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Conformance to Reg Guide 1.97,Fort St Vrain Nuclear Generating Station, Informal Rept
	ML20212Q101
Person / Time
Site:	Fort Saint Vrain
Issue date:	08/31/1986
From:	Udy A; EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	; NRC
Shared Package
ML20209D506	List: ML20212Q101;
References
	CON-FIN-A-6483, RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-NTA-7081, GL-82-33, NUDOCS 8704160066
	Download: ML20212Q101 (19)
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EGG-NTA-7081 August 1986 l

INFORMAL REPORT r

e idaho CONFORMANCE TO REGULATORY GUIDE 1.97, Nationa/ FORT ST VRAIN NUCLEAR GENERATING STATION Engineering Laboratory Managed

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DISCLAIMER This book was prepared as an account of work sponsored by an agency of the United States Government. Neitner the United States Government nor any agency thereof, nor any of therr employees, mates any warranty, express or imphed, or assumes any legal babslity of respons.Dility for the accuracy, Completeness. or usefulness of any enformation, apDaratus. DrocuCt or process d:sclosed, of represents that sts use would not infringe privately owried rights References herein to any specifsc CommerCtal product, process, or sennce by trade name, trademark, manufacturer, or otherwise, does not necessarity Constitute or imply its endorsement, recommendation, or favonng by the United States Government or any agency tnereof. The vows and opinions of authors erDressed herein do not necessarily state or reflect those of the Uruted States Government or any agency thereof.

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EGG-NTA-7081 l', August 1986 s

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...~g INFORMAL REPORT

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Idaho CONFORMANCE TO REGULATORY GUIDE 1.97, National FORT ST. VRAIN NUCLEAR GENERATING STATION Engineering Laboratory Managed

, by the U S. A. C. Udy Department of Energy i

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EGG-NTA-7081

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! CONFORMANCE TO REGULATORY GUIDE 1.97 FORT ST. VRAIN NUCLEAR GENERATING STATION l A. C. Udy i

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Published August 1986 1

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EG&G Idaho, Inc.

l Idaho Falls, Idaho 83415 ;

i Prepared for the U.S. Nuclear Regulatory Connission Washington, D.C. 20555 Uncer 00E Contract Nc. OF-ACQ7,761001%70 I FIN No. A6483

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i j ABSTRACT 1

j This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for the Fort St. Vrain Nuclear Generating Station and identifies l'

areas of nonconformance to the regulatory guide. Exceptions to Regulatory l Guide 1.97 are evaluated.

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Docket No. 50-267 I TAC No. 51092 l

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4 FOREWORD j i

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

Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, ;

j Division of PWR Licensine-A, by EG&G Idaho, Inc., NRC and I&E Support

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l The U.S. Nuclear Regulatory Commission funded the work under ,

authorization 20-19-10-11-3.

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J-Docket No. 50-267 f' TAC No. 51092 i

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CONTENTS

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ABSTRACT .............................................................. 11 FOREWORD .............................................................. 11

1. INTRODUCTION ..................................................... 1 ,

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

3. EVALUATION ....................................................... 4 3.1 Adherence to Regulatory Guide 1.97 ......................... 4 3.2 Type A Variables ........................................... 4 3.3 Exceptions to Regulatory Guide 1.97 ........................ 4

4. CONCLUSIONS ...................................................... 11

5. REFERENCES ....................................................... 12 O

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i CONFORMANCE TO REGULATORY GUIDE 1.97 FORT ST. VRAIN NUCLEAR GENERATING STATION l =

I j- 1. INTRODUCTION l

4 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 l

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 l

response capability. These requirements have been published as Supplement Nc. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

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i ine Public Service Company of Colorado, the licensee for the l

i Fort St. Vrain Nuclear Generating Station, provided a response to the Regulatory Guide 1.97 portion of the generic letter on February 28, 1985 l

i (Reference 4). Additional information was provided on July 14, 1986 f (Reference 5).

This report provides an evaluation of these submittals.

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2. REVIEW REQUIREMENTS

.r 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 following information for each variable shown in the applicable table of Regulatory Guide 1.97:

1. Instrument range

2. Environmental qualification

3. Seismic qualification

4. Quality assurance

5. Redundance and sensor location

6. power supply

7. Location of display

8. Schedule of installation or upgrade.

The submittal should identify deviations from the regulatory guide and provide 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.

Regulatory Guide 1.97 addresses the post-accident instrumentation needed ,

f os light-water-cooled nuclear power plants. As the Fort St. Vrain station is a high temperature gas-cooled reactor, the NRC directed the licensee to adapt Regulatory Guide 1.97 by determining the variables to be monitored in a post-accident situation for this type of plant.

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1 j T j At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Where licensees or applicants l

explicitly state that instrument systems conform to the regulatory guide 1F l*

was noted that no further staff review would be necessary. Therefore, this j.. report only addresses exceptions to Regulatory Guide 1.97. The following evaluation is an audit of the licensee's submittals based on the review l - policy described in the NRC regional meetings.

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i 3. EVALUATION '

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j The licensee provided a response to NRC Generic Letter 82-33 on

] February 28, 1985. Additional information was provided on July 14, 1986. ,

This evaluation is based on these submittals.

i j 3.1 Adherence to Reaulatory Guide 1.97 l The licensee provided a report (Reference 4) describing the methodology used in determining the variables to be monitored at the

Fort St. Vrain station in a post-accident situation. The report details the instrumentation to be used to meet the intent of Regulatory Guide 1.97 4

{ as it applies to a high temperature gas-cooled reactor (the regulatory

] guide addresses itself to light-water-cooled reactors). The licensee compliance with the regulatory guide will be complete by the end of the fourth refueling outage (mid-1987). Additional instrumentation, identified

, in Reference 5, will be installed no later than the start-up following the L j same (fourth) refueling outage. Therefore, we conclude that the licensee l

! has provided an explicit commitment on conformance to Regulatory '

l Guide 1.97. Exceptions to and deviations from the regulatory guide are l noted in Section 3.3.

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j 3.2 Type A variables i, Regulatory Guide 1.97 does not specifically identify Type A variables, ,

t 1.e., these variables that provide the information required to permit the

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l control room operator to take specific manually controlled safety actions. !

i The licensee has identified the variable prestressed concrete reactor vessel pressure as the only Type A variable. This instrumentation meets the Category 1 requirements consistent with the requirements for Type A -

variables.

3.3 IsceDtions to Requlatory Guide 1.97 The licensee has determined those vartables to be utilized in a post-accident situation for monitoring a high temperature gas. cooled l

reactor. Some variables identified are the same as those used for a I

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light-water-cooled reactor (for example, neutren flux, core exit l

l temperature, radiation release monitoring instrumentation and

- meteorological instrumentation). Other variables are unique to the Fort St. Vrain Station. The licensee has utilized the concept of key and l ,.

backup variables as described in Regulatory Guide 1.97 in determining the l l ,

category of the instrumentation to be supplied.

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The licensee has identified instrumentation to be utilized for the

! following plant safety functions:

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i o Reactivity control

! o Core cooling 4

l 0 Reactor coolant system integrity i .

! o Reactor containment integrity !

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! These plant safety functions and their key and backup instrumentation are used to monitor twenty-six Type 8 variables. These functions correspond to

! those in Regulatory Guidt ' 97. Each function has a key variable !

I identified and several backup variables. We find the instrumentation I supp11ed for the Type B variables consistent with the purpose of Regulatory I Guide 1.97 and, therefore, acceptable.

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3.3.2 Type C Variables

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i The licensee has identified instrumentation to be utilized to indicate the potential for being breached or the actual breach of the barriers to l

! fission product releases. This instrumentation is grouped as follows:

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o fuel particle coating integrity 1 o Reactor coolant system pressure boundary r l

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o Reactor containment integrity o Confinement or reactor building (final fission product cleanup system) .

This instrumentation is used to monitor twenty-two Type C variables. The -

fuel particle coating integrity, reactor coolant system pressure boundary and reactor containment pressure boundary groups of key and backup instrumentation correspond to those in Regulatory Guide 1.97. Each group has a key variable identified and several backup variables. The confinement or reactor building group of variables is unique to the high temperature gas-cooled reactor as an additional barrier to fission product release. All instrumentation for this group of variables will be upgraded to Category 2 to monitor the status of the building louvers or the potential for their (i.e., the louvers) opening. This opening of these louvers is in accordance with the design basis accident senerios and additional upgrading of this instrumentation will not augment the post accident monitoring. We find the instrumentation being supplied for the Type C variables consistent with the purpose of Regulatory Guide 1.97 and, therefore, acceptable.

3.3.3 Type D variables l The licensee has identified instrumentation to be utilized to provide information to indicate the operation of individual safety systems or other systems important to safety. This instrumentation is grouped as follows:

o Primary heat removal auxiliaries o Secondary heat removal o Reserve shutdown system o Reactor plant cooling system o Purification cooling water system 6

o Helium purification system o Service water system o Nitrogen system

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o Ventilation system o power supplies This instrumentation is used to monitor thirty-seven Type 0 variables.

These groups of instrumentation correspond to the groups identified in Regulatory Guide 1.97. Three groups of instrumentation identified in Regulatory Guide 1.97 for Type 0 variables are not mentioned by the licensee in Reference 4. These are auxiliary feedwat# system, containment cooling system and raowaste system. There is no auxiliary feedwater system at Fort St. Vrain. Feedwater is provided by the feedwater system at all times, using either a steam turbine or a pelton wheel water turbine, which are on the same shaft as the feed pump, as motive power. Thus.

instrumentation for auxilie.y feedwater is not needed.

There is no containment building at Fort St. Vrain. The containment function is provided by the prestressed concrete reactor vessel which is ccoled by the reactor plant cooling system. This system is provided with appropriate mo.nitoring instrumentation.

The licensee has not identified instrumentation for radwaste systems because these systems are not used in response to the licensee's identified 1 , design basis accident s%nerlos. Therefore, the licensee concludes that this instrumentation does not need to be included as post-accident

. monitoring instrumentation. Since these systems are not used for accident f mitigation at the Fort St. Vrain Station, this exception from Regulatory Guide 1.97 is acceptable.

In Reference 4, the instrumentation groups ventilation system and I

power supplies are identified as being provided with Category 3

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j instrumentation. This instrumentation is as recommended by ,

Regulatory Guide 1.97, except that Category 2 instrumentation is ,.,

, . t r ecomended. The licensee, in Reference 5, commits to provide Category 2 [

4 instrumentation for tnese variables. We find this commitment acceptable, i ;

For the instrumentation groups reserve shutdown system, reactor plant I cooling system, purification cooling water system, helium purification !

q system and nitrogen system, the licensee did not identify, in Reference 4 j the key variables. In Reference 5, the licensee identified the key ;

i variables as follows:

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l Reserve shutdown system--bottle pressure

Reactor plant cooling system -pressurized concrete reactor vessel (PCRV) subheader flows, PCRV bottom penetrations, bottom head, barrel and l top head and circuit inlet header temperatures j

l Purification cooling water system--purification cooling water heat j

exchanger outlet temperature 1

1 j Helium purification system--helium purification dryer outlet

! temperature and helium purification low 1

temperature absorber outlet temperature l 4

l Nitrogen system--nitrogen storage tank level.

The licensee has comitted, in Reference 5, to upgrade this t

instrumentation to Category 2. We find the provided variables and .,

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l comitment to upgrade the instrumentation to Category 2 acceptable.

I l l The Fort St. Vrain station has two steam generators. Regulatory

) Guide 1.97 recommends monitoring the steam generator pressure with j Category 2 instrumentation (the licensee identified Category 3 l instrumentation in Reference 4) and steam generator level from the tube t

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sheet to the separators with Category 1 instrumentation (the licensee does l I

not have instrumentation for this variable). The steam generators at fort St. Vrain are of helical coil, once through design each consisting of o

integral economizer, evaporator, superheater and reheater sections. The

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water / steam flow is through the helically coiled tubes. Level measurements

- in this type of steam generator are not practical. Therefore, the

' exception of instrumentation for the steam generator level is acceptable.

The licensee, in Reference 5, committed to upgrade the Category 3 instrumentation for the variable steam generator presssure to Category 2.

We find this comitment acceptable.

3.3.4 Type f Variables The licensee has identified instrumentation to be utilized in i determining the release of radioactive materials and continually assessing such releases. The instrumentation is grouped as follows.

i o Airborne radioactive material released from plant o [nvirons radiation and radioactivity j o Neteorology o Accident sampling capability This instruraentation is used to monitor nine Type E variables. These groups of instrumentation correspond to the groups identified in Regulatory Guide 1.97. Two groups of instrumentation identified in Regulatory

. Guide 1.97 for Type E variables are not mentioned by the licensee in Reference 4. These are containment radiation and area radiation. As containment is provided by the prestressed concrete reactor vessel, instrumentation for containment radiation is not needed. Area radiation monitors in the turbine and reactor buildings are listed in Reference 5.

i The refueling floor high range area monitor has a range that envelopes the maximum expected radiation level. The instrumentation is mostly Category 3, which is acceptable per Revision 3 of Regulatory Guide 1.97.

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The other instrumentation is used in conjunction with portable instrumentation to evaluate personel entry into areas containing safe shutdown equipment. " '

From a radiological standpoint, if the radiation levels reach or .

exceed the upper limit of the range, personnel would not be permitted into the areas without portable monitoring (except for life saving). Based on the alternate supplemental portable instrumentation used by the licensee for this variable, we find the proposed ranges for the area radiation monitors acceptable.

For environs radiation and radioactivity, Reference 5 lists the portable radiation instrumentation in response to the requirements of NUR[G.0737, Supplement No. 1, Section 6.2. We find that the listed instrumentation in combination with the post-accident sampling system adequately covers these variables.

l For the variable estimation of atmospheric stabt11ty, Regulatory Guice 1.97 recommends a range of -9 to +18'F for this variable. The Itcensee's instrumentation has a range of -5 to +5'F. It is stated to be in corroliance with Regulatory Guide 1.23 (Reference 6) over a dif ferential height of 48 meters. We find this acceptable.

For the variable accident sampling (primary coolant, containment air and sump), Regulatory Guide 1.97 recomends sampilng the following parameters: gross activity, gamma spectrum, boron content, chloride content, hydrogen content, oxygen content and pH. The licensee lists the parameters gross activity and gama spectrum. Some of these parameters may not be necessary because the Fort St. Vrain design does not have a ,

containment building and sump.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident samp1tng capability. This deviation goes beyond the scope of

this review and is being addressed by the NRC as part of the review of NUREG 0137, Item !!.8.3.

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4. CONCLUSIONS

' Based on our review, we find that the licensee either conforms to or z 4

g is justified in deviating from Regulatory Guide 1.97.

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5. REFERENCES i :: - .

j 1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, l t Applicants for Operating Licensees, and Holders of Construction , i j Permits, " Supplement No. I to NUREG-0737--Requirements for Emergency ,

j Response Capability (Generic Letter No. 82-83) " December 17, 1982, i j , l l 2. Instrumentation for Licht-Water-Cooled Nucler Power Plants to Assess i i

Plant and Environs Conditions Durina and Fo'I ' owine an Accident, j Regulatory Guide 1.97, Revision 2, NRC, Office of Standards :

Development, December 1980. !

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3. Clarification c1 TM! Action Plan Reautrements. Reautrements for l j Emeraency Resoonse Capability, NUREG-0737 Supplement No. . NRC, I j Office of Nuclea.* Reactor Regulation, January 1983.

4 Public Service Company of Colorado letter D. W. Warembourg to [

j Regional Administrator, Region IV NRC, " Emergency Response !

> Capabilities, Regulatory Guide 1.97," February 28, 1985, P-85065 l j 5. Public Service Company of Colorado letter. D. W. Warembourg to i

Director of Nuclear Reactor Regulation, Attention H. N. Berkow, NRC, "

! " Emergency Response Capablittles Regulatory Guide 1.97 " July 14, .

1986, P-86460. :

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) 6. Regulatory Guide 1.23 (Safety Guide 23), Onsite Meteoroloaical

Procrams, NRC, February 17, 1972 or Proposed Revision 1 to Regulatory J

Guide 1.23, Meteoroloaical Proarams in Support of Nuclear Power i

Plants, NRC, Of fice of Standards Development, September 1980. !

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SISLIOGRAPHIC DATA SHEET EGG-NTA-7081 l7,*,'d

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. Conformance to Requlatory Guide 1.97 :

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Division of PWR Licensing-A Technical Evaluation Report Office of Nuclear Reactor Regulation U.S. fluclear Regulatory Comission . ** . c o c o e.s o ..- ...

Washingten. DC 20555

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This EGSG Idaho, Inc, report reviews the subtrittals from the Fort St. Vrain fluetear Generating Station and identifies areas of nonconformance tc 9equiatory Guide 1.97. Exceptions to these guidelines are evaluated.

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ML20212Q101

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