ML20238B221

Conformance to Reg Guide 1.97 Shoreham Nuclear Power Station,Unit 1

ML20238B221
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	03/31/1986
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20234B806	List: ML20234B803 ML20234B820 ML20238B221
References
CON-FIN-A-6493, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-7026, NUDOCS 8709090568
Download: ML20238B221 (19)

Text

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CONFORMANCE TO REGULATORY EUIDE 1.97 SHOREHAM NUCLEAR POWER STATION, UNIT NO. 1 l

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A. Udy )

R. VanderBeek l

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Published March 1986 l

EG&G Idaho, Inc.

Idaho falls, Idaho 83415 l

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Prepared for the l U.S. Nuclear Regulatory Commission Washington. 0.C. 20555

~~ Under'00E Contract No. DE-AC07-761001570 i 7" FIN No A6493  !

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l ABSTRACT This EG% Idaho, Inc., report reviews the submittal = for Regulatory l Guide 1.97, Revision 2, for the Shoreham Nuclear Power Station Unit No. 1. Any exceptions to these guidelines are evaluated and those areas wrere sufficient 08515 for acceptability is not provided are identified.

Docket No. 50-322 11

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f0 REWORD 1

This report is supplied as part of the " Program for Evaluating l Licensee / Applicant Conformance to RG 1.97," being conducted for the U.S.

Nuc' ear Regulatory Comission, Of fice of Nuclear Reactor Regulation, Division of IWR Licensing-A, by EG&G Idaho, Inc., NRR and I & E support Branch.

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The U.S. Nuclear Regulatory Comission funded the work under l authorization B&R 20-19 40-41-3.

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f Docket No. 50-3?2 iii l l i

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CONTENTS A8STRACT - .. ... ... . .. .... .... .............. ........... .

ij FOREWORD ....... . .... . ....... . . .. . .. ......I..., 111

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

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2. REVIEW REQUIREMENTS ...... . . . .... ........... . .... .... 2  ;

3. EVALUATION . .

.. ......... .... .... .. ........ ............. 4 3.1 Adherence to Regulatory Guide 1.97 . ... .. . . ........ .. 4 l 1

3.2 Type A Variables ......................... ................ 4 j f

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3.3 Exceptions to Regulatory Guide 1.97 . . ........... .... ... 5 l 4 CONCLUSIONS .....,.. ........... ............... ................. 13

5. REFERENCES .................... ........... ...................... 14 l

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CONFORMANCE TO REGULATORY GUIDE 1.97 SHOREHAM NUCLEAR POWER STATION, UNIT NO. I s

1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response caqability. These requirements have been published as Supplement.

l No. I to NUREG-0737, "TMI Acticn Plan Requirements" (Reference 3). l l The Long Island Lighting Company, the licensee for the Shoreham l

l Nuclear Power Station, provided .a response to the generic letter on '

April 14, 1983 (Referance 4). Additional information was provided on  !

l October 23, 1985 (Reference 5) and February 27, 1986 (Reference 6). This I l report provides an evaluation of these submittals.

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2. REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement No. 1, sets forth the '

documentation to be submitted in a report to the NRC describing how the licensee complies with Regulatory Guide 1,97 as applied to emergency response facilities. The submittal should include documentation that provides the fol!, wing information for each variable shown in the ]

applicable table of Regulatory Guide 1.97.

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1. Instrument range

• 2. Environmental qualification I

3. Seismic qualification

4. Quality assurance

5. Redundance and sensor location

6. Power supply l

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7. Location of display

8. Schedule of installation or upgrade The submittal should identify deviations from the regulatory guide and p ovide supporting justification or alternatives.

Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subject.

At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide, it was noted that no further staff review would be necessary. Therefore, 2

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this report only addresses exceptions to Regulatory Guide 1,97 The following evaluation is an audit of the licensee's submittals based on the .

review policy described in the NRC regional meetings. ,

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3. EVALUATION l l

The licensee provided a response to NRC Generic Letter 82-33 $h April 14, 1983. Additional information was provideo on October 23, 1985 and February 27, 1986. This evaluation is based on '.hese submittals.

3.1 Adherence to Requlatory Guide 1.97 The licensee stated, in Appendix C of Reference 4, that Shoreham is in conformance with Regulatory Guide 1.97 to the extent discussed in the submittal. Within Table I of Appendix C, the licensee has listed the Regulatory Guide 1.97 variable and its status. One of the licensing conditions for Shoreham is that all Regulatory Guide 1.97 modifications t identified are to be installed prior to the startup following the first refueling outage except as otherwise agreed to by the NRC. Therefore, we conclude that the licensee has provided an explicit conmitment to conform l to Regulatory Guide 1.97, except for those exceptions thet were justified as noted in Section 3.3.

l 3.2 Type A Variable Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide the information required to permit the control room operator to take specific manually controlled safety actions.

The licensee classifies the following instrumentation as Type A.

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1. Reactor pressure vessei water level l 2. Reactor pressure vessel pressure

3. Drywell pressure

4. Sump level l

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6. Drywell and suppression chamber oxygen cor.centrat10n

7. Suppression pooi sator temperature 6 The abovt variables meet the Category ? reconnendation! consistent witn the requirements for Type A variables.

3.3 Exceptiens to Requ_latory Guide 1.97 The licensee identified deviations and exceptions from Regulatory Guide 1.97. These are discusse6 in the following paragraphs. l 3.3.1 Neutron Flux ~

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Except1on has been taken by the licensee to the recommendations of Regulatory Guide 1.97 for the variable neutron flux. The licensee I 6

identified, in Reference 4, (a) a range of 16~I to 10 Ci instead of l the ret.ommended 10~ to 100 percent of full power. (b) the source range c.onitor (S9H) instrumentation does not comply with any of the recommendations of Regulatory Guide 1.97, and (c) a non-Clas'- lE power source instead of the Class lE power source recomniended by Regulatory Guide 1.97. The licensee states that the installed equipment is satisf actory for the interim period and that systeia modifications will be 1 l completed after the second refueling outagc.

In Reference 5, the licensee commits to a study of the neutron monitoring system. However, the Outlined requirements to satisfy the study do not meet the recommended Category I requirements, in the process of our review of the neutron flux instrumentation for boiling water reactors, we note that the detectors and their cables have not satisfied the environmental qualification requirement of Gegulatory Guide 1.97. A Category 1 system that meets all the criteria of Requ)atory Guide 1.97 is an industry development item. Based on our review, we conclude that the existing instrumentation is acceptable for in'serim 5

I operation. The licensee should follow industry development of this equipment, evaluate newly developed equipment, and install Catego(y 1 ,

instrumentation when it becomes available.

3.3.2 Reactor Water Level Reguiatory Guide 1.97 specifies that the level range should be from the bottom of the core support plate to the lesser of the top of the vessel or the centerline of the main steamline. The licensee provides a fuel zone

) range of -308 to -108 in. and a wide range of -150 to +60 in. This range is more than seven feet lower than the recommended upper level of 1

indication and, therefore, does not meet the recommendations of Regulatory l Guide 1.97. However, all safety trips based on high level occur within the provided range. The licensee states that further action to this system is (

pending the resolution of the Inadequate Core Cooling (ICC) detection issue.

The licensee de iates from Regulatory Guide 1.97 with respect to the range of this instrumentation. This deviation goes beyond the scope of this review and is being addressel by the NRC as part of their review of l

NUREG-0737, Item II.F.2 (Inadequate Core Cooling). The acceptance criteria '

for Item II.F.2 is the same as Category 1 for Regulatory Guide 1.97. .

3.3.3 Reactor Coolant System Soluble Boron Concentration

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Exception has been taken by the licensee to the recommendation of  !

riegulatory Guide 1.97 for this variable. The licensee specifies a range of 100 to 1100 ppm instead of the recomended range of 0 to 1000 ppm per Regulatory Guide 1.97. The licensee has not provided any justification for this deviation.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, item II.B.3.

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3.3.4 Orywell Sump level Dryweil p ain Sumps Level 6

The licensee states that compliance to the recommended Type B, Category 1, and Type C, Category 1, requirements is not applicable for

'ihese variables. The licensee states that the installed Category 3 equipment is acceptable as is. The supporting justification is that (a) the drywell pres;ure and temperature along with the primary containment area adiation can be used to provide indication of leakage in the drywell, (b) these verlable! are qualified to Category 1 or 2, and (c) the drywell sump systems are isolated for accident conditions. l We corclude that th9 in:trumentation supplied by the ,1censee will j provide the appropriate monitoring for the parameters of concern. This is based on (a) for small leaks, the instrumer,tation is not expected to experience harsh environments during operation, (b) for larger leaks, the I sumps fill promptly and the sump drain lines isola +.e due to the increase in {

drywell pressure, thus neoating the drywell samp level and drywell drain j sumps level instrumentation, (c) the drywell pressure and temperature as j well as the primary containment area radiation instrumentation can be used to detect leakage in the drywell, and (d) this instrumentation neithar automatically initiates nor alerts the operator to initiate the operation l of a safety-related system in a post-accident situation. Therefore, we find the instrumentation provided acceptable. l 3.3.5 Primary Containment Isolation Valve position The licensee states, Reference 4 that the re:ommerdation of Regulatory Guide 1.97 fer the prh'ary conta1, ment isolation valve position range is not applicable. The recommendation of the regulatory guide is a closed-not closed indication. The licensee states, in Reference S. that this is met by red open and green closed position indicating lights. Both lights are illuminated when the valve is in an ir. termed'. ate position. He find this position indication acceptable.

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3.3.6 Radiation level in Circulating Primary Coolant J

Thelicenseestatesthatalternateinstrumentatiot.1savaila$lefor this variable and that the critical actions to be taken to r event and mitigate a gross breach of fuel cladding are (c) shut down the reactor, and (b) maintain the water level. Neither of these a(.tions are influenced by l this variable.

I Tne licensee indic8tes that the post-accident sampling system (PASS),

which is beinc reviewed by the NRC as part of their review of NUREG-0737, l Item II.B.3, provides a means of obtaihirg sarnples of the reactor ceolant and that '.he primary containment atmosphere trossute and the radiation j monitors in the steam jet air ejector Jod the nain steamlines provide information on the status of fuel cladding when the plant is eat isolated.

Based on the alternate instrumentation provided by the licensee, we

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conclude that the instrumentation supplied for thiis variable is adequate i and, therefore, acceptable, f 3.3.7 R_adiation Exposure Rate i i

l l The licensee has elected not to implement tMs Type C variable as recorrrnenced by Revision 2 of Regulatory Guide 1.97, the justification being l

that other means, such as noble gas monitoring, are better suited fe" breach detection.

i Revision 3 of Regulatory Guide 1.97 (Referente 7) deletes this Type C variable f rom the rec 6mtended ins trumentation. Therefore, tne lack of the l

Type C instruirentation is acceptat;le. This variable is still rucomended

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as a Type E variable Refer to Section 3.3.13.

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3.3.8 Suppression Chamber heray Flew Drvwell Sarav i low I

The residual heat removal (RHR) system flow is uses for these variables. Both sprays derive their fiow from the RHR system, with a 1

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throttling valve proportioning the flow between the two sprays. The position of the throttling valve is controlied from the control room.

Pressure and temperature changes in the drywell and suppression cy mber determine the effectiveness of the sprays.

The licensee concludes trat the RHR flow and the suppression chamber and drywell temperature and pressure, accurately and reliably measure the effectiveness of the drywell and suppression chamDer sprays. Additionally, the position of the throttling valve is known in the control room. We find that this instrumentation is adequata for tris variable.

3.3.9 Core Spray System Flow J

The licensee d'd not submit the range of the core spray system flow ins trumen ta tion in Ref erer,ce 4. Reference 5 states the range of this instr umentation as being 0 to 7000 gpm. This is 151 percent of the design e

flow of the core spray pumps. This range treets the regulatory guide reconrnendations, and is acceptable, j 3.3.10 Residual Heat Removal (WHR) Heat Exchanger Outlet Temperature The 11cer.see has provided a range of 40 to 400*F for this variable.

Revision 2 of Regulatory Guide 1.97 recommends 32 to 350*F.

Revision 3 of Regulatory Guide 1.97 changes the reconinended range to 40 to V0*F. Therefore, the range provided by the licensee is acceptable.

3.3.11 Ceolina Water Temperature to Engineered Safety Features (ESF)

S 1 stem Components The licensee has provided a range of 40 to 320*F for this variable.

Wevision 2 of Regulatory Guide 1.97 recommends 32 to 200*F.

Revision 3 of Regulatory Guioe 1.9) changes the recorrnended range to 40 to 200*F. Therefore, the range provided by the licensee is acceptable.

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3.3.12 Secondary Containment Area Radiation Regulatory Guide 1.97 recomends instrumentation for this va table with a range of 10 to 10 R/hr. The licensee has 14 area monitors with a range of 0.1 to 10 mR/hr, 4 with a range of 1 to 10 mR/hr and one with a range af 10' to 10 R/hr.

The licensee states that the instrumentation for this variable is not l needed, as the plant noble gas effluent monitors (which are Cateoory 2 j instrumentation) are more useful and practical in detecting or assessing piping containment leakage. The licensee reports that the use of local ]

radiation exposure rate monitors to detect breach or leakage through I l primary containment penetrations results in ambiguous indications. This is due to the radioactivity in the primary containment, the radioactivity in l

the fluids flowing in emergency core coolant system piping, the amount and location of fluid and electrical penetrations, staircases and equipment shafts and the high ventilation recirculation rate in secondary containment. The licensee concludes that the use of the plant noble gas effluent monitors is the proper way to accomplish the purpose of this variable. Therefore, the licensee concludes that the existing Category 3 l instrumentation for this variable is adequate.

We find the existing Category 3 instrumentation and ranges in concert l with the noble gas effluent monitors acceptable.

l 3.3.13 Radiation Exoosure Rate Revision 2 of Regulatory Guide 1.97 recomends Category 2 instrumentation for this variable. Revision 3 changes this recommendation to Category 3. The range recomended is 10 to 10 R/hr. The licensee has Category 3 area monitors, 14 with a range of 0.1 to 3 4 10 mR/hr, 4 witn a range of 1 to 10 mR/hr and one with a range of ,

10" to 10 R/hr respectively. The licensee states that these will not be used to determine accessibility.

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From a radiological standpoint, if the radiation levels reach or exceed the upper limits of the instrumentation provided, personnel would not be permitted into the areas without portable monitoring (except for life saving). Based on the survivability of this equipment and the alternate portable monitoring instrumentation used by the licensee for this variable, we find the ranges for the radiation exposure rate monitors acceptable.

i 3.3,14 Accident Samplino Capability (primary coolant, containment air and sump) ]

The licensee takes exception to the recommendations of Regulatory l Guide 1.97 for this variable. For the ross activity range, the licensee specifies 8 decades, 0.1 mR/h to 1 x 10 mR/h, instead of the 10 pC'./mi l to 10 C1/mi recommended; and for the boron content, the licensee specifies a range of 100 to 1100 ppm instead of the recommended range of 0 to 1000 ppm. The licensee provided no justification for these deviations.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scepe of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.

3.3.15 Standby Liquid contr_o1 System (SLCS) Flow  !

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The licensee has elected not to implement this variable as recommended in Regulatory Guide 1.97. The justification given by the licensee is that j the SLCS pump (of positive displacement design) output pressure provides l indication that the SLCS pump is operating and that the level indication in f the SLCS storage tank gives indication that flow is occurring. In j addition, operation of this manually actuated system can be verified by changes in the reactivity as measured by the neutron flux instrumentation, the pump motor current and operating lights, and the squib valve continuity j l

indicating lights, i

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I We find the above instrumentation valid as an alternative indication of SLCS flow.

i 3.3.16 Standby Liquid Control System Storage Tank Level i The licensee states that this instrumentation will be operating in a f

mild environment and that the current design basis for the standby liquid i control system recognizes that the system has a classification less than i

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the importance to safety of the reactor protection system. ]

i The licensee conforms to all the criteria (power supply, range, etc.)  ;

i identified under Category 2 instrumentation, except for environmental 1 i

qualification.

This instrumentation is located in a mild environment.

Therefore, we find this instrumentation acceptable.

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4. CONCLUSIONS j

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Based on our review, we find that the licensee either conforms to or is justified in deviating from Regulatory Guide 1.97, with the foilowing {)

exception.  :

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1. Neutron flux--the licensee's present instrumentation is l

acceptable on an interim basis until Category 1 instrumentation is developed and installed; however, the licensee should commit to the installation of Category 1 instrumentation for this variable (Section 3.3.1).

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5. REFERENCES l l

1. NRC letter D. G. Eisenhut, to all Licensees of Operating Reactors, .

Applicants for Operating Licenses, and Holders of Construction l Permits, " Supplement No. I to NUREG-0737--Requirements for Emt'rgency Response Capability (Generic Letter No. 82-33)," December 17, 1982. )

2. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Cenditions During and Followina an Accident, Regulatory Guide 1.97, Revision 2 NRC, Office of Standards Development, December 1980.

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3. Clarification of TMI Action Plan Requirements. Requirements for )

Emergency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January ~,983.

l 4. Long Island Lighting Company letter , J. L. Smith to H. R. Denton,< t'RC, i

" Requirements for Emergency Response Capability--Implementation Schedule, Shoreham Nuclear Power Station--Unit 1, Docket No. 50-322,"

April 14, 1983. j i

5. Long Island Lighting Company letter, J. D. Leonard, Jr. to l H. R. Denton, NRC, " Emergency Response Capability, Regulate,y l Guide 1.97," October 23, 1985, SNRC-1209. j

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6. Long Island Lighting Company letter, J. D. Leonard, Jr. to )

H. R. Denton, NRC, " Emergency Response Capability, Regulatory Guide 1.97, Rev. 2," February 27, 1986, SNRC-1233.

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7. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess l Plant and Environs Conditions 0urina and followina an Accident, I l Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983. ,

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i This EG&G Idaho, Inc. report reviews the submittals for the Shoreham l

l Nuclear Power Station, Unit No.1. and identifies areas of nonconformance to Regulatory Guide 1.97. Exceptions to these guidelines are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

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