ML20214S079

Summary of 861114 Meeting W/Util in Bethesda,Md to Discuss Util 861014 Submittal on Generic Ltr 83-37 for Facility. Util Requested to Amend 861014 Submittal to Address Encl marked-up Draft Technical Evaluation Rept within 60 Days

ML20214S079
Person / Time
Site:	Millstone
Issue date:	11/24/1986
From:	Jaffe D Office of Nuclear Reactor Regulation
To:	Thadani A Office of Nuclear Reactor Regulation
References
GL-83-37, NUDOCS 8612080180
Download: ML20214S079 (34)

Text

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. November 24, 1986 Docket No. 50-336 MEMORANDUM FOR: Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B

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FROM: D. H. Jaffe, Project Manager

} PWR Project Directorate #8 Division of PWR Licensing-B

SUMMARY

OF THE NOVEMBER 14, 1986 MEETING WITH NORTHEAST NUCLEAR ENERGY COMPANY (NNECo) TO DISCUSS GL 83-37 -

(TMI TECH. SPECS.) j On November 14, 1986, representatives from NNECo and the NRC staff met in Room P-422 of the Phillips Building, in Bethesda, Maryland. The purpose of the meeting was to discuss NNECo's October 14, 1986 submittal on GL 83-37 for Millstone Unit No. 2.

The NRC staff informed NNECo that GL 83-37 contained NRC staff-approved Tech. (

Specs. for hardware required for installation by NUREG-0737. In some -

instances, the requirements of GL 83-37 could be relaxed where staff-approved TMI modifications for Millstone Unit 2 proved inconsistent with the GL 83-37 Tech. Specs.

The discussions with the licensee centered upon the draft technical evaluation report (TER) shown in Enclosure 2 which was prepared for the NRC staff by EG&G, Idaho. The draft TER compared the GL 83-37 Tech. Specs. with those submitted by NNECo on October 14, 1986. The most significant issues raised by the comparison were:

High. Point Vents - The NRC staff will allow credit for power operated relief valves (PORV's) provided that NNECo presents information to demonstrate that the PORV's substantially rieet NRC requirements for high point vents. Plant shutdown will still be required for an inoperable head vent and/or pressurizer PORV/ vent.

Post Accident Sampling - NNECo should propose a change to the Administrative Tech Specs, to require training, procedures, and maintenance for post accident sampling. The proposed limiting condition for operation (LCO) and Surveillance Requirements are not adequate in this regard.

Auxiliary Feedwater (AFW) - NNECo should provide a flow value for AFW testing. As an alternative, NNECo may demonstrate that the existing flow testing in " recirculation" corresponds to a unique flow value.

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Noble Gas Effluent Monitor - NNECo should relocate the proposed Noble Gas Effluent Ponitor Tech Spec to the Millstone Unit 2 Tech.

Spec. table containing other, similar, monitors.

Containment High-Range Radiation Monitor - The proposed Tech. Specs.

do not require use of a preplanned alternative monitor. In addition, NNECo proposes a 30-day shutdown requirement instead of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Inadequate Core Cooling Instrumentation - The licensee has proposed a minimum core exit thermocouple (CET) operability of two CET's in each of two core quadrants rather than two operable CET's per quadrant.

Control Room Habitability - The NRC informed NNECo that the NRC staff would review the existing Tech. Specs. (including those issued with License Amendment No. 100) to determine conformance with GL 83-37.

The NRC staff requested NNEco to amend its October 14, 1986 submittal, to address the draft TER, within 60 days following receipt of these meeting minutes.

/S/

D. H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B

Enclosure:

As stated

-cc w/ enclosure:

See next page PBD-8: PB ._ l P .

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, Mr. John F. Opeka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. 2 cc:

Gerald Garfield, Esq. Mr. Wayne D. Romberg Day, Berry & Howard Superintendent Counselors at Law Millstone Nuclear Power Station City Place P. O. Box 128 Hartford, Connecticut 06103-3499 Waterford, Connecticut 06385 Regional Administrator, Region I Mr. Edward J. Mroczka U.S. Nuclear Regulatory Commission Vice President, Nuclear Operations Office of Executive Director for Northeast Nuclear Energy Company Operations P. O. Box 270 631 Park Avenue Hartford, Connecticut 06141-0270 King of Prussia, Pennsylvania 19406 Mr. Charles Brinkman, Manager Washington Nuclear Operations C-E Power Systems Combustion Engineering, Inc.

7910 Woodmont Avenue Bethesda, Maryland 20814 Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Utfitties Service Company ATTN: Mr. Richard R. Laudenat, Manager Generation Facilities Licensing Post Office Box 270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director Radiation Control Unit Department of Environmental Protection '

State Office Building Hartford, Connecticut 06106 Mr. Theodore Rebelowski q

! U.S. NRC P. O. Box 615 Waterford, Connacticut 06385-0615 Office of Policy & Management i ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106

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MEETING SUMARY DISTRIBUTION Licensee: Millstone Unit No. 2 Distribution

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November 24, 1986

' Docket No. 50-336 MEMORANDUM FOR: Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B FROM: D. H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B

SUMMARY

OF THE NOVEMBER 14, 1986 MEETING WITH NORTHEAST NUCLEAR ENERGY COMPANY (NNECo) TO DISCUSS GL 83-37 (TMI TECH. SPECS.)

On November 14, 1986, representatives from NNECo and the NRC staff met in Room P-422 of the Phillips Building, in Bethesda, Maryland. The purpose of the meeting was to discuss NNECo's October 14, 1986 submittal on GL 83-37 for Millstone Unit No. 2.

The NRC staff informed NNEco that GL 83-37 contained NRC staff-approved Tech.

Specs. for hardware required for installation by NUREG-0737. In some instances, the requirements of GL 83-37 could be relaxed where staff-approved TMI modifications for Millstone Unit 2 proved inconsistent with the GL 83-37 Tech. Specs.

The discussions with the licensee centered upon the draft technical evaluation report (TER) shown in Enclosure 2 which was prepared for the NRC staff by EG&G, Idaho. The draft TER compared the GL 83-37 Tech. Specs. with those submitted by NNECo on October 14, 1986. The most significant issues raised by the comparison were:

High Point Vents - The NRC staff will allow credit for power operated relief valves (PORV's) provided that NNECo presents information to demonstrate that the PORV's substantially meet NRC requirements for high point vents. Plant shutdown will still be required for an inoperable head vent and/or pressurizer PORV/ vent.

l Post Accident Sampling - NNECo should propose a change to the Administrative Tech Specs; to require training, procedures, and

maintenance for post accident sampling. The proposed limiting l

condition for operation (LCO) and Surveillance Requirements are not adequate in this regard.

Auxiliary Feedwater (AFW) - NNECo should provide a flow value for AFW testing. As an alternative, NNECo may demonstrate that the existing flow testing in " recirculation" corresponds to a unique flow value.

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r-4 Noble Gas Effluent Monitor - NNECo should relocate the proposed' Noble Gas Effluent Fonitor Tech Spec to the Millstone Unit 2 Tech.

Spec. table containing other, similar, monitors.

Containment High-Range Rad:stion Monitor - The proposed Tech. Specs.

do not require use of a preplanned alternative monitor. In addition, NNECo proposes a 30-day shutdown requirement instead of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Inadequate Core Cooling Instrumentation - The licensee has proposed a minimum core exit thermocouple (CET) operability of two CET's in each of two core quadrants rather than two operable CET's per quadrant.

Control Room Habitability - The NRC informed NNEco that the NRC staff would review the existing Tech. Specs. (including those issued with License Amendment No. 100) to determine conformance with GL 83-37.

The NRC staff requested NNEco to amend its October 14, 1986 submittal, to address the draft TER, within 60 days following receipt of these meeting minutes.

/S/

D. H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B

Enclosure:

As stated cc w/ enclosure:

See next page PBD-8: PB . I P .

PhrAptzer DHo jch ATh ani 11/g /86 11g/86 1 6

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Enclosert 1 List of Attendees NRC D. H. Jaffe G. L. Plumlee J. Stachew (EG&G)

NNEco R. Bates E. Perkins 7

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Enclosure 2 EGG-NTA-7450 WV 1986 MEET 4bab ifSONE4/ 1 TECHNICAL EVALUATION REPORT OF MILL UNIT 2 NUCLEAR POWER STATION RESPONSE TO THE U. S. NUCLEAR REGULATORY COMitISSION, OFFICE OF NUCLEAR REACTOR REGULATION'S GENERIC LETTER 83-37 J. C. Stachew Prepared for the U. S. NUCLEAR REGULATORY COMMISSION

B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _ LEJ EGG-NTA-7450 DRAFT EG&G TECHNICAL EVALUATION REPORT MILLSTONE UNIT 2 NUCLEAR POWER STATION RESPONSE TO THE U.S. NUCLEAR REGULATORY COMMISSION, OFFICE OF NUCLEAR REACTOR REGULATIONS GENERIC LETTER NO. 83-37.

Docket No. 50-336 J. C. STACHEW Published December 1986 Idaho National Engineering Laboratory l

PREPARED FOR THE U.S. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C., 20555 UNDER DOE CONTRACT NO. DE-AC07-76ID01570

! FIN NO. 06022 l

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B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _ LEJ CONTENTS A B S T RA C T . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii FOREWORD .............................................................. iii

1. INTRODUCTION ..................................................... 1

2. DISCUSSION AND EVALUATION ........................................ 2 2.1 Reactor Coolant System Vents ( II .B.1 ) . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Post-Accident Sampling (II.B.3) ............................ 4 2.3 Long Term Auxiliary Feedwater System Evaluation (II.E.1.1) ...................................... 5 2.4 Noble Gas E f fluent Monitors (II .F .1.1 ) . . . . . . . . . . . . . . . . . . . . . 6 2.5 Sampling and Analysis of Plant Effluents (II.F.1.2) ........ 8 2.6 Containment High-Range Radiation Monitor (II .F.1.3) . . . . . . .. 9 2.7 Containment Pressure Monitor (II .F .1.4 ) . . . . . . . . . . . . . . . . . . . . 10 2.8 Containment Water Level Mon itor (II .F .1.5 ) . . . . . . . . . . . . . . . . . 11 2.9 Containment Hydrogen Monitor (II .F .1.6 ) . . . . . . . . . . . . . . . . . . . . 13 2.10 Instrumentation for Determination of Inadequate Cor e C oo l i n g ( I I . F . 2 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.11 Control Room Habitability Requirements (III.D.3.4) ......... 17

3. ADDITIONAL INFORMATION NEEDED TO COMPLETE THE REVIEW ............. 20

4.

SUMMARY

.......................................................... 21

5. REFERENCES ....................................................... 22 ii

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. ABSTRACT This EG&G Idaho, Inc., report evaluates various submittals provided by Northeast Utility for. Millstone Unit 2 Nuclear Power Station. The submittalsare in response to Generic Letters No. 83-37, "NUREG-0737 Technical Specifications (TS)." Applicable sections of the plants' Technical Specifications are evaluated to determine compliance to the guidelines established in the generic letter.

FOREWORD This report is supplied as part of the " Technical Assistance for Operating Reactors Licensing Actions," being conducted for the U.S. Nuclear Regulatory Commission, Washington D.C., by EG&G Idaho, Inc., NRR and I&E Support.

The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11 1, FIN No. 06022.

Docket No. 50-336 TAC No. f45d[

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o B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 LEJ ENCLOSURE 2 TECHNICAL EVALUATION REPORT

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MILLSTONE UNIT 2 NUCLEAR POWER PLANT

1. INTRODUCTION On November 1, 1983, a letter was sent by the Director, Division of Licensing "To All Pressurized Water Reactor Licensees." This Generic Letter (83-37) provided staff guidance on the content of the Technical Specifications associated with certain items in NUREG-0737. On November 25, 1985, October 4, 1985, and October 14, 1986, Northeast Utility filed responses to the generic letter for Millstone Unit 2 Nuclear Power Station. The following report provides the staff evaluation of the NEU submittals and makes recommendations for resolving the remaining issues.

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2. DISCUSSION AND EVALUATION The licensee wat requested to provide Technical Specifications for several different systems. Each of these proposals is discussed and evaluated in an individual subsection below:

2.1 Reactor Coolant System Vents (II.B.1)

The generic letter stated:

"At least one reactor coolant system vent path (consisting of at least two valves in series which are powered from emergency buses) shall be operable and closed at all times (except for cold shutdown and refueling) at each of the following locations:

a. Reactor Vessel Head

b. Pressurizer steam space

c. Reactor coolant system high point".

A typical Technical Specification for reactor coolant system vents was provided. For the plants using a power operated relief valve (PORV) as a reactor coola'nt system vent, the block valve was not required to be closed as long as the PORV was operable.

Evaluation of NEU's Responses

1. LCO 3.4.11 does not have a vent path for " reactor coolant system high point" contrary to the guidance. NNECO states that NRC approved the RCS vent system in SER dated September 20, 1983. Checked SER and this .

is true. Therefore, deviation has been previously approved by the staff.

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2. Action a deviates from the guide in that with an inoperable vent path, after 30-days, NEU chooses to submit a Special Report to the Commission vice shutdown (hot standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown in the following 30. hours). NEU takes advantage of their PORV if the pressurizer vent path is inoperable. That is, they allow continued operation indefinitely if the PORV is operable. Does the PORV meet the appropriate design requirements? It if does, then using the PORV with the inoperable pressurizer vent path closed with powered removed from the valve actuators of all the valves in the inoperable vent path is okay! Otherwise, use of the PORV is not an acceptable alternative. NEU's justification for not shutting down with an inoperable vent path is not acceptable. Stating that shutdown is not required because the vent valves are primarily for use for events beyond design basis accidents y ge g ble as this was an assumed condition by the NRC when the was developed, including shutdown for an inoperable vent path.

3. Action b deviates from the guide in that with an inoperable vent path, after 30 days, NEU chooses to submit a Special Report to the Commission vice shutdown (hot standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />). Again, NEU's justification for this deviation is unacceptable for the same reasons as stated in 2) directly above.

4. Action c deviates from the guide in that it lets progression from hot shutdown (mode 4) to hot standby (mode 3) to startup (mode 2) to power operation (mode 1). The guide only all'ows startup and or power operation to continue but not to commence with inoperable vent path (s). No justification is provided for this deviation.

5. NEU has no action for two inoperable vent paths. The guide requires restoration of at least two vent paths to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Actually NEU's actions allow continued 3

a B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _ LEJ operation indefinitely with two inoperable vent paths as long as the SpeciaHteport is made after 30 days. No justification is provided for this deviation.

As a result of the review of the material cited, it is considered that this issue has not been acceptably resolved.

2.2 Post-Accident Samplino (II.B.3)

The generic letter stated:

" Licensees should ensure that their plant has the capability to obtain and analyze reactor coolant and containment atmosphere samples under accident conditions. An administrative program should be established, implemented and maintained to ensure this capability. The program should include:

a. training of personnel

b. procedures for sampling and analysis, and

c. provisions for maintenance of sampling and analysis equipment.

"It is acceptable to the Staff, if the licensee elects to reference this program in the administrative controls section of the Technical Specifications and include a detailed description of the program in the plant operation manuals. A copy of the program should be easily available to the operating staff during accident and transient conditions.'"

Evaluation of NEUs Responses

1. The proposed TS 3.3.3.11 adequately ensures operability but deviates from the guidance in not emphasizing er even calling out:

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a. training of personnel )

b. procedures for sampling and analysis

c. provisions for maintenance of sampling and analysis equipment.

The review of the proposed changes indicates that the licensee has not adequately resolved this item.

2.3 Long Term Auxiliary Feedwater System Evaluation (II.E.1.1)

The generic letter stated:

"The objective of this item is to improve the reliability and performance of the auxiliary feedwater (AFW) system. Technical l

Specifications depend on the results of the licensee's evaluation and staff review of each plant. The limiting conditions of operation (LCO) and surveillance requirements for the AFW system should be similar to safety-related systems. Typical generic Technical Specifications are provided in ' Enclosure 3. These specifications are for a plant which has three auxiliary feedwater pumps. Plant specific Technical Specifications could be established by using the generic .

Technical Specifications for the AFW system."

Evaluation of NEU's Responses The licensee responded by identifying Amendment No. 53 as containing the required Technical Specifications. The lice,nsee's response deviates from the guidance as follows:

1. In LCO 3.7.12, the three steam generator auxiliary feedwater pumps are not identified as being independent and their ass,ciated flow paths are not called out to be operable.

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l l B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _ LEJ unummemus

4. In SR 4.7.1.2.2 A) and b) the discharge pressure developed should be DRAFI specified at greater than or equal to a minimum flow rate. M l

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5. In SR 4.7.1.2.a.6. remote operated valves are verified for correct l

position versus the guidance of fully open position when the auxiliary feedwater system is placed in automatic control or when above 10%

I rated therrral power.

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As a result of the review of the cited material, this issue is considered to not be adequately resolved.

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2.4 Noble Gas Effluent Monitors (II.F.1.1) l The generic letter stated

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" Noble gas effluent monitors provide information, during and  !

j following an accident, which are considered helpful to the operator in accessing the plant condition. It is desired that these monitors be operable at all times during plant operation, i

but they are not required for safe shutdown of the plant. In case of failure of the monitor, appropriate actions should be taken to restore its operational capability in a reasonable period of time. Considering the importance of the availability 8 of the equipment and possible delays involved in administrative f

ct.ntrols, 7 days is considered to be the appropriate time period to restore the operability of the monitor. An alternate method 1

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B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _, LEJ for monitoring the effluent should be initiated as soon as practical, but no later than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the identification of the failure of the monitor. If the monitor is not restored to operable cqnditions within 7 days after the failure a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the system to operable l status."

Evaluation of NEU's Responses =

1. NEU deviates from the guidance by not specifying:

I Mode applicability (should be modes 1, 2, 3, 4)

Alarm / trip setpoint Measurement range (should be 1-10 2 pCi/cc to 1-105 uCi/cc Multiple monitor locations (specifies stack only; guidance is for I stack and radwaste building exhaust; auxiliary building exhaust, steam safety valve discharge, shield building exhaust, and condenser exhaust).

l 2. NEU deviates from the guidance in their action statement by:

o Not initiating the preplanned alternative monitoring within

72. hours l

o Uses 30 days for restoration'vice guidance of 7 days NEU " justification for deviation" cites consistency with existing l

Section 3.3.3.10 for 30 days. This is not a safety justification and is therefore inadequate, ammmmmmmmme 7

B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 LEJ Un il i 2.5 Sampling and Analysis of Plant Effluents (II.F.1.2)

The generic letter stated:

"Each operating nuclear power reactor should have the capability to collect and analyze or measure representative samples of radioactive iodines and particulates in plant gaseous effluents during and following an accident. An administrative program should be established, implemented and maintained to ensure this capability. The program should include:

a. training of personnel

b. procedures for sampling and analysis, and

c. provisions for maintenance of sampling and analysis equipment.

Evaluation of NEUs Responses

1. Millstone Unit 2 does not call out the requirements explicitly for obtaining and analyzing:

a. Reactor coolant

b. Radioactive iodines and particulates in plant gaseous effluents

c. Containment atmosphere

d. Procedures for sampling and analysis.

No explicit justification is given for these deviations.

Millstone Unit 2 uses vague language in these areas. If in fact, programs are in place to accomplish the explicit GL 83-37 requirements, then the words in the technical specifications should be clarified to indicate this.

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- As a result of the review of the material cited, this item is considered to not be adeqtrately resolved.

2.4 Containment High-Range Radiation Monitor (II.F.1.3)

The generic letter stated:

"A minimum of two incontainment radiation-level monitors with a maximum range of 810 rad /hr(10 7 R/hr for photon only) should be operable at all times except for cold shutdown and refueling outages. In case of failure of the monitor, appropriate actions should be taken to restore its operational capability as soon as possible. If the monitor is not restored to operable condition within 7 days after the failure, a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the equipment to operable status,

" Typical surveillance requirements are shown in Enclosure 3. The setpoint for the high radiation level alarm should be determined such that spurious alarms will be precluded. Note that the acceptable calibration techniques for these monitors are discussed in NUREG-0737."

Evaluation of NEU's Responses ,

1. In Table 3.3-6, the alarm / trip setpoint proposed by NEU is 100R/hr.

This appears too high. The guidance was for a setpoint at 10 rad /hr.

2. Only 1 channel is specified for " minimum channels operable." The guidance is for 2 channels.

3. With less than the required minimum channels operable, NEU deviates from the guidance of initiating the preplanned alternate method of monitoring within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. NEU does not require any alternate method of monitoring.

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l 4. NEU proposes 30 days for restoration of an inoperable channel prior to l submitting a Special Report versus the guidance of 7 days. ,

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l Disagree with NEU argument that can deviate in action because not required I j for shutdown of any design basis accident.

I I 2.7 Containment Pressure Monitor (II.F.1.4) i i

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The generic letter stated:

l l " Containment pressure should be continuously indicated in the  !

! control room of each operating reactor during Power Operation, Startup and Hot Standby modes of operation. Two channels should l be operable at all times when the reactor is operating in any of l the above mentioned modes. Technical Specifications for these monitors should be included with other accident monitoring instrumentation in the present Technical Specifications.

Limiting conditions for operation (including the required Actions) for the containment press'ure monitor should be similar to other accident monitoring instrumentation included in the present Technical Specifications."

Evaluation to NEUs Responses

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2. There is no action statement for less than the " required number of )

channe M . This deviates from the guidance of restore to operable

status within 7 days or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3. For less than the " minimum channels operable"., NEU proposes 30 days for restoration followed by a Special Report versus the guidance of restore within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4. NEU uses " total number of channels" vice the guidance of " required number of channels".

2.8 Containment Water Level Monitor (II.F.1.5)

The generic letter stated:

"A continuous indication of containment water level should be provided in the control room of each reactor during Power Operation, Startup and Hot Standby modes of operation. At least one channel for narrow range and two channels for wide range

- instruments should be operable at all times when the reactor is operating in any of the above modes. Narrow range instruments should cover the range from the bottom to the top of the containment suma. Wide range instruments should cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon (orlessifjustified) capacity.

" Technical Specifications for containment water level monitors ,

should be included with other accident monitoring instrumentation in the present Technical Specifications. LCOs(includingthe required Actions) for wide range monitors should be similar to other accident monitoring instrumentation included in the present Technical Specifications. LOCs for narrow range monitor should include the requirement that the inoperable channel will be restored to operable status within 30 days or the plant will be brought to Hot Shutdown condition as required for other accident l 11

B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _ LEJ monitoring instrumentation. Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3." B-Evaluation of NEU's Responses D '

2. NEU proposes no action for less than the " required number of channels". The guidance is to restore the inoperable channel (s) to operable status within 7 days (30 days for narrow range) or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3. With less than the " minimum channels operable", NEU proposes to restore the inoperable channel (s) to operable status within 30 days or submit a Special Report within the next 10 days versus the guidance of being in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for narrow range instruments.

4. For wide range instruments, NEU with less than the " minimum channels operable" proposes to restore within 30 days or submit a Special Report vice the guidance of restore in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5. NEU uses " total number of channels" versus the guidance of " required number of channels." ,

I disagree with NEU's argu. ment that they can deviate from the guidance action statements because containment water level monitors are not required for shutdown of any design basis accident. The intent of these monitors was for coping with beyond design basis accidents and this is generic to the majority of NUREG-0737 and GL 83-37 requirements.

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- 2.9 Containment Hydrogen Monitor (II.F.1.6)

The generic letter stated:

"Two independent containment hydrogen monitors should be operable at all times when the reactor is operating in Power Operation or Startup modes. LCO for these monitors should include the requirement that with one hydrogen monitor inoperable, the monitor should be restored to operable status within 30 days or the plant should be brought to at least a hot standby condition within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If both monitors are inoperable, at least one monitor should be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant should be brought to at least hot standby condition within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Typical surveillance requirements are provided in Enclosure 3."

Evaluation of NEU's Responses

1. Action statement deviates from the guidance in that with one hydrogen monitor inoperable after 30 days, Millstone Unit 2 requires hot standby within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> versus the guidance of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. No justification is given for this deviation.

2. There is no action statement with both hydrogen monitors inoperable.

No justification is given for this deviation.

3. There is no surveillance requirement for a channel check at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or for an analog channel operational test at least once per 31 days. No justification is given for this deviation.

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The generic lettpr stated: DRAFT "Subcooling margin monitors, core exit thermocouples, and a reactor coolant inventory tracking system (e.g., differential pressure measurement system designed by Westinghouse, Heated Junction Thermocouple System designed by Combustion Engineering, etc.) may be used to provide indication of the approach to, existence of, and recovery from inadequate core c.ooling(ICC). These instrumentation should be operable during Power Operation, Startup, and Hot Shutdown modes of operation for each reactor. "Subcooling margin monitors should have already been included in the present Technical Specifications. Technical Specifications for core exit thermocouples and the reactor coolant inventory tracking system should be included with other accident monitoring instrumentation in the present Technical Specifications. Four core-exit thermocouples in each core quadrant and two channels in the reactor coolant tracking system are required to be operable when the reactor is operating in any of the above mentioned modes. Minimum of two core-exit thermocouples in each quadrant and one channel in the reactor coolant tracking system should be operable at all times when the reactor is operating in any of the above mentio1ed modes.

Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are pro'vided in Enclosure 3."

Evaluation of NEU's Responses

1. NEV proposes for " minimum channels operable" in their Table 3.3-11, 2 core exit thermocouples in each of 2 core quadrants vite the guidance of 2 per core quadrant. No justification is given for this deviation.

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2. NEU has a footnote to the calibration frequency "R" for core exit thermoeottples and reactor vessel coolant level. The footnote indicates " electronic calibration from the ICC cabinets only". Need to check what other CE units do. The CEOG proposal of February 19, 1985, does not have such a footnote (nor does San Onofre 2 and 3; CE unitsalso.)

3. NEU uses " total number of channels" in Table 3.3-11 vice the guidance of " required number of channels".

4. NEU proposes only one action to determine subcooling margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with less than the minimum channels operable. The guide has two actions. One for less than the required number of channels--namely, restore the inoperable channel (s) to operable status within 7 days or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. And one action statement for less than the minimum number of channels--namely, restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. NEU provides no justification for this deviation.

5. For core exit thermocouples.

o NEU' proposes no action statement for less than the required number of channels. The guide requires to restore the inoperable channel (s) within 7 days or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. NEU provides no justification for this deviation, o NEU proposes to restore the inoperable channel (s) within 30 days or be in hot standby within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> versus the guidance of restore within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. NEU's justification for deviation is inadequate.

NEU's justification is that core exit thermocouples are not required to safely shutdown the plant. This condition, although not stated in the generic letter or NUREG-737, is assumed for the post TMI action items. The instrumentation availability is primarily to cope with a beyond design basis accident. Shutdown 15

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l is called for with continued channel (s) inoperability so as to reduce the risk of experiencing a beyond design basis event with

! the instrumentation inoperable.

l 6. For reactor coolant inventory tracking.

o With less than the required number of channels, NEU has no action l

statement. The guide requires to restore the inoperable channel (s) within 7 days or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. NEU provides no justification for this deviation.

! o With less than the minimum number of channels, NEU requires restoration in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or a Special Report in 30 days and restoration at the next refueling. This deviates from the guidance of being in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if not l restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. NEU's justification for the deviation l is that their action is very similar to that proposed by the CE l

Owner's Group, letter dated 2/19/85. However, NEU's failure to provide an alternate method of monitoring is unacceptable, as it

is this feature of CE Owner's Group generic tech. spec. that makes up for not shutting down.

NEU's justification for not initiating an alternate method of monitoring is l that clear step-by-step guidance could not be given to the operators on how to meet this step. NEU should either develop the appropriate guidance or l

l require hot standby within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,if operability has not been l regained after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

NEU's footnote for the operability of the HJTC (four or more sensors i

operable, two or more in the upper four and two or more in the lower four)

{

is consistent with the NRC Staff's evaluation of the Combustion Engineering Owners Group proposed Technical Spe' fications for the HJTC system.3 16

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LEJ 2.11 Control Room Habitability Requirements (III.D.3.4)

The generic letter stated:

" Licensees should assure that control roon operators will be adequately protected against the effects of the accidental release of toxic and/or radioactive gases and that the nuclear power plant can be safely operated or shutdown under design basis accident conditions. If the results of the analyses of postulated accidental release of toxic gases (at or near the plant) indicate any need for installing the toxic gas detection system, it should be included in the Technical Specifications.

Typical acceptable LCO and surveillance requirements for such a detection system (e.g. chlorine detection system) are provided in Enclosure 3. All detection systems should be included in the Technical Specifications.

"In addition to the above requirements, other aspects of the control room habitability requirements should be included in the Technical Specifications for the control room emergency air cleanup system. Two independent control room emergency air cleanup systems should be operable continuously during all modes of' plant operation and capable of meeting design requirements.

Sample Technical Specifications are provided in Enclosure 3."

s Evaluation of NEU's Responses (Amendments Number 100 and 72).

I. CHLORINE DETECTION SYSTEMS

1. Millstone Unit 2 deviates from the guidance on chlorine detection systems in the applicability statement in that only Modes 1, 2, 3, and 4 are identified versus the guidance of all modes. No justification is given for this deviation.

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2. Millstone Unit 2 does not have a surveillance requirement on an armhg channel operational test at least once per 31 days. No justification is given for this deviation.

3. Each chlorine detection system has a surveillance for verifyin'g l

l energization at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> versus the guidance of demonstrate operable by performance of a channel check at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. No justification is given for this deviation.

l 1 II. CONTROL ROOM EMERGENCY VENTILATION SYSTEM

1. Millstone Unit 2 deviates from the guidance in the applicability statement in that only modes 1, 2, 3, and 4 are specified versus the guidance of all modes. No justification is given for this deviation.

2. The action statement to be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after 7 days with one system inoperable deviates from I

the guidance of be in hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in l

cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. No justification is given for this deviation.

3. Since there is no applicability to Modes 5 and 6, there is also a lack of an action statement for these modes.

4. Surveillance requirement 4.7.6.1.a, on control room air temperature $120*F deviates by a large amount from the recommended 80*F. Control room personnel could not function effectively for any length of time at such a high temperature.

No justification is given for this deviation.

5. Surveillance requirement 4.7.6.1.C on operating emergency ventilation flow for at least 15 minutes deviates by a large amount from the guidance of 10 continuous hours with the heaters operating. No justification is given for this devi,ation.

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. 6. In 4.7.6.1.c, "0R" should be "AND" before the "(1)" and "(2)".

TMris a recognized miswording in the guidance which is being corrected.

7. Surveillance requirement 4.7.7.e.1 on total bypass flow of the system to the facility vent, including leakage through the system diverting valves', is missing. This is allowable if the system has no diverting valves. Need to have NEU describe their system l

or check FSAR.

8. Millstone Unit 2 lacks a surveillance requirement like guidance l item 4.7.7.e.3 on maintaining a positive pressure of at least f 1/8 inch W.G. in the control room. No justification is given for this deviation.

9. Millstone Unit 2 lacks a surveillance requirement like guidance item 4.7.7.e.4 on verifying that the heaters dissipate their kW design rating.

l l

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B 11/12/86 TSB Doc 3283H Disk 3300H Job 38503 Proof 3 _ LEJ 3c--ADDITIONAL INFORMATION NEEDED TO COMPLETE THE REVIEW To permit an ind,ependent Safety Evaluation, the licensee should be requested to provide an updated schedule for completion of the Technical Specifications for the following Projects:

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4.

SUMMARY

The following subsections describe those issues that are considered to have not been satisfactorily addressed by the Licensee:

1. Reactor Coolant System Vents (II.B.1)

2. Post-AccidentSampling(II.B.3)

3. Long-Term Auxiliary Feedwater System Evaluation (II.E.1.1)

4. Noble Gas Effluent Monitor (II.F.1.1)

5. Sampling and Analysis of Plant Effluents (II.F.1.2)

6. Containment High-Range Radiation Monitors (II.F.1.3)

7. Containment Pressure Monitor (II.F.1.4) l 8. Containment Water Level Monitor (II.F.1.5)

! Containment Hydrogen Monitor (II.F.1.6) 9.

10. Instrumentation for Determination of Inadequate Core Cooling (II.F.2)

! 11. Control Room Habitability Requirements (III.D.3.4) l l

l 1

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_ 5. REFERENCES

1. D. G. Eisenhut, NRC letter to all Pressurized Power Reactor Licensees, "NUREG/0737 Technical Specifications (Generic Letter 83-37),"

November 1, 1983.'

2. NUREG-0737, Clarification of TMI Action Plan Requirements, published by the Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, November 1980.

3. D. M. Crutchfield letter to R. W. Wells, Safety Evaluation of Generic Technical Specification Proposed by Combustion Engineering Owners Group for the Reactor Vessel Level Monitoring System, U.S. Nuclear Regulatory Commission, October 28, 1986.

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Division of Licensin9 Technical Evaluation Report Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission liashington, DC 20555 13 swP8LSMSNT ARY 40733

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Interim Technical Evaluation Report on the audit-of the tiillstone Unit 2 Nuclear Power Plant Technical Specifications perforned for the NRC in connection with conformance to the requirements of the NRR Generic Letter No. 83437, "NUREG-0737 Technical Specifications".

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