ML20215J116

Technical Evaluation Rept of Crystal River Nuclear Generating Station Unit 3 Response to NRR Generic Ltr 87-37, Interim Rept

ML20215J116
Person / Time
Site:	Crystal River
Issue date:	02/28/1987
From:	Mobley E EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20215J098	List: ML20215J095 ML20215J116
References
CON-FIN-D-6022 EGG-NTA-7443, GL-87-37, NUDOCS 8705070231
Download: ML20215J116 (26)

Text

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EGG-NTA-7443 November 1986 INFORMAL REPORT i

i idaho at/Onal TECHNICAL EVALUATION REPORT OF CRYSTAL RIVER Eng/ nee /ng NUCLEAR GENERATING STATION UNIT 3 Laboratory RESPONSE TO NRR GENERIC LETTER 83-37 Managed by the U S.

Department E. V. Mobley l of Enorgy l 4

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d N8"** Prepared for the l

w ., ro,,,,,,,o,,,, U. S. NUCLEAR REGULATORY COMMISSION Dor contrar No Of Alvf MdQtStu (1705070231 070223 p*DR ADOCK 05000:102 p f*DH

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DISCLAIMER This book was prepared as an account of wort sponsored by an semcy of the Uruted States Govenment. Neitner the United States Govenment not any agency thereof, not any of their empioyees, mates any warranty, empress or emphed, or auumee any leget habihty or responschty for the accuracy, completenees, or usefulnees of any informat on, apparatus, product of process disclosed, or recreeents that its use would not ininnge pervately owned t'gnts Po'erences herein to any specifw commercial product, process et senrice by trade name. trademark, manufacturer, or otherw+ee, does not necessarily constitute or ireply its endorsement, recommendate, or favonng by the United States Govenment or any agency inereof. The viene and opensons of authors espressed eerein do not necessarily state or reflect those of the Urw ed States Govenment of any agency thereof i

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i EGG-NTA-7443 l

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! . TECHNICAL EVALUATION REPORT OF CRYSTAL RIVER UNIT 3 j RESPONSE TO THE U.S. NUCf. EAR REGULATORY COMMISSION, l 0FFICE OF NUCLEAR REACTOR REGULATION'S GENERIC LETTER N0. 83-37 ,

Docket No. 50-302 E. V. Mobley I

k Pub 11shed February 1987 Idaho National Engineering Laboratory l EG6G Idaho, Inc.

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. Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 '

under 00E Contract No. DE-AC07-761001570 FIN No. 06022 1

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t CONTENTS A8STRACT .............................................................. 111 FOREWORD .............................................................. 111

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

) 2. DISCUSSION AND EVALUATION ........................................ 2 2.1 Reactor Coolant System Vents (11.8.1) ...................... 2

] 2.2 Postaccident Sampling (11.8.3) ............................. 3 i 2.3 Long Term Aux 111ary Feedwater System j Evaluation (II.E.1.1) ...................................... 4 j 2.4 Noble Ga s E f f luent Moni tor s ( II .F .1.1 ) . . . . . . . . . . . . . . . . . . . . . 5  !

2.5 Sampilng and Analysts of Plant Effluents (II.F.1.2) ........ 6 1

, 2.6 Containment High-Range Radiation Monitor (II.F.1.3) ........ 7 i

j 2.7 Containment Pressure Monitor (II.F.1.4) .................... 7 l

2.8 Containment Water Level Monitor (II.F.1.5) ................. 8 f

3 2.9 Containment Hydrogen Monitor (II.F.1.6) .................... 9 i

2.10 Instrumentation for Detection of Inadequate

! Core Cooling (II.F.2) ...................................... 10 1

2.11 Control Room Habitability Requirements (III.O.3.4) ......... 11 1

3 3. A00!i!0NAL INFORMATION REQUIRED NEEDED TO COMPLETE THE REVIEW .... 16 ,

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SUMMARY

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l i S. REFERENCES ....................................................... 19  ;

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l A8STRACT l

This EG&G Idaho, Inc., report evaluates the submittals provided by Florida Power Corporation for Crystal River Unit 3. The submittals are in response to Generic Letter No. 83-37, "NUREG-0737 Technical Specifications

• (TS)". Applicable sections of the Technical Specifications are evaluated i

to determine compliance to the guidelines established in the Generic Letter.

1 FOREWORD This report is supplied as part of the " Technical Assistance for 1

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.

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i Docket No. 50-302 TAC Nos. 54527 and 57602 l

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-i TECHNICAL EVALUATION REPORT l i

CRYSTAL RIVER UNIT 3 1

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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 NRC Staff guidance on the contents of the Technical Specifications (TS) associated with certain items in i 1

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NUREG-0737.2 The responses to Generic Letter 83-37 filed to date by the Florida Power Corporation for the Crystal River Unit 3 include i' (a) Technical Specification Change Request (TSCR) No. 36 dated >

January 17, 1983, with supplements on November 1, 1983,  ;

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December 16, 1983, and March 22, 1984, (b) TSCR No. 82 dated I

June 22, 1983,7 with supplements on February 24, 1984, and j December 31, 1984, (c) TSCR No. 120 dated May 31, 1984,10 (d) TSCR j No. 122 dated January 23, 1985,I with supplements on June 6, 1985, and

! June 28, 1985,1 ,13 and (e) TSCR No. 133 dated March 29, 1985.I The l following report provides the evaluation of the FPC submittals and j indicates the information and action required for resolving the renalning issues.

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2. DISCUSSION AND EVALUATION The Licensee was requested to provide Technical Specifications for several different systems. Each of these proposals is discussed and evaluated in an individual section below.

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

The Generic Letter 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 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 is provided in Enclosure 3. For the plants using a power operated relief valve (PORV) as a reactor coolant system vent, the block valve is not required to be closed if the PORV is operable."

Evaluation:

The Licensee has proposed the addition ~' of Technical Specification 8

3.4.11 and deletion of part of the model Technical l

i Specification 4.4.11. Exemption from the requirements of Technical i Specification 3.0.4 was also requested.10 i

This item is being reviewed by the NRC Staff and no further evaluation is being performed as part of this report.

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2.2 Postaccident Samplina (II.8.3) i

The Generic Letter) contains the following statement:

" Licensees should ensure that their plant has the capability to l

j, obtain and analyze reactor coolant and containment atmosphere i samples under accident conditions. An administrative program i should be established, implemented and maintained to ensure this

. capability. The program should include:

! a) training of personnel i

b) procedures for sampling and analysis, and

! c) provisions for maintenance of sampling and analysis equipment ,

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

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A model Technical Specification for postaccident sampling.is 1

4 provided that requires the capability to sample and analyze radioactive- -

! todines and particulates in plant gaseous effluents.

i i Evaluation:

i f The schedule attached to the Licensee letter dated

! November 20, 1984, shows two Technical Specification actions associated ,

1 l with the Generic Letter Item II.B.3. The first (in time) scheduled action is revision of the containment isolation valve Technical l 9 i Specification. The Licensee letter dated December 31, 1984, included proposed revisions to Technical Specification Table 3.6-1, Containment Isolation Valves. The revisions' to Table 3.6-1 are appropriately j associated with report NUREG-0737,2 but do not apply to the Generic l Letter. Because no guidance is furnished in the Generic Letter for t containment isolation valves, the proposed Technical Specification 9

change was not reviewed. ,

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The other action shown for Item II.8.3 in the Licensee letter of November 20, 1984, is a proposed Technical Specification for sampling. To meet this scheduled commitment,15 the Licensee letter 14 dated March 29, 1985, proposed the addition of Technical Specification 14 Section 6.17. The additional specification proposed by the Licensee imposes requirements that are the same ai those in the model.

l 14 The Technical Specification proposed b the Licentee is judged to

' meettherequirementsoftheGenericLetter{forItemII.8.3.

i 2.3 Lono Term Auxiliary Feedwater System Evaluation (II.E.1.11' I The Generic Letter 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 l 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:

The SER transmitted to the Licensee on October 1, 1984," included a I

review of conformance to the Generic Letter of the TS pertaining to Item II.E.1.1. The TS were accepted on condition that a TS be added to require a flow test of the emergency feedwater system following extended cold shutdowns. The Licensee proposed changes, in letter dated .

January 23, 1985, in Technical Specifications Sections 3.3.2.1, '

Ergineered Safety Features Actuation System; 3.3.3.6, Postaccident Instrumentation, and 3.7.1.2, Emergency Feedwater System. The Licensee i letter dated January 23, 1985,11 included the additional TS required by the October 1, 1984, SER. A correction and change to the original proposal" were made in letters dated June 6, 1985,12 and l 4 i

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June 28, 1985.13 The three Licensee submittals H -13 were reviewed by the Staff and found acceptable. The Safety Evaluation is enclosed with the notification of issue of Amendment No. 78.

Item II.E.1.1 of the Generic Letter is closed.

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

The Generic Letter 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 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 identificatic,n 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."

A model Technical Specification for noble gas effluent monitors is provided that specifies monitor locations and measurement ranges.

Evaluation:

The Licensee proposed, in letter dated January 17, 1983, several changes in the Radiological Effluent Technical Specifications. The proposed changes include Section 3/4.3.3, Monitoring Instrumentation.

Changes to the January 17, 1983, proposal were made in supplements 5

dated November 1, 1983 December 16, 1983, and March 22, 1984. These four submittals have been reviewed by the NRC Staff and found acceptable for' Item II.F.1.1 of the Generic Letter.1 This item was previously approved by-the NRC Staff as meeting the ,

requirements.

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

The Generic Letter contains the following statement:

4 "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 i during and following an accident. An administrative program.

should be established, implemented and maintained to ensure this capability. The program should include:

i 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 4

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."

A model Technical Specification for postaccident sampling is provided that requires the capability to sample and analyze radioactive lodines and particulates in plant gaseous effluents.

Evaluation:

The Licensee letterI

• dated March 29, 1985, proposed the addition of

• Technical Specification Section 6.17. The additional Specification proposed by the Licensee imposes requirements that are the same as i

those in the model.

The Technical Specification proposed by the Licensee is judged to 1

! meet the requirements of the Generic Letter for Item II.F.1.2.

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

I The Generic Letter contains the following statement:

"A minimum of two in containment radiation-level monitors with a

, maximum range of 108 rad /hr (107 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:

The Licensee proposed, in letter dated January 17, 1983, several changes in the Radiological Effluent Technical Specifications. The proposed changes include Section 3/4.3.3, Monitoring Instrumentation.

Changes to the January 17, 1983, proposal were made in supplements -

dated November 1, 1983, December 16, 1983, and March 22, 1984. These four submittals - have been reviewed by the NRC Staff and found acceptable for Item II.F.1.3 of the Generic Letter.1 This item was previously approved by the NRC Staff as meeting the requirements.

2.7 Containment pressure Monitor (II.F.1.4) i The Generic Letter contains the following statement:

" 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 7

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 LCO and -

surveillance requirements for accident monitoring instrumentation

- are included in Enclosure 3."

Evaluation:

The Licensee proposed, in letter dated January 17, 1983, several

changes in the Radiological Effluent Technical Specifications. The proposed changes include Section 3/4.3.3, Monitoring Instrumentation.

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Changes to the January 17, 1983, proposal were made in supplements dated November 1, 1983, December 16, 1983, and March 22, 1984. These four submittals - have been reviewed by the NRC Staff and found acceptable for Item II.F.1.4 of the Generic Letter.1 This item was previously approved by the NRC Staff as meeting the requirements.

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

The Generic Letter 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 i from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or less if justified) capacity. ,

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" 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 wioe range monitors should be similar to other accident monitoring instrumentation included in the present ,

Technical Specifications. LCOs 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

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brought to Hot Shutdown condition as required for other accident monitoring instrumentatica. Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3."

Evaluation

The Licensee proposed, in letter dated January 17,l1983, several changes in the Radiological Effluent Technical Specifications. The proposed changes include Section 3/4.3.3, Monitoring Instrumentation.

3 ~ 4-6 Changes to the January 17, 1983, proposal were made'in supplements dated November 1, 1983, December 16, 1983, and March 22, 1984. .These four submittals have been reviewed by the NRC Staff and found acceptable for Item II.F.1.5 of the Generic Letter.

This item was previously approved by the NRC Staff as meeting the requirements.

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I 2.9 Containment Hydroaen Monitor (II.F.1.6)

I The Generic Letter contains the following statement:

"Two independent containment hydrogen monitors should be operable at all times when the reactor is operating in Power 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 i

! , the plant should be brought to at least a hot standby condition l 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 j

, least one monitor should be restored to operable status within l

- 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 '

l 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."

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i Evaluation:

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i The Licensee proposed changes in Technical Specification  :

Section 3.6.4.1 in the letter dated February 24, 1984. The NRC 17 letter dated October 1, 1984, accepted the Technical Specification

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proposed in the Licensee's letter. However, FPC was requested to

! commit to providing test connections that would allow channel calibration -

on a more frequent basis, as intended by the Generic Letter.1 The Licensee's letter dated January 31, 1985, committed to installing test i . .

connections to allow more frequent testing of the Hydrogen Monitoring System by December 31, 1985. The FPC should commit to a date for appropriate Technical Specifications for channel calibration using sample gas, as indicated in the Generic Letter.1 The presently approved j Technical Specifications for this item were intended to be interim-Specifications until installation of the test connections. See also j Section 3 of this report.

l Based upon the review of the references cited, Item II.F.1.6 is considered incomplete, pending Technical Specification changes to allow for channel calibration using sample gas.

1 2.10 Instrumentation for Detection of Inadeauate Core Coolina (II.F.2) 4 I

The Generic Letter contains the following statement:

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"Subcooling margin monitors, core exit thermocouples, and a reactor coolant inventory tracking sytem (e.g., differential pressure measurement system designed by Westinghouse, Heated Junction Thermocouple System designed by Combustion j

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.

l "Subcooling margin monitors ~should have already been included in the present Technical Specifications. Technical Specifications i for core exit thermocouples and the reactor coolant inventory tracking system should be included with other accident monitoring i instrumentation in the present Technical Specifications. Four

! core-exit thermocouples in each core quadrant and two channels in i

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the reactor coolant tracking system are required to be operable 4

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

, Evaluation:

In response to the NRC request for proposed Technical Specifications for instrumentation for detecting inadequate core cooling, the Licensee committed to propose Technical Specifications by Refuel V. Subsequently, after installing th'e upgraded instrumentation, the Licensee stated that "FPC does not see a need for Technical I

Specifications for this instrumentation." The NRC in the Generic Letter has established that Technical Specifications are required for this instrumentation. The FPC has not provided any justification for not complying with the Generic Letter based on plant specific features of j Crystal River Unit 3. See also Section 3 of this report.

i The addition of Technical Specifications for the instrumentation for f

detection of inadequate core cooling is required for compliance with the Generic Letter.

2.11 Control Room Habitability Reouirements (III.D.3.4)

I The Generic Letter contains the following statement:

i " 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 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 1 detection system (e.g. chlorine detection system) are provided in 4

Enclosure 3. All detection systems should be included in the

Technical Specifications.

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"In addition to the above requirements, other aspects of the control room habitability requirements should be included in the l Technical Specifications for the control room emergency air l 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:

The Licensee proposed the addition of Technical Specification 3.3.3.8 on the Toxic Gas Detection System.

The proposed Technical Specification on the toxic gas part of Item III.D.3.4 of the Generic Letter was reviewed by the NRC Staff and found acceptable.

The current control room emergency ventilation system Technical Specification differences from the Generic Letter and the requirements for compliance are detailed below.

2.11.1 The Licensee initial LC0 statement 0 (TS 3.7.7.1, P 3/4 7-20) adds a footnote that is not in the Generic Letter. The footnote in effect requires recirculation mode upon loss of return duct radiation monitoring.

The footnote on operability is a plant specific equipment-related provision and is judged acceptable.

2.11.2 The Licensee TS 0 (page 3/4 7-20) does not include modes 5 and 6 and associated Action as shown in the Generic Letter.

Justification for the omission of modes 5 and 6 under Applicability and Action is required for compliance with the Generic Letter. ,

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2.11.3 The Licensee Surveillance Requirement 4.7.7.la (page 3/4 7-20) requires... ventilation system shall be demonstrated OPERABLE...by verifying 4

that the control room air temperature is 5 20*F. 1 The value of $120*F is consistent with the B&W Standard TS, but it is unlikely that most Operators would consider a control room approaching 120*F to be habitable.

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The Generic Letter requires a more reasonable value of 180*F.

Justification for use of a control room temperature above 80*F to verify ventilation system operability is required for compliance with the l Generic Letter.1 Citation of the Standard TS will not be considered I adequate justification.

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2.11.4 The Licensee Surveillance Requirement 4.7.7.lb (page 3/4 7-20)

I requires... verifying that the system operates for at least 15 minutes, to .

1 i demonstrate ventilation system operability. The Generic Letter is more

] stringent in that it requires... verifying that the system operates for at l 1 east 10 continuous hours with the heaters operating. Ten hours per month I with heaters is the operation recommended in Regulatory Guide 1.52 (cited in the TS) to reduce moisture buildup.

Justification of a system operability demonstration of less than

.I j 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and without heaters on is required for compliance with the Generic-4 l

Letter.1 J

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2.11.5 The Licensee Surveillance Requirement 4.7.7.1c (page 3/4 -20) reads. ..once per 18 months or, (a) af ter. . . maintenance. . .or, (b) following painting...This is consistent with the Generic Letter and the Standard TS, however, Regulatory Guide 1.52 (see further discussion below) uses and, not o_r_, in the corresponding section.

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The use of "and" is preferable to "or" because the most common

interpretation of "or" would negate the 18-months testing interval. An I

increase in the 18-month test interval by reason of another test that is not time-related is not logical.

Wording that restricts the testing interval of the ventilation system to no more than 18 months is required for compliance with the intent of the ,

Generic Letter.1 2.11.6 The Licensee Surveillance Requirements 4.7.7.lc.2, f and g (pages 3/4 7-21 and 22) are keyed to a footnote that allows testing of air flow distribution downstream of the HEPA filters. In the updated test standard, the measurement of velocity is preferably made downstream j from the filters.

The Licensee deviation from the Generic Letter on the air' flow distribution test is judged acceptable.

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! 2.11.7 The Licensee Surveillance Requirements 4.7.7.lc.3 and 4.7.7.1d refer in four places to Regulatory Guide 1.52, Revision 1, 1976. The i Generic Letter refers to Revision 2, March 1978. The discussion in this report is based on Revision 2.

1 A change in the Technical Specifications to call out Revision 2 or the basis for not using Revision 2 of Regulatory Guide 1.52 is required for compliance with the Generic Letter, i

0 2.11.8 The Licensee Surveillance Requirements 4.7.7.le (page 3/4 7-21) does not include the control room positive pressure and heater dissipation tests as shown in the Generic Letter model, 4.7.7.e.3 and 4.

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I heater dissipation tests is required for compliance with the Generic Letter.

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0 2.11.9 The Licensee Surveillance Requirement 4.7.7.lf (page 3/4 7-22) requires 99% removal for the HEPA filter DOP leak test. The Generic Letter allows a value this low if an efficiency of 90% is assumed in the safety analysis. If filter efficiency of 99% was assumed in the safety analysis (See also Regulatory Guide 1.52), then the required leak test removal is 99.95%.

Citation of the FSAR section showing that 90% HEPA filter efficiency was assumed in the safety analysis would prove acceptability of Surveillance Requirement 4.7.7.lc. If 90% filter efficiency was not assumed in the safety analysis, justification of the HEPA filter leak test 99% removal value is required for compliance with the Generic Letter.

2.11.10 The Licensee Surveillance Requirement 4.7.7.1g (page 3/4 7-22) requires 99% removal for the halogenated hydrocarbon leak test of the charcoal adsorbers. The Generic Letter requires equal to or less than 0.05% leakage, corresponding to 99.95% removal.

Justification of a charcoal adsorber leakage greater than 0.05% is required for compliance with the Generic Letter.

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3. ADDITIONAL INFORMATION NEEDED TO COMPLETE THE REVIEW In Section 2, Discussion and Evaluation, it is shown that for compliance with the Generic Letter, additional information from or action by the Licensee is required for some items. Following is a compilation of the needed information or action:

1. Containment Hydrogen Monitor (II.F.1.6): See also Section 2.9.

i Provide a proposed Technical Specification for channel calibration that is equivalent to the model Technical Specification.

2. Instrumentation for Detection of-Inadequate Core Cooling (II.F.2): See also Section 2.10. Provide the proposed Technical Specifications changes that add ICC instrumentation, as shown in the model Technical Specifications.

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3. Control Room Habitability Requirements (III.D.3.4): See also Section 2.11. 1

a. See also Paragraph 2.11.2. Provide a change request to the i 1  :

1 Technical Specifications to conform to the Generic Letter or provide justification for the omission of modes 5 and 6 under Applicability and Action, Limiting Condition for 20 j Operation 3.7.7.1.

I i

b. See also Paragraph 2.11.3. Provide a change request to the 1

Technical Specifications to conform to the Generic Letter

, or provide justification for air temperature greater than 80*F, Surveillance Requirement (SR) 4.7.7.la, ,

I

c. See also Paragraph 2.11.4. Provide a change request to the .

1 Technical Specification to conform to the Generic Letter l or provide justification for the system test being less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with heaters on, SR 4.7.7.lb.

l l'

16 l

d. See also Paragraph 2.11.5. Provide wording clarification such that maintenance-related tests will not be construed to negate the 18-month time interval for testing, SR 4.7.7.lc.

. e. See also Paragraph 2.11.7. Provide a change request to the Technical Specification to call out Revision 2 or provide justification for using Revision 1 instead of Revision 2 of Regulatory Guide 1.52, SR 4.7.7.lc.3 and 4.7.7.ld.

f. See also Paragraph 2.11.8. Provide a change request to the Technical Specifications to conform to the Generic Letter or provide justification for the omission of the control room positive pressure and heater dissipation tests, SR 4.7.7.le,

g. See also Paragraph 2.11.9. Provide one of the following: A FSAR reference showing that HEPA filter efficiency of 90%

(or less) was assumed in the accident calculations; or, justification for the 99% removal in the leakage test, SR 4.7.7.lf.

h. See also Paragraph 2.11.10. Provide justification for the 99% removal in the leakage test, SR 4.7.7.1g.

17

i

4.

SUMMARY

The following are those items considered to be consistent with the Generic Letter:

1. Postaccident. Sampling (II.B.3)

I .

-2. Long Term Auxiliary Feedwater System Evaluation (II.E.1.1)

< 3. Noble Gas Effluent Monitors (II.F.1.1)

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

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

6. Containment Pressure Monitor (II.F.1.4)

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

The following are those items considered to be out of compliance with the Generic Letter:

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

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

3. Control Room Habitability Requirements (111.0.3.4).

Item II.B.1, Reactor Coolant Vents, is being reviewed by the NRC Staff.

I I

18

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 Reautrements, published i

by the Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, November 1980.

. 3. G.R. Westafer letter to Harold R. Denton, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 36, Radiological Effluent Technical Specifications," Florida Power Corporation, January 17, 1983.

1 4. G.R. Westafer letter to Harold R. Denton, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, Supplement 1 to Technical Specification Change Request No. 36, Radiological Effluent

Technical Specifications," florida Power Corporation, November 1, 1983.

5. G.R. Westafer letter to Harold R. Denton, " Crystal River Unit 3,

Docket No. 50-302, Operating License No. DPR-72, Supplement 2 to

Technical Specification Change Request No. 36, Radiological Effluent

Technical Specifications," Florida Power Corporation, December 16, i

1983.

i 6. G.R. Westafer letter to Harold R. Denton, " Crystal River Unit 3 Docket No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 36, Supplement 3, Radiological Effluent Technical Specifications," Florida Power Corporation, March 22, 1984.

7. G.R. Westafer letter to H.R. Denton, " Crystal River Unit-3, Docket No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 82," Florida Power Corporation, June 22, 1983.

8. G.R. Westafer letter to H.R. Denton, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, Technical Specification l

, Change Request No. 82, Supplement to Attachment A and C," Florida j Power Corporation, February 24, 1984.

9. G.R. Westafer letter to H.R. Denton, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 82, Supplement 3, Revision of NUREG-0737 Technical Specifications," Florida Power Corporation, December 31, 1984.

10. G.R. Westafer letter to H.R. Denton, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, Technical Specification i Change Request No. 120, Mode Change," Florida Power Corporatioa, May 31, 1984.

11. G.R. Westafer letter to H.R. Denton, " Crystal River Unit 3, Docket

No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 122," Florida Power Corporation, January 23, 1985.

19

I

12. G.R. Westafer letter to Ms. Brenda Mozafari, " Crystal River Unit 3, Docket No. 50-302, Operating License No OPR-72 Technical Specification Change Request No. 122," Florida Power Corporation, June 6, 1985.

13. G.R. Westafer letter to Ms. Brenda Mozafari, " Crystal River Unit 3 Docket No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 122, Supplement 2." Florida Power .

Corporation, June 28, 1985.

14. G.R. Westafer letter to H.R. Denton, " Crystal River Unit 3, Docket -

No. 50-302, Operating License No. DPR-72, Technical Specification Change Request No. 133," Florida Power Corporation, March 29, 1985.

15. G.R. Westafer letter to John F. Stolz, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, NUREG-0737, Supplement 1, Schedule of Technical Specification for NUREG-0737 Items," Florida Power Corporation, November 20, 1984.

16. Harley Silver letter to Walter S. Wilgus, " Amendment No. 78 to Facility Operating License No. DPR-72 for the Crystal River Unit 3 Nuclear Generating Plant," U.S. Nuclear Regulatory Comission, July 16, 1985.

17. Letter from NRC to W.S. W11gus, " Crystal River Unit 3 - Request for Revision to Proposed NUREG-0737 Technical Specifications," U.S.

Nuclear Regulatory Commission, October 1, 1984.

18. G.R. Westafer letter to John F. Stolz, " Crystal River Unit 3, Docket No. 50-302, Operating License No. OPR-72, NUREG-0737, Item II.F.1.6, Containment Hydrogen Monitor," Florida Power Corporation, January 31, 1985.

19. G.R. Westafer letter to John F. Stolz, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, NUREG-0737, Item II.F.2, Inadequate Core Cooling Instrumentation Implementation Letter Report,"

Florida Power Corporation, October 23, 1985.

20. Peter B. Erikson letter to J.A. Hancock, " Amendment No. 44 to facility Operating License No. DPR-72 for the Crystal River Unit 3 Nuclear Generating Plant," U.S. Nuclear Regulatory Commission, November 24, 1981.

21. NUREG-0103, Standard Technical Specification for Babcock and Wilcox  !

Pressurized Water Reactors, Rev. 4, U.S. Nuclear Regulatory ,

Commission, Fall 1980. '

22. ANSI /ASME N510-1980, Testino of Nuclear Air Cleaning Systems. -

l l

20 1

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RESPONSE TO NRR GENERIC LETTER 83-37 oo~r. .t..

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• *'* ca ca^a r ~w=*ta EG&G Idaho, Inc.

P. O. Box 1625 06022 Idaho Falls, ID 83415 1Ia ivetOPmtpont 10 SPON50 ming omGamitation mawa amo waisimo aoopti$ staciese t e Cases Division of Licensing Technical Evaluation Report Office of Nuclear Reactor Regulation " ""' *' "' " * " ' ~ ~ '

U.S. Nuclear Regulatory Commission Washington, DC 20555 13 lb>>LEMi%rAAv40rt5 a .. r.acr a ,

Interim Technical Evaluation Report on the audit of the Crystal River Nuclear Generating Station Unit 3 Technical Specifications performed for the NRC in connection with conformance to the requirements of the NRR Generic Letter No. 83-37, "NUREG-0737 Technical Specifications".

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