Text

Technical Evaluation Rept for Haddam Neck Plant Response to Us Nrc,Nrr Generic Ltr 83-37
	ML20246F835
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Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	11/30/1988
From:	Stachew J; EG&G IDAHO, INC.
To:	; NRC
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ML20246F811	List: ML20246F813; ML20246F835; ML20246F844;
References
	CON-FIN-D-6022, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.B.3, TASK-2.E.1.1, TASK-2.F.1, TASK-2.F.2, TASK-3.D.3.4, TASK-TM EGG-NTA-8237, GL-83-37, NUDOCS 8905150117
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. 1 EGG-NTA-8237 1

TECHNICAL EVALUATION REPORT FOR HADDAM NECK PLANT RESPONSE TO THE U.S. NUCLEAR REGULATORY COMMISSION, t OFFICE OF NUCLEAR REACTOR REGULATION'S GENERIC LETTER NO. 83-37 l

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Docket No. 50-213 J. C. Stachew Published November 1988 Idaho National Engineering Laboratory EG&G Idaho, Inc.

i Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. D6022 P

8 ABSTRACT r g

-This EG&G Idaho, Inc.., report evaluates the submittal provided by Connecticut Yankee Atomic Power Company for the Haddam Neck Plant. The t submittal is in response to Generic Letter No. 83-37, "NUREG-0737 Technical Specifications (TS)." Applicable sections of the 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., Regulatory and Technical Assistance.

The U. S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11 1, FIN No. D6022.

Docket No. 50-213 TAC Nos. 54393 and 54538 ii

CONTENTS 4

ABSTRACT............................................................. ii i F0 REWORD............................................................. ii

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

2. DISCUSSION AND EVALUATION....................................... 2 2.1 Reactor Coolant System Vents (II.B.1)...................... 2 2.2 Postaccident Sampling (II.B.3)............................. 4 2.3 Long Term Auxiliary Feedwater System '

Evaluation (II.E.1.1)...................................... 5 2.4 Nobl e Gas Ef fl uent Monitors (II .F. I .1) . . . . . . . . . . . . . . . . . . . . . 7 2.5 Sampling and Analysis of Plant Effluents (II.F.1.2).. . . . . . . 9 2.6 Containment High-Range Radiation Monitor (II.F.1.3)........ 10 2.7 Containment Pressure Monitor (II.F.1.4).................... 11 2.8 Containment Water Level Monitor (II.F.1.5)................. 13 2.9 Containment Hydrogen Monitor (II.F.1.6).................... 14 2.10 Instrumentation for Detection of Inadequate Core Cooling (II.F.2)...................................... 15 2.11 Control Room Habitability Requirements (III.D.3.4) . . . . . . . . . 19

3. ADDITIONAL INFORMATION REQUIRED TO COMPLETE THIS REVIEW......... 20 3.1 Reactor Coolant System Vents (II.B.1)...................... 20 3.2 Long Term Auxiliary Feedwater System Evaluation -(II.E.1.1). 20 3.3 Noble Gas Ef fl uent Moni tors (II .F.1.1) . . . . . . . . . . . . . . . . . . . . . 21 3.4 Containment High-Range Radiation Monitor (II.F.1.3)........ 22 3.5 Containment Water Level Monitor (II.F.1.5)................. 22 3.6 Instrumentation for Detection of Inadequate Core Cooling (II.F.2)........................................... 22

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SUMMARY

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5. REFERENCES...................................................... 25 iii

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. j TECHNICAL EVALUATION REPORT FOR (,

'HADDAM NECK PLANT RESPONSE TO THE U.S. NUCLEAR REGULATORY COMMISSION.

OFFICE OF NUCLEAR REACTOR REGULATION'S GENERIC LETTER NO. 83-37

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)I provided NRC Staff guidance on the contents of the Technical Specifications (TSs) associated with certain items in NUREG-0737.2 Connecticut Yankee Atomic Power Company (CYAPCo) filed a response to Generic Letter 83-37 for the Haddam Neck Plant in their letter of July 25, 1984.3 In this letter, CYAPCo' stated that it was inappropriate to propose technical specifications for reactor coolant I system vents, Item II.B.1 of Generic Letter 83-37, until safety concerns connected with hydrogen reactions at the vent discharge were resolved.

CYAPCo further stated that the other Generic Letter 83-37 items were to be addressed in the conversion to the Standard Technical Specifications.4 The NRC Staff subsequently requested CYAPCo to submit technical specifications for all of the eleven items in Generic Letter 83-37 including item II.B.I. The following report provides the evaluation of 5

the CYAPCo submittal of July 1, 1988 on all of the Generic Letter 83-37 items and makes recommendations for resolving the remaining issues.

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2. DISCUSSION AND EVALUATION- .

j. The Licensee was requested to provide Technical Specifications for several different systems. Each of these proposals is discussed and evaluated'in an individual subsection below.

.I 2.1 Reactor Coolant System Vents (II.B.1) i J

The Generic letter I contains the following statement:

"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- J cold shutdown and refueling) at each of the following locations: I

a. Reactor Vessel Head

b. Pressurizer steam space

c. Reactor coolant system high point "A typical Technical Specification for reactor coolant system vents is provided in Enclosure 3. For the plants using a power operated relief valve (p0RV) as a reactor coolant system vent, the block valve is not required to be closed if the PORV is operable."

Evaluation The Licensee responded by letter dated July 1,1988 .5 The Haddam Neck technical specification specifies reactor vessel head and pressurizer steam space vent paths. However, no reactor coolant system high point vents are specified. This lack of reactor coolant system high point vents is considered closed as the NRC Staff accepted this in their SER 6 approving the vent system design. The following deviations from the Generic Letter 83-37 guidance are noted.

1. In Specification 3.3.5.1 Action a. and b. with one reactor coolant system vent path inoperable, the action is to restore the inoperable vent path within 30 days or submit a Special Report to the Commission within 10 days per Specification 6.9.2. This deviates from the guidance of after 2

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. 1 30 days, be in Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown

, ., within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . In NRC letter dated' September I 28, 1987,7 the NRC Staff accepted-this type deviation'in their approval of similar Action statements for the Millstone Nuclear Power Station, Unit No. 2. It was judged there that the major

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failure modes of. the vent valves of leakage or loss of vital 120 V AC power are already separately addressed in the technical specifications. The Haddam Neck technical specification 3.12, Station Service Power, does not require shutdown on loss of one emergency AC bus or one station battery and does not.specify any.

electrical. power requirement for the Hot Standby or Hot Shutdown modes. Per the Haddam Neck FSAR, Revision 0, dated May 15, 1987, P. 62.of 69, the reactor head vents and pressurizer vents- - -

have their control power supplied from the station batteries

'(each of the two parallel valve sets being supplied from different batteries). Haddam Neck technical specification 3.14, Primary System Leakage, does require reactor shutdown following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months with Reactor Coolant System leakage in excess of ten GPM.

However, since Haddam Neck only has two vent paths (pressurizer and reactor ' vessel head) the appropriate action for one vent path inoperable per Generic Letter 83-37 is to restore at least

., two vent paths to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The intent of the Generic Letter 83-37 is to always have at least two vent paths operable so that a single failure in one path (say one valve failure to open) does not defeat the capability for removing trapped steam and noncondensable gases which may inhibit core cooling during natural circulation under beyond design bases events.

Neither the Licensee's justification for deviation nor the NRC Safety Evaluation Report (SER) for the Millstone Nuclear Power Station Unit 2, addresses this concern of maintaining at least 3

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two operable vent paths. Requiring shutdown after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with less than two operable ver.t paths would not be an undue hardship -

on the Licensee since the Licensee's design for each vent path is a parallel set of two solenoid-operated valves in series. I

, Thus there would have to be a loss of.both parallel paths to get f one inoperable vent path. [

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2. The asterisk footnote on technical specification P. 3-4q of j

" Power to the Valves may be removed" appears to be contrary to ' i the following NRC Staff confirmatory requirement identified in their SER6 approving the design of the vent system:

- "The current design does not provide for continuous d positive valve position indication in the control room per the requirement of NUREG-0737, Item II.B.1 subitem A(5) and

subitem A(6) concerning the requirement for operability of the vent system from the control room. An acceptable resolution would be for the licensee to restore continuous control power supply to the PCS vents system by deleting its commitment to rack out the related circuit breakers during normal operation. The staff has evaluated that the related requirements in 10 CFR 50.44(c)(3)(iii) for the inadvertent or irreversible actuation of a vent can be adequately met by the. switching systems proposed for the individual valves on the vent system. .Therefore, removal of power is not necessary. The licensee is required to take the necessary action to meet these requirements. This item must be confirmed by the licensee."

As a result of the review of the material cited, the Licensee needs to submit additional information to justify how the proposed Technical Specifications meet the Generic Letter (Item II.B.1) guidance or submit revised specifications that meet the guidance.

2.2 Postaccident Samolina (II.B.3)

The Generic LetterI contains the following statement:

" 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:

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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."

A typical Technical Specification for postaccident sampling was provided that required the capability to sample and analyze radioactive iodines and particulate in plant gaseous effluents.

Evaluation:

The Licensee responded in letter dated July 1, 1988,5 proposing Administrative Controls Section 6.19. The Licensee proposes in Section 6.19 wording essentially identical to the guidance in the Generic Letter 83-37.

As a result of the review of the material cited, the Licensee's response is judged to meet the guidance of Generic Letter 83-37 for item II.B.3, Postaccident Sampling.

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

The Generic Letterl contains the following statement:

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

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m Evaluation: '

4 -The Licensee responded in l'etter dated' July 1, 19885 by identifying that existing technical specification Section 4.8 already adequately treats operability and surveillance of the auxiliary feedwater system.

The following deviations from the Generic Letter 83-37 guidance are notec:

1. There is no operability statement per se. Thus, so even though the surveillance is for "each" steam turbine driven pump, the number of pumps is not specified.

2. Haddam Neck existing specification 4.8 does not specify any mode applicability whereas the guidance-is for operability in modes 1, 2, 3.

3. There is no operability statement for the auxiliary feedwater system associated flow paths or operable steam supply system.

4. There are no Action statements for one, two, or three auxiliary feedwater pumps inoperable as in the guidance.

5. .The 31 day surveillance testing is not specified as " staggered" as in the guidance. The staggered testing is to prevent failure of both auxiliary feedwater pumps from a common cause maintenance error.

6. Specification 4.8.1.a requires that each steam turbine driven ,

pump develop a discharge pressure of greater than or equal to f

800 psig every 31 days but does~not specify an associated gpm flow as in the guidance. CYAPCo argues that the discharge pressure is a check on minimum recirculation flow rather than at a specified flow. However, the specification makes no mention :

of recirculation flow or of any valve lineup in the recirculation flow path.

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7. There is no 31 day surveilla'nce requirement for a check that each automatic valve in the. flow path is in the fully open position whenever the. auxiliary feedwater system is placed in automatic control or when above 10% rated thermal power.

8. There is no surveillance requirement requiring a dedicated individual to be stationed near any manually realigned valves during surveillance testing when only one auxiliary feedwater.

train is available.

9. There is no surveillance requirement to demonstrate flowpath operability prior to startup after any refueling outage or other cold shutdown of longer tha.n 30 days. .

As a result of the review of the cited material, the Licensee needs to submit additional information to justify how the cited Technical Specifications' meet the Generic Letter for this Item, II.E.1.1, Long Term Auxiliary Feedwater System Evaluation, or submit revised specifications that meet the guidance.

2.4 Noble As Effluent Monitors (II.F.1.1)

The Generic Letterl contains the following statement:

" Noble gas effluent monitors provide information, during and 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, 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 of the equipment and possible delays involved in administrative controls, 7 days is considered to be the appropriate time period' to restore the operability of the monitor. An alternate method for monitoring the effluent should be initiated as soon as i practical, but no later than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the identification l of the failure of the monitor. If the monitor is not restored to operable conditions 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 status."

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4 A' typical Technical Specification for. noble gas effluent' monitors was *

' also providedl that specified monitor locations and measurement ranges., . ,

Evaluation:

'The Licensee responded in letter dated July 1,1988.5 The i following deviations from the Generic Letter' 83-37 guidance are noted:

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1. .The Licensee has specified only one of the seven monitoring locations given in the guidance and this one location has no redundancy. The Licensee did not provide any justification for not providing the noble gas monitor information for the other monitoring locations given in the Generic Letter 83-37 guidance.

2. In Table 3.23-1, no alarm / trip setpoint or measurement range was specified for the noble gas effluent monitor. .The guidance in Generic Letter 83-37 provided for an alarm / trip setpoint and measurement range that were dependent on the monitor location.

CYAPCo provided no justification for this deviation except to state that the alarm / trip setpoint would be administrative 1y controlled in the plant's surveillance procedure. If the Licensee means the alarm / trip setpoint is changed frequently, say per the Offsite Dose Calculation Manual (0DCM) or some similar procedure and that the alarm trip setpoint is set on the lowest range of the moaitor very close to background, then that would be acceptable. Hovsver, if the alarm trip setpoint is fixed, it should be specified in Table 3.23-1.

3. No Action is specified as in the guidance for a radiation monitoring channel alarm / trip setpoint being exceeded. The guidance requires adjustment of the setpoint to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the channel should be declared inoperable.

4. A triple asterisk footnote in Table 3.23-1 allows isolation of the stack wide range noble gas monitor during periods of high 8

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steam generator blowdown. Although mentioned in the CYAPCo

._ , Position,'no justification was given for this deviation.

Because of potential primary to secondary coolant. leakage, some noble gases could be present in steam generator blowdown water and it would appear that the stack noble gas monitor should be operable during blowdown periods to measure any potential . noble.

gases released through the stack.

5. The Action statement for less than the Minimum Channels Operable does not provide for a preplanned alternate method of noble gas monitoring within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months contrary to the guidance of the Generic Letter.

6. CYAPCo requires a calibration quarterly, which is more often than the guidance of once per 18 months but does'not provide for.

a monthly functional'. test. This trade-off is judged adequate as the calibration is a more complete test and a quarterly frequency is comparatively very high.

7. The guidance is for a Channel Check each shift whereas CYAPCo specified one only daily. This small deviation is judged acceptable.

As a result of the review of the material cited, the Licensee needs to submit additional information to justify how the cited Technical Specifications meet tne Generic Letter (Item II.F.1.1) guidance or submit revised specifications that meet the guidance.

L 2.5 Samplina and Analysis of plant Effluents (II.F.1.2)

The Generic Letterl contains the following statement:

"Each operating nuclear power reactor should have the capability to collect and analyze or measure representative samples of radioactive iodines and particulate 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:

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4 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 readily available to the operating staff during accident and transient conditions."

Evaluation:

The Licensee responded in letter dated July 1, 1988,5 proposing Administrative Controls Section 6.19. The Licensee proposes in Section 6.19 wording essent ally identical to the guidance in the Generic Letter 83-37.

As a result of the review of the cited material, this Item, ii.F.1.2, Sampling and Analysis of Plant Effluents, is judged to meet the guidance of Generic Letter 83-37.

2.6 Containment Hiah-Ranoe Radiation Monitor (II.F.1.3)

The Generic Lette'rl contains the following statement:

radiation-level monitors with a "A minimum maximum of of range two10igrad containmen)

/hr (10 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 iroperability, 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."

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Evaluation:

The Licensee responded in letter dated July 1, 19885 proposing Technical Specification 3.23. -The following deviations from the Generic Letter 83-37 guidance are noted: -

1. In Table 3.23-1, no alarm / trip setpoint or measurement range was specified for the high-range radiation monitors. The guidance requires both to be specified. CYAPCo provided no justification for this deviation.

2. No~ Action is specified, as in the guidance when the alarm / trip setpoint is exceeded, to adjust-the setpoint to within~its-limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable. No justification was provided for this deviation.

3. The Channel Check guidance is once per shift whereas the CYAPCo frequency is once per day. This small deviation is judged acceptable.

4. CYAPCo requires I,o functional test but the guidance is for a functional test at least once per month. No justification was provided for.__this deviation.

5. On P 3-46b, Action 7, the word " operable" should be capitalized in two places.

As a result of the review of the cited material, the Licensee needs to submit additional information to justify how the cited Technical Specifications meet the Generic' Letter (Item II.F.1.3) guidance or submit revised specifications that meet the guidance.

2.7 Containment Pressure Monitor (II.F.1.4)

The Generic LetterI contains the following statement: 1 11 I

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" Containment pressure should be continuously indicated.in the -

control room of each operating reas.;or during Power Operation, Startup and Hot Standby modes of operation. Two chan.wls should ,

be operable at all times when the reactor is operating in any of 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 pressure monitor should be similar to.other accident monitoring instrumentation included in the-present Technical Specifications. Typical acceptable LC0 and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3."

Evaluation The Licensee responded in letter dated July 1, 1988,5 proposing technical specification 3.23. The following deviations from the Generic Letter 83-37 are noted:

1. With less than the Total No. of Channels operable beyond 7 days in Table 3.23-1, CYAPCo requires a Special Report to the Commission within 10 days whereas the guidance is to shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Although this is a deviation from the guidance, it is within the bounds of actions previously accepted by the NRC Staff and is, therefore, judged acceptable.

2. With less than the " minimum channels operable," CAPCo requires restoration of operability within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or submittal of a Special Report to the Commission within the next 10 days. The guidance requires restoration in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Again, although this is a deviation from the guidance, it is within the bounds of Actions l 7

previously accepted by the NRC Staff and is, therefore, f judged acceptable.

As a result of the review of the cited material, the Licensee's response is judged to meet the requirements of the Generic Letter for Item i II.F.1.4, Containment Pressure Monitor.

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y 2.8 Containment Water level Monitor (II.F.1.5)

The Generic Letterl contains the following statement:

"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 sump. Wide range instruments should cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or less if justified) capacity.

" Technical specifications for containment water level monitors should be included with other accident' monitoring instrumentation in the present Technical Specifications. LCOs ',

(including the required Actions) for wide range monitors should be similar to other accident monitoring instrumentation included .

in the present Technical Specifications. LCOs for narrow range

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monitor should include the requirement that the inoperable channel will be restored to operable status with 30 days or the plant will be brought to Hot Shutdown condition as required for !

other accident monitoring instrumentation. Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3."

Evaluation:

The Licensee responded in letter dated July 1, 19885 , providing specification 3.23. The following deviations from the Generic Letter 83-37 guidance are noted.

1. CYAPo did not provide a narrow range containment water level

! instrument as required by the guidance. Per the NRC Staff Safety Evaluation8of Item II.F.1.5 per NUREG-0737 requirements,2 a narrow range containment water level was installed and per Generic Letter 83-37 should be specified in the technical specifications.

2. With less than the Total No. of wide range containment water j

level Channels operable beyond 7 days, CYAPCo requires a Special 13

4 Report to the Commission within 10 days whereas the guidance is to shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Although, this is a deviation.1 rom- .

the guidance, it'is within the bounds of Actions previously 7

accepted by the NRC Staff and is, therefore, judged acceptable.

3. With less than the " minimum channels operable"- for the wide range containment water level, CYAPCo requires restoration of operability within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or submittal of a Special Report to the Commission within the next 10 days. The guidance requires restoration in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least Hot Shutdown within the next.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Again, although this is a deviation from the guidance, it.is within the bounds of Actions previously accepted by the NRC Staff7 and is, therefore, judged acceptable.

As a result of the lack of technical specifications for the narrow range containment water level instrument, the Licensee's response is judged to not meet the guidance of the Generic Letter for Item II.F.1.5, 4

-Containment Water Level Monitor.

2.9 Containment Hydroaen Monitor (II.F.1.6)

The~ Generic letterl contains the following statement:

"Two independent containment hydrogen monitors should be l operable at all times when the reactor is operating in Power j Operation or Startup modes. LC0 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If both monitors are inoperable, at least one monitor should be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant should be brought to at least hot standby condition within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Typical ' I surveillance requirements are provided in Enclosure 3."

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The Licensee responded in letter dated July 1, 1988,5 stating that l no specifications are proposed for this item because the need for 14 l

continuous hydrogen monitors is being evaluated in the Haddam Neck Plant

.. 4 Integrated Safety Assessment Program. The NRC Staff is in agreement with this position.

As a result of the review of the cited material, the NRC Staff is evaluating the need for hydrogen monitors, Generic Letter 83-37 Item II.F.1.6, under the Haddam Neck Plant Integrated Safety Assessment Program.

2.10 instrumentation for O_ete_ tion of Lrta.deguate Core Coolina (II.F.2)

The Generic LetterI contains the following statement:

"Subcooling margin monitors, core exit thermocouple, and a reactor coolant inventory tracking sytem (sic) (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 cooling (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 thermocouple and the reactor coolant inventory tracking system should be included with other accident monitoring instrumentation in the present Technical

., Specifications. Four core-exit thermocouple 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 thermocouple 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 mentioned modes.

Typical acceptable LC0 and surveillance requirements for accident monitoring instrumentation are provide in Enclosure 3."

Evaluation The Licensee responded in letter dated July 1, 1988,5 proposing technical specification 3.23. The following deviations from the Generic Letter 83-37 guidance are noted.

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Core Exit Thermocouple (CET):

1. CYAPCo specifies in Table 3.23-1 the Total No. of Channels as 16/ core rather than 4/ quadrant per Generic Letter 83-37 guidance. This issue is assumed closed as the NRC Staff accepted in their SER9 that the 4th quadrant in the core does .

. not meet the NUREG-0737 minimum requirement of 4 CETs. As described in CYAPCo's letter of February 1, 1985,10 the other three quadrants have at least 4 CETs/ quadrant. Therefore, it is recommended that the Total No. of Channels be specified as 4/ quadrant rather than 16/ core, with a footnote that there are only 3/ quadrant in Quadrant IV.

2. The asterisk footnote on Minimum Channels Operable is that j

" Quadrant III may have only one channel Operable" versus the guidance of at least 2. On page 31 of Attachment I to the CYAPCo letter of February 1, 1985,10 it is identified that there are 6-1/4 CETs in Quadrant III. Therefore, it is not obvious why the Minimum Channels Operable is specified as "may have only one channel Operable," and CYAPCo has provided no

^

justification for this deviation even though they stated that there are 6-1/4 CETs in Quadrant III.

3. With less than the Total No. of CET channels operable beyond 7 days, CYAPCo requires a Special Report to the Commission within 10 days whereas the guidance is to shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Although this is a deviation from the guidance, it is within the ,

bounds of Actions previously accepted by the NRC Staff7 and is, therefore, judged acceptable.

4. CYAPCo has now provided no Action for less than the " minimum channels operable" for the CETs. The guidance requires ,

restoration in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . i J

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5. In Table 3.23-2, CYAPCo specifies by a footnote that Channel

.. . Calibration for the CETs is " electronic calibration from the ICC' cabinets only." This is a deviation from the Generic Letter 83-37 which has'no such footnote. As thermocouple themselves--

are not calibrated but if inoperable are usually just replaced during refueling outages, only electronic calibration from the ICC cabinets is judged acceptable.

Reactor Vessel Water Level (RVWL):

1. CYAPCo specifies a footnote-(not present in the Generic Letter 83-37 guidance) for the RVWL channels clarifying operability as one or more sensors operable of the upper two in the probe and three or more sensors operable of the lower six in the probe. A channel is composed of eight sensors in a probe. This deviation is acceptable as it was required by the NRC Staff in their Safety Evaluation Il of the Combustion Engineering Owners Group (CE0G) technical specifications for other reactor designs [other than Combustion Engineering (CE) System 80 and non-System 80 CE designed reactors] using the CE Heated Junction Thermocouple (HJTC). CYAPCo stated that they use the CE HJTC design in Appendix A o'f their letter of February 1, 1985.10

2. CYAPCo provides no Action statement for less than the Total No.

of Channels operable. The Generic Letter guidance is to restore the inoperable channel (s) to operable status within 7 days or be in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Also, although CYAPCo has employed the CE HJTC design for RVWL, the associated CE0G technical specifications 12 have not been followed with regard to an Action statement for less than the Total No. of Channels operable.

3. The Action statement for less than the Minimum Channels Operable" deviates from the Generic letter 83-37 after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months by requiring a Special Report to the Commission within 30 days 17

[

. versus be in at least Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and CYAPCo does not require initiation of an alternate method of monitor.ing, ,

the reactor vessel inventory which is a deviation from the NRC Staff Safety Evaluation Il approving the CE0G technical specifications for reactors using the CE HJTC RVWL design.

Although this is a deviation from the guidance, it is within the bounds of Actions previously accepted by'the NRC Staff7 and is, therefore, judged acceptable.

9

4. In Table 3.23-2, CYAPCo specifies by a footnote that Channel Calibration of the RVWL is " electronic calibration from the ICC cabinets only." This is a deviation from the Generic Letter 83-37 which.has no such footnote. As the CE HJTC design employs thermocouple which themselves are not calibrated but if inoperable are usually just replaced during refueling outages, only electronic calibration from the Inadequate Core Cooling

.(ICC) cabinets is judged acceptable.

Reactor Coolant System Subcooling Margin Monitor (SMM):

1. CYAPCo provides no Action statement for less than the Total No.

of Channels ' operable. The Generic Letter 83-37 guidance is to restore the inoperable channel (s) to operable status within 7 s days or be in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. Rather than shutdown after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months with less than the Minimum Channels Operable, CYAPCo proposes to determine the subcooling margin once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Although determining the subcooling ,

margin from say hot let temperature and reactor vessel pressure is a reasonable alternative, there should be a time limit on the length of such an alternative. Otherwise, the SMM could remain inoperable indefinitely. It is recommended that CYAPCo consider an additional Action such as " Restore the system to Operable j status at the next scheduled refueling," as was done for the :

Reactor Vessel Water Level system.

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.c As-a result of~ the review of the cited material, the Licensee's 4 c. response for Technical Specifications for Item II.F.2, inadequate core cooling instrumentation is judged not to meet the guidance of the Generic Latter.

2.11 Control Room Habitability Requirements (III.D.3.41' ISe Generic LetterI contains the following statement:

" Licensees should assure that control room 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' relea'se 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 LC0 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 lJbitability 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."

Evaluation 5

The Licensee responded in letter dated July 1, 1988 stating that no specifications are proposed for this item because the control room habitability is being evaluated in the Haddam Neck Plant Integrated Safety Assessment Program. The NRC Staff is in agreement with this position.

As a result of reviewing the cited material, the NRC Staff is evaluating the Technical Specifications for Item 111.0.3.4, control room habitability, under the Haddam Neck Plant Integrated Safety Assessment Program.

19

3. ADDITIONAL INF6RMATION REQUIRED TO COMPLETE THIS REVIEW In Section 2, " Discussion and Evaluation," it is shown that for compliance with the Generic Le,tter,I additional information from or action by the Licensee is required for some items. Following is a compilation of the noncompliance _ items with a description for each of the needed information or action. As an alternative to revising the -

specifications as detailed below, the Licensee may provide site specific technical information and or safety analysis to justify why the revisions are not required for the protection of the public health and safety.

~

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

1. Submit a Technical. Specification for LCO 3.3.5.1 Actions a. and

b. that requires restoring at least two vent paths to Operable status in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or to b i Hot Shutdown Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2. Submit a technical specification for 3.3.5.1 that deletes the asterisk footnote in page 3-4q.

3.2 Lona Term Auxiliary Feedwater System Evaluation (II.E.1.1)

1. Specify mode applicability for the auxiliary feedwater system.

2. Specify the number of auxiliary feedwater pumps in an operability statement.

3. Specify the auxiliary feedwater system associated flow path operability.

4. Specify action statements for one, two, or three auxiliary feedwater pumps.

5. Specify " staggered" testing for the 31 day surveillance on the auxiliary feedwater pumps.

1 20

V . .

6 .' Specify an auxiliary feedwater pump flowrate as well as the

'specified discharge pressure of 800 psig in the 31 day cneck or

(

! submit justification for testing using minimum recirculation.

l '. If testing is to be during minimum recirculation that should be stated as part of the surveillance specification.

7. Specify a 31 day check that each automatic valve in the flow path is in the fully open position in automatic control or above 10% rated thermal power.

8. Specify a dedicated individual near any manually realigned valves during surveillance testing when only one auxiliary feedwater train-is available.

4

9. Specify testing to demonstrate flowpath operability prior to startup after any refueling outage or other cold shutdown of longer than 30 days.

3.3 Noble Gas Effluent Monitors (II.F.1.1)

1. Submit a Technical Specification revision for Table 3.23-1 that identifies multiple noble gas monitors for several locations, as recommended in Generic Letter. -

2. Submit a Technical Specification revision for Table 3.23-1 that provides alarm / trip setpoint and measurement range for the noble gas monitors as recommended in the Generic Letter.

i

3. Submit a Technical Specification revision for Section 3.23 for {

the noble gas monitors in the Action for when the alarm / trip 1 setpoint is exceeded, as recommended in the guidance.

4. Submit a Technical Specification revision for Section 3.23 that deletes the triple asterisk footnote in Table 3.23-1 allowing isolation of the stack wide range noble gas monitor during periods of high steam generator blowdown.

21

m

5. Submit a Technical specification revision for Section 3.23 for

' the noble gas monitors in the Action that requires initiating, a , ,

preplanned alternate method of noble gas monitoring within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the number of operable channels.is less than the minimum channels operable requirement.

6. Submit a Technical Specification revision for Section 3.23 for Table 3.23-2 specifying a monthly functional test for the main stack wide range noble gas monitor.

3.4 Containment Hiah-Ranae Radiation Monitor (II.F.1.3)

1. Submit _ a revision to Table 3.23-1 that specifies alarm / trip ff setpointandmeasurementrangeforthehigh-rangeradiation[

monitors.

2. Specify an Action when the alarm / trip setpoint is exceeded, as in the guidance.

3. Specify a Surveillance Requirement for a functional test at least once per month, as in'the guidance.

~

4. Capitalize the word " operable" in two places in Action 7, P 3-46b.

3.5 Containment Water Level Monitor (II.F.1.5) j

1. Provide technical specifications for the containment water level narrow range instrument.

3.6 Instrumentation for Detection of Inadeauate Core Coolina (II.F.2)

1. Submit a revised technical specification that specifies 4 CETs/ quadrant in Table 3.23-1 rather than 16/ core and a footnote to clarify that there are only 3/ quadrant in Quadrant IV.

22 I

___1_ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _

. r ,.

2. Submit a revised technical specification that' deletes the footnote stating that " Quadrant III may have only one channel Operabl e" .

l

3. Submit a revised technical . specification for Table 3.23-1 that provides for an action statement for less than the " minimum channels operable" of core exit thermocouple.

4. Submit a revised technical specification for Table 3.23-1 that provides for an action statement for less than the Total No. of-Channels operable for_ the reactor vessel water level.

5. Submit a revised technical specification for Table 3.23-1.that provides for an action statement for less than the Total No. of Channels operable for the reactor coolant system subcooling margin monitor.

6. Submit a revised technical specification for Table 3.23-1 that-provides in the action statement for less than the Minimum Channels Operable that the inoperable reactor coolant system subcooling margin monitor will be restored to Operable status at the next scheduled refueling, or submit a revised action

- statement that limits the time period with less than the Minimum Channels Operable.

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74.

SUMMARY

The.following subsections describe those issues addressed by.the Licensee that are. considered to be in compliance with the guidance in Generic Letter 83-37:

o' 'Postaccident Sampling (II.B.3)

~

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

o Containmerit Pressure Monitor' (II.F.1.4) .

The following subsections describe those issues addressed by the

' Licensee that are considered to be not in compliance.with the guidance in Gener,ic Letter 83-37: '

o Reactor. Coolant' System Vents (II.B.1) o Long Term Auxiliary Feedwater System Evaluation (II.E.1.1)'

.o Noble. Gas Effluent. Monitors (II.F.1.1)'

o ' Containment High Range Radiation Monitor (II.F.1.3) o ' Containment Water Level Monitor (II.F.1.5) o Instrumentation for Detection of Inadequate: Core Cooling (II.F.2.). .

The Licenseell has not submitted specifications for the following .

items pending separate review by the NRC Staff under the Haddam Neck Plant Integrated Safety Assessment Program.

o Containment Hydrogen Monitor (II.F.1.6) o Control Room Habitability Requirements (III.D.3.4). )

24 J

(

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{

s 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. W. G. Counsil letter to D. G. Eisenhut, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit No. 2, NUREG-0737 Technical Specifications (Generic Letter 83-37).," Northeast Utilities, July.25, 1984.

4. NUREG-0452, Rev. 4, Standard Technical Specifications for Westinghouse Pressurizr> Water Reactors, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Fall 1981.

5. E. J. Mrotzka letter to N'RC, "Haddam Neck Plant Proposed Changes to Technical Specifications Generic Letter 83-37-NUREG-0737 (TAC Nos.

64538,54393)," Connecticut Yankee Atomic Power Company, July 1, 1988.

6. D. M. Crutchfield letter to W. G. Counsil, "NUREG-0737, Item II.B.1, Reactor Coolant System Vents," Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, September 6,1983.

7. J. F. Stolz letter to E. J. Mroczka, " Issuance of Amendment (TAC #s 54546 and 54399)," Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, September 28, 1987.

8. J. A. Zwolinski letter to J. F. Opeka, " Containment Pressure and Water Level Monitors-TMI Action Items II.F.1.4, and II.F.1.5," Office of Nuclear Rector Regulation, U.S. Nuclear Regulatory Commission, July 9, 1985.

25

e .. .

F. M. Akstulewicz letter to J. F. Opeka, "TMI Action Plan Item II.F.2 9.

Inadequate Core Cooling Instrumentation," Office of Nuclear Reactor . . .

Regulation, U.S. Nuclear Regulatory Commission, December 12, 1985.

10. W. G. Counsil letter to J. A. Zwolinski, "Haddam Neck Plant Additional Information in Response to Generic Letter 82-28," -

Connecticut Yankee Atomic Power Company, February 1, 1985.

11. 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," Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, October 28, 1986.

12. R. W. Wells letter to H. L. Thompson, " Technical Specification for the Reactor Vessel Level Monitoring System," Northeast Utilities, February 19, 1985.

26 i l l 1

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, w a ~osu.1,,a a aave.us~a l Technical Evaluation Report for Haddam Neck Plant Response to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation's Generic Letter No. 83-37

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Fina81lPomiTechnical Division of Licensing Office of Nuclear Reactor Regulation Evaluation Report U.S. Nuclear Regulatory Commission * "a'a cov a no<= - ~ ~<

Washington, DC 20555 12 SUPPLlothT ARv moits

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Technical Evaluation Report on the Haddam Neck Plant Technical Specifications conformance to the requirements of the NRR Generic Letter No. 83-37, "NUREG-0737 Technical Specifications".

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SUMMARY

SUMMARY

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