82 A and B Control Building River Water Booster Pump Discharge NR-V85 A and B Control Building Mechanical Room Fan Coil Inlet NR-V88 A and B Control Building Mechanical Room Fan Coil Outlet NR-V144 A and B Inlet to Liquid Chiller Condenser -

Control Building NR-V145 A and B Outlet from Liquid Chiller Condenser -

Control Building AH-V14 A and B Control Building Liquid CooleI Pump Discharge AH-V28 A and B Cable Room Fan Coil Unit Water Inlet AH-V29 A and B Cable Room Fan Coil Unit Water Outlet AH-EP-5182 Controls NR-V144B AH-EP-5205 Controls NR-V144A AH-EP-5222 A Controls Damper D4088B AH-EP-5222 B Controls NR-V85 B AH-EP-5227 A Controls NR-V85 A AH-EP-5227 B Controls Damper D4088A AH-EP-5235 A and B Controls AH-V28 A and Damper D4074A AH-EP-5237 A and B Controls AH-V28 B and Damper D4074B AH-EP-5245 Controls Damper D4076 AH-EP-5246 A and B Controls Damper D4073, ID 4075A, ID 40758, and ED 4075

r ATTACHENT

, 4410-86-L-0075 VI Screen Wash (SW) System The SW system is designed to provide flushing water for the mechanical trash racks and traveling water screens which provide water filtration for the NSRW pumps. However, since flushing of the racks can be performed manually, the operability of the screen wash pumps is not a requisite for operation of the NSRW pumps. Additionally, as noted in Section V above, GPU Nuclear also proposes deletion of the NSRW system from the TMI-2 Technical Specifications. Testing of the following valves is currently not required since the SW system is not essential for plant safety:

Valve Function SW-V1 A and B SW-P-1 A and B Discharge SW-V20 A and B Lube Water to Screen Wash Pumps

• SW-V28 A and B Lube Water to Screen Wash Pumps SW-V35 Domestic Water Supply to Screen Wash Pumps

• These valves are secured in the "open" position; therefore, ISI testing is not required to verify their operability.

VII Control Room HVAC System TMI-2 TSCR No. 49, submitted to the NRC via GPU Nuclear letter 4410-85-L-0110, dated June 18, 1985, proposed deletion of certain functions of the Control Room HVAC System which require diesel generators in the event of a loss of off-site power. However, based on NRC concerns, GPU Nuclear letter 4410-86-L-0033 dated February 26, 1986, proposed retaining operability requirements for this system in the Technical Specifications, and requested only that the requirements for back-up on-site AC power supply be deleted. This request was based on the results of analyses which indicate that probability of a simulatenous occurrence of a Unit 1 LOCA and loss of off-site power is sufficiently low to be considered an incredible event. Therefore, emergerscy diesel generator power backup for the Control Room HVAC system is not required.

Additionally, TMI-2's unique condition is such that no actions are required to be taken from the Unit 2 Control Room to maintain the unit in a safe shutdown (i.e., continuous manning of the Unit 2 Ccntrol Room is not required to maintain a safe shutdown condition).

Additionally, as previously noted, the Control Room HVAC is and will continue to be maintained operable in accordance with the TMI-2 Technical Specifications. While the Technical Specification surveillance does not satisfy the ISI testing requirement, it is noteworthy that in granting GPU Nuclear an exemption from the ISI testing of Category A valves, which a"e the containment isolation valves in the case of TMI-2, the NRC based the exemption on the fact that these valves are maintained in accordance with the Technical Specifications. Thus, the basis for not testing Category A valves should also be applied to the following Control Room HVAC valves:

Valve Function AH-V32 A and B Control Room Fan Coil Unit Water Inlet AH-V33 A and B Control Room Fan Coil Unit Water Outlet

ATTACHMENT

. 4410-86-L-0075 AH-V124 A and B Control Room Recirculation Instrument Air AH-V125 A and B Control Room Recirculation Instrument Air AH-EP-5210 Controls Damper D40920 AH-EP-5216 A/B Controls Damper D4096A and AH-V32A AH-EP-5217 A/B. Controls Damper D40968 and AH-V328 AH-EP-5265 Controls Damper D4091A AH-EP-5266 Controls Damper 040918 VIII Decay Heat Removal System (DHRS)

NRC Amendment of Order dated August 8, 1986, deleted the Technical Specifications requirements for the DHRS based on a safety analysis which concluded that forced borated water recirculation systems are no longer required in TMI-2's unique condition. Accordingly, the TMI-2 Technical Specifications have been modified to replace the DHRS system with a Reactor Building Sump Recirculation System and two (2) operable flowpaths downstream of the Borated Water Storage Tank (BWST) common drop line, i.e., gravity feed from the BWST.

Thus, that portion of the DHRS which is not required for RCS make-up as a portion of the gravity feed flowpath from the BWST is not essential for plant safety. Therefore, testing of the following valves is not required:

Valve Function DH-VlD0 A and B Decay Heat Removal Pump Crosstie DH-V103 A and B Decay Heat Removal Pumps Discharge Check DH-Vl93 A and B Decay Heat Removal Ccolers Crosstie The following valves are associated with gravity feed from the BWST:

Valve Function DH-V5 A/B BWST to Decay Heat Removal Pumps DH-V102 A/B Discharge from Decay Heat Removal Pumps to RV DH-V113 A/B BWST to Decay Heat Removal Pumps Check Valves DH-V119 BWST Vacuum Breaker Valve DH-V128 A/B Discharge from Decay Heat Removal Cooler to RV DH-V227 BWST Vacuum Breaker Valve GPU Nuclear does not consider the above valves to be essential for plant safety. In the event that gravity feed from the BWST cannot be established via the DHRS, gravity feed can be accomplished through either the SPC, MDHRS, or other alternative means, e.g., temporary hose connection via the Fuel Canal Cleanup (FCC) manifold. Additionally, the analyses presented in the GPU Nuclear Seismic Design Criteria (

Reference:

GPU Nuclear letter 4410-85-L-0077,~ dated April 16, 1986), which has been i

reviewed by the NRC, demonstrated that a vessel draindown would not result in'either a criticality or offsite exposures in excess of 10 CFR Part 100 guidelines. Therefore, testing of the above valves is not required.

t